[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 [www.gpo.gov]
[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 https://www.regulations.gov by 11:59 p.m. eastern time on June 15, 2026.
ADDRESSES: Submit your comments, identified by Docket ID NRC-2025-0379,
at https://www.regulations.gov. If your material cannot be submitted
using https://www.regulations.gov, 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 https://www.regulations.gov to
view public comments.
You can read a plain language description of this proposed rule at
https://www.regulations.gov/docket/NRC-2025-0379. 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:
[email protected]; Elijah Dickson, Office of Nuclear Reactor
Regulation, telephone: 301-415-7647, email: [email protected];
Michael Balazik, Office of Nuclear Reactor Regulation, telephone: 301-
415-2856, email: [email protected]; and William Kennedy,
telephone: 301-415-2313, email: [email protected]. 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
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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:
Federal Rulemaking Website: Go to https://www.regulations.gov and search for Docket ID NRC-2025-0379.
NRC's Agencywide Documents Access and Management System
(ADAMS): You may obtain publicly available documents online in the
ADAMS Public Documents collection at https://www.nrc.gov/reading-rm/adams.html. To begin the search, select ``Begin Web-based ADAMS
Search.'' For problems with ADAMS, please contact the NRC's Public
Document Room (PDR) reference staff at 1-800-397-4209, at 301-415-4737,
or by email to [email protected]. For the convenience of the reader,
instructions about obtaining materials referenced in this document are
provided in the ``Availability of Documents'' section.
NRC's PDR: 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
[email protected] 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.
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: https://www.nrc.gov/public-involve/public-meetings/index.cfm.
B. Submitting Comments
Comments must be submitted electronically using https://www.regulations.gov 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
https://www.regulations.gov as well as enter the comment submissions
into ADAMS. The NRC does not routinely edit comment submissions to
remove identifying or contact information.
If you are requesting or aggregating comments from other persons
for submission to the NRC, then you should inform those persons not to
include identifying or contact information that they do not want to be
publicly disclosed in their comment submission. Your request should
state that the NRC does not routinely edit comment submissions to
remove such information before making the comment submissions available
to the public or entering the comment into ADAMS.
II. 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
(MWth) to about 2250 MWth. 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 MWth to about 4400
MWth.
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 (UO2) 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 (BeF2), and uranium tetrafluoride (UF4)
(LiF-BeF2-UF4).
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.
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.
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.
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.
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.
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.
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-
[[Page 23644]]
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 Table C-1 in appendix C of 10 CFR part 51). This conclusion is
based on the Commission's determination, in the 1996 final rule
amending its license renewal environmental review regulations (61 FR
66537), that impacts are of small significance if radiological doses to
individuals and radiological effluent releases do not exceed the
permissible levels in the Commission's regulations. The AEA requires
the NRC to promulgate, inspect, and enforce standards that provide an
adequate level of protection of the public health and safety. Health
impacts on individual humans are the focus of NRC regulations limiting
radiological doses. Numerous environmental assessments developed by the
NRC have concluded no significant impact with respect to human health
if radiological doses to individuals and radiological effluent releases
do not exceed the permissible levels in the Commission's regulations.
Therefore, if radiological doses to individuals and radiological
effluent releases do not exceed the permissible levels in the
Commission's regulations, which is the basis of the findings within
NUREG-2249, the impacts are not significant.
For those environmental impacts outside of human health, when a
SMALL impact is concluded in NUREG-2249, the NRC has determined that
the environmental effects are not detectable or are so minor that they
will neither destabilize nor noticeably alter any important attribute
of the resource, and this determination is comparable to a no
significant impact determination. The practical effect of this
determination is that actions that fall within the bounds of those
generic analyses in NUREG-2249 would meet the criteria for a
categorical exclusion, or the basis for a finding of no significant
impact if the NRC prepares an environmental assessment.
This categorical exclusion for this proposed rule does not rely
upon Category 2 issues in NUREG-2249 because that conclusion is not
generic across all sites. Instead, this proposed rule includes criteria
in Sec. 57.350(b) that, if met, ensure the environmental impacts of
the action would not be significant. The NRC provides guidance in
Chapter 16 of draft NUREG-2271, ``Guidelines for Preparing and
Reviewing Applications Under 10 CFR part 57,'' on addressing these
criteria.
As such, if an application for a proposed microreactor meets the
values and assumptions of the plant parameter envelope and site
parameter envelope for Category 1 issues as defined in NUREG-2249 and
the specific criteria for all other issues that are described in
Chapter 16 of draft NUREG-2271, and there is no new and significant
information that would change these conclusions, then these actions
related to construction permits and operating licenses for
microreactors and other reactors with comparable risk profiles do not
individually or cumulatively have a significant effect on the human
environment under 10 CFR 51.22.
The proposed criteria for the categorical exclusion would revolve
around site-specific considerations that the NRC and other Federal
agencies have established based on environmentally sensitive resources.
An environmentally sensitive resource is typically a resource that has
been identified as needing protection through Executive Order, statute,
or regulation by Federal, State, or local government, or a Federally-
recognized Indian tribe. The NRC is proposing four such criteria based
on past NEPA reviews and being informed on how other Federal agencies,
such as the U.S. Department of Energy (DOE) and the U.S. Department of
War (DOW), have defined environmentally sensitive resources. The four
criteria being proposed and the rational for each are as follows:
1. The Site Will Be Within a Previously Disturbed Area as Defined in
Sec. 57.3
The NRC would define ``previously disturbed areas'' in proposed
Sec. 57.3 as areas that have been changed by development of a prior
facility and remain altered by human activity such that they do not
provide habitat for ecologically important species, such as those
protected under the Endangered Species Act, and no longer have the
potential to yield historic and cultural resources. This definition
would include the lateral and vertical extent of alteration from
natural cover to a managed state. This proposed definition is based on
the definition of ``previously disturbed or developed'' in DOE's NEPA
implementing regulations under paragraph (g)(1) of Sec. 1021.102,
``Application of categorical exclusions (categories of actions that
normally do not require EAs or EISs).''
2. The Cooling System(s) Will Not Require the Use or Consumption of
Water Withdrawn Directly From Surface Water or Groundwater Sources or
Discharge to Surface Water or Groundwater Sources
In NUREG-2249, the NRC identified three water-related issues as
Category 2 issues, which cannot be evaluated generically and must be
evaluated on a case-by-case basis using project-specific information.
Water-based cooling systems discharge waste heat and have the potential
to affect the water bodies from which water is taken and into which it
is discharged. If the cooling system of the facility does not result in
the direct withdraw or discharge of water from surface water or
groundwater resources degradation of surface water quality and impacts
to aquatic biota from chemical and thermal discharges are not
anticipated. The three issues involve (1) surface water quality
degradation due to chemical and thermal discharges, (2) thermal
discharge plume impacts on aquatic biota, and (3) other effects of
cooling-water discharges on aquatic biota. Of specific note regarding
surface water quality degradation due to chemical and thermal
discharges, Clean Water Act section 401 on water quality certification
states that a Federal agency may not issue a license or permit to
conduct any activity, including
[[Page 23656]]
construction or operation of facilities, that may result in any
discharge into navigable waters (i.e., ``waters of the United States'')
unless the State or authorized Tribe where the discharge would
originate issues either a Clean Water Act section 401 water quality
certification or a waiver. The Clean Water Act forbids ``any addition''
of any pollutant from ``any point source'' to ``navigable waters''
without an appropriate permit from the Environmental Protection Agency
(EPA), or EPA-delegated permit authority. Water quality certification
is intended to ensure that the discharge will comply with applicable
effluent limitations and water quality requirements under the Clean
Water Act and with any appropriate requirement of State law. The
Supreme Court of the United States reinforced this in its decision in
County of Maui v. Hawaii Wildlife Fund, 590 U.S. 165 (2020). The Court
held that the statute requires a permit when there is a direct
discharge from a point source into navigable waters. This includes
industrial and stormwater point-source discharges of pollutants to
navigable waters of the United States, which, in the case of many
nuclear power plants, are to surface water bodies. Thus, if a reactor
under proposed part 57 would not require the use or consumption of
water for the cooling system and the site would not have significant
point-source discharges (e.g., stormwater), no project-specific
information or analysis would be necessary. Therefore, meeting this
criterion would support a determination that a categorical exclusion
could be issued. In a similar manner, both DOW and DOE categorical
exclusions include when environmental effects involving water use and
quality are items that must be considered (e.g., DOW's ``Cultural and
natural resources'' categorical exclusion in its ``National
Environmental Policy Act Implementing Procedures,'' appendix A,
``Department of Defense Categorical Exclusions (CATEX),'' paragraph
I.(c)1., and DOE's ``Drop-in Hydroelectric Systems'' categorical
exclusion in 10 CFR part 1021, ``National Environmental Policy Act
Implementing Procedures,'' appendix B, ``Categorical Exclusions
Applicable to Specific Agency Actions,'' paragraph B5.24).
3. Air Emissions Will Be Below de Minimis Threshold Levels in 40 CFR
93.153(b)(1) or (b)(2), as Applicable
The Clean Air Act, as implemented in EPA's enabling regulations,
set de minimis threshold levels for air quality in areas defined as
non-attainment and maintenance areas under 40 CFR 93.153(b)(1) for non-
attainment areas and 40 CFR 93.153(b)(2) for maintenance areas. This
criterion on air emissions would be consistent with categorical
exclusion criteria in other Federal agencies such as the DOW and DOE
where all airborne emissions must be in compliance with existing
applicable Federal, State, and local laws and regulations (e.g.,
paragraph III.16. of appendix A in DOW's NEPA Implementing Procedures)
or would not cause a significant increase in the quantity or rate of
air emissions (e.g., DOE's ``Projects to Reduce Emissions and Waste
Generation'' categorical exclusion in 10 CFR part 1021, appendix B,
paragraph B3.9).
4. The Licensed Activity Will Be in Accordance With Applicable State
and Local Requirements (Such as Land Use Planning, Zoning Requirements,
and Coastal Zone Management Program Requirements Under the Coastal Zone
Management Act) in the Proposed Site or Region
Any commercial construction activity may have to satisfy local land
use planning and zoning requirements as enacted in ordnances outside of
the NRC's licensing actions. Government ordnances could include
radiological liquid effluent discharge restrictions, and land use
planning and zoning requirements. Some States may also have their own
environmental regulations similar to the Federal government's NEPA
(e.g., the State of Washington's State Environmental Policy Act
(https://ecology.wa.gov/regulations-permits/sepa/environmental-review/sepa-guidance/basic-overview)). This categorical exclusion criterion
would be similar to the criterion in DOW and DOE categorical
exclusions. For example, DOW applies the following criterion under a
Missile Defense Agency categorical exclusion in paragraph VI.18.a of
appendix A in its NEPA Implementing Procedures for new construction or
equipment installation: ``The structure and proposed use are compatible
with applicable Federal, tribal, state, and local planning and zoning
standards.'' DOE states in many of their categorical exclusions that
``[c]overed actions would be in accordance with applicable requirements
(such as local land use and zoning requirements) in the proposed
project area'' (e.g., DOE's ``Small-Scale Renewable Energy Research and
Development and Pilot Projects'' categorical exclusion in 10 CFR part
1021, appendix B, paragraph B5.15). Thus, the construction and
operation of microreactors or other reactors with comparable risk
profiles would also need to be in accordance with applicable State and
local requirements in the proposed site or region.
Separately, in response to E.O. 14300, section 5(c), the NRC is
reexamining the NRC's NEPA implementing regulations in 10 CFR part 51.
M. Subpart L--Inspections
Proposed Sec. 57.355, ``Unfettered access for inspections,'' would
establish requirements for the provision of facilities and unfettered
access for inspections. These requirements would be equivalent to Sec.
50.70, ``Inspections,'' with only minor changes proposed to provide
additional flexibilities and address possible differences related to
reactors licensed under proposed part 57. Proposed Sec. 57.355 also
would address inspections for transportation of radioactive material,
storage of nuclear fuel and radioactive waste and would be equivalent
to Sec. Sec. 71.93, ``Inspection and tests,'' 72.82, ``Inspections and
tests,'' and 70.55, ``Inspections,'' respectively.
N. Subpart M--Material Control and Accounting
The NRC would include regulations for material control and
accounting specific to microreactors and other reactors with comparable
risk profiles because the provisions in 10 CFR part 74, ``Material
Control and Accounting of Special Nuclear Material,'' do not explicitly
provide these requirements. The proposed material control and
accounting requirements in proposed Sec. 57.360, ``Material control
and accounting,'' would be equivalent to the requirements of part 74,
subpart B, ``General Reporting and Recordkeeping Requirements,'' which
is applicable to all holders of SNM. Microreactors and other reactors
with comparable risk profiles would not be required to meet the other
requirements in part 74 (except enforcement), the general performance
objectives and system capabilities, because those requirements were
written principally for fabrication and enrichment facilities.
The NRC proposes to employ a risk-informed approach, so the
material control and accounting for a microreactor or reactor with
comparable risk profile would be equivalent to the measures at a large
LWR, recognizing that the total amounts of material would differ. For
the use of high assay low enriched uranium (HALEU) at microreactors or
other reactors with comparable risk profiles, the frequency of physical
inventory would not be greater than 6 months for licensees of
facilities without personnel on site.
[[Page 23657]]
Otherwise, licensees of facilities under proposed part 57 would be
subject to the controls under part 74, subpart B. The increase in
periodicity of the physical inventory from the 12 month subpart B
requirement would provide additional assurance that this higher
enriched material has not been diverted or lost. For these reactors
that will use fuel that is not in item form, equivalent measures to the
material control and accounting under subpart B would be used, but not
the full set of measures used at a fabrication or enrichment facility.
The Nuclear Material Management and Safeguards System provisions,
as described in part 74, subpart B, would be applicable to proposed
part 57 licensees, especially for reporting operation location as the
nuclear reactors move across geographic locations. These licensees
would follow the reporting requirements for nuclear material
transaction reports and material balance reports, as required in part
74, subpart B and submit reports consistent with electronic reporting
instructions provided in NUREG/BR-0006 and NUREG/BR-0007.
O. Subpart N--[Reserved]
Subpart N is reserved for future rulemakings in part 57.
P. Subpart O--Enforcement
Subpart O would contain two provisions, proposed Sec. 57.380,
``Violations,'' and Sec. 57.385, ``Criminal penalties,'' which would
be analogous to provisions contained in other parts of 10 CFR chapter I
that impose requirements on regulated entities. Proposed Sec. 57.380
would provide notice of the Commission's authority under the AEA to
obtain injunctions or other court orders for the enumerated violations.
Proposed Sec. 57.385(a) would provide notice to all persons and
entities subject to proposed part 57 that they would be subject to
criminal sanctions for willful violations, attempted violations, or
conspiracy to violate certain regulations under proposed part 57.
Criminal sanctions would not apply to the regulations listed in
proposed Sec. 57.385(b). The regulations for which criminal penalties
would apply are limited to those that establish either a regulatory
obligation or prohibition.
Q. Subpart P--Operator Licensing and Human Factors
Proposed subpart P of part 57 would include provisions to address
staffing, training, personnel qualifications, and human factors
engineering (HFE) requirements that would be applicable to the
operation of microreactors or other facilities with comparable risk
profiles. These requirements would be adapted from portions of
Sec. Sec. 50.34(f) and 50.54, ``Conditions of licenses,'' and 10 CFR
part 55, ``Operators' Licenses,'' with considerable modification to
reflect the expected reduced role of personnel in preventing and
mitigating events and to be consistent with the licensing framework of
other facilities with comparable risk profiles, like NPUFs. These
requirements also would serve as a component of the required content of
joint applications for CPs and associated OLs under proposed part 57.
The requirements associated with this approach would be in proposed
Sec. Sec. 57.390, ``Definitions,'' through 57.429, ``Training and
qualification for non-licensed personnel.'' These sections would be
divided into four main portions that cover HFE and human interface
system (HSI) design requirements, generally licensed reactor operator
(GLRO) requirements, operator and senior operator requirements, and
training requirements for other nuclear plant personnel.
Proposed Sec. 57.390 would define specific terms. Some definitions
would draw from those in Sec. 55.4, ``Definitions.'' The NRC would
introduce five new definitions for use within the context of proposed
subpart P. These new definitions would be the following: ``Auxiliary
operator,'' ``Generally licensed reactor operator,'' ``Load
following,'' ``Operator-dependent facility,'' and ``Operator-
independent facility.''
To establish uniform conditions for the licensing operators, the
NRC proposes two classes of nuclear power plants in Sec. 57.391(a). An
``operator-dependent facility'' is the classification for a nuclear
plant whose design demonstrates that operator actions are required to
maintain the nuclear plant within the dose criterion of proposed Sec.
57.25(a); the NRC would require the specific licensing of operators and
senior operators to manipulate the controls and direct the licensed
activities of operators at this class of nuclear plant under proposed
Sec. 57.420. This concept would be like provisions for operators and
senior operators at ``interaction-dependent-mitigation facilities''
introduced in part 53.
An ``operator-independent facility'' is the classification for a
nuclear plant whose design demonstrates that no operator actions are
required to maintain the nuclear plant within the criterion of proposed
Sec. 57.25(a). A GLRO would be an individual licensed under the
provisions of proposed Sec. 57.405, ``Generally licensed reactor
operators,'' to manipulate controls of an operator-independent facility
licensed under proposed part 57 and to direct the licensed activities
of GLROs. The concept of general licensing of operators under proposed
part 57 would be similar to provisions for GLROs introduced in part 53.
The term ``auxiliary operator'' would mean any individual who would
operate components of a nuclear plant licensed under proposed part 57
but would not manipulate controls or direct the manipulation of
controls of the plant and would not be required to be licensed under
proposed part 57. This term would distinguish between plant personnel
that operate the controls of the facility and are therefore required to
be licensed and those that are not required to be licensed because they
do not manipulate or direct the manipulation of plant controls. The
term ``load following'' would describe a nuclear plant automatically
changing its output to match expected demand in response to externally
originated instructions or signals.
Certain routine communications are necessary to facilitate the
operator licensing process. The NRC would adapt the requirements of
Sec. Sec. 55.5, ``Communications,'' and 50.74, ``Notification of
change in operator or senior operator status,'' in proposed Sec.
57.392, ``Communications,'' to accomplish this.
Specific information must be collected to facilitate the initial
issuance of operator licenses, as well as to allow for license renewals
and required updates thereafter. Such information collection activities
must also be approved by the OMB. The NRC would adapt the requirements
of Sec. 55.8, ``Information collection requirements; OMB approval,''
to include any needed updates in OMB approval information in proposed
Sec. 57.8 to accomplish this.
The information used within the regulatory processes of the NRC
must be free from omissions and inaccuracies to facilitate effective
regulation. Consistent with this, the NRC would adapt the requirements
of Sec. 55.9, ``Completeness and accuracy of information,'' in
proposed Sec. 57.393, ``Completeness and accuracy of information,'' to
require the completeness and accuracy of material information provided
by individual applicants and license holders.
Proposed Sec. 57.395, ``Human factors engineering requirements,''
would contain the HFE requirements for applicants for or holders of an
OL under proposed part 57. Proposed Sec. 57.395(a) would contain the
human-system interface design requirements. Human-system interfaces
provide vital information to plant operations staff
[[Page 23658]]
across a spectrum of operating conditions that can range from normal
operations through accident conditions. The specific types of
information that must be available to support operations staff during
such conditions would include, in part, those associated with safety
function parameters, safety system status, possible core damage states,
barrier integrity, and radioactive leakage. Due to the importance of
such information, the NRC would require, under proposed Sec.
57.395(a), specific human-system interface design features for all part
57 facilities. Therefore, the NRC would adapt the following post-Three
Mile Island requirements of Sec. 50.34(f) in a technology-inclusive
manner:
Sec. 50.34(f)(2)(iv) would become proposed Sec.
57.395(a)(1).
Sec. 50.34(f)(2)(v) would become proposed Sec.
57.395(a)(2).
Sec. 50.34(f)(2)(xi), 50.34(f)(2)(xii), and
50.34(f)(2)(xxi) would become proposed Sec. 57.395(a)(3).
Sec. 50.34(f)(2)(xvii), 50.34(f)(2)(xviii),
50.34(f)(2)(xix), and 50.34(f)(2)(xxiv) would become proposed Sec.
57.395(a)(4).
Sec. 50.34(f)(2)(xxvi) would become proposed Sec.
57.395(a)(5).
Sec. 50.34(f)(2)(xxvii) would become proposed Sec.
57.395(a)(6).
Sec. 50.34(f)(2)(iii) would become the proposed Sec.
57.395(d) and would only be applicable to locations where operator
actions are required to maintain the reactor within the criterion of
proposed Sec. 57.25(a) or locations where a credible operator or
maintenance error could result in exceeding that criterion.
In addition to the requirements of proposed Sec. 57.395(a)(1)
through (6), the human-system interfaces and operator capabilities
listed in proposed Sec. 57.395(a)(7)(i)-(iv) would be required to
allow GLROs, operators, and senior operators to evaluate plant
conditions and respond appropriately in the event of an emergency. This
would also include the ability to immediately initiate a manual reactor
shutdown. Operating experience provides an important source of
information by which to inform various aspects of facility design and
operations. Accordingly, the NRC would adopt in proposed Sec.
57.395(b) the requirements of Sec. 50.34(f)(3)(i) for requiring an
operating experience program.
The NRC recognizes that the licensed operator staffing requirements
of Sec. 50.54(k) and (m) are prescriptive and in most cases would not
be appropriate for the staffing needs of microreactors and other
reactors with comparable risk profiles. Therefore, proposed Sec.
57.395(c) would allow a performance-based means to determine staffing
levels for proposed part 57 facilities. The staffing plan would need to
be supported by HFE analyses and assessments and approved by the NRC.
Once the appropriate facility staffing plan has been determined and
approved by the NRC, the staffing level would need to be maintained to
ensure that appropriately qualified individuals would be available when
needed to support the safe operation of the facility. Therefore, the
NRC would require under proposed Sec. 57.399(a) that the staffing
described within the approved facility staffing plan be maintained as a
condition of the facility license. Under proposed Sec. 57.395(c), the
staffing plan would be part of the OL and, thus, a license amendment
would be required for any subsequent changes to the plan.
Due to the unique authorities and responsibilities of nuclear power
plant operators, it would be essential that any individual fulfilling
such a role demonstrate compliance with the regulatory requirements for
operator licensing. Section 107 of the AEA authorizes the Commission to
prescribe conditions for the licensing of operators and to issue
licenses consistent with those conditions. The NRC would adapt the
requirements of Sec. 55.3, ``License requirements,'' in proposed Sec.
57.398, ``Operator license requirements,'' to require that any person
performing the function of a GLRO, operator, or senior operator be
authorized by a license issued by the Commission.
The NRC proposes to license individuals to operate proposed part 57
facilities under a general licensing framework or a specific licensing
framework depending on the licensed operators' role in reactor safety.
The GLRO framework would only be applicable to proposed part 57
facilities that do not require operator actions to maintain the reactor
within the criterion of proposed Sec. 57.25(a), or operator-
independent facilities, as required by proposed Sec. 57.405(a),
``Applicability.'' If one or more operator actions are required to
maintain the reactor within the criterion of proposed Sec. 57.25(a),
then the specific licensing framework for operators at operator-
dependent facilities and the requirements in proposed Sec. Sec. 57.420
through 57.427, ``Expiration of operator and senior operator
licenses,'' would apply.
GLROs would perform duties under the provisions of a general
license that would be effective without the filing of an application
with the Commission or the issuance of licensing documents to a
particular person. The NRC proposes requirements for the general
licensing process for GLROs under proposed Sec. 57.400 through Sec.
57.415. The requirements for GLROs would parallel those for senior
operators under part 55 regarding their comparable administrative
responsibilities. However, operator licensing for GLROs would have
fewer requirements compared to the requirements for specifically
licensed operators under part 55 due to the GLROs not having to execute
operator actions to maintain the reactor within the criterion of
proposed Sec. 57.25(a) and unique safety attributes of microreactors
and other reactors with comparable risk profiles.
In order to use GLROs to operate the controls of a proposed part 57
facility, an OL applicant would need to demonstrate that it would
comply with the following requirements on an ongoing basis: maintain
GLRO qualifications for the performance of important functions and
tasks; incorporate relevant programmatic controls into technical
specifications; administer the related programs for training,
examination, and proficiency; and ensure that the relevant provisions
of part 26 would be met. Additionally, to provide for an accurate
accounting of what individuals would be licensed under the general
license, facility licensees would be required to report the identities
of all generally licensed reactor operators to the NRC on an annual
basis. Proposed Sec. 57.400(a) through (f) would establish
requirements for facility licensees that address these topics and
others.
Under the AEA, the NRC is required to license any individuals who
manipulate the controls of a utilization or production facility.
Because the operation of facility controls would directly affect
reactivity or power level of the reactor, only those individuals who
possess appropriate levels of qualification and authorization would be
permitted to operate those controls. The NRC would adapt the
requirements of Sec. 50.54(i) in proposed Sec. 57.399(b) to require
that only GLROs, operators, and senior operators may operate facility
controls, with allowance for specified exceptions for the purposes of
operator training or proficiency.
Proposed Sec. 57.399(c) would require that a GLRO, operator, or
senior operator monitor plant conditions during the manipulation of
apparatus and mechanisms, other than controls, that could affect the
reactivity or power level of the reactor.
Load following occurs when plant output automatically changes in
response to externally originated instructions or signals and is not
permitted under the existing regulations of Sec. 50.54. However, new
technological considerations and concepts of
[[Page 23659]]
operation may justify such an operational approach under appropriate
circumstances. The NRC recognizes that, beyond electrical power
generation, load following may also affect other applications of plant
output, such as hydrogen production, desalination, or district heating.
For load following to be permissible, measures must be in place to
provide assurance that plant output considerations are not permitted to
lead to challenges to safe reactor operations. These measures may
consist of automated control systems, automatic protective features, or
the continuous oversight and immediate intervention capability of an
appropriately qualified and authorized individual. Proposed Sec.
57.399(d) would allow for load following, provided that appropriate
measures in proposed Sec. 57.399(d)(1) were in place.
Core alterations such as refueling are associated with specific
considerations that warrant limiting the oversight of such operations
to appropriately qualified and authorized individuals. Unlike other
types of fuel handling operations, core alterations occur within the
confines of a reactor vessel that is specifically designed to support
and sustain nuclear criticality, thereby justifying the imposition of
higher qualification levels within such contexts. The NRC would adapt
the requirements of Sec. 50.54(m)(2)(iv) in proposed Sec. 57.399(e)
to require the supervision of core alterations by a GLRO, senior
operator, or a senior operator limited to fuel handling, as applicable
to the facility. Because certain reactor designs may be capable of
refueling while at power and, in any event, overall facility oversight
would already be required by a GLRO or senior operator, proposed Sec.
57.399(f) would omit this requirement as redundant during periods where
core alterations occur while the plant is operating.
The NRC cannot predict every possible scenario that a nuclear plant
might potentially encounter. Therefore, it is prudent to grant the
authority for appropriately qualified individuals to depart from
facility license conditions when emergency circumstances dictate that
doing so is in the interest of public health and safety. The NRC would
adapt the requirements of Sec. 50.54(x) and (y) in proposed Sec.
57.399(g) and (h) to permit GLROs or senior operators to authorize
departures from facility license conditions or technical specifications
when emergency conditions warrant doing so for the protection of the
public health and safety. While the NRC does not anticipate that GLROs
will have a role in the fulfillment of safety functions at operator-
independent facilities licensed under part 57 or that operators at such
facilities would be in a position to significantly influence
radiological safety outcomes, the very nature of Sec. 50.54(x) and (y)
and proposed Sec. 57.399(g) and (h) concerns situations that are
unanticipated and, therefore, unforeseeable. Thus, it is appropriate to
propose to grant GLROs a comparable authority to that of senior
licensed operators and certified fuel handlers as it relates to
invoking this provision under emergency conditions as a means of
accounting for such possibilities.
GLROs would be licensed as a class of individuals under the
provision of proposed Sec. 57.405(a) and would be subject to the
conditions specified in proposed Sec. 57.405(b)(1) through (8).
Portions of these conditions are adapted from Sec. 55.53, ``Conditions
of licenses.'' The NRC would retain the ability to suspend or prohibit
individuals from operating under the general license should such action
be warranted.
The NRC proposes overall programmatic requirements for GLRO
training, examination, and proficiency in proposed Sec. 57.410,
``Generally licensed reactor operator training, examination, and
proficiency programs.'' In general, these proposed requirements would
be adapted from those of part 55. These requirements would include
flexibility commensurate with the expected reduced level of operator
actions at microreactor and other reactors with comparable risk
profiles. The requirements in proposed Sec. 57.410 would cover, in
part, the initial training, initial examination, continuing training,
requalification examination, and proficiency of GLROs. Proposed Sec.
57.400(b) would require the facility licensee to develop, implement,
and maintain these programs. Proposed Sec. 57.405(b)(1)-(8), in turn,
would prescribe that the requirements of proposed Sec. 57.400 would
need to be met as a requirement of the general license. The implication
of this structure is that the facility licensee would need to implement
these programs for training, examination, and proficiency, and GLROs
would need to participate in these programs to demonstrate compliance
with the requirements of the general license. The initial training
process would provide GLROs with the knowledge and abilities needed to
fulfill assigned duties as GLROs. The use of a systems approach to
training (SAT)-based training program would serve to ensure that the
training program is based upon job requirements in a manner that can be
adapted to account for differences in plant technology and concepts of
operations. Proposed Sec. 57.410(b) would require facility licensees
to implement an SAT-based training program for the initial training of
GLROs that would be adequate to ensure that they have the necessary
knowledge, skills, and abilities to perform their duties. For
microreactor and other reactors with comparable risk profiles, such
programs would not be subject to NRC approval, however the NRC would
maintain oversight of these licensing programs through inspection.
Examinations would provide a means of assessing that individuals
have achieved a degree of knowledge and ability that would be
sufficient to enable them to carry out assigned duties as GLROs in a
manner that is both safe and reliable. The NRC would adapt the
requirements of Sec. Sec. 55.40, ``Implementation,'' 55.41, ``Written
examination: Operators,'' 55.43, ``Written examination: Senior
operators,'' and 55.45, ``Operating tests,'' in proposed Sec.
57.410(b), ``Requirements,'' to require that facility licensees
establish and implement an initial examination program for GLROs. A key
difference from the current comparable requirements of part 55 would be
that facility licensees under proposed part 57 would have the
flexibility to determine, subject to NRC approval, the examination
methods and criteria to be used in assessing satisfactory individual
performance. Such examination programs (including those used within the
scope of continuing training) would need to provide for acceptable
levels of both test validity and test reliability in order to be
considered acceptable. In contrast with requirements for licensing
examinations in part 55, the NRC would not administer or evaluate these
initial examinations of GLROs. However, the examination processes would
continue to be subject to ongoing NRC oversight including subsequent
review and approval of any substantial changes to approved examination
programs. The NRC plans to develop guidance to facilitate the review of
initial examination programs that are proposed by facility licensees.
Continuing training programs would provide the ongoing training and
examination of GLROs to ensure that they maintain the knowledge and
abilities needed to support the safe and reliable performance of job
duties following the completion of an initial training and examination
program. The NRC would adapt the requirements of Sec. 55.59,
``Requalification,'' in proposed Sec. 57.410(b) to require that
facility licensees implement both an SAT-based continuing training
program and a requalification examination program.
[[Page 23660]]
However, a notable difference from the examinations required under part
55 is that under proposed part 57, distinct annual operating test and
biennial written examination components would not be required. Instead,
the facility licensee would propose examination methods and criteria to
be used in assessing satisfactory performance. The NRC plans to develop
guidance to facilitate the review of the requalification examination
programs that are proposed by facility licensees.
For examinations to provide for valid assessments of the knowledge
and abilities of individuals, the examinations must remain free from
compromises that could affect their underlying integrity. The NRC would
adapt the requirements of Sec. 55.49, ``Integrity of examinations and
tests,'' in proposed Sec. 57.410(d) to require that examinations and
related activities remain free from any compromise that might affect
the integrity of the examination process.
Simulators provide a valuable means of training and evaluating
plant operators, and the NRC is specifically authorized under section
306 of the Nuclear Waste Policy Act of 1982 (42 U.S.C. 10226) to
establish regulations for the use of simulators within such context.
The NRC would adapt the requirements of Sec. 55.46, ``Simulation
facilities,'' in proposed Sec. 57.410(e) to address the use of
simulation facilities for training and examinations, and experience
requirements, as well as to address the maintenance of simulator
fidelity. The use of full scope, plant-referenced simulators would not
be mandatory. The potential use of alternative simulation facilities
consisting of, for example, partial scope simulators or the plant
itself, would be allowed provided that all associated proposed
requirements were demonstrated to be met using alternative approaches
and methods.
There may be situations in which GLROs have previous training and
experience that justify waiving some or all of the initial examination.
Therefore, under proposed Sec. 57.410(f), the NRC would allow facility
licensees to waive some or all portions of initial examinations
provided that such waivers would be consistent with an examination
program that has been approved by the NRC.
For GLROs to safely and reliably perform their assigned duties,
they would need to perform those duties frequently enough to maintain a
sufficient degree of proficiency. However, the NRC recognizes that
facilities that would utilize GLROs may have concepts of operation that
warrant unique proficiency considerations. Therefore, the NRC would
require in proposed Sec. 57.410(g) that facility licensees develop,
implement, and maintain programs to maintain and re-establish, if
needed, the proficiency of GLROs. This could occur, for example, if an
individual's extended absence from watch standing rendered proficiency
requirements unmet.
The NRC would require under proposed Sec. 57.415, ``Cessation of
individual applicability,'' that the general license would cease to be
applicable on an individual basis when the individual would no longer
be employed in a position that might call for the individual to
manipulate the reactivity controls of the facility. However, the NRC
recognizes that for some types of proposed part 57 facilities, very
long periods may elapse between circumstances that necessitate manual
manipulation of reactivity controls. Therefore, the general license
would remain in effect for an individual as long as the individual's
current position could potentially require that individual to
manipulate reactivity controls at some point within the course of the
individual's assigned job duties.
Specifically licensed operators would differ from GLROs because the
former would be directly and independently evaluated by the NRC as part
of their licensing process. This direct and independent evaluation
would remain appropriate at operator-dependent facilities where
operators may reasonably be expected to have a role in public health
and safety outcomes. The NRC would set forth requirements for the use
of a specific licensing process for licensed operators and senior
operators under proposed Sec. Sec. 57.420 through 57.427, with Sec.
57.420 addressing applicability.
Medical fitness is an important component of the overall process of
specifically licensing operators because it provides assurance that
operators will be able to carry out important duties without being
precluded from doing so by health-related issues. Medical fitness also
provides assurance that such issues will not adversely affect the
performance of assigned job duties or cause operational errors that
endanger public health and safety. In addition to a requirement for
medical fitness, a medical examination by a physician to confirm
compliance with this requirement would be necessary. The NRC would
adapt the requirements of Sec. Sec. 55.21, ``Medical examination,''
55.23, ``Certification,'' and 55.27, ``Documentation,'' under proposed
Sec. 57.421, ``Medical requirements,'' to require medical fitness,
examinations by physicians, and medical certification for specifically
licensed operators and senior operators. In recognition of the fact
that GLROs are not expected to have a role in the fulfillment of safety
functions at the facilities at which they are licensed, the NRC would
not extend a comparable medical requirement to GLROs.
The NRC also would adapt the requirements of Sec. Sec. 55.25,
``Incapacitation because of disability or illness,'' and 50.74(c) in
proposed Sec. 57.422, ``Incapacitation because of disability or
illness,'' to require that timely notifications be made to the NRC if a
specifically licensed operator or senior operator develops a permanent
physical or mental condition that adversely affects the performance of
assigned operator job duties or could cause operational errors
endangering public health and safety.
The process of specifically licensing individuals as operators or
senior operators requires the submittal of applications to the NRC for
review. These applications must detail certain elements associated with
licensing, including the demonstration of compliance with examination,
experience, and medical requirements. The NRC would adapt the
requirements of current subpart D, ``Applications,'' of part 55 in
proposed Sec. 57.423, ``Applications for operators and senior
operators,'' to include requirements for the applications associated
with the specific licensing of operators and senior operators at
commercial nuclear plants licensed under proposed part 57.
The NRC proposes programmatic requirements for specifically
licensed operator and senior operator training, examination, and
proficiency in Sec. 57.424, ``Training, examination, and proficiency
programs.'' In general, the requirements are adapted from those in part
55, but with flexibility to support diverse reactor technologies and
concepts of operations. Specifically, the requirements in proposed
Sec. 57.424 would concern the initial training, initial examination,
requalification training, requalification examination, and proficiency
of specifically licensed operators and senior operators.
The initial training process provides individuals with the
knowledge and abilities needed to subsequently fulfill assigned duties
as licensed operators or senior operators in a safe and reliable
manner. The use of an SAT-based training program would ensure that the
training program is based upon job requirements in a manner that can be
adapted to account for differences in plant technology, concepts of
operations, and operator roles in the
[[Page 23661]]
fulfillment of design-specific safety functions. The NRC would require
under proposed Sec. 57.424(a)(1) that facility licensees implement an
SAT-based training program for the initial training of operator and
senior operator applicants. The program would need to be adequate to
ensure that applicants will be capable of performing the duties
necessary to both protect public health and safety and maintain plant
safety functions. The NRC would also require NRC approval of the
training program, including the change process, which would state when
NRC approval is needed for subsequent changes.
Examinations provide a means of assessing that individuals have
achieved a level of knowledge and ability that is sufficient to carry
out assigned duties as specifically licensed operators or senior
operators in a manner that is safe and reliable. The NRC would adapt
the requirements of Sec. Sec. 55.40, 55.41, 55.43, and 55.45 in
proposed Sec. 57.424(b) to require that facilities establish and
implement an initial examination program. However, a key difference
from the comparable requirements of part 55 is that facilities would
have the flexibility to propose, subject to NRC approval, the
examination methods and the criteria for use in assessing applicant
performance. Such examination programs (including those used within the
scope of requalification training) would need to provide for acceptable
levels of both test validity and test reliability in order to be
considered acceptable. The NRC intends that guidance would be available
to facilitate the review of licensing examination programs that are
proposed by facility licensees and that, following NRC approval,
initial examination programs will be subject to an appropriate change
control process. Furthermore, the NRC would allow facility licensees
the option to administer their own NRC-approved licensing examinations.
The NRC would continue to exercise appropriate oversight of the
program, make operator licensing decisions based upon the examination
results, and reserve the right to administer the examinations in lieu
of permitting the facility to do so.
Requalification training programs provide for the continuing
training and examination of specifically licensed operators and senior
operators to ensure that they maintain the knowledge and abilities
needed to support the safe and reliable performance of job duties
following the completion of an initial training and examination
program. The NRC would adapt the requirements of Sec. 55.59 in
proposed Sec. 57.424(c) to require that facilities implement both an
SAT-based requalification training program and a biennial
requalification examination program. However, a notable difference from
the biennial requalification examinations required under part 55 is
that facility licenses would be able to propose examination methods and
criteria to be used in assessing satisfactory performance as part of
their replated programs. The NRC intends that guidance would be
available to facilitate the review of the requalification examination
programs that are proposed by facility licensees and that, following
NRC approval, requalification examination programs would be subject to
an appropriate change control process.
For examinations to provide valid assessments of the knowledge and
abilities of individuals, the examinations must remain free from
compromises that could affect their underlying integrity. The NRC would
adapt the requirements of Sec. 55.49 in proposed Sec. 57.424(d) to
require that examinations and related activities remain free from any
compromise that might affect the integrity of the examination process.
Simulators provide a valuable means of training and evaluating
plant operators, and the NRC is specifically authorized under the
section 306 of the Nuclear Waste Policy Act of 1982, as amended to
establish regulations for the use of simulators within such context.
The NRC would adapt the requirements of Sec. 55.46 in proposed Sec.
57.424(e) to address the use of simulation facilities for training,
examinations, and applicant experience requirements, as well as to
address the maintenance of simulator fidelity. However, the
requirements of proposed part 57 would not mandate that full scope,
plant-referenced simulators be used and would allow the use of
alternative simulation facilities consisting of, for example, partial
scope simulators or the plant itself, provided that all associated
requirements can be demonstrated to be met using alternative approaches
and methods.
There may be situations in which applicants for operator or senior
operator licenses have previous training and experience that justify
reducing some, or all, of the initial examination requirements. The NRC
would adapt the high-level requirements of Sec. 55.47, ``Waiver of
examination and test requirements,'' in proposed Sec. 57.424(f), to
support the evaluation of requests for waivers of examination
requirements.
For licensed operators and senior operators to perform their
assigned duties safely and reliably, it is essential that they perform
those duties frequently enough to maintain proficiency. The NRC would
adapt the requirements of Sec. 55.53(e) and (f) in proposed Sec.
57.424(g) to require that specifically licensed operators and senior
operators maintain proficiency and, if proficiency is not maintained,
regain proficiency prior to resuming licensed duties. However, a major
difference from the part 55 requirements is that the facility licenses
would propose their own program for operator proficiency, subject to
NRC approval. Similar to training and examination program changes,
following NRC approval, proficiency programs would also be subject to
an appropriate change control process.
As the holders of specific licenses, licensed operators and senior
operators would be subject to license conditions on an individual basis
to ensure that the basis upon which the licenses were issued remains
valid. The NRC would adapt the requirements of Sec. 55.53 in proposed
Sec. 57.425, ``Conditions of operator and senior operator licenses,''
to require appropriate conditions of licenses for specifically licensed
operators and senior operators. However, in contrast with the
requirements of Sec. 55.53(e) and (f), the NRC would allow certain
aspects of operator proficiency to be addressed by an NRC-approved
facility proficiency program.
Licenses for specifically licensed operators and senior operators
under part 55 are currently issued by the NRC and must remain subject
to modification or revocation. The NRC would adapt the requirements of
Sec. Sec. 55.51, ``Issuance of licenses,'' and 55.61, ``Modification
and revocation of licenses,'' in proposed Sec. 57.426, ``Issuance,
modification, and revocation of operator and senior operator
licenses,'' to address the issuance, modification, and revocation of
licenses issued to specifically licensed operators and senior
operators.
Finally, proposed Sec. 57.427 would address conditions that would
cause licenses issued to specifically licensed operators and senior
operators to expire.
Section 306 of the Nuclear Waste Policy Act of 1982 authorizes and
directs the NRC to, in part, issue regulations and guidance that
address the training and qualifications of civilian nuclear power plant
operators, supervisors, technicians, and other appropriate operating
personnel. The NRC implements this in part 50 through the requirements
of Sec. 50.120, ``Training and qualification of nuclear power plant
personnel.'' The NRC would adapt under proposed Sec. 57.429 the
requirements of Sec. 50.120 for specific categories of nuclear plant
personnel.
[[Page 23662]]
This list of personnel would be modified from the list of positions in
Sec. 50.120 to be more applicable to facilities licensed under
proposed part 57. The NRC would require under proposed Sec. 57.429
that SAT-based training programs would be established within a
timeframe based upon when the associated personnel would be needed to
support facility-specific needs. The training programs would include
the training and qualification of plant personnel in the general
categories of supervisors, technicians, and other appropriate operating
personnel. The category of supervisors would reflect on-shift
supervisors for the licensed operators, similar to the current
classification in Sec. 50.120(b)(2)(ii). The facility licensee would
not be required to seek NRC approval of a training program prior to
usage. However, the facility licensee would be required to accommodate
NRC inspection of the training programs.
R. Subpart Q--Reporting and Other Administrative Requirements
Proposed part 57 would address various reporting and administrative
requirements in subpart Q.
Proposed Sec. 57.430, ``Maintenance of records, making of
reports,'' would require the maintenance of records and the making of
various reports by the licensee to the NRC. These requirements would be
largely equivalent to Sec. 50.71(a), (c), and (d).
Proposed Sec. 57.430(f) would require licensees to notify the NRC
of successful completion of any startup testing of a nuclear reactor to
support the assessment of annual fees under 10 CFR part 171, ``Annual
Fees for Reactor Licenses and Fuel Cycle Licenses and Materials
Licenses, Including Holders of Certificates of Compliance,
Registrations, and Quality Assurance Program Approvals and Government
Agencies Licensed by the NRC.'' The assessment of annual fees normally
commences upon completion of those testing activities. With respect to
annual fees, the NRC recently modified its annual fee regulations to
address differences between the current fleet of large operating
reactors and potential future smaller reactors. In the Fiscal Year 2023
final fee rule, the NRC amended its annual fee regulations to (1) be
technology-inclusive by expanding the applicability of the small
modular reactor variable fee structure to include non-LWR small modular
reactors (previously it was limited to LWR small modular reactors); and
(2) establish an additional minimum fee and variable rate applicable to
smaller reactors.
Proposed Sec. 57.435, ``Reporting requirements,'' would establish
requirements for immediate notifications by licensees under proposed
part 57. These requirements would be equivalent to Sec. 50.72,
``Immediate notification requirements for operating nuclear power
reactors,'' with minor changes proposed to make the reporting criteria
technology-inclusive and remove the notification of the NRC Operations
Center using the Emergency Notification System.
Proposed Sec. 57.440, ``Licensee event report system,'' would
require each holder of an OL under proposed part 57 to have a licensee
event report system. These requirements would be equivalent to Sec.
50.73, ``Licensee event report system,'' with minor changes to remove
requirements of specific reactor technologies.
Proposed Sec. 57.445(a) and (b) would require periodic reporting
of the quantity of radionuclides released to unrestricted areas in
liquid and gaseous effluents, and doses to members of the public.
Proposed Sec. 57.445, ``Reports of radiation exposure to members of
the public,'' would be similar to Sec. 50.36a(a)(2).
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
This proposed rule would make conforming changes throughout 10 CFR
chapter I by adding ``and part 57'' where appropriate to account for
the addition of the proposed part 57. In addition, this proposed rule
would revise Sec. 2.340(d) in three places to correct the
manufacturing license reference from subpart C to subpart F.
B. 10 CFR part 26
1. Introduction
The NRC proposes to include fitness-for-duty (FFD) requirements for
microreactors and other reactors with comparable risk profiles. This
proposed rule would establish a technology-inclusive, risk-informed,
and performance-based approach for the application of drug and alcohol
testing and fatigue management requirements for facilities licensed
under proposed part 57. The proposed rule would add a new subpart P,
``Fitness-for-Duty Programs for Facilities Licensed Under 10 CFR part
57,'' in 10 CFR part 26, ``Fitness for Duty Programs,'' and make
conforming changes to existing part 26 provisions. The proposed rule
would also provide the option for certain reactors with comparable risk
profiles to implement an FFD program of their specification (i.e., one
that is not subject to the requirements of part 26) if they meet
applicable human reliability criteria.
The NRC would use operating experience to provide regulatory
flexibility to proposed part 57 licensees and other entities in the
part 26 framework to help support a licensee's or other entity's
response to changes in societal drug use, drug testing technologies and
processes, and FFD program performance. The flexibility would also help
in FFD program implementation because of the wide variety of staff
sizes anticipated at nuclear plants licensed under proposed part 57 and
the geographically remote locations in which these nuclear plants may
be sited.
Licensees and other entities would have the option to implement one
of three types of FFD programs at their facilities: one that meets all
the requirements of part 26 except subpart K, ``FFD Program for
Construction,'' of part 26 and proposed subpart P; one that meets the
requirements in proposed subpart P; or an FFD program of their
specification. These requirements would be commensurate with the
potential radiological consequences of reactors licensed under proposed
part 57, and the options available to a licensee would be dependent on
the human reliability considerations associated with the operation of
their facilities. This risk-informed regulatory strategy would be
consistent with the current part 26, which provides a comprehensive set
of deterministic requirements for licensees and other entities at
facilities that are operating plus a more flexible framework under
subpart K for nuclear power reactors under construction.
Proposed subpart P to part 26 would be essentially equivalent to
the requirements in subpart K as supplemented by select requirements
from subparts E, ``Collecting Specimens for Testing,'' of part 26, and
the requirements in subparts A, ``Administrative Provisions,'' I,
``Managing Fatigue,'' and O, ``Inspection, Violations, and Penalties,''
of part 26. These requirements would help deter individuals subject to
proposed subpart P from drug and/or alcohol use and from being impaired
from any cause including fatigue. These requirements also would help
licensees and other entities identify individuals as users of impairing
substances and demonstrate compliance with Sec. 26.23, ``Performance
objectives.''
Proposed subpart P of part 26 would enable a part 57 licensee or
other entity
[[Page 23663]]
to implement innovative drug testing technologies and behavior
observation techniques while continuing to demonstrate compliance with
the part 26 performance objective in Sec. 26.23(b) of providing
reasonable assurance that individuals are not under the influence of
any substance or mentally or physically impaired from any cause, which
in any way adversely affects their ability to safely and competently
perform assigned duties. These technologies would include drug testing
of oral fluid, urine, and hair specimens and non-invasive portal area
screening instruments that would passively test for drugs, alcohol, or
both. Part of the basis to enable the use of innovative drug and
alcohol testing technologies, should they become available, is to
maintain FFD program effectiveness should the staff size at a part 57
nuclear plant be small and challenge the effective implementation of
the behavioral observation and drug and alcohol testing programs. Also,
a proposed part 57 nuclear plant that is sited at a geographically
remote location could present additional challenges not encountered by
traditional LWR facilities licensed under part 50 or 52, such as:
efficiency of postal services for shipping and controlling biological
specimens; proximity to drug and alcohol collection facilities that are
reasonably equivalent to that described in subpart E of part 26;
availability of internet and cellular services to enable same-time
discussions among the Medical Review Officer (MRO), donor, and
laboratory; accessibility to substance abuse treatment services
described in subpart H of part 26; and proximity to an MRO (or
management and clinical staff) to evaluate potential impairment caused
by fatigue and/or substance use or abuse, for-cause and post-event
occurrences, and the individual's potential to return to duty.
A proposed part 57 nuclear plant that is sited in a geographically
remote location and has a small staff size may present implementation
challenges and the potential for small group dynamics that could have
the potential to impact FFD program effectiveness. For example,
behavioral observation may be less effective at a plant that has a
small staff size, which can be subject to greater impacts from
groupthink and other biasing factors.\3\ As such, alternative
approaches to behavior observation programs, such as supplementing
onsite behavior observation activities with video-based observation by
individuals separate from the onsite work unit, could serve to mitigate
potential issues by bringing in independent and objective perspectives.
---------------------------------------------------------------------------
\3\ Groupthink is a psychological phenomenon that can emerge and
is particularly prevalent among cohesive and insulated groups that
experience high levels of decisional stress. Groupthink can impact
individuals' willingness to speak out against practices they deem
unsafe, for fear of deviating from group norms. Research also
indicates that groups make riskier decisions than individuals acting
alone due to the diffusion of responsibility among group members.
For additional information; see, e.g., Irene W[aelig]r[oslash],
Ragnar Rosness, and Stine Skaufel Kilska, ``Human performance and
safety in Arctic environments,'' SINTEF (2018); and see, e.g.,
Mannion and Thompson, ``Systematic biases in group decision-making:
implications for patient safety,'' International Journal for Quality
I Health Care, Vol. 26, No. 6 (2014): 606-612.
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Additionally, random testing may be less effective when applied to
small staff sizes, because it may be easier for staff to communicate
and predict when individuals will be subject to drug and alcohol
testing. Furthermore, if a facility is sited in a remote location,
program implementation could be challenged by the following factors:
limited mail services to laboratories certified by the U.S. Department
of Health and Human Services (HHS), availability of local clinical or
medical options for treatment and determinations of fitness by an MRO
or Substance Abuse Expert, and use of offsite drug and alcohol
collection facilities.
The increased potential for small staff sizes to impact FFD policy
compliance would necessitate additional flexibilities be provided to
implement various FFD program elements. The NRC would require that
facilities with small staff sizes that cannot implement random drug and
alcohol testing without predictability, use a consortium/third-party
administrator (C/TPA) to include the workers from multiple licensees or
other entities into a combined random testing pool under Sec.
26.907(b)(2)(vi). Use of a C/TPA would significantly improve the
effectiveness of the random testing programs of sites with small worker
populations and ensure that individuals would not be able to predict
whether random testing would be conducted in a given period of time.
Use of C/TPAs is not new in Federally-regulated testing, as the U.S.
Department of Transportation has employed the use of C/TPAs in specific
modal administrations, such as the Federal Motor Carrier Safety
Administration under 49 CFR part 382, ``Controlled Substances and
Alcohol Use and Testing,'' which, in part, covers independent owner-
operator truck drivers that must be drug and alcohol tested. The U.S.
Department of Transportation requirements in 49 CFR part 40,
``Procedures for Transportation Workplace Drug and Alcohol Testing
Programs'' also enables the use of C/TPAs to perform a variety of
functions for employers, such operating random testing programs, and
contracting with specimen collection sites and HHS-certified
laboratories for services.
Another flexibility would be proposed Sec. 26.907(g)(2), where the
NRC would enable the virtual collection of oral fluid specimens for
drug and alcohol testing at facilities that must use a C/TPA to
implement random testing under Sec. 26.907(b)(2)(vi). These sites
would have small staff sizes and could be in remote locations where
accessing an in-person specimen collector might be difficult, untimely,
and/or costly. Because all aspects of an oral fluid collection would be
directly observed by the specimen collector, a video teleconference
could accomplish many key elements of the collection process. The use
of video teleconference technology would not be new to the NRC, as some
clinicians complete other required evaluations, such as performing a
psychological assessment under the personnel access authorization
requirements in Sec. 73.56(e)(4) or a determination of fitness
performed under Sec. 26.189(b) by a Substance Abuse Expert when
potentially disqualifying FFD information is discovered about an
individual that is subject to 10 CFR part 26. In addition, existing
Sec. 26.31(b)(1)(iii) enables the use of a monitor to assist a
specimen collector in completing aspects of a urine collection when a
trained collector is not able to complete the activity, and existing
Sec. 26.109(b)(1) permits a hydration monitor to observe a donor
during the shy bladder process in lieu of the collector conducting the
activity. In both cases, the monitor must receive information from the
collector on his or her responsibilities.
Also, the NRC would establish a change control requirement to allow
a licensee or other entity to change its subpart P FFD program while
ensuring that FFD program effectiveness is maintained.
2. Proposed Changes to Part 26, Subparts A Through E, I, and N
Proposed Sec. 26.3(d) is the applicability paragraph for
contractor/vendors (C/Vs) that implement FFD programs or program
elements, to the extent that the licensees and other entities specified
in Sec. 26.3(a) through (c) rely on those C/V FFD programs or program
elements to satisfy the requirements of part 26. Section 26.3(d) would
be amended to address proposed part 57 licensees and other entities in
proposed Sec. 26.3(f).
[[Page 23664]]
Proposed Sec. 26.3(f) would place part 57 licensees or other
entities within the scope of part 26. For applicants for or holders of
a CP or OL under proposed part 57, except a holder of an ML, proposed
Sec. 26.3(f)(1) would require the FFD program to be implemented no
later than the start of construction activities. Proposed Sec.
26.3(f)(2) would require the holder of an ML under proposed part 57 to
implement its FFD program before commencing activities that assemble a
reactor. All three licensees would have three FFD program options:
implement all the requirements of part 26 except subparts K and P, the
requirements in proposed subpart P, or an FFD program of their
specification. Proposed Sec. 26.3(f)(3) would provide the criteria by
which licensees and other entities under proposed part 57 could
implement an FFD program of their specification. That criterion would
be if the licensee's or other entity's reactor manufactured,
constructed, or operated under a part 57 license would not require
operator action to maintain the reactor within the criterion of Sec.
57.25(a) or a credible operator or maintenance error could not result
in exceeding that criterion.
Current Sec. 26.4, ``FFD program applicability to categories of
individuals,'' describes FFD program applicability to categories of
individuals. These categories are based on the duties,
responsibilities, and the types of access an individual may possess.
The NRC proposes to amend Sec. 26.4 to include licensees and other
entities described in proposed Sec. 26.3(f). The NRC expects that not
all categories of individuals described in current Sec. 26.4 would be
applicable to all proposed part 57 facilities.
Section 26.4(a) requires individuals who are granted unescorted
access to nuclear power reactor protected areas by the licensees in
Sec. 26.3(a) and, as applicable, (c) and perform the duties listed in
26.4 to be subject to an FFD program that meets all of the requirements
of part 26, except subpart K. The NRC would amend Sec. 26.4(a) to
except proposed subpart P as well as subpart K.
Section 26.4(a)(1) and (a)(4) would be amended to account for the
possibility that certain individuals may perform or direct the
performance of operational and maintenance activities from a remote
facility (for example, a remote control station) for licensees or other
entities licensed under proposed part 57. The framework of the current
part 26 does not account for individuals who perform operating and
maintenance duties at remote facilities. Although current Sec.
26.4(a)(1) does not limit the operating of applicable SSCs to onsite
operating, Sec. 26.5, ``Definitions,'' limits the definition of
``Maintenance,'' for the purposes of Sec. 26.4(a)(4), to include only
``onsite maintenance activities.'' In the 2008 part 26 final rule
preamble, the NRC explained that the work hour requirements apply to
those individuals who perform maintenance activities within the
licensee's owner-controlled area. Furthermore, regarding the direction
of applicable operations and maintenance activities, current Sec.
26.4(a)(1) and (4) address only individuals who perform ``onsite
direction.''
Under the proposed amendments to part 26, the limitation of
``onsite'' activities to those performed within the owner-controlled
area would still apply to facilities licensed under part 50 or 52.
However, for licensees and other entities described in proposed Sec.
26.3(f), the NRC would remove the ``onsite'' limitation to include
activities performed both within the owner-controlled area as well as
operations and maintenance duties performed at remote facilities where
safety-significant systems and components are expected to be operated
within the design basis of the nuclear plant.
In the 2008 part 26 final rule, the purpose of limiting
``directing'' activities to those ``directing'' activities that are
conducted onsite was to avoid requiring work hour controls for
individuals performing incidental duties, consistent with Sec.
26.205(b)(5), from an offsite location in instances where those duties
might be considered to be ``directive'' in nature. Under the proposed
amendments to part 26, the exclusion of incidental duties while
calculating work hours would still be applicable for licensees and
other entities licensed under proposed part 57. However, for these
licensees and other entities, beyond instances of incidental duties,
the direction of operations and maintenance activities associated with
safety-significant SSCs, when performed at remote facilities, would be
considered in an equivalent fashion as direction performed at non-
remote facilities, for the purposes of administering work hour
controls.
Section 26.4(b) requires individuals who are granted unescorted
access to nuclear power reactor protected areas by the licensees in
Sec. 26.3(a) and, as applicable, (c) and who do not perform the duties
described in Sec. 26.4(a), to be subject to an FFD program that meets
all of the requirements of part 26, except Sec. Sec. 26.205, ``Work
hours,'' through 26.209, ``Self-declarations,'' and subpart K. The NRC
would amend Sec. 26.4(b) to except proposed subpart P as well as
subpart K. Proposed Sec. 26.4(b) also would include in an FFD program
individuals who are granted unescorted access to the protected area of
a facility licensed under proposed part 57 and do not perform or direct
the performance of the duties described in Sec. 26.4(a). This
requirement would contribute to the defense in depth regulatory
framework that helps provide that individuals who have unescorted
access are fit for duty, trustworthy, and reliable.
Section 26.4(c) requires individuals who are required by a licensee
in Sec. 26.3(a) and, as applicable, (c) to physically report to the
licensee's Technical Support Center or Emergency Operations Facility by
licensee emergency plans and procedures to be subject to an FFD program
that meets all of the requirements of part 26, except Sec. Sec. 26.205
through 26.209 and subpart K. The NRC would amend Sec. 26.4(c) to
except proposed subpart P as well as subpart K.
The NRC also would amend Sec. 26.4(c) to include in an FFD program
individuals who are assigned to physically report to the proposed part
57 licensee's emergency response facility (or facilities) or
participate remotely in emergency response activities, and individuals
without unescorted access to the part 57 facility who, remotely or
otherwise, make decisions and/or direct actions regarding plant safety
or security. Proposed part 57 nuclear plants may rely upon offsite
facilities to fulfill the role of a Technical Support Center or
Emergency Operations Facility. Therefore, the proposed rule would
account for such offsite facilities or remotely performed activities.
Further, the use of personnel to operate systems and components,
maintain and surveil SSCs, and respond to plant conditions and security
events may be different than those included in the Technical Support
Center or Emergency Operations Facility team for power reactors
currently licensed under part 50 or part 52.
For the individuals whose duties for the licensees and other
entities in Sec. 26.3(c) require the individuals to have the types of
access or perform the activities listed in Sec. 26.4(e)(1) through (6)
at the location where the nuclear plant will be constructed and
operated, current Sec. 26.4(e) requires them to be subject to an FFD
program that satisfies all the requirements of part 26 except subparts
I and K. The NRC would amend Sec. 26.4(e) to except proposed subpart P
as well as subparts I and K. The NRC would also amend Sec. 26.4(e) to
include in an FFD program the individuals whose duties for the
[[Page 23665]]
licensees and other entities in Sec. 26.3(f) require the individuals
to have the types of access or perform the activities listed in Sec.
26.4(e)(1) through (6) or perform construction activities as defined in
Sec. 26.5.
The proposed rule would amend Sec. 26.4(f) to require individuals
who construct or direct the construction of safety- or security-related
SSCs at facilities licensed under proposed part 57 to be subject to an
FFD program under proposed subpart P of part 26 or an FFD program that
demonstrates compliance with all the requirements of part 26 except for
subparts I, K, and P of part 26, unless the licensee or other entity
meets the criteria in proposed Sec. 26.3(f)(3) and subjects these
individuals to an FFD program of its own specification.
Section 26.4(g) is the applicability paragraph for FFD program
personnel (e.g., the FFD manager, MRO, and technicians) and persons who
perform access authorization determinations (e.g., the licensee- or
other entity-designated Reviewing Official). This section would be
amended to address proposed part 57 licensed facilities. Specifically,
a proposed part 57 licensee or other entity would use FFD program
personnel to implement its FFD program as well as other assigned
individuals who are not involved in the day-to-day operations of the
program to implement specific elements of its FFD program, such as the
collection of a specimen for drug or alcohol testing. These individuals
would be held accountable for program implementation, including
consistent implementation of protections afforded to all individuals
subject to the FFD program.
Section 26.4(h) would be amended to include proposed subpart P of
part 26 unless the licensee or other entity meets the criteria in
proposed Sec. 26.3(f)(3) and subjects these individuals to an FFD
program of its own specification.
The NRC proposes to include several new definitions in Sec. 26.5
and amend some existing definitions. The NRC is proposing to add a
definition for ``Biological marker.'' The proposed definition would be
consistent with ``Biomarker'' defined by the HHS in its Mandatory
Guidelines for Federal Workplace Drug Testing (HHS Guidelines) using
oral fluid as the biological specimen to be tested (84 FR 57554;
October 25, 2019). However, the proposed definition for Sec. 26.5
would add that the endogenous substance used to validate that the
biological specimen ``was produced by the donor'' because subpart P of
part 26 proposes to have the MRO evaluate any discrepant biological
marker identified in a biological specimen collected from a donor.
The NRC is proposing a definition for the word ``Change'' as used
in proposed Sec. 26.903(c), ``FFD program change control,'' process.
The proposed definition would be consistent with the definition of
``Change'' for a part 50 or 52 licensee's emergency plans in Sec.
50.54(q)(1)(i).
The NRC is proposing a definition for ``Consortium/third-party
administrator,'' which would be used in Sec. 26.907(b)(2)(vi), with
respect to administering the random testing pool and random testing
selections for licensees and other entities with facilities with small
staff sizes. A C/TPA also could provide access to, for example,
services of medical review officers, substance abuse experts, employee
assistance programs, and HHS-certified laboratories under contract to
perform drug testing.
The NRC proposes to revise the definition of ``Constructing or
construction activities'' to clarify that for licensees or other
entities in proposed Sec. 26.3(f), the definition of ``Construction''
would be that in proposed Sec. 57.3.
The definitions of ``Contractor/vendor'' (C/V) and ``Other entity''
would be revised to make them applicable to proposed part 57 licensees.
A holder of an ML under part 57 could be a C/V under the proposed C/V
definition.
The NRC is proposing a definition for ``Illicit substance'' because
this phrase would be used in proposed subpart P of part 26 and would
address substances that cause impairment and possible addiction but
would not be an ``illegal drug'' as defined in Sec. 26.5. This
proposal is based on operating experience where individuals have
admitted to using common household, non-drug substances to achieve a
high or satisfy an addiction. These common household items include, but
are not limited to nitrous oxide, butane, propane, glue, paint vapors,
lighter fluid, nail polish remover, degreasers, permanent markers, and
methyl alcohol (which is found in hand sanitizer and mouthwash).
The NRC is proposing a definition for ``Reduction in FFD program
effectiveness'' because this phrase, similar to the proposed definition
for ``Change,'' would be used in proposed Sec. 26.903(c). The proposed
definition is generally consistent with the definition of ``Reduction
in effectiveness'' provided for emergency plans in Sec.
50.54(q)(1)(iv).
The proposed rule would make the current definition of ``Reviewing
official'' applicable to those licenses and other entities in proposed
Sec. 26.3(f).
The current part 26 definition of ``Safety-related structures,
systems, and components'' would be amended to use the NRC's proposed
definition in Sec. 57.3 for the part 57 licensees and other entities
described in proposed Sec. 26.3(d) and (f).
The NRC would amend the definition of ``Security-related SSCs'' in
Sec. 26.5 to make it applicable to a licensee or other entity
described in proposed Sec. 26.3(d) and (f).
The NRC proposes a definition for ``Special nuclear material'' that
would refer to the definition in Sec. 70.4, ``Definitions,'' to ensure
consistency.
The NRC is proposing a revision of the definition of ``Unit
outage'' to account for the potential use of nuclear plants for
purposes other than electricity generation.
The proposed rule would amend Sec. 26.8, ``Information collection
requirements: OMB approval,'' to reflect the addition of proposed
subpart P to part 26.
Section 26.21, ``Fitness-for-duty program,'' an applicability
statement for part 26 FFD programs, would be amended to include
licensees and other entities described in proposed Sec. 26.3(f) that
choose to implement an FFD program that implements all part 26
requirements, except those in subparts K and P of part 26, and do not
implement an FFD program of their own specification if they meet the
criteria in proposed Sec. 26.3(f)(3).
The proposed rule would amend Sec. 26.35(c)(3) to include a
reference to proposed Sec. 26.906(b)(2)(vii), which would ensure that
licensees and other entities take immediate action upon receiving
notice from the EAP that an individual's condition or actions pose or
have posed an immediate hazard to themself or others.
Section 26.51, ``Applicability,'' would be amended to apply to
licensees and other entities described in proposed Sec. 26.3(f) that
elect not to implement the requirements in proposed subpart P of part
26 for the categories of individuals in Sec. 26.4, and do not
implement an FFD program of their own specification if they meet the
criteria in proposed Sec. 26.3(f)(3).
Section 26.53(e) and (g) through (i), which are general provisions
for granting and maintaining authorization, would be amended to apply
to licensees and other entities described in proposed Sec. 26.3(f).
Section 26.63(d), a suitable inquiry requirement, would be amended
to apply to licensees and other entities described in proposed Sec.
26.3(f).
[[Page 23666]]
Section 26.73, ``Applicability,'' the applicability statement for
subpart D of part 26, would be amended to apply to licensees and other
entities described in proposed Sec. 26.3(f) that elect not to
implement the requirements in proposed subpart P of part 26 for the
categories of individuals in Sec. 26.4 and do not implement an FFD
program of their own specification if they meet the criteria in
proposed Sec. 26.3(f)(3).
Section 26.81, ``Purpose and applicability,'' the purpose and
applicability statement for subpart E of part 26, would be amended to
apply to licensees and other entities described in proposed Sec.
26.3(f) that elect not to implement the requirements in proposed
subpart P of part 26 for the categories of individuals in Sec. 26.4
and do not implement an FFD program of their own specification if they
meet the criteria in proposed Sec. 26.3(f)(3). The subpart E
requirements to be implemented are listed in proposed Sec.
26.907(c)(2)(i) and (c)(2)(ii) and (c)(3).
The NRC proposes to revise Sec. 26.97(a) and (b) to enable the
virtual collection of oral fluid specimens for drug and alcohol
testing, as would be permitted under proposed Sec. 26.907(g)(2). The
NRC also would amend Sec. 26.97(a) and (b) to update the oral fluid
specimens collection process requirements.
Section 26.201, ``Applicability,'' the applicability statement for
subpart I of part 26, would be amended to apply to licensees and other
entities described in proposed Sec. 26.3(f). Also, the applicability
statement would be divided into two paragraphs for clarity.
The NRC proposes to add Sec. 26.202, ``General provisions for
facilities licensed under part 57,'' for licensees or other entities
described in proposed Sec. 26.3(f) that elect to implement the
requirements in subpart I of part 26 in accordance with proposed Sec.
26.904, ``FFD program requirements.'' Proposed Sec. 26.202 would
establish requirements equivalent to those in current Sec. 26.203,
``General provisions,'' which is applicable to part 50 and 52
licensees. The NRC would add the separate Sec. 26.202 because Sec.
26.203 would refer to various requirements under subpart B of part 26,
which would not be applicable to facilities licensed under proposed
part 57 that implement proposed subpart P of part 26.
Additionally, proposed Sec. 26.202(c), ``Training and
assessments,'' unlike current Sec. 26.203(c), ``Training and
examinations,'' would not include a comprehensive examination
requirement because trainee assessment is conducted as part of an SAT
that would be required as proposed under the FFD program training
requirements in proposed Sec. 26.908, ``FFD program training.''
Proposed changes in Sec. Sec. 26.205, 26.207, ``Waivers and
exceptions,'' and 26.211, ``Fatigue assessment,'' would add references
to new requirements in subparts I and P of part 26 that would be
applicable specifically to licensees and other entities in proposed
Sec. 26.3(f). The NRC would not change the specific provisions for
work hour requirements in current Sec. 26.205(d).
Proposed changes to Sec. Sec. 26.207(a)(1)(ii) and 26.211(b) would
allow licensees and other entities in proposed Sec. 26.3(f) to perform
face-to-face assessments to support the approval of work hour control
waivers and the conduct of fatigue assessments, respectively, using
electronic communications. These proposals would allow supervisors to
conduct such assessments from a remote location under appropriate
circumstances. Such remotely conducted assessments would need to be
supported by someone who is present in-person with the individual being
assessed and who is trained in accordance with the requirements of
either Sec. 26.29, ``Training,'' and Sec. 26.203(c) or proposed Sec.
26.908 and Sec. 26.202(c). The reasoning for these proposals and the
associated need for in-person support to augment electronic
communications is addressed further in the preamble discussion of
proposed Sec. 26.919, ``Suitability and fitness determinations.''
Proposed Sec. 26.709, ``Applicability,'' would make the
recordkeeping and reporting requirements in subpart N, ``Recordkeeping
and Reporting Requirements,'' of part 26 applicable to licensees and
other entities of facilities licensed under proposed part 57 that elect
not to implement the requirements in proposed subpart P of part 26 and
do not implement an FFD program of their own specification if they meet
the criteria in proposed Sec. 26.3(f)(3).
Proposed Sec. 26.711(c) and (d) would be amended to make these
requirements applicable to licensees or other entities described in
proposed Sec. 26.3(f). Section 26.711(c) provides protection to
individuals subject to part 26 by enabling an individual's right to
review FFD-related information and correct any inaccurate or incomplete
information. Section 26.711(d) requires, in part, that any FFD-related
information shared with other licensees or other entities is correct
and complete.
