[Federal Register Volume 91, Number 60 (Monday, March 30, 2026)]
[Rules and Regulations]
[Pages 15696-15881]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2026-06048]
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Vol. 91
Monday,
No. 60
March 30, 2026
Part II
Nuclear Regulatory Commission
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10 CFR Parts 1, 2, et al.
Risk-Informed, Technology-Inclusive Regulatory Framework for Advanced
Reactors; Final Rule
Federal Register / Vol. 91 , No. 60 / Monday, March 30, 2026 / Rules
and Regulations
[[Page 15696]]
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NUCLEAR REGULATORY COMMISSION
10 CFR Parts 1, 2, 10, 11, 19, 20, 21, 25, 26, 30, 40, 50, 51, 53,
70, 72, 73, 74, 75, 95, 140, 150, 170, and 171
[NRC-2019-0062]
RIN 3150-AK31
Risk-Informed, Technology-Inclusive Regulatory Framework for
Advanced Reactors
AGENCY: Nuclear Regulatory Commission.
ACTION: Final rule.
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SUMMARY: The U.S. Nuclear Regulatory Commission (NRC) is amending its
regulations by adding a risk-informed, performance-based, and
technology-inclusive regulatory framework for commercial nuclear plants
in response to the Nuclear Energy Innovation and Modernization Act
(NEIMA). The current application and licensing requirements were
primarily developed to address license requests concerning light water-
cooled reactors and operational requirements for those types of
reactors. This final rule responds to NEIMA by creating an alternative,
technology-inclusive regulatory framework to accommodate licensing of
future commercial nuclear plants, including advanced reactor designs
that may not employ light-water technology.
DATES: This final rule is effective on April 29, 2026.
ADDRESSES: Please refer to Docket ID NRC-2019-0062 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-2019-0062. Address
questions about NRC dockets to Helen Chang; telephone: 301-415-3228;
email: [email protected]. For technical questions, contact the
individuals listed in the FOR FURTHER INFORMATION CONTACT section of
this document.
NRC's Agencywide Documents Access and Management System
(ADAMS): You may obtain publicly available documents online in the
ADAMS Public Documents collection at https://www.nrc.gov/reading-rm/adams.html. To begin the search, select ``Begin 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.
FOR FURTHER INFORMATION CONTACT: Nicole Fields, Office of Nuclear
Material Safety and Safeguards, telephone: 630-829-9570, email:
[email protected] and Anders Gilbertson, Office of Nuclear Reactor
Regulation, telephone: 301-415-1541, email: [email protected].
Both are staff of the U.S. Nuclear Regulatory Commission, Washington,
DC 20555-0001.
SUPPLEMENTARY INFORMATION: This rulemaking is separate from the NRC's
comprehensive review and reform of its regulations, including those
governing transportation, in accordance with Executive Order (E.O.)
14300, ``Ordering the Reform of the Nuclear Regulatory Commission'' (90
FR 22587; May 29, 2025). The rulemakings associated with that effort
will comprehensively reexamine NRC requirements. While there could be
additional revisions to part 53, ``Risk-Informed, Technology-Inclusive
Regulatory Framework for Commercial Nuclear Plants,'' of title 10 of
the Code of Federal Regulations (10 CFR) as a result of these future
rulemakings, the NRC is moving forward with publication of this final
rule at this time because it is a deregulatory action of high interest
for stakeholders that was in progress before the issuance of E.O.
14300.
Executive Summary
A. Need for the Regulatory Action
On January 14, 2019, the President signed the NEIMA into law (Pub.
L. 115-439). NEIMA section 103(a)(4) directs the NRC to ``complete a
rulemaking to establish a technology-inclusive, regulatory framework
for optional use by commercial advanced nuclear reactor applicants for
new reactor license applications.'' NEIMA defines a ``technology-
inclusive regulatory framework'' as one that is ``developed using
methods of evaluation that are flexible and practicable for application
to a variety of reactor technologies, including, where appropriate, the
use of risk-informed and performance-based techniques.'' NEIMA, as
further amended by the Accelerating Deployment of Versatile, Advanced
Nuclear for Clean Energy Act of 2024 (ADVANCE Act), defines the term
``advanced nuclear reactor'' as ``a nuclear fission reactor or fusion
machine, including a prototype plant (as defined in sections 50.2 and
52.1 of title 10 of the Code of Federal Regulations (10 CFR) (as in
effect on the date of enactment of [NEIMA])), with significant
improvements compared to commercial nuclear reactors under construction
as of the date of enactment of [NEIMA].''
The NRC initially considered establishing the scope of 10 CFR part
53 as being for ``advanced nuclear plants'' consisting of one or more
``advanced nuclear reactors'' as defined in NEIMA. Based on public
discussions on the use of the term, the NRC determined that the NEIMA
definition, although broad, did not define ``significant improvements''
with enough specificity to implement in NRC regulations. Additionally,
a number of stakeholders suggested that the descriptor, ``advanced,''
implied enhanced safety, while the NEIMA definition includes
``significant improvements'' in areas other than safety enhancements.
In response to this feedback, and to be technology-inclusive, the NRC
determined that the broader term ``commercial nuclear plant'' is
preferable.
The current application and licensing requirements in 10 CFR part
50, ``Domestic Licensing of Production and Utilization Facilities,''
and 10 CFR part 52, ``Licenses, Certifications, and Approvals for
Nuclear Power Plants,'' were primarily developed to address license
requests concerning light water-cooled reactors and operational
requirements for those types of reactors. This final rule responds to
NEIMA by creating an alternative, technology-inclusive regulatory
framework to accommodate licensing of future commercial nuclear plants,
including advanced reactor designs that may not employ light-water
technology. The new alternative requirements and implementing guidance
adopt technology-inclusive approaches and use risk-informed and
performance-based techniques to ensure an equivalent level of safety to
that of operating commercial nuclear plants while providing optionality
and flexibility for licensing and regulating a variety of technologies
and designs for commercial nuclear reactors.
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B. Major Provisions
Major provisions of this final rule, supported by accompanying
guidance, include the following:
A new alternative technology-inclusive, risk-informed,
performance-based framework that includes requirements for licensing
and regulating nuclear plants during the various stages of their life
cycles.
A new alternative technology-inclusive, risk-informed, and
performance-based framework in 10 CFR part 26, ``Fitness for Duty
Programs,'' developed from existing requirements in subpart K, ``FFD
Programs for Construction,'' of part 26.
A new alternative technology-inclusive and performance-
based security framework in 10 CFR part 73, ``Physical Protection of
Plants and Materials,'' that includes requirements for protection of
licensed activities at commercial nuclear plants.
C. Costs and Benefits
The NRC prepared a final regulatory analysis to determine the
expected quantitative costs and benefits of this final 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 final rule and associated guidance would result in net averted
costs to the industry and the NRC of $152 million using a 7-percent
discount rate and $203 million using a 3-percent discount rate. The
annualized averted costs at a 7-percent discount rate are approximately
$1.64 million per year to the NRC and $9.1 million per year to
industry, or net annualized averted costs of approximately $10.7
million, over the 66-year analysis period. The number of future
applicants was chosen conservatively, based on information known to the
NRC; with each additional applicant beyond those included in the
regulatory analysis, this final rule becomes even more cost-beneficial.
The final regulatory analysis also considers qualitative factors
such as greater regulatory stability, predictability, and clarity to
the licensing process. These benefits would result, for example, from
incorporating advances in probabilistic risk assessment (PRA) and other
risk-informed analyses into the regulatory framework. 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 final regulatory analysis
(available in the NRC's Agencywide Documents Access and Management
System (ADAMS) Accession No. ML26042A230).
Table of Contents
I. Background
NRC Advanced Reactor Readiness
II. Discussion
A. Objective and Applicability
B. Need for Changes to the Existing Regulatory Framework
C. 10 CFR Part 53 Framework
III. Part 53 Framework
Subpart A--General Provisions
A. Discussion of Definitions in Part 53
B. Other General Provisions
Subpart B--Technology-Inclusive Safety Requirements
Subpart C--Design and Analysis Requirements
Subpart D--Siting Requirements
Subpart E--Construction and Manufacturing Requirements
Subpart F--Requirements for Operation
Subpart G--Decommissioning Requirements
Subpart H--Licenses, Certifications, and Approvals
Subpart I--Maintaining and Revising Licensing-Basis Information
Subpart J--Reporting and Other Administrative Requirements
Subpart M--Enforcement
IV. Changes to Other Parts of 10 CFR Chapter I
10 CFR Part 26
A. Introduction
B. Changes to Part 26, Subparts A Through E and I
C. Requirements for Part 26, Subpart M
D. Changes to Part 26, Subpart N
E. Changes to Part 26, Subpart O
10 CFR Part 50
A. Section 50.160: Emergency Preparedness for Small Modular
Reactors, Non-Light-Water Reactors, and Non-Power Production or
Utilization Facilities
B. Appendix B to Part 50: Quality Assurance Criteria for Nuclear
Power Plants and Fuel Reprocessing Plants
C. Appendix E to Part 50: Emergency Planning and Preparedness
for Production and Utilization Facilities
10 CFR Part 73
A. Section 73.100: Technology-Inclusive Requirements for
Physical Protection of Licensed Activities at Commercial Nuclear
Plants Against Radiological Sabotage
B. Section 73.110: Technology-Inclusive Requirements for
Protection of Digital Computer and Communication Systems and
Networks
C. Section 73.120: Access Authorization Program for Commercial
Nuclear Plants
V. Opportunities for Public Participation
VI. Public Comment Analysis
VII. Regulatory Flexibility Certification
VIII. Regulatory Analysis
IX. Backfitting and Issue Finality
X. Cumulative Effects of Regulation
XI. Plain Writing
XII. Environmental Assessment and Final Finding of No Significant
Environmental Impact
XIII. Paperwork Reduction Act
XIV. 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
XV. Congressional Review Act
XVI. Criminal Penalties
XVII. Voluntary Consensus Standards
XVIII. Availability of Guidance
XIX. Availability of Documents
I. Background
The NRC is amending its regulations by adding an alternative risk-
informed, performance-based, and technology-inclusive regulatory
framework as an option for the licensing and regulation of future
commercial nuclear plants. This section discusses previous activities
that have led to the development of this final rule.
NRC Advanced Reactor Readiness
In its ``Policy Statement on the Regulation of Advanced Nuclear
Power Plants,'' dated July 8, 1986, the Commission stated that it
considered the term ``advanced'' to apply to reactors that are
significantly different from current (i.e., current in 1986) generation
light-water reactors (LWRs) then under construction or in operation,
and that ``advanced'' includes reactors that provide enhanced margins
of safety or utilize simplified inherent or other innovative means to
accomplish their safety functions. At the time, certain high
temperature gas-cooled reactors, liquid metal reactors, and LWRs of
innovative design were considered to be ``advanced.'' The 1986 policy
statement provided the Commission's policy regarding the review of, and
desired characteristics associated with, advanced reactors. The NRC
updated this statement in the ``Policy Statement on the Regulation of
Advanced Reactors,'' dated October 14, 2008 (Advanced Reactor Policy
Statement).
The agency has undertaken many activities related to advanced
reactors, including issuing an advance notice of proposed rulemaking
titled ``Approaches to Risk-Informed and Performance-Based Requirements
for Nuclear Power Reactors,'' dated May 4, 2006 (71 FR 26267). These
efforts were often done in parallel, and sometimes interwoven, with the
NRC's efforts to
[[Page 15698]]
improve risk-informed and performance-based approaches within the
agency (e.g., the Commission's PRA policy statement, ``Use of
Probabilistic Risk Assessment Methods in Nuclear Regulatory
Activities,'' dated August 16, 1995 (60 FR 42622)).
In 2016, the NRC issued ``NRC Vision and Strategy: Safely Achieving
Effective and Efficient Non-Light Water Mission Readiness'' (Advanced
Reactor Vision and Strategy Document), in response to increasing
interest in advanced reactor designs. The NRC considered the Department
of Energy's (DOE's) advanced reactor deployment goals in developing the
Advanced Reactor Vision and Strategy Document. Since publication of the
document, the NRC continues to manage its activities to support the
DOE's deployment goals. The Advanced Reactor Vision and Strategy
Document identified initiating and developing a new risk-informed and
performance-based regulatory framework as a possible long-term goal.
However, the NRC staff's initial efforts were focused on resolving
policy issues and developing guidance for licensing non-LWR
technologies under the existing regulatory frameworks (parts 50 and
52). The NRC staff issues annual Commission papers on the status and
progress of the NRC staff's activities related to advanced reactors
(e.g., SECY-24-0020, ``Advanced Reactor Program Status,'' dated
February 27, 2024). These Commission papers provide status updates for
advanced reactor activities undertaken both prior to and after
initiation of this rulemaking.
In 2017, the NRC staff prioritized activities to support the
development of technology-inclusive, risk-informed, and performance-
based licensing approaches that could be implemented under the existing
regulatory framework in parts 50 and 52. These activities leveraged
previous work described in NUREG-1860, ``Feasibility Study for a Risk-
Informed and Performance-Based Regulatory Structure for Future Plant
Licensing,'' published in 2007. One key element of these efforts was
the Licensing Modernization Project (LMP), a cost-shared initiative led
by nuclear utilities and supported by DOE. The LMP methodology is a
technology-inclusive, risk-informed, and performance-based methodology
developed for non-LWR designs. The LMP methodology provides a
systematic and reproducible process for licensing-basis event (LBE)
selection and evaluation; classification of structures, systems, and
components (SSCs); and assessment of defense in depth. The LMP
methodology refined the DOE's Next Generation Nuclear Plant Program
methodologies to reflect interactions with the NRC, to address feedback
from industry, and to broaden the scope of the approach to ensure
applicability to various non-LWR technologies. The LMP methodology
activities led to the publication and submittal of Nuclear Energy
Institute (NEI) 18-04, Revision 1, ``Risk-Informed Performance-Based
Technology-Inclusive Guidance for Non-Light Water Reactor Licensing
Basis Development,'' issued August 2019. The document indicates that
controlling the frequencies and potential consequences of a wide
spectrum of events is the primary focus of the LMP methodology.
The NRC endorsed the principles and methodology in NEI 18-04, with
clarifications, in RG 1.233, ``Guidance for a Technology-Inclusive,
Risk-Informed, and Performance-Based Methodology to Inform the
Licensing Basis and Content of Applications for Licenses,
Certifications, and Approvals for Non-Light-Water Reactors.'' The NRC
staff sought Commission approval of the use of the LMP methodology and
NEI 18-04 in SECY-19-0117, ``Technology-Inclusive, Risk-Informed, and
Performance-Based Methodology to Inform the Licensing Basis and Content
of Applications for Licenses, Certifications, and Approvals for Non-
Light-Water Reactors,'' dated December 2, 2019. In that paper, the
staff described the relationship between the LMP methodology and NEI
18-04 and previous relevant Commission decisions, including those
described in SECY-93-092, ``Issues Pertaining to the Advanced Reactor
(PRISM, MHTGR, and PIUS) and CANDU 3 Designs and their Relationship to
Current Regulatory Requirements,'' dated April 8, 1993. The Commission
approved the use of the LMP methodology and NEI 18-04 as a reasonable
approach for establishing key parts of the licensing basis and content
of applications for licenses, certifications, and approvals for non-
LWRs in Staff Requirements Memorandum (SRM) SRM-SECY-19-0117, dated May
26, 2020. Although the LMP methodology is technology-inclusive, the
industry and NRC staff initially focused the LMP methodology's
applicability on non-LWRs, both for efficiency and to support near-term
non-LWR applications under the existing regulatory framework, such as
the Advanced Reactor Demonstration Projects supported by DOE.
As stated in the part 53 rulemaking plan, SECY-20-0032, dated April
13, 2020, the NRC staff developed part 53 by building upon recent and
ongoing activities such as the LMP methodology described in SECY-19-
0117. Such an approach supports implementing the NEIMA direction to
establish a technology-inclusive framework as well as the requirement
to use, where appropriate, risk-informed and performance-based
techniques, and it also capitalizes on previous initiatives by the
industry, DOE, and the NRC. The LMP methodology highlights the role of
PRA in risk-informed and performance-based approaches to identifying
enhanced safety margins that can be used to justify operational
flexibilities. The part 53 framework is largely based on the
methodology described in SECY-19-0117 and includes a prominent role for
PRA, other systematic risk evaluations (SREs), or a combination
thereof.
II. Discussion
A. Objective and Applicability
The NRC is adding a new, alternative part to its regulations that
sets out a risk-informed, technology-inclusive framework for the
licensing and regulation of commercial nuclear plants. This new
approach achieves the following: (1) continue to provide reasonable
assurance of adequate protection of public health and safety and the
common defense and security; (2) promote regulatory stability,
predictability, and clarity; (3) reduce requests for exemptions from
the current requirements in parts 50 and 52; (4) establish new
requirements to address non-LWR technologies; (5) recognize
technological advancements in reactor design; and (6) credit the
possible response of some designs of commercial nuclear plants to
postulated accidents, including slower transient response times and
relatively small and slow release of fission products. This final rule
adds 10 CFR part 53; subpart M, ``Fitness-for-Duty Programs for
Facilities Licensed Under 10 CFR part 53,'' to part 26; Sec. 73.100,
``Technology-inclusive requirements for physical protection of licensed
activities at commercial nuclear plants against radiological
sabotage,'' Sec. 73.110, ``Technology-inclusive requirements for
protection of digital computer and communication systems and
networks,'' and Sec. 73.120, ``Access authorization program for
commercial nuclear plants,'' as well as makes conforming changes
throughout 10 CFR chapter I, ``Nuclear Regulatory Commission.''
B. Need for Changes to the Existing Regulatory Framework
The NRC has long recognized that the licensing and regulation of a
variety of nuclear reactor technologies presents
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challenges because the existing regulatory framework has evolved
primarily to address the LWR designs that compose the current operating
fleet. The NRC has had many interactions with designers of various
reactor technologies under development, sometimes collectively referred
to as advanced reactors. The interactions have informed the development
of policies and guidance to support the potential licensing of new and
different types of reactor facilities, some of which may not utilize
LWR designs. The NRC issued its Advanced Reactor Policy Statement to
provide all interested parties, including the public, with the
Commission's views concerning the desired characteristics of advanced
reactor designs. The NRC further described its early efforts to
establish a technology-inclusive approach to the regulation of nuclear
reactors in the advance notice of proposed rulemaking published in
2006. The NRC acknowledged in its ``Report to Congress: Advanced
Reactor Licensing,'' issued August 2012, that ``while the safety
philosophy inherent in the current regulations applies to all reactor
technologies, the specific and prescriptive aspects of those
regulations clearly focus on the current fleet of LWR facilities.''