3. Proposed Requirements for Part 26, Subpart P
The proposed rule would add a new subpart P to part 26 that would
provide alternative FFD requirements for licensees and other entities
licensed under proposed part 57.
Proposed Sec. 26.901, ``Applicability,'' would make subpart P of
part 26 applicable to part 57 licensees and other entities, at their
discretion. As provided for in proposed Sec. 26.3(f), a part 57
licensee or other entity that does not elect to implement an FFD
program that demonstrates compliance with the requirements of proposed
subpart P must implement an FFD program that demonstrates compliance
with all part 26 requirements, except for those requirements in
subparts K and P, or an FFD program of their specification if they meet
the criteria in proposed Sec. 26.3(f)(3).
Proposed Sec. 26.903(a), ``FFD program description,'' would
require a proposed part 57 applicant to include a description of its
FFD program in its FSAR, required by proposed subparts C and D of part
57. Unlike an application for a license, a description of an FFD
program would not receive NRC review for possible approval. The
applicant would provide the NRC with information about the applicant's
proposed FFD program to inform the NRC's inspection program and to
demonstrate that the FFD program would be effectively implemented
before a licensee or other entity commences any activity making
individuals at the NRC-licensed facility subject to the FFD program.
Proposed Sec. 26.903(a)(1) would require a discussion that informs
the NRC of the applicability of the applicant's FFD program to
individuals as specified in Sec. 26.4. This description should
summarize any key differences between the staff at the site and any
remote facility and the categories of individuals in Sec. 26.4. The
principal purpose of providing this description would be to inform the
NRC of any substantial differences in the applicability of the FFD
program to the categories of individuals in Sec. 26.4. Proposed Sec.
26.903(a)(1) would also require the FFD program description to describe
how the program would be implemented at a facility authorized to
assemble or perform non-operational testing of a manufactured reactor
under an ML issued under proposed part 57, if applicable.
Proposed Sec. 26.903(a)(2) would require a description of the drug
and alcohol testing and fitness determination process to be implemented
through the licensee's or other entity's procedures, including the
collection and testing facilities to be used, biological specimens to
be collected and tested, and sanctions to be imposed for FFD policy
violations. This process would
[[Page 23667]]
include how individuals who test positive for a drug or alcohol would
be evaluated before being afforded unescorted access to the protected
area to perform or direct those duties or responsibilities making them
subject to the FFD program.
Proposed Sec. 26.903(b), ``FFD program implementation and
availability,'' would establish the longevity of the FFD program.
Unlike the current part 26 regulations, Sec. 26.903(b) would state
that an FFD program is not applicable during decommissioning under
proposed part 57. Proposed Sec. 26.903(b) would require the holder of
a manufacturing license under proposed part 57 to maintain its FFD
program until expiration of the manufacturing license.
In proposed Sec. 26.903(c), ``FFD program change control,'' the
NRC proposes a change control requirement for subpart P of part 26 FFD
programs. Licensees and other entities would be required to demonstrate
compliance with certain requirements before implementing changes to
their FFD programs. Change control would rely on the licensee or other
entity maintaining its procedures in a manner that details how its FFD
program is to be implemented while incorporating changes, with
documentation that justifies the changes to support audits and NRC
inspection.
Proposed Sec. 26.903(c)(1) would permit the licensee or other
entity to implement changes to its FFD program if the licensee or other
entity performs and retains an analysis demonstrating that the changes
do not reduce the effectiveness of the FFD program or the changes were
necessitated or justified by a change to part 26, laboratory processes,
or guidance issued by the HHS or NRC. The change control requirement
would enable flexibility in program implementation should the NRC or
HHS change its drug testing procedures (as implemented by the licensee
or other entity through its procedures) in response to changes in
societal substance abuse or drug testing technologies.
Proposed Sec. 26.903(c)(2) would require that if a change reduces
FFD program effectiveness, then the licensee or other entity must
implement a mitigating strategy so the FFD program, as revised, would
continue to demonstrate compliance with the performance objectives in
Sec. 26.23 and not result in a reduction in program effectiveness.
Proposed Sec. 26.903(c)(3) would prohibit the use of the change
control process to reduce the minimum panel of drugs to be tested and
would reference the drugs listed in proposed Sec. 26.907(c)(1).
Proposed Sec. 26.907(c)(1) would reference current Sec. 26.31(d)(1),
which states that, at a minimum, licensees and other entities shall
test for marijuana metabolite, cocaine metabolite, opioids (codeine,
morphine, 6-acetylmorphine, hydrocodone, hydromorphone, oxycodone, and
oxymorphone), amphetamines (amphetamine, methamphetamine,
methylenedioxymethamphetamine, and methylenedioxyamphetamine),
phencyclidine, and alcohol. The testing of these drugs and drug
metabolites and alcohol is necessary for the FFD program to remain
effective.
Also, there is no proposed subpart P requirement stating that this
panel of drugs and drug metabolites needs to consist of only scheduled
drugs. This flexibility would account for the situation where an
impairing substance becomes prevalent in society and a licensee or
other entity elects to add the substance to their panel of substances
to be tested prior to it being scheduled by the Drug Enforcement
Administration. Alternatively, if HHS proposes to remove a class of
drugs from the panel of drugs to be tested that is listed in Sec.
26.31(d)(1), then a licensee or other entity may not make a similar
change to its panel of drugs to be tested, because this change would be
a reduction in FFD program effectiveness even with a mitigative
strategy implemented.
Changes in the HHS panel of drugs and drug metabolites to be tested
could potentially shift from one metabolite to a different metabolite
for the same drug. Should HHS issue such a change to its panel, this
would not be expected to result in a reduction in FFD program
effectiveness because HHS would be targeting a more effective
metabolite for identifying an existing drug already being tested in its
panel. This situation could occur as HHS gathers more operating
experience from Federal government implementation of its HHS
Guidelines, or data generated by drug testing laboratories and
Federally mandated drug testing programs required by Federal agencies
such as the NRC and U.S. Department of Transportation.
Proposed Sec. 26.903(c)(4) would require that change control
records be maintained for a 5-year record retention period based on the
current NRC practice to conduct triennial inspections of licensees' and
other entities' FFD programs. This would afford the NRC an opportunity
to review the licensee's or other entity's determination that FFD
program changes have not reduced the effectiveness of their FFD
program. Licensees and other entities would also be required to
summarize each change made under proposed Sec. 26.903(c) in their
annual FFD performance reports required by proposed Sec. 26.917(b)(2)
or Sec. 26.717, ``Fitness-for-duty program performance data,'' as
applicable.
Proposed Sec. 26.904(a) would provide the timing for when a
licensee or other entity under proposed part 57 would be required to
have its subpart P FFD program in place and in effect. The timing of
proposed Sec. 26.904(a) would be equivalent to that for an LWR
licensee or other entity that is performing those same activities at a
facility licensed under part 50 or 52 and would help provide assurance
that those individuals who assemble, conduct non-operational testing,
or perform construction activities as defined in Sec. 26.5 or direct
these activities are fit for duty and trustworthy and reliable. This is
important because assembly and non-operational testing of a
manufactured reactor and the construction and testing of SSCs required
for facility operation require, in part, adherence to procedures,
possible implementation of unique and precise assembly techniques, and
QA and controls. Additionally, SSCs within a manufactured reactor may
not be accessible, testable, or available for quality assurance and
verification after the reactor is assembled. This requirement also
would address solo-assembly activities that may cause latent failures
and passive SSCs located internal to a reactor (for example, a fusible
link designed to melt at a particular temperature to trigger an
actuation mechanism) that would be relied upon for safe operation but
could not be inspected or tested for proper installation,
configuration, or operation after installation. A proposed subpart P
FFD program for these types of activities would be equivalent to the
FFD program applicable to the assembly of the reactor vessel internals
and testing of the SSCs internal to the reactor at an LWR licensed
under part 50 or 52.
The holder of the ML should establish in its procedures when
reactor assembly commences and what constitutes assembly. For example,
the FFD program would not need to be implemented for the receipt,
storage, inspection, and staging of components and systems used to
assemble (i.e., build or fabricate) the reactor because this is not a
current requirement for LWR facilities licensed under part 50 or 52.
Furthermore, the NRC currently does not require that an FFD program be
applied to the assembly or manufacturing of components (or basic
components as defined in Sec. 21.3), or systems that were fabricated
or assembled outside the footprint of a power reactor, and this
regulatory
[[Page 23668]]
position also would apply to a manufacturing facility.
Proposed Sec. 26.904(b) would set out the requirements that each
subpart P FFD program would be required to implement. These
requirements include FFD program elements similar to those in subpart B
of part 26, but the proposed new requirements would be less
prescriptive, enabling more flexibility in program implementation like
that offered in subpart K of part 26. For example, the requirements in
subpart B of part 26 are explicit requirements for, in part, the
collection and testing of urine specimens. Subpart B of part 26 does
not enable the use of oral fluid for drug testing, except under very
limited situations as described in subpart E of part 26, or the use of
hair specimens, unlike proposed subpart P. Proposed subpart P would
require drug and alcohol testing based on either the requirements in
part 26 or the HHS Guidelines. The principal benefits of the proposed
subpart P FFD program would be that it would provide a regulatory
framework that is consistent with the radiological consequences for
microreactors and other reactors with comparable risk profiles, and
would afford flexibilities in the conduct of drug and alcohol testing.
Proposed Sec. 26.906, ``Written policy and procedures,'' would
require licensees and other entities to implement and maintain an FFD
policy and procedures for their FFD programs. Proposed Sec.
26.906(a)(1) would require each licensee and other entity to provide a
written FFD policy statement to individuals subject to the FFD program
before the individuals are subjected to any FFD program drug and
alcohol test. This would be a protection measure afforded to
individuals subject to the FFD program to help ensure that they know
what is expected of them before being subject to the FFD program and
potential consequences should they violate the FFD policy or
procedures. This requirement would also contribute to safety and
security because understanding FFD program responsibilities may enhance
an individual's safety culture or the individual may self-select out of
the licensee's or other entity's hiring process.
Proposed Sec. 26.906(a)(2) would require that the FFD policy
statement describe the performance objectives in Sec. 26.23, which are
the same FFD program performance objectives required for facilities
licensed under part 50, 52, or 70. Having a standard performance
outcome based on a licensee or other entity satisfying the Sec. 26.23
performance objectives would enhance consistency in FFD program
implementation across all entities subject to part 26. It would also
generate confidence that individuals subject to part 26 will safely and
competently perform their duties and responsibilities and use NRC-
licensed materials in a manner that will protect the public health and
safety and common defense and security.
Proposed Sec. 26.906(a)(3) would require that the FFD policy
statement describe the licensee's or other entity's implementation of
the minimum days off requirements in Sec. 26.205(d)(3) or maximum
average work hours requirements in Sec. 26.205(d)(7).
Proposed Sec. 26.906(a)(4) would require the FFD policy statement
be written in sufficient detail to provide affected individuals with
information on what is expected of them and what consequences may
result from a lack of adherence to the policy, including those elements
described in proposed Sec. 26.903(b), part 26-required sanctions, and
required medical/clinical treatment and follow-up testing for FFD
policy violations. This requirement would be equivalent to Sec.
26.403(a) of subpart K but would include an additional description of
what the policy statement must include. For example, the policy would
describe the NRC-required sanctions to help deter substance abuse and
required medical/clinical treatment and follow-up testing for FFD
policy violations. This provision would provide a protection measure by
helping the individual get the assistance they need and help ensure
that the individual refrains from substance abuse.
Proposed Sec. 26.906(a)(5) would require that the FFD policy
statement describes the individual's responsibilities to report for
work in a physiological and psychological condition that enables the
safe and competent performance of assigned duties and responsibilities
and to inform a licensee- or other entity-designated representative
when the individual determines that this cannot be accomplished.
Proposed Sec. 26.906(a)(6) would require the FFD policy statement
to prohibit alcohol consumption within at least 5 hours prior to the
individual's arrival at the licensee's or other entity's facility.
Proposed Sec. 26.906(a)(7) would require the FFD policy statement
to convey that abstaining from alcohol for at least 5 hours before any
scheduled tour of duty is a minimum necessary measure, though it may
not be sufficient to ensure fitness for duty.
Proposed Sec. 26.906(b) would require licensees and other entities
implementing a proposed subpart P FFD program to establish, implement,
and maintain written procedures for their FFD programs. This
requirement would be equivalent to that in Sec. 26.403(b) of subpart
K.
Proposed Sec. 26.906(b)(1) would establish requirements for a
proposed subpart P FFD program to have written procedures for the drug
and alcohol testing program. This provision would be equivalent to the
requirements in current Sec. 26.403(b)(1) of subpart K, but proposed
Sec. 26.906(b)(1)(i) through (iv) proposes additional clarity and
specificity that licensees and other entities would be required to
detail in their procedures to address new testing methods in proposed
subpart P that are not permitted under the current part 26 framework.
Clarity and specificity in procedural instructions would support
consistent program implementation, which protects all individuals
subject to the program.
Proposed Sec. 26.906(b)(1)(iv) would require that if the licensee
or other entity elects to use the HHS Guidelines for the conduct of
drug testing, the FFD program procedures must include the name of the
specific HHS Guideline and revision being implemented by the licensee
or other entity and a description of the specific sections in the
guideline that are being implemented, including specimen collections,
drug testing, laboratory procedures, and evaluation of test results.
This requirement would help ensure the following: the validity and
accuracy of drug testing because the specimens would be subject to
laboratory testing that has been certified by the HHS; protection of
worker rights equivalent to the privacy, information, and due process
protections afforded to Federal workers under the HHS Guidelines
because the HHS Guidelines are used in the Federally mandated drug
testing programs; consistency in program implementation because all
individuals subject to the FFD program would be subject to the same
collection, testing, and evaluation processes; and FFD program
effectiveness because the effectiveness of the HHS Guidelines have been
verified by HHS's National Laboratory Certification Program (NLCP).
Detailed procedures would enhance MRO and FFD program personnel reviews
of individual test results because instructions would be provided for,
in part, the evaluation of specific test results (e.g., positive,
negative, biological markers), the conduct of additional testing for
invalid or dilute specimens, and the assessment of subversion attempts
(e.g., adulterated or substituted). This would benefit FFD program
effectiveness and help prevent
[[Page 23669]]
misunderstanding of program requirements and processes.
Proposed Sec. 26.906(b)(2) would require licensees and other
entities to include in their written procedures the immediate and
follow-up actions that would be taken, and the procedures that would be
used, in certain situations specified in proposed Sec. 26.906(b)(2)(i)
through (vi). Proposed Sec. 26.906(b)(2) would be equivalent to the
requirements in current Sec. 26.403(b)(2), which provides the same
requirement under an FFD program for construction for part 50 or 52
licensees and other entities. This would help ensure the effectiveness
of the FFD program and its consistent implementation, because part 57
licensees and other entities would be implementing procedures to
address the same requirements and with individuals who would understand
what is expected of them no matter what part 57 facility they were
assigned.
The situation specified in proposed Sec. 26.906(b)(2)(i) would
arise when individuals subject to the FFD program have been involved in
the use, sale, or possession of illegal substances, illegal drugs, or
illicit substances. This provision would be equivalent to current Sec.
26.403(b)(2)(i), except that the phrase ``illegal drugs'' would be
replaced with ``illegal substances, illegal drugs, or illicit
substances.'' Illegal substances would include legal substances used in
a manner inconsistent with Federal or State law.
The situation specified in proposed Sec. 26.906(b)(2)(ii) would
arise when individuals are impaired by any substance or the consumption
of alcohol as determined by behavioral observation or a test that
measures blood alcohol concentration, as defined in Sec. 26.5. Except
for a few differences, this provision would be equivalent to current
Sec. 26.403(b)(2)(ii) of subpart K. The NRC would not include the
phrases ``to excess'' and ``accurately'' in proposed Sec.
26.906(b)(2)(ii). Proposed subpart P of part 26 would be a performance-
based framework that focuses on impaired human performance, and for
alcohol, impairment is determined by blood alcohol concentrations
exceeding the limits in Sec. 26.103, ``Determining a confirmed
positive test result for alcohol,'' using an evidentiary breath testing
device (EBT) for alcohol (not whether an individual drank ``to
excess'').
The NRC would include the phrase ``illegal substances, illegal
drugs, and illicit substances'' in proposed Sec. 26.906(b)(2)(ii)
based on operating experience and the terminology in current Sec.
26.23(b). There are far more substances that may cause impairment than
those designated by U.S. Drug Enforcement Administration as controlled
substances (i.e., those that appear on Schedules I through V of section
202 of the Controlled Substances Act), and alcohol. The phrase ``before
or while constructing or directing construction of safety- or security-
related SSCs'' in current Sec. 26.403(b)(2)(ii) is not included in
proposed Sec. 26.906(b)(2)(ii) because proposed Sec. 26.906 would
apply during construction and operation. The NRC would include the term
``behavioral observation'' in proposed Sec. 26.906(b)(2)(ii) because
impairment can be visibly or audibly observed in an individual, and
individuals subject to proposed subpart P would be trained in
behavioral observation under proposed Sec. 26.908.
The situation specified in proposed Sec. 26.906(b)(2)(iii) would
arise when individuals attempt to subvert the testing process by
adulterating or diluting specimens (in vivo or in vitro), substituting
specimens, or by any other means and would be equivalent to current
Sec. 26.403(b)(2)(iii). The purpose underlying this proposed
requirement has increased in significance since the issuance of the
2008 part 26 final rule because subversion attempts have accounted for
about one-third of all drug testing violations of the FFD policy every
year since 2016.
The situation specified in proposed Sec. 26.906(b)(2)(iv) would
arise when individuals refuse to provide a specimen for analysis or
refuse to follow instructions provided by FFD program personnel. Except
for one difference, this provision would be equivalent to current Sec.
26.403(b)(2)(iv). The NRC would include the phrase ``or follow the
instructions provided by FFD program personnel'' based on an existing
requirement in Sec. 26.89(c) that the collector must inform the donor
that if the donor refuses to cooperate in the specimen collection
process, then such refusal will be considered a refusal to test and
sanctions for subverting the testing process will be imposed.
The situation specified in proposed Sec. 26.906(b)(2)(v) would
arise when individuals had legal action taken relating to drug or
alcohol use. This requirement would be equivalent to current Sec.
26.403(b)(2)(v).
The situation specified in proposed Sec. 26.906(b)(2)(vi) would be
when individuals subject to an FFD program demonstrated character or
actions indicating that the individual cannot be trusted or relied upon
to perform those duties and responsibilities or maintain access to NRC-
licensed facilities, SNM, or sensitive information. This includes
character traits beyond those attributed to drug or alcohol use. This
proposal would help ensure that the licensee or other entity will
implement an FFD program designed to demonstrate compliance with the
Sec. 26.23(c) performance objective that FFD programs must provide
``reasonable measures for the early detection of individuals who are
not fit to perform the duties that require them to be subject to the
FFD program.'' An individual who is not trustworthy and reliable is not
fit to perform or direct the performance of those duties and
responsibilities or be afforded those types of access that make the
individual subject to an FFD program.
The phrase ``character or actions'' would be used in proposed Sec.
26.906(b)(2)(vi) to focus on observed examples that indicate an
individual subject to proposed subpart P may not be fit for duty or
trustworthy and reliable. Character traits would include but not be
limited to personality, temperament, honesty, carelessness, apathy,
psychosis, and commitment to safety culture. Assessment of an
individual's character should consider the potential for changes in
these traits when compared to a previous baseline. Actions would
include a physical or verbal demonstration of a character trait that
could call into question an individual's fitness, trustworthiness, or
reliability. For example, the individual does something physically,
verbally, or in writing (e.g., falsifying records, driving while
impaired, or harming or threatening to harm oneself, others, or
property) that compels another individual to conclude that the observed
individual cannot be trusted or relied upon.
Unlike the background investigation and reviews of ``character and
reputation'' in Sec. 73.56(d)(6) and (k)(1)(v), which are principally
retrospective reviews of an individual and may be based on third-party
information (i.e., information from individuals not subject to NRC
requirements), the ``character or action'' focus of proposed Sec.
26.906(b)(2)(vi) would be a present observation of an individual
subject to the FFD program and performed by an individual who is also
subject to the FFD program. Whether the information would be received
from an individual subject to the FFD program or someone who is not
subject to the FFD program, the licensee or other entity would need to
review this information (i.e., determine if the information and its
source are credible) to determine whether the individual should
maintain authorization.
[[Page 23670]]
The situation specified in proposed Sec. 26.906(b)(2)(vii) would
be when an individual's condition or actions pose or have posed an
immediate hazard to themself or others, as notified by EAP personnel
under Sec. 26.35(c)(2).
Proposed Sec. 26.906(b)(3) would require licensees and other
entities to address in their procedures the process, including the
duties and responsibilities of FFD program personnel, to be followed if
an individual's behavior or condition raises an FFD concern. This
provision would also require a process to be conducted when credible
information is received by the licensee or other entity that the
individual is not fit for duty, trustworthy, and reliable.
With a few exceptions, proposed Sec. 26.906(b)(3) would be
equivalent to current Sec. 26.403(b)(3). Instead of the phrase ``while
constructing or directing the construction of safety- or security-
related SSCs'' in current Sec. 26.403(b)(3), the NRC would use ``on
the NRC-licensed facility'' in proposed Sec. 26.906(b)(3) because this
provision would apply during nuclear plant construction and operation
in addition to holders of an ML as described in proposed Sec. 26.3(f).
The requirement that the roles and responsibilities of FFD program
personnel be described was developed from current Sec. Sec. 26.4(g)
and 26.31(b) and operating experience, which has demonstrated that
clear job descriptions help ensure that individuals know who is
designated by the licensee or other entity to make decisions regarding
FFD program implementation and who can be approached when physiological
or psychological help is needed. This is principally a protection
consideration afforded to individuals subject to the FFD program.
Proposed Sec. 26.906(b)(3) would also include two conditions not
found in current Sec. 26.403(b) that would clarify the initiation of
the fitness determination process should an individual's behavior or
condition raise an FFD concern. The phrase, ``impairment from any cause
that in any way could adversely affect the individual's ability to
safely and competently perform the individual's duties,'' would reflect
the Sec. 26.23(b) performance objective. The condition, ``the receipt
of credible information indicating that the individual cannot be
trusted or relied on to perform those duties and responsibilities
making the individual subject to this part,'' would reflect the Sec.
26.23(a) performance objective. In either case, as required by Sec.
26.23(c), the FFD program would have to provide reasonable measures for
the early detection of individuals who are not fit to perform the
duties that require them to be subject to the FFD program.
Proposed Sec. 26.906(b)(4) would require licensees and other
entities to have written procedures that address the operation and
oversight of onsite and offsite collection facilities. This requirement
would be equivalent to current Sec. Sec. 26.403(b) and 26.405(e) and
is developed from Sec. 26.41(b), which states that each licensee and
other entity who is subject to subpart B of part 26, shall ensure that
the entire FFD program is audited, which is part of a licensee's or
other entity's oversight of the facility, and Sec. 26.87(a), which
states that each FFD program must have one or more designated
collection sites that have all necessary personnel, materials,
equipment, facilities, and supervision to collect specimens for drug
testing and to perform alcohol testing. Having procedures for the
operation and oversight of onsite and offsite collection facilities
would enhance consistency in program implementation, protect
individuals subject to testing, and account for the flexibilities
afforded in the types of biological specimens than may be collected
under an FFD program subject to proposed subpart P of part 26. Proposed
Sec. 26.906(b)(4), when used with the audit requirement in proposed
Sec. 26.915, ``Audits,'' would help maintain FFD program effectiveness
and prevent subversion attempts at facilities that may not be under the
direct day-to-day oversight of FFD program personnel.
Proposed Sec. 26.906(b)(5) would require licensees and other
entities to have written procedures that address the fatigue management
requirements in proposed Sec. 26.202(b), ``Procedures,'' and either
Sec. 26.205(d)(3) or (d)(7).
Proposed Sec. 26.906(b)(6) would require licensees and other
entities to have written procedures that provide measures to prevent
subversion of drug and alcohol tests conducted onsite and offsite. This
proposal was developed from Sec. 26.27(c)(1).
Proposed Sec. 26.907, ``Drug and alcohol testing,'' would
establish drug and alcohol testing requirements for licensees and other
entities. Except for a few differences, proposed Sec. 26.907 would be
equivalent to current Sec. 26.405, ``Drug and alcohol testing,'' which
requires licensees and other entities implementing an FFD program under
subpart K of part 26 to have a drug and alcohol testing program that
demonstrates compliance with the requirements in Sec. 26.405(b)
through (g). The differences are commensurate with the risk
consequences presented by a part 57-licensed facility as compared to a
part 50 or 52 nuclear power plant. These proposed requirements would
improve flexibility in the conduct of drug and alcohol testing while
maintaining protections afforded to individuals subject to the FFD
program.
Proposed Sec. 26.907(a), ``Split specimens,'' would require
licensees and other entities to obtain a split specimen for all drug
tests using oral fluid or urine for all test conditions in proposed
Sec. 26.907(b), ``Test conditions,'' and (j), ``Blood testing.''
Neither current subpart K nor current subparts B or E of part 26
require a split specimen. However, many of the LWR fleet uses split
specimens for drug testing, and commercially available drug screening
products use a split specimen technique. Since publication of the 2008
part 26 final rule, the HHS has issued guidelines for urine and oral
fluid specimen testing that require split specimen collections. The
U.S. Department of Transportation regulations under 49 CFR part 40 also
require split specimen collections for urine and oral fluid. The
proposed HHS Guidelines for hair testing also require split specimen
collections.
The required use of a split specimen process would protect the
individual because, upon a donor-alleged discrepant or questionable
test result, the donor may provide permission to test the split
specimen (specimen B) in an effort to refute the laboratory test
results for specimen A. The requirement also would enable the MRO to
direct laboratory testing of specimen B if specimen A were invalid;
though the NRC expects specimens becoming invalid at the laboratory to
be a rare occurrence as testing would be conducted by HHS-certified
laboratories. If a specimen is determined to be invalid, then the
occurrence would likely warrant further investigation by the MRO and
laboratory to identify the cause. This protocol would be equivalent to
the special analysis testing in current Sec. 26.163(a)(2) for dilute
specimens and specimens collected under most directly observed
collection conditions in that additional laboratory analysis is
performed because of a questionable test result.
If a split specimen is tested by an HHS-certified laboratory, then
the test result from specimen B must be used as part of the
determination for an FFD policy violation as required by Sec.
26.185(n), ``Evaluating results from a second laboratory.'' However,
this is not to say that the test results from specimen A should be
discarded. Since the HHS-certified laboratory should report all test
results from all specimens tested to the MRO, like the information
described in Sec. 26.169, ``Reporting
[[Page 23671]]
results,'' test result differences between specimens A and B can be
used to inform the MRO as to what should be reported to the licensee or
other entity to either facilitate medical or clinical assistance for
the individual, inform an FFD policy violation determination, or both.
Proposed Sec. 26.907(a) would state that split specimen
collections of oral fluid or urine must be used for the test conditions
described in proposed Sec. 26.907(b). In addition, testing of the
split specimen (specimen B) would require the donor's permission unless
ordered by the MRO to resolve an invalid test result obtained for
specimen A.
Proposed Sec. 26.907(b) would require the licensee or other entity
to subject individuals identified in Sec. 26.4 to drug and alcohol
testing under the five conditions listed in proposed Sec. 26.907(b)(1)
through (5). Proposed Sec. 26.907(b) would be equivalent to current
Sec. 26.405(c).
Proposed Sec. 26.907(b)(1), ``Pre-access,'' would require pre-
access testing similar to current Sec. 26.405(c)(1), which requires
testing before assignment to construct or direct the construction of
safety- or security-related SSCs. Unlike current Sec. 26.405(c)(1),
the proposed requirement would not include the phrase, ``construct or
direct the construction of safety- or security-related SSCs,'' because,
for licensees or other entities under proposed part 57, the pre-access
test condition would apply to construction and operation to help inform
a licensee's or other entity's authorization determination. The
proposed requirement also would use ``pre-access'' instead of ``pre-
assignment,'' which is used in current Sec. 26.405(c)(1).
A pre-access test would require the collection of an oral fluid or
a urine specimen no more than 14 days before the individual is granted
unescorted access. Although this change has roots in the 2008 part 26
final rule, which reduced the period within which pre-access testing
must be performed from 60 days to 30 days or less, the 14-day proposal
is based on two lessons learned from operating experience.
First, the 14-day period would be a large enough window of time to
collect the specimen and evaluate test results because licensees or
other entities typically receive laboratory test results within 5
business days of laboratory receipt of the biological specimen. At the
same time, the 14-day period would be small enough to help ensure that
the test results are representative of the individual's recent drug use
before being granted authorization.
Second, the NRC does not expect licensees and other entities
licensed under proposed part 57 to have the large and periodic influxes
of individuals (either licensee employees or C/Vs) that large LWRs have
to support facility operation, maintenance, engineering design changes,
or nuclear refueling. Therefore, these licensees or other entities
would not be periodically challenged to in-take a large workforce
within the proposed 14-day pre-access testing window.
Proposed Sec. 26.907(b)(2), ``Random,'' would require the licensee
or other entity to conduct random drug and alcohol testing of all
individuals subject to the FFD program. With some exceptions, this
proposed requirement would be equivalent to current Sec. 26.405(b).
Section 26.405(b) gives licensees and other entities that implement an
FFD program subject to subpart K of part 26 the option to impose random
drug and alcohol testing. Proposed Sec. 26.907(b)(2) would not offer
that option because proposed subpart P of part 26, unlike subpart K,
would not allow a licensee or other entity to implement a fitness
monitoring program under current Sec. 26.406, ``Fitness monitoring,''
instead of a random testing program. The principal reasons for not
allowing this flexibility would be that no licensee or other entity has
ever implemented a fitness monitoring program (i.e., there is no
operating or regulatory experience on which to judge the effectiveness
of a fitness monitoring program), and the proposed subpart P framework
already uses behavioral observation to help ensure FFD program
effectiveness. Supplementing the proposed Sec. 26.909, ``Behavioral
observation,'' behavioral observation program (BOP) with an additional
observation technique (i.e., the fitness monitoring program) would not
result in a level of deterrence or detection equivalent to that which
would be obtained through behavioral observation and random drug and
alcohol testing.
Proposed Sec. 26.907(b)(2)(i) through (v) would provide specific
requirements for the conduct of a random testing program. These
paragraphs would be equivalent to Sec. 26.405(b)(1) through (4),
although with a few differences. The similar provisions would be
proposed Sec. 26.907(b)(2)(i), (b)(2)(iii), and (b)(2)(iv).
The differing provisions would include proposed Sec.
26.907(b)(2)(ii), which would refer to an ``FFD program procedure''
instead of the reference to an ``FFD program policy'' in Sec.
26.405(b)(2) because procedures contain the instructions that implement
FFD program requirements, but the FFD policy need not contain specific
instructions. Proposed Sec. 26.907(b)(2)(ii) also would require
individuals who are selected for random testing to report to the onsite
collection site, as opposed to the collection site in Sec.
26.405(b)(2), because alcohol metabolism necessitates a timely alcohol
test. This change is also proposed because the NRC expects that part 57
licensees and other entities may use a combination of onsite (for
random, for-cause, and post-event testing) and offsite (for pre-access,
post-event, and follow-up testing) collection facilities for drug and
alcohol testing and may have to afford reasonable accommodation to
certain individuals, which would add complexity in the licensee's or
other entity's procedurally determined time period in which an
individual must report to the collection facility.
Another difference from Sec. 26.405(b) is proposed Sec.
26.907(b)(2)(v), which would establish the random testing rate for the
population of individuals subject to testing. Subpart K of part 26 does
not establish a random testing rate. The proposed requirement would be
equivalent to current Sec. 26.31(d)(2)(vii), which requires that the
sampling process used to select individuals for random testing provides
that the number of random tests performed annually is equal to at least
50 percent of the population that is subject to the FFD program at the
NRC-licensed site.
Proposed Sec. 26.907(b)(3), ``For cause,'' would require for-cause
testing equivalent to that used in current FFD programs implementing
Sec. 26.405(c)(2). The NRC is proposing for-cause testing, like random
testing, to be conducted onsite to ensure that the test is conducted as
soon as reasonably practicable. This is an important consideration when
for-cause testing for alcohol or using oral fluid for drug screening or
testing because human metabolism continually lowers the concentrations
of the drugs, drug metabolites, and alcohol perhaps to concentrations
lower than the initial or confirmatory testing cutoffs. Additionally,
for facilities that are sited in geographically remote locations, an
offsite collection facility might be too far away or not readily
accessible.
Proposed Sec. 26.907(b)(4), ``Post-event,'' would require post-
event testing in a manner equivalent to current Sec. 26.405(c)(3),
with a few adjustments. For proposed part 57 licensees or other
entities, the NRC is proposing post-event testing under two conditions:
events involving human errors that may have caused or contributed to
the events (proposed Sec. 26.907(b)(4)(i)), and events
[[Page 23672]]
not involving human error that result in adverse health consequences or
damage to any safety- or security-related SSC (proposed Sec.
26.907(b)(4)(ii)). The word ``significant'' would not be used in
proposed Sec. 26.907(b)(4)(ii)(A) to describe the ``illness or
personal injury'' as used in Sec. 26.405(c)(3)(i) because proposed
Sec. 26.907(b)(4)(ii)(A) would describe which illnesses or injuries
are covered. Proposed Sec. 26.907(b)(4)(ii)(B), unlike Sec.
26.405(c)(3)(ii), would not use the word ``significant'' to describe
the damage to safety- or security-related SSCs because any damage to
safety- or security-related SSCs would require testing within four
hours of the event unless immediate medical intervention precludes the
conduct of the test on the individual(s) who caused or contributed to
the event. Proposed Sec. 26.907(b)(4)(ii)(B) would also not use the
word ``construction'' as in Sec. 26.405(c)(3)(ii) because proposed
Sec. 26.907(b)(4) would apply to construction and operation.
Proposed Sec. 26.907(b)(4)(i) would require the licensee or other
entity to define in its procedures the term ``human error.'' This term
may take on various meanings and it is not defined in the current or
proposed rule, so the licensee or other entity would be required to
describe or define this term to help ensure consistent implementation
of proposed subpart P and that the post-event test condition would be
consistently applied to all individuals subject to the FFD program. The
Sec. 26.405(c)(3)(i) requirement that ``the event is recordable under
the Department of Labor standards contained in 29 CFR 1904.7, and
subsequent amendments thereto,'' would not be carried over to proposed
Sec. 26.907(b)(4). Instead, the NRC proposes to prescribe the post-
event test conditions in proposed Sec. 26.907(b)(4), in part so they
would not change unless the NRC amends the requirement.
Proposed Sec. 26.907(b)(5), ``Follow-up,'' would require follow-up
testing. This requirement would be equivalent to current Sec.
26.405(c)(4), although proposed Sec. 26.907(b)(5) would further
describe follow-up testing. The NRC proposes to describe follow-up
testing as part of a series of tests for drugs, alcohol, or both, which
are performed after an individual subject to part 26 has violated the
FFD policy on substance use or abuse, or the sale, use, or possession
of illegal drugs. Follow-up testing would be used to verify an
individual's continued abstinence from substance abuse. The NRC would
not include a reference to a follow-up plan as in Sec. 26.405(c)(4)
because the intent of a follow-up plan is to conduct a series of drug
tests, alcohol tests, or both, to verify continuing abstinence from
substance abuse. Nevertheless, individuals who violate an FFD policy on
substance use or abuse, or the sale, use, or possession of illegal
drugs, should have a follow-up plan that includes a definition of
``abstinence'' from the medical professional prescribing the plan.
Proposed Sec. 26.907(c), ``Urine and oral fluid specimens,'' would
provide additional testing requirements. The proposed requirement would
be equivalent to Sec. 26.405(d) and would require implementation of
select requirements from current subpart E of part 26. The proposed
requirements would govern directly observed collections, shy bladder
situations, special analysis testing, and alcohol testing. These
requirements would be necessary to maintain FFD program effectiveness
equivalent to that currently implemented by the LWR fleet.
Proposed Sec. 26.907(c)(1) would establish the minimum panel of
drugs and drug metabolites to be tested. This panel would be the same
as those in Sec. Sec. 26.31(d)(1) and 26.405(d) because, based on
operating experience from LWR FFD program implementation, this panel
has been determined to contribute to a licensee or other entity
satisfying the FFD performance objectives in Sec. 26.23(a) through
(d).
Section 26.405(d) requires that urine specimens collected for drug
testing be subject to validity testing. Like Sec. 26.405(d), proposed
Sec. 26.907(c)(1) would require testing of urine specimens for
validity. Oral fluid specimens could also be subject to validity
testing, including a biological marker, as specified in either part 26
or the HHS Guidelines.
Proposed Sec. 26.907(c)(2) would include requirements that already
exist in the part 26 framework that provide protections for individuals
subject to the FFD program and contribute to testing effectiveness when
collecting and assessing a urine specimen. Specifically, current Sec.
26.115, ``Collecting a urine specimen under direct observation,''
describes the exclusive grounds for performing a directly observed
collection and the process to be followed to protect the privacy of the
individual. Section 26.119, ``Determining `shy' bladder,'' establishes
the process to be followed when a donor is not able to produce a
sufficient amount of urine for testing, and Sec. 26.163(a)(2) requires
special analysis testing when a specimen is dilute to help prevent a
subversion attempt.
Proposed Sec. 26.907(c)(3) would require implementation of all the
current alcohol testing requirements in Sec. 26.91, ``Acceptable
devices for conducting initial and confirmatory tests for alcohol and
methods of use,'' through Sec. 26.103. Using the same alcohol testing
framework for parts 50, 52, 57, and 70 licensees and other entities
would provide for regulatory consistency, protections for individuals
subject to the FFD program (e.g., the quality controls and verification
applied to the EBT), and FFD program effectiveness (e.g., accuracy of
test results). For alcohol testing, unlike drug testing, there is a
preponderance of evidence that correlates blood alcohol concentrations
to impairment and intoxication. Furthermore, FFD performance data has
demonstrated that the time-dependent alcohol cutoffs in Sec. 26.103
have increased the detection of individuals who are under the influence
of alcohol. For these reasons, the current alcohol requirements in part
26 would be required for FFD programs under proposed subpart P.
Proposed Sec. 26.907(c)(4) would establish additional testing
requirements. This proposal would be equivalent to current Sec.
26.405(f) for facilities licensed under proposed part 57 for the
conduct of drug testing. Unlike Sec. 26.405(f), proposed Sec.
26.907(c)(4) would not reference validity screening and initial drug
and validity tests at licensee testing facilities. Another minor
difference between Sec. 26.405(f) and proposed Sec. 26.907(c)(4)
would reflect the requirement in proposed subpart P to use an HHS-
certified laboratory for all biological specimens collected and not
just for urine specimens.
Consistent with Sec. 26.405(f), proposed Sec. 26.907(c)(4) would
require the use of an HHS-certified laboratory for all test conditions
listed in proposed Sec. 26.907(b), MRO-directed tests, and the testing
of a split specimen. Further, HHS-certified laboratory test results
using urine or oral fluid would be required for the issuance of an FFD
policy violation and part 26-required sanction.
All drug testing would need to be performed at an HHS-certified
laboratory to help ensure FFD program effectiveness and to protect the
donor from a false positive test result and an unwarranted FFD policy
violation. The donor would be protected because laboratory procedures
for specimen accessioning, testing, custody and control, and evaluation
of test results and the training and qualification of laboratory
personnel are evaluated by HHS as part of the NLCP. This would
[[Page 23673]]
provide assurance that the drug testing results are accurate and
attributed to the donor. Hair specimens could also be pre-access tested
for drugs as described in proposed Sec. 26.907(h), ``Hair testing,''
and positive test results could only be used as potentially
disqualifying information for a licensee's or other entity's
authorization determination (i.e., used to assess the fitness,
trustworthiness, and reliability of the individual). A positive hair
test result could not be used for the administration of an FFD policy
violation and sanction, except as provided for in proposed Sec. Sec.
26.907(h)(3) and 26.910(b)(4) for attempts to subvert the testing
process, as defined in Sec. 26.5.
There are three phrases or requirements in Sec. 26.405(f) that the
NRC does not propose to use in proposed Sec. 26.907(c)(4). The first
is the phrase, ``consistent with its standards and procedures for
certification,'' regarding the operation of an HHS-certified
laboratory, because the laboratory would not be HHS-certified if it
were not following ``its standards and procedures for certification.''
The second is the requirement that urine specimens that yield positive,
adulterated, substituted, or invalid initial validity or drug test
results must be subject to confirmatory testing by the HHS-certified
laboratory, except for invalid specimens that cannot be tested. This
requirement would not be used because, under proposed subpart P of part
26, licensees or other entities would not be required to use an HHS-
certified laboratory. For a laboratory to be HHS-certified, it must
follow the HHS Guidelines and include procedures that describe when a
specimen cannot be tested. Lastly, the Sec. 26.405(f) requirement that
other specimens that yield positive initial drug test results must be
subject to confirmatory testing by a laboratory that demonstrates
compliance with stringent quality control requirements that are
comparable to those required for certification by the HHS, would not be
used because proposed subpart P of part 26 would require the use of an
HHS-certified laboratory.
Proposed Sec. 26.907(c)(5) would require the licensee or other
entity to contract with an HHS-certified laboratory and would specify
the same requirements that current Sec. 26.153(f) requires for
contracts between licensees or other entities who are subject to part
26 and HHS-certified laboratories. Proposed Sec. 26.907(c)(5)(ii)
would state that records and documents must be provided and/or able to
be photocopied and removed from the premises to support the inspection
or audit. This requirement would be equivalent to current Sec.
26.41(d), except that laboratories would not be able to limit the use
and dissemination of documents copied or taken from the laboratory by a
licensee or other entity. This would be necessary to ensure the
continuing effectiveness of FFD programs, because NLCP findings and
audit results could adversely impact FFD program effectiveness.
Pertinent information includes and should not be limited to NLCP-
identified weaknesses (e.g., custody and control, accessioning,
instrumentation, procedures, training, supervision, review of test
results, and resolution of previously identified corrective actions)
that may impact the effectiveness of FFD programs.
Proposed Sec. 26.907(d), ``Privacy and integrity,'' would help
protect the donor from mistakes made during the drug and alcohol
testing processes and help ensure FFD program effectiveness. The NRC
would require the licensee or other entity to protect the individual's
privacy and the integrity of the specimen and to implement quality
controls to ensure that test results are valid and attributable to the
correct individual. This proposed requirement would be equivalent to
the first sentence of current Sec. 26.405(e), except that the word
``stringent'' would be removed from the phrase ``stringent quality
controls,'' because the word ``stringent'' is not defined.
Proposed Sec. 26.907(e), ``Offsite collection facilities,'' would
describe the requirements for licensees and other entities that use
offsite collection facilities. Consistent with current Sec. 26.405(e),
a licensee or other entity would be able to conduct specimen
collections and alcohol testing at a local hospital or other facility,
except for those specimens that must be collected onsite under proposed
Sec. 26.907(b)(3) and (4). Unlike Sec. 26.405(e), proposed Sec.
26.907(e) would not restrict licensees and other entities to use
hospitals and other facilities that meet the U.S. Department of
Transportation requirements in 49 CFR part 40 because proposed subpart
P of part 26 is intended to provide flexibilities beyond those in the
current part 26 framework. Licensees and other entities may use these
Department of Transportation requirements to inform their procedures
under proposed Sec. 26.906(b)(1) as long as the procedures do not
conflict with the requirements in part 26 or the HHS Guidelines.
Proposed Sec. 26.907(e) would also require licensees and other
entities to audit offsite collection facilities before their use and
biennially to confirm that the facility procedures are comparable to
those described in subpart E of part 26 or the HHS Guidelines for urine
and oral fluid. This prosed requirement is based on current Sec.
26.41(a) and (b). The proposed Sec. 26.907(e) audit requirement is a
program effectiveness consideration because offsite collection
facilities may not require vigilance of their collectors (e.g.,
identification of subversion attempts), diligence in the protection of
worker rights (e.g., privacy and specimen custody and control), or
procedural compliance.
The offsite facility used by a licensee or other entity under
proposed Sec. 26.907(e) would have to be licensed to conduct specimen
collections and perform alcohol testing, and be audited, by the State
or a State-designated entity. This requirement would help provide
assurance of adequate collection facility performance and may help
reduce the burden on the licensee or other entity and the collection
facility. Crediting a State audit (or State licensure, oversight, or
regulation) is established in Sec. Sec. 26.4(i)(4) and (j),
26.91(e)(5), 26.153(f)(1), and 26.183(a).
Proposed Sec. 26.907(f), ``Initial testing,'' would provide the
requirements for initial drug testing. This provision would be
equivalent to Sec. 26.405(f) except to account for the testing of
urine and oral fluid specimens under proposed subpart P of part 26. The
initial test would have to use an immunoassay or an alternative
technology, as specified in the HHS Guidelines for the specific
biological specimen that is to be tested. Examples of alternative
technologies include liquid or gas chromatography and mass
spectrometry. Another difference from Sec. 26.405(f) would be changing
the word ``urine'' in Sec. 26.405(f) to ``biological specimens'' in
proposed Sec. 26.907(f). Lastly, proposed Sec. 26.907(f) would
include the phrase ``discrepant biological marker'' as a drug screening
result that would have to be analyzed by an HHS-certified laboratory
and evaluated by the MRO to help inform the MRO's determination of a
subversion attempt.
Proposed Sec. 26.907(g), ``Oral fluid testing,'' would enable a
part 57 licensee to use oral fluid as a biological specimen for
testing. This requirement would be equivalent to Sec. 26.31(d)(5),
which enables the MRO to conduct drug and alcohol testing using
alternative methods, and Sec. 26.405, which does not preclude the use
of oral fluid specimens for FFD programs that implement subpart K of
part 26 requirements. In order to provide assurance that drug testing
is effective and protects the worker, proposed Sec. 26.907(g) would
require that the licensee's or other
[[Page 23674]]
entity's procedures incorporate the HHS Guidelines or the requirements
in part 26 for the conduct of urine or oral fluid testing.
Proposed Sec. 26.907(g) would require that the oral fluid device
must not expire before the date of the collection of the specimen.
Also, the drugs, drug metabolites, initial and confirmatory testing
cutoffs, and biological markers, if applicable, would need to be those
established by the HHS Guidelines for oral fluid drug testing and the
alcohol cutoffs in part 26. If they were not established by the HHS
Guidelines or part 26 for the paneled drugs and drug metabolites, then
they would be determined and documented by a forensic toxicologist
review under Sec. 26.31(d)(1)(i)(D).
Proposed Sec. 26.907(g)(2) would permit the virtual collection of
oral fluid specimens for drug and alcohol testing but only at
facilities that must use a C/TPA to implement random testing under
proposed Sec. 26.907(b)(2)(vi). A virtual collection monitor would be
permitted in the location where the specimen collection is to be
performed to assist the virtual collector, such as by completing
Federal CCF paperwork; observing activities outside the viewable area
of the video teleconference equipment to ensure that the donor does not
attempt to subvert the testing process; providing information to the
virtual collector if/when requested; and ensuring that the oral fluid
specimen(s) once packaged for shipping are secured until picked up for
transportation to the HHS-certified laboratory.
Proposed Sec. 26.907(h) would enable the collection of hair
specimens for drug testing to supplement pre-access testing of urine or
oral fluid specimens. Hair testing would be a new feature in the part
26 framework. The NRC proposes to permit the use of hair testing for
only Schedule I or II drugs or their metabolites to inform a licensee's
or other entity's determination whether the individual is trustworthy
and reliable. For example, if an individual stated no prior use of
illegal drugs, a pre-access hair test could be performed to ascertain
the validity of the individual's statement. However, if the HHS-
certified laboratory were to report a positive test result, an FFD
policy violation could not be administered. This laboratory information
would need to be treated as potentially disqualifying FFD information,
unless the individual were determined to have attempted to subvert the
testing process, in which case a permanent denial of authorization
would be required under proposed Sec. 26.910(b)(4). To provide
assurance of testing effectiveness and protections afforded to
individuals subject to the FFD program, proposed Sec. 26.907(h) would
require that an HHS-certified laboratory must be used to test the hair
specimen. The forensic toxicologist review would be necessary if the
panel of drug or drug metabolites to be tested and their cutoffs were
not established by HHS or part 26 for hair.
Proposed Sec. 26.907(i), ``Portal area screening,'' would enable
the use of portal area screening instruments to test for drugs,
alcohol, or both, should these types of screening tests become
available for use. This technology could substantially contribute to a
licensee or other entity satisfying the Sec. 26.23 performance
objectives by helping ensure that all individuals who arrive at the
NRC-licensed facility to perform or direct those duties and
responsibilities or maintain those types of access making them subject
to the FFD program are fit for duty and deterred from arriving onsite
in a physiological condition that may be adverse to safety and
security. Additionally, screening could be conducted when individuals
exit the NRC-licensed facility to provide assurance that substance
abuse had not occurred onsite (see Sec. 26.23(d)). The screening
instrument could be electronically linked to temporarily prevent
ingress or egress and could automatically inform licensee- or other
entity-designated officials of the portal area alarm. The use of portal
area screening technologies could also represent cost savings because,
for NRC-licensed facilities that have small staff sizes or are
geographically remote, passive drug and alcohol screening technologies
could be an innovative alternative to a random testing program,
although the license or other entity would need to request and receive
an exemption.
Proposed Sec. 26.907(i) would also provide that if the portal area
screening instrument detects a substance that exceeds the instrument's
established setpoint, the individual then would need to be for-cause
tested under proposed Sec. 26.907(b)(3) for drugs, alcohol, or both,
depending on the screening test result received. A portal area
screening test result is to be considered credible use information,
which would strengthen the effectiveness of a licensee's or other
entity's BOP. The requirements would not allow an individual to be
rescreened by the portal area screening instrument following an initial
screening detection that exceeded an established setpoint in order to
prevent a subversion attempt. To ensure the accuracy of any portal area
screening testing performed by a licensee or other entity, a
performance-based approach would need to be used to verify the
continuing accuracy of the testing for each substance tested by the
instrument. A portal area screening test could be used so long as the
accuracy of the test result for a specific substance were confirmed by
the resultant for-cause testing performed on an oral fluid or urine
specimen for drugs, oral fluid or breath specimen for alcohol, or both.
If a portal area screening result for a specific drug or drug
metabolite were confirmed by drug testing performed at an HHS-certified
laboratory, or oral fluid or breath alcohol testing for at least 85
percent of the specimens testing positive on portal area screening in
the past 12-month data reporting period for a specific substance, the
portal area screening test for that substance could continue to be
used. This performance-based measure would balance the use of the
technology with the protection afforded to individuals from unnecessary
testing. If these instruments and alcohol screening devices have the
capability, they could also be used to determine the true identity of
individuals to facilitate the implementation of the FFD BOP, which
could be very practicable at facilities that operate with small staff
sizes.
Proposed Sec. 26.907(j) would enable the use of a blood specimen
for drug, alcohol, or other testing for certain medical conditions as
determined by the licensee- or other entity-designated MRO. This
requirement would be equivalent to current Sec. 26.31(d)(5). The use
of a licensee- or other entity-designated MRO and not one designated by
a third party, such as an MRO employed by an offsite specimen
collection facility, would be important because the MRO must be
familiar with the proposed subpart P requirements. To help ensure
testing effectiveness and protect the worker, the blood test would need
to be conducted by a laboratory that demonstrates compliance with
quality control requirements that are comparable to those required for
certification by the HHS, such as a hospital or clinic certified by the
State, Commonwealth, or territory.
Proposed Sec. 26.907(k), ``Federal custody and control form,''
would require licensee and other entities to use a Federal custody and
control form (Federal CCF) as defined in Sec. 26.5 for the collection
and packaging of hair, oral fluid, and urine specimens for drug
testing. This proposed requirement is based on the Federal CCF
documentation requirements in current subpart E of part 26 because
subpart K of part 26 does not require the use of a Federal CCF under
Sec. 26.117(e).
[[Page 23675]]
Proposed Sec. 26.907(l), ``Medical Review Officer,'' would
establish requirements for the licensee- or other entity-designated
MRO. Proposed Sec. 26.907(l)(1) would be equivalent to Sec.
26.405(g), however, the word ``designated'' would be added to the first
sentence to clarify that the MRO would be designated by the licensee or
other entity, and not by a third party. As stated with regard to
proposed Sec. 26.907(j), this change would clarify that it is the
licensee's or other entity's responsibility, through their designated
MRO, to determine whether an individual is fit for duty and trustworthy
and reliable. This would be consistent with the description of FFD
program personnel in current Sec. 26.31(b) and help provide FFD
program effectiveness and protections to individuals subject to the FFD
program. The paragraph was also modified from Sec. 26.405(g) to
address the determinations of FFD policy violations and fitness
required by subpart H of part 26.
Proposed Sec. 26.907(l)(2) would help ensure that MRO reviews are
consistent with those MRO reviews conducted at other NRC-licensed
facilities subject to part 26 and that the MRO maintains knowledge of
drug collection, testing processes and procedures, and evaluation of
testing results.
The NRC also proposes that if an MRO performed the duties and
responsibilities in Sec. Sec. 26.185, ``Determining a fitness-for-duty
policy violation,'' and 26.187, ``Substance abuse expert,'' for at
least three continuous years in the last 10 years prior to being hired
or contracted by the licensee or other entity, then the MRO would not
need to repeat the initial training and examination requirements. The
basis for 3 years is that the MRO would have experienced three annual
cycles of evaluating drug and alcohol test results, contributed to the
annual FFD program performance data reported to the NRC, experienced a
refueling or maintenance outage, understood the duties and
responsibilities of individuals subject to the FFD program to make
informed determinations of fitness, demonstrated a safety culture that
helps ensure FFD program effectiveness, and been subject to NRC
inspection. The basis for 10 years is the relatively long periods
between significant changes to part 26 and the HHS Guidelines.
Proposed Sec. 26.907(l)(3) would require that the MRO attend a
medical- or clinical-based training session every 5 years. This
proposal was developed, in part, from section 13.1 of the HHS
Guidelines for the testing of urine and oral fluid specimens and 49 CFR
40.121 of the U.S. Department of Transportation's requirements. The NRC
would not include an examination requirement as part of this refresher
training requirement because it could limit the types of trainings that
MROs may attend. The proposed requirements are justified to maintain
currency on changes in societal drug use, forensic toxicology,
determinations of fitness, and other part 26 technical areas necessary
to perform required responsibilities as an MRO performing services
under proposed subpart P.
Proposed Sec. 26.907(l)(4) would require the MRO to evaluate drug
testing results by implementing the requirements in Sec. 26.185 or the
HHS Guidelines through the licensee's or other entity's procedures.
This requirement would help ensure FFD program effectiveness and
enhance consistency across the commercial nuclear industry for the
evaluation of drug testing results. This also would help protect
individuals because they would be subject to the same evaluation
criteria. If Sec. 26.185 provides insufficient information for an MRO
to make a determination on a drug testing result (including adulterant
and discrepant biological markers), the guidance issued by a State
agency in the state in which the NRC-licensed facility is located,
Federal agency, or nationally recognized MRO training and certification
organization may be used to inform an MRO determination. This provision
would ensure that the MRO has the flexibility to inform their
evaluation of the drug testing results and fitness determination, if
necessary, considering the drug- and alcohol-related flexibilities
afforded in subpart P of part 26.
The proposed requirement would also state that an MRO need not
review alcohol test results, including positive confirmatory alcohol
test results determined by an EBT under proposed Sec. 26.907(c)(3)(vi)
and (vii), which are the current requirements in Sec. Sec. 26.101,
``Conducting a confirmatory test for alcohol,'' and 26.103,
respectively. Proposed Sec. 26.907(c)(3)(i) would require the use of
an EBT under Sec. 26.91, which would ensure that confirmatory alcohol
test results are precise and accurate to issue FFD policy violations.
Proposed Sec. 26.907(l)(5) would require the licensee- or other
entity-designated MRO to determine and approve the use of oral fluid or
urine as an alternative biological specimen when the donor cannot
provide a requested specimen for testing. This proposed requirement
would be equivalent to Sec. 26.31(d)(5), which enables the use of an
alternative specimen collection if a medical condition makes the
collection of the biological specimen difficult. This determination and
the retest must be completed as soon as reasonably practicable and
documented to support recordkeeping, auditing, and NRC inspection.
Proposed Sec. 26.907(l)(6) would require that the MRO review all
specimen test results associated with a drug-related FFD policy
violation. This would include split specimens and all specimens taken
to resolve a discrepant condition, such as a possible subversion
attempt, impairment without a known cause, or a donor-requested or MRO-
directed retest. To resolve a discrepant condition, the MRO would be
authorized to test a specimen for a biological marker, adulterants, or
additional drugs. The broad scope of this MRO evaluation would be
necessary because of the variety of different screening and testing
methods that may have been associated with the FFD policy violation.
All information learned from the conduct of part 26 drug and alcohol
screening and testing should be used in the evaluation of an
individual's trustworthiness and reliability, issuance of a sanction,
and development of a follow-up treatment and testing plan, if
administered.
Proposed Sec. 26.907(m), ``Limitations of screening and testing,''
would be equivalent to current Sec. 26.31(d)(6) and would establish
limits on the screening and testing of biological specimens. This would
be a protection consideration afforded to individuals subject to the
FFD program and was not provided in subpart K of part 26. This proposed
requirement would state that specimens collected under NRC regulations
may only be designated or approved for screening and testing as
described in part 26 and may not be used to conduct any other analysis
or test without the written permission of the donor. Analyses and tests
that would not be permissible would include, but would not be limited
to, deoxyribonucleic acid (i.e., DNA) testing, serological typing, or
any other medical or genetic test used for diagnostic or specimen
identification purposes.
The NRC proposes to require that no biological specimens may be
passively sampled and analyzed in a manner different than described in
proposed subpart P of part 26 to ensure workers are protected from non-
consensual passive screening. The proposed subpart P framework would
enable passive detection of drugs and alcohol, whereas passive
detection is not afforded in subparts A through I, N, and O of part 26.
[[Page 23676]]
Proposed Sec. 26.907(n), ``Specimen collectors,'' would be
equivalent to current Sec. Sec. 26.31(b)(1)(iii)(A) and 26.89 and
would require that all specimen collections be conducted by a licensee-
or other entity-designated and -trained individual. For proposed
subpart P of part 26, this would include onsite specimen collections,
except a collection by a portal area screening instrument in proposed
Sec. 26.907(i).
Proposed Sec. 26.908 would require licensees and other entities to
provide FFD program training to individuals subject to the FFD program.
The performance-based proposed Sec. 26.908 requirement was developed
from the prescriptive training requirements in current Sec. 26.29 and
modeled on current Sec. 50.120 because there is no training
requirement in subpart K of part 26.
Proposed Sec. 26.908(a)(1) would require an FFD training program
that includes the licensee's or other entity's FFD policies and
procedures, including fatigue management, and the individuals' FFD
program responsibilities. Individuals who collect specimens for testing
would also need to be trained in specimen collector duties and
responsibilities, including, at a minimum, specimen collection, custody
and control, identification and response to subversion attempts, and
privacy. For individuals specified in Sec. 26.4, a licensee or other
entity of a nuclear plant would be required to use a systems approach
to training as defined in proposed in Sec. 57.390. These requirements
are based on requirements in Sec. 26.29(a)(2), (3), (9), and (10).
Proposed Sec. 26.908(a)(2) would require training on the BOP. This
requirement would be based on Sec. Sec. 26.29(a)(8), (9), and (10) and
26.33, ``Behavioral observation.'' The proposal would require
individuals to be trained in the detection of behaviors or conditions
that may indicate the use of illegal drugs, as in the current Sec.
26.33 BOP requirements, and the use of illicit drugs and substance
abuse onsite and offsite. Also, in reference to impairment from fatigue
or any cause if left unattended, the phrase in Sec. 26.33, ``may
constitute a risk to public health and safety or the common defense and
security,'' would be replaced in proposed Sec. 26.908(a)(2)(iii) with
``could result in inattentiveness or human errors,'' because proposed
subpart P of part 26 would be focused, in part, on ensuring individuals
are fit for duty to perform or direct the performance of assigned
duties and responsibilities safely and competently.
Proposed Sec. 26.908(a)(2)(iv) would focus on training to inform
individuals that they are responsible for their own conduct, as well as
observing others. Specifically, individuals would be trained to
recognize when they feel unable to safely and competently perform
assigned duties and responsibilities, as well as to recognize when
others appear unable to safety and competently perform assigned duties
and responsibilities or act in an untrustworthy and unreliable manner.
The training requirement and the self-reporting requirement in proposed
Sec. 26.906(a)(5) would be in the interest of safety and security
because the individual is proactively announcing that assistance may be
necessary. This would be consistent with the performance objectives in
Sec. 26.23(b) and (c), where certain behavior or stress conditions may
be indicative of an individual not being fit for duty, trustworthy, and
reliable.
Proposed Sec. 26.908(a)(3) would help ensure that individuals
subject to the FFD program understand that FFD policy violations would
result in an FFD program sanction and that program information learned
or generated by FFD program implementation would be used to aid
licensee or other entity authorization determinations and be shared, as
requested, with other licensees or other entities subject to parts 26
and 73. This proposed requirement would be equivalent to Sec.
26.29(a)(1). Proposed Sec. 26.908(a)(3) would be a protection measure
afforded to individuals subject to the FFD program because they would
understand that licensees and other entities subject to parts 26 and 73
would be informed of, in part, an individual's character, reputation,
and ability to follow policies, procedures, and instructions to safely
and competently perform assigned duties and responsibilities in a
trustworthy and reliable manner. Fitness for duty-related information
would include drug and alcohol testing results (not quantitative
testing values), issuance of any sanctions, FFD-determinations
regarding trustworthiness and reliability, testing programs, treatment,
and other remedial or corrective action.
Proposed Sec. 26.908(b), ``Training and assessments,'' would
require individuals to be trained on the FFD program and to receive a
trainee assessment before pre-access testing. Proposed Sec. 26.908(b)
also would require that FFD program refresher training and trainee
assessments be conducted on a nominal 24-month frequency or more
frequently if the need is indicated. These requirements would be
equivalent to Sec. 26.29(c)(1). However, proposed Sec. 26.908(b) was
developed from the systems approach to training-based training
requirements in Sec. 50.120 and training elements from the annual FFD
program refresher training requirements in Sec. 26.29(c)(2). A trainee
assessment would be the same as in currently required systems approach
to training-based training programs.
Proposed Sec. 26.908(c), ``Training program review,'' would
require licensees and other entities to periodically evaluate their FFD
training programs and revise them as appropriate. This training focus
is not required by subpart K of part 26 or Sec. 26.29 but is proposed
to address the flexibilities afforded in proposed subpart P of part 26.
This section would be equivalent to Sec. 50.120(b)(3).
Proposed Sec. 26.909 would require the implementation of a BOP.
The requirement would be equivalent to that in Sec. Sec. 26.33 and
26.407, ``Behavioral observation,'' and would apply during construction
and operation. Under the FFD program, the purpose of the BOP would be
to help ensure that individuals subject to the FFD program are fit for
duty and trustworthy and reliable to perform or direct those duties and
responsibilities and maintain those types of access that make the
individual subject to the FFD program. This assurance would be
accomplished by requiring each individual subject to proposed subpart P
to be subject to behavioral observation, and by requiring all
individuals to perform behavioral observation of others and report FFD
concerns to the licensee- or other entity-designated representative(s).
The intent of the BOP requirement would not be to require that all
individuals be observed at all times by others; NRC-licensed operators,
maintenance professionals, security officers, and others routinely
perform solo operations periodically throughout the day. However,
individuals would need to be subject to observation while they are
performing or directing the performance of duties and responsibilities
or maintaining the types of access making them subject to the FFD
program. Observing behavior only at the beginning of a work shift would
not be sufficient to ascertain whether an individual is fit for duty,
trustworthy, and reliable. Impairing substances may have a delayed
effect between use (e.g., ingestion of a controlled substance) and the
onset of physiological or psychological effects, and fatigue
accumulates with time. Behavior must be continually observed throughout
the work shift to detect any changes from baseline human performance
characteristics, including mental or physical health and mannerisms, or
any activities that may
[[Page 23677]]
indicate that the individual is not trustworthy and reliable.
Proposed Sec. 26.909(a) would differ from Sec. Sec. 26.33 and
26.407 in that it would place the responsibility for performing
behavioral observation on ``all individuals subject to this subpart,''
rather than only those ``individuals specified in Sec. 26.4(f) [who]
are constructing or directing the construction of safety- or security-
related SSCs'' in Sec. 26.407 or ``individuals who are trained under
Sec. 26.29 to detect behaviors'' in Sec. 26.33 to improve clarity.
Proposed Sec. 26.909(b) would require all individuals subject to
the FFD program to report to the licensee- or other entity-designated
representative any onsite or offsite behaviors or activities by
individuals subject to part 26 that could constitute an unreasonable
risk to the safety or security of the NRC-licensed facility or SNM or
may cause harm to others. The NRC would require this description of
reportable conduct because an individual's activities (e.g., use of
illegal substances) and communications (e.g., hate speech or threats of
violence) offsite are a direct indication of the individual's fitness,
trustworthiness, and reliability and must be evaluated as to whether
authorization should be granted or maintained. Proposed Sec. 26.909(b)
would include a description of this conduct instead of the Sec. 26.33
undefined phrase, ``FFD concerns,'' to enhance the clarity of the
requirement. This BOP reporting requirement would include any
information relating to character or reputation of the individual
indicating that the individual cannot be trusted or relied upon to
perform those duties and responsibilities or maintain access to NRC-
licensed facilities, SNM, or sensitive information. Proposed Sec.
26.909(a) and (b) were written broadly to include offsite conduct that
the reporting individual considers serious enough to call into question
the character or reputation of the subject individual.
Proposed Sec. 26.909(c) would require that licensees and other
entities perform behavioral observation visually, in-person, and, when
necessary, remotely by live video and audible streaming and capture.
This requirement was developed from the security observation
requirements in Sec. 73.55(e)(7)(i)(B) and (C), (h)(2)(v), and (i)(2)
and (i)(5)(ii). Conducting an in-person observation of another
individual would be the preferred method to ascertain whether the
observed individual can safely and competently perform assigned duties
and responsibilities. When in-person observations would not be feasible
(e.g., during solo operations), the proposed requirement would enable
the use of video monitoring. This is addressed, for example, in
proposed Sec. 26.909(d) regarding NRC-licensed operator manipulation
of reactor controls. Additionally, certain duties (such as maintenance
activities performed by a single worker outside of a control room) may
not present an opportunity for video monitoring; in these situations,
behavioral observation should be conducted on a sampling basis (i.e., a
planned observation of the work activity) as outlined in a licensee's
or other entity's FFD program.
In situations involving small staff sizes, facilities sited in
geographically remote locations, or both, additional observers would
enhance the effectiveness of a BOP. Technological developments in
automated safety and security systems may enable licensees or other
entities to reduce staff sizes to 10 to 40 percent of the staff size of
an LWR facility licensed under part 50 or 52. Smaller staff sizes may
translate into more solo operations, less teamwork, fewer peer checks,
or infrequent management oversight of field activities, leading to
fewer behavioral observations. Therefore, a licensee or other entity
may have fewer opportunities to observe whether individuals are fit for
duty.
Proposed Sec. 26.909(d) would require that licensees or other
entities perform behavioral observation of NRC-licensed operators who
manipulate the controls of any nuclear plant licensed under proposed
part 57, remotely by live video and audible streaming capture for those
part 57 facilities where individual task loading does not allow for the
effective conduct of behavior observation in addition to assigned
operational tasks. The purpose of this paragraph would be similar to
that of proposed Sec. 26.909(c), where the possibility of in-person
observation is significantly diminished because of solo operations or
because the facility may only require a minimum staff size onsite.
Proposed Sec. 26.910(a) would be similar to Sec. 26.409,
``Sanctions,'' and would require the licensee or other entity to
establish sanctions for FFD policy violations that, at a minimum, would
prohibit the individuals specified in Sec. 26.4 from being assigned to
perform or direct those duties and responsibilities or maintaining
authorization making them subject to proposed subpart P of part 26. To
be consistent with Sec. 26.75, ``Sanctions,'' the severity of the
sanction as described in proposed Sec. 26.910(b) would escalate with
the number of occurrences and severity of the FFD policy violation. The
sanction would be long enough to help deter future FFD policy
violations and facilitate counseling and treatment before the licensee
reinstates the individual's access to the facility.
Proposed Sec. 26.910(b)(1) would require a minimum 14-day denial
of access for an individual's first violation of the FFD policy
involving a confirmed positive drug or alcohol test result. Proposed
Sec. 26.910(b)(2) would require a minimum 3-year denial of access for
an individual's second violation of the FFD policy involving a
confirmed positive drug or alcohol test result.
Equivalent to Sec. 26.75(c), proposed Sec. 26.910(b)(3) would
require a minimum 5-year denial of access for who is determined to have
been involved in the sale, use, or possession of illegal drugs or the
consumption of alcohol within a protected area of any facility licensed
under proposed part 57 or within a transporter's facility or vehicle
used in the conveyance of formula quantities of strategic SNM.
Equivalent to Sec. 26.75(b), proposed Sec. 26.910(b)(4) would require
a permanent denial of authorization be issued for a third violation of
the FFD policy involving a confirmed positive drug or alcohol test
result or a subversion attempt of any drug or alcohol test or screening
process.
Proposed Sec. 26.911, ``Protection of information,'' would protect
information collected from FFD program implementation and would be
equivalent to current Sec. 26.411, ``Protection of information.'' The
protected information would include, but not be limited to, privacy and
medical information. Proposed Sec. 26.911 would not include the Sec.
26.411 requirement that FFD programs must maintain and use the personal
information with the highest regard for individual privacy because such
a requirement would be unnecessary considering the proposed Sec.
26.911(a) requirement that licensees and other entities would have to
establish and maintain a system of files and procedures to prevent
unauthorized disclosure.
Proposed Sec. 26.911(b), although equivalent to Sec. 26.411(b),
would require licensees and other entities to have all individuals sign
a consent to be subject to the FFD program before subjecting the
individual to the FFD program (e.g., before being subject to a pre-
access test in proposed Sec. 26.907(b)(1), unlike Sec. 26.411(b)).
The purpose of this proposal would be to enhance protections afforded
to individuals subject to the FFD program and their knowledge of, in
part, why they are subject to drug and alcohol testing, behavioral
observation, information
[[Page 23678]]
collection, MRO reviews, and other FFD program elements. Like the
consent required by Sec. 26.411(b), the consent would authorize
disclosure of the collected information. Consent would not be needed
for disclosures to the individuals and entities specified in Sec.
26.37(b)(1) through (b)(6), (b)(8), and persons deciding matters under
review in proposed Sec. 26.913, ``Appeals process.''
Proposed Sec. 26.913 would be equivalent to Sec. 26.413, ``Review
process.'' The proposed title would be changed to an appeal process to
clarify that proposed Sec. 26.913 would be the process implemented
when an individual elects to appeal a licensee or other entity
determination that the individual had violated the FFD policy. The
proposal would also require that the process include a schedule for the
completion of the review of the determination that the individual had
violated the FFD policy. The NRC proposes this requirement because
operating experience demonstrates that workers may not be protected
from a continuous review process that does not result in an outcome.
Proposed Sec. 26.915, ``Audits,'' would require licensees and
other entities to perform audits of the FFD program. The proposed
section would be similar to Sec. 26.415, ``Audits.'' Under proposed
Sec. 26.915(a), audits would be performed at a frequency that ensures
the FFD program's continuing effectiveness. Corrective actions would be
taken as soon as reasonably practicable to resolve any problems
identified and preclude recurrence. Proposed Sec. 26.915(b) would
require the subject matter, scope, and frequency of audits to be
revised as necessary to improve or maintain FFD program performance
based on annual FFD program performance data reviews performed under
proposed Sec. 26.917(d) and unsatisfactory performance or programmatic
weaknesses identified under proposed Sec. 26.917(b)(3) and (e).
Proposed Sec. 26.915(c) would be equivalent to Sec. 26.415(b) and
would enable licensees and other entities to conduct joint audits or
accept audits of C/Vs so long as the audit addresses the relevant
services of the C/Vs.
Proposed Sec. 26.915(d) would be equivalent to Sec. 26.415(c) by
establishing requirements for the auditing of HHS-certified
laboratories. Unlike Sec. 26.415(c), the proposal would not contain a
reference to the U.S. Department of Transportation drug and alcohol
testing requirements. This would broaden the regulatory flexibility
afforded to a licensee or other entity in that they may use an offsite
collection or testing facility that does not meet the Department of
Transportation requirements.
Proposed Sec. 26.915(d) would state that licensees and other
entities need not audit an HHS-certified laboratory if the licensee's
or other entity's panel of drugs and drug metabolites to be tested is
equivalent to the panel by which the laboratory is certified by HHS or
is subject to the standards and procedures for drug testing and
evaluation used by the laboratory under the HHS Guidelines. The NRC
would afford this flexibility because the NRC is aware that HHS desires
to streamline changes in its guidelines to its panel of drugs and drug
metabolites to be tested. Therefore, if a licensee or other entity
elects to implement the HHS Guidelines in its procedures and maintains
the minimum panel of drugs and drug metabolites to be tested as
required by proposed subpart P, a licensee or other entity may still
use (and not audit) the HHS-certified laboratory because the proposed
Sec. 26.903(e) change control process would maintain FFD program
effectiveness.
To help ensure FFD program effectiveness, Sec. 26.915(d) would
also require that collection facility procedures are comparable to
those required in subpart E of part 26, including a proposed
requirement that the offsite facility's specimen collection and testing
procedures are audited on a biennial basis, which is also a protection
consideration afforded to individuals subject to the FFD program.
Conducting this audit on a biennial basis would be equivalent to that
required in Sec. 26.41(b) and would help ensure that the specimen
collection process at the facility remains effective.
Proposed Sec. 26.917, ``Recordkeeping, reporting and FFD program
performance,'' would establish recordkeeping, reporting, and FFD
program performance requirements similar to those in current Sec.
26.417, ``Recordkeeping and reporting.'' However, proposed Sec. 26.917
would require retention of records pertaining to administration of the
FFD program and FFD performance data required by Sec. 26.717 until
license termination, which is based on current Sec. 26.711(a) because
Sec. 26.417 does not provide for a retention period.
Proposed Sec. 26.917(b)(1) would be identical to the reporting
requirements in Sec. 26.417(b)(1) regarding the licensee's or other
entity's FFD program.
Proposed Sec. 26.917(b)(2) would require the reporting of annual
(i.e., January through December) FFD program performance data for each
FFD program subject to proposed subpart P. Licensees and other entities
would be required to submit the program performance data to the NRC
before March 1 of the following year. This reporting would be
equivalent to the annual program performance requirement in Sec.
26.417(b)(1), and the March 1 due date is based on the reporting
deadline in Sec. 26.717(e). Licensees and other entities would be
required to report FFD performance information using NRC-provided forms
(e.g., new NRC Forms 893, ``Single Positive Test Form, 10 CFR part 26,
subpart P FFD Program,'' and 894, ``Annual Reporting Form, 10 CFR part
26, subpart P FFD Program.''
Proposed Sec. 26.917(b)(3) would require the reporting of drug and
alcohol testing errors to the NRC within 30 days of completing an
investigation of any testing errors or unsatisfactory performance,
discovered at an HHS-certified laboratory or through the processing of
appeals under proposed Sec. 26.913, or matters that could adversely
reflect on the integrity of the random selection or random testing
process. Licensees and other entities would be required to describe in
the reports the incident and any corrective actions taken or planned.
Proposed Sec. 26.917(c) would require that FFD-related information
be shared within the nuclear industry when requested to support
authorization determinations. This requirement would help individuals
seeking employment by another NRC-licensed facility subject to subpart
C of part 26, complete their NRC-required sanctions and licensee-
administered or -directed drug and/or alcohol abuse treatment plans
before the restoration of authorization by a licensee or other entity.
Information sharing may also enhance FFD program effectiveness because
FFD-related lessons learned from, for example, substance testing,
subversion attempts, and laboratory and MRO performance would have to
be shared when requested.
Proposed Sec. 26.917(d) would require that licensees and other
entities must analyze FFD program performance data at least annually
and take appropriate actions to correct any identified program
weakness.
Proposed Sec. 26.917(e) would require that licensees and other
entities must document, trend, and correct non-reportable indicators of
FFD programmatic weaknesses under the licensee's or other entity's
corrective action program. However, to protect individual privacy, drug
and alcohol test results could not be tracked in a manner that would
permit the identification of any individuals.
Proposed Sec. 26.919, ``Suitability and fitness determinations,''
would require
[[Page 23679]]
licensees or other entities to establish a process to evaluate
individuals when their fitness or trustworthiness and reliability are
in question. Section 26.919 would be equivalent to Sec. 26.419,
``Suitability and fitness determinations,'' but, unlike Sec. 26.419,
would apply during the construction and operation phases. Also,
proposed Sec. 26.919 would require that a suitability or fitness
determination conducted for cause be conducted face-to-face. This
proposed requirement is based on current Sec. 26.189(c); however,
unlike Sec. 26.189(c), proposed Sec. 26.919 would not prohibit
augmenting determinations via electronic means of communication (i.e.,
provide sufficient visual and aural clarity to complete the process).
Instead, proposed Sec. 26.919 would explicitly permit determinations
to be performed via electronic means and would explain when a trained
individual must be present in-person with the individual being assessed
(i.e., only to assist in completing for-cause drug and alcohol testing
determinations and fatigue assessments).
In considering the current restriction on the use of electronic
means of communication for determinations of fitness conducted for
cause, the NRC finds that since publication of the 2008 part 26 final
rule, there have been developments in using electronic means of
communication (i.e., videoconferencing) as an alternative to conducting
face-to-face interactions. To address these considerations, the NRC
contracted the Pacific Northwest National Laboratory to study whether a
medical and mental health assessment via electronic communication could
be an acceptable alternative to an in-person, face-to-face assessment.
Based on this study, if electronic means were to be used to conduct a
face-to-face assessment, an in-person element would still be integral
to the assessment process. However, under certain circumstances, face-
to-face determinations and assessments conducted as part of an FFD
program for an entity licensed under proposed part 57 (i.e., those
determinations and assessments performed in accordance with proposed
Sec. 26.919, Sec. 26.207, or Sec. 26.211) may be augmented via
electronic communications. Such remotely conducted determinations and
assessments would be required to be conducted with someone who is
present in-person with the individual being assessed and who is trained
in accordance with the requirements of either Sec. 26.29 and Sec.
26.203(c) or proposed Sec. 26.908 and Sec. 26.202(c). Permitting the
use of electronic communications would help ensure FFD program
effectiveness, especially in instances where the part 57 nuclear plant
is sited in a geographically remote location, when the facility has a
small staff size, and when an urgent determination is required.
C. 10 CFR Part 73
The NRC proposes several conforming changes to its regulations in
10 CFR part 73. Changes to Sec. Sec. 73.1, 73.2, ``Definitions,''
73.8, ``Information collection requirements: OMB approval,'' 73.50,
``Requirements for physical protection of licensed activities,'' 73.56,
``Personnel access authorization requirements for nuclear power
plants,'' 73.57, ``Requirements for criminal history records checks of
individuals granted unescorted access to a nuclear power facility, a
non-power reactor, or access to Safeguards Information,'' and 73.58,
``Safety/security interface requirements for nuclear power reactors,''
would be needed to incorporate proposed part 57 into these
requirements. Changes to Sec. 73.54, ``Protection of digital computer
and communication systems and networks,'' would require a licensee that
elects to implement the requirements of Sec. 73.54 to establish and
implement cybersecurity reviews to assess the effectiveness of the
implementation of the cybersecurity program. Changes to Sec. 73.77,
``Cyber security event notifications,'' would incorporate proposed
Sec. 73.110 into the cyberattack notification requirement and simplify
the regulation by eliminating specific event notifications and
redirecting licensees to existing notification processes.
Proposed Sec. 73.110 would establish requirements for the
development and maintenance of a cybersecurity program for nuclear
plants licensed under proposed part 57. This section would implement a
graded approach to determine the level of cybersecurity protection
required for digital computers, communication systems, and networks.
The proposed new section is informed by: (1) the operating experience
from power reactors and insights from cyber-related assessments of fuel
cycle facilities; and (2) the existing Sec. 73.54 framework, which
addresses some of the basic issues for cybersecurity regardless of the
type of reactor. Differences between the Sec. 73.54 requirements and
those proposed in Sec. 73.110 are primarily based on the
implementation of a consequence-based approach to cybersecurity that
provides flexibility to accommodate the wide range of reactor
technologies the NRC expects to assess under proposed part 57. A graded
approach based on consequences would account for the differing risk
levels among reactor technologies. Specifically, the proposed new
section would require licensees to demonstrate protection against
cyberattacks in a manner that is commensurate with the potential
consequences from those attacks.
D. 10 CFR Part 140
In this proposed rule, the NRC proposes several conforming changes
to its regulations in part 140 of this chapter. These conforming
changes would be needed to include licenses issued under the proposed
part 57 into the NRC's financial protection requirements and in
accordance with the requirements set forth in the Price-Anderson Act
(42 U.S.C. 2210). During the development of this proposed rule, the NRC
also considered a reduction in the amount of financial protection
required for facilities licensed under proposed part 57. Facilities
that would be licensed under proposed part 57 could pose reduced risks
in comparison to existing facilities, for which the current financial
protection requirements were established, thereby warranting a reduced
amount of required financial protection. Upon receipt of a joint
application under proposed part 57, the NRC would perform the necessary
review(s) in which to make a technical finding of this presumption. If
a lesser amount of financial protection were determined to be
commensurate with the reduced risk profile of the reactor, the NRC
would exercise its regulatory discretion to establish a reduced amount
of financial protection for facilities licensed under part 57, based on
factors such as those specified in the Price-Anderson Act: (A) the cost
and terms of private insurance; (B) the type, size, and location of the
licensed activity and other factors pertaining to the hazard; and (C)
the nature and purpose of the licensed activity.
Similarly, the NRC could also consider reducing indemnification
fees for certain licensees. The Price-Anderson Act establishes
indemnification fees but gives discretion to the NRC to establish lower
indemnification fees for some licensees. During its review of part 57
joint applications, the NRC could consider establishing reduced
indemnification fees for those applicants based on factors such as
those specified in the Price-Anderson Act: (1) the type, size, and
location of facility involved, and other factors pertaining to the
hazard, and (2) the nature and purpose of the facility.
[[Page 23680]]
VII. Specific Requests for Comments
The NRC is seeking advice and recommendations from the public on
this proposed rule. We are particularly interested in comments and
supporting rationale from the public on the following:
1. Entry Criteria. The NRC is proposing both a dose limit and a
material limit in proposed Sec. 57.25 as entry criteria for using
proposed part 57. The technical basis for these criteria are described
in section V.C of this document. During its public meetings on this
proposed rule in July 2025, the NRC received feedback from several
stakeholders requesting that this criterion be removed and the NRC
instead rely on a single entry criterion of a 1 rem (10 mSv) site
boundary dose threshold.
Q1-1: In lieu of applying a deterministic material limit
on the quantity of SNM to ensure safety, should the Commission consider
an alternative performance-based entry criterion? Please explain the
basis for your recommendation.
2. General License for Construction. During the development of this
proposed rule, the NRC considered whether it could use a general
license for rapid deployment of the types of reactors described herein.
The general license topic is discussed in section IV.C of this document
and concludes that the NRC cannot license entire utilization facilities
with a general license because of the limits in the NRC's authority
under the AEA. However, the NRC did determine that the issuance of a
general license for some construction activities for ``nth-of-a-kind''
reactors would be permissible.
Q2-1: Besides the general license approach for certain
construction activities in the proposed rule, are there other general
licensing approaches for important components parts of utilization
facilities that would benefit high-volume licensing or other regulatory
processes for microreactors and other reactors with comparable risk
profiles? Please explain the basis for your recommendation.
Q2-2: Given that the NRC anticipates that a review
timeline for the required part 70 license will align with the timeline
to complete a safety and security review of reactors via proposed part
57, would there be any benefits provided by a general license for a
reactor in addition to the general license for construction activities
proposed in part 57? Please provide your explanation.
3. Improvements to Proposed Part 57 Requirements. The NRC developed
this proposed rule with the intent to establish a risk-informed and
performance-based regulatory framework for high-volume licensing of
microreactors and other reactors with comparable risk profiles. The
proposed rule would provide licensing pathways and streamlined
requirements with increased flexibility, as compared to that of 10 CFR
parts 50 and 52, in meeting certain technical requirements. Examples of
this increased flexibility would include applicants being able to
specify industry-approved standards such as for QA programs and
technical codes and standards.
Q3-1: Should any requirements in proposed part 57 be
eliminated or made less burdensome or more flexible? If so, which ones?
For existing requirements in 10 CFR chapter I that are referenced by
proposed part 57, should any of them be similarly revised to the extent
that they are relied upon by a proposed part 57 requirement? If so,
which ones? Please explain the basis for your recommendation.
Q3-2: Recognizing that part 57 shares similar features
with part 53, are there any provisions in part 57 that should be
adapted for part 53 to enhance their complementary nature? For example,
should the NRC include provisions in part 53 that would provide a
general license for partial reactor construction or allow applicants to
reference a general area for siting? If so, what, if any, modifications
to the language in part 57 would be needed for it to be appropriate in
part 53?
Q3-3: Because the proposed part 57 directs licensees to
use 10 CFR 50.59, which uses the term ``important to safety,'' and that
term is not used in part 57, should the NRC explain in a guidance
document how a part 57 licensee should use 10 CFR 50.59 or should the
final part 57 include its own specific 10 CFR 50.59-like process?
Q3-4: Is a single notice in the Federal Register for each
joint application for a construction permit and associated operating
license(s) sufficient and appropriate for notice for large geographic
areas? Or should additional measures be employed to put the public on
notice of a hearing opportunity for a large geographic area, and if so,
what measures?
Q3-5: Should the NRC look holistically at the duration of
renewals for manufacturing licenses, design certifications, and
standard design approvals across all parts?
Q3-6: Should the NRC consider periodicities other than the
proposed 5-year interval for FSAR updates?
4. Early Site Permit Considerations for Proposed Part 57. Under the
current regulatory framework, applicants pursuing licenses under 10 CFR
part 50 must address site suitability, environmental, and emergency
preparedness issues as part of their CP and OL applications. By
contrast, 10 CFR part 52 provides an early site permit (ESP) process
that allows applicants to resolve site-related issues in advance of
design certification or combined license applications. As interest
grows in deploying a wider range of advanced reactor technologies,
including microreactors and other reactors with comparable risk
profiles, stakeholders have suggested that a similar ESP process for
applicants for licenses for microreactors and other reactors with
comparable risk profiles could increase licensing efficiency. Such a
process could enable early resolution of site issues, reduce
duplicative reviews, and provide greater certainty to project
developers while maintaining the NRC's high standards for safety and
environmental protection.
Q4-1: Should a proposed part 57-compatible early site
permit process be developed? Describe the potential value of creating a
proposed part 57-compatible ESP process, including the benefits and
drawbacks of such an approach for applicants and stakeholders, and
whether this process could facilitate more timely and predictable
licensing outcomes.
Q4-2: What types of site issues (e.g., seismic, emergency
planning, tribal consultations) would benefit most from early
resolution under such a process?
Q4-3: Would a part 52-type ESP process reduce licensing
uncertainty and costs for developers, and if so, how?
5. Decommissioning Considerations for Proposed Part 57. Some
stakeholders shared with the NRC at the July 2025 public meetings that
they envision that microreactors could be transported to a facility at
a different location than the operating site to be decommissioned or
refurbished and refueled. If refurbished and refueled, the reactor
would be redeployed for another operating cycle but eventually it would
permanently cease operation and decommissioning would be necessary.
Q5-1: Besides the volume of waste, would there be
differences in the process for refurbishment versus decommissioning of
the reactor, if both occurred at the same facility, that would be
important to consider with regard to enabling more efficient and safe
streamlining of the decommissioning licensing and the license
termination processes? Please provide a rationale supporting your
comment.
Q5-2: The NRC's current regulations generally restrict the
use of
[[Page 23681]]
decommissioning trust funds to activities conducted after permanent
cessation of operations, unless an exemption is granted. The NRC has
received stakeholder interest in accessing decommissioning funds during
reactor operation for the removal or replacement of major components
when those activities would ultimately be necessary for
decommissioning. The NRC is seeking stakeholder input on whether, and
under what conditions, limited access to decommissioning trust funds
for such activities during reactor operation should be considered. For
example, is there an anticipated need to access radiological
decommissioning funds during operations to facilitate the removal of a
reactor for refurbishment or other major radioactive component
disposal? Please provide a rationale supporting your comment.
6. Release of Part of a Nuclear Plant or Site for Unrestricted Use.
Under this proposed rule, a licensee would be able to release portions
of its nuclear plant or site for unrestricted use before license
termination by license amendment, or by including plans to release
parts of the site in the decommissioning plan. However, the proposed
rule does not include a specific provision for release of a part of a
site for unrestricted use before license termination as licensees can
request under Sec. 50.83 and Sec. 53.1080. Under those provisions,
licensees may request a partial site release by providing specific
information to the NRC, with the extent of the necessary information
depending on whether the area to be released has been designated as
``nonimpacted'' or ``impacted.'' The NRC is considering whether
specific provisions for partial site release, similar to those in parts
50 and 53, should be included in proposed part 57. In addition, because
proposed part 57 would include provisions for the NRC to approve
decommissioning plans well before decommissioning activities would
commence, the NRC is asking whether there should be differences between
a provision for releasing part of a site in proposed part 57 and
similar provisions in parts 50 and 53.
Q6-1: Should the NRC include a specific provision for
releasing a part of a nuclear plant or site for unrestricted use before
license termination in proposed part 57? If so, how should the NRC
consider adapting the approach in Sec. 50.83 and Sec. 53.1080 to make
the provision applicable to licensees under proposed part 57?
7. Transportation Dose Rates for Proposed Part 57. For the
certification of a transportation package, specific dose rate
requirements must be met during normal operations, normal conditions of
transport, and hypothetical accident conditions. For example, under
Sec. 71.47(a), during normal conditions incident to transport, the
maximum dose rate cannot exceed 2 millisieverts/hour (2 mSv/h) (or 200
millirem/hour) (200 mrem/h) at any point on the external surface of the
package, unless prepared for transport as an exclusive use package
pursuant to Sec. 71.47(b). Section 71.47(b) has additional operational
requirements and specified dose rates that include 10 mSv/h (1000 mrem/
h) at any point on the external surface of the package, 2 mSv/h (200
mrem/h) at any point on the outer surface of the vehicle, 0.1 mSv/h (10
mrem/h) at any point 2 meters (80 inches) from the outer lateral
surfaces of the vehicle, and 0.02 mSv/h (2 mrem/h) in any normally
occupied space, except that this provision does not apply to private
carriers, if exposed personnel under their control wear radiation
dosimetry devices in conformance with Sec. 20.1502, ``Conditions
requiring individual monitoring of external and internal occupational
dose.'' The additional requirements for Type B packages during accident
conditions is that no external radiation dose rate may exceed 10 mSv/h
(1 rem/h) at 1 meter (40 inches) from the external surface of the
package. These dose rates were developed in coordination with both the
Department of Transportation and the International Atomic Energy
Agency. The NRC is considering whether existing dose rate limits for
the transportation of radioactive material under 10 CFR part 71 remain
appropriate in light of the anticipated deployment of advanced
reactors, including microreactors. Microreactors may present unique
transportation considerations, such as the movement of fueled or
partially fueled reactors, higher-temperature or higher-burnup fuels,
increased shipment frequency to support rapid deployment, and near-site
transport for demonstration projects.
Q7-1: Provide feedback on the need for alternate dose
rates for transportable microreactors, the technical basis for those
alternate dose rates, and the safety implications for those alternative
dose rates.
Q7-2: Are there cost-benefit considerations beyond the
costs and benefits associated with rulemaking (e.g., the costs of
additional shielding due to lower dose rates) that the NRC should
consider with respect to alternate dose rates for transportable
microreactors? Please provide a basis for your response.
Q7-3: Provide feedback on the impact to international and
interstate shipments if there were alternate transportation package
dose rate limits for transportable microreactors.
Q7-4: What assumptions should the NRC use when estimating
the number of shipments, exposure scenarios, and expected dose rates
for fresh and irradiated transportable microreactors? Please provide a
basis for your response.
8. Fitness For Duty for Proposed Part 57. The proposed rule would
allow a licensee or other entity to implement 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.
Q8-1: To support licensees developing an FFD program
tailored to their own specifications, what core elements (such as
program policy and governance; program scope and applicability;
behavioral observation; specimen collection and testing; substances
tested; pre-employment screening; for-cause and post-event measures;
periodic medical fitness evaluations for licensed reactor operators;
program-related training; program audits and corrective actions; and
supportive resources, such as an employee assistance program or other
equivalent substance abuse counseling) should the NRC include in its
program requirements or guidance to help licensees ensure the
trustworthiness, reliability, and fitness of personnel and to support
FFD program consistency within the industry? Please provide a basis for
your response.
Q8-2: What approach or methodology should be used to
determine whether a credible operator or maintenance error could result
in exceeding the dose-based entry criterion specified in proposed Sec.
57.25(a)? Please provide a basis for your response.
Q8-3: What alternative criteria could be applied to
proposed Sec. 26.3(f)(3) to determine whether a licensee should be
permitted to implement an FFD program of its own specification or be
required to implement either the requirements of part 26 except
subparts K and P or the program described in proposed subpart P of part
26? Please provide a basis for your response.
9. Establishing Schedules for Part 57 Applications in the NRC's
Contested Hearing Process. In response to the Accelerating Deployment
of Versatile, Advanced Nuclear for Clean Energy Act of 2024 and E.O.
14300, section 5(j), the NRC has published a proposed rule to
streamline the NRC's contested hearing process for licensing
proceedings (91 FR 10450; March 3, 2026). As part of that proposed
rule, the NRC proposes to
[[Page 23682]]
establish strict hearing schedules for different types of applications,
including special requirements for highly expedited proceedings to
ensure that they are completed more promptly than they otherwise would
be to support expedited NRC decision-making on the underlying
applications. These special requirements include shorter filing periods
(e.g., for hearing requests, answers to hearing requests, new or
amended contention, and motions) and shorter deadlines for the
completion of evidentiary hearings. The NRC proposes to establish a new
term, ``highly expedited proceeding,'' in Sec. 2.4, ``Definitions,''
to define which proceedings are subject to these special requirements.
The rationale and detailed provisions for this proposal are described
in the proposed rule to streamline the NRC's contested hearing process
for licensing proceedings.
Q9-1: Consistent with the objectives of this proposed rule
to support high-volume licensing of microreactors and other reactors
with comparable risk profiles, should the NRC include certain proposed
part 57 applications within the definition of ``highly expedited
proceeding'' if the NRC issues a final rule modifying the NRC's
contested hearing process with special requirements for highly
expedited proceedings? Specifically, when a proposed part 57
application references an NRC approval providing finality on the design
in the adjudicatory proceeding, the scope of issues for adjudication
would be narrow, supporting an even more expedited schedule for filings
and decisions. Licensee-initiated amendments to proposed part 57
licenses should be similarly narrow. Therefore, should the NRC include
these types of proposed part 57 applications within the Sec. 2.4
definition of ``highly expedited proceeding'' and thereby apply
requirements for highly expedited proceedings to these applications? If
these applications were to be included within the scope of highly
expedited proceedings, should the NRC include the following definition
of ``highly expedited proceeding'' in Sec. 2.4 (underlined and
strikeout text shows potential changes to the definition of this term
in the proposed rule to streamline the NRC's contested hearing
process):
[GRAPHIC] [TIFF OMITTED] TP01MY26.028
Q9-2: What hearing schedule requirements should apply to
proposed part 57 joint applications for construction permits and
operating licenses that would not be included within the proposed
definition of ``highly expedited proceeding''? Under the proposed rule
to streamline the NRC's contested hearing process, 10 CFR part 50 or 52
applications for new reactor licenses with no design finality in the
adjudicatory proceeding would be subject to the longest hearing
schedules because these are considered to be the most complex
applications. However, proposed part 57 is limited to smaller reactors
with less complex designs and operational characteristics and low
potential radiological consequences, which should limit the potential
complexity of the license application. Also, proposed part 57 is
intended to support more expedited reviews. Therefore, should the NRC
treat proposed part 57 applications that are not within the proposed
definition of ``highly expedited proceeding'' in accordance with the
proposed hearing schedules that would apply to most types of license
applications, such as 10 CFR part 54 license renewals, rather than the
longer hearing schedules reserved for the most complex applications?
Please provide a basis for your response.
10. Remote Operations and Autonomous Operations. Proposed part 57
would allow remote operations and autonomous operations, which is
expected to be a paradigm shift for the nuclear industry and the NRC.
Q10-1: Should the NRC allow remote operations and
autonomous operations of nuclear power plants that demonstrate low
consequences? What, if any, additional requirements and guidance are
necessary for the regulatory review of remote operation and autonomous
operation as part of the rapid licensing envisioned under part 57?
Please provide a basis for your response.
11. Application of the Single Failure Criterion. Applicants are
encouraged to balance their selected risk assessment methods between
traditional deterministic approaches such as application of single
failure criterion methodologies (see SECY-77-439) with risk-insights
(see SRM-SECY-19-0036) as the most effective path forward to achieving
rapid and streamlined licensing decisions. While the single failure
criterion is a cornerstone of nuclear safety, the NRC recognizes that
it is not sufficient by itself for ensuring reasonable assurance of
adequate protection. Instead, it serves as just one analytical tool
within a broader, multi-layered framework, designed to achieve reliable
shutdown, cooling, and accident mitigation of a facility. The
Commission's ``Policy Statement on the Regulation of Advanced
Reactors'' (73 FR 60612, October 14, 2008) includes expectations that
advanced reactors will provide enhanced margins of safety and/or use
simplified, inherent, passive, or other innovative means to accomplish
their safety and security functions. The policy statement provides
examples of design attributes that could assist in establishing the
acceptability or licensability of a proposed advanced reactor design
and explains that incorporating these attributes may promote more
efficient and effective design reviews. However, some licensing
problems continue to exist in specific interpretations and applications
of the single failure criterion for advanced reactor designs. Some of
these issues were described in
[[Page 23683]]
SRM-SECY-19-0036, and the Commission directed the NRC staff to apply
risk-informed principles when strict, prescriptive application of
deterministic criteria such as the single failure criterion is
unnecessary to provide for reasonable assurance of adequate protection
of public health and safety.
Q11-1: To what extent should the proposed part 57
implementation guidance consider the single failure criterion as a
desired attribute to enhance reliability and defense in depth, rather
than as a limiting factor in determining whether reasonable assurance
of adequate protection exists for advanced reactor designs with
enhanced margins of safety and/or that use simplified, inherent,
passive, or other innovative means to accomplish their safety and
security functions? Please provide a basis for the response.
Q11-2: Are there criteria or methods that can be included
in the proposed part 57 implementation guidance that provide balance
between the use of deterministic methods such as the single failure
criterion and applicant-derived risk information to provide for
reasonable assurance of adequate protection of public health and
safety? Please provide a basis for the response.
12. Alternatives considered in the Regulatory Analysis. The NRC
invites comment on the alternatives considered and the rationale for
establishing proposed part 57 rather than using other frameworks (i.e.,
part 50, part 52, or part 53).
Q12-1: Are the NRC's conclusions--existing pathways
designed for large or specialized facilities (e.g., part 52 with
inspections, tests, analyses, and acceptance criteria (ITAAC) or part
50 requirements tailored to large LWRs) would impose unnecessary burden
and extend review timelines for microreactors--accurate and
sufficiently supported?
Q12-2: What additional, intermediate, or hybrid
alternatives (e.g., targeted modifications to part 52, streamlined
ITAAC constructs, or scoped use of part 53 elements) should the NRC
evaluate to meet the statutory objectives while minimizing cost and
schedule impacts? Please provide data, examples, or suggested
regulatory text that could enable rapid, high-volume licensing of
microreactors within or alongside existing regulations.
VIII. Regulatory Flexibility Certification
As required by the Regulatory Flexibility Act of 1980, 5 U.S.C.
605(b), the Commission certifies that this rule, if adopted, will not
have a significant economic impact on a substantial number of small
entities. This proposed rule affects only the licensing and operation
of nuclear power plants. The companies that own these plants do not
fall within the scope of the definition of ``small entities'' set forth
in the Regulatory Flexibility Act or the size standards established by
the NRC (10 CFR 2.810).
IX. Regulatory Analysis
The NRC has prepared a draft regulatory analysis on this proposed
regulation. The analysis examines the costs and benefits of the
alternatives considered by the NRC. The conclusion from the analysis is
that this proposed rule and associated guidance would result in net
averted costs (cost savings) 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 due to reductions in exemption
requests. The analysis also assumes 2,235 applicants under part 57 over
the 40 years of the analysis. As the number of applicants increases, so
do the estimated averted costs. The NRC requests public comment on the
draft regulatory analysis, which is available as indicated in the
``Availability of Documents'' section of this document. Comments on the
draft analysis may be submitted to the NRC as indicated under the
ADDRESSES caption of this document.
X. Backfitting and Issue Finality
This section describes the backfitting and issue finality
implications of this proposed rule and the draft guidance document
described in section XVIII, ``Availability of Guidance,'' of this
document, as applied to pertinent NRC approvals and certain applicants
that reference NRC approvals in their applications. The NRC's current
backfitting provisions relevant to this proposed rule appear in Sec.
50.109, Sec. 70.76, and Sec. 72.62, all entitled ``Backfitting,'' and
apply to holders of construction permits and operating licenses for
commercial and industrial purposes under part 50, holders of licenses
under part 70, and holders of general or specific licenses under part
72, respectively. Issue finality provisions (analogous to the
backfitting provisions in Sec. 50.109) for approvals under part 52 are
in various provisions of part 52. The NRC Management Directive 8.4,
``Management of Backfitting, Forward Fitting, Issue Finality, and
Information Requests,'' describes the Commission's policies on
backfitting and issue finality.
This proposed rule would provide a regulatory scheme for entities
to apply for approvals under parts 30, 40, 57, 70, 71, and 72. The
parts 50, 70, and 72 backfitting provisions and part 52 issue finality
provisions apply to actions taken by the NRC under parts 50, 70, 72,
and 52, respectively, or actions taken by the NRC under other parts of
10 CFR chapter I that, for holders of certain approvals under part 50,
70, 72, or 52, inextricably affect their activities regulated under
part 50, 70, 72, or 52, respectively. Issuance and implementation of
proposed part 57 would not constitute actions taken under part 50, 70,
72, or 52. Therefore, the issuance and implementation of proposed part
57 would not affect part 50, 70, 72, or 52 entities' activities
regulated under those parts. The addition of part 57 through this
proposed rule would not be within the scope of the part 50, 70, or 72
backfitting or part 52 issue finality provisions.
The NRC also proposes conforming changes to parts 1, 2, 10, 11, 19,
20, 21, 25, 26, 30, 40, 50, 51, 70, 72, 73, 74, 75, 95, 140, and 150 to
reflect the addition of part 57 (see section VI.A of this document).
These changes would not meet the definition of ``backfitting'' in Sec.
50.109 or Sec. 70.76 because the proposed changes would not modify or
add to the systems, structures, components, or design of a facility or
to the procedures or organization required to operate a facility under
part 50 or 70. These changes would not meet the definition of
``backfitting'' in Sec. 72.62 because the proposed changes would not
add, eliminate, or modify the SSCs of an independent spent fuel storage
installation or the procedures or organization required to operate an
independent spent fuel storage installation. These proposed changes
would not inextricably affect activities regulated under parts 50, 52,
70, or 72. Therefore, the proposed changes to parts 1, 2, 10, 11, 19,
20, 21, 25, 26, 30, 40, 50, 51, 70, 72, 73, 74, 75, 95, 140, and 150
would not constitute backfitting under parts 50, 70, or 72 or affect
the issue finality of an approval under part 52.
The NRC is issuing one draft guidance document that, if issued as a
final guidance document, would provide guidance on the methods
acceptable to the NRC for complying with aspects of this proposed rule.
This guidance would not apply to holders of approvals issued under part
50 or part 52. Although the guidance could apply to holders of part 70
or part 72 licenses, the guidance would apply to them only in relation
to a part 57 license, and there would be no
[[Page 23684]]
part 57 licenses at the time the final guidance is issued. Further, as
discussed in the guidance documents, applicants and licensees would not
be required to comply with the positions set forth in the guidance.
Therefore, issuance of the guidance documents as final guidance would
not constitute backfitting under part 50, 70, or 72 or affect the issue
finality of any approval issued under part 52.
XI. Cumulative Effects of Regulation
The NRC seeks to minimize potential negative consequences resulting
from the cumulative effects of regulation (CER). The NRC believes that
the de-regulatory impacts of this rulemaking activity are unlikely to
cause implementation challenges for stakeholders. In addition, during
the pendency of this rulemaking, the NRC is deprioritizing issuance of
regulatory actions that might influence the implementation date for the
new rule requirements (e.g., orders, generic communications, license
amendment requests, and inspection findings of a generic nature).
To fully understand any potential CER implications that could
result from this rulemaking, the NRC is asking the following questions.
Response to these questions is voluntary and any input will be
considered during development of the final rule.
1. The NRC is proposing an effective date that will be 30 days
after the date of publication of a final rule. Does this provide
sufficient time to implement the proposed requirements? Please provide
a rationale for your response.
2. Are there unintended consequences related to this rulemaking and
how should they be addressed? Please provide a rationale for your
response.
3. Please comment on the NRC's cost and benefit estimates in the
regulatory analysis that supports this proposed rule.
XII. Plain Writing
The Plain Writing Act of 2010 (Pub. L. 111-274) requires Federal
agencies to write documents in a clear, concise, and well-organized
manner. The NRC has written this document to be consistent with the
Plain Writing Act as well as the Presidential Memorandum, ``Plain
Language in Government Writing,'' published June 10, 1998 (63 FR
31885). The NRC requests comment on this document with respect to the
clarity and effectiveness of the language used.
XIII. Environmental Assessment and Proposed Finding of No Significant
Environmental Impact
A. Introduction
The NRC has prepared this environmental assessment (EA) in
compliance with the agency's environmental review requirements in 10
CFR part 51, ``Environmental Protection Regulations for Domestic
Licensing and Related Regulatory Functions,'' which implement the
National Environmental Policy Act of 1969, as amended. This EA
evaluates and documents the potential environmental impacts resulting
from the proposed rulemaking related to amending the regulations by
creating an alternative regulatory framework for licensing
microreactors and other reactors with comparable risk profiles.
Sections III and IV of this document provide the background
regarding E.O. 14300, the proposed action, along with the purpose of
and need for the proposed action. Section V of this document describes
the structure of the proposed part 57. Further discussion of these
topics does not need to be repeated in this EA. The organization of
this EA addresses the conforming changes under this proposed part 57
rule, environmental impacts of the proposed action, the environmental
impacts of the alternative to the proposed action, agencies and persons
consulted, proposed finding of no significant environmental impacts,
stakeholder interactions, and the references noted in this EA.
B. Conforming Changes
This rulemaking would make conforming changes throughout 10 CFR
chapter I. Table B.1-1 lists the chapter I parts with conforming
changes for this proposed rule. Most of these changes would only insert
the appropriate part 57 cross-reference and are considered
administrative changes.
[[Page 23685]]
[GRAPHIC] [TIFF OMITTED] TP01MY26.007
C. Environmental Impacts of the Proposed Action
Most of the subparts to the proposed part 57 are related to
establishing programs and related procedures rather than actions
requiring technical analysis with approved methodologies or guidance.
Additionally, many of these subparts establish technical requirements
that would be equivalent to companion regulations under 10 CFR part 21,
part 50, part 52, part 70, part 71, part 72, part 73, and part 74.
Thus, these subparts are procedural provisions or incorporate similar
requirements as existing regulations and are not substantive
environmentally different regulations. Therefore, since this group of
subparts would generally address administrative, procedural processes,
and technical requirements equivalent to ones under various parts under
10 CFR, their implementation would result in no significantly different
environmental impacts under this rule. The proposed part 57 subpart
regulations with their equivalent regulations are listed in Table C-1.
BILLING CODE 7590-01-P
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BILLING CODE 7590-01-C
C.1 Part 57 Subparts Not Related to Existing 10 CFR Regulations
Subpart C--Construction Permits and Operating Licenses
As noted in Table C-1, proposed subpart C would contain several
sections that would be similar to existing regulations in part 50.
Proposed part 57 also relies upon regulatory radiological limits under
10 CFR part 20 and an entry criterion of 1 rem (10 mSv) dose threshold
to any individual located in the unrestricted area. Similarly, for
other facilities with comparable risk profiles, such as research
reactors licensed under 10 CFR part 50, the NRC applies the
radiological limit requirements in 10 CFR part 20 and a comparable
accident dose criterion of 1 rem (10 mSv), as specified in Sec.
50.34(a)(1)(i). Other regulations under proposed subpart C would be new
regulations that would also provide assurance that the complete design
had been reviewed and approved by the NRC or are related to required
processes and procedures necessary to further support such reasonable
assurance.
Proposed Sec. 57.45(d) would establish a general license for
construction activities on a site that is specified in a joint
application for a construction permit and associated operating
license(s) under proposed part 57, subject to certain conditions. These
conditions would include requirements that the application references a
reactor of the same design that had been constructed and placed into
operation under Commission oversight and that had met the criteria for
a categorical exclusion under the regulations in proposed subpart K of
part 57. In addition, the proposed Sec. 57.45(d)(4) would require that
the general licensee must not allow special nuclear material or
radioactive material that would be associated with operation under an
operating license issued pursuant to proposed part 57 to be brought to
the site under the general license. Therefore, the activities that
would be conducted under the general license would not give rise to
nuclear or radiological hazards. Also, proposed Sec. 57.45(d)(1)(iii)
would require that the application submitted by the general licensee
would include a plan for redress of any adverse environmental impact
from conduct of activities under the general license should such
redress be necessary.
Therefore, the regulations under proposed subpart C would provide
the
[[Page 23692]]
same level of protection of public health and safety as existing
regulations, and there would be no significantly different
environmental effects with implementing this new regulation.
Subpart D--Manufacturing Licenses
As previously noted, the proposed subpart D would address
applications for, issuance of, and other provisions related to MLs
covering manufacturing activities at one or more licensee facilities.
These proposed requirements would be largely equivalent to those in
part 52 for MLs. The most significant change proposed for MLs in part
57 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. Because
the 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 same level of reasonable assurance of
adequate protection of public health and safety would be maintained as
for currently licensed operating plants for their receipt of SNM. Thus,
implementation of subpart D would provide an equivalent level of
reporting, administrative, and safety requirements as the current ML
and fuel possession and loading regulatory framework with no
significant environmental impacts.
Subpart K--Categorical Exclusions
Categorical exclusions provide a mechanism to identify types of
Federal actions that normally do not have significant environmental
effects to the human environment and for which neither an environmental
assessment nor environmental impact statement is normally required.
This ensures that resources are not expended conducting environmental
analysis of proposals that do not present potential for significant
environmental impacts. The proposed Sec. 57.350 establishes the
criteria for determining whether a categorial exclusion applies in
support of a license under this part. Additionally, determining whether
a categorical exclusion applies is a NEPA process to inform the
decision-maker of the environmental impacts for issuing a license and,
thus, an administrative step. Therefore, there would be no significant
environmental effects with implementing this proposed new categorical
exclusion regulation.
C.2 Changes to Other Parts of Chapter 10 of the CFR
C.2.1 10 CFR Part 25
The conforming changes to part 25 for activities in connection with
the proposed part 57 are a revision to a definition, the addition of a
reference regarding access authorizations for individuals who need
access to classified information, and an update to apply the
requirements for classified visits to licensees and applicants under
proposed part 57. Therefore, these changes to part 25 would be at a
level equivalent to the current 10 CFR part 25 regulatory framework and
its implementation would have no significantly different environmental
impacts.
C.2.2 10 CFR Part 26
As stated in section VI.B. of this document, proposed part 57 would
add a new subpart P in 10 CFR part 26, ``Fitness for Duty Programs,''
and make other conforming changes to existing part 26 provisions. The
NRC proposes a flexible, technology-inclusive, risk-informed, and
performance-based approach with options to the application of drug and
alcohol testing and fatigue management requirements for facilities
licensed under proposed part 57. Proposed part 57 licensees and other
entities could implement requirements in proposed subpart P of part 26,
all the requirements of part 26 except subparts K and P, or an FFD
program of their specification. Notwithstanding the type of FFD program
a licensee or other entity would implement, the licensees and other
entity that would apply for or have been issued an OL or CP under
proposed part 57 would be required, no later than the start of
construction activities, to implement the FFD program. Holders of an ML
under proposed part 57 would be required to implement their FFD program
before commencing activities that assemble a manufactured reactor.
Concerning an FFD program of their specification, licensees and
other entities that would apply for or would have been issued an OL or
CP under proposed part 57, and holders of an ML under proposed part 57,
could elect to implement an FFD program of their specification only if
the licensee's or other entity's reactor manufactured under an ML
issued under proposed part 57, constructed under a construction permit
issued under proposed part 57, or operated under an OL issued under
proposed part 57, as applicable, would not require operator action 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.
The FFD requirements would be commensurate with the radiological
risks presented by the facilities in question (i.e., reactors with
comparable risk profiles). The NRC used operating experience to propose
regulatory flexibility in the new FFD framework to help support a
licensee's or other entity's response to changes in societal drug use,
drug testing technologies and processes, and FFD program performance.
The flexibility would also help in FFD implementation because of the
wide variety of staff sizes anticipated at different facilities
licensed under proposed part 57 and the geographically remote locations
in which these facilities may be sited. Therefore, an FFD program
implemented under this proposed rule would be at a level of risk-
informed equivalency to the current 10 CFR part 26 regulatory framework
ensuring adequate protection of the public health and safety while
providing flexibility to a proposed part 57 license, and its
implementation would have no significantly different environmental
impacts.
C.2.3 10 CFR Part 51
Additional text under 10 CFR 51.4, ``Definitions,'' for
``construction'' would point to the definition of ``construction''
under Sec. 57.3 to account for differences among that definition and
the definitions of ``construction'' under 10 CFR parts 50 and 52. This
proposed change to the definition of ``construction'' in 10 CFR part 51
would be administrative in application and, as such, would not have a
significant environmental impact.
C.2.4 10 CFR Part 73
Changes to part 73 in support of proposed part 57 address
cybersecurity programs by implementing a graded approach to determine
the level of cybersecurity protection required for digital computers,
communication systems, and networks. The changes are based on (1) the
operating experience from power reactors and insights from cyber-
related assessments of fuel cycle facilities; and (2) the existing
Sec. 73.54 framework. Differences between the Sec. 73.54 requirements
and those proposed by part 57 changes to part 73 are primarily based on
the implementation of a consequence-based approach to cybersecurity.
This consequence-based approach would provide flexibility to
accommodate the wide range of reactor technologies and would account
for the differing risk
[[Page 23693]]
levels among reactor technologies. Specifically, the proposed new
section would require licensees to demonstrate reasonable assurance of
protection against cyberattacks in a manner that is commensurate with
the potential consequences from those attacks. Thus, the part 73
changes in this rule would provide a similar level of protection from
cyberattack as the current regulations and its implementation would
have no significantly different environmental impacts.
C.3 Summary of the Environmental Impacts of the Proposed Action
With regard to potential environmental effects, implementation of
the proposed part 57 rule would not have a significant environmental
impact. Proposed requirements would be administrative in application, a
matter of procedure, or would provide an equivalent level of safety and
security for protection of public health and safety as existing
regulations with no significant environmental effects to the human
environment with implementing this new regulation.
In addition, requirements under proposed part 57 would not affect
any threatened or endangered species or historic properties since this
proposed rule would result in no physical changes to the environment.
Accordingly, the NRC finds that this proposed rulemaking action
would not have a significant effect on the quality of the human
environment.
D. Environmental Impacts of the Alternative to the Proposed Agency
Action
Under the no-action alternative (i.e., the status quo), the
regulations would not change. Licensees would continue to be required
to meet current regulations (namely, 10 CFR part 50 and 10 CFR part 52)
or seek relief using the existing regulatory framework. As stated in
section C of this EA, the proposed rule would not result in a
significant impact on the environment because reactors licensed under
the proposed part 57 are expected to have a smaller impact on the
affected environment than plants licensed under the current
regulations, and the proposed rule would offer an equivalent level of
safety as provided by the current regulations. This rulemaking provides
an additional option to existing processes to license a microreactor or
other reactor with a comparable risk profile and does not add any
additional environmental requirements. Therefore, there would be no
difference in environmental impacts between the no-action alternative
and the proposed rule. The NRC would analyze the environmental impacts
of a license application under existing regulations and guidance for
the no-action alternative and would continue to analyze the
environmental impacts of applications, exemptions, and license
amendment requests on a case-by-case basis. The NRC describes the costs
and benefits of the no-action alternative and the proposed action in
the regulatory analysis for the proposed rule.
E. Agencies and Persons Consulted
The NRC developed the proposed rule and is requesting public
comment on this draft EA. The NRC intends to hold a public meeting
during the proposed rule comment period to allow stakeholders to ask
questions about the proposed rule and this EA. The agency will consider
comments received on the docket as it develops the final rule and the
final EA. The NRC will issue the final EA when it publishes the final
rule.
The proposed rule is one step in the rulemaking process. During the
development of this proposed rule, the NRC conducted public meetings
and other interactions with stakeholders related to the development of
the part 57 regulations. Table G-1 in Section G of this EA provides
details about stakeholder interactions.
The proposed rule would provide an equivalent level of safety as
the current regulations in 10 CFR part 50 and 10 CFR part 52 and would
result in no significant impact on the environment. As such, the
rulemaking would not impact threatened or endangered species or
critical habitat; the NRC has determined that a section 7 consultation
under the Endangered Species Act is not necessary. Likewise, the NRC
has determined that the proposed rulemaking would not cause any adverse
effects to historic properties. Therefore, the NRC has determined that
no consultation is required under section 106 of the National Historic
Preservation Act.
F. Proposed Finding of No Significant Environmental Impacts
The Commission has determined under the National Environmental
Policy Act of 1969, as amended, and the Commission's regulations in 10
CFR part 51, that this rule, if adopted, would not be a major Federal
action significantly affecting the quality of the human environment,
and an environmental impact statement is not required. The basis of
this determination is that NRC's proposed action (rulemaking) would
provide adequate protection of the public health and safety and common
defense and security for microreactors and reactors with comparable
risk profiles without the need to grant specific exemptions or license
amendments in certain regulatory areas. Rulemaking would reduce the
need for exemptions from existing regulations and license amendment
requests and would support the principles of good regulation, including
openness, clarity, and reliability. Therefore, the proposed rulemaking
meets the need for the proposed agency action.
The determination of this EA is that this proposed agency action
would not have a significant effect on the quality of the human
environment. Public stakeholders should note, however, that comments on
any aspect of this EA may be submitted to the NRC as indicated under
the ADDRESSES caption.
The NRC has sent a copy of the EA and this proposed rule to every
State Liaison Officer and has requested comments.
G. Stakeholder Interactions
The stakeholder interactions for part 57 thus far are listed in
Table G-1 for interactions between the NRC and stakeholders during
public meetings and communications on issues related to the part 57
rulemaking. The NRC received feedback from various stakeholders on part
57 during or as a result of these interactions.
[[Page 23694]]
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H. Environmental Assessment References
U.S. Department of Defense (DOD). 2025a. Department of Defense
National Environmental Policy Act Implementing Procedures. https://www.denix.osd.mil/nepa/denix-files/sites/55/2025/06/DoD-NEPA-Procedures-FINAL.pdf. June 30, 2025.
U.S. Department of Defense (DOD). 2025b. Department of Defense
National Environmental Policy Act Implementing Procedures: Appendix
A Department of Defense Categorical Exclusions (CATEX). https://www.denix.osd.mil/nepa/denix-files/sites/55/2025/06/DOD-NEPA-Procedures-APPENDIX-A_FINAL.pdf. June 30, 2025.
U.S. Department of Energy (DOE). 2025. Revision of National
Environmental Policy Act Implementing Procedures. Interim final
rule; request for comments. DOE-HQ-2025-0026, RIN 1990-AA52. https://federalregister.gov/d/2025-12383. July 3, 2025.
XIV. Paperwork Reduction Act
This proposed rule contains new or amended collections of
information subject to the Paperwork Reduction Act of 1995 (44 U.S.C.
3501 et seq). This proposed rule has been submitted to the Office of
Management and Budget (OMB) for review and approval of the information
collections.
Type of submission: New.
The title of the information collection: Licensing Requirements for
Microreactors and Other Reactors with Comparable Risk Profiles.
OMB Approval Numbers: (3150-0002, 3150-0024, 3150-0090, 3150-0104,
3150-0146, 3150-0238, 3150-0272, and 3150-XXXX).
The form number if applicable: NRC Forms 361T, 366, 366A, 366B,
396, 398, 893, and 894.
How often the collection is required or requested: Once, on
occasion, every 30 days, biannually, annually, biennially, every four
years, every five years, every ten years.
Who will be required or asked to respond: Part 57 licensees and
license applicants for reactors to be licensed under part 57.
An estimate of the number of annual responses:
10 CFR part 26: 1,576.6 (13 reporting responses + 5.7 recordkeepers
+ 1,557.9 third-party disclosures).
10 CFR part 57: 376.4 (33.9 reporting responses + 9 recordkeepers +
333.5 third-party disclosures).
10 CFR part 73: 2.7 (0 reporting responses + 2.7 recordkeepers + 0
third-party disclosures).
NRC Form 361T: 18 reporting responses.
NRC Forms 366, 366A, and 366B: 13 reporting responses.
NRC Form 396: 68 (34 reporting responses + 34 recordkeepers).
NRC Form 398: 34 reporting responses.
NRC Forms 893 and 894: 312 reporting responses.
The estimated number of annual respondents:
10 CFR Part 26: 5.7 respondents.
10 CFR Part 57: 9 respondents.
10 CFR Part 73: 2.7 respondents.
NRC Form 361T: 3.7 respondents.
NRC Forms 366, 366A, and 366B: 3.7 respondents.
NRC Form 396: 2.3 respondents.
NRC Form 398: 2.3 respondents.
NRC Forms 893 and 894: 3.7 respondents.
An estimate of the total number of hours needed annually to comply
with the information collection requirement or request:
10 CFR Part 26: 7,458.7 (113.5 reporting + 6,320.3 recordkeeping +
1,024.9 third-party disclosures).
10 CFR Part 57: 1,013,327.8 (971,607.4 reporting + 41,637.0
recordkeeping + 83.4 third-party disclosures).
10 CFR Part 73: 3,898.2 (0 reporting + 3,898.2 recordkeeping + 0
third-party disclosures).
NRC Form 361T: 9.
NRC Forms 366, 366A, and 366B: 832.
NRC Form 396: 42.5.
NRC Form 398: 87.
NRC Forms 893 and 894: 578.
Abstract: The NRC is proposing 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, consistent with the licensing
framework for non-power production or utilization facilities. 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. The NRC's goal in this rulemaking is to expedite the
licensing process for microreactors and other reactors with comparable
risk profiles.
The proposed rule covers diverse topics, which result in
recordkeeping and reporting requirements related to construction and
manufacturing, contents of applications, plant design and analysis,
facility operations, decommissioning, FFD, physical security,
cybersecurity, siting, programs, staffing, and quality assurance.
In addition to the new information collections in the proposed
regulations, proposed part 57 would result in new collections via NRC
Forms 361T, 366, 366A, 366B, 396, 398, 893, and 894. A new version of
NRC Form 361 (NRC Form 361T) would be created for use by proposed part
57 licensees, covering an equivalent scope as the requirements in Sec.
50.72, but without LWR-specific terminology to ensure technology
inclusiveness. NRC Forms 366, 366A, and 366B would be modified to
include reportable events in proposed part 57, subpart Q, covering an
equivalent scope as the requirements in Sec. 50.73, but without LWR-
specific terminology to ensure technology inclusiveness. NRC Forms 396
and 398 would be modified to satisfy requirements in proposed part 57,
subpart P, to certify the medical fitness of an applicant for an
operator or senior operator license. Finally, the proposed rule would
require part 57 licensees to use NRC Forms 893 and 894 to report on
positive drug and alcohol test results (NRC Form 893) and annual
fitness-for-duty program performance
[[Page 23695]]
(NRC Form 894), as required in proposed Sec. 26.917.
The NRC is seeking public comment on the potential impact of the
information collections contained in this proposed rule and on the
following issues:
1. Is the proposed information collection necessary for the proper
performance of the functions of the NRC, including whether the
information will have practical utility? Please explain your response.
2. Is the estimate of the burden of the proposed information
collection accurate? Please explain your response.
3. Is there a way to enhance the quality, utility, and clarity of
the information to be collected? Please explain your response.
4. How can the burden of the proposed information collection on
respondents be minimized, including the use of automated collection
techniques or other forms of information technology? Please explain
your response.
A copy of the OMB clearance package and proposed rule are available
in the ``Availability of Documents'' section of this document or may be
viewed free of charge by contacting the NRC's Public Document Room
reference staff at 1-800-397-4209, at 301-415-4737, or by email to
[email protected]. You may obtain information and comment on
submissions related to the OMB clearance package by searching on
https://www.regulations.gov under Docket ID NRC-2025-0379.
You may submit comments on any aspect of these proposed information
collection(s), including suggestions for reducing the burden and on
these issues, by the following method:
Federal Rulemaking Website: Go to https://www.regulations.gov and search for Docket ID NRC-2025-0379.
Submit comments by June 1, 2026.
Public Protection Notification
The NRC may not conduct or sponsor, and a person is not required to
respond to, a collection of information unless the document requesting
or requiring the collection displays a currently valid OMB control
number.
XV. Executive Orders
The following are Executive orders that are related to this
proposed rule:
A. Executive Order 12866: Regulatory Planning and Review (as Amended by
Executive Order 14215, Ensuring Accountability for All Agencies)
The Office of Information and Regulatory Affairs (OIRA) has
determined that this proposed rule is an economically significant
regulatory action. Accordingly, the NRC submitted this proposed rule to
OIRA for review. The NRC is required to conduct an economic analysis in
accordance with section 6(a)(3)(B) of E.O. 12866. More information can
be found in section IX, ``Regulatory Analysis,'' of this document.
B. Executive Order 14154: Unleashing American Energy
The NRC has examined this proposed rule and has determined that it
is consistent with the policies and directives outlined in E.O. 14154.
C. Executive Order 14192: Unleashing Prosperity Through Deregulation
This action is tentatively determined to be a deregulatory action
as defined by E.O. 14192. Details on the estimated costs of this
proposed rule can be found in Section IX, of this document,
``Regulatory Analysis.''
D. Executive Order 14270: Zero-Based Regulatory Budgeting To Unleash
American Energy
E.O. 14270, ``Zero-Based Regulatory Budgeting to Unleash American
Energy,'' requires the NRC to insert a conditional sunset date into all
new or amended NRC regulations provided the regulations are (1)
promulgated under the Atomic Energy Act of 1954, as amended (AEA), the
Energy Reorganization Act of 1974, as amended (ERA), or the Nuclear
Waste Policy Act of 1982, as amended (NWPA); (2) not statutorily
required; and (3) not part of the NRC's permitting regime. The NRC
determined that the regulatory changes proposed in this rule are
required because they would be necessary for providing reasonable
assurance of adequate protection of public health and safety and
provide for the common defense and security, and would be part of the
NRC's permitting regime authorized by the AEA. Therefore, the NRC views
this rulemaking to be outside the scope of E.O. 14270 and does not
propose to insert conditional sunset dates for the regulatory changes
in this proposed rule.
E. Executive Order 14294: Fighting Overcriminalization in Federal
Regulations
This proposed rule includes Federal regulations that, if adopted,
would be enforceable by criminal penalty, as authorized by section 223
of the AEA. Therefore, per E. O. 14294, those regulations constitute
``criminal regulatory offenses.''
For the purposes of section 223 of the AEA, the NRC is issuing this
proposed rule that would add a new part 57 and amend 10 CFR parts 19,
20, 21, 25, 26, 30, 40, 50, 70, 72, 73, 74, 95, and 140 under one or
more of sections 161(b), 161(i), or 161(o) of the AEA, except as noted
in Sec. Sec. 19.40(b), 20.2402(b), 21.62(b), 25.39(b), 26.825(b),
30.64(b), 40.82(b), 50.111(b), 57.385(b), 70.92(b), 72.86(b), 73.81(b),
74.84(b), 95.63(b), and 140.89(b). The applicability of criminal
penalties to regulations in parts 19, 20, 21, 25, 26, 30, 40, 50, 57,
70, 72, 73, 74, 95, and 140 is set forth in Sec. Sec. 19.40, 20.2402,
21.62, 25.39, 26.825, 30.64, 40.82, 50.111, 57.385, 70.92, 72.86,
73.81, 74.84, 95.63, and 140.89, respectively. Willful violations of
the 10 CFR parts 19, 20, 21, 25, 26, 30, 40, 50, 57, 70, 72, 73, 74,
95, and 140 regulations, other than those listed in Sec. Sec.
19.40(b), 20.2402(b), 21.62(b), 25.39(b), 26.825(b), 30.64(b),
40.82(b), 50.111(b), 57.385(b), 70.92(b), 72.86(b), 73.81(b), 74.84(b),
95.63(b), and 140.89(b) (including as updated by this proposed rule),
would be subject to criminal enforcement.
XVI. Voluntary Consensus Standards
The National Technology Transfer and Advancement Act of 1995,
Public Law 104-113, requires that Federal agencies use technical
standards that are developed or adopted by voluntary consensus
standards bodies unless the use of such a standard is inconsistent with
applicable law or otherwise impractical. In this proposed rule, the NRC
will revise regulations by adding a regulatory framework for
microreactors and other reactors with comparable risk profiles. This
action does not constitute the establishment of a standard that
contains generally applicable requirements.
XVII. Availability of Guidance
The NRC is issuing draft guidance in NUREG-2271, ``Guidelines for
Preparing and Reviewing Applications Under 10 CFR part 57,'' for
implementation of the proposed requirements in this rulemaking. The
draft guidance is available in ADAMS under Accession No. ML25259A304.
When finalized, NUREG-2271 would provide stakeholders with guidance for
implementing the final requirements contemplated by this proposed rule.
You may submit comments on the draft regulatory guidance by the methods
outlined in the ADDRESSES section of this document.
XVIII. Public Meeting
The NRC will conduct a public meeting on the proposed rule for the
purpose of describing the proposed rule to the public and answering
questions
[[Page 23696]]
from the public to facilitate public comments on the proposed rule.
The NRC will publish a notice of the location, time, and agenda of
the meeting in the Federal Register, on Regulations.gov, and on the
NRC's public meeting website within at least 10 calendar days before
the meeting. Stakeholders should monitor the NRC's public meeting
website for information about the public meeting at: https://www.nrc.gov/public-involve/public-meetings/index.cfm.
XIX. Availability of Documents
The documents identified in the following table are available to
interested persons through one or more of the following methods, as
indicated.
BILLING CODE 7590-01-P
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BILLING CODE 7590-01-C
The NRC may post materials related to this document, including
public comments, on the Federal rulemaking website at https://www.regulations.gov under Docket ID NRC-2025-0379. In addition, the
Federal rulemaking website allows members of the public to receive
alerts when changes or additions occur in a docket folder. To
subscribe: (1) navigate to the docket folder (NRC-2025-0379); (2) click
the ``Subscribe'' link; and (3) enter an email address and click on the
``Subscribe'' link.
List of Subjects
10 CFR Part 1
Flags, Organization and functions (Government Agencies), Seals and
insignia.
10 CFR Part 2
Administrative practice and procedure, Antitrust, Byproduct
material, Classified information, Confidential business information,
Freedom of information, Environmental protection, Hazardous waste,
Nuclear energy, Nuclear materials, Nuclear power plants and reactors,
Penalties, Reporting and recordkeeping requirements, Sex
discrimination, Source material, Special nuclear material, Waste
treatment and disposal.
10 CFR Part 10
Administrative practice and procedure, Classified information,
Government employees, Security measures.
10 CFR Part 11
Hazardous materials transportation, Investigations, Nuclear energy,
Nuclear materials, Penalties, Reporting and recordkeeping requirements,
Security measures, Special nuclear material.
[[Page 23702]]
10 CFR Part 19
Criminal penalties, Environmental protection, Nuclear Energy,
Nuclear materials, Nuclear power plants and reactors, Occupational
safety and health, Penalties, Radiation protection, Reporting and
recordkeeping requirements, Sex discrimination.
10 CFR Part 20
Byproduct material, Criminal penalties, Fusion, Hazardous waste,
Licensed material, Nuclear energy, Nuclear materials, Nuclear power
plants and reactors, Occupational safety and health, Packaging and
containers, Penalties, Radiation protection, Reporting and
recordkeeping requirements, Source material, Special nuclear material,
Waste treatment and disposal.
10 CFR Part 21
Nuclear power plants and reactors, Penalties, Radiation protection,
Reporting and recordkeeping requirements.
10 CFR Part 25
Classified information, Criminal penalties, Investigations,
Penalties, Reporting and recordkeeping requirements, Security measures.
10 CFR Part 26
Administrative practice and procedure, Alcohol abuse, Alcohol
testing, Appeals, Drug abuse, Drug testing, Employee assistance
programs, Fitness for duty, Management actions, Nuclear power plants
and reactors, Privacy, Protection of information, Radiation protection,
Reporting and recordkeeping requirements.
10 CFR Part 30
Byproduct material, Criminal penalties, Fusion, Government
contracts, Intergovernmental relations, Isotopes, Nuclear energy,
Nuclear materials, Penalties, Radiation protection, Reporting and
recordkeeping requirements, Whistleblowing.
10 CFR Part 40
Criminal penalties, Exports, Government contracts, Hazardous
materials transportation, Hazardous waste, Nuclear energy, Nuclear
materials, Penalties, Reporting and recordkeeping requirements, Source
material, Uranium, Whistleblowing.
10 CFR Part 50
Administrative practice and procedure, Antitrust, Backfitting,
Classified information, Criminal penalties, Education, Emergency
planning, Fire prevention, Fire protection, Intergovernmental
relations, Nuclear power plants and reactors, Penalties, Radiation
protection, Reactor siting criteria, Reporting and recordkeeping
requirements, Whistleblowing.
10 CFR Part 51
Administrative practice and procedure, Environmental impact
statements, Hazardous waste, Nuclear energy, Nuclear materials, Nuclear
power plants and reactors, Reporting and recordkeeping requirements.
10 CFR Part 57
Administrative practice and procedure, Antitrust, Backfitting,
Atomic energy, Construction permit, Combined license, Classified
information, Criminal Penalties, Early site permit, Emergency planning,
Fees, Fire prevention, Fire protection, Inspection, Intergovernmental
relations, Limited work authorization, Manufacturing license, Nuclear
energy, Nuclear materials, Nuclear power plants and reactors, Nuclear
safety, Operating license, Penalties, Prototype, Radiation Protection,
Radioactive materials, Reactor siting criteria, Reporting and
recordkeeping requirements, Standard design, Standard design
certification, Training programs.
10 CFR Part 70
Classified information, Criminal penalties, Emergency medical
services, Hazardous materials transportation, Material control and
accounting, Nuclear energy, Nuclear materials, Packaging and
containers, Penalties, Radiation protection, Reporting and
recordkeeping requirements, Scientific equipment, Security measures,
Special nuclear material, Whistleblowing.
10 CFR Part 72
Administrative practice and procedure, Hazardous waste, Indians,
Intergovernmental relations, Nuclear energy, Penalties, Radiation
protection, Reporting and recordkeeping requirements, Security
measures, Spent fuel, Whistleblowing.
10 CFR Part 73
Criminal penalties, Exports, Hazardous materials transportation,
Imports, Nuclear energy, Nuclear materials, Nuclear power plants and
reactors, Penalties, Reporting and recordkeeping requirements, Security
measures.
10 CFR Part 74
Accounting, Criminal penalties, Hazardous materials transportation,
Material control and accounting, Nuclear energy, Nuclear materials,
Packaging and containers, Penalties, Radiation protection, Reporting
and recordkeeping requirements, Scientific equipment, Special nuclear
material.
10 CFR Part 75
Criminal penalties, Intergovernmental relations, Nuclear energy,
Nuclear materials, Nuclear power plants and reactors, Penalties,
Reporting and recordkeeping requirements, Security measures, Treaties.
10 CFR Part 95
Classified information, Criminal penalties, Penalties, Reporting
and recordkeeping requirements, Security measures.
10 CFR Part 140
Insurance, Intergovernmental relations, Nuclear materials, Nuclear
power plants and reactors, Penalties, Reporting and recordkeeping
requirements.
10 CFR Part 150
Criminal penalties, Hazardous materials transportation,
Intergovernmental relations, Nuclear energy, Nuclear materials,
Penalties, Reporting and recordkeeping requirements, Security measures,
Source material, Special nuclear material.
For the reasons set out in the preamble and under the authority of
the Atomic Energy Act of 1954, as amended; the Energy Reorganization
Act of 1974, as amended; and 5 U.S.C. 552 and 553, the NRC is proposing
to amend 10 CFR parts 1, 2, 10, 11, 19, 20, 21, 25, 26, 30, 40, 50, 51,
70, 72, 73, 74, 75, 95, 140, and 150 and add 10 CFR part 57:
PART 1--STATEMENT OF ORGANIZATION AND GENERAL INFORMATION
0
1. The authority citation for part 1 continues to read as follows:
Authority: Atomic Energy Act of 1954, secs. 23, 25, 29, 161, 191
(42 U.S.C. 2033, 2035, 2039, 2201, 2241); Energy Reorganization Act
of 1974, secs. 201, 203, 204, 205, 209 (42 U.S.C. 5841, 5843, 5844,
5845, 5849); Administrative Procedure Act (5 U.S.C. 552, 553);
Reorganization Plan No. 1 of 1980, 5 U.S.C. Appendix (Reorganization
Plans).
Sec. 1.43 [Amended]
0
2. In Sec. 1.43, in paragraph (a)(2), add the number ``57,'' in
sequential order.
PART 2--AGENCY RULES OF PRACTICE AND PROCEDURE
0
3. The authority citation for part 2 continues to read as follows:
[[Page 23703]]
Authority: Atomic Energy Act of 1954, secs. 29, 53, 62, 63, 81,
102, 103, 104, 105, 161, 181, 182, 183, 184, 186, 189, 191, 234 (42
U.S.C. 2039, 2073, 2092, 2093, 2111, 2132, 2133, 2134, 2135, 2201,
2231, 2232, 2233, 2234, 2236, 2239, 2241, 2282); Energy
Reorganization Act of 1974, secs. 201, 206 (42 U.S.C. 5841, 5846);
Nuclear Waste Policy Act of 1982, secs. 114(f), 134, 135, 141 (42
U.S.C. 10134(f), 10154, 10155, 10161); Administrative Procedure Act
(5 U.S.C. 552, 553, 554, 557, 558); National Environmental Policy
Act of 1969 (42 U.S.C. 4332); 44 U.S.C. 3504 note. Section 2.205(j)
also issued under 28 U.S.C. 2461 note.
0
4. In Sec. 2.1, revise paragraph (e) to read as follows:
Sec. 2.1 Scope.
* * * * *
(e) Standard design approvals under part 52 or part 57 of this
chapter.
* * * * *
0
5. In Sec. 2.4, revise the definition for ``Facility'' to read as
follows:
Sec. 2.4 Definitions.
* * * * *
Facility means a production facility or a utilization facility as
defined in Sec. Sec. 50.2 and 57.3 of this chapter.
* * * * *
Sec. 2.100 [Amended]
0
6. In Sec. 2.100, remove the phrase ``subpart E of part 52'' and add
in its place the phrase ``subpart E of part 52 or subpart E of part
57''.
0
7. In Sec. 2.101, revise paragraph (a)(3)(i) to read as follows:
Sec. 2.101 Filing of application.
(a) * * *
(3) * * *
(i) Submit to the Director, Office of Nuclear Reactor Regulation,
or Director, Office of Nuclear Material Safety and Safeguards, as
appropriate, such additional copies as the regulations in part 50, part
51, and part 57 of this chapter require;
* * * * *
0
8. In Sec. 2.105, revise paragraphs (a)(4) and (a)(13) to read as
follows:
Sec. 2.105 Notice of proposed action.
(a) * * *
(4) An amendment to an operating license, combined license, or
manufacturing license for a facility of the type described in Sec.
50.21(b) or Sec. 50.22 of this chapter, as applicable, or for a
testing facility, as follows:
(i) If the Commission determines under Sec. 50.58 or Sec. 57.130
of this chapter that the amendment involves no significant hazards
consideration, though it will provide notice of opportunity for a
hearing pursuant to this section, it may make the amendment immediately
effective and grant a hearing thereafter; or
(ii) If the Commission determines under Sec. Sec. 50.58 and 50.91
or Sec. 57.130 of this chapter, as applicable, that an emergency
situation exists or that exigent circumstances exist and that the
amendment involves no significant hazards consideration, it will
provide notice of opportunity for a hearing pursuant to Sec. 2.106 (if
a hearing is requested, it will be held after issuance of the
amendment);
* * * * *
(13) A manufacturing license under subpart F of part 52 or subpart
D of part 57 of this chapter.
* * * * *
0
9. In Sec. 2.109, revise paragraphs (b) and (d) to read as follows:
Sec. 2.109 Effect of timely renewal application.
* * * * *
(b) If the licensee of a nuclear power plant of the type described
in Sec. 50.21(b) or Sec. 50.22 of this chapter files a sufficient
application for renewal of either an operating license or a combined
license at least 5 years before the expiration of the existing license,
the existing license will not be deemed to have expired until the
application has been finally determined.
* * * * *
(d) If the licensee of a manufacturing license under subpart F of
part 52, or under subpart D of part 57 of this chapter files a
sufficient application for renewal under Sec. 52.177 or Sec. 57.190
of this chapter at least 12 months before the expiration of the
existing license, the existing license will not be deemed to have
expired until the application has been finally determined.
* * * * *
0
10. In Sec. 2.110, revise paragraphs (a)(1) and (b) to read as
follows:
Sec. 2.110 Filing and administrative action on submittals for
standard design approval or early review of site suitability issues.