Congress similarly recognized the potential benefits of developing
a regulatory infrastructure to support the development and
commercialization of advanced nuclear reactors. Consequently, Congress
passed NEIMA in late 2018, and the President signed it into law in
January 2019. NEIMA directed the NRC to undertake a rulemaking to
establish a technology-inclusive regulatory framework for optional use
by applicants for new commercial advanced nuclear reactor licenses. In
addition, on July 9, 2024, the President signed into law the
Accelerating Deployment of Versatile, Advanced Nuclear for Clean Energy
Act of 2024, also referred to as the ADVANCE Act. The NRC has evaluated
the ADVANCE Act, including how NRC regulations, such as part 53 or
future revisions to it, could be used to address provisions in the
ADVANCE Act. The ADVANCE Act contains provisions on a variety of
nuclear-related topics, such as microreactors, nuclear reactor license
application reviews, and nuclear fuel. Finally, in 2025, the President
signed E.O. 14300, ``Ordering the Reform of the Nuclear Regulatory
Commission,'' which builds on the provisions in the ADVANCE Act. E.O.
14300 will complement this rulemaking by providing additional
mechanisms for streamlining the agency's efforts to provide an
efficient licensing pathway for advanced reactors.
The requirements in part 53 support a wide variety of potential
commercial nuclear reactor technologies. The current regulatory
framework in parts 50 and 52 evolved in the context of the current
operating reactor fleet dominated by LWRs and as a result includes
provisions specific to LWR technologies. While the NRC can license
other reactor technologies under the current framework by using
existing regulatory flexibilities and the exemption process, there is
significant interest in developing a regulatory framework that is
flexible enough to accommodate multiple technologies and robust enough
to ensure a level of safety equivalent to parts 50 and 52, consistent
with the Commission's Advanced Reactor Policy Statements. The
Commission reiterated its safety expectations for new reactors in the
SRM for SECY-10-0121, ``Modifying the Risk-Informed Regulatory Guidance
for New Reactors,'' dated March 2, 2011:
Because new plant designs incorporate operating experience from
current generation reactors, severe accident research, and risk
insights from design probabilistic risk assessments, the Commission
expects that the advanced technologies incorporated in new reactors
will result in enhanced margins of safety. However, the Commission
continues to expect (consistent with the 2008 Advanced Reactor
Policy Statement), as a minimum, at least the same degree of
protection of the public and the environment that is required for
current-generation light-water reactors. New reactors with these
enhanced margins and safety features should have greater operational
flexibility than current reactors.
However, developing a regulatory framework that can accommodate a
wide range of technologies while maintaining an acceptable level of
safety presents significant regulatory challenges. The existing
regulations have been developed over the course of decades and reflect
changes to address events discovered through operating experience. As a
result, the existing regulations have benefited from a focused and
tailored treatment of safety issues as issues arose and evolved. In
contrast, part 53 is being developed to accommodate technologies that,
in some cases, lack significant operating experience. This lack of
operating experience makes it challenging to develop technology-
inclusive regulatory requirements when it is less well-known which
issues may be more or less important to safety for any given set of
technologies. To address these challenges, the NRC drew on well-
developed approaches to licensing to produce a technology-neutral and
robust regulatory framework. The regulatory framework uses PRAs, other
SREs, or a combination thereof, to assess risks and focus on the issues
most important to safety, help establish technical requirements, and
manage operations. The framework builds on the LMP methodology, which
is a technology-inclusive approach to licensing that leverages risk
insights to provide applicants with significant design and operation
flexibilities.
C. 10 CFR Part 53 Framework
This final rule consists of several major components, including a
new part 53, to be added to 10 CFR chapter I, revisions for part 26,
part 50, and part 73, and conforming changes throughout 10 CFR chapter
I. The major features of this final rule include the following:
(1) Technology-inclusiveness. This rule provides a broad and
flexible regulatory framework that can be used for any reactor
technology, any size reactor, and any reactor end use.
(2) Risk-informed framework to support safety-focused decision-
making. Part 53 provides a holistic, risk-informed framework that
offers substantial flexibility in leveraging safety margins and
focusing on design features and programmatic controls important to
protecting public health and safety. The framework allows for explicit
consideration of risk through the use of PRAs or other SRE techniques,
or a combination thereof, to generate risk insights, and to assess and
manage those risks. This approach departs from traditional
deterministic methods, notably the use of the single-failure criterion,
by enabling applicants to propose comprehensive risk metrics and
associated risk performance objectives, appropriate systematic risk
assessment techniques, and to demonstrate how their design and
associated programmatic controls protect public health and safety.
(3) Performance-based approach. Part 53 is a performance-based
framework that provides flexibility in establishing appropriate high-
level safety objectives and demonstrating how a reactor design or
specific commercial nuclear plant meets those objectives. Rather than
prescribing specific methods or processes, the performance-based
approach in part 53 promotes efficiency and innovation by allowing
applicants to propose design features to meet safety objectives and
achieve safety outcomes. This will support novel concepts such as
leveraging functional containment concepts, alternative siting criteria
for commercial nuclear reactors in relation to population centers,
reduced staffing
[[Page 15700]]
levels, and remote operations, while eliminating traditional,
prescriptive requirements, such as general design criteria and aircraft
impact assessments.
(4) Licensing pathways that accommodate a broad spectrum of design
maturities and deployment models. Part 53 provides several licensing
options for applicants to choose from to meet their deployment model or
business case needs, including the licenses, certifications, and
approvals provided by parts 50 and 52. This final rule provides
additional flexibility for manufacturing licenses (MLs), including the
possible factory loading of fuel into manufactured reactors with
appropriate features to prevent criticality for deployment to another
location for operation.
(5) Operator licensing. Part 53 introduces the concept of self-
reliant-mitigation facilities and the use of generally licensed reactor
operators (GLROs) for those facilities. The allowance for GLROs
provides flexibility for the types and locations of staffing needed
under part 53.
(6) Efficiency. Part 53 provides opportunities to improve
regulatory efficiency by including provisions for licensing first-of-a-
kind proposals as well as provisions that benefit those proposing
standardized and repetitious applications. Part 53 provides finality to
designs for which an operating license has been issued to improve its
incorporation into a standardized design approval or certification.
Part 53 also provides for a risk-informed approach for managing plant
equipment and programmatic controls that reduce the future need for
regulatory approvals.
(7) Codes and standards. Part 53 does not incorporate by reference
specific codes and standards as is done in Sec. 50.55a, ``Codes and
standards.'' Instead, part 53 allows the use of generally accepted
codes and standards to be tailored to the assessed safety significance
of SSCs, such as the use of non-nuclear codes and standards for SSCs
composed of commercial grade components.
Part 53 is comprised of subparts A through M. These provisions are
organized to provide high-level performance criteria and to specify
requirements to demonstrate compliance with those performance criteria
throughout major stages of the life cycle of commercial nuclear plants.
This organization reflects a systems-engineering style approach to the
design, licensing, operation, and ultimately decommissioning of future
commercial nuclear plants. Organizing requirements in this manner also
supports performance-based approaches. Required programs (e.g.,
radiation protection) and monitoring (e.g., technical specification
(TS) surveillance) during the operations phase that are similar to
those required by part 50 complement the design and analysis
requirements in subpart C. The performance-based approach adopted in
part 53 also includes regulatory requirements that allow applicants to
use a flexible and graded approach to the performance of safety
functions based on the role of a particular SSC, human action, or
program in limiting the overall risks to the public below accepted
standards through balanced measures to prevent and mitigate possible
events.
Subpart M of part 26 is new and is largely consistent with the
objective-based 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. Subpart M
of part 26 is 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
requirements are not entirely equivalent because current subpart K of
part 26 only applies during construction of the commercial nuclear
plant, whereas subpart M of part 26 applies during construction,
operation, and decommissioning. Furthermore, subpart M of part 26
allows 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.
Revisions to part 73 establish a new technology-inclusive,
consequence-based approach for a range of security areas, including
physical security, cybersecurity, and access authorization (AA) for
commercial nuclear reactors. The NRC used operating experience to
include additional regulatory flexibility for a part 53 licensee's
implementation of security requirements.
In addition, this final rule makes conforming changes throughout 10
CFR chapter I, by adding ``and part 53'' where appropriate to account
for the addition of part 53.
III. Part 53 Framework
Subpart A--General Provisions
Subpart A provides the general provisions applicable to all
applicants and licensees that are established in part 53 for the
issuance, amendment, and termination of licenses, permits,
certifications, and approvals for commercial nuclear plants licensed
under section 103 of the Atomic Energy Act of 1954, as amended (the
AEA) and title II of the Energy Reorganization Act of 1974 (88 Stat.
1242). Subpart A includes purpose, scope, definitions, written
communications, employee protections, completeness and accuracy of
information, exemptions, standards for review, jurisdictional limits,
consideration of attacks and destructive acts by enemies of the United
States, and information collection requirements.
The requirements in subpart A are largely equivalent to the general
requirements in part 50 that are applicable to all part 50 applicants
and licensees (specifically, Sec. Sec. 50.1 through 50.13) but
reference the corresponding regulations in part 53 in place of
references to part 50.
A. Discussion of Definitions in Part 53
This final rule includes a definition section in Sec. 53.020. The
definitions of most terms in Sec. 53.020 are equivalent to the
corresponding terms defined in: (1) Sec. Sec. 50.2, 52.1, and other
NRC regulations; (2) NEI 18-04, as endorsed by RG 1.233; or (3)
American Society of Mechanical Engineers (ASME)/American Nuclear
Society Risk Assessment Standard (RA-S)-1.4-2021, as endorsed for trial
use by RG 1.247, ``Acceptability of Probabilistic Risk Assessment
Results for Non-Light-Water Reactor Risk-Informed Activities.'' This is
intended to provide clarity and consistency in terminology where
possible and to utilize past and ongoing NRC initiatives to support the
licensing of new reactors. Specific deviations from existing
definitions are further explained in the following paragraphs.
Regarding the definition of ``Commercial nuclear plant'' and
``Commercial nuclear reactor'' in Sec. 53.020, as noted previously,
the NRC initially considered establishing the scope of part 53 as being
for ``advanced nuclear plants.'' The preliminary proposed rule language
defined ``advanced nuclear plant'' as ``a utilization facility
consisting of one or more advanced nuclear reactors'' as defined in
NEIMA. NEIMA defines the term ``advanced nuclear reactor'' as ``a
nuclear fission reactor or fusion machine, including a prototype plant
(as defined in sections 50.2 and 52.1 of 10 CFR (as in effect on the
date of enactment of this Act)), with significant improvements compared
to commercial nuclear reactors under construction as of the date of
enactment of this Act,
[[Page 15701]]
including improvements such as--(A) additional inherent safety
features; (B) significantly lower levelized cost of electricity; (C)
lower waste yields; (D) greater fuel utilization; (E) enhanced
reliability; (F) increased proliferation resistance; (G) increased
thermal efficiency; or (H) ability to integrate into electric and
nonelectric applications.''
Based on public discussions on the use of the term, the NRC
determined that the NEIMA definition, although broad, did not define
``significant improvements'' with enough specificity to implement in
NRC regulations. Additionally, a number of stakeholders suggested that
the descriptor ``advanced'' implied enhanced safety, while the NEIMA
definition includes ``significant improvements'' in areas other than
safety enhancements. In response to this feedback, and to be
technology-inclusive, the NRC determined that the broader term
``commercial nuclear plant'' is preferable. The NEIMA definition of
advanced nuclear reactor also includes fusion technologies. Fusion
energy systems have not been included in the scope of part 53 but are
the subject of a separate rulemaking activity, ``Regulatory Framework
for Fusion Systems.'' See NRC docket ID NRC-2023-0017 on the Federal
rulemaking website https://www.regulations.gov.
The NRC allows the use of part 53 by any ``commercial nuclear
plant.'' The use of the term ``plant'' versus ``reactor,'' as used in
existing regulations (i.e., Sec. 50.2), recognizes that co-located
support facilities and radionuclide sources need to be considered in
the licensing of a facility. The phrase ``commercial purposes,'' as
used in the definition of ``commercial nuclear plant,'' includes
purposes such as providing process heat for a variety of industrial
applications (e.g., desalination, oil refining, hydrogen production).
The NRC has not compiled a complete list of such commercial purposes.
The definition of ``Commercial nuclear plant'' refers to a ``Commercial
nuclear reactor,'' which is defined based on the definition of
``Nuclear reactor'' in Sec. 50.2. However, the phrase ``in a self-
supporting chain reaction'' is not included in the definition of
Commercial nuclear plant to enable applying part 53 to accelerator
driven systems that use special nuclear material (SNM) but that do not
involve self-sustaining chain reactions. Relatedly, ``Utilization
facility'' is also defined in Sec. 53.020 based on the definition of
that term in Sec. 50.2 and refers to a ``Commercial nuclear plant'' as
defined in Sec. 53.020.
The definition of ``Construction'' is different from the definition
in Sec. 50.10. Because the regulatory framework in part 53 uses risk-
informed, less prescriptive, and performance-based requirements as
compared to part 50, the part 53 definition takes a different approach
in determining what activities are prohibited without an NRC license.
Under the part 53 approach, the definition of ``Construction''
specifies a variety of activities that are applicable to safety-related
(SR) and non-safety-related but safety-significant (NSRSS) SSCs and are
credited or relied upon for demonstrating compliance with safety
criteria defined in subpart B of part 53 as well as SSCs necessary to
comply with part 73 and onsite emergency facilities necessary to comply
with Sec. 53.855. By listing the activities for SR and NSRSS SSCs that
are credited or relied upon for demonstrating compliance with safety
criteria defined in subpart B, this definition describes activities
related to SSCs subject to some sort of special treatment, as that term
is defined in Sec. 53.020. These special treatment requirements, which
include quality assurance, design criteria, and programmatic controls,
apply to safety-related SSCs and the set of non-safety-related SSCs for
which a license is required to authorize construction activities. The
latter category includes a facility's NSRSS SSCs. The non-safety-
significant SSCs not subject to special treatment and NSRSS SSCs for
which special treatments are limited to operational controls are, in
general, identified as ``commercial grade'' and may be designed,
procured, and installed in accordance with the usual practices employed
for industrial plants. Importantly, under the part 53 definition, an
SSC that falls outside the definition of construction may still be
subject to the NRC's statutory authority during operations. In view of
the foregoing, the definition of ``Construction'' in Sec. 53.020 is
consistent with the provisions of the AEA related to construction
permits, while simultaneously allowing activities related to SSCs that
are commercial grade but which could still be subject to the NRC's
jurisdiction during operations. This definition also includes the
listed activities which are for SSCs necessary to comply with part 73
or onsite emergency facilities necessary to comply with Sec. 53.855.
The inclusion of the listed activities which are for these SSCs is
consistent with Sec. 50.10(a)(1)(v) and (vii), which include
activities for corresponding SSCs. Including these activities in the
definition of ``Construction'' is appropriate because, in both
instances, part 53 points back to the relevant existing frameworks in
part 73 and the relevant part 50 requirements, respectively, rather
than creating an entirely new framework. Section 53.020 also adds
definitions for terms related to event selection (LBEs, design-basis
accidents (DBAs), anticipated event sequences, unlikely event
sequences, and very unlikely event sequences); equipment
classifications (SR, NSRSS, and non-safety-significant SSCs);
performance metrics (e.g., safety criteria and functional design
criteria); and special treatment.
The regulation defines ``Safety criteria'' in terms of the plant-
level performance-based metrics that are provided in Sec. Sec. 53.210
and 53.220. The term ``Functional design criteria'' is defined as
metrics for the performance of specific SSCs that are determined from
the role of the SSC in meeting the safety criteria. These are new terms
that have not previously been defined or used in NRC regulation.
The term ``Safety-related SSCs'' refers to those SSCs needed to
meet the safety criteria in Sec. 53.210. The term ``Non-safety-related
but safety-significant SSCs'' means those SSCs that are not SR because
they are not relied upon to perform any function necessary to
demonstrate compliance with Sec. 53.210 but warrant special treatment
because they are relied on to achieve adequate defense in depth or
perform risk-significant functions. The term ``Non-safety-significant
SSCs'' means those SSCs that are not SR or NSRSS.
The term ``Programmatic controls'' means administrative measures
that govern human action in implementing programs and operating,
monitoring, and maintaining SSCs and equipment of a commercial nuclear
plant.
The terms ``Design-basis accidents,'' ``Anticipated event
sequences,'' ``Unlikely event sequences,'' and ``Very unlikely event
sequences'' are defined to be different types of ``Licensing-basis
events'' and are also largely equivalent to the LMP methodology's
definitions of DBAs, anticipated operational occurrences (AOOs),
design-basis events (DBEs), and beyond-design-basis events,
respectively. The term ``Design-basis accidents'' is defined as
postulated event sequences that are used to set functional design
criteria and performance objectives for the design of SR SSCs through
deterministic analyses. Design-basis accidents are derived from the
unlikely event sequences from the PRA, a type of SRE, other SREs, or a
combination thereof, and then analyzed in a conservative approach by
[[Page 15702]]
prescriptively assuming that only SR SSCs are available to mitigate
postulated accident scenarios. Within the LMP methodology, event
sequences with mean frequencies of 1x10\-2\/plant-year and greater are
classified as anticipated event sequences. Within the LMP methodology,
infrequent event sequences with mean frequencies of 1x10\-4\/plant-year
to 1x10\-2\/plant-year are classified as unlikely event sequences.