(a)(1) A submittal for a standard design approval under subpart E
of part 52 or under subpart E of part 57 of this chapter shall be
subject to Sec. Sec. 2.101(a) and 2.390 to the same extent as if it
were an application for a permit or license.
* * * * *
(b) Upon initiation of review by the NRC staff of a submittal for
an early review of site suitability issues under appendix Q to part 50
of this chapter, or for a standard design approval under subpart E of
part 52 or under subpart E of part 57 of this chapter, the Director,
Office of Nuclear Reactor Regulation, shall publish in the Federal
Register a notice of receipt of the submittal, inviting comments from
interested persons within 60 days of publication or other time as may
be specified, for consideration by the NRC staff and ACRS in their
review.
* * * * *
0
11. In Sec. 2.202, revise paragraphs (e)(1), (e)(5), and (6) to read
as follows:
Sec. 2.202 Orders.
* * * * *
(e)(1) If the order involves the modification of a part 50 or a
part 57 license and is a backfit, the requirements of Sec. 50.109 or
Sec. 57.16 of this chapter, as applicable, shall be followed, unless
the licensee has consented to the action required.
* * * * *
(5) If the order involves a change to a standard design approval
referenced by that plant's application, the requirements of Sec.
52.145 or Sec. 57.220 of this chapter, as applicable, must be followed
unless the applicant or licensee has consented to follow the action
required.
(6) If the order involves a modification of a manufacturing license
under subpart F of part 52 or under subpart D of part 57 of this
chapter, the requirements of Sec. 52.171 or Sec. 57.175 of this
chapter, as applicable, must be followed, unless the applicant or
licensee has consented to the action required.
0
12. In Sec. 2.309, revise paragraph (h)(2) to read as follows:
Sec. 2.309 Hearing requests, petitions to intervene, requirements for
standing, and contentions.
* * * * *
(h) * * *
(2) If the proceeding pertains to a production or utilization
facility (as defined in Sec. 50.2 or Sec. 57.3 of this chapter)
located within the boundaries of the State, local governmental body, or
Federally-recognized Indian Tribe seeking to participate as a party, no
further demonstration of standing is required. If the production or
utilization facility is not located within the boundaries of the State,
local governmental body, or Federally-recognized Indian Tribe seeking
to participate as a party, the State, local governmental body, or
Federally-recognized Indian Tribe also must demonstrate standing.
* * * * *
Sec. 2.310 [Amended]
0
13. In Sec. 2.310, add the number ``57,'' in sequential order to
paragraph (a) and paragraph (h) introductory text.
* * * * *
[[Page 23704]]
0
14. In Sec. 2.339, revise paragraph (d) to read as follows:
Sec. 2.339 Expedited decisionmaking procedure.
* * * * *
(d) The provisions of this section do not apply to an initial
decision directing the issuance of a limited work authorization under
10 CFR 50.10; an early site permit under subpart A of part 52 of this
chapter; a construction permit or construction authorization under part
50 or part 57 of this chapter; a combined license under subpart C of
part 52 of this chapter; or a manufacturing license under subpart F of
part 52 or under subpart D of part 57.
0
15. In Sec. 2.340, revise paragraphs (d), (f), and the introductory
text of paragraph (i) to read as follows:
Sec. 2.340 Initial decision in certain contested proceedings;
immediate effectiveness of initial decisions; issuance of
authorizations, permits, and licenses.
* * * * *
(d) Initial decision--manufacturing license under 10 CFR part 52 or
part 57.
(1) Matters in controversy; presiding officer consideration of
matters not put in controversy by parties. In any initial decision in a
contested proceeding on an application for a manufacturing license
under subpart F of part 52 or subpart D of part 57 of this chapter
(including an amendment to or renewal of a manufacturing license), the
presiding officer shall make findings of fact and conclusions of law on
the matters put into controversy by the parties and any matter
designated by the Commission to be decided by the presiding officer.
The presiding officer also shall make findings of fact and conclusions
of law on any matter not put into controversy by the parties, but only
to the extent that the presiding officer determines that a serious
safety, environmental, or common defense and security matter exists,
and the Commission approves of an examination of and decision on the
matter upon its referral by the presiding officer under, inter alia,
the provisions of Sec. Sec. 2.323 and 2.341.
(2) Presiding officer initial decision and issuance of permit or
license.
(i) In a contested proceeding for the initial issuance or renewal
of a manufacturing license under subpart F of part 52 or subpart D of
part 57 of this chapter, or the amendment of a manufacturing license,
the Commission or the Director, Office of Nuclear Reactor Regulation,
as appropriate, after making the requisite findings, shall issue, deny,
or appropriately condition the permit or license in accordance with the
presiding officer's initial decision once that decision becomes
effective.
(ii) In a contested proceeding for the initial issuance or renewal
of a manufacturing license under subpart F of part 52 or subpart D of
part 57 of this chapter, or the amendment of a manufacturing license,
the Commission or the Director, Office of Nuclear Reactor Regulation,
as appropriate (appropriate official), may issue the license, permit,
or license amendment in accordance with Sec. 2.1202(a) or Sec.
2.1403(a) before the presiding officer's initial decision becomes
effective. If, however, the presiding officer's initial decision
becomes effective before the license, permit, or license amendment is
issued under Sec. 2.1202 or Sec. 2.1403, then the Commission or the
Director, Office of Nuclear Reactor Regulation, as appropriate, shall
issue, deny, or appropriately condition the license, permit, or license
amendment in accordance with the presiding officer's initial decision.
* * * * *
(f) Immediate effectiveness of certain presiding officer decisions.
A presiding officer's initial decision directing the issuance or
amendment of a limited work authorization under Sec. 50.10 of this
chapter; an early site permit under subpart A of part 52 of this
chapter; a construction permit or construction authorization under part
50 or part 57 of this chapter; an operating license under part 50 or
part 57 of this chapter; a combined license under subpart C of part 52
of this chapter; a manufacturing license under subpart F of part 52 or
subpart D of part 57 of this chapter; a renewed license under part 54
or part 57 of this chapter; or a license under part 72 of this chapter
to store irradiated fuel in an independent spent fuel storage
installation (ISFSI) or a monitored retrievable storage installation
(MRS); an initial decision directing issuance of a license under part
61 of this chapter; or an initial decision under Sec. 52.103(g) of
this chapter that acceptance criteria in a combined license have been
met, is immediately effective upon issuance unless the presiding
officer finds that good cause has been shown by a party why the initial
decision should not become immediately effective.
* * * * *
(i) Issuance of authorizations, permits, and licenses--production
and utilization facilities. The Commission or the Director, Office of
Nuclear Reactor Regulation, as appropriate, shall issue a limited work
authorization under Sec. 50.10 of this chapter; an early site permit
under subpart A of part 52 of this chapter; a construction permit or
construction authorization under part 50 or part 57 of this chapter; an
operating license under part 50 or part 57 of this chapter; a combined
license under subpart C of part 52 of this chapter; or a manufacturing
license under subpart F of part 52 or subpart D of part 57 of this
chapter within 10 days from the date of issuance of the initial
decision:
* * * * *
Sec. 2.400 [Amended]
0
16. In Sec. 2.400, remove the phrase ``parts 50 or 52'' and add in its
place the phrase ``part 50 or part 52 or part 57''.
0
17. In Sec. 2.401, revise the section heading and paragraph (a) to
read as follows:
Sec. 2.401 Notice of hearing on construction permit application
pursuant to 10 CFR part 57 or appendix N of 10 CFR part 50 or combined
license application pursuant to appendix N of 10 CFR part 52.
(a) In the case of applications under appendix N of part 50 of this
chapter for construction permits for nuclear power reactors of the type
described in Sec. 50.22 of this chapter, or applications under
appendix N of part 52 of this chapter for combined licenses, or
applications under part 57 of this chapter for construction permits,
the Secretary will issue notices of hearing pursuant to Sec. 2.104.
* * * * *
0
18. In Sec. 2.402, revise paragraph (a) to read as follows:
Sec. 2.402 Separate hearings on separate issues; consolidation of
proceedings.
(a) In the case of applications under appendix N of part 50 of this
chapter for construction permits for nuclear power reactors of a type
described in 10 CFR 50.22, or applications pursuant to appendix N of
part 52 of this chapter for combined licenses, or applications under
part 57 of this chapter for construction permits and operating
licenses, the Commission or the presiding officer may order separate
hearings on particular phases of the proceeding, such as matters
related to the acceptability of the design of the reactor in the
context of the site parameters postulated for the design or
environmental matters.
* * * * *
0
19. In Sec. 2.403, revise the section heading and section to read as
follows:
Sec. 2.403 Hearings on applications for operating licenses pursuant
to 10 CFR part 57 or appendix N of 10 CFR part 50.
In the case of applications pursuant to appendix N of part 50 or
part 57 of this chapter for operating licenses for nuclear power
reactors, if the
[[Page 23705]]
Commission has not found that a hearing is in the public interest, the
Commission or the Director, Office of Nuclear Reactor Regulation, as
appropriate will, prior to acting thereon, cause to be published in the
Federal Register, pursuant to Sec. 2.105, a notice of proposed action
with respect to each application as soon as practicable after the
applications have been docketed.
0
20. Revise Sec. 2.404 to read as follows:
Sec. 2.404 Hearings on applications for operating licenses pursuant
to appendix N of 10 CFR part 50 or 10 CFR part 57.
If a request for a hearing and/or petition for leave to intervene
is filed within the time prescribed in the notice of proposed action on
an application for an operating license pursuant to appendix N of part
50 or part 57 of this chapter with respect to a specific reactor(s) at
a specific site, and the Commission, the Chief Administrative Judge, or
a presiding officer has issued a notice of hearing or other appropriate
order, then the Commission, the Chief Administrative Judge, or the
presiding officer may order separate hearings on particular phases of
the proceeding and/or consolidate for hearing two or more proceedings
in the manner described in Sec. 2.402.
0
21. Revise Sec. 2.406 to read as follows:
Sec. 2.406 Finality of decisions on separate issues.
Notwithstanding any other provision of this chapter, in a
proceeding conducted pursuant to this subpart and appendix N to part 50
or 52 or part 57 of this chapter, no matter which has been reserved for
consideration in one phase of the hearing shall be considered at
another phase of the hearing except on the basis of significant new
information that substantially affects the conclusion(s) reached at the
other phase or other good cause.
0
22. Revise Sec. 2.500 to read as follows:
Sec. 2.500 Scope of subpart.
This subpart prescribes procedures applicable to licensing
proceedings that involve the consideration in separate hearings of an
application for a license to manufacture nuclear power reactors under
subpart F of part 52 or subpart D of part 57 of this chapter.
0
23. In Sec. 2.501, revise the section heading, introductory text of
paragraph (a), and footnote 1 to read as follows:
Sec. 2.501 Notice of hearing on application under subpart F of 10 CFR
part 52 or subpart D of part 57 for a license to manufacture nuclear
power reactors.
(a) In the case of an application under subpart F of part 52 or
subpart D of part 57 of this chapter for a license to manufacture
nuclear power reactors of the type described in Sec. 50.22 or part 57
of this chapter to be operated at sites not identified in the license
application, the Secretary will issue a notice of hearing to be
published in the Federal Register at least 30 days before the date set
for hearing in the notice.\1\ The notice shall be issued as soon as
practicable after the application has been docketed. The notice will
state:
* * * * *
\1\ The thirty-day (30) requirement of this paragraph is not
applicable to a notice of the time and place of hearing published by
the presiding officer after notice of hearing described in this
section has been published.
* * * * *
Sec. 2.813 [Amended]
0
24. In Sec. 2.813, in paragraph (a), remove the phrase ``and 100'' and
add in its place the phrase ``57, and 100''.
Sec. 2.1103 [Amended]
0
25. In Sec. 2.1103, in the first sentence, remove the phrase ``of this
chapter'' and add in its place ``or 57 of this chapter''.
0
26. In Sec. 2.1202, revise paragraphs (a)(3) and (a)(6) to read as
follows:
Sec. 2.1202 Authority and role of NRC staff.
(a) * * *
(3) An application for a manufacturing license under subpart F of
10 CFR part 52 or under subpart D of 10 CFR part 57;
* * * * *
(6) Production or utilization facility licensing actions that
involve significant hazards considerations as defined in Sec. 50.92 or
subpart H of part 57 of this chapter.
* * * * *
Sec. 2.1301 [Amended]
0
27. In Sec. 2.1301, in paragraph (b), remove the phrase ``and part
52'' and add in its place ``, 52, and 57''.
Sec. 2.1403 [Amended]
0
28. In Sec. 2.1403, in paragraph (a)(3), remove ``.'' and add in its
place ``or 57.310.''.
PART 10--CRITERIA AND PROCEDURES FOR DETERMINING ELIGIBILITY FOR
ACCESS TO RESTRICTED DATA OR NATIONAL SECURITY INFORMATION OR AN
EMPLOYMENT CLEARANCE
0
29. The authority citation for part 10 continues to read as follows:
Authority: Atomic Energy Act of 1954, secs. 145, 161 (42 U.S.C.
2165, 2201); Energy Reorganization Act of 1974, sec. 201 (42 U.S.C.
5841); E.O. 10450, 18 FR 2489, 3 CFR, 1949-1953 Comp., p. 936, as
amended; E.O. 10865, 25 FR 1583, 3 CFR, 1959-1963 Comp., p. 398, as
amended; E.O. 12968, 60 FR 40245, 3 CFR, 1995 Comp., p. 391.
Sec. 10.1 [Amended]
0
30. In Sec. 10.1, in paragraph (a)(3), remove the phrase ``of this
chapter'' and add in its place the phrase ``or part 57 of this
chapter''.
Sec. 10.2 [Amended]
0
31. In Sec. 10.2, in paragraph (b), wherever it may appear, remove the
phrase ``of this chapter'' and add in its place the phrase ``or part 57
of this chapter''.
PART 11--CRITERIA AND PROCEDURES FOR DETERMINING ELIGIBILITY FOR
ACCESS TO OR CONTROL OVER SPECIAL NUCLEAR MATERIAL
0
32. The authority citation for part 11 continues to read as follows:
Authority: Atomic Energy Act of 1954, secs. 161, 223 (42 U.S.C.
2201, 2273); Energy Reorganization Act of 1974, sec. 201 (42 U.S.C.
5841); 44 U.S.C. 3504 note. Section 11.15(e) also issued under 31
U.S.C. 9701; 42 U.S.C. 2214.
Sec. 11.7 [Amended]
0
33. In Sec. 11.7, in the introductory text, add the number ``57,'' in
sequential order.
PART 19--NOTICES, INSTRUCTIONS AND REPORTS TO WORKERS: INSPECTION
AND INVESTIGATIONS
0
34. The authority citation for part 19 continues to read as follows:
Authority: Atomic Energy Act of 1954, secs. 53, 63, 81, 103,
104, 161, 223, 234, 1701 (42 U.S.C. 2073, 2093, 2111, 2133, 2134,
2201, 2273, 2282, 2297f); Energy Reorganization Act of 1974, secs.
201, 211, 401 (42 U.S.C. 5841, 5851, 5891); 44 U.S.C. 3504 note.
0
35. In Sec. 19.2, revise paragraphs (a)(1) through (3) to read as
follows:
Sec. 19.2 Scope.
(a) * * *
(1) All persons who receive, possess, use, or transfer material
licensed by the NRC under the regulations in parts 30 through 36, 39,
40, 60, 61, 63, 70, or 72 of this chapter, including persons licensed
to operate a production or utilization facility under part 50, part 52,
or part 57 of this chapter, persons licensed to possess power reactor
spent fuel in an independent spent fuel storage installation (ISFSI)
under part 72 of this chapter, and in accordance with 10 CFR 76.60 to
persons required to obtain a certificate of compliance or an approved
compliance plan under part 76 of this chapter;
(2) All applicants for and holders of licenses (including
construction permits
[[Page 23706]]
and early site permits) under parts 50, 52, 54, and 57 of this chapter;
(3) All applicants for and holders of a standard design approval
under subpart E of part 52 or under subpart E of part 57 of this
chapter; and
* * * * *
0
36. In Sec. 19.3, revise the definitions for ``License'' and
``Regulated entities'' to read as follows:
Sec. 19.3 Definitions.
* * * * *
License means a license issued under the regulations in part 30
through 36, 39, 40, 60, 61, 63, 70, or 72 of this chapter, including
licenses to manufacture, construct and/or operate a production or
utilization facility under part 50, 52, 54, or 57 of this chapter.
* * * * *
Regulated entities means any individual, person, organization, or
corporation that is subject to the regulatory jurisdiction of the NRC,
including (but not limited to) an applicant for or holder of a standard
design approval under subpart E of part 52 or under subpart E of part
57 of this chapter or a standard design certification under subpart B
of part 52 of this chapter.
* * * * *
0
37. In Sec. 19.11, revise paragraphs (a), (b), and (e)(1) to read as
follows:
Sec. 19.11 Posting of notices to workers.
(a) Each licensee (except for a holder of an early site permit
under subpart A of part 52 of this chapter, or a holder of a
manufacturing license under subpart F of part 52 or subpart D of part
57 of this chapter) shall post current copies of the following
documents:
* * * * *
(b) Each applicant for and holder of a standard design approval
under subpart E of part 52 or subpart E of part 57 of this chapter,
each applicant for an early site permit under subpart A of part 52 of
this chapter, each applicant for a standard design certification under
subpart B of part 52 of this chapter, and each applicant for and holder
of a manufacturing license under subpart F of part 52 or subpart D of
part 57 of this chapter shall post:
* * * * *
(e)(1) Each licensee, each applicant for a specific license, each
applicant for or holder of a standard design approval under subpart E
of part 52 or subpart E of part 57 of this chapter, each applicant for
an early site permit under subpart A of part 52 of this chapter, and
each applicant for a standard design certification under subpart B of
part 52 of this chapter shall prominently post NRC Form 3, ``Notice to
Employees,'' dated August 1997. Later versions of NRC Form 3 that
supersede the August 1997 version shall replace the previously posted
version within 30 days of receiving the revised NRC Form 3 from the
Commission.
* * * * *
0
38. In Sec. 19.14, revise paragraph (a) to read as follows:
Sec. 19.14 Presence of representatives of licensees and regulated
entities, and workers during inspections.
(a) Each licensee, applicant for a license, applicant for or holder
of a standard design approval under subpart E of part 52 or subpart E
of part 57 of this chapter, applicant for an early site permit under
subpart A of part 52 of this chapter, and applicant for a standard
design certification under subpart B of part 52 of this chapter shall
afford to the Commission at all reasonable times opportunity to inspect
materials, activities, facilities, premises, and records under the
regulations in this chapter.
* * * * *
Sec. 19.20 [Amended]
0
39. In Sec. 19.20, add the number ``57,'' in sequential order.
PART 20--STANDARDS FOR PROTECTION AGAINST RADIATION
0
40. The authority citation for part 20 continues to read as follows:
Authority: Atomic Energy Act of 1954, secs. 11, 53, 63, 65, 81,
103, 104, 161, 170H, 182, 186, 223, 234, 274, 1701 (42 U.S.C. 2014,
2073, 2093, 2095, 2111, 2133, 2134, 2201, 2210h, 2232, 2236, 2273,
2282, 2021, 2297f); Energy Reorganization Act of 1974, secs. 201,
202 (42 U.S.C. 5841, 5842); Low-Level Radioactive Waste Policy
Amendments Act of 1985, sec. 2 (42 U.S.C. 2021b); 44 U.S.C. 3504
note.
Sec. 20.1002 Scope.
0
41. In Sec. 20.1002, add the number ``57,'' in sequential order.
0
42. In Sec. 20.1003, revise the definition for ``License'' to read as
follows:
Sec. 20.1003 Definitions.
* * * * *
License means a license issued under the regulations in part 30
through 36, 39, 40, 50, 57, 60, 61, 63, 70, or 72 of this chapter.
* * * * *
Sec. 20.1406 [Amended]
0
43. In Sec. 20.1406, in paragraphs (a) and (b), wherever it may
appear, remove the phrase ``of this chapter'' and add in its place the
phrase ``or part 57 of this chapter''.
Sec. 20.1905 [Amended]
0
44. In Sec. 20.1905, in paragraph (g) introductory text, remove the
phrase ``of this chapter'' and add in its place the phrase ``, or part
57 of this chapter''.
0
45. In Sec. 20.2004, revise paragraph (b)(1) to read as follows:
Sec. 20.2004 Treatment or disposal by incineration.
* * * * *
(b)(1) Waste oils (petroleum derived or synthetic oils used
principally as lubricants, coolants, hydraulic or insulating fluids, or
metalworking oils) that have been radioactively contaminated in the
course of the operation or maintenance of a nuclear power reactor
licensed under part 50 or part 57 of this chapter may be incinerated on
the site where generated provided that the total radioactive effluents
from the facility, including the effluents from such incineration,
conform to the requirements of appendix I to part 50 of this chapter
and the effluent release limits contained in applicable license
conditions other than effluent limits specifically related to
incineration of waste oil. The licensee shall report any changes or
additions to the information supplied under Sec. 50.34, Sec. 50.34a,
or subpart C of part 57 of this chapter associated with this
incineration pursuant to Sec. 50.71 or 57.315 of this chapter, as
appropriate. The licensee shall also follow the procedures of Sec.
50.59 of this chapter with respect to such changes to the facility or
procedures.
* * * * *
0
46. In Sec. 20.2201, revise paragraphs (b)(2)(i) and (c) to read as
follows:
Sec. 20.2201 Reports of theft or loss of licensed material.
* * * * *
(b) * * *
(2) * * *
(i) For holders of an operating license for a nuclear power plant,
the events included in paragraph (b) of this section must be reported
under the procedures described in Sec. 50.73(b), (c), (d), (e), and
(g) or Sec. 57.440(b), (c), (d), and (e) of this chapter and must
include the information required in paragraph (b)(1) of this section,
and
* * * * *
(c) A duplicate report is not required under paragraph (b) of this
section if the licensee is also required to submit a report pursuant to
Sec. 30.55(c), Sec. 37.57, Sec. 37.81, Sec. 40.64(c), Sec. 50.72,
Sec. 50.73, subpart Q of part 57, Sec. 70.52, Sec. 73.27(b), Sec.
73.67(e)(3)(vii), Sec. 73.67(g)(3)(iii), Sec. 73.1205, or Sec.
150.19(c) of this chapter.
* * * * *
[[Page 23707]]
Sec. 20.2202 [Amended]
0
47. In Sec. 20.2202, in paragraph (d)(1), remove the phrase ``of this
chapter'' and add in its place the phrase ``or Sec. 57.435 of this
chapter''.
0
48. In Sec. 20.2203, revise paragraph (c) to read as follows:
Sec. 20.2203 Reports of exposures, radiation levels, and
concentrations of radioactive material exceeding the constraints or
limits.
* * * * *
(c) For holders of an operating license or a combined license for a
nuclear power plant, the occurrences included in paragraph (a) of this
section must be reported under the procedures described in Sec.
50.73(b), (c), (d), (e), and (g) or Sec. 57.440(b), (c), (d), and (e)
of this chapter, and must include the information required by paragraph
(b) of this section. Occurrences reported under Sec. 50.73 or Sec.
57.440(b), (c), (d), and (e) of this chapter need not be reported by a
duplicate report under paragraph (a) of this section.
* * * * *
0
49. In Sec. 20.2206, revise paragraph (a)(1) to read as follows:
Sec. 20.2206 [Amended]
* * * * *
(a) * * *
(1) Operate a nuclear reactor that is both designed to produce
electrical or heat energy and of the type described in Sec. 50.21(b)
or Sec. 50.22 of this chapter, or is a testing facility as defined in
Sec. 50.2 of this chapter; or
* * * * *
PART 21--REPORTING OF DEFECTS AND NONCOMPLIANCE
0
50. The authority citation for part 21 continues to read as follows:
Authority: Atomic Energy Act of 1954, secs. 53, 63, 81, 103,
104, 161, 223, 234, 1701 (42 U.S.C. 2073, 2093, 2111, 2133, 2134,
2201, 2273, 2282, 2297f); Energy Reorganization Act of 1974, secs.
201, 206 (42 U.S.C. 5841, 5846); Nuclear Waste Policy Act of 1982,
secs. 135, 141 (42 U.S.C. 10155, 10161); 44 U.S.C. 3504 note.
0
51. In Sec. 21.2, revise paragraphs (a)(2), (a)(4), (b), and (c) to
read as follows:
Sec. 21.2 Scope.
(a) * * *
(1) * * *
(2) Each individual, corporation, partnership, or other entity
doing business within the United States, and each director and
responsible officer of such an organization, that constructs a
production or utilization facility licensed for manufacture,
construction, or operation under part 50, part 52, or part 57 of this
chapter, an ISFSI for the storage of spent fuel licensed under part 72
of this chapter, an MRS for the storage of spent fuel or high-level
radioactive waste under part 72 of this chapter, or a geologic
repository for the disposal of high-level radioactive waste under part
60 or part 63 of this chapter; or supplies basic components for a
facility or activity licensed, other than for export, under parts 30,
40, 50, 52, 57, 60, 61, 63, 70, 71, or 72 of this chapter;
* * * * *
(4) Each individual, corporation, partnership, or other entity
doing business within the United States, and each director and
responsible officer of such an organization, applying for or holding a
standard design approval under part 52 or part 57 of this chapter; or
supplying basic components with respect to a standard design approval
under part 52 or part 57 of this chapter.
(b) For persons licensed to construct a facility under either a
construction permit issued under Sec. 50.23 or Sec. 57.95 of this
chapter or a combined license under part 52 of this chapter (for the
period of construction until the date that the Commission makes the
finding under Sec. 52.103(g) of this chapter), or to manufacture a
facility under part 52 or part 57 of this chapter, evaluation of
potential defects and failures to comply and reporting of defects and
failures to comply under Sec. 50.55(e) or Sec. 57.270 of this chapter
satisfies each person's evaluation, notification, and reporting
obligation to report defects and failures to comply under this part and
the responsibility of individual directors and responsible officers of
these licensees to report defects under Section 206 of the Energy
Reorganization Act of 1974.
(c) For persons licensed to operate a nuclear power plant under
part 50, part 52, or part 57 of this chapter, evaluation of potential
defects and appropriate reporting of defects under Sec. 50.72, Sec.
50.73, Sec. 57.270, or Sec. Sec. 73.1200 and 73.1205 of this chapter,
satisfies each person's evaluation, notification, and reporting
obligation to report defects under this part, and the responsibility of
individual directors and responsible officers of these licensees to
report defects under Section 206 of the Energy Reorganization Act of
1974.
* * * * *
0
52. In Sec. 21.3, revise the definitions for ``Commercial grade
item'', ``Critical characteristicsv'', ``Dedicating entity'',
``Defect'', and ``Substantial safety hazard'' to read as follows:
Sec. 21.3 Definitions.
* * * * *
Commercial grade item.
(1) When applied to nuclear power plants licensed under 10 CFR part
50, commercial grade item means a structure, system, or component, or
part thereof that affects its safety function, that was not designed
and manufactured as a basic component. Commercial grade items do not
include items where the design and manufacturing process require in-
process inspections and verifications to ensure that defects or
failures to comply are identified and corrected (i.e., one or more
critical characteristics of the item cannot be verified).
(2) When applied to facilities and activities licensed pursuant to
10 CFR parts 30, 40, 50 (other than nuclear power plants), 57, 60, 61,
63, 70, 71, or 72, commercial grade item means an item that is:
(i) Not subject to design or specification requirements that are
unique to those facilities or activities;
(ii) Used in applications other than those facilities or
activities; and
(iii) To be ordered from the manufacturer/supplier on the basis of
specifications set forth in the manufacturer's published product
description (for example, a catalog)
* * * * *
Critical characteristics. When applied to nuclear power plants
licensed under part 50 or part 57 of this chapter, critical
characteristics are those important design, material, and performance
characteristics of a commercial grade item that, once verified, will
provide reasonable assurance that the item will perform its intended
safety function.
Dedicating entity. When applied to nuclear power plants licensed
under part 50 or part 57 of this chapter, dedicating entity means the
organization that performs the dedication process. Dedication may be
performed by the manufacturer of the item, a third-party dedicating
entity, or the licensee itself. The dedicating entity, under Sec.
21.21(c) of this part, is responsible for identifying and evaluating
deviations, reporting defects and failures to comply for the dedicated
item, and maintaining auditable records of the dedication process.
* * * * *
Defect means:
(1) A deviation in a basic component delivered to a purchaser for
use in a facility or an activity subject to the regulations in this
part if, on the basis of an evaluation, the deviation could create a
substantial safety hazard;
(2) The installation, use, or operation of a basic component
containing a defect as defined in this section;
(3) A deviation in a portion of a facility subject to the early
site permit, standard design certification, standard
[[Page 23708]]
design approval, construction permit, combined license or manufacturing
licensing requirements of part 50, part 52, or part 57 of this chapter,
provided the deviation could, on the basis of an evaluation, create a
substantial safety hazard and the portion of the facility containing
the deviation has been offered to the purchaser for acceptance;
(4) A condition or circumstance involving a basic component that
could contribute to the exceeding of a safety limit, as defined in the
technical specifications of a license for operation issued under part
50, part 52, or part 57 of this chapter; or
(5) An error, omission or other circumstance in a design
certification, or standard design approval that, on the basis of an
evaluation, could create a substantial safety hazard.
* * * * *
Substantial safety hazard means a loss of safety function to the
extent that there is a major reduction in the degree of protection
provided to public health and safety for any facility or activity
licensed or otherwise approved or regulated by the NRC, other than for
export, under part 30, 40, 50, 52, 57, 60, 61, 63, 70, 71, or 72 of
this chapter.
* * * * *
0
53. In Sec. 21.21, revise paragraphs (a)(3), (d)(1)(i) and (ii) to
read as follows:
Sec. 21.21 Notification of failure to comply or existence of a defect
and its evaluation.
(a) * * *
(3) Ensure that a director or responsible officer subject to the
regulations of this part is informed as soon as practicable, and, in
all cases, within the 5 working days after completion of the evaluation
described in paragraphs (a)(1) or (a)(2) of this section if the
manufacture, construction, or operation of a facility or activity, a
basic component supplied for such facility or activity, the design
certification or design approval under part 52 of this chapter, or the
design approval under part 57 of this chapter--
* * * * *
(d)(1) * * *
(i) The manufacture, construction or operation of a facility or an
activity within the United States that is subject to the licensing
requirements under parts 30, 40, 50, 52, 57, 60, 61, 63, 70, 71, or 72
of this chapter and that is within his or her organization's
responsibility; or
(ii) A basic component that is within his or her organization's
responsibility and is supplied for a facility or an activity within the
United States that is subject to the licensing, design certification,
or approval requirements under parts 30, 40, 50, 52, 57, 60, 61, 63,
70, 71, or 72 of this chapter.
* * * * *
Sec. 21.51 [Amended]
0
54. In Sec. 21.51, in paragraph (a)(5), remove the phrase ``of this
chapter'' and add in its place the phrase ``or part 57 of this
chapter''.
0
55. In Sec. 21.61, revise paragraph (b) to read as follows:
Sec. 21.61 Failure to notify.
* * * * *
(b) Any NRC licensee or applicant for a license (including an
applicant for, or holder of, a permit), applicant for a design
certification under part 52 of this chapter during the pendency of its
application, applicant for a design certification after Commission
adoption of a final design certification rule for that design, or
applicant for or holder of a standard design approval under part 52 or
part 57 of this chapter subject to the regulations in this part who
fails to provide the notice required by Sec. 21.21, or otherwise fails
to comply with the applicable requirements of this part shall be
subject to a civil penalty as provided by Section 234 of the Atomic
Energy Act of 1954, as amended.
* * * * *
PART 25--ACCESS AUTHORIZATION
0
56. The authority citation for part 25 continues to read as follows:
Authority: Atomic Energy Act of 1954, secs. 145, 161, 223, 234
(42 U.S.C. 2165, 2201, 2273, 2282); Energy Reorganization Act of
1974, sec. 201 (42 U.S.C. 5841); 44 U.S.C. 3504 note; E.O. 10865, 25
FR 1583, as amended, 3 CFR, 1959-1963 Comp., p. 398; E.O. 12829, 58
FR 3479, 3 CFR, 1993 Comp., p. 570; E.O. 13526, 75 FR 707, 3 CFR,
2009 Comp., p. 298; E.O. 12968, 60 FR 40245, 3 CFR, 1995 Comp., p.
391. Section 25.17(f) and Appendix A also issued under 31 U.S.C.
9701; 42 U.S.C. 2214.
0
57. In Sec. 25.5, revise the definition for ``License'' to read as
follows:
Sec. 25.5 Definitions.
* * * * *
License means a license issued pursuant to 10 CFR parts 50, 52, 57,
60, 63, 70, or 72.
* * * * *
Sec. 25.17 [Amended]
0
58. In Sec. 25.17, in paragraph (a), add the number ``57,'' in
sequential order.
0
59. In Sec. 25.35, revise paragraph (a) to read as follows:
Sec. 25.35 Classified visits.
(a) The number of classified visits must be held to a minimum. The
licensee, certificate holder, applicant for a standard design
certification under part 52 of this chapter (including an applicant
after the Commission has adopted a final standard design certification
rule under part 52 of this chapter), or other facility, or an applicant
for or holder of a standard design approval under part 52 or part 57 of
this chapter shall determine that the visit is necessary and that the
purpose of the visit cannot be achieved without access to, or
disclosure of, classified information. All classified visits require
advance notification to, and approval of, the organization to be
visited. In urgent cases, visit information may be furnished by
telephone and confirmed in writing.
* * * * *
PART 26--FITNESS FOR DUTY PROGRAMS
0
60. The authority citation for part 26 continues to read as follows:
Authority: Atomic Energy Act of 1954, secs. 53, 103, 104, 107,
161, 223, 234, 1701 (42 U.S.C. 2073, 2133, 2134, 2137, 2201, 2273,
2282, 2297f); Energy Reorganization Act of 1974, secs. 201, 202 (42
U.S.C. 5841, 5842); 44 U.S.C. 3504 note.
0
61. In Sec. 26.3, revise paragraphs (a), (b), (c) introductory text,
and (d) and add new paragraph (f) to read as follows:
Sec. 26.3 Scope.
(a) Licensees who are authorized to operate a nuclear power reactor
under 10 CFR 50.57, and holders of a combined license under 10 CFR part
52 after the Commission has made the finding under 10 CFR 52.103(g)
shall comply with the requirements of this part, except for subparts K
and P of this part. Licensees who receive their authorization to
operate a nuclear power reactor under 10 CFR 50.57 after the date of
publication of this final rule in the Federal Register and holders of a
combined license under 10 CFR part 52 after the Commission has made the
finding under 10 CFR 52.103(g) shall implement the FFD program before
the receipt of special nuclear material in the form of fuel assemblies.
(b) Licensees who are authorized to possess, use, or transport
formula quantities of strategic special nuclear material (SSNM) under
part 70 of this chapter, and any corporation, firm, partnership,
limited liability company, association, or other organization who
obtains a certificate of compliance or an approved compliance plan
under part 76 of this chapter, only if the entity elects to engage in
activities involving formula quantities of SSNM shall comply with the
requirements of this
[[Page 23709]]
part, except for subparts I, K, and P of this part.
(c) Before the receipt of special nuclear material in the form of
fuel assemblies, the following licensees and other entities shall
comply with the requirements of this part, except for subparts I and P
of this part; and, no later than the receipt of special nuclear
material in the form of fuel assemblies, the following licensees and
other entities shall comply with the requirements of this part, except
for subpart P of this part:
* * * * *
(d) Contractor/vendors (C/Vs) who implement FFD programs or program
elements, to the extent that the licensees and other entities specified
in paragraphs (a) through (c) and (f) of this section rely on those C/V
FFD programs or program elements to meet the requirements of this part,
shall comply with the requirements of this part.
* * * * *
(f) Applicants for and holders of licenses, permits, and approvals
under part 57 of this chapter, as applicable, must implement their FFD
programs as follows:
(1) No later than the start of construction activities, licensees
and other entities that have applied for or have been issued an
operating license or construction permit under part 57 of this chapter
must implement the requirements in subpart P of this part, all the
requirements of this part except subparts K and P, or an FFD program of
their specification.
(2) Holders of a manufacturing license under part 57 of this
chapter must implement the requirements in subpart P, all the
requirements of this part except subparts K and P, or an FFD program of
their specification, before commencing activities that assemble a
manufactured reactor.
(3) Licensees and other entities that have applied for or have been
issued an operating license or construction permit under part 57 of
this chapter, and holders of a manufacturing license under part 57 of
this chapter, may elect to implement an FFD program of their
specification only if the licensee's or other entity's reactor
manufactured under a manufacturing license issued under part 57 of this
chapter, constructed under a construction permit issued under part 57
of this chapter, or operated under an operating license issued under
part 57 of this chapter, as applicable, would not require operator
action to maintain the reactor within the criterion of Sec. 57.25(a)
of this chapter or a credible operator or maintenance error could not
result in exceeding that criterion.
0
62. In Sec. 26.4, revise paragraph (a) introductory text, (a)(1) and
(4), (b), (c), (e) introductory text, (f), (g) introductory text, and
(h) introductory text to read as follows:
Sec. 26.4 FFD program applicability to categories of individuals.
(a) All persons who are granted unescorted access to nuclear power
reactor protected areas by the licensees in Sec. 26.3 (a) and, as
applicable, (c) and perform the following duties shall be subject to an
FFD program that meets all of the requirements of this part, except
subparts K and P of this part, and those persons who are granted
unescorted access to either nuclear power reactor protected areas or
remote facilities where safety-significant systems or components may be
operated within the design basis of a nuclear plant, by the licensees
and other entities in Sec. 26.3(f) and perform the following duties
must be subject to an FFD program that satisfies either the
requirements in subpart P of this part or all of the requirements of
this part except subparts K and P, unless the licensee or other entity
meets the criteria in Sec. 26.3(f)(3) and subjects these individuals
to an FFD program of its own specification:
(1) For persons who are granted unescorted access by the licensees
in Sec. 26.3(a) and, as applicable, (c), operating or onsite directing
of the operation of systems and components that a risk-informed
evaluation process has shown to be significant to public health and
safety; for those persons who are granted unescorted access by the
licensees and other entities in Sec. 26.3(f), operating or directing
of the operation of systems and components that a risk-informed
evaluation process has shown to be significant to public health and
safety;
* * * * *
(4) For persons who are granted unescorted access to nuclear power
reactor protected areas by the licensees in Sec. 26.3(a) and, as
applicable, (c), performing maintenance or onsite directing of the
maintenance of SSCs that a risk-informed evaluation process has shown
to be significant to public health and safety; for those persons who
are granted unescorted access to nuclear power reactor protected areas
by the licensees and other entities in Sec. 26.3(f), performing
maintenance or directing of the maintenance of SSCs that a risk-
informed evaluation process has shown to be significant to public
health and safety; and
* * * * *
(b) All persons who are granted unescorted access to nuclear power
reactor protected areas by the licensees in Sec. 26.3(a) and, as
applicable, (c) and who do not perform the duties described in
paragraph (a) of this section shall be subject to an FFD program that
meets all of the requirements of this part, except Sec. Sec. 26.205
through 26.209 and subparts K and P of this part. All persons who are
granted unescorted access to a facility licensed under part 57 of this
chapter, and who do not perform or direct the performance of the duties
described in Sec. 26.4(a), must be subject to either the requirements
in subpart P of this part or all the requirements of this part, except
Sec. Sec. 26.205 through 26.209 and subparts K and P, unless the
licensee or other entity meets the criteria in Sec. 26.3(f)(3) and
subjects these individuals to an FFD program of its own specification.
(c) All persons who are required by a licensee in Sec. 26.3(a)
and, as applicable, (c) to physically report to the licensee's
Technical Support Center or Emergency Operations Facility by licensee
emergency plans and procedures shall be subject to an FFD program that
meets all of the requirements of this part, except Sec. Sec. 26.205
through 26.209 and subparts K and P of this part. For licensees or
other entities in Sec. 26.3(f), all persons without unescorted access
to the facility who make decisions and/or direct actions regarding
plant safety and security, and all persons who participate remotely in
emergency response activities or physically report to the Technical
Support Center or Emergency Operations Facility (or an equivalent
facility), must be subject to an FFD program that satisfies either all
of the requirements described in subpart P of this part or all the
requirements of this part, except Sec. Sec. 26.205 through 26.209 and
subparts K and P, unless the licensee or other entity meets the
criteria in Sec. 26.3(f)(3) and subjects these individuals to an FFD
program of its own specification.
* * * * *
(e) When construction activities, as defined in Sec. 26.5, begin,
any individual whose duties for the licensees and other entities in
Sec. 26.3(c) require him or her to have the following types of access
or perform the following activities at the location where the nuclear
power plant will be constructed and operated shall be subject to an FFD
program that meets all of the requirements of this part, except
subparts I, K, and P of this part, and for any individual whose duties
for the licensees and other entities in Sec. 26.3(f) require him or
her to have the
[[Page 23710]]
following types of access, perform construction activities as defined
in Sec. 26.5, or perform the following activities must be subject to
an FFD program as described in subpart P or an FFD program that
satisfies all the requirements of this part, except subparts I, K, and
P, unless the licensee or other entity meets the criteria in Sec.
26.3(f)(3) and subjects these individuals to an FFD program of its own
specification:
* * * * *
(f) Any individual who is constructing or directing the
construction of safety- or security-related SSCs shall be subject to an
FFD program that meets the requirements of subpart K, or, if
applicable, subpart P of this part or all the requirements of this
part, except for subparts I, K, and P of this part, unless the licensee
or other entity meets the criteria in Sec. 26.3(f)(3) and subjects
these individuals to an FFD program of its own specification.
(g) All FFD program personnel who are involved in the day-to-day
operations of the program, as defined by the procedures of the
licensees and other entities in Sec. 26.3(a) through (c), and, as
applicable, (d) and whose duties require them to have the following
types of access or perform the following activities shall be subject to
an FFD program that meets all of the requirements of this part, except
subparts I, K, and P of this part, and, at the licensee's or other
entity's discretion, subpart C of this part. All personnel whose duties
require them to have the following types of access or perform the
following activities at facilities licensed under part 57 of this
chapter must be subject to the requirements in either subpart P or all
the requirements of this part, except subparts I, K, and P, and, at the
licensee's or other entity's discretion, subpart C of this part, unless
the licensee or other entity meets the criteria in Sec. 26.3(f)(3) and
subjects these individuals to an FFD program of its own specification:
* * * * *
(h) Individuals who have applied for authorization to have the
types of access or perform the activities described in paragraphs (a)
through (d) of this section shall be subject to Sec. Sec. 26.31(c)(1),
26.35(b), 26.37, 26.39, and the applicable requirements of subparts C,
E through H, and P of this part, unless the licensee or other entity
meets the criteria in Sec. 26.3(f)(3) and subjects these individuals
to an FFD program of its own specification.
* * * * *
0
63. In Sec. 26.5, add, in alphabetical order, definitions for
``Biological marker'', ``Change'', ``Consortium/third-party
administrator'', ``Illicit substance'', ``Reduction in FFD program
effectiveness'', and ``Special nuclear material''; and revise the
definitions for ``Constructing or construction activities'',
``Contractor/vendor (C/V)'', ``Other entity'', ``Reviewing official'',
``Safety-related structures, systems, and components (SSCs)'',
``Security-related SSCs'', and ``Unit outage'' to read as follows:
Sec. 26.5 Definitions.
* * * * *
Biological marker means, for a part 57 licensee implementing
subpart P of this part, an endogenous substance that is used to
validate that the biological specimen collected for testing was
produced by the donor.
* * * * *
Change as used in Sec. 26.903 (c) means an action that results in
a modification of, addition to, or removal from the licensee's or other
entity's FFD program.
* * * * *
Constructing or construction activities mean, for the purposes of
this part, the tasks involved in building a nuclear power plant that
are performed at the location where the nuclear power plant will be
constructed and operated. These tasks include fabricating, erecting,
integrating, and testing safety- and security-related SSCs, and the
installation of their foundations, including the placement of concrete.
For a licensee or other entity described in Sec. 26.3(f), construction
is defined in Sec. 57.3 of this chapter.
Consortium/third-party administrator means a contractor/vendor that
provides or coordinates one or more FFD program elements for a group of
licensees or other entities, such as administering a collective random
testing pool and random testing selections under Sec.
26.907(b)(2)(vi), that otherwise could not be independently implemented
by those licensees or other entities. A consortium/third-party
administrator also could provide access to, for example, the services
of medical review officers, substance abuse experts, employee
assistance programs, and HHS-certified laboratories under contract to
perform drug testing.
Contractor/vendor (C/V) means any company, or any individual not
employed by a licensee or other entity specified in Sec. 26.3(a)
through (c) and (f), who is providing work or services to a licensee or
other entity covered in Sec. 26.3(a) through (c) and (f), either by
contract, purchase order, oral agreement, or other arrangement.
* * * * *
Illicit substance means a substance that causes impairment and
possible addiction but is not an illegal drug as defined in this
section.
* * * * *
Other entity means any corporation, firm, partnership, limited
liability company, association, C/V, or other organization who is
subject to this part under Sec. 26.3(a) through (c) and (f) but is not
licensed by the NRC.
* * * * *
Reduction in FFD program effectiveness means, for a part 57
licensee or other entity implementing subpart P of this part, a change
or series of changes to an element of the FFD program that reduces or
eliminates the licensee's ability to satisfy or maintain site-specific
FFD program performance when compared to historical site-specific
performance, the licensee's fleet-level program performance, or
industry performance.
* * * * *
Reviewing official means an employee of a licensee or other entity
specified in Sec. 26.3(a) through (c) and (f), who is designated by
the licensee or other entity to be responsible for reviewing and
evaluating any potentially disqualifying FFD information about an
individual, including, but not limited to, the results of a
determination of fitness, as defined in Sec. 26.189, in order to
determine whether the individual may be granted or maintain
authorization.
Safety-related structures, systems, and components (SSCs) means,
for part 50 or part 52 licensees and other entities described in Sec.
26.3(a) through (d), those SSCs that are relied on to remain functional
during and following design basis events to ensure the integrity of the
reactor coolant pressure boundary, the capability to shut down the
reactor and maintain it in a safe shutdown condition, or the capability
to prevent or mitigate the consequences of accidents that could result
in potential offsite exposure comparable to the guidelines in Sec.
50.34(a)(1) of this chapter. For part 57 licensees and other entities
described in Sec. 26.3(d) and (f), safety-related has the same meaning
as that in Sec. 57.3 of this chapter.
Security-related SSCs means, for the purposes of this part, those
structures, systems, and components that the licensee will rely on to
implement the licensee's physical security and safeguards contingency
plans that either are required under part 73 of this chapter if the
licensee is a construction
[[Page 23711]]
permit applicant or holder or an early site permit holder, as described
in Sec. 26.3(c)(3) through (c)(5), respectively, or are included in
the licensee's application if the licensee is a combined license
applicant or holder, as described in Sec. 26.3(c)(1) and (c)(2),
respectively, or a licensee or other entity described in Sec. 26.3(d)
or (f).
* * * * *
Special nuclear material (SNM) has the same meaning as that in
Sec. 70.4 of this chapter.
* * * * *
Unit outage means, for the purposes of this part, for electricity-
generation units, that the reactor unit is disconnected from the
electrical grid. Unit outage means, for the purposes of this part, for
non-electricity-generation units, that the reactor unit is disconnected
from the loads to which its output is supplied under normal operating
conditions.
* * * * *
0
64. In Sec. 26.8, revise paragraph (b) to read as follows:
Sec. 26.8 Information collection requirements: OMB approval.
* * * * *
(b) The approved information collection requirements contained in
this part appear in Sec. Sec. 26.9, 26.27, 26.29, 26.31, 26.33, 26.35,
26.37, 26.39, 26.41, 26.53, 26.55, 26.57, 26.59, 26.61, 26.63, 26.65,
26.67, 26.69, 26.75, 26.77, 26.85, 26.87, 26.89, 26.91, 26.93, 26.95,
26.97, 26.99, 26.101, 26.103, 26.107, 26.109, 26.111, 26.113, 26.115,
26.117, 26.119, 26.125, 26.127, 26.129, 26.135, 26.137, 26.139, 26.153,
26.157, 26.159, 26.163, 26.165, 26.167, 26.168, 26.169, 26.183, 26.185,
26.187, 26.189, 26.202, 26.203, 26.205, 26.207, 26.211, 26.401, 26.403,
26.405, 26.406, 26.407, 26.411, 26.413, 26.415, 26.417, 26.711, 26.713,
26.715, 26.717, 26.719, 26.821, 26.903, 26.904, 26.906, 26.907, 26.908,
26.909, 26.911, 26.913, 26.917, and 26.919.
0
65. Revise Sec. 26.21 to read as follows:
Sec. 26.21 Fitness-for-duty program.
The licensees and other entities specified in Sec. 26.3(a) through
(c) and (f) (for those licensees and other entities that do not
implement the requirements in subparts P and K of this part, and do not
implement an FFD program of their own specification if they meet the
criteria in Sec. 26.3(f)(3) shall establish, implement, and maintain
FFD programs that, at a minimum, comprise the program elements
contained in this subpart. The individuals specified in Sec. 26.4(a)
through (e) and (g), and, at the licensee's or other entity's
discretion, Sec. 26.4(f), and, if necessary, Sec. 26.4(j) shall be
subject to these FFD programs. Licensees and other entities may rely on
the FFD program or program elements of a C/V, as defined in Sec. 26.5,
if the C/V's FFD program or program elements satisfy the applicable
requirements of this part.
0
66. In Sec. 26.35, revise paragraph (c)(3) to read as follows:
Sec. 26.35 Employee assistance programs.
* * * * *
(c) * * *
(3) If a licensee or other entity receives a report from EAP
personnel under paragraph (c)(2) of this section, the licensee or other
entity must ensure that the requirements of Sec. Sec. 26.69(d) and
26.77(b), or the procedures and actions required by Sec.
26.906(b)(2)(vii) are implemented, as applicable.
0
67. Revise Sec. 26.51 to read as follows:
Sec. 26.51 Applicability.
The requirements in this subpart apply to the licensees and other
entities identified in Sec. 26.3(a), (b), and, as applicable, (c) for
the categories of individuals in Sec. 26.4(a) through (d), and, at the
licensee's or other entity's discretion, in Sec. 26.4(g) and, if
necessary, Sec. 26.4(j). The requirements in this subpart also apply
to the licensees and other entities specified in Sec. 26.3(c), as
applicable, for the categories of individuals in Sec. 26.4(e). At the
discretion of a licensee or other entity in Sec. 26.3(c), the
requirements of this subpart also may be applied to the categories of
individuals identified in Sec. 26.4(f). In addition, the requirements
in this subpart apply to the entities in Sec. 26.3(d) to the extent
that a licensee or other entity relies on the C/V to satisfy the
requirements of this subpart. Certain requirements in this subpart also
apply to the individuals specified in Sec. 26.4(h). The requirements
in this subpart apply to the FFD programs of licensees and other
entities identified in Sec. 26.3(f) that elect not to implement the
requirements in subpart P for the categories of individuals in Sec.
26.4 and do not implement an FFD program of their own specification if
they meet the criteria in Sec. 26.3(f)(3).
0
68. In Sec. 26.53, revise paragraph (e) introductory text, paragraph
(g), and introductory text of paragraphs (h) and (i) to read as
follows:
Sec. 26.53 General provisions.
* * * * *
(e) Licensees and other entities in Sec. 26.3(a) through (c) and
(f) may also rely on a C/V's FFD program or program elements when
granting or maintaining the authorization of an individual who is or
has been subject to the C/V's FFD program, if the C/V's program or
program elements meet the applicable requirements of this part.
* * * * *
(g) The licensees and other entities specified in Sec. 26.3(a)
and, as applicable, (c), (d), and (f), shall identify any violation of
any requirement of this part to any licensee who has relied on or
intends to rely on the FFD program element that is determined to be in
violation of this part.
(h) The licensees and other entities specified in Sec. 26.3(a)
and, as applicable, (c), (d), and (f), may not initiate any actions
under this subpart without the knowledge and written consent of the
subject individual. The individual may withdraw his or her consent at
any time. If an individual withdraws his or her consent, the licensee
or other entity may not initiate any elements of the authorization
process specified in this subpart that were not in progress at the time
the individual withdrew his or her consent, but shall complete and
document any elements that are in progress at the time consent is
withdrawn. The licensee or other entity shall record the individual's
application for authorization; his or her withdrawal of consent; the
reason given by the individual for the withdrawal, if any; and any
pertinent information gathered from the elements that were completed
(e.g., the results of pre-access drug tests, information obtained from
the suitable inquiry). The licensee or other entity to whom the
individual has applied for authorization shall inform the individual
that--
* * * * *
(i) The licensees and other entities specified in Sec. 26.3(a)
and, as applicable, (c), (d), and (f), shall inform, in writing, any
individual who is applying for authorization that the following actions
related to providing and sharing the personal information required
under this subpart are sufficient cause for denial or unfavorable
termination of authorization:
* * * * *
0
69. In Sec. 26.63, revise paragraph (d) to read as follows:
Sec. 26.63 Suitable inquiry.
* * * * *
(d) When any licensee or other entity in Sec. 26.3(a) through (d)
and (f) is legitimately seeking the information required for an
authorization decision under this subpart and has obtained a signed
release from the subject individual authorizing the disclosure of
information, any licensee or other entity subject to this part shall
disclose whether the subject individual's authorization was denied or
terminated
[[Page 23712]]
unfavorably as a result of a violation of an FFD policy and shall make
available the information on which the denial or unfavorable
termination of authorization was based, including, but not limited to,
drug or alcohol test results, treatment and follow-up testing
requirements or other results from a determination of fitness, and any
other information that is relevant to an authorization decision.
* * * * *
0
70. Revise Sec. 26.73 to read as follows:
Sec. 26.73 Applicability.
The requirements in this subpart apply to the licensees and other
entities identified in Sec. 26.3(a), (b), and, as applicable, (c) for
the categories of individuals specified in Sec. 26.4(a) through (d)
and (g). The requirements in this subpart also apply to the licensees
and other entities specified in Sec. 26.3(c), as applicable, for the
categories of individuals in Sec. 26.4(e). At the discretion of a
licensee or other entity in Sec. 26.3(c), the requirements of this
subpart also may be applied to the categories of individuals identified
in Sec. 26.4(f). In addition, the requirements in this subpart apply
to the entities in Sec. 26.3(d) to the extent that a licensee or other
entity relies on the C/V to satisfy the requirements of this subpart.
The regulations in this subpart also apply to the individuals specified
in Sec. 26.4(h) and (j), as appropriate. The requirements in this
subpart apply to the FFD programs of licensees and other entities
identified in Sec. 26.3(f) that elect not to implement the
requirements in subpart P for the categories of individuals in Sec.
26.4 and do not implement an FFD program of their own specification if
they meet the criteria in Sec. 26.3(f)(3).
0
71. Revise Sec. 26.81 to read as follows:
Sec. 26.81 Purpose and applicability.
This subpart contains requirements for collecting specimens for
drug testing and conducting alcohol tests by or on behalf of the
licensees and other entities in Sec. 26.3(a) through (d) for the
categories of individuals specified in Sec. 26.4(a) through (d) and
(g). At the discretion of a licensee or other entity in Sec. 26.3(c),
specimen collections and alcohol tests must be conducted either under
this subpart for the individuals specified in Sec. 26.4(e) and (f) or
the licensee or other entity may rely on specimen collections and
alcohol tests conducted under the requirements of 49 CFR part 40 for
the individuals specified in Sec. 26.4(e) and (f). The requirements of
this subpart do not apply to specimen collections and alcohol tests
that are conducted under the requirements of 49 CFR part 40, as
permitted in this paragraph and under Sec. Sec. 26.4(j) and
26.31(b)(2) and subpart K. The requirements in this subpart apply to
the FFD programs of licensees and other entities identified in Sec.
26.3(f) that elect not to implement the requirements in subpart P for
the categories of individuals in Sec. 26.4 and do not implement an FFD
program of their own specification if they meet the criteria in Sec.
26.3(f)(3).
0
72. In Sec. 26.97, revise paragraphs (a) and (b) to read as follows:
Sec. 26.97 Collecting oral fluid specimens for alcohol and drug
testing.
(a) The collector, with the assistance of a virtual collection
monitor as permitted under Sec. 26.907(g)(2) of this chapter, if
applicable, shall perform the oral fluid specimen collection consistent
with the device manufacturer's instructions. At a minimum, the
collector shall--
(1) Check the expiration date on the device and show it to the
donor (the device cannot be used after its expiration date).
(2) Explain the collection process to the donor, including any
actions the donor must perform during the collection process, and that
a failure to cooperate with the specimen collection process will be
considered a refusal to test and sanctions for subverting the testing
process will be imposed.
(3) Instruct the donor to wash and dry their hands before providing
a specimen. If a sink is not available in the area where the collection
is to be conducted, another equivalent method to clean the donor's
hands must be provided (e.g., provide the donor with single use
examination gloves to wear during the collection process).
(4) Ensure that the donor's mouth is free of any items that could
impede or interfere with the collection of an oral fluid specimen, such
as food or tobacco.
(5) Open in the presence of the donor, or direct the donor to open
an individually wrapped or sealed package containing the device.
(6) Instruct the donor to insert the device into their mouth to
gather oral fluids in the manner described in the device manufacturer's
instructions.
(7) When the device is ready to be removed from the donor's mouth,
follow the device manufacturer's instructions to complete the
collection process.
(b) If all steps in paragraph (a) of this section could not be
completed successfully (e.g., the device breaks, the device is dropped
on the floor, the device fails to activate), the collector, with the
assistance of a virtual collection monitor as permitted under Sec.
26.907(g)(2), if applicable, shall--
(1) Discard the oral fluid specimen device;
(2) Document the reason(s) that a new specimen collection is
required, or the reasons that a donor has been determined to have
refused the test; and
(3) If a new specimen collection is required, collect a new
specimen under paragraph (a) of this section.
* * * * *
0
73. Revise Sec. 26.201 to read as follows:
Sec. 26.201 Applicability.
(a) The requirements in this subpart, with the exception of Sec.
26.202, apply to the licensees and other entities identified in Sec.
26.3(a); if applicable, (c), (d), and (f), for licensees and other
entities not implementing the requirements in subparts K and P and that
do not implement an FFD program of their own specification if they meet
the criteria in Sec. 26.3(f)(3). For the licensees and other entities
to whom the requirements in this subpart, with the exception of Sec.
26.202, apply, the requirements in Sec. Sec. 26.203 and 26.211 apply
to the individuals identified in Sec. 26.4(a) through (c). In
addition, the requirements in Sec. 26.205 through Sec. 26.209 apply
to the individuals identified in Sec. 26.4(a).
(b) The requirements in this subpart, with the exception of Sec.
26.203, apply to the licensees or other entities identified in Sec.
26.3(f) implementing this subpart under Sec. 26.904. For these
licensees and other entities, the requirements in Sec. Sec. 26.202 and
26.211 apply to the individuals identified in Sec. 26.4(a) through (c)
and any person licensed to operate under 10 CFR part 57; and the
requirements in Sec. Sec. 26.205 through 26.209 apply to the
individuals identified in Sec. 26.4(a).
0
74. Add Sec. 26.202 to read as follows:
Sec. 26.202 General provisions for facilities licensed under part 57.
(a) Policy. Licensees must establish a policy for the management of
fatigue for all individuals who are subject to the licensee's FFD
program and incorporate it into the written policy required in Sec.
26.906(a).
(b) Procedures. In addition to the procedures required in Sec.
26.906(b), licensees must develop, implement, and maintain procedures
that--
(1) Describe the process to be followed when any individual
identified in Sec. 26.4(a) through (c) makes a self-declaration that
the individual is not fit to safely and competently perform his or her
duties for any part of a working tour as a result of fatigue. The
procedure must--
[[Page 23713]]
(i) Describe the individual's and licensee's rights and
responsibilities related to self-declaration;
(ii) Describe requirements for establishing controls and conditions
under which an individual may be permitted or required to perform work
after that individual declares that he or she is not fit due to
fatigue; and
(iii) Describe the process to be followed if the individual
disagrees with the results of a fatigue assessment that is required
under Sec. 26.211(a)(2);
(2) Describe the process for implementing the controls required
under Sec. 26.205 for the individuals who are performing the duties
listed in Sec. 26.4(a);
(3) Describe the process to be followed in conducting fatigue
assessments under Sec. 26.211; and
(4) Describe the disciplinary actions that the licensee may impose
on an individual following a fatigue assessment, and the conditions and
considerations for taking those disciplinary actions.
(c) Training and assessments. Licensees must include the following
knowledge and abilities in the content of the training and trainee
assessments required in Sec. 26.908:
(1) Knowledge of the contributors to worker fatigue, circadian
variations in alertness and performance, indications and risk factors
for common sleep disorders, shiftwork strategies for obtaining adequate
rest, and the effective use of fatigue countermeasures; and
(2) Ability to identify symptoms of worker fatigue and contributors
to decreased alertness in the workplace.
(d) Recordkeeping. Licensees must retain the following records for
at least 3 years or until the completion of all related legal
proceedings, whichever is later:
(1) Records of work hours for individuals who are subject to the
work hour controls in Sec. 26.205;
(2) For licensees implementing the requirements of Sec.
26.205(d)(3), records of shift schedules and shift cycles, or, for
licensees implementing the requirements of Sec. 26.205(d)(7), records
of shift schedules and records showing the beginning and end times and
dates of all averaging periods, of individuals who are subject to the
work hour controls in Sec. 26.205;
(3) The documentation of waivers that is required in Sec.
26.207(a)(4), including the bases for granting the waivers;
(4) The documentation of work hour reviews that is required in
Sec. 26.205(e)(3) and (e)(4); and
(5) The documentation of fatigue assessments that is required in
Sec. 26.211(g).
(e) Reporting. Licensees must include the following information in
a standard format in the annual FFD program performance report required
under Sec. 26.917(b)(2):
(1) A summary for each nuclear power plant site of all instances
during the previous calendar year when the licensee waived one or more
of the work hour controls specified in Sec. 26.205(d)(1) through
(d)(5)(i) and (d)(7) for individuals described in Sec. 26.4(a). The
summary must include only those waivers under which work was performed.
If it was necessary to waive more than one work hour control during any
single extended work period, the summary of instances must include each
of the work hour controls that were waived during the period. For each
category of individuals specified in Sec. 26.4(a), the licensee must
report--
(i) The number of instances when each applicable work hour control
specified in Sec. 26.205(d)(1)(i) through (iii), (d)(2)(i) and (ii),
(d)(3)(i) through (v), and (d)(7) was waived for individuals not
working on outage activities;
(ii) The number of instances when each applicable work hour control
specified in Sec. 26.205(d)(1)(i) through (iii), (d)(2)(i) and (ii),
(d)(3)(i) through (v), (d)(4) and (d)(5)(i), and (d)(7) was waived for
individuals working on outage activities; and
(iii) A summary that shows the distribution of waiver use among the
individuals applicable within each category of individuals identified
in Sec. 26.4(a) (e.g., a table that shows the number of individuals
who received only one waiver during the reporting period, the number of
individuals who received a total of two waivers during the reporting
period).
(2) A summary of corrective actions, if any, resulting from the
analyses of these data, including fatigue assessments.
(f) Audits. Licensees must audit the management of worker fatigue
under Sec. 26.915.
0
75. In Sec. 26.205, revise paragraphs (d)(7)(iii) and (d)(8) to read
as follows:
Sec. 26.205 Work hours.
* * * * *
(d) * * *
(7) * * *
(iii) Each licensee shall state, in its FFD policy and procedures
required by either Sec. 26.27 and Sec. 26.203(a) and (b) or Sec.
26.202(a) and (b) and Sec. 26.906, the work hour counting system in
Sec. 26.205(d)(7)(ii) the licensee is using.
(8) Each licensee shall state, in its FFD policy and procedures
required by either Sec. 26.27 and Sec. 26.203(a) and (b) or Sec.
26.202(a) and (b) and Sec. 26.906, the requirements with which the
licensee is complying: the minimum days off requirements in Sec.
26.205(d)(3) or maximum average work hours requirements in Sec.
26.205(d)(7).
* * * * *
0
76. In Sec. 26.207, revise paragraph (a)(1)(ii) to read as follows:
Sec. 26.207 Waivers and exceptions.
(a) * * *
(1) * * *
(ii) A supervisor assesses the individual face to face and
determines that there is reasonable assurance that the individual will
be able to safely and competently perform his or her duties during the
additional work period for which the waiver will be granted. The
supervisor performing the assessment shall be trained as required by
either Sec. 26.29 and Sec. 26.203(c) or Sec. 26.202(c) and Sec.
26.908 and shall be qualified to direct the work to be performed by the
individual. If there is no supervisor on site who is qualified to
direct the work, the assessment may be performed by a supervisor who is
qualified to provide oversight of the work to be performed by the
individual. At a minimum, the assessment must address the potential for
acute and cumulative fatigue considering the individual's work history
for at least the past 14 days, the potential for circadian degradations
in alertness and performance considering the time of day for which the
waiver will be granted, the potential for fatigue-related degradations
in alertness and performance to affect risk-significant functions, and
whether any controls and conditions must be established under which the
individual will be permitted to perform work. For licensees and other
entities in Sec. 26.3(f), the assessment may be performed remotely
using electronic communications. In such instances, the assessment must
be supported by someone who is present in-person with the individual
whose alertness may be impaired, and that supporting person must be
trained under the requirements of either Sec. 26.29 and Sec.
26.203(c) or Sec. 26.202(c) and Sec. 26.908.
* * * * *
0
77. In Sec. 26.211, revise paragraphs (a)(1) and (3) and the
introductory text of paragraph (b) to read as follows:
Sec. 26.211 Fatigue assessments.
(a) * * *
(1) For-cause. In addition to any other test or determination of
fitness that may be required under Sec. Sec. 26.31(c), 26.77,
26.907(b), and 26.919, a fatigue
[[Page 23714]]
assessment must be conducted in response to an observed condition of
impaired individual alertness creating a reasonable suspicion that an
individual is not fit to safely and competently perform his or her
duties, except if the condition is observed during an individual's
break period. If the observed condition is impaired alertness with no
other behaviors or physical conditions creating a reasonable suspicion
of possible substance abuse, then the licensee need only conduct a
fatigue assessment. If the licensee has reason to believe that the
observed condition is not due to fatigue, the licensee need not conduct
a fatigue assessment;
* * * * *
(3) Post-event. A fatigue assessment must be conducted in response
to events requiring post-event drug and alcohol testing as specified in
Sec. 26.31(c) or post-event tests in Sec. 26.907(b)(4). Licensees may
not delay necessary medical treatment in order to conduct a fatigue
assessment; and
* * * * *
(b) Only supervisors and FFD program personnel who are trained
under either Sec. Sec. 26.29 and 26.203(c) or Sec. Sec. 26.202(c) and
26.908 may conduct a fatigue assessment. The fatigue assessment must be
conducted face to face with the individual whose alertness may be
impaired. For licensees and other entities in Sec. 26.3(f), a fatigue
assessment may be performed remotely using electronic communications.
In such instances, the fatigue assessment must be supported by someone
who is present in-person with the individual whose alertness may be
impaired, and that supporting person must be trained in accordance with
the requirements of either Sec. Sec. 26.29 and 26.203(c) or Sec. Sec.
26.202(c) and 26.908.
* * * * *
0
78. Revise Sec. 26.709 to read as follows:
Sec. 26.709 Applicability.
(a) The requirements of this subpart apply to the FFD programs of
licensees and other entities specified in Sec. 26.3(a) through (d),
except for FFD programs that are implemented under subpart K of this
part.
(b) The requirements in this subpart apply to the FFD programs of
licensees and other entities specified in Sec. 26.3(f) that elect not
to implement the requirements in subpart P and do not implement an FFD
program of their own specification if they meet the criteria in Sec.
26.3(f)(3).
0
79. In Sec. 26.711, revise paragraphs (c) and (d) to read as follows:
Sec. 26.711 General provisions.
* * * * *
(c) The licensees and other entities specified in Sec. 26.3(a)
and, as applicable, (c), (d), and (f), shall inform each individual of
his or her right to review information about the individual that is
collected and maintained under this part to assure its accuracy.
Licensees and other entities shall provide the individual with an
opportunity to correct any inaccurate or incomplete information that is
documented by licensees and other entities about the individual.
(d) Licensees and other entities shall ensure that only correct and
complete information about individuals is retained and shared with
other licensees and entities. If, for any reason, the shared
information used for determining an individual's eligibility for
authorization under this part changes or new information is developed
about the individual, licensees and other entities shall correct or
augment the shared information contained in the records. If the changed
or developed information has implications for adversely affecting an
individual's eligibility for authorization, a licensee and other entity
specified in Sec. 26.3(a) and, as applicable, (c), (d), and (f), who
has discovered the incorrect information, or develops new information,
shall inform the reviewing official of any FFD program under which the
individual is maintaining authorization of the updated information on
the day of discovery. The reviewing official shall evaluate the
information and take appropriate actions, which may include denial or
unfavorable termination of the individual's authorization.
0
80. Add Subpart P, consisting of Sec. Sec. 26.901 through 26.919, to
read as follows:
Subpart P--Fitness-for-Duty Programs for Facilities Licensed Under
10 CFR Part 57
Sec. 26.901 Applicability.
Sec. 26.903 General provisions.
Sec. 26.904 FFD program requirements.
Sec. 26.906 Written policy and procedures.
Sec. 26.907 Drug and alcohol testing.
Sec. 26.908 FFD program training.
Sec. 26.909 Behavioral observation.
Sec. 26.910 Sanctions.
Sec. 26.911 Protection of information.
Sec. 26.913 Appeals process.
Sec. 26.915 Audits.
Sec. 26.917 Recordkeeping, reporting, and FFD program performance.
Sec. 26.919 Suitability and fitness determinations.
Sec. 26.901 Applicability.
A licensee or other entity in Sec. 26.3(f) that elects to
implement the requirements of this subpart must establish, implement,
and maintain a fitness-for-duty (FFD) program that satisfies the
requirements of this subpart for those categories of individuals in
Sec. 26.4, as applicable, and any person licensed to operate under 10
CFR part 57.
Sec. 26.903 General provisions.
(a) FFD program description. An applicant's description of the FFD
program in its final safety analysis report, required by subparts C and
D of 10 CFR part 57, must include--
(1) A discussion of the applicability of the FFD program to those
individuals described in Sec. 26.4 and how the program will be
implemented at a facility authorized to assemble or perform non-
operational testing of a manufactured reactor under a manufacturing
license issued under part 57 of this chapter, if applicable; and
(2) A description of the drug and alcohol testing and fitness
determination process to be implemented through the licensee's or other
entity's procedures, including the collection and testing facilities to
be used, biological specimens to be collected and tested, and sanctions
to be imposed for FFD policy violations.
(b) FFD program implementation and availability. For the licensees
and other entities implementing the requirements of this subpart, the
FFD program must be implemented as stated in Sec. 26.904(a). For the
holder of an operating license under part 57 of this chapter, the FFD
program must be maintained until the NRC's docketing of the license
holder's certifications described in Sec. 57.305 of this chapter. For
the holder of a manufacturing license under part 57 of this chapter,
the FFD program must be maintained until expiration of the
manufacturing license.
(c) FFD program change control.
(1) The licensee or other entity may make changes to its FFD
program under this subpart if--
(i) The licensee or other entity performs and retains an analysis
demonstrating that the changes do not reduce the effectiveness of the
FFD program; or
(ii) The change was necessitated or justified by a change to part
26, laboratory processes or procedures, or guidance issued by the HHS
or NRC, as implemented by the licensee or other entity though its
procedures.