``Very unlikely event sequences'' are less likely to occur than
unlikely event sequences. Within the LMP methodology, rare event
sequences with frequencies of 5x10\-7\/plant-year to 1x10\-4\/plant-
year are classified as very unlikely event sequences. While the
terminology for these event sequences creates some differences between
part 53 and the LMP methodology, part 53 uses new terms for these event
sequences specifically to avoid conflicts with terms already used
within part 50 and part 52 to represent different concepts. Further,
because some stakeholder comments demonstrated confusion related to the
history of beyond-design-basis accidents terminology, these definitions
seek to clarify the event categories in part 53. Finally, although the
term ``event sequence'' is often used in the context of a PRA, that
term is used generically in part 53 and does not imply the use of a
specific type of SRE, such as a PRA. The sections of this preamble
related to subparts B and C provide additional discussion of LBEs.
Section 53.020 includes a definition of ``Special treatment'' to
explain that it means those requirements, such as quality assurance
(QA), design criteria, and programmatic controls, that are taken beyond
the procurement, installation, and maintenance of commercial grade
products. Routine commercial practices may include the use of selected
consensus codes and standards that are cited in applications to support
the identification of special treatments that may go beyond what would
otherwise be required by those selected commercial codes and standards.
The special treatments increase confidence that SR and NSRSS SSCs will
provide defense in depth, or perform risk-significant functions, under
service conditions and with SSC reliabilities that are consistent with
the analysis required in subpart C. Structures, systems, and components
designated as SR also contribute to defense in depth and risk-
significant functions and may warrant special treatments beyond those
defined for the SR functions needed for compliance with Sec. 53.210.
To maintain alignment with definitions in part 52, the NRC has
added a definition of early site permit (ESP). The NRC proposed
definitions for ``Consensus code or standard'' and ``probabilistic risk
assessment'' but is not including a definition for these terms in this
final rule because these terms were determined not to be essential for
the framework and including the definitions could introduce issues with
consistency given alternative definitions developed by other
organizations.
B. Other General Provisions
Section 53.040 governs written communications and how applications
and other required information must be submitted to the NRC. These
requirements are equivalent to those in Sec. 50.4.
Section 53.050 establishes 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 are
equivalent to those in Sec. 50.5.
Section 53.060 prohibits discrimination against an employee of a
holder or applicant for an NRC license, permit, design certification
(DC), or design approval, or a contractor or subcontractor of a holder
or applicant for an NRC license, permit, DC, or design approval for
engaging in certain protected activities. Section 53.060 also
prescribes a procedure for seeking a remedy for employees who believe
they have been discriminated against for engaging in such protected
activities. These requirements are equivalent to those in Sec. Sec.
50.7 and 52.5.
Section 53.070 governs the completeness and accuracy of information
provided to the NRC. These requirements are equivalent to those in
Sec. Sec. 50.9 and 52.6.
Section 53.080 governs exemptions from the requirements of the
regulations in part 53. These requirements are equivalent to those in
Sec. Sec. 50.12 and 52.7.
Paragraphs (a) through (d) of Sec. 53.090 establish requirements
for standards that the NRC will consider in determining whether a
construction permit (CP), operating license (OL), ESP, combined
license, or ML under part 53 will be issued to an applicant. These
requirements are equivalent to those in Sec. Sec. 50.40, 50.42, 50.43
and 50.22, respectively. Requirements equivalent to those in Sec. Sec.
50.41 and 50.21 are not included in part 53 because they apply to Class
104 licenses, and part 53 does not apply to those licenses.
Section 53.100 requires that no license issued under part 53 may
cover activities that are not under or within the jurisdiction of the
United States. These requirements are equivalent to those in Sec.
50.53.
Section 53.110 states that licensees and applicants are not
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 are equivalent to those in Sec. 50.13.
Section 53.115 establishes requirements for rights related to SNM.
These requirements are equivalent to those in Sec. 50.54(b) and (c).
Section 53.117 establishes 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 are equivalent to those in Sec. 50.54(d).
Section 53.120 establishes requirements for information collection
requirements that have received Office of Management and Budget (OMB)
approval. These requirements are equivalent to those in Sec. 50.8.
Subpart B--Technology-Inclusive Safety Requirements
Subpart B, ``Technology-Inclusive Safety Requirements,'' provides
technology-inclusive safety criteria that serve as performance
standards for the subsequent performance-based requirements used
throughout part 53. Subsequent subparts define how specific activities
during various stages of the life cycle of a commercial nuclear plant
contribute to satisfying these high-level performance standards. The
performance standards in subpart B also establish a means to determine
appropriate regulatory controls for SSCs, human actions, and programs
in the following subparts. For example, the classification of SR SSCs
is built upon the safety criteria in Sec. 53.210, ``Safety criteria
for design-basis accidents.'' The more detailed requirements for those
SSCs are then further defined in the design and analysis requirements
in subpart C, ``Design and Analysis Requirements.'' The activities for
manufacturing, constructing, and maintaining the SR SSCs are governed
by subpart E, ``Construction and Manufacturing Requirements,'' and
subpart F, ``Requirements for Operation.''
Requirements for NSRSS SSCs warranting special treatment are
[[Page 15703]]
likewise determined under Sec. 53.220, ``Safety criteria for
licensing-basis events other than design-basis accidents,'' in subpart
B and Sec. 53.460, ``Safety categorization and special treatment,'' in
subpart C. Regulatory requirements related to the NSRSS SSCs are
distinguished from the regulatory requirements for SR SSCs throughout
part 53. Part 53 affords more flexibility to applicants and licensees
regarding how NSRSS SSCs are used in the design and maintained during
plant operations, as compared to SR SSCs.
The collective set of performance-based requirements in part 53 are
sufficient, if met, for the NRC to make the findings required to grant
an application for a utilization facility under section 182 of the AEA
that the utilization of SNM will be in accord with the common defense
and security and will provide adequate protection to the health and
safety of the public. This construct is similar to existing NRC
regulations, which the Commission has said on many occasions do not
specifically define ``adequate protection.'' However, compliance with
NRC regulations may be presumed to assure adequate protection at a
minimum. The requirements throughout part 53 that support demonstrating
compliance with Sec. 53.220 are similar to current regulations that
both contribute to assuring adequate protection of public health and
safety and are desirable to promote the common defense and security or
to protect health or to minimize danger to life or property under
section 161 of the AEA.
Consistent with historical practice, sections 182 and 161 of the
AEA are cited as authorizing legislation within this final rule.
However, specific language from the AEA is not incorporated into the
safety objectives or safety criteria in part 53. This is because, again
consistent with historical practice, the NRC is not defining ``adequate
protection'' through the individual safety requirements in part 53.
Rather, part 53 enables the NRC to make its required findings under the
AEA by providing sufficient performance standards, safety criteria, and
related requirements on how applicants must demonstrate compliance with
subpart B and other subparts.
Section 53.210 provides safety criteria for DBAs that are required
to be identified under Sec. 53.240 and analyzed under Sec. 53.450(f)
in subpart C of part 53. Subsequent sections in part 53 require that
the SSCs relied upon to demonstrate compliance with the criteria in
Sec. 53.210 be classified as SR. The use of SR SSCs and the 25 rem
reference values for potential radiological consequences aligns with
traditional deterministic approaches for LWRs from Sec. Sec. 50.34,
52.79, and 100.11 for evaluating the effectiveness of plant design
features with respect to postulated reactor accidents. A footnote
similar to that included in Sec. 50.34(a)(1)(ii)(D)(1) and Sec.
52.79(a)(1)(vi)(A) is included in Sec. 53.210 to explain that the use
of the 25 rem value is not intended to imply that this number
constitutes an acceptable limit for an emergency dose to the public
under accident conditions. Rather, this dose value has been set forth
in this section as a reference value that is used in the evaluation of
plant design features with respect to DBAs to verify that the proposed
designs would provide assurance of low risk of public exposure to
radiation in the event of an accident. The inclusion of the safety
criteria for DBAs in subpart B provides a logical structure supporting
the identification and treatment of SR SSCs and establishing the
corresponding functional design criteria for those SSCs.
Section 53.220 provides safety criteria for LBEs other than DBAs
that are required to be identified under Sec. 53.240 and analyzed
under Sec. 53.450(e) in subpart C. Whereas Sec. 53.210 and the
related requirements for SR SSCs provide that a defined success path
exists for DBAs, the safety criteria for LBEs other than DBAs establish
the connections between SSC design, human actions, and programmatic
controls and a broader set of potential internal and external hazards.
These safety criteria also address defense-in-depth matters such as a
balanced consideration of prevention and mitigation.
The safety criterion in Sec. 53.220(b) includes a requirement to
use a comprehensive risk metric or set of metrics and associated risk
performance objectives against which calculated values of the risk
metrics are compared. The comprehensive risk metrics or set of metrics
and associated risk performance objectives support a performance-based
approach to developing an appropriate combination of design features
and programmatic controls to prevent or mitigate LBEs other than DBAs.
The applicant must propose the comprehensive risk metric or set of
metrics and associated risk performance objectives, and the
comprehensive risk metric or set of metrics and associated risk
performance objectives must provide an appropriate level of safety.
Comprehensive risk metrics should consist of a proposed plant risk
metric or set of proposed risk metrics that approximate the total,
overall risk from the facility and that address the range of possible
plant configurations and associated internal and external hazards to
the extent practicable. The associated risk performance objectives are
pre-established, indicative values of the comprehensive risk metrics
that are used as part of risk-informed decision-making. The methodology
for developing and using proposed comprehensive risk metrics and
associated risk performance objectives is defined by the requirements
for analyses in Sec. 53.450. Therefore, the application must include a
description of that methodology and, among other things, should explain
the initial conditions, boundary conditions, and key assumptions used
to develop and calculate the risk metrics. Screening tools and bounding
or simplified methods may be used for any mode or hazard, provided that
the applicant provides an acceptable technical basis. As with all risk-
informed methodologies, treatment of uncertainties must be addressed.
The risk performance objectives established under this methodology
are likely to involve assessing and averaging the risks over a period
of time (e.g., plant year) and do not constitute a real-time
requirement that must be continuously demonstrated by the licensee. The
use of a comprehensive risk metric or set of risk metrics and risk
performance objectives that reflect an average risk to establish
performance goals for SR and NSRSS SSCs is consistent with current
practices that use other risk assessment techniques to address short-
term plant configurations during plant maintenance activities.
It is worth noting that the evaluation of plant risks, as
represented by a comparison of analysis results to acceptable risk
performance objectives for comprehensive risk metrics, is one of
several performance standards used in subpart B. The use of multiple
performance standards, including deterministic criteria and defense-in-
depth measures, reflects an integrated decision-making process similar
to that described in RG 1.174, ``An Approach for Using Probabilistic
Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to
the Licensing Basis,'' Revision 3. The NRC's approval of using a
comprehensive risk metric or set of metrics with associated risk
performance objectives is not, by itself, an indicator of adequate
protection. Rather, the comparison of comprehensive risk metrics to
associated risk performance objectives that are acceptable to the NRC
is part of a suite of regulatory requirements that,
[[Page 15704]]
when considered holistically, form the basis for the NRC's decision-
making. This is analogous to the approach used for plants licensed
under part 50 and part 52, where no single regulatory requirement
governs whether a plant is ``safe enough.''
The RG 1.233, ``Guidance for a Technology-Inclusive, Risk-Informed,
and Performance-Based Methodology to Inform the Licensing Basis and
Content of Applications for Licenses, Certifications, and Approvals for
Non-Light-Water Reactors,'' describes an example of an acceptable
approach for identifying and analyzing LBEs under part 50 and part 52,
including the use of the quantitative health objectives (QHOs) stated
in the NRC's policy statement, ``Safety Goals for Nuclear Power Plant
Operation,'' dated August 4, 1986 (51 FR 28044), as corrected and
republished August 21, 1986 (51 FR 30028) (Safety Goals Policy
Statement), as acceptable performance objectives for comprehensive risk
metrics. The use of comprehensive risk metrics, such as the individual
early fatality risk (IEFR) and the individual latent cancer fatality
risk (ILCFR), and associated risk performance objectives, such as the
QHOs, from the Safety Goals Policy Statement, could form the basis for
one approach to meet Sec. 53.220(b). The requirement for comprehensive
risk metrics, in combination with the other requirements in subparts B
and C, brings the approach endorsed in RG 1.233 for parts 50 and 52
into part 53. Additionally, the use of comprehensive risk metrics and
associated risk performance objectives provides a logical performance
objective to support the risk management approaches in the various
subparts comprising part 53.
The Commission stated in the introduction of the Safety Goals
Policy Statement that improvements to then-current regulatory practices
could lead to a more coherent and consistent regulation of nuclear
power plants, a more predictable regulatory process, a better public
understanding of the regulatory criteria that the NRC applies, and
public confidence in the safety of operating plants. Accordingly, the
Commission announced the safety goals with a focus on the risks to the
public from nuclear power plant operation. Following the issuance of
the Safety Goals Policy Statement, the NRC has used the comprehensive
risk metrics and performance objectives provided in the safety goals
within the criteria for many decisions involving safety judgments
during the licensing and regulation of operating reactors and proposed
nuclear reactor designs. Consistent with NUREG-0880, the proposed
comprehensive risk metrics and associated risk performance objectives
required under Sec. 53.220(b) can be expressed in terms of a
biologically average individual in terms of age and other risk factors.
Although some comprehensive risk objectives such as the IEFR and ILCFR
are defined in terms of fatality risks, the Commission continues to
make clear that no death attributable to nuclear power plant operation
will ever be ``acceptable'' in the sense that the Commission would
regard it as a routine or permissible event. Comprehensive risk metrics
and associated risk performance objectives as used in this final rule
establish acceptable risks, not acceptable deaths.
Applicants under part 53 may choose to develop and seek NRC
approval of comprehensive risk metrics or sets of risk metrics and
associated risk performance objectives beyond those previously
discussed, including the use of surrogate measures for use in specific
analyses to satisfy the requirements in Sec. 53.220(b). Such surrogate
measures for comprehensive risk metrics and associated risk performance
objectives could be used in a manner similar to the use of core damage
frequency and conditional containment failure probability for LWRs
within the safety goal evaluation process in NUREG/BR-0058,
``Regulatory Analysis Guidelines of the U.S. Nuclear Regulatory
Commission,'' and other assessments of LWRs using the NRC's safety
goals. The NRC will, as appropriate, review novel approaches for
comprehensive risk metrics and associated risk performance objectives
proposed by applicants, industry organizations, or standard development
organizations and will engage stakeholders during the development of
the related regulatory guidance or specific licensing actions.
Section 53.230 requires safety functions needed to ensure that the
safety criteria under Sec. Sec. 53.210 and 53.220 can be met if an
assumed LBE were to occur at a commercial nuclear plant. Section 53.230
specifies that limiting the release of radioactive materials from the
facility is the primary safety function, and therefore, limiting
potential offsite consequences (i.e., dose to a hypothetical
individual) can be used as the primary performance metric throughout
part 53. The additional or subsidiary safety functions needed to limit
the release of radionuclides may include, without limitation,
controlling processes related to reactivity, heat generation, heat
removal, and chemical interactions. This final rule provides
flexibility to applicants and licensees in identifying, implementing,
and maintaining the safety functions supporting retention of
radionuclides for commercial nuclear plants of varying sizes and
technologies.
Section 53.240 requires applicants to identify and address LBEs.
LBEs are unplanned events, resulting from both internal and external
hazards, that are used in the design and analyses required under part
53 for licensing commercial nuclear plants. This ensures estimates of
offsite consequences from analyses performed under Sec. 53.450 are
below the safety criteria identified under Sec. Sec. 53.210 and 53.220
and that SSCs, personnel, and programs address the safety functions
from Sec. 53.230. Including a high-level performance requirement
related to the identification of LBEs to address appropriate risk-
informed combinations of malfunctions of plant SSCs, human errors,
facility hazards, and the effects of external hazards and analysis
thereof in subpart B reflects the historical and continuing importance
of evaluating unplanned events as part of the licensing of commercial
nuclear plants. Section 53.240 requires identification and analysis of
LBEs under Sec. 53.450 using a PRA, other SREs, or a combination
thereof. An example of acceptable methods of using PRAs to identify and
assess LBEs is the methodology in RG 1.233, as discussed in RG 1.254,
``Technology-Inclusive Identification of Licensing Events for
Commercial Nuclear Plants.''
Section 53.250 establishes defense-in-depth requirements based on
the longstanding philosophy of providing defense in depth to address
uncertainties about the design, operation, and performance of
commercial nuclear plants. For example, parts 50 and 52 address defense
in depth through layered prescriptive technical requirements (e.g.,
fuel performance, cladding integrity, reactor coolant system integrity,
containment performance) for LWRs. In contrast, the flexibility
afforded to applicants in how they propose to demonstrate compliance
with the high-level safety criteria within part 53 necessitates this
specific requirement to ensure defense in depth is provided. The
requirements in this section state that no single engineered design
feature, human action, or programmatic control, no matter how robust,
should be exclusively relied upon to address the range of LBEs other
than DBAs. The requirement under Sec. 53.250(c) is different from the
single failure criterion described in appendix A to part 50. The Sec.
53.250(c) requirement does not allow the safety analysis to exclusively
rely upon a
[[Page 15705]]
single engineered design feature, human action, or programmatic control
to address the range of LBEs other than DBAs (i.e., ranging from very
unlikely event sequences to anticipated event sequences). In contrast,
the single failure criterion under appendix A to part 50 relates, in
part, to the failure of a component to perform its intended safety
function, regardless of whether that component was exclusively relied
upon to address the range of LBEs. This means the requirement under
Sec. 53.250(c) does not strictly disallow single failures, as defined
in appendix A to part 50, because a component could experience such a
single failure and, if it is not otherwise exclusively relied upon to
address the range of LBEs other than DBAs, its failure alone does not
preclude being able to satisfy Sec. 53.250(c). In that regard, Sec.
53.250 allows for greater flexibility such that other measures could be
taken to ensure appropriate defense in depth without needing to
accommodate single failures, as defined in appendix A to part 50. The
phrase ``engineered design feature'' does not preclude the possible
crediting of inherent characteristics within the design and analysis
for commercial nuclear reactors. While defense in depth is only
assessed for LBEs other than DBAs, the need to ensure dedicated success
paths for DBAs contributes to the overall defense in depth for each
commercial nuclear plant under part 53.