(2) A licensee or other entity desiring to make a change that
decreases FFD program effectiveness must implement a mitigating
strategy so the FFD program, as revised, will continue to satisfy the
performance objectives in Sec. 26.23 and
[[Page 23715]]
will not result in a reduction in FFD program effectiveness.
(3) Notwithstanding Sec. 26.903(c)(1)(ii), the change control
process may not be used to reduce the minimum panel of drugs to be
tested in Sec. 26.907(c)(1).
(4) The licensee must retain a record of each change made under
this section for a period of at least 5 years from the date the change
was implemented and summarize this change in its annual FFD performance
report required by Sec. 26.917(b)(2) or Sec. 26.717, as applicable.
Sec. 26.904 FFD program requirements.
(a) The licensee or other entity must establish, implement, and
maintain an FFD program under this subpart before the start of--
(1) for a holder of a manufacturing license, activities authorized
by the manufacturing license;
(2) for a holder of a construction permit, construction activities
as defined in Sec. 26.5;
(3) for the holder of an operating license--
(i) operational testing of a manufactured reactor at a
manufacturing facility; and
(ii) the earliest occurrence of the following at the operating
site, as applicable:
(A) the loading of fuel into a reactor vessel;
(B) the receipt of a fueled manufactured reactor; and
(C) individuals subject to part 26 operate, test, perform
maintenance of, or direct the maintenance or surveillance of security-
related equipment or equipment that a risk-informed evaluation process
has shown to be significant to public health and safety; and
(4) for a general licensee under Sec. 57.45(d), construction
activities as defined in Sec. 26.5.
(b) The FFD program required by this subpart must:
(1) Apply to those individuals described in Sec. 26.4, as
applicable; and
(2) Implement the following requirements and subparts:
(i) Section 26.23, Performance objectives;
(ii) Section 26.35, Employee assistance programs;
(iii) Section 26.903, General provisions;
(iv) Section 26.906, Written policies and procedures;
(v) Section 26.907, Drug and alcohol testing;
(vi) Section 26.908, FFD program training;
(vii) Section 26.909, Behavioral observation;
(viii) Section 26.910, Sanctions;
(ix) Section 26.911, Protection of information;
(x) Section 26.913, Appeals process;
(xi) Section 26.915, Audits;
(xii) Section 26.917, Recordkeeping, reporting, and FFD program
performance;
(xiii) Section 26.919, Suitability and fitness determinations;
(xiv) Subpart A--Administrative Provisions;
(xv) Subpart I--Managing Fatigue; and
(xvi) Subpart O--Inspections, Violations, and Penalties.
Sec. 26.906 Written policy and procedures.
(a) Licensees and other entities that implement an FFD program
under this subpart must ensure that--
(1) A written FFD policy statement is provided to each individual
who is subject to the program before the individual is subject to drug
and alcohol testing.
(2) The FFD policy statement describes the performance objectives
in Sec. 26.23.
(3) The FFD policy statement describes the minimum days off
requirements in Sec. 26.205(d)(3) or maximum average work hours
requirements in Sec. 26.205(d)(7).
(4) The FFD policy statement must be written in sufficient detail
to provide affected individuals with information on what is expected of
them and what consequences may result from a lack of adherence to the
policy, including those elements described in Sec. 26.906(b), part 26-
required sanctions, and required medical/clinical treatment and follow-
up testing for FFD policy violations.
(5) The FFD policy statement describes the individual's
responsibilities to report for work in a physiological and
psychological condition that enables the safe and competent performance
of assigned duties and responsibilities and inform a licensee- or other
entity-designated representative when the individual determines that
this cannot be accomplished.
(6) The FFD policy statement must prohibit the consumption of
alcohol, at a minimum, within an abstinence period of 5 hours preceding
the individual's arrival at the licensee's or other entity's facility.
(7) The FFD policy statement must convey that abstinence from
alcohol for the 5 hours preceding any scheduled tour of duty is
considered to be a minimum that is necessary, but may not be
sufficient, to ensure that the individual is fit for duty.
(b) Licensees and other entities must establish, implement, and
maintain written procedures that address the following topics:
(1) For the drug and alcohol testing program under this subpart,
(i) The methods and techniques to collect and test for drugs and
alcohol and for the shipping and temporary storage of biological
specimens used for drug testing at HHS-certified laboratories,
(ii) The urine specimen volumes, techniques for split specimen
collections, and the acceptability of a urine specimen as described in
Sec. 26.111 or as described in the HHS Guidelines,
(iii) Protecting the privacy of an individual who provides a
specimen, protecting the integrity of the specimen, and ensuring that
the test results are valid and attributable to the correct individual,
and
(iv) If the licensee or other entity elects to use the HHS
Guidelines, the name of the specific HHS Guideline and revision being
implemented by the licensee or other entity and a description of the
specific sections in the guideline that are being implemented in the
procedure, including specimen collections, drug testing, and evaluation
of test results.
(2) The immediate and follow-up actions that will be taken, and the
procedures to be used, in those cases in which individuals who are
subject to the FFD program:
(i) Have been involved in the use, sale, or possession of illegal
substances, illegal drugs, or illicit substances;
(ii) Are impaired by any illegal substances, illegal drugs, or
illicit substances or the consumption of alcohol as determined by
behavioral observation or a test that measures blood alcohol
concentration;
(iii) Attempted to subvert the testing process by adulterating or
diluting specimens (in vivo or in vitro), substituting specimens, or by
any other means;
(iv) Refused to provide a specimen for analysis or follow
instructions provided by FFD program personnel;
(v) Had legal action taken relating to drug or alcohol use;
(vi) Demonstrated character or actions indicating that the
individual cannot be trusted or relied upon to perform those duties and
responsibilities or maintain access to NRC-licensed facilities, special
nuclear material (SNM), or sensitive information; or
(vii) Have a condition or have taken actions that pose or have
posed an immediate hazard to themselves or others, as notified by EAP
personnel under Sec. 26.35(c)(2).
(3) The process, including the duties and responsibilities of FFD
program
[[Page 23716]]
personnel, to be followed if an individual's behavior or condition
raises a concern regarding the possible use, sale, or possession of
illegal drugs on- or offsite; the possible use or possession of alcohol
on the NRC-licensed facility; impairment from any cause that in any way
could adversely affect the individual's ability to safely and
competently perform the individual's duties; or the receipt of credible
information indicating that the individual cannot be trusted or relied
on to perform those duties and responsibilities making the individual
subject to this part.
(4) Operation and oversight of any onsite or offsite collection
facility.
(5) The fatigue management requirements in Sec. Sec. 26.202(b) and
either 26.205(d)(3) or (d)(7).
(6) Measures to prevent subversion of drug and alcohol tests
conducted onsite and offsite.
Sec. 26.907 Drug and alcohol testing.
Licensees and other entities must perform drug and alcohol testing
that complies with the following requirements--
(a) Split specimens. Split specimen collections of oral fluid or
urine must be used for the test conditions described in paragraph (b)
of this section. Testing of the split specimen (specimen B) requires
the donor's permission unless ordered by the MRO to resolve an invalid
test result obtained for specimen A.
(b) Test conditions. Individuals identified in Sec. 26.4 must be
subject to drug and alcohol testing under the following conditions:
(1) Pre-access. A pre-access test must be conducted for drugs and
alcohol before performing or directing the conduct of roles and
responsibilities making the individual subject to this subpart or being
granted unescorted access to the protected areas of the NRC-licensed
facility. A pre-access test must have been conducted no more than 14
days before the individual is granted unescorted access.
(2) Random. Random testing for drugs and alcohol must--
(i) Be administered in a manner that provides reasonable assurance
that individuals are unable to predict the time periods during which
specimens will be collected;
(ii) Require individuals who are selected for random testing to
report to the onsite collection site as soon as reasonably practicable
after notification, within the time period specified in the FFD program
procedure;
(iii) Ensure that all individuals in the population that is subject
to random testing on a given day have an equal probability of being
selected and tested;
(iv) Ensure that an individual completing a test is immediately
eligible for another random test; and
(v) Ensure that the sampling process used to select individuals for
random testing provides that the number of random tests performed
annually is equal to at least 50 percent of the population that is
subject to the FFD program at the NRC-licensed site.
(vi) If the number of individuals subject to random testing at an
NRC-licensed site is such that paragraph (b)(2)(v) of this section
cannot be implemented without predictable outcomes, the licensee must
use a consortium/third-party administrator to manage the random testing
pool and make selections for testing throughout the year. In such
instances, the consortium/third-party administrator must ensure that
the testing rate for the random testing pool from which they sample
meets the requirement in paragraph (b)(2)(v).
(3) For-cause. For-cause drug and alcohol tests must be conducted
onsite in response to an individual's observed behavior or physical
condition indicating possible substance abuse, as defined in Sec.
26.5. A for-cause drug test, alcohol test, or both, must be conducted
onsite after receiving credible information either that an individual
is engaging in substance abuse or in response to a portal area
screening test result under paragraph (i) of this section.
(4) Post-event. A post-event test for drugs and alcohol must be
conducted--
(i) As soon as practical after an event involving a human error
that was committed by an individual specified in Sec. 26.4, where the
human error may have caused or contributed to the event. This test must
be conducted onsite unless the individual requires offsite medical
care. The licensee or other entity must test the individual(s) who
committed or directed the error and need not test individuals who were
affected by the event and whose actions likely did not cause or
contribute to the event. The licensee or other entity must describe in
its procedures what constitutes a human error.
(ii) Within 4 hours of an event unless immediate medical
intervention precludes the conduct of the test on the individual(s) who
caused or contributed to the accident(s), if the event results in--
(A) An illness or personal injury to any individual which results
in death, days away from work, restricted work, transfer to another
job, medical treatment beyond first aid, loss of consciousness, or
other significant illness or injury, as diagnosed by a licensee- or
other entity-designated physician or other licensed health care
professional, even if the illness or injury does not result in death,
days away from work, restricted work or job transfer, medical treatment
beyond first aid, or loss of consciousness; or
(B) Damage to any safety- or security-related structures, systems,
and components; and
(5) Follow-up. An individual subject to part 26 who has violated
the FFD policy for substance use or abuse, or the sale, use, or
possession of illegal drugs must be subject to a follow-up series of
tests for drugs, alcohol, or both to verify an individual's continued
abstinence from substance abuse.
(c) Urine and oral fluid specimens.
(1) All urine or oral fluid specimens must be tested for the
substances listed in Sec. 26.31(d)(1), except as allowed by Sec.
26.903(c)(3). All urine specimens must be subject to validity testing
as specified in either this part or the HHS Guidelines. All oral fluid
specimens may be subject to validity testing, including a biological
marker, as specified in either this part or the HHS Guidelines.
(2) For the use of urine as the biological specimen to be tested,
the following requirements must be implemented--
(i) Section 26.115, Collecting a urine specimen under direct
observation;
(ii) Section 26.119, Determining ``shy'' bladder; and
(iii) Section 26.163, Cutoff levels for drugs and drug metabolites.
(3) For alcohol testing onsite, the following requirements must be
implemented--
(i) Section 26.91, Acceptable devices for conducting initial and
confirmatory tests for alcohol and methods of use;
(ii) Section 26.93, Preparing for alcohol testing;
(iii) Section 26.95, Conducting an initial test for alcohol using a
breath specimen;
(iv) Section 26.97, Collecting oral fluid specimens for alcohol and
drug testing;
(v) Section 26.99, Determining the need for a confirmatory test for
alcohol;
(vi) Section 26.101, Conducting a confirmatory test for alcohol;
and,
(vii) Section 26.103, Determining a confirmed positive test result
for alcohol.
(4) For all test conditions in Sec. 26.907(b), MRO-directed tests
under Sec. 26.185, and the testing of a split specimen, drug testing
must be performed at an HHS-certified laboratory for the specific
biological
[[Page 23717]]
specimen to be tested. Only HHS-certified laboratory test results from
urine and oral fluid specimens may be used for the issuance of a part
26-required sanction.
(5) The licensee or other entity must establish and maintain a
contract with an HHS-certified laboratory for each specimen to be
tested. Each contract must stipulate the following:
(i) The laboratory must comply with the applicable provisions of
any State licensor requirements;
(ii) Laboratory records and documents must be provided and/or able
to be photocopied and removed from the premises to support an
inspection or audit;
(iii) The laboratory must make available qualified personnel to
testify in an administrative or disciplinary proceeding against an
individual when that proceeding is based on test results reported by
the HHS-certified laboratory;
(iv) The laboratory must maintain test records in confidence,
consistent with the requirements of Sec. 26.37, and use them with the
highest regard for individual privacy;
(v) Consistent with the principles established in section 503 of
Public Law 100-71, any employee of a licensee or other entity who is
the subject of a drug test (or his or her representative designated
under Sec. 26.37(d)) must, on written request, have access to the
laboratory's records related to his or her validity and drug test and
any records related to the results of any relevant certification,
review, or revocation-of-certification proceedings;
(vi) The laboratory may not enter into any relationship with the
licensee's or other entity's MRO(s) that may be construed as a
potential conflict of interest, including, but not limited to, the
relationships described in Sec. 26.183(b), and may not derive any
financial benefit by having a licensee or other entity use a specific
MRO; and
(vii) The laboratory must permit representatives of the NRC and any
licensee or other entity using the laboratory's services to inspect the
laboratory at any time, including unannounced inspections.
(d) Privacy and integrity. The specimen collection and drug and
alcohol testing procedures of FFD programs must protect the donor's
privacy and the integrity of the specimen and implement quality
controls to ensure that test results are valid and attributable to the
correct individual.
(e) Offsite collection facilities. At the licensee's or other
entity's discretion, except for those specimens that must be collected
onsite under Sec. 26.907(b)(3) and (4), specimen collections and
alcohol testing may be conducted at a local hospital or other facility
licensed to conduct specimen collections and perform alcohol testing
and audited by the State or a State-designated entity. The licensee or
other entity must audit these facilities, if used, before their initial
use and then on a biennial basis to confirm that the facility
procedures are comparable to those described in subpart E of this part
or the HHS Guidelines for urine and oral fluid.
(f) Initial testing. A licensee or other entity subject to this
subpart performing an initial test must use an immunoassay, or an
alternative technology as specified in the HHS Guidelines for the
specific biological specimen that is to be tested. Specimens that yield
positive, positive and dilute, adulterated, substituted, or invalid
initial validity or drug test results or discrepant biological markers
must be subject to confirmatory testing by an HHS-certified laboratory,
certified for that biological specimen, except for invalid specimens
that cannot be tested.
(g) Oral fluid testing.
(1) If the licensee or other entity elects to use oral fluid for
drug or alcohol testing, the collection, packaging, temporary storage
and shipment of an oral fluid specimen to an HHS-certified laboratory
for drug testing, or the collection of an oral fluid specimen for
alcohol testing must be performed in accordance with licensee- or other
entity-established procedures based either on the requirements in this
part or the procedures in HHS Guidelines identified by the licensee or
other entity in Sec. 26.906(b)(1)(iv). The oral fluid device must not
expire before the date of the collection of the specimen for testing.
The drugs, drug metabolites, initial and confirmatory testing cutoffs,
and biological markers, if applicable, must be those established by the
HHS Guidelines for oral fluid testing and the alcohol cutoffs in this
part or, if not established by the HHS Guidelines or this part for the
panel of drugs and drug metabolites to be tested, as determined and
documented by a forensic toxicologist review conducted pursuant to
Sec. 26.31(d)(1)(i)(D).
(2) The virtual collection of oral fluid specimens for drug and
alcohol testing is only permitted for sites that must use a consortium/
third-party administrator to implement random testing under Sec.
26.907(b)(2)(vi). For a licensee or other entity to utilize a virtual
oral fluid specimen collection process, the following must apply or
should be considered, as applicable:
(i) The specimen collector completing the virtual collection must
meet the requirements in 10 CFR 26.85, ``Collector qualifications and
responsibilities.''
(ii) The oral fluid specimen collection process must be completed
as described in Sec. 26.97, ``Collecting oral fluid specimens for
alcohol and drug testing,'' and Sec. 26.99, ``Determining the need for
a confirmatory test for alcohol.''
(iii) An individual other than the donor (i.e., a virtual
collection monitor) may be needed in the location where the specimen
collection is to be performed to assist the virtual collector in
completing activities, performing observations, or both.
(iv) If a virtual collection monitor is used to assist the specimen
collector in completing an oral fluid specimen collection, then the
virtual specimen collector must explain the collection process to the
monitor and provide instruction to the monitor on required activities
to be performed during the collection process. The monitor's name must
be recorded on the Federal CCF for drug testing specimens, or an
analogous document for alcohol testing.
(v) Video teleconference communication method(s) must provide
sufficient visual and aural clarity to complete the process and ensure
that a donor is not able to subvert the testing process.
(vi) Collection kit materials must be maintained in a secure
fashion until the virtual collector initiates the virtual collection
process with the donor.
(vii) The licensee or other entity's written FFD procedures must
describe in detail the virtual collection process and when and how it
is to be implemented.
(viii) The virtual collection procedure must address problem
collections, such as the video teleconference becomes inoperable during
the collection process or the donor is unable to provide an oral fluid
specimen of sufficient quantity to complete the specimen collection
process for drug or alcohol testing.
(ix) The virtual collection procedure must include steps to collect
a breath specimen using an EBT if the oral fluid specimen test result
under Sec. 26.99(b) requires a confirmatory testing for alcohol under
Sec. 26.101. At a minimum, a donor with an oral fluid specimen test
result requiring confirmatory testing for alcohol must be removed from
duty pending additional testing.
(h) Hair testing. The testing of hair specimens may only be used to
inform a licensee's or other entity's determination of whether the
individual is trustworthy and reliable under the test condition in
Sec. 26.907(b)(1) to
[[Page 23718]]
supplement the information gained from a pre-access test using oral
fluid or urine as the test specimen and must be conducted at an HHS-
certified laboratory certified to test hair specimens.
(1) If used, this process must be described in the licensee's or
other entity's FFD policy and described in detail in its procedure. The
panel of drugs and drug metabolites to be evaluated must only include
those listed as Schedule I or II of section 202 of the Controlled
Substances Act [21 U.S.C. 812]. The collection, packaging, and
temporary storage of a hair specimen and shipment of the specimen to an
HHS-certified laboratory must be conducted in accordance with the HHS
Guidelines. The licensee- or other entity-designated FFD program
personnel must conduct the collection, packaging, temporary storage,
shipping, and custody and control of the specimen.
(2) Before the licensee or other entity begins to conduct hair
testing, the initial and confirmatory testing cutoffs must be the
cutoffs established by the HHS Guidelines for hair testing or, if not
established by the HHS Guidelines or this part, as determined by a
forensic toxicologist review conducted pursuant to Sec.
26.31(d)(1)(i)(D).
(3) Confirmed positive test results must be considered potentially
disqualifying FFD information until proven otherwise by a review under
Sec. 26.913. Sanctions under this subpart must not be issued for any
FFD policy violation involving a drug test using a hair specimen unless
the licensee or other entity determines that the individual has
attempted to subvert the testing process, as defined in Sec. 26.5, for
the hair test.
(i) Portal area screening. A non-invasive testing instrument may be
used to screen individuals for drugs, drug metabolites, and alcohol
before the individuals' entry into or exit from a protected or vital
area.
(1) The instrument must be operated in accordance with the
manufacturer's specifications. If screening detects the presence of any
drug, drug metabolite, or alcohol at or above the instrument set
point(s), the individual screened by the instrument must be subject to
for-cause testing under Sec. 26.907(b)(3).
(2) Annually, the licensee or other entity must verify the accuracy
of the portal area screening test for each substance with any positive
results. If at least 85 percent of the positive portal area screening
test results for a substance in the past 12 months do not subsequently
confirm positive on for-cause testing performed under paragraph (i)(1)
of this section, the licensee or other entity cannot continue to use
the screening test for the particular substance until such time as
corrective actions have been implemented to improve the testing
accuracy.
(3) A part 26 sanction may not be issued to an individual based
solely on a portal area screening instrument detection that drugs or
alcohol exceed the instrument's established setpoint.
(j) Blood testing. The testing of blood specimens may only be
conducted under the order of the licensee- or other entity-designated
MRO for a valid medical reason as confirmed by the MRO pursuant to
Sec. 26.31(d)(5). This specimen must be subject to testing by a
laboratory that satisfies quality control requirements that are
comparable to those required for certification by the HHS.
(k) Federal custody and control form. For the collection and
packaging of urine, oral fluid, and hair specimens for drug testing,
the licensee or other entity must use a Federal CCF.
(l) Medical Review Officer. Licensees or other entities must--
(1) Require their designated MRO to review positive, positive and
dilute, adulterated, substituted, and invalid confirmatory drug and
validity test results to determine whether the donor has violated the
FFD policy. The review must be completed before reporting the results
to the individual designated by the licensee or other entity to assess
authorization or perform the suitability and fitness determinations
required under Sec. 26.919, or, if required, that are described in
subpart H of this part.
(2) Require their MRO to satisfy the requirements in Sec. 26.183
and, prior to conducting any activities under this part, attend and
pass a medical- or clinical-based training session to improve his/her
knowledge of MRO duties and responsibilities, drug and alcohol testing
processes and procedures, and evaluation of drug testing results. This
training session must be conducted by a nationally recognized MRO
training and certification organization that has been assessed by the
licensee's or other entity's FFD program personnel to include the
technical elements an MRO must implement under Sec. 26.185. An MRO who
performed the duties and responsibilities in Sec. Sec. 26.185 and
26.187 for at least 3 continuous years in the last 10 years prior to
being hired or contracted by the licensee or other entity satisfies the
requirements in this paragraph (l)(2).
(3) Require their MRO to attend a medical- or clinical-based
training session at least every 5 years to improve his/her knowledge of
changes in drug and alcohol testing processes and procedures and
evaluation of drug testing results.
(4) Require their MRO to determine whether a biological specimen is
positive, positive and dilute, adulterated, substituted, or invalid by
implementing the requirements in Sec. 26.185 or the HHS Guidelines
through the licensee's or other entity's procedures.
(i) If Sec. 26.185 or the HHS Guidelines, as used by the licensee
or other entity in its procedures, are insufficient to make this
determination, then guidance issued by a State agency in the State in
which the NRC-licensed facility is located, Federal agencies, or
nationally recognized MRO training and certification organizations may
be used to inform an MRO determination.
(ii) An MRO need not review alcohol test results, including
positive confirmatory alcohol test results determined by an EBT under
Sec. 26.907(c)(3)(vi) and (vii).
(5) Require their MRO to determine and approve the use of oral
fluid or urine as an alternative biological specimen when the donor
cannot provide a specimen for testing. This determination and the
retest must be documented and completed as soon as reasonably
practicable.
(6) Require the MRO to review all specimen test results associated
with drug-related FFD policy violations. This review includes split
specimens and all specimens taken to resolve a discrepant condition,
such as a possible subversion attempt, impairment without a known
cause, or a donor-requested or MRO-directed retest. To resolve a
discrepant condition, the MRO is authorized to test a specimen for a
biological marker, adulterants, or additional drugs.
(m) Limitations of screening and testing. Specimens collected under
NRC regulations may only be designated or approved for screening and
testing as described in this part and may not be used to conduct any
other analysis or test without the written permission of the donor.
Analyses, screens, and tests that may not be conducted include, but are
not limited to, DNA testing, serological typing, or any other medical
or genetic test used for diagnostic or specimen identification
purposes. No biological specimens may be passively sampled and analyzed
in a manner different than described in this subpart.
(n) Specimen collectors. All onsite specimen collections, except a
collection by a portal area screening instrument in Sec. 26.907(i),
must be
[[Page 23719]]
conducted by licensee- or other entity-designated and -trained
personnel.
Sec. 26.908 FFD program training.
(a) FFD program training.
(1) Individuals must be trained in the FFD policy and procedure,
including fatigue management, and their FFD program responsibilities.
Individuals who collect specimens for testing must also be trained in
specimen collector duties and responsibilities, including, at a
minimum, specimen collection, custody and control, identification and
response to subversion attempts, and privacy. For licensees and other
entities of nuclear plants, the FFD program training program must use a
systems approach to training as described in Sec. 57.390 of this
chapter for those individuals in Sec. 26.4.
(2) FFD program training must include training on the behavioral
observation program. The behavioral observation program training must
include the detection of physiological behaviors or conditions that may
indicate--
(i) Possible use, sale, or possession of illegal drugs or illicit
drugs, or substance abuse on- or offsite;
(ii) Use or possession of alcohol onsite or use while on duty
offsite;
(iii) Impairment from fatigue or any cause that, if left
unattended, could result in inattentiveness or human errors; and
(iv) Any individual's inability to safely and competently perform
assigned duties and responsibilities or act in a trustworthy and
reliable manner while having access to protected areas, SNM, or
sensitive information.
(3) Training must explain that an individual's FFD policy violation
will--
(i) Subject the individual to an FFD program-required sanction
designed to preclude recurrence of an FFD policy violation;
(ii) Contribute to the licensee's or other entity's assessment of
whether the individual can be trusted and relied upon to safely and
competently perform the assigned duties and responsibilities making the
individual subject to this subpart;
(iii) Be used to inform the licensee's or other entity's insider
mitigation program under Sec. 57.325 of this chapter and access
authorization program under Sec. 73.56 of this chapter; and
(iv) Be used to inform other NRC licensees and other entities
subject to part 26 when FFD program information is requested to support
authorization determinations under subpart C of part 26 or Sec. 73.56
of this chapter.
(b) Training and assessments. Training and a trainee assessment
must be conducted before pre-access testing, and FFD program refresher
training and trainee assessments must be conducted on a nominal 24-
month frequency, or more frequently where the need is indicated.
Indications of the need for more frequent training include, but are not
limited to, an individual's failure to properly implement FFD program
procedures and the frequency, nature, or severity of problems
discovered through audits or the administration of the program.
(c) Training program review. The licensee or other entity must
periodically evaluate its FFD training program and revise it as
appropriate to reflect industry experience as well as applicable
changes to the regulations in this part, the HHS Guidelines, if used,
and specimen collection and testing processes implemented by the
licensee or other entity.
Sec. 26.909 Behavioral observation.
(a) Licensees and other entities must ensure that the individuals
who are subject to this subpart are subject to behavioral observation
and that behavioral observation is performed by all individuals subject
to this subpart.
(b) Licensees and other entities must require all individuals
subject to the FFD program to report to the licensee- or other entity-
designated representative any onsite or offsite behaviors or activities
by individuals subject to this part that may constitute an unreasonable
risk to the safety or security of the NRC-licensed facility or SNM or
may cause harm to others. This reporting must include any information
relating to character or reputation of the individual indicating that
the individual cannot be trusted or relied upon to perform those duties
and responsibilities or maintain access to NRC-licensed facilities,
SNM, or sensitive information that makes them subject to part 26.
(c) Behavioral observation must be performed visually, in-person,
and, when necessary, remotely by live video and audible streaming and
capture, to observe the behavior of individuals in the workforce
subject to the requirements in this subpart.
(d) Not withstanding Sec. 26.909(c), for a reactor facility where
individual task loading does not allow for the effective conduct of
behavior observation in addition to assigned operational tasks, the
licensee or other entity must implement a live video and audible
streaming and capture system to conduct behavioral observation of
persons licensed to operate under 10 CFR part 57 who manipulate the
controls of any nuclear plant licensed under 10 CFR part 57.
Sec. 26.910 Sanctions.
(a) Licensees and other entities that implement an FFD program
under this subpart must establish sanctions for FFD policy violations
that, at a minimum, prohibit the individuals specified in Sec. 26.4
from being assigned to perform or direct those duties and
responsibilities or maintaining authorization making them subject to
this subpart.
(b) The severity of the sanction must escalate with the number of
occurrences and severity of the FFD policy violation. The sanction must
be long enough to act as a deterrent and, if the individual is retained
as a licensee employee or contractor/vendor, facilitate the individual
to complete counseling or treatment. The sanctions must include an
immediate unfavorable termination of the individual's authorization as
follows:
(1) A minimum 14-day denial of access for a first violation of the
FFD policy involving a confirmed positive drug or alcohol test result;
(2) A minimum 3-year denial of access for a second violation of the
FFD policy involving a confirmed positive drug or alcohol test result;
(3) A minimum 5-year denial of access for any individual who is
determined to have been involved in the sale, use, or possession of
illegal drugs or the consumption of alcohol within a protected area of
any facility licensed under part 57 of this chapter or within a
transporter's facility or vehicle used in the conveyance of formula
quantities of strategic SNM while the individual is subject to this
subpart; and
(4) A permanent denial of access for a third violation of the FFD
policy involving a confirmed positive drug or alcohol test result or a
subversion attempt of any drug or alcohol test or screening process.
Sec. 26.911 Protection of information.
(a) Licensees and other entities that collect personal information
about an individual for the purpose of complying with this subpart must
establish and maintain a system of files and procedures to prevent
unauthorized disclosure.
(b) Licensees and other entities must obtain a signed consent that
documents the individual's acceptance of being subject to the FFD
program and authorizes the disclosure of the personal information
collected and maintained under this subpart, except for disclosures to
the individuals and entities specified in Sec. 26.37(b)(1)
[[Page 23720]]
through (b)(6), (b)(8), and persons deciding matters under review in
Sec. 26.913. This signed and dated consent must be obtained before
making the individual subject to the FFD program.
Sec. 26.913 Appeals process.
Licensees and other entities that implement an FFD program under
this subpart must establish and implement procedures for the review of
a determination that an individual in Sec. 26.4 has violated the FFD
policy. The procedure must provide for an objective and impartial
review of the facts related to the determination that the individual
has violated the FFD policy and a schedule for the completion of the
review.
Sec. 26.915 Audits.
(a) Licensees and other entities that implement an FFD program
under this subpart must audit their programs at a frequency that
ensures the continuing effectiveness of their FFD program, FFD program
elements that are provided by C/Vs, and the FFD programs of C/Vs that
are accepted by the licensee or other entity. Corrective actions must
be taken as soon as reasonably practicable to resolve any problems
identified in an audit and preclude recurrence.
(b) The subject matter, scope, and frequency of audits must be
revised as necessary to improve or maintain program performance based
on annual FFD program performance data reviews performed under Sec.
26.917(d) and unsatisfactory performance or programmatic weaknesses
identified under Sec. 26.917(b)(3) and (e).
(c) Licensees and other entities may conduct joint audits or accept
audits of C/Vs so long as the audit addresses the relevant services of
the C/Vs.
(d) Licensees and other entities must audit HHS-certified
laboratories unless the licensee's or other entity's panel of drugs and
drug metabolites to be tested is equivalent to the panel by which the
laboratory is certified by HHS or is subject to the standards and
procedures for drug testing and evaluation used by the laboratory under
the HHS Guidelines. Licensees and other entities must audit any
hospital or other facility licensed by the State (or State-designated
entity) if used to conduct specimen collections and perform alcohol
testing under this part on a biennial basis to confirm that the
facility procedures are comparable to those described in subpart E of
this part, for urine and oral fluid.
Sec. 26.917 Recordkeeping, reporting, and FFD program performance.
(a) Licensees and other entities that implement FFD programs under
this subpart must ensure that records pertaining to the administration
of their program, which may be stored and archived electronically, are
maintained so that they are available for NRC inspection purposes and
for any legal proceedings resulting from the administration of the
program. Records pertaining to the administration of the FFD program
and FFD performance data required by Sec. 26.717 must be retained
until license termination.
(b) Licensees and other entities must make the following reports:
(1) Reports to the NRC Headquarters Operations Center by telephone
within 24 hours after the licensee or other entity discovers any
intentional act that casts doubt on the integrity of the FFD program
and any programmatic failure, degradation, or discovered vulnerability
of the FFD program that may permit undetected drug or alcohol use or
abuse by individuals who are subject to this subpart. These events must
be reported under this subpart, rather than under the provisions of
Sec. 73.1200 of this chapter;
(2) Annual FFD program performance data under Sec. 26.717(b) for
each FFD program subject to this subpart. Licensees and other entities
must submit FFD program performance data (for January through December)
to the NRC annually, before March 1 of the following year and must use
unexpired NRC-provided forms for the electronic submission of FFD
information to the NRC; and
(3) Reports on drug and alcohol testing errors within 30 days of
completing an investigation of any testing errors or unsatisfactory
performance discovered at an HHS-certified laboratory or through the
processing of appeals under Sec. 26.913, or errors or matters that
could adversely reflect on the integrity of the random selection or
random testing process. The reports must describe the incident and any
corrective actions taken or planned.
(c) Licensees and other entities subject to this subpart must
describe in sufficient detail to support an authorization
determination, an individual's FFD policy violation (while protecting
privacy information under Sec. 26.911) and FFD program weakness to the
NRC, licensees, and other entities subject to part 26 when requested to
support authorization determinations under subpart C of this part or to
support licensee or other entity performance monitoring.
(d) Licensees and other entities must analyze FFD program
performance data at least annually and take appropriate actions to
correct any identified program weakness.
(e) Licensees and other entities must document, trend, and correct
non-reportable indicators of FFD programmatic weaknesses under the
licensee's or other entity's corrective action program, but may not
track or trend drug and alcohol test results in a manner that would
permit the identification of any individuals.
Sec. 26.919 Suitability and fitness determinations.
Licensees and other entities that implement FFD programs under this
subpart must develop, implement, and maintain procedures for evaluating
whether to assign individuals to perform or direct those duties and
responsibilities making them subject to this subpart. A suitability or
fitness determination conducted for cause must be performed face to
face. A suitability or fitness determination conducted for cause may be
performed remotely using electronic communications that provide
sufficient visual and aural clarity to complete the assessment. A
fitness determination may be supported by someone who is present in-
person with the individual being assessed only during for-cause drug
and alcohol testing determinations under Sec. 26.907(b)(3) and fatigue
assessments performed under Sec. 26.211(a)(1). The supporting person
must be trained in accordance with the requirements of either Sec.
26.29 or Sec. 26.908.
PART 30--RULES OF GENERAL APPLICABILITY TO DOMESTIC LICENSING OF
BYPRODUCT MATERIAL
0
81. The authority citation for part 30 continues to read as follows:
Authority: Atomic Energy Act of 1954, secs. 11, 81, 161, 181,
182, 183, 184, 186, 187, 223, 234, 274 (42 U.S.C. 2014, 2111, 2201,
2231, 2232, 2233, 2234, 2236, 2237, 2273, 2282, 2021); Energy
Reorganization Act of 1974, secs. 201, 202, 206, 211 (42 U.S.C.
5841, 5842, 5846, 5851); 44 U.S.C. 3504 note.
0
82. In Sec. 30.4, revise the definition for ``Utilization facility''
to read as follows:
Sec. 30.4 Definitions.
* * * * *
Utilization facility means a utilization facility as defined in the
regulations contained in part 50 or part 57 of this chapter;
83. In Sec. 30.50, revise paragraph (c)(3) to read as follows:
Sec. 30.50 Reporting requirements.
* * * * *
(c) * * *
(3) The provisions of Sec. 30.50 do not apply to licensees subject
to the notification requirements in Sec. 50.72 or
[[Page 23721]]
Sec. 57.435 of this chapter. They do apply to those part 50 or part 57
licensees possessing material licensed under part 30, who are not
subject to the notification requirements in Sec. 50.72 or Sec. 57.435
of this chapter, respectively.
PART 40--DOMESTIC LICENSING OF SOURCE MATERIAL
0
84. The authority citation for part 40 continues to read as follows:
Authority: Atomic Energy Act of 1954, secs. 62, 63, 64, 65, 69,
81, 83, 84, 122, 161, 181, 182, 183, 184, 186, 187, 193, 223, 234,
274, 275 (42 U.S.C. 2092, 2093, 2094, 2095, 2099, 2111, 2113, 2114,
2152, 2201, 2231, 2232, 2233, 2234, 2236, 2237, 2243, 2273, 2282,
2021, 2022); Energy Reorganization Act of 1974, secs. 201, 202, 206,
211 (42 U.S.C. 5841, 5842, 5846, 5851); Uranium Mill Tailings
Radiation Control Act of 1978, sec. 104 (42 U.S.C. 7914); 44 U.S.C.
3504 note.
0
85. In Sec. 40.60, revise paragraph (c)(3) to read as follows:
Sec. 40.60 Reporting requirements.
* * * * *
(c) * * *
(3) The provisions of Sec. 40.60 do not apply to licensees subject
to the notification requirements in Sec. 50.72 or Sec. 57.435, of
this chapter. They do apply to those part 50 or part 57 licensees
possessing material licensed under part 40 of this chapter who are not
subject to the notification requirements in Sec. 50.72 or Sec. 57.435
of this chapter, respectively.
PART 50--DOMESTIC LICENSING OF UTILIZATION AND PRODUCTION
FACILITIES
0
86. The authority citation for part 50 is revised to read as follows:
Authority: Atomic Energy Act of 1954, secs. 11, 101, 102, 103,
104, 105, 108, 122, 147, 149, 161, 181, 182, 183, 184, 185, 186,
187, 189, 223, 234 (42 U.S.C. 2014, 2131, 2132, 2133, 2134, 2135,
2138, 2152, 2167, 2169, 2201, 2231, 2232, 2233, 2234, 2235, 2236,
2237, 2239, 2273, 2282); Energy Reorganization Act of 1974, secs.
201, 202, 206, 211 (42 U.S.C. 5841, 5842, 5846, 5851); Nuclear Waste
Policy Act of 1982, sec. 306 (42 U.S.C. 10226); National
Environmental Policy Act of 1969 (42 U.S.C. 4332); 44 U.S.C. 3504
note.
0
87. In Sec. 50.44, revise the introductory texts of paragraphs (c) and
(d) to read as follows:
Sec. 50.44 Combustible gas control for nuclear power reactors.
* * * * *
(c) Requirements for future water-cooled reactor applicants and
licensees.\1\ * * *
* * * * *
(d) Requirements for future non-water-cooled reactor applicants and
licensees and certain water-cooled reactor applicants and licensees.
The requirements in this paragraph apply to all construction permits
and operating licenses under this part, and to all design approvals,
design certifications, combined licenses, or manufacturing licenses
under part 52 or construction permits, operating licenses,
manufacturing licenses, or standard design approvals under part 57 of
this chapter, for non-water-cooled reactors and water-cooled reactors
that do not fall within the description in paragraph (c), footnote 1 of
this section, any of which are issued after October 16, 2003.
Applications subject to this paragraph must include:
* * * * *
\[1]\ The requirements of this paragraph apply only to water-cooled
reactor designs with characteristics (e.g., type and quantity of
cladding materials) such that the potential for production of
combustible gases is comparable to light water reactor designs licensed
as of October 16, 2003.
* * * * *
0
88. In Sec. 50.59, revise paragraphs (b), (c)(3), and (d)(2) to read
as follows:
Sec. 50.59 Changes, tests, and experiments.
* * * * *
(b) This section applies to each holder of an operating license
issued under this part, or a combined license issued under part 52 of
this chapter, or a manufacturing license, construction permit, or
operating license issued under part 57 of this chapter, including the
holder of a license authorizing the operation of a nuclear power
reactor that has submitted the certification of permanent cessation of
operations required under Sec. 50.82(a)(1) or Sec. 52.110 or 57.305
of this chapter, a reactor licensee whose license has been amended to
allow possession of nuclear fuel but not operation of the facility, or
a non-power production or utilization facility that has permanently
ceased operations.
* * * * *
(c) * * *
(3) In implementing this paragraph, the FSAR (as updated) is
considered to include FSAR changes resulting from evaluations performed
pursuant to this section and analyses performed pursuant to Sec. 50.90
or Sec. 57.312 of this chapter since submittal of the last update of
the final safety analysis report pursuant to Sec. 50.71 or Sec.
57.315 of this chapter.
* * * * *
(d) * * *
(2) The licensee shall submit, as specified in Sec. 50.4 or Sec.
52.3 or Sec. 57.4 of this chapter, as applicable, a report containing
a brief description of any changes, tests, and experiments, including a
summary of the evaluation of each. A report must be submitted at
intervals not to exceed 24 months. For combined licenses, the report
must be submitted at intervals not to exceed 6 months during the period
from the date of application for a combined license to the date the
Commission makes its findings under 10 CFR 52.103(g).
* * * * *
0
89. In Sec. 50.68, revise paragraph (a), to read as follows:
Sec. 50.68 Criticality accident requirements.
(a) Each holder of a construction permit or operating license for a
nuclear power reactor issued under this part or part 57 of this
chapter, or a combined license for a nuclear power reactor issued under
part 52 of this chapter, shall comply with either 10 CFR 70.24 of this
chapter or the requirements in paragraph (b) of this section.
* * * * *
PART 51--ENVIRONMENTAL PROTECTION REGULATIONS FOR DOMESTIC
LICENSING AND RELATED REGULATORY FUNCTIONS
0
90. In Sec. 51.4, revise the definition for ``Construction'' to read
as follows:
Sec. 51.451.4 Definitions
* * * * *
Construction means:
(1) For production and utilization facilities licensed under 10 CFR
part 50 or 10 CFR part 52, the activities in paragraph (1)(i) of this
definition, and does not mean the activities in paragraph (1)(ii) of
this definition.
* * * * *
(3) For utilization facilities licensed under 10 CFR part 57, the
activities in the definition of construction in 10 CFR 57.3.
* * * * *
0
91. Add part 57, consisting of Sec. Sec. 57.1 through 57.445, to read
as follows:
PART 57--LICENSING REQUIREMENTS FOR MICROREACTORS AND OTHER
REACTORS WITH COMPARABLE RISK PROFILES
Subpart A--General Provisions
Sec.
57.157.1 Scope.
57.357.3 Definitions.
57.457.4 Written communications.
57.557.5 Deliberate misconduct.
57.657.6 Employee protection.
57.757.7 Completeness and accuracy of information.
57.857.8 Information collection requirements: OMB approval.
[[Page 23722]]
57.957.9 Specific exemptions.
57.11 Jurisdictional limits.
57.12 Attacks and destructive acts.
57.13 Rights related to special nuclear material.
57.14 License suspension and rights of recapture.
57.15 Agreement limiting access to Classified Information.
57.16 Backfitting and issue finality.
57.17 Referral to the Advisory Committee on Reactor Safeguards
(ACRS).
57.18 Combining licenses; elimination of repetition; relationships
between subparts.
57.19 Filing of applications.
Subpart B--Eligibility
57.20 Scope.
57.25 Applicability.
57.30 Design criteria attributes.
57.35 Licensing requirements.
Subpart C--Construction Permits and Operating Licenses
57.40 Scope.
57.45 License required; exceptions from licensing.
57.55 Contents of applications; general information.
57.60 Contents of applications; technical information.
57.80 Standards for review of applications.
57.90 Common standards for licenses.
57.95 Issuance of construction permit.
57.100 Issuance of operating license.
57.105 Continuation of license.
57.110 Transfer of licenses.
57.115 Application for renewal.
57.120 Criteria for renewal.
57.130 Hearings.
57.135 Duration of renewal.
57.142 Finality for construction permits and operating licenses.
Subpart D--Manufacturing Licenses
57.145 Scope.
57.150 Contents of applications for manufacturing licenses; general
information.
57.155 Contents of applications; technical information in final
safety analysis report.
57.160 Contents of applications; additional information.
57.165 Standards for review of applications.
57.170 Administrative review of applications; hearings.
57.172 Issuance of manufacturing license.
57.175 Finality of manufacturing licenses; information requests.
57.180 Duration of manufacturing license.
57.185 Transfer of manufacturing license.
57.190 Renewal of manufacturing licenses.
57.197 Manufacturing.
Subpart E--Standard Design Approvals
57.200 Scope.
57.205 Contents of applications; general information.
57.210 Contents of applications; technical information.
57.213 Standards for review of applications.
57.215 Staff approval of design.
57.220 Finality of standard design approvals; information requests.
57.225 Duration of design approval.
Subpart F--Reporting of Defects and Noncompliance
57.230 Purpose.
57.235 Scope.
57.240 Definitions.
57.255 Posting requirements.
57.260 Exemptions.
57.270 Notification of failure to comply or existence of a defect
and its evaluation.
57.275 Procurement documents.
57.280 Inspections.
57.285 Maintenance and inspection of records.
57.290 Failure to notify.
Subpart G--Irradiated Fuel Storage, Decommissioning, and Termination of
License Requirements
57.300 Irradiated fuel storage.
57.305 Decommissioning and license termination.
Subpart H--Maintaining and Revising Licensing Basis Information
57.310 Amendment of license.
57.312 Changes to facility as described in final safety analysis
reports.
57.315 Maintenance and submittal of the final safety analysis, as
updated.
57.317 Updated decommissioning report.
Subpart I--Transportation Package Design Certification
57.319 Purpose.
57.320 Applicability.
Subpart J--Physical Security Requirements
7.325 Physical security requirements.
Subpart K--Categorical Exclusion
57.350 Categorical exclusion.
Subpart L--Inspections
57.355 Unfettered access for inspections.
Subpart M--Material Control and Accounting
57.360 Material control and accounting.
Subpart N [Reserved]
Subpart O--Enforcement
57.380 Violations.
57.385 Criminal penalties.HD1>Subpart P--Operator Licensing and
Human Factors
57.390 Definitions.
57.391 General requirements for operator licensing and human
factors.
57.392 Communications.
57.393 Completeness and accuracy of information.
57.395 Human factors engineering requirements.
57.398 Operator license requirements.
57.399 Facility licensee requirements--General.
57.400 Facility licensee requirements related to GLROs.
57.405 Generally licensed reactor operators.
57.410 Generally licensed reactor operator training, examination,
and proficiency programs.
57.415 Cessation of individual applicability.
57.420 Operator licensing for operator-dependent facilities.
57.421 Medical requirements.
57.422 Incapacitation because of disability or illness.
57.423 Applications for operators and senior operators.
57.424 Training, examination, and proficiency programs.
57.425 Conditions of operator and senior operator licenses.
57.426 Issuance, modification, and revocation of operator and senor
operator licenses.
57.427 Expiration of operator and senior operator licenses.
57.429 Training and qualification for non-licensed personnel.
Subpart Q--Reporting and Other Administrative Requirements
57.430 Maintenance of records, making of reports.
57.435 Reporting requirements.
57.440 Licensee event report system.
57.445 Reports of radiation exposure to members of the public.
Authority: Atomic Energy Act of 1954, secs. 11, 101, 103, 108,
122, 147, 161, 181, 182, 183, 184, 185, 186, 187, 189, 223, 234 (42
U.S.C. 2014, 2131, 2132, 2133, 2134, 2135, 2138, 2152, 2167, 2169,
2201, 2231, 2232, 2233, 2234, 2235, 2236, 2237, 2239, 2273, 2282);
Energy Reorganization Act of 1974, secs. 201, 202, 206, 211 (42
U.S.C. 5841, 5842, 5846, 5851); Nuclear Waste Policy Act of 1982,
sec. 306 (42 U.S.C. 10226); National Environmental Policy Act of
1969 (42 U.S.C. 4332); 44 U.S.C. 3504 note; Pub. L. 118-67, div. B,
July 9, 2024, 138 Stat. 1448.
Subpart A--General Provisions
Sec. 57.157.1 Scope.
Subpart A provides general provisions applicable to all applicants
and licensees subject to the requirements of this part.
Sec. 57.357.3 Definitions.
For the purposes of this part, the following definitions apply,
although an applicant may provide its own definitions of these terms in
an application submitted under this part if the definitions are
supported by the applicant's safety analysis, except for those terms
defined in the Atomic Energy Act of 1954 (68 Stat. 919), as amended
(AEA).
Applicant means a person applying for a license, construction
permit, or other form of Commission permission or approval under this
part.
Autonomous operation means the performance of operational and
safety functions without reliance on human intervention, external
command, or active control system input under normal, abnormal, and
accident conditions.
Certified fuel handler means a non-licensed operator who
demonstrates compliance with the following criteria:
[[Page 23723]]
(1) Has qualified in accordance with a fuel handler training
program that demonstrates compliance with the same requirements as
training programs for non-licensed operators required by Sec. 57.420,
and
(2) Is responsible for decisions on--
(i) Safe conduct of decommissioning activities,
(ii) Safe handling and storage of spent fuel, and
(iii) Appropriate response to plant emergencies.
Commission means the Nuclear Regulatory Commission or its duly
authorized representatives.
Construction means the driving of piles, subsurface preparation,
placement of backfill, concrete, or permanent retaining walls within an
excavation, installation of foundations, or in-place assembly,
erection, fabrication, or testing, which are for safety-related
structures, systems, or components (SSCs) of a facility or SSCs that
are relied upon to implement the requirements in Sec. 57.60(a)(8)(v)
or subpart J of this part.
Controls means an apparatus and mechanisms, the manipulation of
which directly affects the reactivity or power level of the reactor.
Control room means a location either inside or outside the site
boundary where actions can be taken to operate the nuclear reactor
safely under normal conditions and to maintain it in a safe condition
under accident conditions.
Decommission means to remove an individually licensed nuclear
reactor, a nuclear plant, or a site safely from service and reduce
residual radioactivity to a level that permits--
(1) Release of the property for unrestricted use and termination of
the license; or
(2) Release of the property under restricted conditions and
termination of the license.
Defense in depth means inclusion of 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.
Department and Department of Energy means the Department of Energy
established by the Department of Energy Organization Act (Pub. L. 95-
91, 91 Stat. 565, 42 U.S.C. 7101 et seq.), to the extent that the
department, or its duly authorized representatives, exercises functions
formerly vested in the Atomic Energy Commission, its Chairman, members,
officers and components and transferred to the U.S. Energy Research and
Development Administration and to the Administrator thereof pursuant to
sections 104 (b), (c) and (d) of the Energy Reorganization Act of 1974
(Pub. L. 93-438, 88 Stat. 1233 at 1237, 42 U.S.C. 5814) and
retransferred to the Secretary of Energy pursuant to section 301(a) of
the Department of Energy Organization Act (Pub. L. 95-91, 91 Stat. 565
at 577-578, 42 U.S.C. 7151).
Design bases means 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 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.
Design features mean the active and passive safety-related SSCs and
inherent characteristics of those safety-related 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.
Director means an individual, appointed or elected according to
law, who is authorized to manage and direct the affairs of a
corporation, partnership or other entity. In the case of an individual
proprietorship, director means the individual.
Electric utility means any entity within the U.S. Nuclear
Regulatory Commission's (NRC's) jurisdiction that generates or
distributes electricity and which recovers the cost of this
electricity, either directly or indirectly, through rates established
by the entity itself or by a separate regulatory authority. Investor-
owned utilities, including generation or distribution subsidiaries,
public utility districts, municipalities, rural electric cooperatives,
and State and Federal agencies, including associations of any of the
foregoing, are included within the meaning of ``electric utility.''
Fission product release means the amount and composition of
radioactive material released to the environment, after accounting for
any retention of radionuclides provided by reactor design features.
Fuel means special nuclear material (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.
Government agency means any executive department, commission,
independent establishment, corporation, wholly or partly owned by the
United States of America which is an instrumentality of the United
States, or any board, bureau, division, service, office, officer,
authority, administration, or other establishment in the executive
branch of the Government.
License means a license, including a construction permit, operating
license, or manufacturing license, issued by the Commission under this
part.
Licensee means a person who is authorized to conduct activities
under a license issued by the Commission.
Licensing basis information means information contained in
regulations, orders, licenses, certifications, or approvals issued by
the NRC for a nuclear plant licensed under this part and that
information submitted to the NRC by an applicant or licensee in a final
safety analysis report, program description, or other licensing-related
document required under this part.
Manufactured reactor means the essential portions of a nuclear
reactor that are manufactured under a manufacturing license and
subsequently incorporated into a nuclear plant under a construction
permit issued under subpart C of this part.
Manufacturing license means a license issued under subpart D of
this part that authorizes the manufacture of manufactured reactors but
not their construction, installation, or operation.
Notification means communication to the NRC Operations Center or
written transmittal of information to the NRC Document Control Desk.
Nuclear plant means one or more nuclear reactors and the supporting
safety-related SSCs and other SSCs used together to generate thermal
energy to produce electricity or process heat, or for other
applications.
Nuclear reactor means an apparatus, other than an atomic weapon,
designed or used to sustain nuclear fission in a self-supporting chain
reaction.
Operating or operation means the operation of a facility or the
conduct of a licensed activity which is subject to the regulations in
this part and consulting services related to operations that are safety
related.
Person means:
(1) Any individual, corporation, partnership, firm, association,
trust, estate, public or private institution, group, government agency
other than the Commission or the Department,
[[Page 23724]]
except that the Department will be considered a person to the extent
that its facilities are subject to the licensing and related regulatory
authority of the Commission pursuant to section 202 of the Energy
Reorganization Act of 1974, any State or any political subdivision of,
or any political entity within a State, any foreign government or
nation or any political subdivision of any such government or nation,
or other entity; and
(2) Any legal successor, representative, agent, or agency of the
foregoing.
Previously disturbed area means areas that have been changed by
development of a prior facility and remain altered by human activity
such that they do not provide habitat for ecologically important
species, such as those protected under the Endangered Species Act, and
no longer have the potential to yield historic and cultural resources.
This definition will include the lateral and vertical extent of
alteration from natural cover to a managed state.
Programmatic controls means administrative procedures that govern
human action in implementing programs and operating, monitoring, and
maintaining safety-related SSCs and equipment of a nuclear plant.
Quality assurance means those planned and systematic actions during
design, construction, and modification necessary to provide adequate
confidence that the structure, system, or component will perform
satisfactorily in service.
Remote monitoring means observation of plant data from a location
outside of the site boundary.
Remote operation means command and control of the nuclear reactor
or nuclear plant from a location outside of the site boundary.
Restricted data means all data concerning:
(1) Design, manufacture, or utilization of atomic weapons;
(2) The production of special nuclear material; or
(3) The use of special nuclear material in the production of energy
but must not include data declassified or removed from the Restricted
Data category pursuant to section 142 of the AEA.
Safe shutdown means, under design basis accident conditions with
loss of emergency power and off site power, bringing the nuclear
reactor to safe, stable conditions specified in plant technical
specifications.
Safety function means a purpose served by a design feature, human
action, or programmatic control to prevent or mitigate unplanned events
and thereby demonstrate compliance with requirements in this part for
limiting risks to public health and safety. Safety functions can be
performed by any combination of the elements supported by the safety
analysis and can be specified at the plant level or at the level of a
particular barrier or system. Multiple plant-level safety functions are
assumed to apply to all reactor designs based on established
requirements and historical practices. These fundamental safety
functions 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 may also be referred to as a safety function. Subpart B
provides qualitative information of design criteria attributes for
control of reactivity, removal of heat, and limiting the release of
radioactive materials.
Safety-related SSCs means those SSCs of a nuclear plant that are
relied upon to remain functional during and following design basis
accidents to ensure:
(1) The capability to adequately control thermodynamic conditions
and reactivity, and to retain radioactive material;
(2) The capability to shut down the reactor and maintain it in a
safe shutdown condition; or
(3) The capability to prevent or mitigate the consequences of
accidents analyzed to meet the entry criteria in subpart B of this
part.
Source material means source material as defined in section 11(z)
of the AEA and in the regulations contained in part 40 of this chapter.
Source term means 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.
Special nuclear material means:
(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
special nuclear material, but does not include source material; or
(2) Any material artificially enriched by any of the foregoing, but
does not include source material.
Standard design approval or design approval means an NRC staff
approval, issued under subpart E of this part of a final standard
design for a nuclear reactor. The approval may be for either the final
design for the entire nuclear reactor or the final design of major
portions thereof.
Total effective dose equivalent (TEDE) means the sum of the
effective dose equivalent (for external exposures) and the committed
effective dose equivalent (for internal exposures).
Unrestricted area means a location where the public can be present
without restrictions related to radiation exposure. These areas are
characterized by the absence of controls to limit access specifically
for radiation protection purposes.
Utilization facility means any nuclear reactor other than one
designed or used primarily for the formation of plutonium or U-233.
Sec. 57.457.4 Written communications.
(a) General requirements. All correspondence, reports,
applications, and other written communications from the applicant or
licensee to the NRC concerning the regulations in this part or
individual license conditions must be sent either by mail addressed:
ATTN: Document Control Desk, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001; by hand delivery to the NRC's offices at
11555 Rockville Pike, Rockville, Maryland, between the hours of 8:15
a.m. and 4 p.m. eastern time; or, where practicable, by electronic
submission, for example, via Electronic Information Exchange, email, or
CD-ROM. Electronic submissions must be made in a manner that enables
the NRC to receive, read, authenticate, distribute, and archive the
submission, and process and retrieve it a single page at a time.
Detailed guidance on making electronic submissions can be obtained by
visiting the NRC's website at https://www.nrc.gov/site-help/e-submittals.html; by email to [email protected]; or by writing the
Office of the Chief Information Officer, U.S. Nuclear Regulatory
Commission, Washington, DC 20555-0001. The guidance discusses, among
other topics, the formats the NRC can accept, the use of electronic
signatures, and the treatment of nonpublic information. If the
communication is on paper, the signed original must be sent. If a
submission due date falls on a Saturday, Sunday, or Federal holiday,
the next Federal working day becomes the official due date.
(b) Distribution requirements. Copies of all correspondence,
reports, and other written communications concerning the regulations in
this part, individual license conditions, or the terms and conditions
of a standard design approval, must be submitted to the persons listed
in this section (addresses for the NRC Regional Offices are listed in
appendix D to 10 CFR part 20).
[[Page 23725]]
(1) Applications for amendment of construction permits and
licenses, reports, and other communications. All written communications
(including responses to generic letters, bulletins, information
notices, regulatory information summaries, inspection reports, and
miscellaneous requests for additional information) that are required of
holders of licenses, construction permits, or design approvals issued
pursuant to this part, must be submitted as follows, except as
otherwise specified in paragraphs (b)(2) through (7) of this section:
to the NRC's Document Control Desk (if on paper, the signed original),
with a copy to the appropriate Regional Office, and a copy to the
appropriate NRC Resident Inspector if one has been assigned to the site
of the facility or the place of manufacture of a reactor licensed under
this part.
(2) Applications for construction permits and licenses, and
amendments to applications. Applications for licenses, construction
permits, and design approvals and amendments to any of these types of
applications must be submitted to the NRC's Document Control Desk, with
a copy to the appropriate Regional Office, and a copy to the
appropriate NRC Resident Inspector if one has been assigned to the
facility or the place of manufacture of a reactor licensed under this
part, except as otherwise specified in paragraphs (b)(3) through (9) of
this section. If the application or amendment is on paper, the
submission to the Document Control Desk must be the signed original.
(3) Acceptance review application. Written communications required
for an application for determination of suitability for docketing must
be submitted to the NRC's Document Control Desk, with a copy to the
appropriate Regional Office. If the communication is on paper, the
submission to the Document Control Desk must be the signed original.
(4) Security plan and related submissions. Written communications,
as defined in paragraphs (b)(4)(i) through (v) of this section, must be
submitted to the NRC's Document Control Desk, with a copy to the
appropriate Regional Office. If the communication is on paper, the
submission to the Document Control Desk must be the signed original.
Submissions should include the following as appropriate:
(i) Physical security plan;
(ii) Safeguards contingency plan;
(iii) Cybersecurity plan;
(iv) Application for amendment of the physical security plan,
safeguards contingency plan, or cybersecurity plan as part of an
application for amendment of the license; and
(v) Changes to the physical security plan, safeguards contingency
plan, or cybersecurity plan made without prior Commission approval if
the changes do not decrease the safeguards effectiveness of these
plans.
(5) Security plan and related changes and records.
(i) The licensee must maintain records of changes to the
submissions in paragraphs (b)(4)(i) through (iii) of this section made
without prior approval for a period of three years from the date of the
change, and must, within two months after the change is made, submit a
report addressed to Director, Office of Nuclear Security and Incident
Response, U.S. Nuclear Regulatory Commission, in accordance with this
section, containing a description of each change.
(ii) A copy of the report must be sent to the Regional
Administrator of the appropriate NRC Regional Office specified in
appendix A to part 73 of this chapter.
(6) Emergency plan and related submissions. Written communications
as defined in paragraphs (b)(5)(i) through (ii) of this section must be
submitted to the NRC's Document Control Desk, with a copy to the
appropriate Regional Office, and a copy to the appropriate NRC Resident
Inspector if one has been assigned to the site of the facility. If the
communication is on paper, the submission to the Document Control Desk
must be the signed original. Submissions should include the following
as appropriate:
(i) Emergency plan; and
(ii) Change to an emergency plan.
(7) Updated final safety analysis report. An updated final safety
analysis report (FSAR) or replacement pages under Sec. 57.315 must be
submitted to the NRC's Document Control Desk every 5 years beginning 5
years after the date of issuance of an operating license or
manufacturing license under this part to ensure that the information
included in the report contains the latest information developed. This
submittal must contain all the changes necessary to reflect information
and analyses submitted to the Commission by the applicant or licensee
or prepared by the applicant or licensee pursuant to Commission
requirement since the submittal of the original FSAR, or as
appropriate, the last update to the FSAR under this section. The
submittal must include the effects of all changes made in the facility
or procedures as described in the FSAR; all safety analyses and
evaluations performed by the applicant or licensee either in support of
approved license amendments or in support of conclusions that changes
did not require a license amendment in accordance with Sec.
50.59(c)(2) of this chapter and all analyses of new safety issues
performed by or on behalf of the applicant or licensee at Commission
request. Effects of changes include appropriate revisions of
descriptions in the FSAR such that the updated FSAR is complete and
accurate. The updated information must be appropriately located within
the FSAR (as updated). If the communication is on paper, the submission
to the Document Control Desk must be the signed original. If the
communications are submitted electronically, see Guidance for
Electronic Submissions to the Commission.
(8) Quality assurance related submissions. Changes to the final
safety analysis report quality assurance program description under
Sec. 57.60(a)(3), or a change to a licensee's NRC-accepted quality
assurance topical report, must be submitted to the NRC's Document
Control Desk, with a copy to the appropriate Regional Office, and a
copy to the appropriate NRC Resident Inspector if one has been assigned
to the site of the facility or the place of manufacture of a reactor
licensed under this part. If the communication is on paper, the
submission to the Document Control Desk must be the signed original
copy.
(9) Certification of permanent cessation of operations. The
licensee's certification of permanent cessation of operations, under
subpart G of this part, must state the date on which operations have
ceased or will cease, and must be submitted to the NRC's Document
Control Desk. This submission must be under oath or affirmation.
(10) Certification of permanent fuel removal. The licensee's
certification of permanent fuel removal, under subpart G of this part,
must state the date on which the fuel was removed from the reactor
vessel and the disposition of the fuel, and must be submitted to the
NRC's Document Control Desk. This submission must be under oath or
affirmation.
(c) Form of communications. All paper copies submitted to
demonstrate compliance with the requirements set forth in paragraph (b)
of this section must be typewritten, printed, or otherwise reproduced
in permanent form on unglazed paper. Exceptions to these requirements
imposed on paper submissions may be granted for the submission of
micrographic, photographic, or similar forms.
[[Page 23726]]
(d) Regulation governing submission. Licensees and applicants under
this part submitting correspondence, reports, and other written
communications under the regulations of this part are requested but not
required to cite whenever practical, in the upper right corner of the
first page of the submission, the specific regulation or other basis
requiring submission.
Sec. 57.557.5 Deliberate misconduct.
(a) Any licensee or applicant for a license, or holder of or
applicant for a standard design approval, under this part; employee of
a licensee or holder of a standard design approval, or applicant for a
license or standard design approval under this part; or any contractor
(including a supplier or consultant), subcontractor, employee of a
contractor or subcontractor of any licensee or applicant for a license,
or holder of or applicant for a standard design approval under this
part, who knowingly provides to any licensee, applicant, contractor, or
subcontractor, any components, equipment, materials, or other goods or
services that relate to a licensee's or applicant's activities in this
part, may not--
(1) Engage in deliberate misconduct that causes or would have
caused, if not detected, a licensee or applicant to be in violation of
any rule, regulation, or order; or any term, condition, or limitation
of any license issued by the Commission; or
(2) Deliberately submit to the NRC, a licensee, an applicant, or a
licensee's or applicant's contractor or subcontractor, information that
the person submitting the information knows to be incomplete or
inaccurate in some respect material to the NRC.
(b) A person who violates paragraph (a)(1) or (2) of this section
may be subject to enforcement action in accordance with the procedures
in subpart B of 10 CFR part 2.
(c) For the purposes of paragraph (a)(1) of this section,
deliberate misconduct by a person means an intentional act or omission
that the person knows--
(1) Would cause a licensee or applicant to be in violation of any
rule, regulation, or order; or any term, condition, or limitation, of
any license issued by the Commission; or
(2) Constitutes a violation of a requirement, procedure,
instruction, contract, purchase order, or policy of a licensee,
applicant, contractor, or subcontractor.
Sec. 57.657.6 Employee protection.
(a) Discrimination by a Commission licensee, a holder of a standard
design approval, an applicant for a license or standard design
approval, or a contractor or subcontractor of a Commission licensee,
holder of a standard design approval, or an applicant for a license or
standard design approval, against an employee for engaging in certain
protected activities is prohibited. Discrimination includes discharge
and other actions that relate to compensation, terms, conditions, or
privileges of employment. The protected activities are established in
section 211 of the Energy Reorganization Act of 1974, as amended, and
in general are related to the administration or enforcement of a
requirement imposed under the AEA or the Energy Reorganization Act of
1974, as amended.
(1) The protected activities include but are not limited to--
(i) Providing the Commission or his or her employer information
about alleged violations of either of the statutes named in paragraph
(a) of this section or possible violations of requirements imposed
under either of those statutes;
(ii) Refusing to engage in any practice made unlawful under either
of the statutes named in paragraph (a) of this section or under these
requirements if the employee has identified the alleged illegality to
the employer;
(iii) Requesting the NRC to institute action against his or her
employer for the administration or enforcement of these requirements;
(iv) Testifying in any Commission proceeding, or before Congress,
or at any Federal or State proceeding regarding any provision (or
proposed provision) of either of the statutes named in paragraph (a) of
this section; and
(v) Assisting or participating in, or being about to assist or
participate in, these activities.
(2) These activities are protected even if no formal proceeding is
actually initiated as a result of the employee assistance or
participation.
(3) This section does not apply to any employee alleging
discrimination prohibited by this section who, acting without direction
from his or her employer (or the employer's agent), deliberately causes
a violation of any requirement of the Energy Reorganization Act of
1974, as amended, or the AEA.
(b) Any employee who believes that they have been discharged or
otherwise discriminated against by any person for engaging in protected
activities specified in paragraph (a)(1) of this section may seek a
remedy for the discharge or discrimination through an administrative
proceeding in the Department of Labor. The administrative proceeding
must be initiated within 180 days after an alleged violation occurs.
The employee may do this by filing a complaint alleging the violation
with the Department of Labor, Wage and Hour Division. The Department of
Labor may order reinstatement, back pay, and compensatory damages.
(c) A violation of paragraph (a), (e), or (f) of this section by a
Commission licensee, a holder of a standard design approval, an
applicant for a Commission license or standard design approval, or a
contractor or subcontractor of a Commission licensee or holder of a
standard design approval, or any applicant may be grounds for--
(1) Denial, revocation, or suspension of the license or standard
design approval;
(2) Imposition of a civil penalty on the licensee, holder of a
standard design approval, or applicant, or a contractor or
subcontractor of the licensee, holder of a standard design approval or
applicant; or
(3) Other enforcement action.
(d) Actions taken by an employer, or others, which adversely affect
an employee may be predicated upon nondiscriminatory grounds. The
prohibition applies when the adverse action occurs because the employee
has engaged in protected activities. An employee's engagement in
protected activities does not automatically render him or her immune
from discharge or discipline for legitimate reasons or from adverse
action dictated by nonprohibited considerations.
(e) To ensure employees are informed of their rights, each license
holder or applicant must follow the guidelines for posting NRC Form 3,
``Notice to Employees,'' as follows:
(1) Each holder or applicant for a license or design approval must
prominently post the revision of NRC Form 3, ``Notice to Employees,''
referenced in Sec. 19.11(e)(1) of this chapter. This form must be
posted at locations sufficient to permit employees protected by this
section to observe a copy on the way to or from their place of work.
Premises must be posted no later than 30 days after an application is
docketed and remain posted while the application is pending before the
Commission, during the term of the license, and for 30 days following
license termination.
[[Page 23727]]
(2) Copies of NRC Form 3 may be obtained by writing to the Regional
Administrator of the appropriate NRC Regional Office listed in appendix
D to 10 CFR part 20, via email to [email protected], or by
visiting the NRC's online library at https://www.nrc.gov/reading-rm/doc-collections/forms/.
(f) No agreement affecting the compensation, terms, conditions, or
privileges of employment, including an agreement to settle a complaint
filed by an employee with the Department of Labor pursuant to section
211 of the Energy Reorganization Act of 1974, as amended, may contain
any provision which would prohibit, restrict, or otherwise discourage
an employee from participating in protected activity as defined in
paragraph (a)(1) of this section including, but not limited to,
providing information to the NRC or to his or her employer on potential
violations or other matters within NRC's regulatory responsibilities.
(g) Part 19 of 10 CFR sets forth requirements and regulatory
provisions applicable to licensees, holders of a standard design
approval, applicants for a license or standard design approval, and
contractors or subcontractors of a Commission licensee or holder of a
standard design approval, and are in addition to the requirements in
this section.
Sec. 57.757.7 Completeness and accuracy of information.
(a) Information provided to the Commission by a holder of a
license, construction permit, or standard design approval under this
part or an applicant for a license, construction permit, or standard
design approval under this part, and information required by statute or
by the Commission's regulations, orders, license conditions, or terms
and conditions of a standard design approval to be maintained by the
applicant or the licensee must be complete and accurate in all material
respects.
(b) Each applicant or licensee and each holder of a standard design
approval under this part must notify the Commission of information
identified by the applicant or licensee as having for the regulated
activity a significant implication for public health and safety or
common defense and security. An applicant, licensee, or holder violates
this paragraph only if the applicant, licensee, or holder fails to
notify the Commission of information that the applicant, licensee, or
holder has identified as having a significant implication for public
health and safety or common defense and security. Notification must be
provided to the Administrator of the appropriate Regional Office within
2 working days of identifying the information. This requirement is not
applicable to information which is already required to be provided to
the Commission by other reporting or updating requirements.
Sec. 57.857.8 Information collection requirements: OMB approval.
(a) The NRC has submitted the information collection requirements
contained in this part to the Office of Management and Budget (OMB) for
approval as required by the Paperwork Reduction Act (44 U.S.C. 3501 et
seq.). The NRC may not conduct or sponsor, and a person is not required
to respond to, a collection of information unless it displays a
currently valid OMB control number. OMB has approved the information
collection requirements contained in this part under control number
3150-XXXX.
(b) The approved information collection requirements contained in
this part appear in Sec. Sec. 57.7, 57.9, 57.15, 57.45, 57.55, 57.60,
57.95, 57.110, 57.115, 57.145, 57.150, 57.155, 57.160, 57.190, 57.197,
57.205, 57.210, 57.220, 57.255, 57.270, 57.285, 57.300, 57.305, 57.310,
57.315, 57.317, 57.325, 57.395, 57.399, 57.400, 57.405, 57.410, 57.424,
57.425, 57.429, 57.430, 57.435, 57.445.