Section 53.260 governs normal operations and establishes a level of
safety based on requirements in 10 CFR part 20, ``Standards for
Protection Against Radiation,'' which limit doses to members of the
public and dose rates in unrestricted areas.
Section 53.270 provides for the protection of plant workers and
establishes a level of safety based on requirements in 10 CFR part 20,
which limit occupational dose.
Subpart C--Design and Analysis Requirements
This subpart provides requirements for the design of commercial
nuclear plants and the supporting analyses, including the analyses of
LBEs, to demonstrate that the performance standards in subpart B can be
satisfied. The sections within subpart C reflect the overall hierarchy
throughout part 53, which covers: (1) plant-level safety criteria
(Sec. Sec. 53.210 and 53.220); (2) safety functions (Sec. 53.230)
needed to demonstrate compliance with the safety criteria; (3) design
features (Sec. 53.400), human actions, and programmatic controls
needed to fulfill the safety functions; and (4) functional design
criteria (Sec. Sec. 53.410 and 53.420) that must be defined for each
design feature relied upon to demonstrate the safety criteria
(Sec. Sec. 53.210 and 53.220) are met. Subpart C also contributes to
the logic and structure of part 53 by distinguishing between SR SSCs
and NSRSS SSCs and licensee-controlled programs that address LBEs other
than DBAs. Specifically, SR SSCs, human actions, and programmatic
controls needed to protect against DBAs are used to satisfy the safety
criteria in Sec. 53.210. NSRSS SSCs, human actions, and licensee-
controlled programs that address LBEs other than DBAs generally
contribute to the appropriate measures considering potential risks to
public health and safety.
Section 53.400 establishes a requirement that design features be
provided for each commercial nuclear plant to satisfy the safety
criteria and fulfill safety functions from subpart B during LBEs. Other
sections in subpart C, in turn, further address the necessary
capabilities and reliabilities for SSCs by establishing functional
design criteria, fulfilling design requirements, performing analyses of
LBEs, performing other supporting analyses, and categorizing SSCs based
on their roles in preventing or mitigating LBEs.
Section 53.410 requires that functional design criteria be defined
for safety-related design features relied upon to demonstrate that the
consequences from DBAs would be below the criteria in Sec. 53.210
through analyses performed under Sec. 53.450(f), which includes
insights from both PRAs and deterministic analyses. Other sections
within part 53 establish appropriate controls on these design features
(e.g., safety classification, protection from external hazards, quality
assurance, and TS) to ensure the functional design criteria are
satisfied. The performance requirements for the SSCs needed to address
DBAs and the consideration of human actions and programmatic controls
in the identification of special treatments associated with the design
of SR SSCs will contribute to ensuring that a commercial nuclear plant
licensed under part 53 would meet the safety criteria in Sec. 53.210.
Section 53.415 requires that SR SSCs be protected against or
designed to withstand the effects of natural phenomena (e.g.,
earthquakes, tornadoes, hurricanes, floods, tsunami, and seiches) and
constructed hazards (e.g., from dams, transportation routes, and
military or industrial facilities). Specifically, Sec. 53.415 requires
that SR SSCs remain capable of performing the safety functions stated
in Sec. 53.230 for which they are credited up to the design-basis
external hazard levels as determined under Sec. 53.510. As used in
Sec. 53.415 and subpart D of part 53, a hazard level refers to such
things as the magnitude and recurrence rate of an earthquake and the
resultant ground motions, the height of a flood, the force of hurricane
winds, or the concentrations of chemicals resulting from a release from
a nearby facility. These requirements will support either traditional
deterministic approaches for determining and protecting against
external hazards or probabilistic approaches that are being developed
for seismic and some other external hazards.
Section 53.420 requires that functional design criteria be defined
for design features that play a significant role in demonstrating that
the safety criteria for LBEs other than DBAs are satisfied. The
analyses required for this demonstration are described in Sec.
53.450(e), which requires that those events be identified and assessed
using a PRA, other SREs, or a combination thereof, together with other
generally accepted approaches for systematically evaluating engineered
systems. The SSCs determined to be safety significant (i.e., either SR
or NSRSS) will have associated special treatment requirements as
specified in Sec. 53.460. Special treatment is defined in subpart A of
part 53 and generally refers to measures (e.g., quality assurance,
testing, monitoring) taken beyond normal commercial practices related
to the procurement, installation, and maintenance of commercial grade
products to provide confidence that SR and NSRSS SSCs will perform
under the service conditions and with the reliability assumed in the
analysis under Sec. 53.450(e) and will comply with the applicable
functional design criteria. Such normal commercial practices include
the use of consensus codes and standards, as identified in an
application, to support the identification of special treatments that
include measures that may go beyond the use of commercial codes and
standards. The inclusion of a systematic approach to identifying the
functional design criteria for SSCs and tailoring the special
treatments to specific LBEs and safety functions is an important
contributor to satisfy the safety criteria in subpart B. Therefore,
designers and licensees for commercial nuclear plants are provided
flexibility on how LBEs other than DBAs are either prevented or
mitigated and how the calculated comprehensive plant risks satisfy the
safety criterion established under Sec. 53.220(b).
[[Page 15706]]
Section 53.425 establishes requirements for design features and
related functional design criteria limiting doses to members of the
public during normal operations to satisfy the criteria in part 20.
Section 53.430 provides similar requirements for design features and
related functional design criteria for protection of plant workers to
meet the safety criteria in part 20. Section 53.425 provides applicants
and licensees flexibility to define design objectives for design
features related to controlling liquid, gaseous, and solid wastes as
required under part 20. The design objective will assist designers,
applicants, and licensees in performing the evaluations of possible
reductions in public dose from routine effluents when considering costs
and other factors.
The requirements in Sec. Sec. 53.425 and 53.430 for design
features and functional design criteria to support radiation protection
activities have parallels in existing regulations such as Sec.
50.34(a) and (b)(3), which require in part that the means be provided
for meeting the requirements of part 20 and General Design Criterion
60, 61, 63, and 64 in appendix A to part 50, which provide radiation
protection related design criteria.
Section 53.440 addresses various design requirements that warrant
specific mention to ensure that the design features required by Sec.
53.400 comply with the functional design criteria required by
Sec. Sec. 53.410 and 53.420. These requirements will be met through
design practices, consideration of testing and operating experience,
and various assessments of LBEs and other potential challenges to
commercial nuclear plants. Discussion of some of the key design
requirements included in section 53.440 follow.
(1) Sec. 53.440(a): An essential element to ensuring a proposed
design can comply with the performance criteria in part 53 is that the
ability of design features to fulfill their safety functions is
demonstrated by a combination of analyses, test programs, prototype
testing, and operating experience. This requirement closely aligns with
the language in Sec. 50.43(e) and is included in part 53 as the same
foundational requirement. In addition, Sec. 53.440(a) requires the
design processes for SSCs under this section to include administrative
procedures for evaluating operating, design, and construction
experience for considering applicable important industry experiences in
the design of those SSCs. This requirement corresponds to the existing
requirement under Sec. 50.34(f)(3)(i) that was developed in response
to the 1979 accident at Three Mile Island Nuclear Generating Station.
(2) Sec. 53.440(b): The design and licensing of commercial nuclear
plants should use generally accepted consensus codes and standards for
design features classified as safety-related. Such codes and standards
ensure sufficient testing and qualification of materials and equipment
and provide defined processes, specifications, and acceptance criteria
for use by designers and suppliers. The NRC will indicate acceptance of
consensus codes and standards used in the design and licensing of a
specific commercial nuclear plant either through the NRC's generic
endorsement of a code or standard (i.e., through regulatory guidance),
including any limitations or conditions, that can be referenced within
an application, or through the review of a referenced code or standard
as part of the review of a specific application.
(3) Sec. 53.440(c): The design requirements in subpart C require
the materials used for SR and NSRSS SSCs to be qualified for their
service conditions over the design life of the SSC as appropriate to
satisfy the special treatments established for the SSC under Sec.
53.460.
(4) Sec. 53.440(d): The requirements in Sec. 53.440 include the
need to consider possible degradation mechanisms for materials and
equipment to inform both the design process and the development of
integrity assessment programs to be executed during plant operations in
accordance with subpart F of part 53. The inclusion of requirements
related to designing and monitoring for possible degradation mechanisms
reflects important lessons learned from the history of LWRs as well as
operating experience with structures and systems in countless other
engineering endeavors.
(5) Sec. 53.440(e) and (f): The design requirements in subpart C
state specific design requirements similar to existing requirements in
parts 50, 52, and 73 for protections against fires and explosions and
consideration of safety and security together in the design process.
Under Sec. 53.440(f), 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.
This approach ensures considerations are given for safety and security
together throughout the plant's lifetime, including the design process
and prior to implementing changes to plant configurations, to ensure
risks are effectively managed. The implementation of a security
strategy and design features early in the design process has the
potential to be more efficient and cost-effective rather than
implementing these features after the plant has been designed and
constructed.
(6) Sec. 53.440(g) and (h): Specific design requirements will
ensure that commercial nuclear reactors under part 53 have the
capability to achieve and maintain subcriticality and long-term
cooling. The requirements are included to address the potential that
some reactor designs may be able to achieve a stable end state for the
purpose of event analyses but might need further actions to completely
shut down and service the facility.
(7) Sec. 53.440(i): The design, analysis, and development of
programmatic controls under part 53 will consider the number of reactor
units and other significant inventories of radioactive materials
contributing to the risks to public health and safety. This reflects
the definition of ``Commercial nuclear plant'' in subpart A and
reinforces that the evaluation of LBEs is performed on a plant-wide
basis. This aspect of part 53 is different from parts 50 and 52, which
generally define safety requirements on the assumption of events
involving only individual reactor units.
(8) Sec. 53.440(k): The inclusion of a specific requirement for
design features and related functional design criteria, including
associated programmatic controls or a combination thereof, to address
the risks to public health from potential chemical hazards of licensed
material is appropriate given the diversity of reactor technologies and
designs that might be licensed under part 53. The requirement in part
53 is similar to the existing requirements in 10 CFR part 70,
``Domestic Licensing of Special Nuclear Material,'' that address both
potential radiological and chemical hazards for licensed materials at
fuel cycle facilities.
(9) Sec. 53.440(l): These provisions require that measures be
taken during the design of commercial nuclear plants to minimize
contamination of the facility and the environment, facilitate eventual
decommissioning, and minimize the generation of radioactive waste in
accordance with Sec. 20.1406.
(10) Sec. 53.440(m): This design requirement provides a
technology-inclusive equivalent to the requirements in Sec. 50.68 by
including options for commercial nuclear plants to either have a
monitoring system capable of detecting a criticality as described in
Sec. 70.24 or to have restrictions on SNM
[[Page 15707]]
handling and storage that would prevent inadvertent criticality events.
(11) Sec. 53.440(n): The design needs to reflect state-of-the-art
human factors principles for safe and reliable performance in all
settings that human activities are expected for performing or
supporting the continued availability of plant safety or emergency
response functions.
One notable exclusion from the design requirements in the part 53
proposed rule is an explicit requirement to consider and address the
potential impact of a large, commercial aircraft, as is currently
required of parts 50 and 52 applicants under Sec. 50.150, ``Aircraft
impact assessment.'' When the Commission promulgated the aircraft
impact final rule on June 12, 2009 (74 FR 28112), it noted that ``the
impact of a large aircraft on the nuclear power plant is regarded as a
beyond-design-basis event'' and it was ``the NRC's view that effective
mitigation of the effects of events causing large fires and explosions
(including the impact of a large, commercial aircraft) can be provided
through operational actions,'' which were covered by other
requirements. In light of this view, the Commission stated that ``the
mitigation of the effects of aircraft impacts through design should be
regarded as a safety enhancement which is not necessary for adequate
protection.'' In the Regulatory Analysis that accompanied the aircraft
impact rule, the NRC quantified the costs of the rule, but did not
quantify the benefits of the rule, stating that the ``benefits of the
final rule can be evaluated only on a qualitative basis.'' The NRC
concluded that the key benefit of the rule was ``improvement in
knowledge.'' The Commission acknowledged that ``it is difficult to
quantify the safety enhancement gained through implementation of the
aircraft impact rule,'' but stated that ``the NRC nevertheless believes
that the cost of performing the assessment and incorporating the
results into the design . . . is justified in view of the increased
safety provided by implementation of the aircraft impact rule.''
It has been over 15 years since the promulgation of the aircraft
impact rule in 2009. Events like the terrorist attacks of September 11,
2001, are now much less likely due to significant increases in security
at commercial aviation facilities as well as hardened access to
aircraft cockpits. In addition, it is not clear that the Commission's
previous belief that the cost of implementation of the aircraft impact
rule was justified by the increase in safety provided by the rule would
hold true for future reactors licensed under part 53. As stated
previously, the NRC concluded that the key benefit of the rule was
``improvement in knowledge'' achieved by performing the aircraft impact
assessment. It's worth noting that licenses issued under parts 50 and
52 were largely based on deterministic analyses of the safety of the
facility relying on the General Design Criteria. The technical
requirements in part 50 were supplemented over the years to address
specific beyond-design-basis events, such as the loss of large areas of
the plant due to fires and explosions. In contrast, under part 53,
applicants will be required to perform a comprehensive assessment of
their reactor design to identify potential failures, susceptibility to
internal and external hazards, and other contributing factors that
could pose a risk to public health and safety. The spectrum of events
and hazards considered will include those that have traditionally been
considered design-basis events and those that have been considered
beyond-design-basis events. Although part 53 does not include
prescriptive requirements to assess a licensing-basis event comprising
an intentional act that could cause large fires or explosions, it does
require applicants to assess a full spectrum of unplanned events, to
include anticipated events, unlikely events, and very unlikely events.
The NRC believes that the systematic evaluations of internal hazards,
external hazards, and security threats under part 53 and part 73
sufficiently address the potential loss of large areas of the plant due
to explosions or fire currently addressed under Sec. 50.155(b)(2).
Therefore, part 53 applicants will have considered how to mitigate
the broader potential plant impacts that may result from an event such
as the impact of a large aircraft. As a result, applicants and
licensees under part 53 will have substantially more information about
the design of their facilities than applicants and licensees did before
the promulgation of the aircraft impact rule. Accordingly, the
``improvement in knowledge'' to be gained by requiring a separate
assessment of the impact of a large commercial aircraft under part 53
is expected to be significantly less than the improvements in knowledge
for part 50 or 52 applicants the Commission estimated when it
promulgated the aircraft impact rule. Because the potential impact of
beyond-design-basis events are considered in other ways under part 53,
the NRC concludes that the cost of performing a separate aircraft
impact assessment and incorporating the results into the design of a
commercial nuclear plant licensed under part 53 would not be justified.
For these reasons, this final rule does not contain requirements for
applicants to assess the impact of a large, commercial aircraft on the
design of the facility.
Section 53.450 establishes analysis requirements and centers upon
the use of a PRA, other SREs, or a combination thereof with other
generally accepted approaches for systematically evaluating engineered
systems. The use of PRA, other SREs, or a combination thereof as a key
component in the analysis requirements for part 53 reflects the decades
of improvements in the use of such methodologies and their increasing
use in the design, licensing, and oversight of both operating and
future nuclear reactors. Part of the Commission's PRA Policy Statement
is that the use of PRA technology should be increased in all regulatory
matters to the extent supported by the state-of-the-art in PRA methods
and data and in a manner that complements the NRC's deterministic
approach and supports the NRC's traditional defense-in-depth
philosophy. This policy statement also acknowledges the variability in
the characteristics of events considered and the associated complexity
of engineered systems related to different regulatory activities and
that risk-informed analysis techniques of varying complexity may be
employed to yield meaningful insights and results. In that regard, the
use of PRA, other SREs, or a combination thereof under part 53 needs to
be commensurate with the complexity of the analyzed systems and their
behaviors, with consideration of all aspects of operations. The need to
supplement PRA insights with other engineering approaches and judgments
reflects the NRC's longstanding policy described in the SRM to SECY-98-
144, ``Staff Requirements--SECY-98-144--White Paper on Risk-Informed
and Performance-Based Regulations,'' dated February 24, 1999, for
regulatory decision-making to be risk-informed but not solely based on
numerical results of a risk assessment (i.e., not a risk-based
approach). Part 53 maintains a role for NRC's traditional deterministic
approaches (particularly for DBAs) and defense-in-depth philosophy by
including specific requirements utilizing these regulatory tools in
subparts B and C.
PRA, other SREs, or a combination thereof will be used together
with other techniques in part 53 to identify and categorize LBEs,
classify SSCs, evaluate defense in depth, and inform the appropriate
special treatments for SSCs. This increased role for PRA and SREs
[[Page 15708]]
necessitates that they be developed, performed, and maintained in
accordance with NRC approved standards and practices (see Sec.
53.450(c) and (d)). The computer codes used to model the plant response
and the behavior of the barriers to the release of radionuclides must
be qualified for the range of conditions being simulated across a wide
range of unplanned events. These analyses must use realistic approaches
and address uncertainties associated with states of knowledge,
modeling, and performance of SSCs.
While industry consensus PRA standards and PRA peer review
processes endorsed in RGs 1.200 and 1.247 remain acceptable for
developing a PRA, they are not regulatory requirements and an
application under part 53 need not follow every aspect of the
applicable consensus PRA standard. Existing processes for defining the
scope and capability of a PRA supporting an application offer
flexibility in determining the degree to which the PRA needs to be
developed and may be informed by other factors such as design
complexity and the needed degree of realism and level of detail,
consistent with the use of the PRA with SREs and the substance of the
application. Such processes are currently available for appropriately
defining the scope of the PRA and determining applicability of
supporting requirements in consensus PRA standards needed to satisfy
the regulatory requirements for the specific uses of analyses under
Sec. 53.450(b). The specific uses of analyses in Sec. 53.450(b) are
to inform LBE selection; inform classification of SSCs according to
safety significance; evaluate adequacy of defense in depth; identify
and assess all plant operating states with a potential for uncontrolled
release of radioactivity to the environment; identify and assess events
that challenge plant control and safety systems whose failure could
lead to the uncontrolled release of radioactive material to the
environment; and inform the establishment and updating of appropriate
measures for plant operations, including availability controls, to
ensure configurations and special treatments for SR SSCs and NSRSS SSCs
provide the capabilities, availability, and reliability consistent with
satisfying the high-level safety criteria in Sec. 53.220.
Likewise, NRC determinations of the acceptability of such PRAs
would include consideration of the appropriateness of the applicant-
defined scope as part of determining the applicability of and
conformance to consensus PRA standard supporting requirements
consistent with the current state of practice. In addition, these
determinations would include consideration of other aspects of the
development of the PRA, such as PRA peer reviews. An NRC determination
of the acceptability of a PRA includes but is not limited to assessing
the initial and boundary conditions and key assumptions used in the
analysis, treatment of uncertainties, and the use of screening tools
and bounding or simplified methods for any mode or hazard, provided the
use of those tools and methods is justified by an acceptable technical
basis. In that regard, the consensus PRA standards would not be applied
by the NRC as a strict checklist of requirements for part 53 PRA
acceptability determinations.
For risk contributors that are excluded from PRA logic models or
PRA screening processes and are otherwise analyzed by an SRE--also
referred to as supplementary analyses--the NRC plans to develop
guidance for determining the acceptability of such SREs.
Section 53.450(c) requires periodic maintenance and upgrading of
the PRA, other SREs, or a combination thereof to maintain an alignment
between the supporting analyses and the design and performance of plant
equipment, programs and procedures, and other factors associated with
meeting the safety criteria of Sec. 53.220 and the evaluation criteria
of Sec. 53.450(e)(2). The periodic maintenance of the PRA, other SREs,
or a combination thereof is also a means to consider new or revised
information related to external hazards, industry operating experience,
performance issues with or degradation of SSCs, and other contributors
to the frequency and potential consequences of various event sequences.
The periodic assessments performed by licensees to support the
maintenance of the PRA, other SREs, or a combination thereof and other
requirements in part 53 will be complemented by NRC inspections and
programs to assess new or revised information related to topics such as
natural hazards, operating experience, and potential generic safety
issues.
Section 53.450(d) provides requirements for the qualification of
the analytical codes used in modeling the physical behavior of plant
systems and that those codes must be qualified for the range of
conditions for which they are to be used.
The categories of LBEs used in part 53 include anticipated event
sequences, unlikely event sequences, and very unlikely event sequences.
The unlikely event sequences include those events with estimated
frequencies well below the frequency of events expected to occur during
the lifetime of a commercial nuclear plant. An important aspect of the
analysis requirements is that, under Sec. 53.450(e), the analyses of
LBEs other than DBAs will be used not only to show the performance
criteria of Sec. 53.220 are satisfied but also to show that evaluation
criteria defined for each LBE or category of LBEs are satisfied. Such
evaluation criteria for specific LBEs or categories of LBEs are defined
in terms of limits on the release of radionuclides or maintaining the
integrity of one or more barriers used to limit the release of
radionuclides and reflect a graded approach of allowing lesser
potential consequences from more frequent events. An example of such
evaluation criteria for a range of LBEs that could likely be expanded
for part 53 is provided in RG 1.233. An applicant's or licensee's
defining of evaluation criteria under Sec. 53.450(e) and the risk
performance objectives under Sec. 53.220(b) are also part of the
integrated approach within part 53 where the analyses from subpart C
are used for decisions on design, siting, and operations. As an
example, an applicant or licensee could propose to justify siting
proposals by defining their evaluation criteria such that the
calculated consequences for an individual at the exclusion area
boundary are less than the total effective dose equivalent (TEDE)
values used in graded approaches to assessing population densities
under subpart D. Another requirement for the Sec. 53.450(e) analyses
is that the methodology must include a means to identify event
sequences deemed risk-significant such that those event sequences can
be given special attention within other sections of part 53.
Part 53 maintains an important role for a deterministic analysis of
DBAs in the performance criteria of Sec. 53.210 and the related
analytical requirements in Sec. 53.450(f). The analysis of DBAs will
be required to address event sequences drawn from those with estimated
frequencies below the expected lifetime of a generation of reactors
(e.g., event sequences with frequencies as low as one in ten thousand
years). As set forth in this section, DBAs must be analyzed using
deterministic methods and ensure a safe, stable end state with reliance
upon only SR SSCs and human actions, if needed, to be performed by
operators licensed under the provisions of Sec. Sec. 53.760 through
53.795.
While the DBAs analyzed under part 53 are similar to the
traditional DBAs analyzed under parts 50 and 52, there are important
distinctions between the overall role of DBA analyses in part 50 and
part 53. In part 53, the role of the
[[Page 15709]]
DBA analysis is more narrowly focused on selecting SR SSCs and
determining functional design criteria for those SSCs to ensure the
commercial nuclear plant meets the safety criteria in Sec. 53.210. The
overall control of risks posed by commercial nuclear plants under part
53 is provided by the analyses of and measures taken for both DBAs and
other LBEs, including very unlikely event sequences. This contrasts
with the traditional deterministic approach in part 50 wherein the
analyses of DBEs such as DBAs were used to provide bounding
assessments, to incorporate standard design rules such as assumptions
related to single failures, and to define conservative performance
requirements for SR SSCs. Limitations related to the traditional
deterministic approach were addressed in part 50 through case-by-case
assessments and specific actions for beyond-design-basis events such as
anticipated transients without scram and station blackout.
Section 53.450(g) includes provisions to ensure that analyses are
performed to support the design requirements of Sec. 53.440(e) on fire
protection and Sec. 53.425 on using design features and plant programs
to control doses to members of the public from routine effluents and
direct radiation from contained sources. The analysis requirements
related to fire protection support either a traditional, deterministic
approach or a more risk-informed approach where the risks from fires
are addressed within the identification and analyses of LBEs.
Section 53.460 establishes criteria for the safety classification
of SSCs and determination of appropriate special treatments. As noted
in subpart A, the term ``Special treatments'' is defined to mean those
items, such as measures taken to satisfy functional design criteria,
quality assurance, and programmatic controls, that provide assurance
that certain SSCs will provide defense in depth or perform risk-
significant functions. These requirements also provide confidence that
the SSCs will perform under the service conditions and with the
reliability credited in the analysis performed in accordance with Sec.
53.450 to satisfy the safety criteria in Sec. Sec. 53.210 and 53.220.
The terminology used in part 53 includes the following categories for
SSC classification: (1) SR; (2) NSRSS; and (3) non-safety significant.
Requirements for SR SSCs are defined in other sections of part 53 and
include using TSs for controls during operation and the application of
quality assurance requirements from appendix B to part 50.
Requirements for NSRSS SSCs include the need to identify necessary
special treatments such as performance measures on reliability.
Licensees will generally be afforded flexibility in maintaining and
changing special treatments for SSCs categorized as NSRSS. Non-safety-
significant SSCs will be addressed under normal licensee programs for
commercial grade equipment and typical industry practices for general
plant design and maintenance. Safety-related SSCs also contribute to
defense in depth and risk-significant functions and may warrant special
treatments beyond those defined for their SR functions to reflect their
role in meeting the safety criteria in Sec. 53.220 and the evaluation
criteria in Sec. 53.450(e).
Section 53.480 establishes seismic design considerations. This
section relates to the safety criteria in subpart B, the analytical
requirements related to external hazards in Sec. 53.450, and subpart
D, ``Siting Requirements.'' For licenses issued under part 53, this
section in subpart C will support a variety of approaches to seismic
design. For example, a design for a commercial nuclear plant could show
that SSCs are able to withstand the effects of earthquakes by adopting
an approach similar to that in appendix S to part 50. Alternatively, an
applicant could follow the more recent risk-informed alternatives
afforded by standards development organizations (e.g., American Society
of Civil Engineers (ASCE)/Structural Engineering Institute (SEI) 43-19,
``Seismic Design Criteria for Structures, Systems, and Components in
Nuclear Facilities''). Because the agency has not endorsed ASCE/SEI-43-
19, an applicant can propose to use ASCE/SEI 43-19 on an application-
specific basis to meet Sec. 53.480 and the NRC would evaluate the
adequacy of the standard as applied in that application. The design
could also be done with the full integration of seismic PRAs into the
design and licensing of a particular commercial nuclear plant. This
section has been developed to accommodate a variety of potential risk-
informed, performance-based seismic design approaches. The analyses
required by Sec. 53.450 must address seismic hazards as well as other
external hazards. The expected responses of SSCs to a range of seismic
events must be included in the analyses when ensuring that the safety
criteria defined under Sec. 53.220 will be met. The potential SSC
responses to seismic hazards could be addressed in the analyses using a
fragility model (conditional probability of its failure at a given
hazard input level), a high confidence of low probability of failure
value, or other method endorsed or otherwise found acceptable by the
NRC.
Subpart D--Siting Requirements
Subpart D in part 53 states requirements for the siting of
commercial nuclear plants and serves the role provided by 10 CFR part
100, ``Reactor Site Criteria,'' for nuclear reactors licensed under
parts 50 and 52. As reflected in Sec. 53.500, the reason for
establishing siting requirements remains 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 commercial nuclear plant
(e.g., external hazards) and, conversely, what potential adverse health
and safety impacts a commercial nuclear plant may have on nearby
populations in view of the site characteristics.
Section 53.510 requires that design-basis external hazard levels be
identified and characterized based on site-specific assessments of
natural and constructed hazards with the potential to adversely affect
plant functions. The site-specific assessments are used in Sec.
53.415, which requires that SR SSCs be designed to withstand the
effects of natural phenomena and constructed hazards of levels or
severities up to design-basis external hazard levels. The design-basis
levels for external hazards relevant to a site need to account for
uncertainties and variabilities in data, models, and methods used to
characterize those hazards. Existing approaches can be used to
demonstrate compliance with this requirement. The historical importance
of assessing seismic events as risks to commercial nuclear plants and
the associated development of risk-informed approaches to address
seismic events are reflected in Sec. 53.480, ``Earthquake
engineering,'' and specific requirements in subpart C. The NRC is
developing a graded approach for seismic design by grouping SSCs into
different seismic design categories (SDCs) based on their risk
significance. While the agency has not endorsed ASCE/SEI-43-19, an
applicant can propose to use ASCE/SEI 43-19 on an application-specific
basis to meet Sec. 53.480 and the NRC will evaluate the adequacy of
the standard as applied in that application. The NRC staff will
continue to review ASCE/SEI-43-19 as part of its efforts to further
develop guidance in this area. The approach described in RG 1.208, ``A
Performance-Based Approach to Define the Site-Specific Earthquake
Ground Motion,'' is an acceptable way to develop site-specific ground
motion
[[Page 15710]]
response spectra for SSCs under appendix S to part 50, which
corresponds to SSCs that are categorized as the highest SDC (SDC-5) in
ASCE/SEI 43-19.
The evaluation of seismic hazards under subpart D needs to be
sufficient to inform a site-specific design (e.g., a CP or custom
combined license (COL)) or confirm the use of a standard design for a
commercial nuclear plant under Sec. 53.480 and other sections of
subpart C. A risk-informed approach can use several design-basis ground
motions (DBGMs) to assess SSCs in various SDCs (i.e., one DBGM per
SDC). Section 53.510(d) states that geologic and seismic siting factors
must also include related hazards such as seismically induced flooding
and volcanic activity that may affect the design and operation of a
proposed commercial nuclear plant for the proposed site.
Section 53.520 requires applicants to identify and assess site
characteristics related to topics that include meteorology, geology,
hydrology, or other areas in the design and analyses required under
subpart C.
Section 53.530 sets requirements for population-related
considerations and largely maintains requirements and definitions
similar to those currently in part 100 for an exclusion area, low
population zone, and population center distance. The NRC recognizes
that some applicants may propose to essentially collapse the exclusion
area and low population zone to the site boundary. This approach would
rest on a demonstration that the calculated consequences of DBAs remain
below the dose guidelines used in Sec. 53.210, which are the same as
those in the existing regulations in parts 50, 52, and 100. The
definitions in Sec. 53.020 allow such configurations, assuming they
were justified by the design and analyses from subpart C. This approach
should provide flexibility to justify alternative exclusion areas and
low population zones without foreclosing the option for an applicant to
define more conventional exclusion areas and low population zones
outside of a defined site boundary. The NRC's long-standing preference
for siting reactors in areas of low population density is maintained in
part 53 by using the current language from part 100 as one option under
Sec. 53.530(c). The NRC revised guidance related to population
densities surrounding a commercial nuclear plant in Revision 4 to RG
4.7, ``General Site Suitability Criteria for Nuclear Power Stations''
to reflect Commission direction in SRM-SECY-20-0045, ``Population
Related Siting Considerations for Advanced Reactors.'' The NRC
recognizes that safety, environmental, economic, or other factors may
justify siting commercial nuclear plants in areas with higher
population densities or within a densely populated center containing
more than about 25,000 residents. Therefore, an option is included
within Sec. 53.530 for such sites to be proposed using assessments of
additional societal risks associated with siting a reactor in areas of
higher population density (e.g., potential increases in population dose
or economic consequences from reactor accidents) in comparison to the
societal benefits of a specific site (e.g., ability to use existing
infrastructure for a retired fossil fuel power plant). Site-related
requirements in part 20 (restricted area) and part 73 (protected and
owner-controlled areas) remain applicable to commercial nuclear plants
licensed under part 53.
Section 53.540 requires that site characteristics be appropriately
considered in other activities such as the design and analysis
performed under subpart D of part 53 and the emergency planning and
security programs under subpart F of part 53.
Subpart E--Construction and Manufacturing Requirements
The part 53 language establishes construction and manufacturing
requirements in subpart E. The language for construction-related
activities largely reflects current requirements in part 50 without any
fundamental changes. Limited changes were made in several places, as
described in the following paragraphs, to be technology-neutral and for
consistency with the organization and language of part 53. The language
for requirements for manufacturing activities largely mirrors those for
construction-related activities. However, the manufacturing
requirements have been updated from the current requirements in subpart
F of part 52 to better accommodate the possible factory fabrication of
manufactured reactors. The manufacturing of specific components outside
the scope of an ML is not addressed by these subparts.
Section 53.600 establishes the overall construction and
manufacturing requirements for CPs, OLs, COLs, MLs, and limited work
authorizations (LWAs). This section connects the construction and
manufacturing requirements to the safety criteria, quality assurance
requirements, and other requirements located in other subparts. These
requirements require that construction and manufacturing activities be
managed and conducted such that when combined with associated design
features and programmatic controls, the constructed plant will satisfy
the relevant requirements in subpart B.
Section 53.605 establishes requirements for the reporting of
defects and instances of noncompliance during construction. This
section provides equivalent requirements to those in Sec. 50.55(e).
Section 53.610(a) establishes the requirement to have in place a
well-defined command and control structure to manage construction
activities. The requirements generally reflect current requirements,
with an emphasis on the quality assurance programs for complying with
the requirements in appendix B to part 50. Section 53.610(a)(6)
requires programmatic controls for implementing special treatment for
NSRSS SSCs to align with requirements in other subparts in part 53. The
section also refers to other NRC regulations to address matters such as
requirements to have an FFD program, a radiation protection program if
radioactive materials are brought onto the site, and security programs
to protect sensitive information and protect against cyber threats.
Section 53.610(b) provides requirements governing construction
activities, including the equivalent of the requirement in Sec.
50.10(e) that prohibits starting construction until the NRC has
authorized the activities by issuing a CP, COL, ESP, or LWA. Section
53.610(b)(1)(iii) requires procedures to be in place prior to beginning
construction to ensure that construction-related activities do not
undermine important features such as slope stability and that
construction-related activities such as backfilling of excavated
portions of the site appropriately address potential pre-construction
activities such as the emplacement of retaining walls or drainage
systems. Other requirements in these paragraphs are equivalent to
requirements in parts 50 and 52 with appropriate references to other
parts for items such as possession of byproduct material or SNM,
protecting operating units from construction activities for commercial
nuclear plants with multiple reactor units, and having a redress plan
in case LWA activities are terminated.
Section 53.610(c) addresses inspection and acceptance activities by
including requirements in part 53 equivalent to specific quality
assurance criteria in appendix B to part 50 and inspections, tests,
analyses, and acceptance criteria (ITAAC) in part 52 for COLs.
[[Page 15711]]
Section 53.620(a) includes requirements covering the activities
performed under an ML issued under part 53. 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.\1\ 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 DC 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.
---------------------------------------------------------------------------
\1\ On December 17, 1982, the NRC issued ``Manufacturing License
ML-1 to Offshore Power Systems for the manufacture of a maximum of
eight floating nuclear plants,'' dated September 30, 1982, but the
project was subsequently canceled.
---------------------------------------------------------------------------
The part 53 sections in subpart E for manufacturing and in subpart
H for licensing matters maintain requirements largely equivalent to
those in part 52 for MLs. The NRC approval of a standard design and
related manufacturing processes, coupled with a stable workforce and
established procedures, has the potential for maintaining and even
improving the quality and consistency of manufacturing, as compared to
the traditional method of constructing reactors onsite by a variety of
contractors and subcontractors.
Subpart E includes requirements that apply to portions of a
manufactured reactor in recognition that some activities covered by an
ML may occur at different fabrication facilities. As with the preceding
sections on construction, Sec. 53.620 establishes the requirements to
have in place programs, procedures, and a well-defined command and
control structure to manage manufacturing-related activities.
Section 53.620(b) in subpart E includes requirements for executing
the manufacturing activities following receipt of an ML under part 53.
Information about the design and manufacturing processes should be
provided by the applicant. The importance of the ML is reflected in
several of the requirements in Sec. 53.620(b) that refer to complying
with the ML, including conducting manufacturing processes within
facilities for which the license holder can control activities. The
essential role of post-manufacturing inspections is also incorporated
into this section by requiring the holder of the ML to perform
inspections and have acceptance processes for manufactured reactors or
portions of a manufactured reactor.
Section 53.620(c) provides requirements for the control of
radioactive materials if the holder of an ML plans to possess and use
source, byproduct, or SNM as part of the manufacturing process. By and
large, subpart E refers to NRC regulations in 10 CFR part 30, ``Rules
of General Applicability to Domestic Licensing of Byproduct Material,''
10 CFR part 40, ``Domestic Licensing of Source Material,'' and part 70
for the requirements on controlling radioactive materials. Several
specific requirements to address the potential hazards of radioactive
materials are included in areas such as having a fire protection
program, an emergency plan, training programs, and procedures to
minimize contamination.