(c) This part contains information collection requirements in
addition to those approved under the control number specified in
paragraph (a) of this section. The information collection requirement
and the control numbers under which it is approved are as follows:
(1) In Sec. Sec. 57.421, 57.422, and 57.423, NRC Form 396 is
approved under control number 3150-0024.
(2) In Sec. Sec. 57.423 and 57.424, NRC Form 398 is approved under
control number 3150-0090.
(3) In Sec. 57.435, NRC Form 361 is approved under control number
3150-0238.
(4) In Sec. 57.440, NRC Form 366 is approved under control number
3150-0104.
Sec. 57.957.9 Specific exemptions.
(a) The Commission may, upon application by any interested person
or upon its own initiative, grant exemptions from the requirements of
the regulations of this part, which are authorized by law, will not
present an undue risk to the public health and safety, and are
consistent with the common defense and security.
(b) The Commission will not consider granting an exemption unless
special circumstances are present. Special circumstances are present
whenever--
(1) Application of the regulation in the particular circumstances
conflicts with other rules or requirements of the Commission;
(2) Application of the regulation in the particular circumstances
would not serve the underlying purpose of the rule or is not necessary
to achieve the underlying purpose of the rule;
(3) Compliance would result in undue hardship or other costs that
are significantly in excess of those contemplated when the regulation
was adopted, or that are significantly in excess of those incurred by
others similarly situated;
(4) The exemption would result in benefit to the public health and
safety that compensates for any decrease in safety that may result from
the grant of the exemption;
(5) The exemption would provide only temporary relief from the
applicable regulation and the licensee or applicant has made good faith
efforts to comply with the regulation; or
(6) There is present any other material circumstance not considered
when the regulation was adopted for which it would be in the public
interest to grant an exemption. If such condition is relied on
exclusively for demonstrating compliance with paragraph (b) of this
subsection, the exemption may not be granted until the Executive
Director for Operations has consulted with the Commission.
(c) Any person may request an exemption permitting the conduct of
construction before the issuance of a construction permit. The
Commission may grant such an exemption upon considering and balancing
the following factors:
(1) Whether conduct of the proposed activities will give rise to a
significant adverse impact on the environment and the nature and extent
of such impact, if any;
(2) Whether redress of any adverse environment impact from conduct
of the proposed activities can reasonably be effective should such
redress be necessary;
(3) Whether conduct of the proposed activities would foreclose
subsequent adoption of alternatives; and
(4) The effect of delay in conducting such activities on the public
interest, including whether the power needs to be used by the proposed
facility, the availability of alternative sources, if any, to meet
those needs on a timely basis and delay costs to the applicant and to
consumers.
(d) Issuance of such an exemption will not be deemed to constitute
a
[[Page 23728]]
commitment to issue a construction permit. During the period of any
exemption granted pursuant to paragraph (c) of this section, any
activities conducted must be carried out in such a manner as will
minimize or reduce their environmental impact.
(e) The Commission's consideration of requests for exemptions from
requirements of the regulations of other parts in this chapter that are
applicable by virtue of this part will be governed by the exemption
requirements of those parts.
Sec. 57.11 Jurisdictional limits.
No license or standard design approval under this part may be
deemed to have been issued for activities that are not under or within
the jurisdiction of the United States.
Sec. 57.12 Attacks and destructive acts.
Licensees, holders of a standard design approval, applicants for
licenses and design approvals, and applicants for an amendment to any
license or design approval under this part are not required to provide
for design features or other measures for the specific purpose of
protection against the effects of--
(a) Attacks and destructive acts, including sabotage, directed
against the facility by an enemy of the United States, whether a
foreign government or other person; or
(b) Use or deployment of weapons incident to U.S. defense
activities.
Sec. 57.13 Rights related to special nuclear material.
(a) No right to the SNM will be conferred by a license issued under
this part except as may be defined by the license.
(b) Neither a license issued under this part, nor any right
thereunder, nor any right to utilize or produce SNM may be transferred,
assigned, or disposed of in any manner, either voluntarily or
involuntarily, directly or indirectly, through transfer of control of
the license to any person, unless the Commission, after securing full
information, finds that the transfer is in accordance with the
provisions of the AEA and gives its consent in writing.
Sec. 57.14 License suspension and rights of recapture.
Any license issued under this part will be subject to suspension
and to the rights of recapture of the material or control of the
facility reserved to the Commission under section 108 of the AEA in a
state of war or national emergency declared by Congress.
Sec. 57.15 Agreement limiting access to classified information.
As part of its application under this part and in any event before
the receipt of Restricted Data or classified National Security
Information or the issuance of a license or standard design approval,
the applicant must agree in writing that it will not permit any
individual to have access to, or any facility to possess, Restricted
Data or classified National Security Information until the individual
and/or facility has been approved for access under the provisions of 10
CFR parts 25 and/or 95. The agreement of the applicant becomes part of
the license or standard design approval.
Sec. 57.16 Backfitting and issue finality.
(a) Backfitting.
(1) Assessment.
(i) Definition. Backfitting is defined as the modification of or
addition to systems, structures, components, or design of a facility;
or the standard design approval or manufacturing license for a
facility; or the procedures or organization required to design,
construct or operate a facility; any of which may result from a new or
amended provision in the Commission's regulations or the imposition of
a regulatory staff position interpreting the Commission's regulations
that is either new or different from a previously applicable staff
position after:
(A) The date of issuance of a construction permit under subpart C
of this part;
(B) The date of issuance of an operating license under subpart C of
this part;
(C) The date of issuance of a manufacturing license under subpart D
of this part; or
(D) The date of issuance of a standard design approval under
subpart E of this part.
(ii) Proposed backfitting. Except as provided in paragraph
(a)(1)(iv) of this section, the Commission must require a systematic
and documented analysis pursuant to paragraph (a)(2) of this section
for backfits which it seeks to impose.
(iii) Backfit analysis. Except as provided in paragraph (a)(1)(iv)
of this section, the Commission must require the backfitting of a
facility only when it determines, based on the analysis described in
paragraph (a)(2) of this section, that there is a substantial increase
in the overall protection of the public health and safety or the common
defense and security to be derived from the backfit and that the direct
and indirect costs of implementation for that facility are justified in
view of this increased protection.
(iv) Exceptions. The provisions of paragraphs (a)(1)(ii) and (iii)
of this section are inapplicable and, therefore, backfit analysis is
not required and the standards in paragraph (a)(1)(iii) of this section
do not apply where the Commission or staff, as appropriate, finds and
declares, with appropriate documented evaluation for its finding,
either:
(A) That a modification is necessary to bring a facility into
compliance with a license or the rules or orders of the Commission, or
into conformance with written commitments by the licensee; or
(B) That regulatory action is necessary to ensure that the facility
provides adequate protection to the health and safety of the public and
is in accord with the common defense and security; or
(C) That the regulatory action involves defining or redefining what
level of protection to the public health and safety or common defense
and security should be regarded as adequate.
(v) Mandatory backfitting. The Commission will always require the
backfitting of a facility if it determines that such regulatory action
is necessary to ensure that the facility provides adequate protection
to the health and safety of the public and is in accord with the common
defense and security.
(vi) Documented evaluation. The documented evaluation required by
paragraph (a)(1)(iv) of this section must include a statement of the
objectives of and reasons for the modification and the basis for
invoking the exception. If immediately effective regulatory action is
required, then the documented evaluation may follow rather than precede
the regulatory action.
(vii) Implementation. If there are two or more ways to achieve
compliance with a license or the rules or orders of the Commission, or
with written licensee commitments, or there are two or more ways to
reach a level of protection which is adequate, then ordinarily the
applicant or licensee is free to choose the way which best suits its
purposes. However, should it be necessary or appropriate for the
Commission to prescribe a specific way to comply with its requirements
or to achieve adequate protection, then cost may be a factor in
selecting the way, provided that the objective of compliance or
adequate protection is met.
(2) Backfit analysis factors. In reaching the determination
required by paragraph (a)(1)(iii) of this section, the Commission will
consider how the backfit should be scheduled in light of
[[Page 23729]]
other ongoing regulatory activities at the facility and, in addition,
will consider information available concerning any of the following
factors as may be appropriate and any other information relevant and
material to the proposed backfit:
(i) Statement of the specific objectives that the proposed backfit
is designed to achieve;
(ii) General description of the activity that would be required by
the licensee or applicant in order to complete the backfit;
(iii) Potential change in the risk to the public from the
accidental off site release of radioactive material;
(iv) Potential impact on radiological exposure of facility
employees;
(v) Installation and continuing costs associated with the backfit,
including the cost of facility downtime or the cost of construction
delay;
(vi) The potential safety impact of changes in plant or operational
complexity, including the relationship to proposed and existing
regulatory requirements;
(vii) The estimated resource burden on the NRC associated with the
proposed backfit and the availability of such resources;
(viii) The potential impact of differences in facility type, design
or age on the relevancy and practicality of the proposed backfit;
(b) Issue finality. In the proceedings for issuance of a standard
design approval, manufacturing license, construction permit, or
operating license under this part--
(1) For which a construction permit or operating license issued
under part 50 of this chapter, or a standard design approval or
combined license issued under part 52 of this chapter, is referenced,
the NRC staff and the Advisory Committee on Reactor Safeguards will use
and rely on the reactor design and any operational programs or
requirements with generic applicability that were approved in the
proceeding on the application for issuance or renewal of the
construction permit or operating license under part 50 of this chapter
or the standard design approval or combined license under part 52 of
this chapter, unless there exists significant new information that
substantially affects the earlier determination or other good cause.
(2) For which an early site permit, standard design certification,
or manufacturing license issued under part 52 of this chapter is
referenced, the Commission will treat as resolved those matters
resolved in the proceeding on the application for issuance or renewal
of the early site permit, standard design certification, or
manufacturing license under part 52 of this chapter.
(c) Requests for departures. An applicant or licensee under this
part who references a construction permit or operating license for a
nuclear reactor or nuclear plant that was afforded generic finality
under Sec. 57.142(e) or references a manufacturing license under this
chapter must include in the application analysis of each departure,
both individually and cumulatively, from the design characteristics,
site parameters, terms and conditions, or approved design of the
nuclear reactor, nuclear plant, or manufactured reactor. An applicant
is not required to provide analysis of departures from operational
programs or requirements approved with the referenced construction
permit, operating license, or manufacturing license that are not
material to the adequacy of the design, if the applicant proposes
alternative operational programs or requirements. Departures will be
subject to litigation in the same manner as other issues in the
construction permit or operating license hearing.
Sec. 57.17 Referral to the Advisory Committee on Reactor Safeguards
(ACRS).
The Commission will refer a copy of each initial joint application
submitted under this part for a construction permit and associated
operating license(s) and each application for a manufacturing license
or standard design approval to the ACRS. The ACRS must apply the
standards in Sec. Sec. 57.80, 57.165, and 57.213 in accordance with
the finality provisions for any construction permit, operating license,
manufacturing license, or standard design approval referenced in the
application. The ACRS review will focus on aspects of the design that
are principally unique, novel, and noteworthy. Any report will be made
part of the record of the application and available to the public,
except to the extent that security classification prevents disclosure.
Sec. 57.18 Combining licenses; elimination of repetition;
relationships between subparts.
(a) Applicants under this part may combine applications for
multiple and different kinds of licenses, certifications, and approvals
under the regulations of this part and parts 30, 40, 70, 71, and 72 of
this chapter.
(1) In situations in which applications are filed under this part
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, reference may be made to a single final safety
analysis report other than for applicant- or site-specific information.
(2) An applicant may include in its joint application for a
construction permit and operating licenses for a nuclear reactor or
nuclear plant under this part the information required by Sec.
57.60(a)(5) and 10 CFR part 51 for multiple sites at which the
applicant proposes to construct and operate the reactor or plant.
(3) An application under this part for multiple types of permits,
licenses, or certifications must clearly indicate to which permit,
license, or certification information in the application pertains.
(4) Holders of operating licenses under this part that reference
the same manufacturing license may combine applications for a license
amendment under Sec. 57.310 that would affect the facility or the
procedures described in the final safety analysis report for the
manufacturing license, and may combine, with the holder of the
manufacturing license that is referenced in the operating licenses,
applications for a license amendment submitted by the holder of the
manufacturing license under Sec. 57.310.
(5) An applicant may include in a joint application a request for a
construction permit for any number of nuclear reactors of essentially
the same design to be built at a specific site and requests for
operating licenses for those reactors, provided that the application
states the earliest and latest dates for completion of the construction
of each nuclear reactor as required by Sec. 57.55(g) and includes the
information specified in Sec. 57.60(a)(4).
(b) An applicant may incorporate by reference in its application
information contained in previous applications, statements, or reports
filed with the Commission and applicable Commission approvals issued
under part 50 or 52 of this chapter, provided that such references are
clear and specific. For an application under this part that references
an approval issued under part 50 or 52 of this chapter, the scope and
nature of matters resolved for that application are governed by Sec.
57.16(b).
(c) The Commission may combine in a single license the activities
of an applicant that would otherwise be licensed separately.
(d) A joint application for a construction permit and associated
operating license(s) filed under this part may reference a standard
design approval, construction permit, operating license, manufacturing
license, or combination thereof, issued under this part. An application
for a manufacturing license under this part may reference a
[[Page 23730]]
standard design approval issued under this part.
(e) An application for a standard design approval or manufacturing
license or a joint application for a construction permit and associated
operating license(s) filed under this part may reference 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. Any reference must identify how aspects of
the authorization address applicable NRC regulations in this part.
(f) Subparts in this part may be used independently.
Sec. 57.19 Filing of applications.
(a) Any person, except one excluded by 10 CFR 50.38, may file a
joint application for a construction permit and associated operating
license(s), or an application for a manufacturing license under this
part with the Director, Office of Nuclear Reactor Regulation.
(b) Any person may submit a proposed standard design for a nuclear
reactor of the type described in this part to the NRC staff for its
review. The submittal may consist of either the final design for the
entire nuclear reactor or the final design of major portions thereof.
(c) The application must comply with the applicable filing
requirements of 10 CFR 50.30 and subpart A of 10 CFR part 2.
(d) The submittal for review of a proposed standard design must be
made in the same manner as provided in 10 CFR 50.30 for license
applications.
(e) The fees associated with the filing and review of applications
under this part are set forth in 10 CFR part 170.
(f) An applicant for licenses to construct and operate one or more
nuclear reactors under subpart C of this part must file a joint
application for a construction permit and associated operating
license(s). The joint application must include the information
specified in Sec. 57.55 and Sec. 57.60 and be complete enough to
permit all evaluations necessary for the issuance of the requested
construction permit and the associated operating license(s) upon the
NRC making the finding required by Sec. 57.100(b)(1).
Subpart B--Eligibility
Sec. 57.20 Scope.
This subpart specifies the applicability criteria for construction
permit, operating license, and manufacturing license applicants and the
design criteria attributes for these applicants and standard design
approval applicants, under which these entities may be considered
eligible to use the provisions of this part.
Sec. 57.25 Applicability.
To be eligible for a construction permit and operating license or a
manufacturing license under this part, an applicant must demonstrate
that its nuclear reactor or nuclear plant design and operation meets
the following entry criteria:
(a) An evaluation of the applicable radiological consequences shows
with reasonable assurance that any individual located in the
unrestricted area following the onset of a postulated accident that
bounds a broad range of design basis accidents would not exceed 1 rem
(0.01 Sv) TEDE for the duration of the accident; and
(b) The total inventory of thorium, uranium, and plutonium
contained in the nuclear reactor or any individual nuclear reactor that
is part of the nuclear plant must not exceed 10 metric tons.
Sec. 57.30 Design criteria attributes.
The applicant for a license or design approval under this part must
provide information that demonstrates that the nuclear reactor or
nuclear plant design has design criteria attributes that satisfy the
following:
(a) Reactivity control. The design must provide for the following:
(1) Control of the power level during normal operations;
(2) Rapid insertion of reactivity control devices to immediately
shut down the reactor and maintain it in a safe shutdown state under
accident conditions; and
(3) Net negative reactivity feedback as a result of increased
reactor power.
(b) Heat removal. The design must provide for highly reliable
passive decay heat removal to limit core coolant and fuel temperatures
during accident conditions to within design limits to protect the fuel
and, as appropriate, the reactor coolant and fission product
boundaries.
(c) Fission product retention. The design must provide for the
protection of engineered fission product boundaries to limit the
fission product release of radionuclides during normal and accident
conditions.
(d) Shielding. The design must provide the following:
(1) Adequate permanent and temporary shielding to comply with 10
CFR part 20 for the protection of workers and the public from direct
radiation exposure from the reactor and radioactive sources during
operation, shutdown, and transport, including during abnormal
conditions; and
(2) Sufficient robustness and heat removal to prevent loss of
shielding integrity during normal and accident conditions
(e) Radioactive effluents control. The design must meet the
requirements of part 20 of this chapter for control, monitoring, and
release of radioactive materials to the environment.
(f) Security by design. Safety and security must be considered
together in the design process such that, where possible, security
issues are effectively resolved through design and engineered security
features.
Sec. 57.35 Licensing requirements.
(a) If an applicant for a construction permit, license, or standard
design approval under this part can demonstrate that its reactor design
meets the applicable eligibility requirements of Sec. Sec. 57.25 and
57.30, then the applicant must comply with the applicable application
and procedural requirements set forth in this part.
(b) Notwithstanding the requirements of part 50 or 52 of this
chapter, if an applicant is issued a construction permit, license, or
design approval under this part, then that entity is subject to the
requirements of this part and not part 50 or 52 of this chapter unless
specifically required by this part.
Subpart C--Construction Permits and Operating Licenses
Sec. 57.40 Scope.
This subpart sets forth the requirements and procedures applicable
to Commission issuance of construction permits and operating licenses
for utilization facilities of the type described in Sec. 50.22 of this
chapter.
Sec. 57.45 License required; exceptions from licensing.
(a) Except as provided for in paragraph (b) of this section, no
person within the United States may transfer or receive in interstate
commerce, manufacture, produce, transfer, acquire, possess, or use any
utilization facility under this part except as authorized by a license
issued under this part by the Commission.
(b) Nothing in this part may be deemed to require a license for the
transportation or possession of a utilization facility by a common or
contract carrier or warehousemen in the regular course of carriage for
another or storage incident thereto.
(c) Except as provided for in paragraph (d) of this section, no
person may begin the construction of a utilization facility on a site
on which
[[Page 23731]]
the facility is to be operated until that person has been issued a
construction permit under this part.
(d) A general license is hereby issued for construction activities
on a site that is specified in a joint application for a construction
permit and associated operating license(s) under this part, subject to
the following conditions:
(1) The general licensee has submitted and the Commission docketed
a joint application for a construction permit and associated operating
license(s) under this part that meets the following criteria:
(i) The joint application references a manufacturing license issued
by the Commission under this chapter;
(ii) The joint application references a construction permit and
operating license issued pursuant to this part that the Commission
afforded generic finality under Sec. 57.142(e), that referenced the
same manufacturing license as the general licensee in its joint
application, and that met the criteria for a categorical exclusion
under subpart K of this part.
(iii) The joint application includes a plan for redress of any
adverse environmental impact from conduct of activities under the
general license should such redress be necessary.
(2) The general licensee has notified the NRC under Sec. 57.4 that
all applicable permits, licenses, approvals, and other entitlements in
connection with the proposed action have been obtained.
(3) All applicable Federal environmental consultations have been
completed.
(4) The general licensee must not allow special nuclear material or
radioactive material that would be associated with operation under an
operating license issued pursuant to this part to be brought to the
site under the general license;
(5) The general licensee must not allow a manufactured reactor to
be brought to the site under the general license.
(6) The general licensee must allow for NRC inspections that the
Commission deems necessary related to activities performed under the
general license.
(7) Any activities undertaken by the general licensee or on its
behalf under the general license are entirely at the risk of the
general licensee and have no bearing on the issuance of a construction
permit with respect to the requirements of the AEA, and rules,
regulations, or orders issued under the AEA.
Sec. 57.55 Contents of applications; general information.
Each application must state:
(a) Name of applicant;
(b) Address of applicant;
(c) Description of business or occupation of applicant;
(d) Organization information of applicant, including the following
information:
(1) If applicant is an individual, the citizenship of applicant.
(2) If applicant is a partnership, the name, citizenship and
address of each partner and the principal location where the
partnership does business.
(3) If applicant is a corporation or an unincorporated association,
the following information:
(i) The state where it is incorporated or organized and the
principal location where it does business;
(ii) The names, addresses and citizenship of its directors and of
its principal officers;
(iii) Whether it is owned, controlled, or dominated by an alien, a
foreign corporation, or foreign government, and if so, give details.
(4) If the applicant is acting as agent or representative of
another person in filing the application, identify the principal and
furnish information required under this paragraph with respect to such
principal.
(e) The type of license(s) applied for, the use to which the
facility will be put, the period of time for which the license(s) are
sought, and a list of other licenses, issued or applied for in
connection with the proposed facility.
(f) Except for an electric utility applicant for a license to
operate a utilization facility, information sufficient to demonstrate
to the Commission the financial qualification of the applicant to carry
out, in accordance with regulations in this chapter, the activities for
which the construction permit and operating license is sought. As
applicable, the following must be provided:
(1) For a construction permit under this section, the applicant
must submit information that demonstrates that the applicant appears to
be financially qualified to cover estimated construction costs and
related fuel cycle costs. The applicant must submit estimates of the
total construction costs of the facility and related fuel cycle costs,
a financial capacity plan, and any source(s) of funds available at the
time of application to cover these costs. If available funding at the
time of application is 50 percent or less, the applicant should include
proposed license conditions to facilitate verification that funding is
available prior to the start of construction.
(2) For an operating license under this section, the applicant must
submit information that demonstrates the applicant appears to be
financially qualified to cover estimated operation costs for the period
of the license. The applicant must submit estimates for total annual
operating costs for each of the first 5 years of operation of the
facility and a financial capacity plan and indicate any source(s) of
funds available at the time of application to cover these costs. If
available funding at the time of application is 50 percent or less, the
applicant should include proposed license conditions to facilitate
verification that funding is available prior to the start of
operations. An applicant seeking to renew or extend the term of an
operating license need not submit the financial information that is
required in an application for an initial license.
(g) If the applicant proposes to construct or materially alter a
utilization facility, the application must state the earliest and
latest dates for completion of the construction or material alteration.
(h) If the proposed activity is the generation and distribution of
electric energy under a license under this part, a list of the names
and addresses of such regulatory agencies as may have jurisdiction over
the rates and services incident to the proposed activity, and a list of
trade and news publications that circulate in the area where the
proposed activity will be conducted and that are considered appropriate
to give reasonable notice of the application to those municipalities,
private utilities, public bodies, and cooperatives, which might have a
potential interest in the facility.
(i) Information in the form of a report, as described in 10 CFR
50.75, indicating how reasonable assurance will be provided that funds
will be available to decommission the facility.
(j) If the application contains Restricted Data or classified
National Security Information, confirmation that all Restricted Data
and classified National Security Information are separated from the
unclassified information.
Sec. 57.60 Contents of applications; technical information.
(a) Final safety analysis report. Each application must include a
final safety analysis report that consists of the following:
(1) A description and safety assessment of the site and a safety
assessment of the facility, including the following:
(i) Intended use of the reactor including the maximum power level
[[Page 23732]]
and the nature and inventory of contained radioactive materials;
(ii) The safety features that are to be engineered into the
facility and those barriers that must be breached as a result of an
accident before a release of radioactive material to the environment
can occur. Special attention must be directed to design features
intended to prevent and mitigate the radiological consequences of
accidents.
(iii) An evaluation that meets the dose-based entry criterion of
Sec. 57.25(a). In performing this evaluation, an applicant must assume
a fission product release utilizing a postulated accident source term
that represents the most limiting fission product inventory during the
lifetime of the nuclear reactor while assuming that the reactor is
operated at the ultimate power level contemplated.
(iv) As applicable, a description and assessment of SSCs for remote
operation of the reactor from outside the site boundary that
demonstrates that the reactor can be safely operated and can reach and
maintain a safe shutdown state, including under abnormal conditions.
(v) As applicable, a description and assessment of design features
for remote monitoring of the nuclear reactor or nuclear plant from
outside the site boundary and protecting the integrity of important
safety parameters and safety function data needed to perform human
actions that protect public health and safety, and to protect sensitive
plant data that could be used to aid in an attack (physical or cyber)
against the reactor.
(vi) As applicable, a description and assessment of design features
for autonomous performance of operations and safety functions without
reliance on human intervention, external command, or active control
system input under normal, abnormal, and accident conditions.
(vii) Analysis, appropriate test programs, prototype testing,
operating experience, or a combination thereof that demonstrates that
each of the design criteria attributes described in Sec. 57.30 are
met. This demonstration must consider interdependent effects throughout
the nuclear plant for the duration of the nuclear plant's lifetime.
(2) The design basis of the facility, including:
(i) The principal design criteria.
(ii) Relation of the design bases to the principal design criteria.
(iii) Relation of the principal design criteria to the design
criteria attributes described in Sec. 57.30.
(3) A description of the quality assurance program to be applied to
the design, fabrication, manufacture (as applicable), construction, and
testing of the safety-related SSCs of the facility.
(4) For sites at which multiple nuclear reactors may be built or
installed under a construction permit under this part, the application
must--
(i) specify limitations on and provide an analysis of the number
and configuration of nuclear reactors that may be in various stages of
construction, operation, shutdown, and decommissioning at any time from
the commencement of construction of the first reactor to the
termination of the last operating license;
(ii) include an assessment of potential hazards to safety-related
SSCs of the operating reactors at the site posed by activities related
to the construction, operation, and decommissioning of other reactors
at the site;
(iii) include a description of the portions of the nuclear plant
that will be shared by multiple reactors over the lifetime of the plant
and specify functional requirements and measures to meet the
requirements for any safety-related SSCs of the nuclear plant that will
be shared by multiple reactors over the lifetime of the plant; and
(iv) include technical specifications in accordance with Sec.
57.60(a)(8)(vi), as appropriate, for the portions of the nuclear plant
that will be shared with one or more other reactors over the lifetime
of the plant.
(5) Information relating to current and projected population
distributions in the surrounding area and applicable site evaluation
factors for seismic, meteorological, hydrologic, and geologic
characteristics with appropriate consideration of natural phenomena,
including, as applicable, information demonstrating that the site
characteristics are bounded by the site parameters postulated for the
design.
(6) An evaluation of the safety-related SSCs of the facility, with
emphasis upon performance requirements; the bases, with their technical
justifications, upon which such requirements have been established; and
the evaluations required to show that safety functions will be
accomplished. The evaluation must be sufficient to permit understanding
of the system designs and their relationship to safety analyses.
(7) The kinds and quantities of radioactive materials expected to
be produced by operation of the nuclear reactor or nuclear plant and
the means for controlling and limiting radioactive effluents and
radiation exposures within the limits set forth in part 20 of this
chapter, including:
(i) 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 operations;
(ii) 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 operations; and
(iii) 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.
(8) Information related to operational programs concerning facility
operation. Implementation milestones for each operational program must
be described depending on whether the program will be implemented all
at once or on a phased basis. Programs concerning facility operations
include:
(i) The applicant's organizational structure, allocations of
responsibilities and authorities, personnel qualifications and training
requirements, and conduct of operations.
(ii) Plans for preoperational testing and initial operations.
(iii) Plans for conduct of normal operations, including
maintenance, surveillance, and periodic testing of safety-related SSCs.
(iv) An emergency plan for responding to events that could lead to
an accidental release or loss of control of radioactive material, and
to any associated hazards directly incident thereto. Each applicant and
licensee under this part must coordinate response needs with local
emergency planning and offsite response organizations. The applicant
must provide the offsite response organizations that are expected to
respond in an emergency with the opportunity to provide input on the
emergency plan before submitting it to the NRC. The application must
contain any input on the emergency plan received from offsite
organizations.
(v) Security programs.
(A) Physical Security.
(1) Each applicant and licensee under this part must implement
security requirements for the protection of special nuclear material
based on the type, enrichment, and quantity in accordance with part 73
of this chapter, as applicable.
(2) Each applicant and licensee under this part must implement
security requirements for the protection of Category 1 and Category 2
quantities of radioactive material in accordance with part 37 of this
chapter, as applicable.
(3) Each applicant and licensee under this part must implement
security requirements for radiological sabotage
[[Page 23733]]
set forth in subpart J of this part, unless the applicant and licensee
demonstrates that the radiological consequences from a design basis
threat-initiated event do not exceed the dose reference values defined
in Sec. 50.34(a)(1)(ii)(D)(1) of this chapter. To satisfy this
requirement, the design must be assessed against the design basis
threat of radiological sabotage as stated in Sec. 73.1 of this
chapter. The analysis must assume that licensee mitigation and recovery
actions, including any operator actions, are unavailable or
ineffective.
(B) Cybersecurity. Each applicant and licensee under this part must
develop, implement, and maintain a cybersecurity program under Sec.
73.54 or Sec. 73.110 of this chapter.
(C) Information Security. Each applicant and licensee under this
part must develop, implement, and maintain an information protection
system under Sec. Sec. 73.21, 73.22, and 73.23 of this chapter, as
applicable.
(D) Access Authorization. Each applicant and licensee under this
part must establish, implement, and maintain an access authorization
program under Sec. 73.56 of this chapter and must describe the program
in the physical security plan.
(vi) Proposed technical specifications prepared in accordance with
the requirements of Sec. 50.36 of this chapter.
(vii) As applicable, procedures to be used to provide assurance
that the limiting conditions for operation of any operating reactor
will not be exceeded as a result of activities associated with
construction of additional reactors at the same site.
(viii) The radiation protection program.
(ix) The fire protection program.
(A) Each application must include a fire protection plan that
describes the overall fire protection program for the facility;
identifies the various positions within the licensee's organization
that are responsible for the program; states the authorities that are
delegated to each of these positions to implement those
responsibilities; and outlines the plans for fire protection, fire
detection and suppression capability, and limitation of fire damage.
(B) The fire protection plan must also describe specific features
necessary to implement the program described in paragraph (a)(8)(ix)(A)
of this section such as the following: administrative controls and
personnel requirements for fire prevention and manual fire suppression
activities; training requirements for any fire brigade members;
automatic and manually operated fire detection and suppression systems,
as appropriate; and the means to limit fire damage to safety-related
SSCs.
(C) The fire protection plan must include an analysis to
demonstrate that a fire or explosion in any plant area would not
prevent safety-related SSCs from fulfilling safety functions.
(D) Safety-related SSCs must be designed, located, and maintained
to minimize, consistent with other safety requirements, the likelihood
and effect of fires and explosions.
(E) Noncombustible and fire-resistant materials must be used
wherever practical in locations with safety-related SSCs.
(F) Fire detection and fire suppression systems of appropriate
capacity and capability must be provided and designed and maintained to
minimize the adverse effects of fires on safety-related SSCs.
(G) Fire suppression systems must be designed and maintained to
ensure that their rupture or inadvertent operation does not
significantly impair the ability of safety-related SSCs to perform
their safety functions.
(H) Fire detection and fire suppression systems must also consider
and address, as appropriate, any impact from collocated facilities
within the site boundary.
(x) A description of how the human factors engineering requirements
of Sec. 57.395 are addressed and the training, examination, and
proficiency programs necessary to meet the requirements of subpart P of
this part.
(xi) As applicable, a description and plans for implementation of a
remote operation or monitoring program.
(xii) Program(s), and their implementation, necessary to ensure
that the systems and components meet the requirements in the codes or
standards identified in the application in accordance with Sec.
57.60(a)(9).
(xiii) The program, and its implementation, for the environmental
qualification of safety-related electric equipment.
(xiv) A fitness-for-duty program under part 26 of this chapter.
(xv) A staffing plan and supporting analysis in accordance with
Sec. 57.395(c).
(xvi) If the applicant seeks, with its application, approval of a
plan for storage of irradiated fuel after termination of an operating
license, then a plan that demonstrates compliance with all applicable
irradiated fuel possession, safety, and environmental requirements;
includes a plan for funding the management of the fuel; and addresses,
as applicable, transport of the fuel to a designated storage site.
(xvii) If the applicant seeks, with its application, approval of a
decommissioning plan, then a decommissioning plan prepared using the
framework of Sec. 50.82(b)(4) of this chapter, limited to those
provisions applicable to the design characteristics of the nuclear
reactor or nuclear plant, that addresses, as applicable, transport of a
nuclear reactor to a designated facility for final decommissioning,
final decommissioning of individual nuclear reactors, and final
decommissioning of the entire nuclear plant, and ensures compliance
with all applicable safety and environmental requirements.
(xviii) Managerial and administrative controls to be used to assure
safe operation.
(9) Information related to the use of codes and standards. In the
case that generally recognized consensus codes or standards are used
and applied to the design of the facility, they must be named and
evaluated for applicability, adequacy, and sufficiency. Justification
must be provided if they are to be supplemented or modified in keeping
with the safety importance of the function to be performed. Criteria
from these consensus codes or standards must be clearly stated and must
be shown to provide the appropriate level of reliability, safety, and
performance capability. The applicability of these criteria must be
determined from the safety assessment.
(10) The analyses and descriptions of the equipment and systems for
combustible gas control required by Sec. 50.44(d) of this chapter.
(11) The technical qualifications of the applicant to engage in the
proposed activities in accordance with the regulations of this chapter.
(12) A description of the design-specific risk analysis methods
applied to demonstrate adequate defense in depth and safety margin and
the results of the analysis; and
(13) Information demonstrating how the applicant will comply with
requirements for criticality accidents in Sec. 50.68 of this chapter,
except that the maximum nominal U-235 enrichment of the fresh fuel
assemblies is limited to less than twenty (20.0) weight percent for the
purposes of Sec. 50.68(b)(7).
(b) Environmental information. Each application must include
information justifying application of a categorical exclusion, or if a
categorical exclusion is not applicable, an environmental report or
applicant-prepared environmental assessment or environmental impact
statement, in accordance with part 51 of this chapter.
(c) Request for generic finality. An applicant may include in its
joint application a request that the Commission afford generic
finality, in
[[Page 23734]]
accordance with Sec. 57.142(e), to the construction permit, associated
operating license(s), or both. The joint application must include 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 Sec. 57.60(a)(8), consistent
with the scope of the request for generic finality.
(d) Large designated areas. If a joint application for a
construction permit and associated operating license(s) under this part
designates large geographical areas within which the applicant proposes
to construct and operate one or more utilization facilities specified
in the application, then it must include the following additional
information:
(1) Under Sec. 57.60(a)(1), descriptions and safety assessments of
the designated areas, including maps showing the boundaries of the
areas;
(2) Under Sec. 57.60(a)(4), any restrictions on the relative
specific locations of the nuclear reactors proposed in the application
within a designated area;
(3) Under Sec. 57.60(a)(5), information covering the entirety of
the designated areas, including information demonstrating that the site
parameters postulated for the design bound the maximum values for site
evaluation factors within the designated areas;
(4) A plan for storage of irradiated fuel after termination of an
operating license as described in Sec. 57.60(a)(8)(xvi);
(5) A decommissioning plan as described in Sec. 57.60(a)(8)(xvii);
(6) A procedure covering activities that will be conducted in
connection with constructing each utilization facility and placing it
into operation at a specific location, including considerations related
to Sec. 57.60(a)(8)(vii) and NRC inspections required by Sec.
57.100(b)(1);
(7) A procedure that describes how the applicant will 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 Sec. 57.4, prior to beginning construction; and
(8) Under Sec. 57.60(b), information in the environmental report
or applicant-prepared environmental assessment or environmental impact
statement, required by part 51 of this chapter pertaining to the
entirety of the designated areas, as appropriate.
Sec. 57.80 Standards for review of applications.
(a) Applications filed under this part will be reviewed according
to the standards set out in this part and 10 CFR parts 20, 50, 51, 54,
70, 71, 72, 73, 74, and 140, as applicable.
(b) The Commission must perform an environmental review during
review of the application in accordance with the applicable provisions
of subpart K of this part and 10 CFR part 51.
Sec. 57.90 Common standards for licenses.
In determining that a construction permit or operating license in
this part will be issued to an applicant, the Commission will be guided
by the following considerations:
(a) The processes to be performed, the operating procedures, the
facility and equipment, the use of the facility, and other technical
specifications, or the proposals, in regard to any of the foregoing
collectively provide reasonable assurance that the applicant will
comply with the regulations in this chapter, including the regulations
in part 20 of this chapter, and that the health and safety of the
public will not be endangered.
(b) The applicant for a construction permit and operating license
is technically and financially qualified to engage in the proposed
activities in accordance with the regulations in this chapter. However,
no consideration of financial qualification is necessary for an
electric utility applicant for an operating license under this part.
(c) The issuance of a construction permit or operating license to
the applicant will not, in the opinion of the Commission, be inimical
to the common defense and security or to the health and safety of the
public.
(d) Any applicable requirements of 10 CFR part 51 have been
satisfied.
(e) In determining whether a construction permit or operating
license will be issued to an applicant, the Commission will consider
whether the proposed activities will serve a useful purpose
proportionate to the quantities of special nuclear material or source
material to be utilized.
(f) Upon determination that an application for a license meets the
standards and requirements of the AEA and regulations, and that
notifications, if any, to other agencies or bodies have been duly made,
the Commission will issue a construction permit or operating license in
such form and containing such conditions and limitations including
technical specifications, as it deems appropriate and necessary.
(g) An applicant for an operating license or an amendment of an
operating license who proposes to construct or materially alter a
utilization facility will be initially granted a construction permit if
the application is in conformity with and acceptable under the criteria
of Sec. Sec. 57.55 through 57.80, and the standards of this section as
applicable.
(h) A construction permit under this part for the construction of
one or more utilization facilities will be issued before the issuance
of any license to operate a utilization facility if the application is
otherwise acceptable. The construction permit will be converted into
one or more operating licenses upon the completion of construction and
Commission action. A construction permit for a material alteration of a
utilization facility will be issued before the issuance of an amendment
of a license, if the application for amendment is otherwise acceptable,
as provided in Sec. 57.310.
(i) In the case of a construction permit or operating license under
this part for a facility for the generation of commercial power:
(1) The NRC will--
(i) Give notice in writing of each application to the regulatory
agency or State as may have jurisdiction over the rates and services
incident to the proposed activity;
(ii) Publish notice of the application in trade or news
publications as it deems appropriate to give reasonable notice to
municipalities, private utilities, public bodies, and cooperatives
which might have a potential interest in the utilization facility; and
(iii) Publish notice of the application once each week for four
consecutive weeks in the Federal Register. No license will be issued by
the NRC prior to the giving of these notices and until four weeks after
the last notice is published in the Federal Register.
(2) If there are conflicting applications for a limited opportunity
for such license, the Commission will give preferred consideration in
the following order: first, to applications submitted by public or
cooperative bodies for facilities to be located in high cost power
areas in the United States; second, to applications submitted by others
for facilities to be located in such areas; third, to applications
submitted by public or cooperative bodies for facilities to be located
in areas other than high cost power areas; and, fourth, to all other
applicants.
(3) The licensee who transmits electric energy in interstate
commerce, or sells it at wholesale in interstate commerce, will be
subject to the regulatory provisions of the Federal Power Act.
[[Page 23735]]
(4) Nothing must preclude any government agency, now or hereafter
authorized by law to engage in the production, marketing, or
distribution of electric energy, if otherwise qualified, from obtaining
a construction permit or operating license under this part for a
utilization facility for the primary purpose of producing electric
energy for disposition for ultimate public consumption.
Sec. 57.95 Issuance of construction permit.
(a) After conducting a hearing in accordance with Sec. 57.130 and
receiving the report submitted by the ACRS, the Commission may issue a
construction permit if the Commission finds that:
(1) The applicable standards and requirements of the AEA and the
Commission's regulations have been met;
(2) Any required notifications to other agencies or bodies have
been duly made;
(3) There is reasonable assurance that the facility will be
constructed in conformity with the construction permit, the provisions
of the AEA, and the Commission's regulations.
(4) The applicant is technically and financially qualified to
engage in the activities authorized;
(5) Issuance of the construction permit will not be inimical to the
common defense and security or to the health and safety of the public;
and
(6) The findings required by part 51 of this chapter have been
made.
(b) A construction permit will constitute an authorization to the
applicant to proceed with construction but will not constitute
Commission approval of the operational programs or requirements, other
than those material to the adequacy of the design, unless the applicant
specifically requests such approval and such approval is incorporated
in the construction permit. The applicant, at its option, may request
such approvals in the construction permit or, from time to time, by
amendment of its construction permit. The Commission may, in its
discretion, incorporate in any construction permit provisions requiring
the applicant to furnish periodic reports of the progress.
(c) Any construction permit must state the earliest and latest
dates for the completion of the construction of each nuclear reactor or
modification authorized by the permit.
Sec. 57.100 Issuance of operating license.
(a) Upon completion of the construction or material alteration of a
facility, in compliance with the terms and conditions of the
construction permit and subject to any necessary testing of the
facility for health or safety purposes, the Commission will, in the
absence of good cause shown to the contrary, issue an operating license
or an appropriate amendment of the license, as the case may be.
(b) An operating license may be issued by the Commission, up to the
full term authorized by Sec. 57.105(a), upon finding that:
(1) Construction of the facility has been substantially completed,
in conformity with the construction permit and the application as
amended, the provisions of the AEA, and the rules and regulations of
the Commission;
(2) The facility will operate in conformity with the application as
amended, the provisions of the AEA, and the rules and regulations of
the Commission;
(3) There is reasonable assurance that the activities authorized by
the operating license can be conducted without endangering the health
and safety of the public and will be conducted in compliance with the
regulations in this chapter;
(4) The applicant is technically and financially qualified to
engage in the activities authorized by the operating license in
accordance with the regulations in this chapter. However, no finding of
financial qualification is necessary for an electric utility applicant
for an operating license for a utilization facility;
(5) The applicable provisions of part 140 of this chapter have been
satisfied; and
(6) The issuance of the operating license will not be inimical to
the common defense and security or to the health and safety of the
public.
(c) Each operating license will include appropriate provisions with
respect to any uncompleted items of construction and such limitations
or conditions as are required to ensure that operation during the
period of the completion of such items will not endanger public health
and safety.
(d) An applicant may, in a case where a hearing is held in
connection with a pending proceeding under this section make a motion
in writing, under this paragraph, for an operating license authorizing
low power testing, and further operations less than full power
operation. Action on such a motion by the presiding officer will be
taken with due regard to the rights of the parties to the proceedings,
including the right of any party to be heard to the extent that his
contentions are relevant to the activity to be authorized. Before
taking any action on such a motion that any party opposes, the
presiding officer must make findings on the matters specified in
paragraph (b) of this section as to which there is a controversy, in
the form of an initial decision with respect to the contested activity
sought to be authorized. The Director of Nuclear Reactor Regulation
will make findings on all other matters specified in paragraph (b) of
this section. If no party opposes the motion, the presiding officer
will issue an order in accordance with Sec. 2.319(p) of this chapter
authorizing the Director of Nuclear Reactor Regulation to make
appropriate findings on the matters specified in paragraph (b) of this
section and to issue a license for the requested operation.
(e) Each operating license for a nuclear reactor issued under this
part that references a manufacturing license issued under subpart D of
this part must include, as applicable, a condition that--
(1) The authorization to operate the reactor is suspended while the
features to prevent criticality described in the manufacturing license
are in place; and
(2) Removal of the features to prevent criticality may not be
initiated unless--
(i) All conditions of an operating license under this part are met,
or
(ii) The reactor has been defueled in accordance with an
appropriate license issued by the Commission.
(f) The operating license for a nuclear reactor that is part of a
nuclear plant at which portions of the nuclear plant will be shared by
multiple reactors over the lifetime of the plant as described in Sec.
57.60(a)(4)(iii), must include a condition specifying that the shared
portions of the plant are part of the facility as described in the
operating license's final safety analysis report and any related
technical specifications under Sec. 57.60(a)(4)(iv) are incorporated
in the license.
Sec. 57.105 Continuation of license.
(a) Each construction permit and operating license will be issued
for a fixed period of time to be specified in the license but in no
case to exceed 40 years from date of issuance. Where the operation of a
facility is involved, the Commission will issue the operating license
for the term requested by the applicant or for the estimated useful
life of the nuclear reactor or nuclear plant if the Commission
determines that the estimated useful life is less than the term
requested. Licenses may be renewed by the Commission upon the
expiration of the period. Renewal of operating licenses requirements
are provided in Sec. 57.115 and Sec. 57.120. Application for
termination of license is to be made pursuant to Sec. 57.305.
(b) Each operating license for a facility that has permanently
ceased operations,
[[Page 23736]]
continues in effect beyond the expiration date to authorize ownership
and possession of the facility, until the Commission notifies the
licensee in writing that the operating license is terminated. During
such period of continued effectiveness the licensee must--
(1) Take actions necessary to decommission and decontaminate the
facility and continue to maintain the facility, including, where
applicable, the storage, control and maintenance of irradiated fuel, in
a safe condition, and
(2) Conduct activities in accordance with all other restrictions
applicable to the facility in accordance with the NRC regulations and
the provisions of the specific license for the facility.
Sec. 57.110 Transfer of licenses.
(a) No construction permit or license under this part, or any right
thereunder, may be transferred, assigned, or in any manner disposed of,
either voluntarily or involuntarily, directly or indirectly, through
transfer of control of the license or construction permit to any
person, unless the Commission gives its consent in writing.
(b) Contents of license transfer applications.
(1) An application for transfer of a license or construction permit
must include:
(i) For a construction permit or operating license under this part,
as much of the information described in Sec. Sec. 57.55 and 57.60 with
respect to the identity and technical and financial qualifications of
the proposed transferee as would be required by those sections if the
application were for an initial construction permit or license.
(ii) For a manufacturing license under this part, as much of the
information described in Sec. Sec. 57.150 and 57.155 with respect to
the identity and technical qualifications of the proposed transferee as
would be required by those sections if the application were for an
initial license.
(2) For a construction permit or operating license under this part,
the Commission may require additional information such as data
respecting proposed safeguards against hazards from radioactive
materials and the applicant's qualifications to protect against such
hazards.
(3) The application must include a statement of the purposes for
which the transfer of the construction permit or license is requested,
the nature of the transaction necessitating or making desirable the
transfer, and an agreement to limit access to Restricted Data and
classified National Security Information pursuant to Sec. 57.15. The
Commission may require any person who submits an application for a
construction permit or license pursuant to the provisions of this
section to file a written consent from the existing licensee or a
certified copy of an order or judgment of a court of competent
jurisdiction attesting to the person's right (subject to the licensing
requirements of the AEA and these regulations) to possession of the
facility or site involved.
(c) After appropriate notice to interested persons, including the
existing licensee, and observance of such procedures as may be required
by the AEA or regulations or orders of the Commission, the Commission
will approve an application for the transfer of a construction permit
or license, if the Commission determines:
(1) That the proposed transferee is qualified to be the holder of
the license; and
(2) That transfer of the construction permit or license is
otherwise consistent with applicable provisions of law, regulations,
and orders issued by the Commission pursuant thereto.
Sec. 57.115 Application for renewal.
(a) The filing of an application for renewal must be in accordance
with subpart A of part 2 of this chapter, Sec. 57.4, and Sec. 57.7.
(b) Each application for renewal must include the information
described in Sec. 57.55(a) through (e), (g), and (h).
(c) Each application must include any information required by 10
CFR part 51.
(d) Each application must include any technical specification
changes or additions necessary to manage the effects of aging during
the period of extended operation as part of the renewal application.
The justification for changes or additions to the technical
specifications must be contained in the operating license renewal
application.
(e) Each application for renewal must include technical information
as follows:
(1) Identify safety-related SSCs subject to an aging management
review, excluding those that are not subject to replacement based on a
qualified life or specified time period.
(2) For each safety-related SSC identified in paragraph (e)(1) of
this section, demonstrate that the effects of aging will be adequately
managed so that the intended safety function(s) will be maintained
consistent with the licensing basis for the period of extended
operation.
(3) At least 3 months before scheduled completion of the NRC
review, an amendment to the renewal application must be submitted that
identifies any change to the licensing basis of the facility that
materially affects the contents of the license renewal application,
including the FSAR supplement.
(4) A list of time-limited aging analyses to demonstrate the
following:
(i) The analyses remain valid for the period of extended operation;
(ii) The analyses have been projected to the end of the period of
extended operation; or
(iii) The effects of aging on the safety function(s) will be
adequately managed for the period of extended operation.
(5) An FSAR supplement for the facility that contains a summary
description of the programs and activities for managing the effects of
aging and the evaluation of time-limited aging analyses for the period
of extended operation.
(f) A notice of an opportunity for a hearing will be published in
the Federal Register in accordance with 10 CFR 2.105 and 2.309. In the
absence of a request for a hearing filed within 60 days by a person
whose interest may be affected, the Commission may issue a renewed
operating license without a hearing upon a 30-day notice and
publication in the Federal Register of its intent to do so.
Sec. 57.120 Criteria for renewal.
A renewed license may be issued by the Commission up to the full
term authorized by Sec. 57.135 if the Commission finds that:
(a) Actions have been identified and have been or will be taken
with respect to the matters identified in paragraphs (a)(1) and (a)(2)
of this section, such that there is reasonable assurance that the
activities authorized by the renewed license will continue to be
conducted in accordance with the current licensing basis, and that any
changes made to the plant's current licensing basis in order to comply
with this paragraph are in accord with the AEA and the Commission's
regulations. These matters are:
(1) Managing the effects of aging during the period of extended
operation on the functionality of structures and components that have
been identified to require review under Sec. 57.115(e)(1); and
(2) Time-limited aging analyses that have been identified to
require review under Sec. 57.115(e)(4).
(b) Any applicable requirements of 10 CFR part 51 have been
satisfied.
(c) Any matters raised under 10 CFR 2.335 have been addressed.
Sec. 57.130 Hearings.
(a) A notice of an opportunity for a hearing will be published in
the Federal Register in accordance with 10 CFR
[[Page 23737]]
2.105 and 2.309 for each application for a renewed operating license.
In the absence of a request for a hearing filed within 30 days by a
person whose interest may be affected, the Commission may issue a
renewed operating license or without a hearing upon a 30-day notice and
publication in the Federal Register of its intent to do so.
(b) Hearings procedure.
(1) The Commission will hold a hearing after at least 30 days'
notice and publication once in the Federal Register on each application
for a construction permit filed under this part.
(2) When an application is made for an amendment to a construction
permit or operating license, the Commission may hold a hearing after at
least 30 days' notice and publication once in the Federal Register, or,
in the absence of a request therefor by any person whose interest may
be affected, may issue an amendment to a construction permit or
operating license without a hearing, upon 30 days' notice and
publication once in the Federal Register of its intent to do so.
(3) If the Commission finds, in an emergency situation, as defined
in Sec. 50.91 of this chapter, that no significant hazards
consideration is presented by an application for an amendment to an
operating license, it may dispense with public notice and comment and
may issue the amendment. If the Commission finds that exigent
circumstances exist, as described in Sec. 50.91, it may reduce the
period provided for public notice and comment.
(4) Both in an emergency situation and in the case of exigent
circumstances, the Commission will provide 30 days' notice of
opportunity for a hearing, though this notice may be published after
issuance of the amendment if the Commission determines that no
significant hazards consideration is involved.
(5) The Commission will use the standards in subpart H of this part
to determine whether a significant hazards consideration is presented
by an amendment to an operating license and may make the amendment
immediately effective, notwithstanding the pendency before it of a
request for a hearing from any person, in advance of the holding and
completion of any required hearing, where it has determined that no
significant hazards consideration is involved.
(6) No petition or other request for review of or hearing on the
staff's significant hazards consideration determination will be
entertained by the Commission. The staff's determination is final,
subject only to the Commission's discretion, on its own initiative, to
review the determination.
(7) If an applicant requests generic finality under Sec. 57.60(c),
then the Commission will include a request for generic finality as a
proposed action in the joint notice of hearing and proposed action
under Sec. Sec. 2.104 and 2.105 of this chapter.
Sec. 57.135 Duration of renewal.
A renewed license will be issued for a fixed period of time, which
is the sum of the additional amount of time beyond the expiration of
the operating license that is requested in a renewal application plus
the remaining number of years on the operating license currently in
effect. The term of any renewed license may not exceed 40 years.
Sec. 57.142 Finality for construction permits and operating licenses.
(a) Notwithstanding any provision in Sec. 57.16, during the term
of a construction permit or operating license issued under this part,
the Commission may not modify, rescind, or impose new requirements on
the terms and conditions of the construction permit or operating
license afforded generic finality pursuant to paragraph (e) of this
section, unless the Commission determines that a modification is
necessary to bring the construction permit or operating license into
compliance with the Commission's requirements applicable and in effect
at the time the construction permit or operating license was issued, or
to provide reasonable assurance of adequate protection to public health
and safety or common defense and security.
(b) In the proceedings for issuance of a construction permit or
operating license, or in any enforcement hearing other than one
initiated by the Commission under paragraph (a) of this section, in
which a construction permit or operating license issued under this
subpart is referenced, the Commission must treat as resolved those
matters resolved in the proceeding on the application for issuance or
renewal of the referenced construction permit or operating license,
including, if applicable, the adequacy of a reactor design and any
generic aspects of operational programs or requirements, where the
referenced construction permit or operating license was afforded
finality pursuant to paragraph (e) of this section.
(c) The holder of a construction permit or operating license
afforded generic finality pursuant to paragraph (e) of this section may
make changes to the facility or procedures as described in the FSAR
associated with the construction permit or operating license without
obtaining a license amendment pursuant to Sec. 57.310 if the change
meets the criteria in Sec. 50.59(c) of this chapter. If the change
does not meet the criteria in Sec. 50.59(c) of this chapter, then the
request for a change must be in the form of an application for a
license amendment under Sec. 57.310.
(d) Except for information requests seeking to verify compliance
with the current licensing basis of the construction permit or
operating license, the NRC must prepare the reason or reasons for each
information request to the holder of a construction permit or operating
license under this part before issuance to ensure that the burden to be
imposed on respondents is justified in view of the potential safety
significance of the issue to be addressed in the requested information.
Each such justification provided for an evaluation performed by the NRC
staff must be approved by the Executive Director for Operations or
designee before issuance of the request.
(e) 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 aspects of operational programs
and requirements submitted pursuant to 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 aspects of
operational programs and requirements, without undue risk to the health
and safety of the public.
Subpart D--Manufacturing Licenses.
Sec. 57.145 Scope.
This subpart sets out the requirements and procedures applicable to
Commission issuance of a license under this part authorizing the
manufacture of manufactured reactors. This subpart also sets out
requirements for manufacturing, loading fuel into, and transportation
of manufactured reactors.
Sec. 57.150 Contents of applications for manufacturing licenses;
general information.
Each application for a manufacturing license under this part must
include the information required by Sec. 57.55(a) through (e) and (j).
Sec. 57.155 Contents of applications; technical information in final
safety analysis report.
The application must include a final safety analysis report
containing the
[[Page 23738]]
information set forth below, with a level of design information
sufficient to enable the Commission to judge the applicant's proposed
means of assuring that the manufacturing conforms to the design of the
reactor to be manufactured and to reach a final conclusion on all
safety questions associated with the design, permit the preparation of
construction and installation specifications by an applicant who seeks
to use the manufactured reactor, and permit the preparation of
acceptance and inspection requirements by the NRC. The application must
include the following information:
(a) Other than site-specific information, the information required
by Sec. 57.60(a)(1) through (3), (6), (7), and (9) through (12)
relevant to the manufactured reactor;
(b) The site parameters postulated for the design of the reactor to
be manufactured under this subpart, 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
(c) Information necessary to establish that the design of the
reactor to be manufactured under this subpart complies with the
technical requirements in 10 CFR chapter I, including:
(1) A description and analysis of the fire protection design
features for the manufactured reactor necessary to comply with Sec.
57.60(a)(8)(ix)(B);
(2) Information demonstrating how the applicant will comply with
requirements for criticality accidents in Sec. 50.68(b)(2) through (4)
of this chapter;
(3) The information required by Sec. 20.1406 of this chapter;
(4) The technical qualifications of the applicant to engage in the
proposed activities in accordance with the regulations in this chapter;
(5) Proposed technical specifications applicable to the
manufactured reactor, prepared in accordance with the requirements of
Sec. 57.60(a)(8)(vi);
(6) The interface requirements between the manufactured reactor and
the remaining portions of the nuclear plant or connections to other
facilities outside of the nuclear plant. These requirements must be
sufficiently detailed to allow for applicants for construction permits
and operating licenses under this part that reference the manufactured
reactor to complete the final safety analysis;
(7) A representative conceptual design for a nuclear plant using
the manufactured reactor, to aid the NRC in its review of the final
safety analysis report required by this section and to permit
assessment of the adequacy of the interface requirements;
(8) As an alternative to the information required by paragraphs
(c)(6) and (7) of this section, the application may include all non-
site-specific information on the remaining portions of the nuclear
plant that would be included in a joint application for a construction
permit and associated operating license(s) under subpart C of this
part;
(9) Justification that compliance with the interface requirements
of paragraph (c)(6) of this section or the information provided under
paragraph (c)(8) is verifiable through inspections, testing, or
analysis; and
(10) Unless the application includes essentially complete plans for
preoperational testing and initial operation under Sec. 57.160(a),
necessary parameters to be used in developing such plans.
Sec. 57.160 Contents of applications; additional information.
(a) An applicant may include in its application descriptions of
generic operational programs and requirements of the types described in
Sec. 57.60(a)(8). The NRC may afford finality to such programs in
accordance with Sec. Sec. 57.16 and 57.175.
(b) The application must include information justifying application
of a categorical exclusion or, if a categorical exclusion is not
applicable, an environmental report or applicant-prepared environmental
assessment, in accordance with part 51 of this chapter.
(c) The application must contain a description of the program to
protect safeguards information against unauthorized disclosure in
accordance with the requirements in Sec. Sec. 73.21 and 73.22 of this
chapter, as applicable.
(d) The application must include the following information related
to the manufacturing processes, organization, controls, and
inspections:
(1) A description, including references to relevant codes and
standards, of the processes that will be used to procure, fabricate,
and assemble components that make up the manufactured reactor. The
description must clearly define which activities are proposed to be
within the scope of the manufacturing license and those, such as the
making of a component to be procured from a separate company for
installation in the manufactured reactor, that are not considered to be
within the scope of the manufacturing license;
(2) A description of the organizational and management structure
singularly responsible for direction of the design and manufacture of
the manufactured reactor. The information should include a description
of the management plans, technical qualifications, and controls in
place to demonstrate compliance with the requirements of Sec. 57.197.
(3) A description of the inspections and tests to be performed as
part of the manufacturing process, including the inspection of procured
components, inspection and testing of fabrication processes, and
inspections and testing of the assembled manufactured reactor;
(4) A description of the fitness-for-duty program required by part
26 of this chapter and its implementation.
(e) The application must include a description of the following
information related to the deployment of a manufactured reactor:
(1) Procedures governing the preparation of the manufactured
reactor or portions of the manufactured reactor for shipping to the
site where it is to be operated; the conduct of shipping; and verifying
the condition of the shipped items upon receipt at the site;
(2) Details of the interaction of the design, manufacture, and
installation of a manufactured reactor within the applicant's
organization and how the applicant will ensure integration between the
designer, contractors, and any facility in which the manufactured
reactor is to be installed; and
(3) Measures to be used for the control of interfaces, including
the consideration of significant site parameters, between the holder of
the manufacturing license and the holder of the construction permit for
the nuclear plant at which the manufactured reactor is to be installed.
(f) An application for a manufacturing license for a manufactured
reactor that will be fueled at the manufacturing facility under a 10
CFR part 70 license must include the following information related to
loading fuel and the required features to prevent criticality and to
otherwise provide reasonable assurance that the fueled manufactured
reactor can be transported to and installed at a site for which the
Commission has issued a construction permit that authorizes
construction of a nuclear plant using the manufactured reactor and
operated in accordance with an operating license issued under this
part:
(1) A description of the procedures used during the fueling of the
manufactured reactor that ensure that the configuration of fuel within
the fueled manufactured reactor is consistent with the design and
analyses supporting operation of the manufactured reactor under the
operating license at the place of operation. The description may
reference the applicable 10 CFR part 70
[[Page 23739]]
application and other sections of the final safety analysis report
supporting the manufacturing license application.
(i) The application must describe the measures taken for
inspections and non-nuclear testing performed to ensure that the
configuration of fuel within the fueled manufactured reactor is
consistent with the design and analyses supporting operation of the
manufactured reactor under the operating license at the place of
operation.
(ii) The application must describe the design features included in
the manufactured reactor to prevent criticality, the associated
functional design criteria applied to those design features, and the
physical and programmatic controls implemented during manufacturing,
storage, and transport that are credited to ensure the features
function as designed when subject to potential hazards and human
errors. The descriptions must include how those measures will be
controlled during installation under the manufacturing license and
removal under the operating license at the place of operation.
(2) A description of the procedures governing the transfer of
responsibilities for the fueled manufactured reactor from the holder of
the manufacturing license to the holder of the construction permit for
the installation site.
(3) If available at the time of filing the manufacturing license
application or, if not available at the time of filing the
manufacturing license application, submitted as an amendment to the
manufacturing license or manufacturing license application at the time
of filing the 10 CFR part 70 application, a description of the programs
needed to demonstrate compliance with the requirements of Sec.
57.197(d) and 10 CFR parts 70, 71, and 73 for the receipt, storage, and
loading of SNM into a manufactured reactor and the transport of the
fueled manufactured reactor to a site for which the Commission has
issued a construction permit that authorizes construction of a nuclear
plant using the manufactured reactor, including the following:
(i) A physical security program in accordance with Sec.
57.197(d)(3)(i).
(ii) A cybersecurity program in accordance with Sec.
57.197(d)(3)(i).
Sec. 57.165 Standards for review of applications.
Applications for manufacturing licenses under this part will be
reviewed according to the applicable standards set out in this subpart
as well as applicable standards in 10 CFR parts 20, 25, 26, 50, 51, 57,
70, 71, 73, and 75.
Sec. 57.170 Administrative review of applications; hearings.
A proceeding on a manufacturing license under this part is subject
to all applicable procedural requirements contained in 10 CFR part 2,
including the requirements for docketing in Sec. 2.101(a)(1) through
(4) of this chapter, and the requirements for issuance of a notice of
proposed action in Sec. 2.105 of this chapter, provided, however, that
the designated sections may not be construed to require that the
environmental report or applicable environmental review by the NRC
include an assessment of the benefits of constructing and/or operating
the manufactured reactor or an evaluation of alternative energy
sources. All hearings on manufacturing licenses are governed by the
hearing procedures contained in 10 CFR part 2, subparts C, E, G, L, and
N.
Sec. 57.172 Issuance of manufacturing license.
(a) After completing any hearing under Sec. 57.170, and receiving
the report submitted by the ACRS under Sec. 57.17, the Commission may
issue a manufacturing license if the Commission finds that:
(1) Applicable standards and requirements of the AEA and the
Commission's regulations have been met;
(2) There is reasonable assurance that the manufactured reactor
will be manufactured, and can be transported, incorporated into a
nuclear plant, and operated in conformity with the manufacturing
license, the provision of the AEA, and the Commission's regulations;
(3) The proposed manufactured reactor can be incorporated into a
nuclear plant, including, as applicable, the nuclear plant described in
the manufacturing license application, and operated at sites having
characteristics that fall within the site parameters postulated for the
design of the manufactured reactors in conformity with the requirements
in subpart B of this part and without undue risk to the health and
safety of the public;
(4) The applicant is technically qualified to design and
manufacture the proposed manufactured reactor;
(5) The proposed parameters to be used in developing plans for
preoperational testing and initial operation, or the essentially
complete plans provided in the application, are necessary and
sufficient, within the scope of the manufacturing license, to provide
reasonable assurance that the manufactured reactor will be manufactured
and operated in conformity with the license, the provisions of the AEA,
and the Commission's regulations;
(6) The generic operational programs and requirements proposed for
the manufactured reactor 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.
(7) The issuance of a manufacturing license to the applicant will
not be inimical to the common defense and security or to the health and
safety of the public; and
(8) The findings required by 10 CFR part 51 have been made.
(b) Each manufacturing license issued under this subpart must
specify:
(1) Terms and conditions as the Commission deems necessary and
appropriate;
(2) Technical specifications for operation of the manufactured
reactor, as the Commission deems necessary and appropriate;
(3) Significant site parameters and significant design
characteristics for the manufactured reactor; and
(4) The interface requirements to be met by the site-specific
elements of the facility not within the scope of the manufactured
reactor, or that the portions of the nuclear plant other than the
manufactured reactor must be as described in the application.
Sec. 57.175 Finality of manufacturing licenses; information requests.
(a) Notwithstanding any provision in Sec. 57.16, during the term
of a manufacturing license issued under this part the Commission may
not modify, rescind, or impose new requirements on the design of the
nuclear reactor being manufactured under the manufacturing license, or
the requirements for the manufacture of the nuclear reactor, unless the
Commission determines that a modification is necessary to bring the
design of the reactor or its manufacture into compliance with the
Commission's requirements applicable and in effect at the time the
manufacturing license was issued, or to provide reasonable assurance of
adequate protection to public health and safety or common defense and
security.
(b) Any modification to the design of a manufactured reactor that
is imposed by the Commission under paragraph (a) of this section will
be applied to all reactors manufactured under the license, including
those that have already been transported and sited, except those
reactors to which the modification has been rendered
[[Page 23740]]
technically irrelevant by action taken under paragraph (d) of this
section.
(c) In the proceedings for issuance of a construction permit or
operating license, or in any enforcement hearing other than one
initiated by the Commission under paragraph (a) of this section, in
which a manufacturing license under this part is referenced, the
Commission must treat as resolved those matters resolved in the
proceeding on the application for issuance or renewal of the
manufacturing license, including the adequacy of design of the
manufactured reactor, the adequacy of the design of the remaining
portions of a nuclear plant described in the manufacturing license
application, and any essentially complete operational programs or
requirements.
(d) The holder of a manufacturing license under this part may make
changes to the facility or procedures as described in the FSAR
associated with the manufacturing license without obtaining a license
amendment pursuant to Sec. 57.310 if the change meets the criteria in
Sec. 50.59(c) of this chapter. If the change does not meet the
criteria in Sec. 50.59(c) of this chapter, then the request for a
change must be in the form of an application for a license amendment
under Sec. 57.310.
(e) Except for information requests seeking to verify compliance
with the current licensing basis of either the manufacturing license or
the manufactured reactor, the NRC must prepare the reason or reasons
for each information request to the holder of a manufacturing license
under this part or an applicant or licensee using a manufactured
reactor before issuance to ensure that the burden to be imposed on
respondents is justified in view of the potential safety significance
of the issue to be addressed in the requested information. Each such
justification provided for an evaluation performed by the NRC staff
must be approved by the Executive Director for Operations or designee
before issuance of the request.
Sec. 57.180 Duration of manufacturing license.
A manufacturing license issued under this subpart may be valid for
up to 40 years from the date of issuance. Upon expiration of the
manufacturing license, the manufacture of any uncompleted reactors must
cease unless a timely application for renewal has been docketed with
the NRC.
Sec. 57.185 Transfer of manufacturing license.
A manufacturing license may be transferred in accordance with Sec.
57.110.
Sec. 57.190 Renewal of manufacturing licenses.
(a) Not less than 12 months, nor more than 5 years before the
expiration of the manufacturing license, or any later renewal period,
the holder of the manufacturing license issued under this part may
apply for a renewal of the license. An application for renewal must
contain all information necessary to bring up to date the information
and data contained in the previous application. The filing of an
application for a renewed license must be in accordance with subpart A
of 10 CFR part 2 of this chapter and Sec. 57.19.
(b) A manufacturing license issued under this part, either original
or renewed, for which a timely application for renewal has been filed,
remains in effect until the Commission has made a final determination
on the renewal application.
(c) Any person whose interest may be affected by renewal of the
license may request a hearing on the application for renewal. The
request for a hearing must comply with Sec. 2.309 of this chapter. If
a hearing is granted, notice of the hearing will be published in
accordance with Sec. 2.104 of this chapter.
(d) The Commission may grant the renewal if the Commission
determines--
(1) The manufacturing license complies with the AEA and the
Commission's regulations and orders applicable and in effect at the
time the manufacturing license was originally issued; and
(2) Any new requirements the Commission may wish to impose are--
(i) Necessary for adequate protection to public health and safety
or common defense and security;
(ii) Necessary for compliance with the Commission's regulations and
orders applicable and in effect at the time the manufacturing license
was originally issued; or
(iii) A substantial increase in overall protection of the public
health and safety or the common defense and security to be derived from
the new requirements, and the direct and indirect costs of
implementation of those requirements are justified in view of this
increased protection.
(e) A renewed manufacturing license may be issued for a term up to
40 years, plus any remaining years on the manufacturing license then in
effect before renewal. The renewed license will be subject to the
requirements of Sec. 57.175.
Sec. 57.197 Manufacturing.
(a) Holders of manufacturing licenses must ensure that the
following plans, programs, and organizational units are developed and
implemented to manage and control the manufacturing activities within
the scope of the manufacturing license:
(1) Programs to ensure that the manufacturing of a reactor complies
with the design and analysis requirements in this part. The entity with
design authority for the manufactured reactor covered by the
manufacturing license must be identified in the license.
(2) An organizational and management structure responsible for
managing, controlling, and evaluating the adequacy of the reactor
design and manufacturing activities.
(3) Procedures describing the qualifications for personnel in key
positions in the licensee's management and control organization and the
organizational responsibilities, authority, and interfaces with other
parts of the licensee's organization.
(4) A fitness-for-duty program, in accordance with part 26 of this
chapter.
(5) A quality assurance program to be applied to the design,
fabrication, construction, and testing of the safety-related SSCs of
the manufactured reactor.
(6) A radiation protection program, in accordance with 10 CFR part
20, that includes measures for monitoring the dose to individuals if
the manufacturing activities include working with radioactive
materials.
(7) An information security program in accordance with Sec. Sec.
73.21, 73.22 and 73.23 of this chapter, as applicable.
(b) Holders of manufacturing licenses must satisfy the following
requirements:
(1) The manufacturing process must be conducted within facilities
for which the manufacturing license holder has the authority to
establish controls on any activity that might affect manufacturing. The
licensee must establish access controls to the portions of each
facility involved in the manufacturing processes governed by the
manufacturing license.
(2) Manufacturing processes must be performed in accordance with
the manufacturing license, including the codes or standards described
in the manufacturing license application under Sec. 57.160(d) and
found acceptable by the NRC.
(3) A post-manufacturing inspection and acceptance process to
verify that manufacturing activities have been completed in accordance
with the manufacturing license must be established and implemented
before transporting a manufactured reactor or portions of a
manufactured reactor for installation at a nuclear plant.
[[Page 23741]]
(c) As appropriate considering the types and quantities of
radioactive materials being brought into the manufacturing facility--
(1) Procedures must be in place to receive, transfer, possess, and
use source, byproduct, and special nuclear material in accordance with
the applicable portions of 10 CFR parts 30, 40 and 70.
(2) A fire protection program must be established and implemented
before the initial receipt of byproduct, source, or non-fuel special
nuclear material (excluding exempt quantities as described in Sec.
30.18 of this chapter).
(3) An emergency plan appropriate for responding to the facility-
specific hazards of an accidental release of radioactive material and
to limit the health effects of the associated chemical hazards of
licensed material must be approved and implemented prior to the receipt
of byproduct, source, or special nuclear material (excluding exempt
quantities as described in Sec. 30.18 of this chapter).