The most significant change for MLs in part 53 as compared to MLs
under part 52 relates to Sec. 53.620(d) in subpart E and the
associated licensing provisions in subpart H. These provisions allow
and establish requirements for the loading of fuel into a manufactured
reactor at the manufacturing site for subsequent transport to a
commercial nuclear facility that will operate pursuant to a COL or OL.
The first requirement in Sec. 53.620(d) establishes limitations on
when a license under part 70 would authorize the loading of fuel into a
reactor manufactured under an ML. The regulation requires the
manufactured reactor to be configured during its loading, storage, and
transport with features to prevent criticality and that those features
be specified in the ML. The requirement provides 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, storage, 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, such
as during transport. This requirement contributes to the NRC's
longstanding practice of requiring defense in depth for preventing
accidents in any facility dealing with SNM, including requirements in
Sec. 70.64 for certain part 70 licensees to adhere to the ``double
contingency principle.''
The requirements to have in place features to prevent criticality
could likewise support meeting other provisions in subpart H to part
70, such as those related to having a safety program and integrated
safety assessment. The features to prevent criticality in the part 53
requirements will 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 part 53 and all
criticality safety controls required by 10 CFR 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 10 CFR part 70.
Collectively, these measures will reasonably ensure that the
manufactured reactor is not capable of operations, thereby obviating
the need for a COL under Sec. Sec. 53.1416 and 53.1440 to authorize
fuel loading. Additionally, this approach focuses 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, will be regulated primarily under the
part 70 license. The reference to the requirements in subpart H of part
70 in Sec. 53.620(d) assures that the activities involving the
receipt, storage, and loading of a variety of possible fuel forms and
enrichments at the manufacturing facility will be analyzed in a
systematic manner and appropriate protection will be provided against
equipment malfunctions, human errors, external hazards, and other
adverse conditions. The regulations in 10 CFR part 51, ``Environmental
Protection Regulations for Domestic Licensing and Related Regulatory
Functions,'' provide a flexible approach for environmental review to
address the range of regulated activities under part 70. The
flexibility in part 51 will 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.
Section 53.620(d) cites the requirements in parts 70, 71, and 73 to
ensure important features and programs
[[Page 15712]]
are in place prior to the receipt of SNM. The features and programs
required to be in place prior to receipt of SNM 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. Section 53.620(d)(2)(i) includes 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 part 53 regulations will 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 requires 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 is consistent with other controls for
MLs, including reactivity and criticality controls.
The requirements also require a holder of an ML and 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.
The regulations in part 53 covering the activities related to the
storage, movement, and loading of fresh fuel into a manufactured
reactor in the manufacturing facility likewise refer to the applicable
regulations in part 70. Section 53.620(d) also requires the loading or
unloading of unirradiated 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 meeting the requirements in
subpart F. The NRC is aware of proposals to introduce reprocessing of
existing or future spent nuclear fuel into the fuel cycle for some
potential commercial nuclear plants. This final rule does not address
the loading of spent nuclear fuel or fuel resulting from reprocessing
of spent nuclear fuel into a manufactured reactor.
Section 53.620(e) only allows 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 Commission has issued a
COL or CP authorizing the construction of a commercial 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 requirement is similar to the
limitations in Sec. 52.153. An additional paragraph in Sec. 53.620(e)
provides requirements for protecting fueled manufactured reactors
during transport to the site of the commercial nuclear plant by
referencing the transportation and security requirements in 10 CFR part
71, ``Packaging and Transportation of Radioactive Material,'' and part
73. As noted previously, Sec. 53.620(e) includes an additional
provision that allows a manufactured reactor or portions of a
manufactured reactor to be removed from the place of manufacture for
export in accordance with part 110, which represents another difference
from the similar provision in Sec. 52.153.
Section 53.620(f) includes requirements for the acceptance and
installation of a manufactured reactor at the site of a commercial
nuclear plant. The requirements reference the construction requirements
in Sec. 53.610 to govern the integration of the manufactured reactor
into the construction of a commercial nuclear plant. Other requirements
in the section address required receipt inspections and verification
that interface requirements between the manufactured reactor and the
balance of the commercial nuclear plant have been met.
Subpart F--Requirements for Operation
Subpart F provides the requirements for the operations phase of a
commercial nuclear plant to ensure that the safety criteria in subpart
B are satisfied throughout the plant's lifetime and during all modes of
normal operation and unplanned events. Section 53.700 provides the
general organization and overall objectives of subpart F, which are to
establish requirements during operations for (1) plant SSCs; (2)
personnel; and (3) plant programs.
Section 53.710 provides the requirements for maintaining
capabilities, availability, and reliability of SSCs to demonstrate
compliance with the safety criteria and design requirements for
unplanned events that are described in subparts B and C. The basic
structure of this section is that measures for SR SSCs are provided by
TS and measures for NSRSS SSCs are required to be addressed with
licensee-controlled documents and procedures.
The general content and control of TS under part 53 are similar to
the requirements in part 50. The requirements for TS include limits on
the inventories of radioactive materials, plant operating limits, and
specific requirements for each SR SSC, including limiting conditions
for operation (LCO) and required surveillances. The requirements for TS
also include a section on important design elements, which is similar
to design features in Sec. 50.36, and a section for administrative
controls. A provision addressing the development and submittal of TS to
address decommissioning activities is also included in subpart G.
The requirements for TS under part 53 do not carry over safety
limits or associated limiting safety system settings from Sec. 50.36,
which contains TS requirements for operating reactors under parts 50
and 52. As discussed in SECY-18-0096, systematic assessments and more
mechanistic approaches to evaluating source terms support an
alternative approach to establishing barrier-based safety limits. An
example provided in that paper is a comparison of: (1) the traditional
specified acceptable fuel design limits (SAFDL) that support protecting
a specific barrier from potential failure mechanisms (e.g., departure
from nucleate boiling to protect fuel cladding); and (2) the specified
acceptable system radionuclide release design limit (SARRDL) concept,
which limits the possible increase in circulating radionuclide
inventory during normal operations or an AOO as part of an integrated
or ``functional containment'' approach. Additional discussion of the
use of SARRDL in the design and licensing of advanced reactors is
provided in RG 1.232. The SARRDL could be addressed as an operating
limit within this construct of requirements for TS. In cases, such as
LWRs, where a SAFDL approach might be used as part of a mechanistic
approach to meeting the design and analysis requirements in subpart C,
the associated functional design criteria in Sec. 53.410 and TS under
Sec. 53.710(a) define similar requirements as those provided by the
safety limit and limiting safety system setting requirements in Sec.
50.36.
[[Page 15713]]
The requirements for TS under part 53 do not include specific
criteria for identifying when LCOs must be established (i.e., do not
include an equivalent to Sec. 50.36(c)(2)(ii)). Instead, consistent
with subparts B and C, the TS requirements in subpart F of part 53
define TS LCOs as providing limits on SR SSCs. The SR SSCs protect
against DBAs to demonstrate compliance with the safety criteria in
Sec. 53.210. In the construct for part 53, risk-significant SSCs are
addressed through a combination of TS for SR SSCs and establishment and
monitoring of performance standards for NSRSS SSCs.
In addition to addressing TS for SR SSCs, Sec. 53.710 requires
appropriate control measures be developed and implemented for NSRSS
SSCs. Examples include appropriate surveillances and controls
established through reliability assurance programs. Configuration
management and other special treatments provide that the capabilities,
availabilities, and reliabilities of NSRSS SSCs are maintained
consistent with the underlying risk assessments while providing
flexibility to licensees through maintaining the management functions
within licensee-controlled programs. Controls on NSRSS SSCs are
appropriate as part of the overall performance-based approach within
part 53. Special treatments beyond those defined for their SR functions
may also be warranted for SR SSCs to reflect their role in meeting the
safety criteria in Sec. 53.220 and the evaluation criteria in Sec.
53.450(e). The performance objectives for NSRSS SSCs reflect that the
comprehensive risk metrics and related risk performance objectives
established under Sec. 53.220 may involve assessing and averaging the
risks over a defined period (e.g., plant year) and do not constitute a
real-time requirement that must be continuously demonstrated by the
licensee. The controls under Sec. 53.710(b) justify changes in part 53
from the traditional or deterministic approaches in parts 50 and 52 in
areas such as replacing the single-failure criterion with a
probabilistic reliability criterion (see SRM-SECY-03-0047, ``Policy
Issues Related to Licensing Non-Light-Water Reactor Designs,'' dated
June 26, 2003). This approach can also support the incorporation of
risk insights and analytical margins to gain operational flexibilities
in areas such as siting and staffing requirements described in
subsequent sections of subpart F.
Section 53.715 provides the requirements for developing and
implementing a program to do the following: (1) control maintenance
activities; (2) take appropriate corrective action when performance
issues are identified; (3) conduct routine evaluations of
effectiveness; and (4) assess and manage risks resulting from
maintenance activities. These requirements are similar to those
included in Sec. 50.65 (maintenance rule), including the need to
assess and manage the increase in risk that may result from the
maintenance activities. While, for the maintenance rule, specific
criteria must be developed to capture both SR and non-SR but otherwise
important SSCs, Sec. 53.715 covers SR SSCs and NSRSS consistent with
other subparts in part 53.
Section 53.720 provides the requirements for responding to a
seismic event during the operating phase of the life cycle of a
commercial nuclear plant and is equivalent to the requirements in
paragraph IV(a)(3) of appendix S, ``Earthquake Engineering Criteria for
Nuclear Power Plants,'' to part 50.
Part 53 includes provisions to address staffing, training,
personnel qualifications, and human factors engineering (HFE) in a
manner that is risk-informed, technology-inclusive, performance-based,
and flexible in nature. During the development of part 53, the staff
prepared a draft white paper on ``Risk-Informed and Performance-Based
Human-System Considerations for Advanced Reactors,'' to support
interactions with stakeholders and the Advisory Committee on Reactor
Safeguards (ACRS). Key considerations include the recognition that
staffing, operator qualifications, and HFE are interconnected areas
that must be approached in an integrated manner and, furthermore, that
safety functions, including the means by which they are fulfilled,
provide an effective method for informing technology-inclusive
requirements.
The requirements associated with this approach are in Sec. Sec.
53.725 through 53.830. Section 53.725 discusses applicability and
defines specific terms. Some definitions draw from those in Sec. 55.4.
Several new definitions are introduced for use within the context of
subpart F. These new definitions are the following: ``Automation,''
``Auxiliary operator,'' ``Generally licensed reactor operator,''
``Interaction-dependent-mitigation facility,'' ``Load following,'' and
``Self-reliant-mitigation facility.''
Sections 53.725 through 53.830 are divided into four portions that
cover general operational requirements, operator and senior operator
licensing requirements, GLRO requirements, and general training
requirements for plant staff. The NRC intends to provide guidance
addressing the review of operator staffing plans; the review of
operator, senior operator, and GLRO examination programs; and the
implementation of scalable HFE reviews. Licensees will be required to
use GLROs upon demonstrating compliance with the criteria in Sec.
53.800.
Certain routine communications are necessary to facilitate the
operator licensing process. The NRC adapts the requirements of
Sec. Sec. 55.5 and 50.74 to Sec. 53.726 to accomplish this.
Specific information must be collected in order 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 adapts the
requirements of Sec. 55.8, to include any needed updates in OMB
approval information, to Sec. 53.120 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 adapts the requirements of
Sec. 55.9 to Sec. 53.728 to require the completeness and accuracy of
material information provided by individual applicants and license
holders.
Section 53.730 provides performance-based and technology-inclusive
requirements for assessing the role of personnel in facility safety,
applying human system considerations within facility design, and
incorporating operational approaches that are consistent with design-
specific safety considerations. Most of these requirements are adapted
from portions of Sec. Sec. 50.34(f) and 50.54 and 10 CFR part 55,
``Operators' Licenses,'' with considerable modification in order to
reflect the introduction of new technologies and possible changes in
the roles of personnel in preventing and mitigating events. The NRC
intends that these technical requirements will, together, serve as a
component of the required content of applications for OLs and COLs
under part 53. Additionally, the NRC intends that the specific
technical requirements associated with HFE, human-system interface
design, concept of operations, functional requirements analysis, and
function allocation will serve as a component of the required content
of applications for standard DCs, standard design approvals, MLs, and
CPs, as well.
Human factors engineering is essential to facilitate the role of
personnel in facility safety in a manner that is both effective and
reliable. The
[[Page 15714]]
NRC adapts Sec. 53.730(a) from the HFE design requirements of Sec.
50.34(f)(2)(iii). A key difference is that the requirement is now
focused on settings where personnel fulfill their safety or emergency
response roles wherever they may occur. The NRC additionally includes
within the scope of this requirement activities for assuring the
continued availability of plant equipment that is needed for safety,
and the NRC envisions that these activities may encompass relevant
maintenance, inspections, and testing as well. This requirement is
associated with the staff guidance for conducting scalable reviews of
HFE in DRO-ISG-2023-03, ``Development of Scalable HFE Review Plans''
that accompanies part 53.
Human-system interfaces provide vital information to operators
across a spectrum of operating conditions that can range from normal
operations through severe accident conditions. The specific types of
information that must be available to support operations staff during
such conditions 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 requires under Sec. 53.730(b) such human-system
interface design features for all facilities, irrespective of other
flexibilities under part 53. Therefore, the NRC adapts specific post-
Three Mile Island requirements of Sec. 50.34(f) in a technology-
inclusive manner as detailed in the following:
Paragraph (b)(1) is adapted from Sec. 50.34(f)(2)(iv).
Paragraph (b)(2) is adapted from Sec. 50.34(f)(2)(v).
Paragraph (b)(3) is adapted from Sec. 50.34(f)(2)(xi),
50.34(f)(2)(xii), and 50.34(f)(2)(xxi).
Paragraph (b)(4) is adapted from Sec. 50.34(f)(2)(xvii),
50.34(f)(2)(xviii), 50.34(f)(2)(xix), and 50.34(f)(2)(xxiv).
Paragraph (b)(5) is adapted from Sec. 50.34(f)(2)(xxvi).
Paragraph (b)(6) is adapted from Sec. 50.34(f)(2)(xxvii).
In addition to the requirements of Sec. 53.730(b)(1) through (6),
a further set of human-system interface design requirements applicable
only to those facilities that will be staffed by GLROs is provided
under Sec. 53.730(b)(7). This prescriptive set of design requirements
for those facilities that demonstrate compliance with the criteria of
Sec. 53.800 recognizes that the application of HFE under Sec.
53.730(a) is anticipated to be significantly streamlined at such
facilities in the absence of an expected operator role for the
fulfillment of safety functions. However, it should be noted that the
capability for an immediately initiated, manual reactor shutdown is
conservatively mandated irrespective of any other design considerations
for both interaction-dependent and self-reliant mitigation facilities,
as required under Sec. 53.730(b)(8).
The NRC requires under Sec. 53.730(c) the submittal of a concept
of operations that is of sufficient scope and detail to appropriately
inform the staff. The development of a concept of operations can
facilitate a clear understanding on the part of the NRC for potential
novel operating concepts. Additionally, such information is likely to
reduce the degree of resources and interactions needed for the NRC to
obtain the understanding necessary to enable flexible requirements in
areas such as staffing, operator qualifications, and HFE.
The NRC requires under Sec. 53.730(d) the submittal of both a
Functional Requirements Analysis and a Function Allocation. The
identification of design-specific safety functions and how they are
fulfilled serves as a primary means for achieving technology-inclusive
requirements within areas such as staffing, operator qualifications,
and HFE. The Functional Requirements Analysis and Function Allocation
processes (which are both HFE methods derived from systems engineering
principles), provide an effective means to identify both how safety
functions will be satisfied and how to characterize any associated
operator role in doing so. A Functional Requirements Analysis shows
what features, systems, and human actions are relied upon to
demonstrate safety (i.e., fulfill safety functions). A Function
Allocation then describes how safety functions are assigned to both
personnel and automatic systems. However, an important adaptation of
the Function Allocation for use under this final rule is the further
need not only to describe allocations of safety functions to human
action and automation, but also to identify allocations made to active
safety features, passive safety features, or inherent safety
characteristics as well.
Operating experience provides an important source of information by
which to inform various aspects of facility design and operations.
Accordingly, the NRC adapts in Sec. 53.730(e) the requirements of
Sec. 50.34(f)(3)(i) for requiring an operating experience program.
New technologies may involve concepts of operations that are more
conducive to customizable licensed operator staffing requirements than
the prescriptive requirements of Sec. 50.54(m). Analyses and
assessments that are based on HFE principles provide a performance-
based means of determining licensed operator and senior operator
staffing needed to support safe operations. In contrast, for those
facilities required to be staffed by GLROs, the NRC anticipates that
the operator staffing plans will reflect a simpler approach of showing
that a continuity of responsibility will be maintained for facility
operations throughout the operating phase, with at least one GLRO
providing continuous oversight and remaining immediately available when
any units are fueled. Additionally, a revised approach to the
traditional position of the shift technical advisor that focuses on the
availability of engineering expertise as a means of addressing
uncertainties and abnormal circumstances is more suitable within the
context of part 53 and is intended to be applicable to all facilities,
irrespective of other design and staffing considerations.
Consistent with this approach, the NRC requires under Sec.
53.730(f) the submittal of a staffing plan that details operations
staffing, how engineering expertise will be provided, and what staffing
will be available to provide other needed support functions. The
staffing plan description of how engineering expertise will be provided
should include details of the position, such as location, expected
response time, access to plant status information, and methods of
communication. The staffing plan description should contain information
on how the described response time has been or will be determined to be
adequate based on the facility design. This requirement is associated
with the staff guidance for reviewing operations staffing plans in DRO-
ISG-2023-02, ``Interim Staff Guidance Augmenting NUREG-1791, `Guidance
for Assessing Exemption Requests from the Nuclear Power Plant Licensed
Operator Staffing Requirements Specified in 10 CFR 50.54(m),' for
Licensing Commercial Nuclear Plants under 10 CFR part 53'' that
accompanies part 53. Following NRC approval of the OL or COL, the
staffing plan will become a condition of the facility license.