(4) A plant staff training program associated with the receipt of
radioactive material must be approved and implemented before initial
receipt of byproduct, source, or special nuclear material (excluding
exempt quantities as described in Sec. 30.18 of this chapter).
(5) Security requirements must be implemented for the protection of
SNM based on the type, enrichment, and quantity in accordance with 10
CFR part 73, as applicable, and for the protection of Category 1 and
Category 2 quantities of radioactive material in accordance with 10 CFR
part 37, as applicable.
(d) Fuel loading.
(1) The Commission has determined that a fueled manufactured
reactor in which features to prevent criticality are in place is not in
operation.
(i) A holder of a manufacturing license may load fuel into a
manufactured reactor pursuant to a license issued under part 70 of this
chapter only if the manufactured reactor is configured before its fuel
loading and during storage and transport with features to prevent
criticality that are specified in the manufacturing license.
(ii) Upon issuance of an operating license for a nuclear plant that
incorporates the manufactured reactor, the features to prevent
criticality may be removed. Upon initiating the removal of the features
to prevent criticality, the fueled manufactured reactor has commenced
operation.
(2) Holders of 10 CFR part 70 licenses authorizing the possession
and loading of fuel into reactors manufactured under a manufacturing
license issued under this part must comply with the requirements of 10
CFR part 70 for the facilities and activities related to the storage,
movement, and loading of fuel in the manufactured reactors. Procedures,
equipment, and personnel required by the 10 CFR part 70 license must be
in place before the receipt of SNM at the manufacturing facility.
(3) Before the receipt of SNM, the licensee must have security
programs in place that meet the performance objectives of 10 CFR 73.67,
with the following additions and exceptions:
(i) A physical security plan describing the physical security
program must be maintained and a cybersecurity program must be
established for the possession and loading of fresh fuel into a
manufactured reactor authorized by a 10 CFR part 70 license, regardless
of fuel type, enrichment, and quantity.
(ii) The physical security program must be designed to prevent
unintended and uncontrolled criticality events.
(iii) The cybersecurity program must provide reasonable assurance
that a cyberattack would not adversely impact the functions performed
by digital assets necessary for implementing the physical security
requirements of this section, or the radiation monitoring and
criticality requirements in this section or in 10 CFR part 70.
(iv) All holders of a 10 CFR part 70 license that authorizes
loading of fresh fuel into a manufactured reactor must perform the
screening required in Sec. 73.67(d)(4) of this chapter to confirm the
identity, trustworthiness, and reliability of individuals prior to
granting unescorted access to special nuclear material, and these
determinations must be documented.
(4) The loading or unloading of fresh fuel into or from a
manufactured reactor and any changes to the configuration of reactivity
control and prevention systems for the fueled manufactured reactor must
be performed by a certified fuel handler.
(e) Transportation.
(1) A holder of a manufacturing license under this part may not
transport or allow to be removed from the places of manufacture the
reactor manufactured under the manufacturing license except for either
transport to a site for which the Commission has issued a construction
permit that references the subject manufacturing license or for export
in accordance with 10 CFR part 110.
(2) A holder of a manufacturing license must include in any
contract governing the transport of a manufactured reactor or portions
thereof as defined in the manufacturing license from the places of
manufacture to any other location, a provision requiring that the
person transporting the manufactured reactor comply with all shipping
requirements in applicable NRC regulations, certificates of compliance,
and NRC-issued licenses.
(3) Procedures governing the preparation of the manufactured
reactor or portions thereof as defined in the manufacturing license for
transport and the conduct of the transport must be issued prior to
transport. The procedures must implement the protective measures and
restrictions described in NRC regulations and NRC-issued licenses to
protect the reactor from potential conditions that would adversely
affect the safe operation of a nuclear plant.
(4) For a manufactured reactor that is to be loaded with fresh fuel
before transport to the place of operation, the manufacturing license
must specify that transportation will be in accordance with parts 71
and 73 of this chapter.
(f) Acceptance and installation at the site for which the
Commission has issued a construction permit that references the subject
manufacturing license.
(1) Installation must be in accordance with the construction permit
that references the subject manufacturing license.
(2) Upon arrival at the site, the manufactured reactor may not be
installed in its place of operation unless the construction permit
holder performs inspections sufficient to verify the reactor is in
compliance with the manufacturing license and has not been damaged in
transit. The construction permit holder must perform these inspections
in accordance with documented procedures subject to quality assurance
measures commensurate with their importance to safety. In addition,
inspections must confirm that the interface requirements between the
manufactured reactor or portions of a manufactured reactor and the
remaining portions of the nuclear plant are met.
Subpart E--Standard Design Approvals
Sec. 57.200 Scope.
This subpart sets out procedures for the filing and NRC staff
review of standard designs, or major portions thereof, for a nuclear
reactor of the type to which this part is applicable.
Sec. 57.205 Contents of applications; general information.
The application must contain all of the information required by 10
CFR 57.55(a) through (c) and (j).
[[Page 23742]]
Sec. 57.210 Contents of applications; technical information.
(a) If the applicant seeks review of a major portion of a standard
design, the application need only contain the information required by
this section to the extent the requirements are applicable to the major
portion of the standard design for which NRC staff approval is sought.
If an applicant seeks approval of a major portion of the design, the
application must demonstrate compliance with the design criteria
attributes in Sec. 57.30, as applicable, for the major portion of the
standard design for which NRC staff approval is sought. Such applicants
must identify conditions related to interfaces with systems outside the
scope of the major portion of the standard design for which NRC staff
approval is sought, and functional or physical boundary conditions
between the major portion of the standard design for which NRC staff
approval is sought and the remainder of the standard design. These
conditions must be demonstrated when the standard design approval is
incorporated into a subsequent joint application for a construction
permit and associated operating license(s) or a manufacturing license
application under this part.
(b) The application must contain a final safety analysis report
that describes the facility, presents the design bases and the limits
on its operation, and presents a safety analysis of the safety-related
SSCs and of the facility, or major portion thereof, and must include
the following information:
(1) Other than site-specific information, the information required
by Sec. 57.60(a)(1) through (3), (a)(6) and (7), and (a)(9) through
(13) relevant to the standard design;
(2) A description and analysis of the fire protection design
features for the standard plant necessary to limit fire damage to
safety-related SSCs as required by Sec. 57.60(a)(8)(ix)(B);
(3) The information necessary to demonstrate that the standard
plant complies with the environmental information relating to
applicable site evaluation factors for seismic, meteorological,
hydrologic, and geologic characteristics with appropriate consideration
of natural phenomena;
(4) A description, analysis, and evaluation of the interfaces
between the standard design and the balance of the nuclear plant; and
(5) The information required by Sec. 20.1406 of this chapter.
Sec. 57.213 Standards for review of applications.
Applications filed under this part will be reviewed under the
standards set out in 10 CFR parts 20, 57, and 73.
Sec. 57.215 Staff approval of design.
Upon completion of its review of a submittal under this subpart and
receiving any report submitted by the ACRS under Sec. 57.17, the NRC
staff must publish a determination in the Federal Register as to
whether the design is acceptable, subject to appropriate terms and
conditions, and make an analysis of the design in the form of a report
available at the NRC website, https://www.nrc.gov.
Sec. 57.220 Finality of standard design approvals; information
requests.
(a) An approved design must be used by and relied upon by the NRC
staff and the ACRS in their review of any joint application for a
construction permit and associated operating license(s) or a
manufacturing license application under this part that incorporates by
reference a standard design approved in accordance with this paragraph
unless there exists significant new information that substantially
affects the earlier determination or other good cause.
(b) The determination and report by the NRC staff do not constitute
a commitment to issue a construction permit, operating license, or
manufacturing license in any way affect the authority of the
Commission, Atomic Safety and Licensing Board Panel, or presiding
officers in any proceeding under part 2 of this chapter.
(c) Except for information requests seeking to verify compliance
with the current licensing basis of the standard design approval, the
NRC must prepare the reason or reasons for each information request to
the holder of a standard design approval under this part before
issuance to ensure that the burden to be imposed on respondents is
justified in view of the potential safety significance of the issue to
be addressed in the requested information. Each such justification
provided for an evaluation performed by the NRC staff must be approved
by the Executive Director for Operations or designee before issuance of
the request.
(d) The Commission will require, before granting a construction
permit, operating license, or manufacturing license that references a
standard design approval, that engineering documents, such as analyses,
drawings, procurement specifications, or construction and installation
specifications, be completed and available for audit if the more
detailed information is necessary for the Commission to verify the
information in the application and make its safety determination,
including the determination that the application is consistent with the
design approval information. This information may be acquired by
appropriate arrangements with the design approval applicant.
Sec. 57.225 Duration of design approval.
A standard design approval issued under this subpart has no
expiration date.
Subpart F--Reporting of Defects and Noncompliance
Sec. 57.230 Purpose.
The regulations in this subpart 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, who obtains information reasonably
indicating:
(a) that 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
(b) that the facility, activity, or basic component supplied to
such facility or activity contains defects, which could create a
substantial safety hazard, to immediately notify the Commission of such
failure to comply or such defect, unless the individual has actual
knowledge that the Commission has been adequately informed of such
defect or failure to comply.
Sec. 57.235 Scope.
(a) The regulations in this subpart apply to:
(1) Each individual, partnership, corporation, or other entity
applying for or holding a license or construction permit under this
part to construct, manufacture, possess, own, operate, or transfer
within the United States, a utilization facility; and each director and
responsible officer of such a licensee;
(2) Each individual, corporation, partnership, or other entity
doing business within the United States, and each director and
responsible officer of such an organization, that constructs a
utilization facility licensed for manufacture, construction, or
operation under this part; or supplies basic
[[Page 23743]]
components for a facility or activity licensed under this part; and
(3) Each individual, corporation, partnership, or other entity
doing business within the United States, and each director and
responsible officer of such an organization, applying for or holding a
standard design approval under this part; or supplying basic components
with respect to a standard design approval under this part.
(b) For persons licensed to construct a facility under subpart C of
this part, or to manufacture a facility under subpart D of this part,
evaluation of potential defects and failures to comply and reporting of
defects and failures to comply satisfies each person's evaluation,
notification, and reporting obligation to report defects and failures
to comply under this part and the responsibility of individual
directors and responsible officers of these licensees to report defects
under Section 206 of the Energy Reorganization Act of 1974.
(c) For persons licensed to operate a nuclear plant under subpart C
of this part, evaluation of potential defects and appropriate reporting
of defects under this subpart satisfies each person's evaluation,
notification, and reporting obligation to report defects under this
part, and the responsibility of individual directors and responsible
officers of these licensees to report defects under Section 206 of the
Energy Reorganization Act of 1974.
(d) Nothing in these regulations should be deemed to preclude
either an individual, a manufacturer, or a supplier of a commercial
grade item (as defined in Sec. 57.240) not subject to the regulations
in this part from reporting to the Commission, a known or suspected
defect or failure to comply and, as authorized by law, the identity of
anyone so reporting will be withheld from disclosure. NRC regional
offices and headquarters will accept collect telephone calls from
individuals who wish to speak to NRC representatives concerning nuclear
safety-related problems. The location and telephone numbers of the four
regions (answered during regular working hours) are listed in appendix
D to part 20 of this chapter. The telephone numbers of the NRC
Headquarters Operations Center (answered 24 hours a day--including
holidays) are listed in appendix A to part 73 of this chapter.
Sec. 57.240 Definitions.
For purposes of this subpart, the definitions in Sec. 57.3 of this
part apply, except the term ``construction.'' The following definitions
also apply for the purposes of this subpart.
Basic component means--
(1) a structure, system, or component, or part thereof necessary to
ensure:
(i) The capability to adequately control thermodynamic conditions
and reactivity, and to retain radioactive material;
(ii) The capability to shut down the reactor and maintain it in a
safe shutdown condition; or
(iii) The capability to prevent or mitigate the consequences of
accidents that could result in potential offsite exposures comparable
to those referred to in Sec. 57.25(a).
(2) Basic components are items designed and manufactured under a
quality assurance program complying with Sec. 57.60(a)(3) of this
part, or commercial grade items which have successfully completed the
dedication process.
(3) In all cases, basic components include safety related design,
analysis, inspection, testing, fabrication, replacement parts, or
consulting services that are associated with the component hardware,
whether these services are performed by the component supplier or other
supplier.
Commercial grade item means an item that is:
(1) Not subject to design or specification requirements that are
unique to facilities or activities licensed pursuant to this part;
(2) Used in applications other than facilities or activities
licensed pursuant to this part; and
(3) To be ordered from the manufacturer/supplier on the basis of
specifications set forth in the manufacturer's published product
description (for example, a catalog).
Constructing or construction, as used in this subpart, means the
analysis, design, manufacture, fabrication, placement, erection,
installation, modification, inspection, or testing of a facility or
activity that is subject to the regulations in this part and safety-
related consulting services related to the facility or activity.
Critical characteristics means those important design, material,
and performance characteristics of a commercial grade item that, once
verified, will provide reasonable assurance that the item will perform
its intended safety function.
Dedicating entity means the organization that performs the
dedication process.
Dedication means an acceptance process undertaken to provide
reasonable assurance that a commercial grade item to be used as a basic
component will perform its intended safety function and, in this
respect, is deemed equivalent to an item designed and manufactured
under a Sec. 57.60(a)(3) quality assurance program. This assurance is
achieved by identifying the critical characteristics of the item and
verifying their acceptability by inspections, tests, or analyses
performed by the purchaser or third-party dedicating entity after
delivery, supplemented as necessary by one or more of the following:
commercial grade surveys; product inspections or witness at hold points
at the manufacturer's facility, and analysis of historical records for
acceptable performance. In all cases, the dedication process must be
conducted in accordance with the applicant's applicable provisions for
their Sec. 57.60(a)(3) quality assurance program. The process is
considered complete when the item is designated for use as a basic
component.
Defect means:
(1) A deviation in a basic component delivered to a purchaser for
use in a facility or an activity subject to the regulations in this
part if, on the basis of an evaluation, the deviation could create a
substantial safety hazard;
(2) The installation, use, or operation of a basic component
containing a defect as defined in this part;
(3) A deviation in a portion of a facility subject to the
construction permit or manufacturing licensing requirements of this
part, provided the deviation could, on the basis of an evaluation,
create a substantial safety hazard and the portion of the facility
containing the deviation has been offered to the purchaser for
acceptance;
(4) A condition or circumstance involving a basic component that
could contribute to the exceeding of a safety limit, as defined in the
technical specifications of a license for operation issued under this
part; or
(5) An error, omission or other circumstance in a standard design
approval that, on the basis of an evaluation, could create a
substantial safety hazard.
Deviation means departure from the technical requirements included
in a procurement document or specified in standard design approval.
Discovery means the completion of the documentation first
identifying the existence of a deviation or failure to comply
potentially associated with a substantial safety hazard within the
evaluation procedures discussed in Sec. 57.270.
Evaluation means the process of determining whether a particular
deviation could create a substantial hazard or determining whether a
failure to comply is associated with a substantial safety hazard.
[[Page 23744]]
Procurement document means a contract that defines the requirements
which facilities or basic components must meet in order to be
considered acceptable by the purchaser.
Responsible officer means the president, vice-president, or other
individual in the organization of a corporation, partnership, or other
entity who is vested with executive authority over activities subject
to this part.
Substantial safety hazard means a loss of safety function to the
extent that there is a major reduction in the degree of protection
provided to public health and safety for any facility or activity
authorized under this part.
Sec. 57.255 Posting requirements.
(a) Posting of documents.
(1) Each individual, partnership, corporation, dedicating entity,
or other entity subject to the regulations in this part must post
current copies of --
(i) The regulations in this part;
(ii) Section 206 of the Energy Reorganization Act of 1974; and
(iii) Procedures adopted pursuant to the regulations in this part.
(2) These documents must be posted in a conspicuous position on any
premises within the United States where the activities subject to this
part are conducted.
(b) If posting of the regulations in this part or the procedures
adopted pursuant to the regulations in this part is not practicable,
the licensee or firm subject to the regulations in this part may, in
addition to posting Section 206 of the Energy Reorganization Act of
1974, post a notice that describes the regulations or procedures,
including the name of the individual to whom reports may be made, and
states where they may be examined.
Sec. 57.260 Exemptions.
Suppliers of commercial grade items are exempt from the provisions
of this part to the extent that they supply commercial grade items.
Sec. 57.270 Notification of failure to comply or existence of a
defect and its evaluation.
(a) Each individual, corporation, partnership, dedicating entity,
or other entity subject to the regulations in this part must adopt
appropriate procedures to--
(1) Evaluate deviations and failures to comply to identify defects
and failures to comply associated with substantial safety hazards as
soon as practicable, and, except as provided in paragraph (a)(2) of
this subsection, in all cases within 60 days of discovery, in order to
identify a reportable defect or failure to comply that could create a
substantial safety hazard, were it to remain uncorrected.
(2) Ensure that if an evaluation of an identified deviation or
failure to comply potentially associated with a substantial safety
hazard cannot be completed within 60 days from discovery of the
deviation or failure to comply, an interim report is prepared and
submitted to the Commission through a director or responsible officer
or designated person as discussed in Sec. 57.270(d)(5). The interim
report should describe the deviation or failure to comply that is being
evaluated and should also state when the evaluation will be completed.
This interim report must be submitted in writing within 60 days of
discovery of the deviation or failure to comply.
(3) Ensure that a director or responsible officer subject to the
regulations of this part is informed as soon as practicable, and, in
all cases, within the 5 working days after completion of the evaluation
described in paragraphs (a)(1) or (a)(2) of this section if the
manufacture, construction, or operation of a facility or activity, a
basic component supplied for such facility or activity, or standard
design approval of this part--
(i) Fails to comply with the AEA or any applicable rule,
regulation, order, or license of the Commission, relating to a
substantial safety hazard, or
(ii) Contains a defect.
(iii) For construction permit and manufacturing license holders,
undergoes any significant breakdown in any portion of the quality
assurance program conducted under the requirements of Sec.
57.60(a)(3), which could have produced a defect in a basic component.
These breakdowns in the quality assurance program are reportable
whether the breakdown actually resulted in a defect in a design
approved and released for construction, installation, or manufacture.
(b) If the deviation or failure to comply is discovered by a
supplier of basic components, or services associated with basic
components, and the supplier determines that it does not have the
capability to perform the evaluation to determine if a defect exists,
then the supplier must inform the purchasers or affected licensees
within five working days of this determination so that the purchasers
or affected licensees may evaluate the deviation or failure to comply,
pursuant to Sec. 57.270(a).
(c) A dedicating entity is responsible for--
(1) Identifying and evaluating deviations and reporting defects and
failures to comply associated with substantial safety hazards for
dedicated items; and
(2) Maintaining auditable records for the dedication process.
(d) Notifications to the NRC.
(1) A director or responsible officer subject to the regulations of
this part or a person designated under Sec. 57.270(d)(5) must notify
the Commission when he or she obtains information reasonably indicating
a failure to comply or a defect affecting--
(i) The manufacture, construction, or operation of a facility or an
activity within the United States that is subject to the licensing
requirements under this part and that is within his or her
organization's responsibility; or
(ii) A basic component that is within his or her organization's
responsibility and is supplied for a facility or an activity within the
United States that is subject to the licensing or approval requirements
under this part;
(iii) For construction permit and manufacturing license holders, a
quality assurance program that undergoes any significant breakdown that
could have produced a defect in a basic component.
(2) The notification to the NRC of a failure to comply or of a
defect under paragraph (d)(1) of this section and the evaluation of a
failure to comply or a defect under paragraphs (a)(1) and (a)(2) of
this section, are not required if the director or responsible officer
has actual knowledge that the Commission has been notified in writing
of the defect or the failure to comply.
(3) Notification required by paragraph (d)(1) of this section must
be made as follows--
(i) Initial notification to the NRC Headquarters Operations Officer
email address: [email protected], which is the preferred method of
notification, or by telephone to the NRC Operations Center at (301)
816--5100 within two days following receipt of information by the
director or responsible corporate officer under paragraph (a)(1) of
this section, on the identification of a defect or a failure to comply.
Verification that the email has been received should be made by calling
the NRC Operations Center. This paragraph does not apply to interim
reports described in Sec. 21.21(a)(2) of this chapter.
(ii) Written notification to the NRC at the address specified in
Sec. 57.4 within 30 days following receipt of information by the
director or responsible corporate officer under paragraph (a)(3) of
this subsection, on the identification of a defect or a failure to
comply.
(4) The written report required by paragraph (d)(1) must include,
but need not be limited to, the following information, to the extent
known:
[[Page 23745]]
(i) Name and address of the individual or individuals informing the
Commission.
(ii) Identification of the facility, the activity, or the basic
component supplied for such facility or such activity within the United
States that fails to comply or contains a defect.
(iii) Identification of the firm constructing the facility or
supplying the basic component that fails to comply or contains a
defect.
(iv) Nature of the defect or failure to comply and the safety
hazard that is created or could be created by such defect or failure to
comply.
(v) The date on which the information of such defect or failure to
comply was obtained.
(vi) In the case of a basic component that contains a defect or
fails to comply, the number and location of these components in use at,
supplied for, being supplied for, or may be supplied for, manufactured,
or being manufactured for one or more facilities or activities subject
to the regulations in this part.
(vii) The corrective action that has been, is being, or will be
taken; the name of the individual or organization responsible for the
action; and the length of time that has been or will be taken to
complete the action.
(viii) Any advice related to the defect or failure to comply about
the facility, activity, or basic component that has been, is being, or
will be given to purchasers or licensees.
(5) The director or responsible officer may authorize an individual
to provide the notification required by this paragraph, provided that,
this must not relieve the director or responsible officer of his or her
responsibility under this paragraph (d).
(e) Individuals subject to this part may be required by the
Commission to supply additional information related to a defect or
failure to comply. Commission action to obtain additional information
may be based on reports of defects from other reporting entities.
Sec. 57.275 Procurement documents.
Each individual, corporation, partnership, dedicating entity, or
other entity subject to the regulations in this part must ensure that
each procurement document for a facility, or a basic component issued
by him, her or it specifies, when applicable, that the provisions of 10
CFR part 57, subpart F apply.
Sec. 57.280 Inspections.
Each individual, corporation, partnership, dedicating entity, or
other entity subject to the regulations in this part must permit the
Commission to inspect records, premises, activities, and basic
components as necessary to accomplish the purposes of this part.
Sec. 57.285 Maintenance and inspection of records.
(a) Each individual, corporation, partnership, dedicating entity,
or other entity subject to the regulations in this part must prepare
and maintain records necessary to accomplish the purposes of this part,
specifically --
(1) Retain evaluations of all deviations and failures to comply for
a minimum of five years after the date of the evaluation;
(2) Suppliers of basic components must retain any notifications
sent to purchasers and affected licensees for a minimum of five years
after the date of the notification.
(3) Suppliers of basic components must retain a record of the
purchasers of basic components for 10 years after delivery of the basic
component or service associated with a basic component.
(4) Applicants for or holders of a standard design approval under
subpart E of this part and others providing a design that is the
subject of a design approval must retain any notifications sent to
purchasers and affected licensees for a minimum of 5 years after the
date of the notification, and retain a record of the purchasers for 15
years after delivery of the design which is the subject of the design
approval or service associated with the design.
(b) The holder of a construction permit or manufacturing license
must prepare and maintain records necessary to accomplish the purposes
of this part, specifically--
(1) Retain procurement documents, which define the requirements
that facilities or basic components must meet in order to be considered
acceptable, for the lifetime of the facility or basic component.
(2) Retain records of evaluations of all deviations and failures to
comply for the longer of:
(i) Ten (10) years from the date of the evaluation; or
(ii) Five (5) years from the date of the delivery of a manufactured
reactor.
(3) Suppliers of basic components must retain records of:
(i) All notifications sent to affected licensees or purchasers for
a minimum of 10 years following the date of the notification;
(ii) The facilities or other purchasers to whom basic components or
associated services were supplied for a minimum of 15 years from the
delivery of the basic component or associated service.
(c) Each individual, corporation, partnership, dedicating entity,
or other entity subject to the regulations in this part must permit the
Commission the opportunity to inspect records pertaining to basic
components that relate to the identification and evaluation of
deviations, and the reporting of defects and failures to comply,
including (but not limited to) any advice given to purchasers or
licensees on the placement, erection, installation, operation,
maintenance, modification, or inspection of a basic component.
Sec. 57.290 Failure to notify.
(a) Any director or responsible officer of an entity (including
dedicating entity) that is not otherwise subject to the deliberate
misconduct provisions of this chapter but is subject to the regulations
in this part who knowingly and consciously fails to provide the notice
required by Sec. 57.270 will be subject to a civil penalty equal to
the amount provided by section 234 of the AEA.
(b) Any NRC licensee or applicant for a license (including an
applicant for, or holder of, a construction permit), or applicant for
or holder of a standard design approval under subpart E, subject to the
regulations in this part who fails to provide the notice required by
Sec. 57.270, or otherwise fails to comply with the applicable
requirements of this part will be subject to a civil penalty as
provided by section 234 of the AEA.
(c) The dedicating entity, pursuant to Sec. 57.270(c) of this
part, is responsible for identifying and evaluating deviations,
reporting defects and failures to comply for the dedicated item, and
maintaining auditable records of the dedication process. NRC
enforcement action can be taken for failure to identify and evaluate
deviations, failure to report defects and failures to comply, or
failure to maintain auditable records.
Subpart G--Irradiated Fuel Storage, Decommissioning, and License
Termination Requirements
Sec. 57.300 Irradiated fuel storage.
While an irradiated fuel transportation package certified under 10
CFR part 71 of this chapter or irradiated fuel storage system certified
under 10 CFR part 72 is in the SNM handling or storage area, the
requirements in 10 CFR part 71 or 72, as applicable, and the
requirements of the certificate of compliance for that package or
storage system, are the applicable requirements for the fuel within
that package or storage system.
(a) Operating licensee. After cessation of operations of a nuclear
reactor
[[Page 23746]]
licensed under this part, the holder of the operating license may store
the fuel irradiated in the reactor at the operating site by either in-
reactor storage governed by the provisions of the operating license or
transfer of the irradiated fuel to an NRC-certified irradiated fuel
storage system pursuant to the provisions of 10 CFR part 72. If the
operating license is no longer in effect, a 10 CFR part 72 site-
specific license is required to maintain a storage installation at the
operating site location.
(b) Manufacturing licensee. A holder of a manufacturing license
under this part and a license under 10 CFR part 70 for possession of
the special nuclear material contained in a reactor manufactured under
the manufacturing license may store the reactor's irradiated fuel at
the manufacturing site by either in-reactor storage if the reactor has
been certified as a 10 CFR part 72 irradiated fuel storage system, or
transfer of the reactor's irradiated fuel to an NRC-certified
irradiated fuel storage system pursuant to the provisions of 10 CFR
part 72. The manufacturing license holder may temporarily allow
irradiated fuel to remain within the reactor after operational testing
and before shipment to an operating site or when a reactor containing
irradiated fuel is returned to the manufacturing facility site. The
manufacturing license holder must demonstrate that the irradiated fuel
in the reactor is maintained in a safe condition and that radiological
dose to the workers and the public is consistent with the provisions in
10 CFR part 72.
(c) Site-specific licensee. A holder of a 10 CFR part 70 license
for possession of SNM and a site-specific license under 10 CFR part 72
for irradiated fuel storage may store irradiated fuel from a reactor at
the licensed storage site after transfer of the reactor's irradiated
fuel to an NRC-certified irradiated fuel storage system pursuant to the
provisions of 10 CFR part 72.
(d) Irradiated fuel storage plan. Licensees that do not have an
approved plan for storage of irradiated fuel must submit, for NRC
review and approval under Sec. 57.310, a plan describing how the
licensee intends to manage and provide funding for the management of
all irradiated fuel at the designated storage site following permanent
cessation of operations of the reactor.
(1) Submission of this plan must occur (1) within 1 year following
permanent cessation of operations of the reactor, (2) more than 2 years
before expiration of the reactor operating license if storage occurs at
the reactors site, or (3) more than 2 years before expiration of the
manufacturing license if storage occurs at the manufacturing site,
whichever occurs first.
(2) The licensee must demonstrate to the NRC that the storage
management and funding plan is in compliance with all applicable
possession, safety, and environmental requirements for storage of
irradiated fuel, and must address, as applicable, transport to a
designated storage site.
Sec. 57.305 Decommissioning and license termination.
(a)(1) When a licensee has determined to permanently cease
operations, the licensee must, within 30 days, submit a written
certification to the NRC, consistent with the requirements of Sec.
57.4(b)(8);
(2) If the fuel has been permanently removed from the reactor on
site or transferred to a licensed remediation or storage facility, the
licensee must submit a written certification to the NRC that meets the
requirements of Sec. 57.4(b)(9);
(3) A licensee that permanently ceases site operations must make
notification of the permanent cessation of operations no later than 1
year prior to the expiration of the operating license.
(b) Licensees that do not have an approved decommissioning plan at
the time of permanent cessation of operations are subject to the
requirements of Sec. 50.82(b) of this chapter. These licensees'
decommissioning plans may be limited to those provisions applicable to
the design characteristics of the nuclear reactors or nuclear plants
and must address, as applicable, transport of nuclear reactors to
designated facilities for final decommissioning, final decommissioning
of individual nuclear reactors, or final decommissioning of entire
nuclear plants, and ensure compliance with all applicable safety and
environmental requirements.
(c)(1) Decommissioning trust funds may be used by licensees that
meet the following requirements:
(i) The withdrawals are for expenses for legitimate decommissioning
activities consistent with the definition of decommissioning in Sec.
57.3;
(ii) The expenditure would not reduce the value of the
decommissioning trust below an amount necessary to place and maintain
the reactor in a safe storage condition if unforeseen conditions or
expenses arise; and
(iii) The withdrawals would not inhibit the ability of the licensee
to complete funding of any shortfalls in the decommissioning trust
needed to ensure the availability of funds to ultimately release the
site and terminate the license.
(2) Unless otherwise noted in a licensee's NRC-approved
decommissioning plan, and until the licensee has completed its final
radiation survey and demonstrated that residual radioactivity has been
reduced to a level that permits termination of its license, the
licensee must annually submit to the NRC, by March 31, a financial
assurance status report. The report must include the following
information, current through the end of the previous calendar year:
(i) The amount spent on decommissioning, both cumulative and over
the previous calendar year, the remaining balance of any
decommissioning funds, and the amount provided by other financial
assurance methods being relied upon;
(ii) An estimate of the costs to complete decommissioning,
reflecting any difference between actual and estimated costs for work
performed during the year, and the decommissioning criteria upon which
the estimate is based;
(iii) Any modifications occurring to a licensee's current method of
providing financial assurance since the last submitted report; and
(iv) Any material changes to trust agreements or financial
assurance contracts.
(3) If the sum of the balance of any remaining decommissioning
funds, plus earnings on such funds calculated at not greater than a 2
percent real rate of return, together with the amount provided by other
financial assurance methods being relied upon, does not cover the
estimated cost to complete the decommissioning, the financial assurance
status report must include additional financial assurance to cover the
estimated cost of completion.
(d) Licensees may not perform any decommissioning activities that--
[hyphen]
(1) Foreclose release of the site for possible unrestricted use;
(2) Result in significant environmental impacts not previously
reviewed; or
(3) Result in there no longer being reasonable assurance that
adequate funds will be available for decommissioning.
(e) If the operating license is the only operating license for a
nuclear reactor using the shared portions of the plant described in
Sec. 57.60(a)(4)(iii), then the entire nuclear plant must be
decommissioned before termination of the operating license.
(f) All holders of operating licenses are subject to the license
termination provisions of Sec. 50.82(b) of this chapter.
[[Page 23747]]
Subpart H--Maintaining and Revising Licensing Basis Information
Sec. 57.310 Amendment of license.
(a) Whenever a holder of a construction permit, operating license,
or manufacturing license desires to amend the license, application for
an amendment must be filed with the Commission, as specified in Sec.
57.4, as applicable. The application must fully describe the changes
desired and follow, as far as applicable, the form prescribed for
original applications.
(b) In determining whether an amendment to a license issued under
this part will be issued to the applicant, the Commission will be
guided by the considerations that govern the issuance of initial
licenses to the extent applicable and appropriate. If the application
involves the material alteration of a licensed facility, a construction
permit will be issued before the issuance of the amendment to the
license. However, no application for a construction permit is required
if the application involves a material alteration to a nuclear reactor
manufactured under a manufacturing license issued under this part
before the reactor is installed at a site. If the amendment involves a
significant hazards consideration, the Commission will give notice of
its proposed action according to the following:
(1) Under Sec. 2.105 of this chapter before acting thereon; and
(2) As soon as practicable after the application has been docketed.
(c) The Commission will be particularly sensitive to a license
amendment request that involves irreversible consequences (such as one
that permits a significant increase in the amount of effluents or
radiation emitted by a nuclear plant).
(d) The Commission may make a final determination, under the
procedures in Sec. 50.91 of this chapter, that a proposed amendment to
an operating license under this part involves no significant hazards
consideration, if operation of the facility in accordance with the
proposed amendment would not:
(1) Involve a significant increase in the likelihood or
consequences of an accident previously evaluated; or
(2) Create the possibility of a new or different kind of accident
from any accident previously evaluated; or
(3) Involve a significant reduction in a margin of safety.
(e) For an application requesting an amendment to an operating
license under this part, the Commission will use the procedures in
Sec. 50.91 of this chapter for notifying the public and consulting the
State.
Sec. 57.312 Changes to facility as described in final safety analysis
reports.
(a) A licensee under this part may make changes in the facility as
described in the final safety analysis report, make changes in the
procedures as described in the final safety analysis report, and
conduct tests or experiments not described in the final safety analysis
report without obtaining a license amendment pursuant to Sec. 57.310
in accordance with the requirements in Sec. 50.59 of this chapter.
(b) The holder of an operating license issued under this part that
authorizes operation of a manufactured reactor may make changes in the
facility as described in the final safety analysis report (as updated)
and make changes in the procedures as described in the final safety
analysis report (as updated) if the changes are identical to changes
approved by the Commission by amendment to the manufacturing license
for the manufactured reactor and upon determining that implementation
of the changes will be consistent with the basis for the Commission's
approval of the amendment to the manufacturing license and not involve
any additional changes that would require an amendment to its operating
license.
Sec. 57.315 Maintenance and submittal of the final safety analysis,
as updated.
(a) Each holder of an operating license issued under this part and
each holder of a manufacturing license issued under this part must
update periodically the FSAR originally submitted as part of the
application for the license, to ensure that the information included in
the report contains the latest information developed. This submittal
must contain all the changes necessary to reflect information and
analyses submitted to the Commission by the applicant or licensee or
prepared by the applicant or licensee pursuant to Commission
requirement since the submittal of the original FSAR, or as
appropriate, the last update to the FSAR under this section. The
submittal must include the effects of all changes made in the facility
or procedures as described in the FSAR; all safety analyses and
evaluations performed by the applicant or licensee either in support of
approved license amendments or in support of conclusions that changes
did not require a license amendment in accordance with Sec.
50.59(c)(2) or (e) of this chapter and all analyses of new safety
issues performed by or on behalf of the applicant or licensee at
Commission request. Effects of changes include appropriate revisions of
descriptions in the FSAR such that the FSAR (as updated) is complete
and accurate. The updated information must be appropriately located
within the update to the FSAR.
(b) The licensee must submit revisions containing updated
information to the Commission, as specified in Sec. 57.4, on a
replacement-page basis that is accompanied by a list which identifies
the current pages of the FSAR following page replacement. Each
submittal must reflect all changes made to the FSAR up to a maximum of
6 months prior to the date of filing the submittal.
(c) The submittal must include:
(1) a certification by a duly authorized officer of the licensee
that either the information accurately presents changes made since the
previous submittal, necessary to reflect information and analyses
submitted to the Commission or prepared pursuant to Commission
requirement, or that no such changes were made; and
(2) an identification of changes made under the provisions of Sec.
50.59 of this chapter but not previously submitted to the Commission.
(d) Each replacement page must include both a change indicator for
the area changed, e.g., a bold line vertically drawn in the margin
adjacent to the portion actually changed, and a page change
identification (date of change or change number or both).
(e) The updated FSAR must be retained by the licensee until the
Commission terminates their license.
Sec. 57.317 Updated decommissioning report.
The report required by Sec. 57.55(i) must be updated and submitted
to the NRC as specified in Sec. 57.4 before issuance of any operating
license associated with an approved construction permit, within 3 years
following issuance of an operating license, and no more than every 3
years thereafter for that operating license. The updated information
must include the amount of decommissioning funds estimated to be
required; the amount of decommissioning funds accumulated to the end of
the calendar year preceding the date of the report; a schedule of the
annual amounts remaining to be collected; and the assumptions used
regarding rates of escalation in decommissioning costs, rates of
earnings on decommissioning funds, and rates of other factors used in
funding projections.
Subpart I--Transportation Package Design Certification
Sec. 57.319 Purpose.
This subpart sets forth the requirements and procedures applicable
[[Page 23748]]
to certificates of compliance for packaging and shipping of one or more
reactors manufactured or operated under a license issued under this
part.
Sec. 57.320 Applicability.
While an irradiated fuel transportation package approved under 10
CFR part 71 of this chapter is in the SNM handing or storage area at
the licensee's site, the requirements in 10 CFR part 71, as applicable,
and the requirements of the certificate of compliance for that package,
are the applicable requirements for the fuel within that package.
(a) Reactor as transportation package. A licensee under this part
may transport a reactor loaded with fuel, either irradiated or
unirradiated, under a certificate of compliance issued pursuant to 10
CFR part 71 if the licensee meets the following criteria:
(1) The requirements of 10 CFR part 71 considering the reactor as
the transportation package have been met. In lieu of an evaluation of
the effects of the tests required by 10 CFR 71.41(a) and specified in
10 CFR 71.71, 71.73 and 71.61 on a package, a risk methodology or other
risk-informed approach for evaluating normal and/or accident conditions
that has been endorsed or otherwise approved by the Commission may be
used to evaluate a package for certification, and
(2) Features to prevent criticality that meet the requirements of
Sec. 57.160(f)(1)(ii) are in place.
(b) Reactor as transportation package contents. A 10 CFR part 71
general license is issued to any licensee of the Commission to
transport, or to deliver to a carrier for transport, licensed material
in a package for which a license, certificate of compliance, or other
approval has been issued by the NRC. The fueled reactor as
transportation package contents must have been identified as authorized
contents in the transportation package certificate of compliance in the
application for a new package certification or through an amendment of
an existing transportation package under 10 CFR 71.19(c) before the
licensee's first use of the transportation package to transport a
reactor.
(1) A general licensee must meet the requirements of 10 CFR 71.17,
and
(2) Features to prevent criticality that meet the requirements of
Sec. 57.160(f)(1)(ii) must be in place before the first use of the
package.
Subpart J--Physical Security Requirements
Sec. 57.325 Physical security requirements.
(a) Introduction.
(1) Each licensee that is licensed to operate a nuclear reactor
under this part and did not meet the requirement in Sec.
57.60(a)(8)(v)(A)(3) must implement the requirements of this section
through its physical security plan, training and qualification plan,
safeguards contingency plan, and cybersecurity plan, referred to
collectively hereafter as ``security plans,'' before initial fuel load
into the reactor (or, for a fueled manufactured reactor, before
initiating the removal of any of the features to prevent criticality
required under Sec. 57.160(f)(1)(ii)).
(2) The security plans must identify, describe, and account for
site-specific conditions that affect the licensee's capability to
satisfy the requirements of this section.
(b) General performance objective and requirements.
(1) The licensee must establish, implement, and maintain a physical
protection program and a security organization, which will have as
their objective to provide reasonable assurance that activities
involving special nuclear material are not inimical to the common
defense and security and do not constitute an unreasonable risk to the
public health and safety.
(2) The physical protection program must be designed to prevent a
release of radionuclides from any source from exceeding the dose
reference values defined in Sec. 50.34(a)(1)(ii)(D)(1) of this
chapter.
(3) To satisfy the general performance objective of paragraph
(b)(1) of this section, the physical protection program must protect
against the design basis threat of radiological sabotage as stated in
Sec. 73.1 of this chapter.
(4) The physical protection program must be designed and
implemented to achieve and maintain the reliability and availability of
SSCs required for demonstrating compliance with the following
performance requirements at all times:
(i) Intrusion detection. The licensee must be capable of detecting
attempted and actual unauthorized access to interior and exterior areas
containing SSCs needed to implement safety and security functions.
(ii) Intrusion assessment. The licensee must be capable of timely
assessment for determining the cause of a detected intrusion.
(iii) Security communication. The licensee must be capable of
continuous security communications. Communication systems must account
for design basis threats that can interrupt or interfere with
continuity or integrity of communications.
(iv) Security response. The physical protection program must be
designed to provide timely security response to interdict and
neutralize adversary attacks up to and including the design basis
threat of radiological sabotage.
(5) The licensee must provide necessary information about the
facility and make available periodic training to law enforcement or
other offsite armed responders who will fulfill the interdiction and
neutralization functions for threats up to and including the design
basis threat of radiological sabotage.
(6) The licensee must be capable of detecting and denying
unauthorized access to persons and pass-through of contraband materials
to protected areas.
(7) The licensee must document and maintain the process used to
develop and identify target sets, to include the site-specific analyses
and methodologies used to determine and group the target set equipment
or elements.
(8) The licensee must implement a process for the oversight of
target set equipment and systems to ensure that changes to the
configuration of the identified equipment and systems are considered in
the licensee's protective strategy. Where appropriate, changes must be
made to documented target sets.
(9) The licensee must establish, implement, and maintain a
performance evaluation program to assess the effectiveness of the
licensee's implementation of the physical protection program to protect
against the design basis threat of radiological sabotage.
(10) The licensee must establish, implement, and maintain a
cybersecurity program under Sec. 73.54 or Sec. 73.110 of this chapter
and must describe the program in the cybersecurity plan.
(11) The licensee must establish, implement, and maintain an
insider mitigation program and must describe the program in the
physical security plan.
(12) The licensee must have the capability to track, trend,
correct, and prevent recurrence of failures and deficiencies in the
implementation of the requirements of this section.
(13) Implementation of security plans and associated procedures
must be coordinated with other onsite plans and procedures to preclude
conflict during both normal and emergency conditions and ensure the
adequate management of the safety and security interface.
(14) The licensee must ensure that the firearms background check
requirements of Sec. 73.17 of this chapter are met for all members of
the security organization whose official duties
[[Page 23749]]
require access to covered weapons or who inventory enhanced weapons.
The provisions of this paragraph are only applicable to licensees
subject to this section that are also subject to the firearms
background check provisions of Sec. 73.17 of this chapter.
(c) Protection of records. The licensee must retain, in accordance
with paragraph (h) of this section, all analyses, assessments,
calculations, and descriptions of the technical basis for demonstrating
compliance with the performance requirements of paragraph (b) of this
section. The licensee must protect these records in accordance with the
requirements for protecting safeguards information in Sec. Sec. 73.21
and 73.22 of this chapter.
(d) Search requirements. The licensee must establish and implement
searches of individuals, vehicles, and materials to detect and prevent
the introduction into the protected area of firearms, explosives,
incendiary devices, or other items and material which could be used to
commit radiological sabotage.
(e) Training and qualification program. The licensee must establish
and maintain a training and qualification program that ensures
personnel who are responsible for the physical protection of the
facility against radiological sabotage are able to effectively perform
their assigned security-related job duties for implementing the
requirements of this section and must describe the program in the
training and qualification plan.
(f) Performance evaluation. Licensee performance evaluations must
include methods appropriate and necessary to assess, test, and
challenge the integration of the physical protection program's
functions to protect against the design basis threat, including
measures to protect against cyberattack and engineered systems designed
to protect against the design basis threat standalone ground vehicle
bomb attack.
(g) Suspension of security measures.
(1) The licensee may suspend implementation of affected
requirements of this section in accordance with Sec. 57.399(g)-(h) of
this chapter under the following conditions:
(i) In an emergency, when action is immediately needed to protect
the public health and safety; and
(ii) During severe weather, when the suspension of affected
security measures is immediately needed to protect the personal health
and safety of personnel.
(2) Suspended security measures must be reinstated as soon as
conditions permit.
(3) The suspension of security measures must be reported and
documented in accordance with the provisions of Sec. Sec. 73.1200 and
73.1205 of this chapter.
(h) Records.
(1) The Commission may inspect, copy, retain, and remove all
reports, records, and documents required to be kept by Commission
regulations, orders, or license conditions, whether the reports,
records, and documents are kept by the licensee or a contractor.
(2) The licensee must maintain all records required to be kept by
Commission regulations, orders, or license conditions, until the
Commission terminates the license for which the records were developed
and must maintain superseded portions of these records for at least 3
years after the record is superseded, unless otherwise specified by the
Commission.
(3) If a contracted security force is used to implement the onsite
physical protection program, the licensee's written agreement with the
contractor must be retained by the licensee as a record for the
duration of the contract.
(4) Review and audit reports must be available for inspection, for
a period of 3 years.
Subpart K--Categorical Exclusion
Sec. 57.350 Categorical exclusion.
(a) The NRC has determined that the categories of actions
identified in paragraph (b) of this section meet the criteria for
categorical exclusion pursuant to 10 CFR 51.22.
(b) The issuance of an initial or renewed license for a
microreactor or other reactor with a comparable risk profile, and all
forms of related NRC actions, including amendments, exemptions and
orders, under this part, are categorically excluded from the
requirement to prepare an environmental assessment or environmental
impact statement, provided that the following criteria are met:
(1) The application for the initial or renewed license, amendment,
or exemption, or the order, demonstrates that the licensed action is
within the environmental plant parameter and site parameter envelope
for Table C-1 of Appendix C of 10 CFR part 51, which may include the
siting of multiple reactors across a region or at one site.
(2) The application for the initial or renewed license, amendment,
or exemption, or the order, demonstrates the following:
(i) The site will be within a previously disturbed area as defined
in Sec. 57.3;
(ii) The cooling system(s) will not require the use or consumption
of water withdrawn directly from surface water or groundwater sources
or discharges to surface water or groundwater sources;
(iii) Air emissions will be below de minimis threshold levels in 40
CFR 93.153(b)(1) or (b)(2), as applicable; and
(iv) The licensed activity will be in accordance with applicable
State and local requirements (such as land use planning, zoning
requirements, and coastal zone management program requirements under
the Coastal Zone Management Act) in the proposed site or region.
Subpart L--Inspections
Sec. 57.355 Unfettered access for inspections.
(a) Each applicant for or holder of a construction permit,
operating license, or manufacturing license, and each general licensee
under Sec. 57.45(d), must permit inspection, by duly authorized
representatives of the Commission, of its records, premises, and
activities, and of licensed materials in possession or use, related to
the license or construction permit as may be necessary to effectuate
the purposes of the AEA and the Energy Reorganization Act of 1974, as
amended.
(b) Each holder of a construction permit, operating license, or
manufacturing license must provide adequate facilities and access for
Commission inspection personnel as follows:
(1) Each holder of a construction permit, operating license, or
manufacturing license must provide temporary office space for the
exclusive use of the Commission inspection personnel. Heat, air
conditioning, light, and electrical outlets must be furnished by each
licensee and each holder of a construction permit. The office space
must be convenient to and have full access to the facility and must
provide the inspectors with both visual and acoustic privacy. The
office space must be generally commensurate with other office
accommodations at the site.
(2) The licensee or permit holder must afford any NRC inspectors
identified by the Regional Administrator as likely to inspect the
facility, immediate unfettered access, equivalent to access provided
regular plant employees, following proper identification and compliance
with applicable access control measures for security, radiological
protection, and personal safety.
(3) The licensee or permit holder must ensure that the arrival and
presence of an NRC inspector, who has been properly authorized facility
access as described in paragraph (b)(2) of this section, is not
announced or otherwise
[[Page 23750]]
communicated by its employees or contractors to other persons at the
facility unless specifically requested by the NRC inspector.
(c) For fuel cycle facilities licensed under part 70, NRC
inspections are conducted in accordance with 10 CFR 70.55.
(d) For a licensee, certificate holder, and applicant for a
certificate of compliance, NRC transportation inspections are conducted
in accordance with 10 CFR 71.93.
(e) For a holder of a license to receive, possess, package, or
transfer irradiated fuel, high-level radioactive waste, or reactor-
related greater than Class C waste, NRC inspections are conducted in
accordance with 10 CFR 72.82.
Subpart M--Material Control and Accounting
Sec. 57.360 Material control and accounting.
(a) Licensees of facilities licensed under this part and containing
special nuclear material (SNM) are subject to the material control and
accounting requirements found in 10 CFR 74.11, 74.13, 74.15, and 74.19.
(b) Licensees of facilities under this part with initial
unirradiated fuel load that averages greater than 10% uranium-235 (U-
235) enrichment but less than 20% U-235 enrichment and that do not have
personnel on site must perform the physical inventory with not greater
than 6 months periodicity.
(c) Each licensee under this part that possesses more than 1 gm of
SNM must report location changes in accordance with 10 CFR 74.15.
Subpart N [Reserved]
Subpart O--Enforcement
Sec. 57.380 Violations.
(a) The Commission may obtain an injunction or other court order to
prevent a violation of the provisions of--
(1) The AEA;
(2) Title II of the Energy Reorganization Act of 1974, as amended;
or
(3) A regulation or order issued under those Acts.
(b) The Commission may obtain a court order for the payment of a
civil penalty imposed under section 234 of the AEA:
(1) For violations of--
(i) Sections 53, 57, 62, 63, 81, 82, 101, 103, 104, 107, or 109 of
the AEA;
(ii) Section 206 of the Energy Reorganization Act of 1974, as
amended;
(iii) Any rule, regulation, or order issued under the sections
specified in paragraph (b)(1)(i) of this section;
(iv) Any term, condition, or limitation of any license issued under
the sections specified in paragraph (b)(1)(i) of this section.
(2) For any violation for which a license may be revoked under
section 186 of the AEA.
Sec. 57.385 Criminal penalties.
(a) Section 223 of the AEA provides for criminal sanctions for
willful violation of, attempted violation of, or conspiracy to violate,
any regulation issued under sections 161(b), 161(i), or 161(o) of the
AEA. For purposes of section 223, all the regulations in part 57 are
issued under one or more of sections 161(b), 161(i), or 161(o), except
for the sections listed in paragraph (b) of this section.
(b) The regulations in 10 CFR part 57 that are not issued under
sections 161b, 161i, or 161o for the purposes of section 223 are as
follows: Sec. Sec. 57.1, 57.2, 57.3, 57.4, 57.8, 57.9, 57.11, 57.12,
57.15, 57.16, 57.17, 57.18, 57.19, 57.20, 57.25, 57.30, 57.35, 57.40,
57.55, 57.60, 57.80, 57.90, 57.95, 57.100, 57.105, 57.115, 57.120,
57.130, 57.135, 57.142, 57.145, 57.150, 57.155, 57.160, 57.165, 57.170,
57.172, 57.175, 57.180, 57.185, 57.190, 57.200, 57.205, 57.210, 57.213,
57.215, 57.220, 57.225, 57.230, 57.235, 57.240, 57.260, 57.290, 57.310,
57.319, 57.350, 57.380, 57.385, 57.390, 57.415.
Subpart P--Operator Licensing and Human Factors
Sec. 57.390 Definitions.
For the purposes of this subpart, the following definitions apply:
Auxiliary operator means any individual who operates components of
a nuclear plant under this part but does not manipulate controls or
direct the manipulation of controls of the plant and is not required to
be licensed under the provisions of this part.
Facility licensee means the holder of an operating license under
this part for the nuclear plant where a generally licensed reactor
operator, operator, or senior operator would be licensed or is
licensed.
Generally licensed reactor operator (GLRO) means any individual
licensed under the provisions of Sec. 57.405 to manipulate controls of
an operator-independent facility licensed under this part and to direct
the licensed activities of GLROs.
Licensed medical examiner means an individual licensed by a State
or territory of the United States, the District of Columbia, or the
Commonwealth of Puerto Rico to conduct medical examinations for the
purpose of determining an individual's medical condition and general
health.
Load following means a nuclear plant automatically changing its
output to match expected demand in response to externally originated
instructions or signals.
Operator means any individual licensed under the provisions of
Sec. Sec. 57.420 through 57.427 to manipulate controls of an operator-
dependent facility licensed under this part.
Operator-dependent facility means a nuclear plant whose design
demonstrates that operator actions are required to maintain the nuclear
plant within the criterion of Sec. 57.25(a).
Operator-independent facility means a nuclear plant whose design
demonstrates that no operator actions are required to maintain the
nuclear plant within the criterion of Sec. 57.25(a).
Performance testing means testing conducted to verify a simulation
facility's performance as compared to actual or predicted reference
plant performance.
Physician means an individual licensed by a State or territory of
the United States, the District of Columbia or the Commonwealth of
Puerto Rico to dispense drugs in the practice of medicine.
Reference plant means the specific nuclear power plant from which a
simulation facility's control room configuration, system control
arrangement, and design data are derived. The reference plant may or
may not be constructed.
Senior operator means any individual licensed under the provisions
of Sec. Sec. 57.420 through 57.427 to manipulate controls of an
operator-dependent facility licensed under this part and to direct the
licensed activities of operators.
Simulation facility or simulator means an interface designed to
provide a realistic imitation of the operation of a nuclear plant and
used for the administration of examinations, for training, and/or to
demonstrate compliance with experience prerequisites for applicants or
GLROs, operators, or senior operators. A simulation facility may rely,
in whole or part, upon the physical utilization of the reference plant
itself.
Systems approach to training means a training program that includes
the following five elements:
(1) Systematic analysis of the jobs to be performed.
(2) Learning objectives derived from the analysis which describe
desired performance after training.
[[Page 23751]]
(3) Training design and implementation based on the learning
objectives.
(4) Evaluation of trainee mastery of the objectives during
training.
(5) Evaluation and revision of the training based on the
performance of trained personnel in the job setting
Sec. 57.391 General requirements for operator licensing and human
factors.
(a) Two classes of nuclear plants. Nuclear plants licensed under
this part are of the class of either operator-independent facilities or
operator-dependent facilities, based upon the similarity of operating
and technical characteristics of the plants in the class. A nuclear
plant is an operator-independent facility if the NRC determined as part
of its approval of the operating license for that plant that its design
demonstrates that no operator actions are required to maintain the
reactors within the criterion of Sec. 57.25(a). Otherwise, the nuclear
plant is an operator-dependent facility.
(b) Purpose and applicability. The regulations in Sec. Sec. 57.390
through 57.429 address areas related to staffing, training, personnel
qualifications, human factors engineering, generally licensed reactor
operators, operators, and senior operators, for applicants for or
holders of operating licenses under this part. These regulations are
organized as follows:
(1) Sections 57.391 through 57.399 address staffing, training,
personnel qualifications, and human factors engineering requirements.
The regulations within these sections are applicable to all applicants
for or holders of operating licenses under this part, except where
specifically stated otherwise.
(2) Sections 57.400 through 57.415 address generally licensed
reactor operator requirements. The regulations within these sections
are applicable to those applicants for or holders of operating licenses
under this part for operator-independent facilities that have not yet
certified the permanent cessation of operations and permanent removal
of fuel from the reactor vessel as described under Sec. 57.305(a).
(3) Sections 57.420 through 57.427 address operator and senior
operator requirements. The regulations within these sections are in
lieu of Sec. Sec. 57.400 through 57.415 for those applicants for or
holders of operating licenses under this part for operator-dependent
facilities that have not yet certified the permanent cessation of
operations and permanent removal of fuel from the reactor vessel as
described under Sec. 57.305(a).
(4) Section 57.429 provides general personnel training and
qualification requirements. The regulations within this section are
applicable to all applicants for or holders of operating licenses under
this part.
Sec. 57.392 Communications.
(a) Except as provided under a regional licensing program
identified in paragraph (b) of this section, an applicant or licensee
or facility licensee must submit any communication or report required
by the regulations contained within Sec. Sec. 57.391 through 57.429
and any application filed under these regulations to the Commission
using any of the methods specified in Sec. 57.4(a).
(b) (1) The Director, Office of Nuclear Reactor Regulation, has
delegated to the Regional Administrators of Regions I, II, III, and IV
authority and responsibility under the regulations in this part for the
issuance of licenses for operators and senior operators of nuclear
power reactors licensed under this part and located in these regions.
(2) Any application for an operator or senior operator license
filed under the regulations in Sec. 57.420 and any related inquiry,
communication, information, or report must be submitted to the
appropriate Regional Administrator listed in appendix D to 10 CFR part
20 by a method specified in Sec. 57.4(a). The Regional Administrator
or their designee will transmit to the Director, Office of Nuclear
Reactor Regulation, any matter that is not within the scope of the
Regional Administrator's delegated authority.
(c) Each facility licensee that is required to comply with the
requirements of Sec. Sec. 57.420 through 57.427 must notify the
appropriate Regional Administrator regarding an operator or senior
operator within 30 days of the following events:
(1) Permanent reassignment from the position for which the facility
licensee has certified the need for an operator or senior operator
under Sec. 57.423(a)(1);
(2) Termination of any operator or senior operator; or
(3) Permanent disability or illness as required under Sec. 57.422.
Sec. 57.393 Completeness and accuracy of information.
Information provided to the Commission by an applicant for an
operator or senior operator license or by a licensee or information
required by statute or the Commission's regulations, orders, or license
conditions to be maintained by the applicant or the licensee must be
complete and accurate in all material respects.
Sec. 57.395 Human factors engineering requirements.
Applicants for or holders of an operating license for a nuclear
plant licensed under this part must comply with the following:
(a) Human-system interface design requirements. The plant design
must provide for the following to support operating personnel in
monitoring plant conditions and responding to plant events:
(1) Features for displaying to operating personnel a minimum set of
parameters that define the safety status of the plant and are capable
of displaying both the full range of important plant parameters and
data trends on demand, as well as indicating when process limits are
being approached or exceeded;
(2) Automatic indication of the bypassed and operable status of
safety systems;
(3) Direct indication of SSC status that relates to the ability of
the SSC to perform its safety function, such as relief and safety valve
position (i.e., open or closed), and ultimate heat sink and cooling
system status and availability;
(4) Instrumentation to measure, record, and display key plant
parameters related to the performance of SSCs and the integrity of
barriers important to fulfilling safety functions to support operators
in monitoring plant conditions and responding to plant events.
(5) Leakage control and detection in the design of systems that
pass through barriers important to fulfilling safety functions for the
release of radionuclides.
(6) Monitoring of in-plant radiation and airborne radioactivity as
appropriate for a broad range of normal operating and accident
conditions; and
(7) The capability for GLRO, operator, or senior operator to do the
following:
(i) Receive plant operating data, including reactor parameters and
information needed for the evaluation of emergency conditions.
(ii) Promptly dispatch operations and maintenance personnel.
(iii) Immediately implement responsibilities under the facility
emergency plan, as applicable.
(iv) Immediately initiate a reactor shutdown from their location.
(b) Operating experience. A program, during construction and during
operation, as applicable, for evaluating and applying operating
experience must be developed, implemented, and maintained.
[[Page 23752]]
(c) Staffing plan. A staffing plan must be developed and comply
with the following:
(1) The staffing plan must include a description of how the
proposed numbers, positions, and qualifications of GLROs, operators, or
senior operators will be sufficient to ensure that plant safety
functions will be maintained across all modes of plant operations. The
staffing plan must be supported by human factors engineering analyses
and assessments.
(2) The staffing plan must include a description of how the
positions and responsibilities of personnel contained within those
plans will adequately satisfy necessary support functions within areas
such as plant operations, equipment surveillance and maintenance,
radiological protection, chemistry control, fire brigades, engineering,
security, and emergency response.
(3) The staffing plan must be approved by the NRC as part of its
approval of the operating license for the plant. The approved staffing
plan is subject to the requirements of Sec. 57.312.
(d) Human factors engineering design requirements. The nuclear
plant design must reflect state-of-the-art human factors engineering
principles for safe and reliable performance in all locations that
operator actions are required to maintain the reactor within the
criterion of Sec. 57.25(a) or locations where a credible operator or
maintenance error could result in exceeding that criterion.
Sec. 57.398 Operator license requirements.
A person must be authorized by a license issued by the Commission
to perform the function of a GLRO, operator, or senior operator, as
defined in this part.
Sec. 57.399 Facility licensee requirements--General.
(a) The facility licensee must maintain the staffing complement
described under its approved staffing plan until such time as the
permanent cessation of operations and permanent removal of fuel from
the reactor vessel has been certified as described under Sec.
57.305(a). The facility licensee must develop, implement, and maintain
facility technical specifications that provide the necessary
administrative controls to ensure the implementation of the approved
staffing complement.
(b) The facility licensee may not permit the manipulation of the
controls of any facility by anyone who is not a GLRO, operator, or
senior operator, as appropriate, except in cases where a non-licensed
operator manipulates the controls under the direction and in the
presence of a GLRO, operator, or senior operator as part of the
individual's training as part of the operator training program or to
load or unload fuel into, out of, or within the reactor vessel while
the reactor is not operating.
(c) Apparatus and mechanisms other than controls, the operation of
which may affect the reactivity or power level of a reactor, must be
manipulated only while plant conditions are being monitored by an
individual who is a GLRO, operator, or senior operator, as appropriate.
(d) Load following operations.
(1) Load following is permitted if at least one of the following is
immediately capable of refusing demands when they could challenge the
safe operation of the plant or when precluded by the plant equipment
conditions:
(i) The actuation of an automatic protection system that utilizes
setpoints more conservative than those otherwise credited for the
purposes of reactor protection;
(ii) An automated control system; or
(iii) GLRO, operator, or senior operator, as appropriate,
(2) The provisions of paragraph (c) of this section do not apply
during load following operations.
(e) Facility licensees must have present during alteration of the
core (including fuel loading or transfer) an individual holding a GLRO
license, a senior operator license, or a senior operator license
limited to fuel handling to directly supervise the activity and, during
this time, the facility licensee must not assign other duties to this
person.
(f) The provisions of paragraph (e) of this section do not apply to
core alterations performed as part of refueling operations while a
facility that is capable of online refueling is operating at power.
(g) A facility licensee may take reasonable action that departs
from a license condition or a technical specification (contained in a
license issued under this part) in an emergency when this action is
immediately needed to protect the public health and safety and no
action consistent with license conditions and technical specifications
that can provide adequate or equivalent protection is immediately
apparent.
(h) Facility licensee action permitted by subparagraph (g) of this
section must be approved, as a minimum, by a GLRO or senior operator,
or, at a nuclear plant for which the certifications required under
Sec. 57.305(a) have been submitted, by either a GLRO or a certified
fuel handler, prior to taking the action.
Sec. 57.400 Facility licensee requirements related to GLROs.
Licensees of operator-independent facilities that have not yet
certified the permanent cessation of operations and permanent removal
of fuel from the reactor vessel as described under Sec. 57.305(a) must
demonstrate compliance with the following requirements:
(a) Ensure that, in addition to being qualified to perform those
items identified by the facility-specific systems approach to training
conducted under Sec. 57.410, GLROs are qualified to safely and
competently--
(i) Perform administrative tasks, including compliance with
technical specifications, and perform operability determinations;
(ii) Implement maintenance and configuration controls;
(iii) Comply with radioactive release limitations;
(iv) Understand plant operating data, including reactor parameters,
and evaluate emergency conditions;
(v) Initiate a reactor shutdown from necessary locations;
(vi) Dispatch and direct operations and maintenance personnel;
(vii) Implement any applicable responsibilities under the facility
emergency plan; and
(viii) Make required notifications to local, State, participating
Tribal, and Federal authorities.
(b) Develop, implement, and maintain the GLRO training,
examination, and proficiency programs required under Sec. 57.410.
(c) Ensure that GLROs are subject to the facility's GLRO training,
examination, and proficiency programs required under Sec. 57.410.
Ensure that GLROs are subject to and comply with the applicable
programmatic requirements for plant personnel required under 10 CFR
parts 26 and 73 of this chapter. An individual that is not in
compliance with any of these programs is not qualified to be in a
position that may involve the manipulation of the controls of the
nuclear plant.
(d) Report annually to the NRC the identity of all GLROs at the
nuclear plant, including all additions and deletions since the previous
report.
(e) Develop, implement, and maintain facility technical
specifications that provide the necessary administrative controls to
ensure the implementation of the requirements of Sec. 57.399(a) and
paragraphs (a) through (d) of this section.
(f) Ensure that the facility design and operation continue to not
rely on
[[Page 23753]]
operator actions to maintain the reactor within the criterion of Sec.
57.25(a).
Sec. 57.405 Generally licensed reactor operators.
(a) Applicability. The requirements of this section apply to each
holder on a GLRO license for an operator-independent facility licensed
under this part.
(b) Requirements.
(1) A general license to manipulate the controls of a facility
licensed under this part and to direct the licensed activities of
generally licensed reactor operators is hereby issued to any individual
employed in a position that may involve the manipulation of the
controls of that facility and who observes the restrictions of this
section.
(2) A GLRO must comply with the operating procedures and other
conditions specified in the license authorizing operation of the
facility.
(3) The general license is limited to the facility or facilities at
which the operator is employed.
(4) The Commission will suspend the general license on an
individual basis for violations of any provision of the AEA or any rule
or regulation issued thereunder whenever the Commission deems such
suspension desirable, including--
(i) For willful violation of, or failure to observe, any of the
terms and conditions of the AEA or the general license, or of any rule,
regulation, or order of the Commission;
(ii) For any conduct determined by the Commission to be a hazard to
safe operation of the facility; or
(iii) For the sale, use, or possession of illegal drugs, or refusal
to participate in the facility drug and alcohol testing program, or a
confirmed positive test for drugs, drug metabolites, or alcohol in
violation of the conditions and cutoff levels established by Sec.
57.405(b)(6) or the consumption of alcoholic beverages where the
individual perform activities requiring a general license, or a
determination of unfitness for scheduled work as a result of the
consumption of alcoholic beverages.
(5) The Commission may require information from a GLRO to determine
whether a general license should be revoked or suspended with respect
to that operator.
(6) The GLRO must not consume or ingest alcoholic beverages in any
location where they perform activities requiring a general license. The
GLRO must not use, possess, or sell any illegal drugs. The GLRO must
not perform activities requiring a general license while under the
influence of alcohol or any prescription, over-the-counter, or illegal
substance that could adversely affect his or her ability to safely and
competently perform these activities. For the purpose of this
paragraph, with respect to alcoholic beverages and drugs, the term
``under the influence'' means the GLRO exceeded, as evidenced by a
confirmed test result, the lower of the cutoff levels for drugs or
alcohol contained in 10 CFR part 26, or as established by the facility
licensee. The term ``under the influence'' also means the GLRO could be
mentally or physically impaired as a result of substance use including
prescription and over-the-counter drugs, as determined under the
provisions, policies, and procedures established by the facility
licensee for its fitness-for-duty program, in such a manner as to
adversely affect his or her ability to safely and competently perform
GLRO duties.
(7) The GLRO must notify the Commission within 30 days about a
conviction for a felony.
(8) The GLRO must complete a training and examination program as
described in Sec. 57.410.
Sec. 57.410 Generally licensed reactor operator training,
examination, and proficiency programs.
(a) Applicability. The requirements of this section apply to each
licensee of an operator-independent facility that has not yet certified
the permanent cessation of operations and permanent removal of fuel
from the reactor as described under Sec. 57.305(a).
(b) Requirements.
(1) The facility licensee must develop, implement, and maintain
training and examination programs that demonstrate compliance with the
requirements of paragraphs (b)(2) and (3) of this section.
(2) The training program must provide for both the initial and
continuing training of GLROs and be derived from a systems approach to
training as defined in Sec. 57.390.
(3) Training and examination program requirements.
(i) The training program must incorporate the instructional
requirements necessary to provide qualified GLROs to operate and
maintain the facility in a safe manner in all modes of operation. The
training program must comply with the facility license, including all
technical specifications and applicable regulations. The facility
licensee must periodically evaluate and revise the training program as
appropriate to reflect industry experience and relevant changes,
including changes to the facility, procedures, regulations, and quality
assurance requirements. Facility licensee management must periodically
review the training program for effectiveness.
(ii) The training program must ensure that GLROs have and maintain
the knowledge, skills, and abilities necessary to operate and maintain
the facility in a safe manner.
(iii) The training program must include the GLROs manipulating the
controls of either the facility or a simulation facility that
demonstrates compliance with the requirements of Sec. 57.410(e).
(iv) The training program must include an initial examination
program for testing a representative sample of the knowledge, skills,
and abilities needed to safely perform GLRO duties, to include both the
examination methods and criteria to be used to assess passing
performance. The facility licensee must provide the opportunity for a
representative of the Commission to be present during initial
examination administration.
(v) The training program must include a requalification examination
program for testing a sample of the topics included under the systems
approach to training and include the examination methods and criteria
to assess passing performance. The requalification examination program
must specify an appropriate periodicity for administering a complete
requalification examination to each GLRO, and the facility licensee
must provide the opportunity for a representative of the Commission to
be present during requalification examination administration.
(A) The facility licensee must ensure that any GLRO who either
demonstrates unsatisfactory performance on, or fails to complete, the
requalification examination is removed from the performance of GLRO
duties until any necessary remedial training has been completed and a
retake examination has been passed.
(B) [Reserved]
(vi) The initial and requalification examination programs must
provide valid and reliable examinations and must be approved by the
Commission prior to their first use.
(c) Records. The following is required regarding the documentation
of the GLRO training and examination programs:
(1) Sufficient records must be maintained by the facility licensee
to maintain the integrity of the programs and kept available for NRC
inspection to verify the adequacy of the programs.
(2) The facility licensee must maintain records documenting the
participation of each GLRO in the
[[Page 23754]]
training and examination programs. The records must contain copies of
examinations administered, the answers given by the GLRO, documentation
of the grading of examinations, and documentation of any additional
training administered in areas in which a GLRO exhibited deficiencies.
The facility licensee must retain these records while the associated
GLROs remain employed at the facility.
(3) Each record required by this part must be legible throughout
the retention period. The record may be the original, a reproduced
copy, or an electronic copy provided that the copy is authenticated by
authorized personnel.
(d) Examination integrity. Generally licensed reactor operators and
facility licensees must not engage in any activity that compromises the
integrity of any examination conducted under the GLRO training and
examination programs. The integrity of an examination is considered
compromised if any activity, regardless of intent, affected or, but for
detection, could have affected the consistent administration of the
examination. This includes all activities related to the preparation,
administration, and grading of examinations.
(e) Simulation facilities.
(1) Simulation facilities used for training purposes, for
maintaining proficiency, or for the conduct of examinations must
demonstrate compliance with the following criteria as they relate to
the facility licensee's reference plant:
(i) The simulation facility must be of sufficient scope and
fidelity for individuals to acquire and demonstrate the necessary
knowledge, skills, and abilities to safely perform GLRO duties.
(ii) The simulation facility must utilize models relating to
nuclear, thermal-hydraulic, and other applicable design-specific
characteristics that either replicate the most recent fuel load in the
reference nuclear plant or, prior to initial fuel load (or, for a
fueled manufactured reactor, prior to initiating the removal of the
features to prevent criticality), replicate the intended initial fuel
load for the reference nuclear plant, with the exception of those
portions of the simulation facility that utilize the reference plant
itself.
(iii) Simulator fidelity must be demonstrated so that significant
control manipulations are completed without procedural exceptions,
simulator performance exceptions, or deviation from the approved
training scenario sequence.