The NRC intends that, at a minimum, the approved licensed operator
and senior operator (or, if applicable, GLRO) staffing, positions, and
personnel locations will be incorporated into corresponding
requirements within the facility TS and that a license amendment would
therefore be required for any subsequent changes.
Operator training and qualification programs provide an essential
[[Page 15715]]
component of supporting human performance in implementing tasks with
safety implications. Such programs must include components that cover
the stages of initial training, examination, and continuing training.
Additionally, recognizing the potential for varying concepts of
operations to affect traditional, prescriptive approaches to operator
proficiency, under part 53 the NRC allows facilities to develop
operator proficiency programs based on facility-specific
considerations.
Therefore, the NRC requires in Sec. 53.730(g)(1), as part of its
approval of the OL or COL, approval of the programs that will be used
for the initial training, initial examination, requalification training
and examination, and proficiency of both licensed operators and senior
operators. In a corresponding manner, the NRC requires in Sec.
53.730(g)(2) approval of the programs that will be used for the GLRO
equivalents of each of these programs for facilities with such
staffing. The NRC intends that examination program requirements will be
associated with staff guidance for the review of tailored examination
processes that are planned to accompany part 53. Following the
completion of an initial training program, continuing training programs
provide an important means of sustaining the knowledge and abilities of
individuals. The NRC adapts the requirements of Sec. 50.54(i-1) in
Sec. 53.730(g)(3) to require that operator continuing training
programs be in effect to support operator performance. Under part 53,
the NRC requires these programs to be in effect concurrent with when
the initial operator examinations first commence, in effect putting the
programs in place only when they are needed. This represents a
modification of the comparable requirement of Sec. 50.54(i-1), which
links the commencement of these programs to a timeline driven by the
licensing of the facility.
The authorization to manipulate controls of the facility that
directly affect reactivity or power level is restricted to individuals
who are either licensed operators, licensed senior operators, or GLROs.
However, for practical purposes, situations in which an individual is
participating in an approved training program or reestablishing
proficiency may also call for them to operate the controls of the
facility under the cognizance of a licensed individual. The NRC adapts
the requirements of Sec. 55.13 in Sec. 53.735 to accomplish this,
with a notable difference being the incorporation of GLROs.
Section 53.740 provides requirements for OL and COL holders under
part 53. Portions of Sec. 53.740 are adapted from the conditions of
Sec. 50.54. In general, the conditions for operations staffing under
part 53 reflect considerations for potential technological differences
and varying concepts of operation that are expected among part 53
facility licensees. Additionally, certain requirements are specific to
the operating phase while others remain in effect following the
permanent cessation of facility operations during the decommissioning
phase.
All commercial nuclear plants licensed under part 53 require some
form of licensed operator staffing, whether it be by specifically or
generally licensed operators. Consistent with this, the NRC requires
under Sec. 53.740(a) that facility licensees demonstrate compliance
with the programmatic requirements for either specifically licensed
operators and senior operators or for GLROs, as applicable to the
facility.
The NRC recognizes that technology-inclusive facility staffing will
need to account for a potentially wide range of concepts of operations;
for this reason, flexible and performance-based approaches for
establishing required facility staffing are appropriate. However, once
the appropriate facility staffing has been determined and approved by
the NRC, such staffing must be maintained to ensure that the
appropriately qualified individuals will be available when needed to
support the safe operation of the facility. Therefore, the NRC requires
under Sec. 53.740(b) that the staffing described within the approved
facility staffing plan be maintained as a condition of the facility
license as opposed to prescriptive staffing requirements like those of
Sec. 50.54(k) and (m).
Because operation of facility controls directly affects reactivity
or power level, only those individuals who possess appropriate levels
of qualification and authorization are permitted to operate those
controls. The NRC adapts the requirements of Sec. 50.54(i) in Sec.
53.740(c) to require that only specifically licensed operators and
senior operators or, alternatively, GLROs, may operate facility
controls, with allowance for specified exceptions for the purposes of
operator training or proficiency.
Senior operators, by virtue of their license level, are qualified
and authorized both to perform certain important responsibilities and
to direct the licensed activities of licensed operators. Therefore,
facilities that are required to be staffed by specifically licensed
operators must also include senior operators within their staffing. In
contrast, facilities staffed with GLROs only have a single license
level available and, therefore, there is no equivalent provision for
such facilities. The NRC adapts the requirements of Sec. 50.54(l) in
Sec. 53.740(d) to require the licensing and designation of senior
operators at facilities staffed by specifically licensed operators.
In contrast with control manipulations that directly affect reactor
power and reactivity (e.g., control rod movement, control drum
rotation, recirculation pump speed adjustment, reactor coolant system
boration or dilution, etc.) and are therefore restricted to performance
only by licensed operators, other types of plant operations that may
result in reactor power and reactivity changes via means that are
indirect in nature (e.g., electrical generation changes, turbine bypass
valve operation, steam usage by process heat applications, etc.) may be
implemented by non-licensed personnel. However, due to the potential
influence of such operations on reactor power and reactivity, the
continuous oversight of reactor parameters by a licensed operator is
necessary during these operations. The NRC therefore adapts the
requirements of Sec. 50.54(j) in Sec. 53.740(e) to require
appropriate oversight of operations, other than those associated with
the controls themselves, that may affect reactivity or power level.
Load following where plant output automatically changes in response
to externally originated instructions or signals is not permitted under
the existing regulations of Sec. 50.54. However, new technological
considerations and concepts of 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. Section 53.740(f) allows for load following,
provided that appropriate measures are in place. In considering the
acceptability of the measures associated with load following, the NRC
expects that any automatic protection relied
[[Page 15716]]
upon would be separate from that credited for reactor protection
purposes and would employ setpoints that are set so as to prevent
actuation of the reactor protection system while accomplishing its
functions to the extent practical.
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 adapts the
requirements of Sec. 50.54(m)(2)(iv) in Sec. 53.740(g) to require the
supervision of core alterations by either a specifically licensed
senior operator, a specifically licensed senior operator whose license
is limited to fuel handling, or by a GLRO, as applicable to the
facility. Because certain commercial reactor designs may be capable of
refueling while at power and, in any event, overall facility oversight
will already be required by either a specifically licensed senior
operator or by a GLRO, the NRC omits this requirement as redundant
during periods where core alterations occur while the plant is
operating.
It is impossible to predict every possible scenario that a
commercial 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 adapts the requirements of Sec. 50.54(x) and (y) in Sec.
53.740(h) to permit specific individuals to authorize departures from
facility license conditions or TSs when emergency conditions warrant
doing so for the protection of the public health and safety.
Recognizing that certain facilities licensed under part 53 may be
staffed by GLROs in lieu of specifically licensed senior operators, the
NRC extends this authority to GLROs. While it is not anticipated that
GLROs will have a role in the fulfillment of safety functions at self-
reliant-mitigation facilities, nor is it anticipated that operators at
such facilities would be in a position by which to significantly
influence radiological safety outcomes, the very nature of the Sec.
50.54(x) and (y) and Sec. 53.740(h) provisions concern situations that
are unanticipated and, therefore, unforeseeable. Thus, it is
appropriate 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.
Due to the unique authorities and responsibilities of both
specifically and generally licensed reactor operators, it is 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 adapts the requirements of Sec. 55.3 in Sec. 53.745 to
require that any person performing the function of an operator, senior
operator, or GLRO must be authorized by a license issued by the
Commission.
The NRC will license individuals as operators under both specific
and general licensing frameworks. Specific licenses will be for
licensed operators (i.e., reactor operators) and senior operators
(i.e., senior reactor operators) and will be issued to a named person
upon approval by the Commission of an application for that named
person. In contrast, GLROs will perform duties under the provisions of
a general license that is effective without the filing of an
application with the Commission or the issuance of licensing documents
to a particular person. The NRC sets forth requirements for the use of
a specific licensing process for licensed operators and senior
operators under Sec. Sec. 53.760 through 53.795, with Sec. 53.760
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 is necessary. The NRC adapts the
requirements of Sec. Sec. 55.21, 55.23, and 55.27 under Sec. 53.765
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 does not extend a comparable medical requirement
to GLROs.
The NRC also adapts the requirements of Sec. Sec. 55.25 and
50.74(c) in Sec. 53.770 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.
Notwithstanding this requirement related to permanent medical
conditions, the NRC continues to recognize that it is appropriate for
facility licenses to impose administrative restrictions and conditions
upon specifically licensed operators and senior operators in response
to temporary medical conditions.
The process of specifically licensing individuals as licensed
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 adapts
the requirements of Sec. Sec. 55.31 through 55.35 in Sec. 53.775 to
include requirements for the applications associated with the specific
licensing of licensed operators and senior operators at commercial
nuclear plants licensed under part 53. In contrast with the part 55
requirements, the NRC provides additional flexibility by locating
certain details associated with the preparation and submittal of these
applications within guidance in lieu of placement within this final
rule itself.
The NRC includes overall programmatic requirements for specifically
licensed operator and senior operator training, examination, and
proficiency in Sec. 53.780. In general, the requirements are adapted
from those in part 55, with several additional flexibilities being
incorporated to better account for potential variations in reactor
technologies and concepts of operations. The requirements in Sec.
53.780 cover, in part, 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 a systems approach to training (SAT) ensures that
the
[[Page 15717]]
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 fulfillment of design-specific
safety functions. The NRC requires under Sec. 53.780(a) that facility
licensees implement a SAT-based training program for the initial
training of licensed operator and senior operator applicants. The
program must be adequate to ensure that applicants will be capable of
performing the duties necessary both to protect public health and
safety and to maintain plant safety functions. The NRC further requires
that such programs be subject to NRC approval and subsequent change
control processes of an appropriate nature.
Examinations provide a means of assessing that individuals have
achieved a degree of knowledge and ability that is sufficient to carry
out assigned duties as licensed operators or senior operators in a
manner that is safe and reliable. The NRC adapts the requirements of
Sec. Sec. 55.40, 55.41, 55.43, and 55.45 in Sec. 53.780(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 have the flexibility to propose, subject to NRC
approval, the examination methods and criteria to be used in assessing
satisfactory applicant performance. Such examination programs
(including those used within the scope of requalification training)
must provide for acceptable levels of both test validity and test
reliability in order to be considered acceptable. The NRC intends that
staff guidance will 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 provides
holders of licenses to operate commercial nuclear plants under part 53
the alternative of administering their own approved licensing
examinations. The NRC will 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. However,
irrespective of the provided flexibilities in examination format and
structure, at a minimum, topics from the following general categories
of knowledge and abilities should be sampled in such examinations:
Reactor Theory, Thermodynamics, and Chemical Interactions
Plant Systems and Components
Reactivity Management and Manipulations
Radiation Control and Safety
Emergency, Abnormal, and Normal Operations
Administrative Requirements and Conditions of the Facility
License
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 adapts the requirements of Sec. 55.59 in Sec.
53.780(c) to require that facilities implement both a 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 distinct
annual operating test and biennial written examination components are
not mandated, with the facility licensee instead proposing the
examination methods and criteria to be used in assessing satisfactory
performance. The NRC intends that guidance will 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 will 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
adapts the requirements of Sec. 55.49 in Sec. 53.780(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
Nuclear Waste Policy Act of 1982, as amended (NWPA), section 306 (42
U.S.C. 10226) to establish regulations for the use of simulators within
such context. The NRC adapts the requirements of Sec. 55.46 in Sec.
53.780(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 part 53 do not mandate that full scope, plant-
referenced simulators be used and will 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. Additionally, in allowing for the possibility that an
applicant or licensee might demonstrate compliance with training,
examination, or experience requirements using the plant itself, the NRC
is not allowing the initiation of transients on the actual plant.
Consistent with this, aside from controlled reactivity manipulations
that are conducted for the purposes of demonstrating compliance with
experience requirements, actual plant components may not be operated
for these purposes. Rather, the NRC perspective is that the use of the
plant for training and examination purposes should be restricted to
techniques such as walkthroughs, job performance measures, simulated
tasks, use of augmented reality technology, and similar approaches that
provide training and examination value while avoiding the operation of
actual plant components.
There may be situations in which applicants for operator or senior
operator licenses have previous training and experience that justifies
waiving some, or all, of the initial examination requirements. The NRC
adapts the requirements of Sec. 55.47 in Sec. 53.780(f) to allow for
consideration of requests for waivers of examinations requirements. In
contrast with the part 55 requirements, the NRC locates certain details
associated with such waiver requests within guidance documentation in
lieu of placement within this final rule itself.
For licensed operators and senior operators to perform their
assigned duties safely and reliably, it is essential that they perform
those duties frequently enough so as to maintain a sufficient degree of
proficiency. The NRC adapts the requirements of Sec. 55.53(e) and (f)
in Sec. 53.780(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, in recognition of the fact that varying concepts of operations
are possible for advanced reactor facilities, the NRC, in contrast with
the requirements of part 55, is allowing facility licensees to
establish their own programs for operator proficiency, subject to NRC
approval.
As the holders of specific licenses, licensed operators and senior
operators
[[Page 15718]]
must be subject to license conditions on an individual basis to ensure
that the basis upon which the licenses were issued remains valid. The
NRC adapts the requirements of Sec. 55.53 in Sec. 53.785 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 is allowing certain aspects of operator
proficiency to be addressed by an NRC-approved facility proficiency
program.
Licenses for specifically licensed operators and senior operators
are issued by the NRC and must remain subject to modification or
revocation. The NRC adapts the requirements of Sec. Sec. 55.51 and
55.61 in Sec. 53.790 to address the issuance, modification, and
revocation of licenses issued to specifically licensed operators and
senior operators.
The licenses issued to specifically licensed operators and senior
operators are valid for a period of 6 years, after which they expire,
unless otherwise renewed. The NRC adapts the requirements of Sec. Sec.
55.55 and 55.57 in Sec. 53.795 to address the expiration and renewal
of licenses issued to specifically licensed operators and senior
operators.
In developing this final rule, the NRC has discussed with
stakeholders the considerations that might justify the omission of the
specifically licensed operators and senior operators. However, even for
an inherently safe reactor with autonomous operation features, certain
important administrative functions (e.g., compliance with TS,
operability determinations, NRC notifications, emergency declarations,
risk assessment, maintenance oversight, and radiological release limit
compliance) would still need to be accomplished by appropriately
qualified and authorized individuals. Additionally, the NRC recognized
that manual manipulations of facility reactivity controls must only be
performed by individuals who have been appropriately licensed by the
Commission. The NRC therefore establishes under Sec. 53.800 a new
class of facility (defined as a self-reliant-mitigation facility),
according to the criteria contained in Sec. 53.800 for part 53. These
facilities will employ GLROs rather than specifically licensed
operators and senior operators. The GLRO regulations offer enhanced
flexibilities and targeted relaxations in a manner that is commensurate
with the modified role of such operators to ensure the safe operation
of the associated facilities. In contrast, those facilities not meeting
the criteria of Sec. 53.800 will instead be considered interaction-
dependent-mitigation facilities and will require staffing by
specifically licensed operators and senior operators. The terminology
used to designate these facility types reflects differences in how
operators are anticipated to need to interact with their plant systems
in mitigating events and achieving safe outcomes; such systems may
either need operators to interact with them in some manner (i.e., be
interaction-dependent) or may instead be able to rely fully upon their
own capabilities independent of operator interaction (i.e., be self-
reliant).
Generally licensed reactor operators differ from specifically
licensed operators because the latter will be directly and
independently evaluated by the NRC as part of their licensing process.
This direct and independent evaluation remains appropriate when
operators may reasonably be expected to exert a significant influence
on public health and safety outcomes. Therefore, a key determinant as
to whether generally licensed reactor operators can be utilized in
facility staffing is the assessment of the operator's role in
maintaining and fulfilling safety functions at the facility, such as
through the performance of credited actions for the mitigation of plant
events.
The criteria in Sec. 53.800 designate self-reliant-mitigation
facilities. These criteria are derived from the following set of
considerations:
no human action needed to satisfy radiological consequence
criteria;
no human action needed to address LBEs;
safety functions not allocated to human action;
reliance upon robust and highly reliable safety features; and
appropriate defense in depth achieved without reliance on
important human action.
It should be noted that those facilities not meeting the criteria in
Sec. 53.800 will instead be classified as interaction-dependent-
mitigation facilities and will require staffing by specifically
licensed operators and senior operators instead.
Generally licensed reactor operators will perform duties under the
provisions of a general license that is effective without the filing of
an application with the Commission or the issuance of licensing
documents to a particular person. The NRC sets forth requirements for
the general licensing process for GLROs under Sec. Sec. 53.805 through
53.820. The requirements for GLROs parallel those for senior operators
in regard to their comparable administrative responsibilities.
Nonetheless, the requirements for GLROs are relaxed and incorporate
greater flexibilities compared to the requirements for specifically
licensed operators in a manner that is consistent with the GLRO's role
in safety at self-reliant-mitigation facilities.
In order to use GLROs in lieu of specifically licensed operators
and senior operators, a OL/COL applicant must demonstrate that its
proposed facility is a self-reliant-mitigation facility, i.e., that it
will comply with the following requirements on an ongoing basis:
maintaining GLRO qualifications for the performance of important
functions and tasks; incorporating relevant programmatic controls into
TS; administering the related programs for training, examination, and
proficiency; and ensuring that the relevant provisions of parts 26 and
73 are met. Additionally, to provide for an accurate accounting of what
individuals are licensed under the general license, facility licensees
are required to report the identities of all generally licensed reactor
operators to the NRC on an annual basis. Furthermore, a facility
licensee must ensure that the facility design and performance continue
to meet the technological criteria to be classified as a self-reliant-
mitigation facility (i.e., the criteria of Sec. 53.800) on a continual
basis during the operating phase, as the relaxations afforded to such
facilities in the areas of operator licensing, staffing, and HFE are
predicated on this assumption. The NRC therefore establishes under
Sec. 53.805 requirements for facility licensees that address issues
such as these. Finally, the failure of a self-reliant-mitigation
facility to subsequently meet the criteria of Sec. 53.800 after the
issuance of an OL or COL will constitute a reportable event (i.e., an
unanalyzed condition that significantly degrades plant safety) under
the provisions of Sec. 53.1630.
The NRC sets forth the general license for GLROs under Sec.