(2) Facility licensees that maintain a simulation facility for
training purposes, for maintaining proficiency, or for the conduct of
examinations must--
(i) Conduct performance testing throughout the life of the
simulation facility in a manner sufficient to ensure that paragraph
(e)(1) of this section is met;
(ii) Retain the results of performance testing for 4 years after
the completion of each performance test or until superseded by updated
test results;
(iii) Promptly correct modeling and hardware discrepancies and
discrepancies identified from scenario validation and from performance
testing or provide justification for why the presence of such
discrepancies will not adversely affect the criteria of paragraph
(e)(1) of this section;
(iv) Make the results of any uncorrected performance test failures
that may exist at the time of an inspection available for NRC review;
and
(v) Maintain the provisions for examination integrity consistent
with Sec. 57.410(d).
(f) Waiver of examination requirement. The facility licensee may
waive any or all the requirements for an examination in accordance with
the facility licensee's Commission-approved GLRO examination program.
(g) Proficiency. The facility licensee must develop, implement, and
maintain a proficiency program to allow GLROs to maintain proficiency
regarding position functions and familiarity with plant status. This
program must include those steps that will be taken to re-establish
proficiency when it cannot be maintained.
Sec. 57.415 Cessation of individual applicability.
The general license ceases to be applicable on an individual basis
once a GLRO is no longer being employed in a position that may involve
the manipulation of the controls of the operator-independent facility.
Sec. 57.420 Operator licensing for operator-dependent facilities.
(a) Applicability. Sections 57.420 through 57.427 address operator
and senior operator licensing requirements. The regulations within
these sections are applicable to those applicants for or holders of
operating licenses under this part for operator-dependent facilities
that have not yet certified the permanent cessation of operations and
permanent removal of fuel from the reactor vessel as described under
Sec. 57.305(a).
(b) [Reserved]
Sec. 57.421 Medical requirements.
(a) An applicant for an operator or senior operator license must
have a medical examination by a physician or other licensed medical
examiner. An operator or senior operator must have a medical
examination by a physician or other licensed medical examiner every 2
years. The physician or other licensed medical examiner shall determine
that the applicant or licensee meets the requirements of Sec.
57.423(b)(1)(i).
(b) To certify the medical fitness of an applicant for an operator
or senior operator license, an authorized representative of the
facility licensee must complete and sign NRC Form 396, ``Certification
of Medical Examination by Facility Licensee,'' which can be obtained by
writing the Office of the Chief Information Officer, U.S. Nuclear
Regulatory Commission, Washington, DC 20555-0001, by calling 301-415-
7232, or by visiting the NRC's website at https://www.nrc.gov and
selecting forms from the index found on the home page, or by other
means provided by the NRC.
(1) NRC Form 396 must certify that a physician or other licensed
medical examiner has conducted the medical examination of the applicant
as required in paragraph (a) of this section.
(2) When the medical certification requests a conditional license
based on medical evidence, the medical evidence must be submitted on
NRC Form 396 to the Commission to enable the Commission to make a
determination in accordance with Sec. 57.425(b).
(c) The facility licensee must document and maintain the results of
medical qualifications data, test results, and each operator's or
senior operator's medical history for the current license period and
provide the documentation to the Commission upon request. The facility
licensee must retain this documentation while an individual performs
the functions of an operator or senior operator.
Sec. 57.422 Incapacitation because of disability or illness.
If, during the term of the operator or senior operator license, the
licensee develops a permanent physical or mental condition that causes
the licensee to fail to demonstrate compliance with the requirements of
Sec. 57.423(b)(1)(i), the facility licensee must notify the Commission
within 30 days of learning of the diagnosis. For conditions for which a
conditional license (as described in Sec. 57.423(b)) is requested, the
facility licensee must provide medical certification on NRC Form 396 to
the Commission (as described in Sec. 57.421(b)).
[[Page 23755]]
Sec. 57.423 Applications for operators and senior operators.
(a) How to apply.
(1) The applicant for an operator or senior operator license must--
(i) Complete NRC Form 398, ``Personal Qualification Statement--
Licensee,'' which can be obtained by writing the Office of the Chief
Information Officer, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001, by calling 301-415-5877, or by visiting the NRC's website
at https://www.nrc.gov and selecting forms from the index found on the
home page, or by other means provided by the NRC;
(ii) File an original of NRC Form 398, or an equivalent electronic
submittal, together with the information required in paragraphs
(a)(1)(iii) and (a)(1)(iv) of this section, with the appropriate
Regional Administrator.
(iii) Provide evidence that the applicant, as a trainee, has
successfully demonstrated competence in manipulating the controls of
either the facility for which a license is sought or a simulation
facility that demonstrates compliance with the requirements of Sec.
57.424(e). For operators applying for a senior operator license,
certification that the operator has successfully operated the controls
of the facility as an operator will be accepted; and
(iv) Provide certification by the facility licensee of medical
condition and general health on NRC Form 396, to comply with Sec.
57.421.
(2) The Commission may at any time after the application has been
filed, and before the license has expired, require further information
under oath or affirmation to enable it to determine whether to grant or
deny the application or whether to revoke, modify, or suspend the
license.
(3) An applicant whose application has been denied because of a
medical condition or their general health may submit a further medical
report at any time as a supplement to the application.
(4) Each application and statement must contain complete and
accurate disclosure as to all matters required to be disclosed. The
applicant must sign statements required by paragraphs (a)(1)(i) and
(a)(1)(ii) of this section.
(b) Disposition of an initial application.
(1) License approval. The Commission will approve an initial
application if it finds that the following criteria are met:
(i) Health. The applicant's medical condition and general health
will not adversely affect the performance of assigned operator or
senior operator job duties or cause operational errors endangering
public health and safety. The Commission will base its finding upon the
certification by the facility licensee as detailed in Sec. 57.421(b).
(ii) Examination. The applicant has passed the requisite
examination in accordance with Sec. 57.424(b). The examination
determines whether the applicant for an operator's or senior operator's
license has learned to operate a facility competently and safely, and,
in the case of a senior operator, whether the applicant has learned to
supervise the licensed activities of operators competently and safely.
(2) Conditional license. If an applicant's general medical
condition does not demonstrate compliance with the minimum standards
under Sec. 57.423(b)(1)(i), the Commission may approve the application
and include conditions in the license to accommodate the medical
condition. The Commission will consider the recommendations and
supporting evidence of the facility licensee and of the examining
physician (provided on NRC Form 396) in arriving at its decision.
(c) Re-applications.
(1) An applicant whose application for a license has been denied
because of failure to pass the examination may file a new application.
The application must be submitted on NRC Form 398 and include a
statement signed by an authorized representative of the facility
licensee by whom the applicant will be employed that states in detail
the extent of the applicant's additional training and remediation since
the denial and certifies that the applicant is ready for re-
examination.
(2) An applicant who has passed a portion of the examination and
failed another may request in a new application on NRC Form 398 to be
excused from re-examination on the portions of the examination that the
applicant has passed. The Commission may in its discretion grant the
request if it determines that sufficient justification is presented.
Sec. 57.424 Training, examination, and proficiency program.
(a) Operator licensing initial training program.
(1) A program that is based upon a systems approach to training, as
defined by Sec. 57.390, must be utilized for the training of
applicants for operator and senior operator licenses. The program must
ensure that applicants at the facility will possess the knowledge,
skills, and abilities necessary to protect public health and safety and
maintain plant safety functions specific to the facility design. The
program must be approved by the Commission prior to its use for
training applicants.
(2) The facility licensee must maintain operator licensing initial
training program records documenting the initial operator licensing
training administered and completed by each applicant. The facility
licensee must retain these records during the period in which any
trainees subsequently remain licensed as operators or senior operators
at the facility.
(b) Operator licensing initial examination program.
(1) The facility licensee must establish and implement an
examination program for testing a representative sample of the
knowledge, skills, and abilities needed to safely perform operator and
senior operator duties, to include both the examination methods and
criteria to be used to assess passing performance. The program must
provide for valid and reliable examinations and be approved by the
Commission prior to its use for examining applicants.
(2) The facility licensee must submit prepared examinations to the
Commission for review and approval in advance of their administration.
(3) The Commission will either administer an approved examination
or allow the facility licensee to administer the examination. The
facility licensee must ensure that sufficient advance notification is
provided to the Commission to either administer the examination or
allow for a representative of the Commission to be afforded the
opportunity to be present when the facility licensee administers the
examination.
(4) Graded examination documentation for each applicant must be
provided to the Commission for review in making operator licensing
decisions.
(5) The facility licensee must maintain operator licensing initial
examination program records documenting the participation of each
operator and senior operator applicant in the initial examination. The
records must contain copies of examinations administered, the answers
given by the applicant, documentation of the grading of examinations,
and documentation of any additional training administered in areas in
which an applicant exhibited deficiencies. The facility licensee must
retain these records during the period in which the associated
operators or senior operators remain licensed at the facility.
(c) Operator licensing requalification program.
(1) A program based upon a systems approach to training must be
utilized for the continuing training of operators and senior operators.
[[Page 23756]]
(i) The program must ensure that operators and senior operators at
the facility maintain the knowledge, skills, and abilities necessary to
protect the public health and safety and maintain plant safety
functions specific to the facility design. The program must be
conducted for a continuous period not to exceed 24 months in duration.
(ii) The program must be approved by the Commission prior to its
use for continuing training and implemented upon commencing the
administration of initial examinations under the operator licensing
examination program required under Sec. 57.424(b).
(2) The following requirements apply to operator licensing
requalification programs:
(i) The facility licensee must propose a requalification
examination program for testing, for each requalification period, a
sample of the topics included under the systems approach to training,
to include both the examination methods and criteria to be used to
assess passing performance. The program must provide for valid and
reliable examinations and be approved by the Commission prior to its
use for examining operators and senior operators.
(ii) The following requirements apply to the requalification
examination program:
(A) The facility licensee must make prepared requalification
examinations available to the Commission for review.
(B) The facility licensee must ensure that a representative of the
Commission is afforded the opportunity to be present during
requalification examination administration.
(C) The facility licensee must ensure that each operator and senior
operator is administered a complete requalification examination on a
periodicity not to exceed 24 months. Additionally, the facility
licensee must ensure that any operator or senior operator who either
demonstrates unsatisfactory performance on, or fails to complete, this
biennial requalification examination is removed from the performance of
operator and senior operator duties until any necessary remedial
training has been completed and a retake examination has been passed.
(D) The facility licensee must promptly provide a summary of
examination results to the NRC for each operator and senior operator
following the completion of the requalification examination.
(3) The facility licensee must maintain operator licensing
requalification program records documenting the participation of each
operator and senior operator in the requalification program. The
records must contain copies of examinations administered, the answers
given by the operator or senior operator, documentation of the grading
of examinations, and documentation of any additional training
administered in areas in which an operator or senior operator exhibited
deficiencies. The facility licensee must retain these records until the
operator's or senior operator's license is renewed.
(d) Examination integrity. Applicants, operators, senior operators,
and facility licensees must not engage in any activity that compromises
the integrity of any application or examination required by Sec. Sec.
57.420 through 57.427. The integrity of an examination is considered
compromised if any activity, regardless of intent, affected or, but for
detection, could have affected the consistent administration of the
examination. This includes activities related to the preparation and
certification of applications and all activities related to the
preparation, administration, and grading of examinations required by
Sec. Sec. 57.420 through 57.427.
(e) Simulation facilities.
(1) This section addresses the use of a simulation facility for the
administration of examinations, for training, or to demonstrate
compliance with experience requirements for applicants for operator and
senior operator licenses.
(2) Simulation facilities used for training purposes, for
demonstrating compliance with experience requirements, or for the
conduct of examinations under Sec. 57.424(b) and (c) must demonstrate
compliance with the following criteria as they relate to the facility
licensee's reference plant:
(i) The simulation facility must be of sufficient scope and
fidelity for individuals to acquire and demonstrate the necessary
knowledge, skills, and abilities to safely perform operator and senior
operator duties.
(ii) The simulation facility must utilize models relating to
nuclear, thermal-hydraulic, and other applicable design-specific
characteristics that either replicate the most recent fuel load in the
reference nuclear plant or, prior to initial fuel load (or, for a
fueled manufactured reactor, prior to initiating the removal of the
features to prevent criticality), replicate the intended initial fuel
load for the reference nuclear plant, with the exception of those
portions of the simulation facility that utilize the reference plant
itself.
(iii) Simulation facility fidelity must be demonstrated so that
significant control manipulations are completed without procedural
exceptions, simulator performance exceptions, or deviation from the
approved training scenario sequence.
(3) Facility licensees that maintain a simulation facility that has
been approved by the Commission for training purposes, demonstrating
compliance with experience requirements, or the conduct of examinations
under Sec. 57.424(b) and (c) for the facility licensee's reference
plant must:
(i) Conduct performance testing throughout the life of the
simulation facility in a manner sufficient to ensure that paragraph
(e)(2) of this section is met;
(ii) Retain the results of performance testing for 4 years after
the completion of each performance test or until superseded by updated
test results;
(iii) Promptly correct modeling and hardware discrepancies and
discrepancies identified from scenario validation and performance
testing or provide justification as to why the presence of such
discrepancies will not adversely affect simulator performance with
respect to the criteria of paragraph (e)(2) of this section;
(iv) Make the results of any uncorrected performance test failures
that may exist at the time of the initial license examination or
requalification examination available for NRC review, prior to or
concurrent with preparations for each initial license examination or
requalification examination; and
(v) Maintain the provisions for license application and examination
integrity consistent with Sec. 57.424(d).
(4) A simulation facility must demonstrate compliance with the
requirements of paragraphs (e)(2) and (e)(3) of this section for the
Commission to accept the simulation facility for conducting initial
examinations as described in Sec. 57.424(b), requalification training
as described in Sec. 57.424(c), or performing control manipulations
that affect reactivity to establish eligibility for an operator or
senior operator license as described in Sec. 57.423(a).
(f) Waiver of examination requirement. On application, the
Commission may waive any or all of the requirements for an initial
licensing examination if it finds that the applicant has demonstrated
the required knowledge, skills, and abilities to safely operate the
plant, and is capable of continuing to do so. The Commission may make
such a finding based on demonstration of the following:
(1) Recent operating experience at a comparable facility;
[[Page 23757]]
(2) Proof of the applicant's past competent and safe performance;
and
(3) Proof of the applicant's current qualifications.
(g) Proficiency. The facility licensee must develop, implement, and
maintain a proficiency program to ensure that operators and senior
operators will actively perform the functions of an operator or senior
operator, respectively, as needed to maintain proficiency with on-shift
duties and familiarity with plant status. This program must include
those steps that will be taken to re-establish proficiency when it
cannot be maintained. This program must be approved by the Commission
as part of its approval of the operating license for the plant.
(h) Records. Each record required by this section must be legible
throughout the retention period specified by each Commission
regulation. The record may be the original, a reproduced copy, or an
electronic copy provided that the copy is authenticated by authorized
personnel.
Sec. 57.425 Conditions of operator and senior operator licenses.
Each operator and senior operator license contains and is subject
to the following conditions whether stated in the license or not:
(a) Neither the license nor any right under the license may be
assigned or otherwise transferred.
(b) The license is limited to the facility or facilities for which
it is issued.
(c) The license is limited to those controls of the facility or
facilities specified in the license.
(d) The license is subject to, and the licensee must observe, all
applicable rules, regulations, and orders of the Commission.
(e) The licensee must maintain or re-establish proficiency in
accordance with the facility licensee's Commission-approved proficiency
program required under Sec. 57.424(g).
(f) The licensee must be subject to the facility's Commission-
approved operator licensing requalification and requalification
examination programs required under Sec. 57.424(c).
(g) The licensee must have a biennial medical examination as
described by Sec. 57.421.
(h) The licensee must notify the Commission within 30 days about a
conviction for a felony.
(i) The licensee must not consume or ingest alcoholic beverages
within the protected area of nuclear plants. The licensee must not use,
possess, or sell any illegal drugs. The licensee must not perform
activities authorized by a license issued under this part while under
the influence of alcohol or any prescription, over-the-counter, or
illegal substance that could adversely affect his or her ability to
safely and competently perform his or her licensed duties. For the
purpose of this paragraph (i), with respect to alcoholic beverages and
drugs, the term ``under the influence'' means the licensee exceeded, as
evidenced by a confirmed test result, the lower of the cutoff levels
for drugs or alcohol contained in 10 CFR part 26, or as established by
the facility licensee. The term ``under the influence'' also means the
licensee could be mentally or physically impaired as a result of
substance use including prescription and over-the-counter drugs, as
determined under the provisions, policies, and procedures established
by the facility licensee for its fitness-for-duty program, in such a
manner as to adversely affect his or her ability to safely and
competently perform licensed duties.
(j) Each licensee must participate in the drug and alcohol testing
programs as required under 10 CFR part 26.
(k) The licensee must comply with any other conditions that the
Commission may impose to protect health or to minimize danger to life
or property.
Sec. 57.426 Issuance, modification, and revocation of operator and
senior operator licenses.
(a) Issuance of operator and senior operator licenses. If the
Commission determines that an applicant for an operator license or a
senior operator license demonstrates compliance with the requirements
of the AEA and its regulations, it will issue a license in the form and
containing any conditions and limitations it considers appropriate and
necessary.
(b) Modification and revocation of operator and senior operator
licenses.
(1) The terms and conditions of all operator and senior operator
licenses are subject to amendment, revision, or modification by reason
of rules, regulations, or orders issued in accordance with the AEA or
any amendments thereto.
(2) Any license may be revoked, suspended, or modified, in whole or
in part--
(i) For any material false statement in the application or in any
statement of fact required under section 182 of the AEA;
(ii) Because of conditions revealed by the application or statement
of fact or any report, record, inspection, or other means that would
warrant the Commission to refuse to grant a license on an original
application;
(iii) For willful violation of, or failure to observe, any of the
terms and conditions of the AEA or the license, or of any rule,
regulation, or order of the Commission;
(iv) For any conduct determined by the Commission to be a hazard to
safe operation of the facility; or
(v) For the sale, use, or possession of illegal drugs, or refusal
to participate in the facility drug and alcohol testing program, or a
confirmed positive test for drugs, drug metabolites, or alcohol in
violation of the conditions and cutoff levels established by Sec.
57.425(i) or the consumption of alcoholic beverages within the
protected area of nuclear plants, or a determination of unfitness for
scheduled work as a result of the consumption of alcoholic beverages.
Sec. 57.427 Expiration of operator and senior operator licenses.
Each operator license and senior operator license expires upon
termination of employment with the facility licensee, or upon
determination by the facility licensee that the licensed individual no
longer needs to maintain a license. The facility licensee shall notify
the Commission, as described in Sec. 57.392, within 30 days of either
occurrence. An operator license or senior operator license also expires
upon the Commission's determination that a licensed individual's
general medical condition does not meet the minimum standards under
Sec. 57.423(b)(1)(i) and that the medical condition cannot be
accommodated.
Sec. 57.429 Training and qualification for non-licensed personnel.
(a) The regulations within this section address personnel training
requirements and are applicable to all applicants for or holders of an
operating license under this part.
(b) Prior to initial fuel load (or, for a fueled manufactured
reactor, prior to initiating the removal of the features to prevent
criticality), each holder of an operating license under this part must,
with sufficient time to provide trained and qualified personnel to
operate the facility, establish, implement, and maintain a training
program that demonstrates compliance with the requirements of
paragraphs (c) and (d) of this section.
(c) The training program must be derived from a systems approach to
training as defined in Sec. 57.390 and must provide, at a minimum, for
the training and qualification of the following categories of nuclear
plant personnel:
(1) Supervisors (e.g., shift supervisors);
[[Page 23758]]
(2) Technicians (e.g., maintenance, chemistry, and radiological);
and
(3) Other appropriate operating personnel (e.g., auxiliary
operators and certified fuel handlers).
(d) The training program must incorporate the instructional
requirements necessary to provide qualified personnel to operate
components of a nuclear plant and maintain the facility in a safe
manner in all modes of operation. The training program must be
developed to be in compliance with the facility license, including all
technical specifications and applicable regulations.
(1) The training program must be periodically evaluated and revised
as appropriate to reflect industry experience and relevant changes,
including changes to the facility, procedures, regulations, and quality
assurance requirements. The training program must be periodically
reviewed by facility licensee management for effectiveness.
(2) Sufficient records must be maintained by the facility licensee
to maintain program integrity and kept available for NRC inspection to
verify the adequacy of the training program.
Subpart Q--Reporting and Other Administrative Requirements
Sec. 57.430 Maintenance of records, making of reports.
(a) Each holder of a manufacturing license, operating license, or
construction permit must maintain all records and make all reports, in
connection with the activity, as may be required by the conditions of
the license or permit or by the regulations and orders of the
Commission in effectuating the purposes of the AEA and the Energy
Reorganization Act of 1974, as amended. Reports must be submitted in
accordance with Sec. 57.4.
(b) Records that are required by this part, by license condition,
or by technical specifications must be retained for the period
specified by the appropriate regulation, license condition, or
technical specification. If a retention period is not otherwise
specified, these records must be retained until the Commission
terminates the facility license.
(c) Records that must be retained under this part may be the
original or a reproduced copy or a microform if the reproduced copy or
microform is duly authenticated by authorized personnel and the
microform is capable of producing a clear and legible copy after
storage for the period specified by Commission regulations. The record
may also be stored in electronic media with the capability of producing
legible, accurate, and complete records during the required retention
period. Records such as letters, drawings, and specifications, must
include all pertinent information such as stamps, initials, and
signatures. The licensee must maintain adequate safeguards against
tampering with and loss of records.
(d) Each licensee must keep records of information important to the
decommissioning of the facility in accordance with the requirements of
10 CFR 50.75(g).
(e) If there is a conflict between the Commission's regulations in
this part, license condition, or technical specification, or other
written Commission approval or authorization pertaining to the
retention period for the same type of record, the retention period
specified in the regulations of this part for such records must apply
unless the Commission, pursuant to Sec. 57.9 of this part, has granted
a specific exemption from the record retention requirements in the
regulations of this part.
(f) Each licensee must notify the Commission as specified in Sec.
57.4, of successfully completing startup testing, as applicable, within
30 calendar days of completing the testing.
Sec. 57.435 Reporting requirements.
(a) Reporting methods. Licensees under this part must make reports
required by paragraphs (b) and (c) of this section by telephone or any
other method that will ensure that a report is made as soon as possible
to the NRC Headquarters Operations Center at the numbers specified in
appendix A to part 73 of this chapter.
(b) Events for notification--
(1) One-hour reports. The licensee must notify the NRC as soon as
possible and in all cases within 1 hour of the occurrence of any of the
following:
(i) Any event resulting in activation of the emergency plan.
(ii) Any deviation from the plant's Technical Specifications
authorized pursuant to Sec. 57.399(g) of this part.
(2) Four-hour reports. If not reported under paragraph (b)(1) of
this section, the licensee must notify the NRC as soon as possible, and
in all cases, within 4 hours of the occurrence of any of the following:
(i) The initiation of any nuclear plant shutdown required by the
plant's Technical Specifications.
(ii) Any event or condition that results in actuation of the
reactor protection system when the reactor is critical except when the
actuation results from and is part of a pre-planned sequence during
testing or reactor operation.
(iii) Any event or condition that results in an unplanned actuation
of a safety-related cooling system.
(iv) Any event or condition that results in an unplanned movement
of, change of state in, or chemical interaction involving a significant
amount of radioactive material within the nuclear plant.
(v) Any event or situation, related to the health and safety of the
public or onsite personnel, or protection of the environment, for which
a news release is planned or notification to other government agencies
has been or will be made. Such an event may include an onsite fatality
or inadvertent release of radioactively contaminated materials.
(3) Eight-hour reports. If not reported under paragraphs (b)(1) or
(b)(2) of this section, the licensee must notify the NRC as soon as
possible and in all cases within 8 hours of the occurrence of any of
the following:
(i) Any event or condition that results in--
(A) The condition of the nuclear plant, including its principal
safety barriers, being seriously degraded; or
(B) The nuclear plant being in an unanalyzed condition that
significantly degrades plant safety.
(ii) Any event or condition that results in valid actuation of a
safety-related system, except when the actuation results from and is
part of a pre-planned sequence during testing or reactor operation.
(iii) Any event or condition that at the time of discovery could
have prevented the fulfillment of the safety function of structures or
systems that are needed to--
(A) Shut down the reactor and maintain it in a safe shutdown
condition;
(B) Remove residual heat;
(C) Control the release of radioactive material; or
(D) Mitigate the consequences of an accident.
(iv) Events covered in paragraph (b)(3)(iii) of this section may
include one or more procedural errors, equipment failures, and/or
discovery of design, analysis, fabrication, construction, and/or
procedural inadequacies. However, individual component failures need
not be reported pursuant to paragraph (b)(3)(iii) of this section if
redundant equipment in the same system was operable and available to
perform the required safety function.
(v) Any event requiring the transport of a radioactively
contaminated person to an offsite medical facility for treatment.
[[Page 23759]]
(c) Follow-up notification: With respect to the notifications made
under paragraph (b) of this section, in addition to making the required
initial notification, each licensee must, during the course of the
event--
(1) Immediately report:
(i) Any further degradation in the level of safety of the plant or
other worsening plant conditions, including those that require
activation of the emergency plan, if such a declaration has not been
previously made,
(ii) Any escalation in emergency response measures has been
necessitated, and
(iii) Termination of an emergency event.
(2) Immediately Report:
(i) The results of ensuing evaluations or assessments of plant
conditions,
(ii) The effectiveness of response or protective measures taken,
and
(iii) Important information related to plant behavior that is not
understood.
(3) Maintain an open, continuous communication channel with the NRC
Operation Center upon request by the NRC. *Other requirements for
immediate notification of the NRC by licensed operating nuclear plants
are contained elsewhere in this chapter, in particular, Sec. Sec.
20.1906, 20.2202, 72.216, 73.71, and 73.77 of this chapter.
Sec. 57.440 Licensee event report system.
(a) Reportable events.
(1) Each licensee holding an operating license under this part must
submit a licensee event report for any event of the type described in
this section within 60 days after discovery of the event. In the case
of an invalid actuation reported under Sec. 57.440(a)(2)(iv)(B), other
than automatic reactor shutdown when the reactor is critical, the
licensee may, at its option, provide a telephone notification to the
NRC Operations Center within 60 days after discovery of the event
instead of submitting a written licensee event report. Unless otherwise
specified in this section, the licensee must report an event if it
occurred within 3 years of the date of discovery regardless of the
plant mode or power level, and regardless of the significance of the
structure, system, or component that initiated the event.
(2) The licensee must report--
(i) The completion of any nuclear plant shutdown required by the
plant's Technical Specifications.
(ii) Any operation or condition that was prohibited by the plant's
Technical Specifications except when--
(A) The Technical Specification is administrative in nature;
(B) The event consisted solely of a case of a late surveillance
test where the oversight was corrected, the test was performed, and the
equipment was found to be capable of performing its specified safety
functions; or
(C) The Technical Specification was revised prior to discovery of
the event such that the operation or condition was no longer prohibited
at the time of the event.
(iii) Any deviation from the plant's Technical Specifications
authorized pursuant to Sec. 57.399(g) of this part.
(iv) Any event or condition that resulted in--
(A) The condition of the nuclear plant, including its principal
safety barriers, being seriously degraded; or
(B) The nuclear plant being in an unanalyzed condition that
significantly degraded plant safety.
(v) Any natural phenomena or other external condition that posed an
actual threat to the safety of the nuclear plant or significantly
hampered site personnel in the performance of duties necessary for the
safe operation of the nuclear plant.
(vi) Any event or condition that resulted in manual or automatic
actuation of a safety-related system, except when--
(A) The actuation resulted from and was part of a pre-planned
sequence during testing; or
(B) The actuation was invalid and--
(1) Occurred while the system was properly removed from service; or
(2) Occurred after the safety function had been already completed.
(vii) Any event or condition that could have prevented the
fulfillment of the safety function of structures or systems that are
needed to--
(A) Shut down the reactor and maintain it in a safe shutdown
condition;
(B) Remove residual heat;
(C) Control the release of radioactive material; or
(D) Mitigate the consequences of an accident.
(viii) Events covered in paragraph (a)(2)(v) of this section may
include one or more procedural errors, equipment failures, and/or
discovery of design, fabrication, construction, and/or procedural
inadequacies. However, individual component failures need not be
reported pursuant to paragraph (a)(2)(v) of this section if any other
equipment was operable and available to perform the required safety
function.
(ix) Any event where a single cause or condition caused at least
one independent train or channel to become inoperable in multiple
systems or two independent trains or channels to become inoperable in a
single system designed to--
(A) Shut down the reactor and maintain it in a safe shutdown
condition;
(B) Remove residual heat;
(C) Control the release of radioactive material; or
(D) Mitigate the consequences of an accident.
(x) Any of the following types of releases--
(A) Airborne radioactive release that, when averaged over a time
period of 1 hour, resulted in airborne radionuclide concentrations in
an unrestricted area that exceeds 20 times the applicable concentration
limits specified in appendix B to part 20 of this chapter, table 2,
column 1.
(B) Liquid effluent release that, when averaged over a time period
of 1 hour, exceeds 20 times the applicable concentrations specified in
appendix B to part 20 of this chapter, table 2, column 2, at the point
of entry into the receiving waters (i.e., unrestricted area) for all
radionuclides except tritium and dissolved noble gases.
(xi) Any event or condition that as a result of a single cause
could have prevented the fulfillment of a safety function for two or
more trains or channels in different systems that are needed to--
(A) Shut down the reactor and maintain it in a safe shutdown
condition;
(B) Remove residual heat;
(C) Control the release of radioactive material; or
(D) Mitigate the consequences of an accident.
(xii) Events covered in paragraph (a)(2)(ix)(A) of this section may
include cases of procedural error, equipment failure, and/or discovery
of a design, analysis, fabrication, construction, and/or procedural
inadequacy. However, licensees are not required to report an event
pursuant to paragraph (a)(2)(ix)(A) of this section if the event
results from--
(A) A shared dependency among trains or channels that is a natural
or expected consequence of the approved plant design; or
(B) Normal and expected wear or degradation.
(xiii) Any event that posed an actual threat to the safety of the
nuclear plant or significantly hampered site personnel in the
performance of duties necessary for the safe operation of the plant,
including fires, toxic gas releases, or radioactive releases.
(b) Contents. The licensee event report must contain--
(1) A brief abstract describing the major occurrences during the
event, including all component or system failures that contributed to
the event
[[Page 23760]]
and significant corrective action taken or planned to prevent
recurrence.
(2) A specific description of the event as follows:
(i) A clear, specific narrative description of what occurred so
that knowledgeable readers conversant with the design of nuclear
plants, but not familiar with the details of a particular plant, can
understand the complete event.
(ii) The narrative description must include the following specific
information as appropriate for the particular event:
(A) Plant operating conditions before the event.
(B) Status of structures, components, or systems that were
inoperable at the start of the event and that contributed to the event.
(C) Dates and approximate time of the occurrences.
(D) The cause of each component or system failure or personnel
error, if known.
(E) The failure mode, mechanism, and effect of each failed
component, if known.
(F) For failures of components with multiple functions, include a
list of systems or secondary functions that were also affected.
(G) For failure that rendered a train of a safety system
inoperable, an estimate of the elapsed time from the discovery of the
failure until the train was returned to service.
(H) The method of discovery of each component or system failure or
procedural error.
(I) For each human performance related root cause, the licensee
must discuss the cause(s) and circumstances.
(J) Automatically and manually initiated safety system responses.
(K) The manufacturer and model number (or other identification) of
each component that failed during the event.
(3) An assessment of the safety consequences and implications of
the event. This assessment must include--
(i) The availability of systems or components that could have
performed the same function as the components and systems that failed
during the event, and
(ii) For events that occurred when the reactor was shut down, the
availability of systems or components that are needed to shut down the
reactor and maintain safe shutdown conditions, remove residual heat,
control the release of radioactive material, or mitigate the
consequences of an accident.
(4) A description of any corrective actions planned as a result of
the event, including those to reduce the likelihood of similar events
occurring in the future.
(5) Reference to any previous similar events at the same plant that
are known to the licensee.
(6) The name and contact information of a person within the
licensee's organization who is knowledgeable about the event and can
provide additional information concerning the event and the plant's
characteristics.
(c) Supplemental Information: The Commission may require the
licensee to submit specific additional information beyond that required
by paragraph (b) of this section if the Commission finds that
supplemental material is necessary for complete understanding of an
unusually complex or significant event. These requests for supplemental
information will be made in writing and the licensee must submit, as
specified in Sec. 57.4, the requested information as a supplement to
the initial licensee event report.
(d) Submission of Reports: Licensee event reports must be prepared
on Form NRC 366 and submitted to the NRC, as specified in Sec. 57.4.
(e) Report Legibility: The reports and copies that licensees are
required to submit to the Commission under the provisions of this
section must be of sufficient quality to permit legible reproduction
and micrographic processing.
Sec. 57.445 Reports of radiation exposure to members of the public.
(a) Each holder of an operating license must submit a report to the
Commission annually that specifies the quantity of each of the
principal radionuclides released to unrestricted areas in liquid and in
gaseous effluents during the previous 12 months. In addition, the
report must include an estimate of the dose received by the maximally
exposed member of the public in an unrestricted area from effluents and
direct radiation from contained sources during the previous 12 months
and include any other information as may be required by the Commission
to estimate maximum potential annual radiation doses to the public. If
the TEDE to members of the public in unrestricted areas during the
reporting period is greater than 10 mrem/year TEDE, the report must
specify the causes for exceedance and describe any corrective actions.
(b) The reports required by this section must be submitted as
specified in Sec. 57.4, and the time between submission of the reports
must be no longer than 12 months.
PART 70--DOMESTIC LICENSING OF SPECIAL NUCLEAR MATERIAL
0
95. The authority citation for 10 CFR part 70 continues to read as
follows:
Authority: Atomic Energy Act of 1954, secs. 51, 53, 57(d), 108,
122, 161, 182, 183, 184, 186, 187, 193, 223, 234, 274, 1701 (42
U.S.C. 2071, 2073, 2077(d), 2138, 2152, 2201, 2232, 2233, 2234,
2236, 2237, 2243, 2273, 2282, 2021, 2297f); Energy Reorganization
Act of 1974, secs. 201, 202, 206, 211 (42 U.S.C. 5841, 5842, 5846,
5851); Nuclear Waste Policy Act of 1982, secs. 135, 141 (42 U.S.C.
10155, 10161); 44 U.S.C. 3504 note.
Sec. 70.20a [Amended]
0
96. In Sec. 70.20a, in paragraph (b) add the number ``57,'' in
sequential order.
0
97. In Sec. 70.22, revise paragraphs (b), (h)(1), (j)(1), and (k) to
read as follows:
Sec. 70.22 Contents of applications.
* * * * *
(b) Each application for a license to possess special nuclear
material, to possess equipment capable of enriching uranium, to operate
an uranium enrichment facility, to possess and use at any one time and
location special nuclear material in a quantity exceeding one effective
kilogram, except for applications for use as sealed sources and for
those uses involved in the operation of a nuclear reactor licensed
pursuant to part 50 or part 57 of this chapter and those involved in a
waste disposal operation, must contain a full description of the
applicant's program for control and accounting of such special nuclear
material or enrichment equipment that will be in the applicant's
possession under license to show how compliance with the requirements
of Sec. 74.31, 74.33, 74.41, or 74.51 of this chapter, as applicable,
will be accomplished.
* * * * *
(h)(1) Each application for a license to possess or use, at any
site or contiguous sites subject to licensee control, a formula
quantity of strategic special nuclear material, as defined in Sec.
70.4, other than a license for possession or use of this material in
the operation of a nuclear reactor licensed pursuant to part 50 or part
57 of this chapter, must include a physical security plan. The plan
must describe how the applicant will meet the applicable requirements
of part 73 of this chapter in the conduct of the activity to be
licensed, including the identification and description of jobs as
required by 10 CFR 11.11(a). The plan must list tests, inspections,
audits, and other means to be used to demonstrate compliance with the
requirements of 10 CFR parts 11 and 73, if applicable.
* * * * *
(j)(1) Each application for a license to possess or use at any site
or contiguous sites subject to control by the licensee uranium-235
(contained in uranium enriched to 20 percent or more in the uranium-235
isotope), uranium-233, or
[[Page 23761]]
plutonium alone or in any combination in a quantity of 5,000 grams or
more computed by the formula, grams = (grams contained U--235) + 2.5
(grams U-233 + grams plutonium) other than a license for possession or
use of this material in the operation of a nuclear reactor licensed
pursuant to part 50 or part 57 of this chapter, must include a licensee
safeguards contingency plan for dealing with threats, thefts, and
radiological sabotage, as defined in part 73 of this chapter, relating
to nuclear facilities licensed under part 50 of this chapter or to the
possession of special nuclear material licensed under this part.
* * * * *
(k) Each application for a license to possess or use at any site or
contiguous sites subject to licensee control, special nuclear material
of moderate strategic significance or 10 kg or more of special nuclear
material of low strategic significance as defined under Sec. 70.4,
other than a license for possession or use of this material in the
operation of a nuclear power reactor licensed pursuant to part 50 or
part 57 of this chapter, must include a physical security plan that
demonstrates how the applicant plans to meet the requirements of
paragraphs (d), (e), (f), and (g) of Sec. 73.67 of this chapter, as
appropriate. The licensee shall retain a copy of this physical security
plan as a record for the period during which the licensee possesses the
appropriate type and quantity of special nuclear material under each
license, and if any portion of the plan is superseded, retain that
superseded portion of the plan for 3 years after the effective date of
the change.
* * * * *
0
98. In Sec. 70.32, revise the introductory text of paragraph (c)(1)
and paragraph (d) to read as follows:
Sec. 70.32 Conditions of licenses.
* * * * *
(c)(1) Each license authorizing the possession and use at any one
time and location of uranium source material at an uranium enrichment
facility or special nuclear material in a quantity exceeding one
effective kilogram, except for use as sealed sources and those uses
involved in the operation of a nuclear reactor licensed pursuant to
part 50 or part 57 of this chapter and those involved in a waste
disposal operation, shall contain and be subject to a condition
requiring the licensee to maintain and follow:
* * * * *
(d) The licensee shall make no change which would decrease the
effectiveness of the plan for physical protection of special nuclear
material in transit prepared pursuant to Sec. 70.22(g) or Sec.
73.20(c) of this chapter without the prior approval of the Commission.
A licensee desiring to make such changes shall submit an application
for a change in the technical specifications incorporated in his or her
license, if any, or for an amendment to the license pursuant to Sec.
50.90, Sec. 57.310, or Sec. 70.34 of this chapter, as appropriate.
The licensee may make changes to the plan for physical protection of
special nuclear material without prior Commission approval if these
changes do not decrease the effectiveness of the plan. The licensee
shall retain a copy of the plan as a record for the period during which
the licensee possesses a formula quantity of special nuclear material
requiring this record under each license and each change to the plan
for three years from the effective date of the change. Within two
months after each change, a report containing a description of the
change must be furnished to the Director of the NRC's Office of Nuclear
Material Safety and Safeguards, using an appropriate method listed in
Sec. 70.5(a); and a copy must be sent to the appropriate NRC Regional
Office shown in appendix A to part 73 of this chapter.
* * * * *
0
99. In Sec. 70.50, revise paragraph (d) to read as follows:
Sec. 70.50 Reporting requirements.
* * * * *
(d) The provisions of Sec. 70.50 do not apply to licensees subject
to Sec. 50.72 or Sec. 57.435 of this chapter. They do apply to those
10 CFR part 50 or part 57 licensees possessing material licensed under
10 CFR part 70 that are not subject to the notification requirements in
Sec. 50.72 or Sec. 57.435 of this chapter, respectively.
PART 72--LICENSING REQUIREMENTS FOR THE INDEPENDENT STORAGE OF
SPENT NUCLEAR FUEL AND HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-
RELATED GREATER THAN CLASS C WASTE
0
100. The authority citation for 10 CFR part 72 continues to read as
follows:
Authority: Atomic Energy Act of 1954, secs. 51, 53, 57, 62, 63,
65, 69, 81, 161, 182, 183, 184, 186, 187, 189, 223, 234, 274 (42
U.S.C. 2071, 2073, 2077, 2092, 2093, 2095, 2099, 2111, 2201, 2210e,
2232, 2233, 2234, 2236, 2237, 2238, 2273, 2282, 2021); Energy
Reorganization Act of 1974, secs. 201, 202, 206, 211 (42 U.S.C.
5841, 5842, 5846, 5851); National Environmental Policy Act of 1969
(42 U.S.C. 4332); Nuclear Waste Policy Act of 1982, secs. 117(a),
132, 133, 134, 135, 137, 141, 145(g), 148, 218(a) (42 U.S.C.
10137(a), 10152, 10153, 10154, 10155, 10157, 10161, 10165(g), 10168,
10198(a)); 44 U.S.C. 3504 note.
0
101. In Sec. 72.3, revise the definition for ``Independent spent fuel
storage installation or ISFSI'' to read as follows:
Sec. 72.372.3 Definitions.
* * * * *
Independent spent fuel storage installation or ISFSI means a
complex designed and constructed for the interim storage of spent
nuclear fuel, solid reactor-related GTCC waste, and other radioactive
materials associated with spent fuel and reactor-related GTCC waste
storage. An ISFSI that is located on the site of another facility
licensed under this part or a facility licensed under part 50 or part
57 of this chapter and shares common utilities and services with that
facility or is physically connected with that other facility may still
be considered independent.
* * * * *
0
102. In Sec. 72.30, revise paragraph (e)(5) to read as follows:
Sec. 72.30 Financial assurance and recordkeeping for decommissioning.
* * * * *
(e) * * *
(5) In the case of licensees who are issued a power reactor license
under part 50 or part 57 of this chapter or ISFSI licensees who are an
electric utility, as defined in part 50 or part 57 of this chapter,
with a specific license issued under this part, the methods of Sec.
50.75(b), (e), and (h) or Sec. 57.55(i) of this chapter, as
applicable. In the event that funds remaining to be placed into the
licensee's ISFSI decommissioning external sinking fund are no longer
approved for recovery in rates by a competent rate making authority,
the licensee must make changes to provide financial assurance using one
or more of the methods stated in paragraphs (1) through (4) of this
section.
* * * * *
Sec. 72.40 [Amended]
0
103. In Sec. 72.40, in paragraph (c), remove the phrase ``of this
chapter,'' and add in its place the phrase ``or part 57 of this
chapter,''.
0
104. In Sec. 72.75, revise paragraph (i)(1)(ii) to read as follows:
Sec. 72.75 Reporting requirements for specific events and conditions.
* * * * *
(i) * * *
(1) * * *
(ii) Licensees issued a general license under Sec. 72.210, after
the licensee has
[[Page 23762]]
placed spent fuel on the ISFSI storage pad (if the ISFSI is located
inside the collocated protected area, for a reactor licensed under part
50 or part 57 of this chapter) or after the licensee has transferred
spent fuel waste outside the reactor licensee's protected area to the
ISFSI storage pad (if the ISFSI is located outside the collocated
protected area, for a reactor licensed under part 50 or part 57 of this
chapter).
* * * * *
Sec. 72.184 [Amended]
0
105. In Sec. 72.184, in paragraph (a) remove the phrase ``of this
chapter'' and add in its place the phrase ``or part 57 of this
chapter''.
0
106. Revise Sec. 72.210 to read as follows:
Sec. 72.210 General license issued.
A general license is hereby issued for the storage of spent fuel in
an independent spent fuel storage installation at power reactor sites
to persons authorized to possess or operate nuclear power reactors
under 10 CFR part 50, 52, or 57.
0
107. In Sec. 72.212, revise paragraph (b)(8) to read as follows:
Sec. 72.212 Conditions of general license issued under Sec. 72.210.
* * * * *
(b) * * *
(8) Before use of the general license, determine whether activities
related to storage of spent fuel under this general license involve a
change in the facility Technical Specifications or require a license
amendment for the facility pursuant to Sec. 50.59(c) or Sec. 57.312
of this chapter. Results of this determination must be documented in
the evaluations made in paragraph (b)(5) of this section.
* * * * *
0
108. In Sec. 72.218, revise paragraphs (a) and (b) to read as follows:
Sec. 72.218 Termination of licenses.
(a) The notification regarding the program for the management of
spent fuel at the reactor required by Sec. 50.54(bb) or Sec. 57.300
of this chapter must include a plan for removal of the spent fuel
stored under this general license from the reactor site. The plan must
show how the spent fuel will be managed before starting to decommission
systems and components needed for moving, unloading, and shipping this
spent fuel.
(b) An application for termination of a reactor operating license
issued under 10 CFR part 50 and submitted under Sec. 50.82 of this
chapter, or a combined license issued under 10 CFR part 52 and
submitted under Sec. 52.110 of this chapter, or an operating license
issued under 10 CFR part 57 and submitted under Sec. 57.305 of this
chapter must contain a description of how the spent fuel stored under
this general license will be removed from the reactor site.
* * * * *
PART 73--PHYSICAL PROTECTION OF PLANTS AND MATERIALS
0
109. The authority citation for 10 CFR part 73 continues to read as
follows:
Authority: Atomic Energy Act of 1954, secs. 53, 147, 149, 161,
161A, 170D, 170E, 170H, 170I, 223, 229, 234, 1701 (42 U.S.C. 2073,
2167, 2169, 2201, 2201a, 2210d, 2210e, 2210h, 2210i, 2273, 2278a,
2282, 2297f); Energy Reorganization Act of 1974, secs. 201, 202 (42
U.S.C. 5841, 5842); Nuclear Waste Policy Act of 1982, secs. 135, 141
(42 U.S.C. 10155, 10161); 44 U.S.C. 3504 note.
Section 73.37(b)(2) also issued under sec. 301, Pub. L. 96-295,
94 Stat. 789 (42 U.S.C. 5841 note).
0
110. In Sec. 73.1, revise paragraph (b)(1)(i) to read as follows:
Sec. 73.173.1 Purpose and scope.
* * * * *
(b) * * *
(1) * * *
(i) The physical protection of production and utilization
facilities licensed under part 50, 52, or 57 of this chapter,
* * * * *
0
111. In Sec. 73.2, revise paragraph (a) to read as follows:
Sec. 73.273.2 Definitions.
* * * * *
(a) Terms defined in parts 50, 52, 57, 70, and 95 of this chapter
have the same meaning when used in this part.
* * * * *
0
112. In Sec. 73.8 revise paragraph (b) to read as follows:
Sec. 73.873.8 Information collection requirements: OMB approval.
* * * * *
(b) The approved information collection requirements contained in
this part appear in Sec. Sec. 73.5, 73.15, 73.17, 73.20, 73.21, 73.24,
73.25, 73.26, 73.27, 73.37, 73.40, 73.45, 73.46, 73.50, 73.54, 73.55,
73.56, 73.57, 73.58, 73.60, 73.67, 73.70, 73.72, 73.73, 73.74, 73.77,
73.110, 73.1200, 73.1205, 73.1210, 73.1215, and appendices B and C to
this part.
* * * * *
0
113. In Sec. 73.50, revise the introductory text to read as follows:
Sec. 73.50 Requirements for physical protection of licensed
activities.
Each licensee who is not subject to Sec. 73.51, but who possesses,
uses, or stores formula quantities of strategic special nuclear
material that are not readily separable from other radioactive material
and which have a total external radiation level in excess of 1 gray
(100 rad) per hour at a distance of 1 meter (3.3 feet) from any
accessible surfaces without intervening shielding other than at a
nuclear reactor facility licensed under part 50, 52, or 57 of this
chapter, shall comply with the following:
* * * * *
0
114. In Sec. 73.54, revise paragraph (g) to read as follows:
Sec. 73.54 Protection of digital computer and communication systems
and networks.
* * * * *
(g) Each licensee that is licensed to operate a nuclear plant under
10 CFR part 50 or 52 after [INSERT THE EFFECTIVE DATE OF THE FINAL
RULE] and elects to implement the requirements of this section, and
each licensee that is licensed to operate a nuclear plant under 10 CFR
part 57 and elects to implement the requirements of this section, must
establish and implement cybersecurity reviews to assess the
effectiveness of the implementation of the cybersecurity program.
(1) The licensee must review each element of the cybersecurity
program at a frequency commensurate with the importance or significance
to safety of plant operations to ensure timely identification and
documentation of vulnerabilities, improvements, and corrective actions.
(2) Cybersecurity reviews must be performed by individuals
independent of those personnel responsible for program management and
any individual who has direct responsibility for implementing the
cybersecurity program.
(3) The licensee must establish and perform self-assessments to
ensure the effective implementation of the cybersecurity program.
(4) The results and recommendations of the cybersecurity program
reviews, management's findings regarding program effectiveness, and any
actions taken as a result of recommendations from prior program
reviews, must be documented in a report and must be maintained in an
auditable form and available for inspection.
* * * * *
0
115. In Sec. 73.56, revise paragraph (a)(3) to read as follows:
Sec. 73.56 Personnel access authorization requirements for nuclear
power plants.
(a) * * *
(3) Each applicant for an operating license under the provisions of
part 50
[[Page 23763]]
of this chapter, each holder of a combined license under the provisions
of part 52 of this chapter, and each applicant for an operating license
under the provisions of part 57 of this chapter that must meet the
requirements of subpart J of this part, shall implement the
requirements of this section before fuel is allowed on site (protected
area).
* * * * *
0
116. In Sec. 73.57, revise paragraph (a)(3) to read as follows:
Sec. 73.57 Requirements for criminal history records checks of
individuals granted unescorted access to a nuclear power facility, a
non-power reactor, or access to Safeguards Information.
(a) * * *
(3) Before receiving its operating license under part 50 or part 57
of this chapter or before the Commission makes its finding under Sec.
52.103(g) of this chapter, each applicant for a license to operate a
nuclear power reactor (including an applicant for a combined license)
or a non-power reactor may submit fingerprints for those individuals
who will require unescorted access to the nuclear power facility or
non-power reactor facility.
* * * * *
0
117. In Sec. 73.58, revise paragraph (a) to read as follows:
Sec. 73.58 Safety/security interface requirements for nuclear power
reactors.
(a) Each operating nuclear power reactor licensee with a license
issued under part 50, 52, or 57 of this chapter shall comply with the
requirements of this section.
* * * * *
0
118. In Sec. 73.77, revise paragraphs (a) and (b) to read as follows:
Sec. 73.77 Cyber security event notifications.
(a) Each licensee subject to the provisions of Sec. 73.54 or Sec.
73.110 must notify the NRC Headquarters Operations Center of a
cyberattack that adversely impacted a safety or security function using
the procedures of Sec. 50.72 or Sec. 57.435 of this chapter or Sec.
73.1200 based on the function adversely impacted (safety or security).
(b) If it is later determined that the cause of a previously
reported event was from a cyberattack, the licensee must inform the NRC
using one of the following applicable methods:
(1) Follow-up notification process as specified in Sec. 50.72 or
Sec. 57.435 of this chapter;
(2) Significant supplemental information process as specified in
Sec. 73.1200; or
(3) Submission of a Licensee Event Report as specified in Sec.
50.73 or Sec. 57.440 of this chapter.
* * * * *
0
119. Add Sec. 73.110 to subpart I to read as follows:
Sec. 73.110 Cybersecurity program.
(a) Each licensee that is licensed to operate a nuclear plant under
10 CFR part 57 and elects to implement the requirements of this
section, and each licensee that is licensed to operate a nuclear plant
under 10 CFR part 50 or 52 after [INSERT THE EFFECTIVE DATE OF THE
FINAL RULE] and elects to implement the requirements of this section,
must establish, implement, and maintain a cybersecurity program that is
commensurate with the potential consequences resulting from
cyberattacks, up to and including the design basis threat as described
in Sec. 73.1. The cybersecurity program must provide reasonable
assurance that digital computer and communication systems and networks
are adequately protected against cyberattacks that are capable of
causing the following consequences:
(1) Adversely impacting the safety, security, and emergency
preparedness functions performed by digital assets that prevent a
postulated fission product release resulting in offsite doses exceeding
the values in Sec. 50.34(a)(1)(ii)(D) or 52.47(a)(2)(iv) of this
chapter, as applicable.
(2) Adversely impacting the security functions performed by digital
assets necessary for implementing the physical security requirements in
Sec. 57.60(a)(8)(v)(A) of this chapter or Sec. 73.55, as applicable.
(b) To protect digital computer and communication systems and
networks associated with the functions described in paragraphs (a)(1)
and (2) of this section (including support systems and equipment that,
if compromised, adversely impact these functions), the licensee must--
(1) Analyze the potential consequences resulting from cyberattacks
on digital computer and communication systems and networks and identify
those assets that must be protected to demonstrate compliance with
paragraph (a) of this section; and
(2) Implement the cybersecurity program in accordance with
paragraph (d) of this section.
(c) The licensee must protect the systems and networks identified
in paragraph (b)(1) of this section in a manner that is commensurate
with the potential consequences resulting from cyberattacks that:
(1) Adversely impact the integrity or confidentiality of data and/
or software;
(2) Deny access to systems, services, and/or data; and
(3) Adversely impact the operation of systems, networks, and
associated equipment.
(d) The cybersecurity program must be designed in a manner that is
commensurate with the potential consequences resulting from
cyberattacks through the following steps:
(1) Implement security controls to protect the assets identified
under paragraph (b)(1) of this section from cyberattacks, commensurate
with the assets' safety and security significance;
(2) Apply and maintain defense in depth protective strategies to
ensure the capability to detect, delay, respond to, and recover from
cyberattacks capable of causing the consequences identified in
paragraph (a) of this section;
(3) Mitigate the adverse effects of cyberattacks capable of causing
the consequences identified in paragraph (a) of this section; and
(4) Ensure that the functions of protected assets identified under
paragraph (b)(1) of this section are not adversely impacted due to
cyberattacks.
(e) The licensee must implement the following requirements in a
manner that is commensurate with the potential consequences resulting
from cyberattacks:
(1) As part of the cybersecurity program, the licensee must comply
with the requirements in Sec. 73.54(d)(1), (2), and (4), and must
ensure that modifications to assets, identified under paragraph (b)(1)
of this section are evaluated before implementation to ensure that the
cybersecurity performance objectives identified in paragraph (a) of
this section are maintained.
(2) The licensee must establish, implement, and maintain a
cybersecurity plan that implements the cybersecurity program
requirements of this section.
(i) The cybersecurity plan must describe how the requirements of
this section will be implemented and must account for the site-specific
conditions that affect implementation.
(ii) The cybersecurity plan must include measures for incident
response and recovery for cyberattacks. The cybersecurity plan must
include the analysis identified under paragraph (b)(1) of this section
and describe how the licensee will--
(A) Apply and maintain defense in depth protective strategies as
required in paragraph (d)(2) of this section;
[[Page 23764]]
(B) Maintain the capability for timely detection and response to
cyberattacks;
(C) Mitigate the consequences of cyberattacks;
(D) Correct exploited vulnerabilities; and
(E) Restore affected systems, networks, and/or equipment affected
by cyberattacks.
(3) The licensee must develop and maintain written policies and
implementing procedures to implement the cybersecurity plan. Policies,
implementing procedures, and other supporting technical information
used by the licensee need not be submitted for Commission review and
approval as part of the cybersecurity plan but are subject to
inspection by NRC staff on a periodic basis.
(4) The licensee must establish and implement cybersecurity reviews
to assess the effectiveness of the implementation of the cybersecurity
program.
(i) The licensee must review each element of the cybersecurity
program at a frequency commensurate with the importance or significance
to safety of plant operations to ensure timely identification and
documentation of vulnerabilities, improvements, and corrective actions.
(ii) Cybersecurity reviews must be performed by individuals
independent of those personnel responsible for program management and
any individual who has direct responsibility for implementing the
cybersecurity program.
(iii) The licensee must establish and perform self-assessments to
ensure the effective implementation of the cybersecurity program.
(iv) The results and recommendations of the cybersecurity program
reviews, management's findings regarding program effectiveness, and any
actions taken as a result of recommendations from prior program
reviews, must be documented in a report and must be maintained in an
auditable form and available for inspection.
(5) The licensee must retain all records and supporting technical
documentation required to demonstrate compliance with the requirements
of this section as a record until the Commission terminates the license
for which the records were developed and must maintain superseded
portions of these records for at least three (3) years after the record
is superseded, unless otherwise specified by the Commission.
0
120. In Sec. 73.1200, revise introductory text of paragraphs (a),
(c)(1), and (e)(1), revise paragraph (e)(4), and introductory text of
paragraph (g)(1) to read as follows:
Sec. 73.1200 Notification of physical security events.
(a) 15-minute notifications--facilities. Each licensee subject to
the provisions of Sec. 73.20, Sec. 73.45, Sec. 73.46, Sec. 73.51,
Sec. 73.55, or subpart J of part 57 of this chapter, must notify the
NRC Headquarters Operations Center, as soon as possible but within 15
minutes after--
* * * * *
(c) * * *
(1) Each licensee subject to the provisions of Sec. 73.20, Sec.
73.45, Sec. 73.46, Sec. 73.50, Sec. 73.51, Sec. 73.55, Sec. 73.60,
Sec. 73.67, or subpart J of part 57 of this chapter, must notify the
NRC Headquarters Operations Center as soon as possible but no later
than 1 hour after the time of discovery of the following significant
facility security events involving--
* * * * *
(e) * * *
(1) Each licensee subject to the provisions of Sec. 73.20, Sec.
73.45, Sec. 73.46, Sec. 73.50, Sec. 73.51, Sec. 73.55, Sec. 73.60,
Sec. 73.67, or subpart J of part 57 of this chapter, must notify the
NRC Headquarters Operations Center within 4 hours after time of
discovery of the following facility security events involving--
* * * * *
(4) For licensees subject to the provisions of Sec. 73.55 or
subpart J of part 57 of this chapter, an event involving the licensee's
suspension of security measures.
* * * * *
(g) * * *
(1) Each licensee subject to the provisions of Sec. 73.20, Sec.
73.45, Sec. 73.46, Sec. 73.50, Sec. 73.51, Sec. 73.55, Sec. 73.60,
Sec. 73.67, or subpart J of part 57 of this chapter, must notify the
NRC Headquarters Operations Center within 8 hours after time of
discovery of the following facility security program failures
involving--
* * * * *
(iv) For licensees subject to the provisions of Sec. 73.77, a
cybersecurity event that impacted the ability of the facility's SSCs to
perform their intended security functions.
* * * * *
0
121. In Sec. 73.1205, revise paragraph (b)(2) to read as follows:
Sec. 73.1205 Written follow-up reports of physical security events.
* * * * *
(b) * * *
(2)(i) Licensees subject to Sec. 50.73 or subpart J of part 57 of
this chapter must prepare the written follow-up report on NRC Form 366.
(ii) Licensees not subject to Sec. 50.73 or subpart J of part 57
of this chapter must prepare the written follow-up report in a letter
format.
* * * * *
0
122. In Sec. 73.1210, revise paragraphs (a)(1) and (b)(3)(i) to read
as follows:
Sec. 73.1210 Recordkeeping of physical security events.
(a) * * *
(1) Licensees with facilities or shipment activities subject to the
provisions of Sec. 73.20, Sec. 73.25, Sec. 73.26, Sec. 73.27, Sec.
73.37, Sec. 73.45, Sec. 73.46, Sec. 73.50, Sec. 73.51, Sec. 73.55,
Sec. 73.60, Sec. 73.67, or subpart J of part 57 of this chapter, must
record the physical security events and conditions adverse to security
that are specified in paragraphs (c) through (f) of this section.
* * * * *
(b) * * *
(3)(i) Licensees must record these physical security events and
conditions adverse to security in either a stand-alone safeguards event
log or as part of the licensee's corrective action program, as
specified under the applicable quality assurance program provisions of
parts 50, 52, 57, 60, 63, 70, and 72 of this chapter, or both.
* * * * *
0
123. In Sec. 73.1215, revise introductory text of paragraph (d)(1) to
read as follows:
Sec. 73.1215 Suspicious activity reports.
* * * * *
(d) * * *
(1) For licensees subject to the provisions of Sec. 73.20, Sec.
73.45, Sec. 73.46, Sec. 73.50, Sec. 73.51, Sec. 73.55, Sec. 73.60,
Sec. 73.67, or subpart J of part 57 of this chapter, the licensees
must report activities they assess are suspicious. Examples include,
but are not limited to, the following:
* * * * *
0
124. In appendix B to part 73, revise Definitions introductory text to
read as follows:
APPENDIX B TO PART 73--GENERAL CRITERIA FOR SECURITY PERSONNEL
* * * * *
Definitions
Terms defined in parts 50, 57, 70, and 73 of this chapter have
the same meaning when used in this appendix.
* * * * *
PART 74--MATERIAL CONTROL AND ACCOUNTING OF SPECIAL NUCLEAR
MATERIAL
0
125. The authority citation for 10 CFR part 74 continues to read as
follows:
Authority: Atomic Energy Act of 1954, secs. 53, 57, 161, 182,
223, 234, 1701 (42 U.S.C. 2073, 2077, 2201, 2232, 2273, 2282,
[[Page 23765]]
2297f); Energy Reorganization Act of 1974, secs. 201, 202 (42 U.S.C.
5841, 5842); 44 U.S.C. 3504 note.
0
126. In Sec. 74.31, revise the introductory text of paragraph (a) to
read as follows:
Sec. 74.31 Nuclear material control and accounting for special
nuclear material of low strategic significance.
(a) General performance objectives. Each licensee who is authorized
to possess and use more than one effective kilogram of special nuclear
material of low strategic significance, excluding sealed sources, at
any site or contiguous sites subject to control by the licensee, other
than a production or utilization facility licensed pursuant to part 50,
part 57, or part 70 of this chapter, or operations involved in waste
disposal, shall implement and maintain a Commission-approved material
control and accounting system that will achieve the following
objectives:
* * * * *
0
127. In Sec. 74.41, revise the introductory text of paragraph (a) to
read as follows:
Sec. 74.41 Nuclear material control and accounting for special
nuclear material of moderate strategic significance.
(a) General performance objectives. Each licensee who is authorized
to possess special nuclear material (SNM) of moderate strategic
significance or SNM in a quantity exceeding one effective kilogram of
strategic special nuclear material in irradiated fuel reprocessing
operations other than as sealed sources and to use this material at any
site other than a nuclear reactor licensed pursuant to part 50 or part
57 of this chapter; or as reactor irradiated fuels involved in
research, development, and evaluation programs in facilities other than
irradiated fuel reprocessing plants; or an operation involved with
waste disposal, shall establish, implement, and maintain a Commission-
approved material control and accounting (MC&A) system that will
achieve the following performance objectives:
* * * * *
0
128. In Sec. 74.51, revise the introductory text of paragraph (a) to
read as follows:
Sec. 74.51 Nuclear material control and accounting for strategic
special nuclear material.
(a) General performance objectives. Each licensee who is authorized
to possess five or more formula kilograms of strategic special nuclear
material (SSNM) and to use such material at any site, other than a
nuclear reactor licensed pursuant to part 50 or part 57 of this
chapter, an irradiated fuel reprocessing plant, an operation involved
with waste disposal, or an independent spent fuel storage facility
licensed pursuant to part 72 of this chapter shall establish,
implement, and maintain a Commission-approved material control and
accounting (MC&A) system that will achieve the following objectives:
* * * * *
PART 75--SAFEGUARDS ON NUCLEAR MATERIAL--IMPLEMENTATION OF
SAFEGUARDS AGREEMENTS BETWEEN THE UNITED STATES AND THE
INTERNATIONAL ATOMIC ENERGY AGENCY
0
129. The authority citation for 10 CFR part 75 continues to read as
follows:
Authority: Atomic Energy Act of 1954, secs. 53, 63, 103, 104,
122, 161, 223, 234, 1701 (42 U.S.C. 2073, 2093, 2133, 2134, 2152,
2201, 2273, 2282, 2297f); Energy Reorganization Act of 1974, sec.
201 (42 U.S.C. 5841); Nuclear Waste Policy Act of 1982, secs. 135,
141 (42 U.S.C. 10155, 10161); 44 U.S.C. 3504 note.
0
130. In Sec. 75.4, revise the introductory text and the definition for
``Facility'', paragraph (6), to read as follows:
Sec. 75.475.4 Definitions.
As used in this part:
Unless otherwise defined in this section, the terms defined in
Sec. Sec. 40.4, 50.2, 57.3, and 70.4 of this chapter have the same
meaning when used in this part.
* * * * *
Facility means:
(1) * * *
(6) Any plant or location where the possession of more than 1
effective kilogram of nuclear material is licensed pursuant to 10 CFR
part 40, 50, 57, 60, 61, 63, 70, 72, 76, or 150 of this chapter or an
Agreement State license.
* * * * *
PART 95--FACILITY SECURITY CLEARANCE AND SAFEGUARDING OF NATIONAL
SECURITY INFORMATION AND RESTRICTED DATA
0
131. The authority citation for 10 CFR part 95 continues to read as
follows:
Authority: Atomic Energy Act of 1954, secs. 145, 161, 223, 234
(42 U.S.C. 2165, 2201, 2273, 2282); Energy Reorganization Act of
1974, sec. 201 (42 U.S.C. 5841); 44 U.S.C. 3504 note; E.O. 10865, as
amended, 25 FR 1583, 3 CFR, 1959-1963 Comp., p. 398; E.O. 12829, 58
FR 3479, 3 CFR, 1993 Comp., p. 570; E.O. 12968, 60 FR 40245, 3 CFR,
1995 Comp., p. 391; E.O. 13526, 75 FR 707, 3 CFR, 2009 Comp., p.
298.
0
132. In Sec. 95.5, revise the definition for ``License'' to read as
follows:
Sec. 95.595.5 Definitions.
* * * * *
License means a license issued under 10 CFR part 50, 52, 54, 57,
60, 63, 70, or 72.
* * * * *
0
133. In Sec. 95.39, revise paragraph (a) to read as follows:
Sec. 95.39 External transmission of documents and material.
(a) Restrictions. Documents and material containing classified
information received or originated in connection with an NRC license,
certificate, standard design approval or standard design certification
under part 52 of this chapter, or NRC license or standard design
approval under part 57 of this chapter, must be transmitted only to CSA
approved security facilities.
* * * * *
PART 140--FINANCIAL PROTECTION REQUIREMENTS AND INDEMNITY
AGREEMENTS
0
134. The authority citation for 10 CFR part 140 continues to read as
follows:
Authority: Atomic Energy Act of 1954, secs. 161, 170, 223, 234
(42 U.S.C. 2201, 2210, 2273, 2282); Energy Reorganization Act of
1974, secs. 201, 202 (42 U.S.C. 5841, 5842); 44 U.S.C. 3504 note.
0
135. In Sec. 140.2, revise paragraphs (a)(1) and (2) to read as
follows:
Sec. 140.2140.2 Scope.
(a) * * *
(1) To each person who is an applicant for or holder of a license
issued under 10 CFR part 50, 52, 54, or 57 to operate a nuclear
reactor, and
(2) With respect to an extraordinary nuclear occurrence, to each
person who is an applicant for or holder of a license to operate a
production facility or a utilization facility (including an operating
license issued under part 50 or part 57 of this chapter and a combined
license under part 52 of this chapter), and to other persons
indemnified with respect to the involved facilities.
* * * * *
0
136. Revise Sec. 140.10 to read as follows:
Sec. 140.10 Scope.
This subpart applies to each person who is an applicant for or
holder of a license issued under 10 CFR part 50, 54, or 57 to operate a
nuclear reactor, or is the applicant for or holder of a combined
license issued under 10 CFR part 52 or 54, except licenses held by
persons found by the Commission to be Federal agencies or nonprofit
educational institutions licensed to
[[Page 23766]]
conduct educational activities. This subpart also applies to persons
licensed to possess and use plutonium in a plutonium processing and
fuel fabrication plant.
0
137. In Sec. 140.11, revise paragraph (b) to read as follows:
Sec. 140.11 Amounts of financial protection for certain reactors.
* * * * *
(b) In any case where a person is authorized under 10 CFR part 50,
52, 54, or 57 to operate two or more nuclear reactors at the same
location, the total primary financial protection required of the
licensee for all such reactors is the highest amount which would
otherwise be required for any one of those reactors; provided, that
such primary financial protection covers all reactors at the location.
0
138. In Sec. 140.12, revise paragraph (c) to read as follows:
Sec. 140.12 Amount of financial protection required for other
reactors.
* * * * *
(c) In any case where a person is authorized under 10 CFR part 50,
52, 54, or 57 to operate two or more nuclear reactors at the same
location, the total financial protection required of the licensee for
all such reactors is the highest amount which would otherwise be
required for any one of those reactors; provided, that such financial
protection covers all reactors at the location.
* * * * *
0
139. Revise Sec. 140.13 to read as follows:
Sec. 140.13 Amount of financial protection required of certain
holders of construction permits and combined licenses under 10 CFR part
52.
Each holder of a 10 CFR part 50 or part 57 construction permit, or
a holder of a combined license under part 52 of this chapter before the
date that the Commission had made the finding under Sec. 52.103(g) of
this chapter, who also holds a license under part 70 of this chapter
authorizing ownership, possession and storage only of special nuclear
material at the site of the nuclear reactor for use as fuel in
operation of the nuclear reactor after issuance of either an operating
license under 10 CFR part 50 or part 57, or a combined license under 10
CFR part 52, shall, during the period before issuance of a license
authorizing operation under 10 CFR part 50 or part 57, or the period
before the Commission makes the finding under Sec. 52.103(g) of this
chapter, as applicable, have and maintain financial protection in the
amount of $1,000,000. Proof of financial protection shall be filed with
the Commission in the manner specified in Sec. 140.15 before issuance
of the license under part 70 of this chapter.
0
140. In Sec. 140.20, revise paragraph (a)(1)(i) to read as follows:
Sec. 140.20 Indemnity agreements and liens.
(a) * * *
(1)(i) The effective date of the license (issued under part 50 or
part 57 of this chapter) authorizing the licensee to operate the
nuclear reactor involved; or
* * * * *
PART 150--EXEMPTIONS AND CONTINUED REGULATORY AUTHORITY IN
AGREEMENT STATES AND IN OFFSHORE WATERS UNDER SECTION 274
0
141. The authority citation for 10 CFR part 150 continues to read as
follows:
Authority: Atomic Energy Act of 1954, secs. 11, 53, 81, 83, 84,
122, 161, 181, 223, 234, 274 (42 U.S.C. 2014, 2201, 2231, 2273,
2282, 2021); Energy Reorganization Act of 1974, sec. 201 (42 U.S.C.
5841); Nuclear Waste Policy Act of 1982, secs. 135, 141 (42 U.S.C.
10155, 10161); 44 U.S.C. 3504 note.
0
142. In Sec. 150.15, revise paragraphs (a)(7)(iii) and (a)(8) to read
as follows:
Sec. 150.15 Persons not exempt.
(a) * * *
(7) * * *
(iii) Greater than Class C waste, as defined in part 72 of this
chapter, in an ISFSI or an MRS licensed under part 72 of this chapter;
the Greater than Class C waste must originate in, or be used by, a
facility licensed under part 50, part 52, or part 57 of this chapter.
(8) Greater than Class C waste, as defined in part 72 of this
chapter, that originates in, or is used by, a facility licensed under
part 50, part 52, or part 57 of this chapter and is licensed under part
30 and/or part 70 of this chapter.
* * * * *
For the Nuclear Regulatory Commission.
Dated: April 29, 2026
Tomas Herrera,
Acting Secretary of the Commission.
[FR Doc. 2026-08550 Filed 4-30-26; 8:45 am]
BILLING CODE 7590-01-P