53.810. GLROs will be licensed as a class of individuals under the
provision of Sec. 53.810(a) and will be subject to the conditions
specified in Sec. 53.810(b) through (g). Portions of these conditions
are adapted from Sec. 55.53 and from those conditions currently
included in the licenses issued to specifically licensed operators and
senior operators. The NRC retains the ability to suspend or prohibit
individuals from operating under the general license should such action
be warranted.
The NRC includes overall programmatic requirements for GLRO
training, examination, and proficiency under Sec. 53.815. In general,
these
[[Page 15719]]
requirements are adapted from those of part 55 and parallel those also
included for specifically licensed senior operators in Sec. 53.780.
These requirements include increased flexibilities and several targeted
relaxations that reflect the limited role of GLROs in facility safety.
The requirements under Sec. 53.815 cover, in part, the initial
training, initial examination, continuing training, requalification
examination, and proficiency of GLROs. Section 53.805 requires the
facility licensee to develop, implement, and maintain these programs.
Section 53.810, in turn, prescribes that the requirements of Sec.
53.805 must be met as a requirement of the general license. The
implication of this structure is that the facility licensee must
implement these programs for training, examination, and proficiency,
and GLROs must participate in these programs to demonstrate compliance
with the requirements of the general license.
The initial training process provides GLROs with the knowledge and
abilities needed to fulfill assigned duties as GLROs. The use of an SAT
serves 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. The NRC requires under
Sec. 53.815(b) that facility licensees implement a SAT-based training
program for the initial training of GLROs that is adequate to ensure
that they have the necessary knowledge, skills, and abilities to
perform their duties. The NRC further requires that such programs be
subject to NRC approval, oversight, and appropriate change control
processes. The training program must ensure that GLROs maintain the
necessary knowledge, skills, and abilities.
Examinations provide a means of assessing that individuals have
achieved a degree of knowledge and ability that will be sufficient to
enable them to carry out assigned duties as GLROs in a manner that is
both safe and reliable. The NRC adapts the requirements of Sec. Sec.
55.40, 55.41, 55.43, and 55.45 in Sec. 53.815(b) to require that
facility licensees establish and implement an initial examination
program. A key difference from the comparable requirements of part 55
is that facility licensees are afforded the flexibility to propose,
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)
must provide for acceptable levels of both test validity and test
reliability in order to be considered acceptable. The NRC intends that
staff guidance will be available to facilitate the review of initial
examination programs that are proposed by facility licensees and that
approved initial examination programs will be subject to an appropriate
change control process. In contrast with both the requirements of part
55 and the requirements of Sec. 53.780, the NRC does not intend to
administer or evaluate these initial examinations. However, the
examination processes themselves will continue to be subject to ongoing
NRC oversight. Irrespective of the provided flexibilities in
examination format and structure, topics from the following general
categories of knowledge and abilities should be sampled in such
examinations:
Reactor Theory, Thermodynamics, and Chemical Interactions
Plant Systems and Components
Reactivity Management and Manipulations
Radiation Control and Safety
Emergency, Abnormal, and Normal Operations
Administrative Requirements and Conditions of the Facility
License
Continuing training programs 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 adapts the requirements of Sec. 55.59 in Sec.
53.815(b) to require that facility licensees implement both an SAT-
based continuing training program and a requalification examination
program. However, a notable difference from the examinations required
under part 55 is that distinct annual operating test and biennial
written examination components are not mandated. The facility licensee
will instead propose examination methods and criteria to be used in
assessing satisfactory performance. Furthermore, unlike the comparable
requirements of part 55 and those for specifically licensed operators
and senior operators, a biennial periodicity for requalification
examinations is not prescribed. However, adequate justification for the
proposed periodicity of requalification examinations is required. The
NRC intends that staff guidance will be available to facilitate the
review of the requalification examination programs that are proposed by
facility licensees. Approved requalification examination programs will
be subject to an appropriate change control process.
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
adapts the requirements of Sec. 55.49 in Sec. 53.815(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 NWPA,
section 306 (42 U.S.C. 10226) to establish regulations for the use of
simulators within such context. The NRC adapts the requirements of
Sec. 55.46 in Sec. 53.815(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 is not mandated. The potential
use of alternative simulation facilities consisting of, for example,
partial scope simulators or the plant itself, is allowed provided that
all associated requirements are demonstrated to be met using
alternative approaches and methods. Additionally, in allowing for the
possibility that an applicant or licensee might demonstrate compliance
with training and examination requirements using the plant itself, the
NRC is not allowing the initiation of transients on the actual plant.
Consistent with this, aside from controlled reactivity manipulations
that are conducted for the purposes of demonstrating compliance with
experience requirements, actual plant components may not be operated
for these purposes. Rather, the use of the plant for training and
examination purposes should be restricted to techniques such as
walkthroughs, job performance measures, simulated tasks, use of
augmented reality technology, and similar approaches that provide
training and examination value while avoiding the operation of actual
plant components.
There may be situations in which GLROs have previous training and
experience that justifies waiving some, or all, of the initial
examination. Therefore, under Sec. 53.815(f) the NRC allows facility
licensees to waive some, or all, portions of initial examinations
provided that such waivers are consistent with a program that has been
approved by the NRC.
For GLROs to safely and reliably perform their assigned duties, it
is essential that they perform those duties frequently enough so as to
maintain a sufficient degree of proficiency.
[[Page 15720]]
However, the NRC recognizes that facilities that utilize GLROs may have
concepts of operation that warrant unique proficiency considerations.
Therefore, the NRC requires in Sec. 53.815(g) that facility licensees
develop, implement, and maintain programs to maintain and reestablish,
if needed, the proficiency of GLROs. This could occur, for example, if
an individual's extended absence from watch standing has rendered
proficiency requirements unmet.
The general license should remain in effect for an individual only
while that individual remains employed in a position that may call for
the individual to manipulate the reactivity controls of the facility.
The NRC requires under Sec. 53.820 that the general license ceases to
be applicable on an individual basis when an individual's employment
status becomes such that this is no longer the case. However, the NRC
recognizes that for some types of self-reliant-mitigation facilities,
very long periods may elapse between circumstances that necessitate
manual manipulation of reactivity controls. Therefore, the general
license remains 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.
The NWPA, section 306 (42 U.S.C. 10226) 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 adapts under Sec. 53.830, with modifications, the
requirements of Sec. 50.120 for use in part 53 to provide more
flexible personnel training and qualification requirements than those
in Sec. 50.120 and better reflect diverse concepts of operations.
The NRC recognizes that the categories of nuclear power plant
personnel in Sec. 50.120 may not be needed for the diverse concepts of
operations, staffing models, and non-traditional personnel roles and
responsibilities anticipated under part 53; conversely, and for the
same reasons, additional categories of personnel may need to be covered
by part 53. The NRC also recognizes that the timeframe prescribed in
Sec. 50.120 for the establishment of training programs may not be
aligned with the schedules associated with the startup of certain types
of commercial nuclear plant facilities. However, the NRC also
recognizes that the SAT-based training required under Sec. 50.120
remains an appropriate means by which training programs should continue
to be developed and implemented. Therefore, the approach taken by the
NRC in addressing the training of certain plant staff under part 53
reflects greater flexibilities in personnel categories and programmatic
timeframes, while still retaining the requirement that such training
programs be based on SAT.
The NRC requires under Sec. 53.830 SAT-based training programs
with the timeframe for when such programs are required being based upon
when the associated personnel are needed to support facility-specific
needs. The training programs will cover the training and qualification
of personnel in the general categories of supervisors, technicians, and
other appropriate operating personnel. Regarding the category of
supervisors, this is intended to reflect on-shift supervisors for the
licensed operators, similar to the current classification in Sec.
50.120(b)(2)(iii), but Sec. 53.830 uses language that is less specific
to account for different conduct of operations and organizational
structures for commercial nuclear plants which may require greater
regulatory flexibility. The licensee is not required to seek NRC
approval of a training program prior to usage. However, the licensee is
required to accommodate NRC inspection of the training program. The NRC
intends to develop guidance to facilitate the inspection of these
training programs but does not intend for such guidance to preclude the
potential for the training programs to be maintained by a separate,
NRC-approved accreditation process.
Section 53.845 requires programs to be developed, implemented, and
maintained to help ensure that design features and human actions have
the capabilities and reliabilities necessary to demonstrate compliance
with the safety criteria in subpart B throughout the operating life of
each commercial nuclear plant. The programmatic requirements in subpart
F also address areas such as radiation protection needed to control
routine effluents during normal operations. Sections 53.850 through
53.910 require programs to support specific activities needed to ensure
the prevention or mitigation of unplanned events or to support normal
operations for any reactor design. However, each holder of an OL or COL
is required to assess whether additional programs are needed for the
specific reactor design and location of the commercial nuclear plant.
Licensees are able to combine, separate, and otherwise organize
programs and related documents as appropriate for the technologies and
organizations associated with the commercial nuclear plant.
Section 53.850 requires a radiation protection program associated
with the requirements in subparts B and C for public doses resulting
from normal operations and the protection of plant workers. The
requirements related to doses from normal operations, including routine
effluents, are similar to those specified in Sec. 50.36a, ``Technical
specifications on effluents from nuclear power reactors,'' and related
requirements in standard TS for offsite dose calculation manuals. While
the section includes requirements that are technically and
programmatically similar to part 50, Sec. 53.850 does not include a
requirement for effluent-related TS as is required in Sec. 50.36a. A
requirement similar to that found in the administrative controls
section of TS for operating reactors licensed under parts 50 and 52 is
included for programmatic controls of solid wastes to complement the
design requirements in Sec. 53.425.
Section 53.855 requires an emergency response plan that
demonstrates compliance with the requirements in appendix E to part 50
and Sec. 50.47(b) or Sec. 50.160. The regulations in Sec. 50.47
stating that the NRC will not issue certain licenses unless it finds
that there is reasonable assurance that adequate protective measures
can and will be taken to protect public health and safety in the event
of a radiological emergency apply equally to applications under part 53
complying with the applicable standards set forth in either Sec.
50.160 or the requirements in appendix E to part 50 and Sec. 50.47(b).
In its 2008 Advanced Reactor Policy Statement, the Commission
stated their expectation that ``the safety features of advanced reactor
designs will be complemented by the operational program for Emergency
Planning (EP). This EP operational program, in turn, must be
demonstrated by inspections, tests, analyses, and acceptance criteria
to ensure effective implementation of established measures.''
Consistent with this policy statement, emergency plans and emergency
planning zones are not safety features in the design. In SECY-97-020,
``Results of Evaluation of Emergency Planning for Evolutionary and
Advanced Reactors,'' dated January 27, 1997, the staff indicated that
the rationale upon which EP for current reactor designs is based, that
is, potential consequences from a spectrum of accidents, is appropriate
for use as the basis for EP for evolutionary and
[[Page 15721]]
passive advanced LWR designs and is consistent with the Commission's
defense-in-depth safety philosophy. Also, in its Safety Goals Policy
Statement the Commission stated that: ``A defense-in-depth approach has
been mandated in order to prevent accidents from happening and to
mitigate their consequences. Siting in less populated areas is
emphasized. Furthermore, emergency response capabilities are mandated
to provide additional defense-in-depth protection to the surrounding
population.'' Consistent with this policy statement, Sec. 53.855
contributes an additional independent layer of defense in depth for
commercial nuclear plants. Therefore, the emergency plans and emergency
planning zones under Sec. 53.855 are not used to demonstrate
compliance with subpart B and subpart C of part 53. Rather, compliance
with the requirements in Sec. 53.855 provides reasonable assurance
that adequate protective measures can and will be taken to protect
public health and safety in the event of a radiological emergency.
Section 53.860 identifies the applicable regulations for part 53
applicants related to the programs for physical security,
cybersecurity, FFD, AA, and information security. These programs are
discussed in more detail in section IV, ``Changes to Other Parts of 10
CFR,'' of this document.
Section 53.860(a) requires licensees to develop, implement, and
maintain a physical protection program that meets either Sec. 73.55 or
Sec. 73.100, and includes physical protection of SNM and Category 1
and Category 2 radioactive material, if applicable.
Section 53.860(b) requires licensees to establish, implement, and
maintain an FFD program under part 26. Section 53.860(c) requires
licensees to establish, implement, and maintain an AA program in
accordance with either Sec. 73.56 or Sec. 73.120, as appropriate.
Section 53.860(d) requires licensees to establish, implement, and
maintain a cybersecurity program in accordance with either Sec. 73.54
or Sec. 73.110. Section 53.860(e) requires licensees to establish,
implement, and maintain an information protection system that complies
with the requirements of Sec. Sec. 73.21, 73.22, and 73.23, as
applicable.
Section 53.865 establishes requirements for quality assurance and
refers to appendix B to part 50 for the part 53 requirements for SR
design features. Requirements related to evaluating and reporting
changes to the quality assurance program are included in subpart I and
are equivalent to those found in Sec. 50.54.
Section 53.870 requires licensees to actively assess possible
degradation of SSCs from the effects of aging, fatigue, and
environmental conditions. The inclusion of requirements related to
designing and monitoring for possible degradation mechanisms reflects
important lessons learned from the history of LWRs and the likely
introduction of new design features and materials in future commercial
nuclear plants. The allowable combinations of design features,
operating experience, testing, and monitoring during operations support
performance-based approaches to the initial licensing of new
technologies. The performance-based approach to integrity assessment
programs also allows for the subsequent consideration of operating
experience and appropriate corrective actions or allowable relaxations
for ensuring that design features comply with the functional design
criteria of Sec. Sec. 53.410 and 53.420. The program is based upon a
comprehensive and integrated evaluation of the aging and other
degradation mechanisms applicable to the design; identification of the
affected SSCs; the allowances provided in the design of the SSCs for
degradation; and schedules and procedures for determining if and at
what rate degradation is occurring, as well as its cause. Risk insights
can be used to prioritize the monitoring, evaluation, and management of
degradation based upon the importance of the SSC to safety and the time
frame for when the effects of degradation could be of concern.
Section 53.875 establishes requirements for a fire protection
program supporting operations similar to Sec. 50.48. The fire
protection program during operations will work in concert with specific
fire protection requirements in subpart C for design and analyses and
in subpart E for construction and manufacturing.
Section 53.880 establishes requirements for an inservice inspection
(ISI) and inservice testing (IST) program, which are historically
important activities conducted in accordance with ASME codes and
regulations in Sec. 50.55a. While part 53 does not incorporate
specific consensus codes and standards into the regulations, Sec.
53.880 allows for the use of generally accepted codes and standards.
The requirement for an ISI and IST program reinforces the need to
develop monitoring programs to be conducted during a plant's operations
phase to complement the design process and address inherent
uncertainties. The NRC encourages the continued use of consensus codes
and standards supporting design, testing, and inspections to support
integrated and performance-based approaches in demonstrating compliance
with the requirements in part 53.
Section 53.910 establishes requirements for developing,
implementing, and maintaining procedures (e.g., operations and
emergency operating procedures) and guidelines (e.g., accident
management guidelines). The programmatic requirements for many of the
procedures listed in this section are similar to the requirements found
in the administrative controls section of TS for plants licensed under
parts 50 and 52. The inclusion, where appropriate, of accident
management guidelines in these requirements is intended to ensure that
an integrated set of procedures and guidelines is established by
licensees to ensure command and control across the spectrum of possible
event sequences. The required procedures also include those needed to
complement the design requirements in Sec. 53.440(m) related to
criticality alarms and the equivalent of the procedures required in
Sec. 50.54(hh) to address notifications of potential aircraft threats.
Subpart G--Decommissioning Requirements
Subpart G provides the regulatory requirements for the
decommissioning phase of the life cycle of a commercial nuclear plant.
The requirements in subpart G for the decommissioning of a commercial
nuclear plant are adapted from the current regulations in Sec. 50.75,
``Reporting and recordkeeping for decommissioning planning,'' Sec.
50.82, ``Termination of license,'' and Sec. 50.83, ``Release of part
of a power reactor facility or site for unrestricted use.'' Although
the requirements from those sections of part 50 have been copied into
subpart G with relatively few changes, the requirements are reorganized
to fit within the part 53 structure. The few changes made were
primarily to make the requirements more technology-inclusive by adding
alternatives within sections, whereas some requirements in part 50 were
developed specifically for LWRs.
As an example, Sec. 50.75 provides minimum amounts of
decommissioning funds required to demonstrate reasonable assurance of
funds for decommissioning LWRs. Such generic amounts have not been
developed for all reactor technologies that may be licensed under part
53. Therefore, a requirement is included in Sec. 53.1020, ``Cost
estimates for decommissioning,'' for site-specific cost estimates for
decommissioning to be developed
[[Page 15722]]
considering costs in such areas as engineering, labor, and waste
disposal. The derivation of the generic cost estimates for LWRs in
Sec. 50.75 is provided in NUREG/CR-5884, ``Revised Analyses of
Decommissioning for the Reference Pressurized Water Reactor Power
Station,'' and NUREG/CR-6187, ``Revised Analyses of Decommissioning for
the Reference Boiling Water Reactor Power Station.'' Similar to part
50, a provision for an annual adjustment of decommissioning cost
estimates is included in Sec. 53.1030.
The NRC is currently pursuing another rulemaking, ``Regulatory
Improvements for Production and Utilization Facilities Transitioning to
Decommissioning,'' which was published as a proposed rule for public
comment on March 3, 2022 (87 FR 12254). As these rulemakings progress,
the NRC will consider revisions to part 53 to align the two rulemaking
efforts. For example, Sec. 53.1075 could be expanded to include or
reference requirements for decommissioning in areas such as EP and
security in addition to the decommissioning fire protection plans that
provide an equivalent to Sec. 50.48(f).
Subpart H--Licenses, Certifications, and Approvals
Subpart H provides requirements related to applications under part
53 for NRC licenses, certifications, or approvals for commercial
nuclear plants.
Subpart H specifies requirements applicable to all part 53
applications as well as requirements specific to part 53 applications
for LWAs, ESPs, standard design approvals, standard DCs, MLs, CPs, OLs,
and COLs. Subpart H is equivalent to and includes all existing
licensing, certification, and approval processes currently covered
under parts 50 and 52, with the exception of the process for early
review of site suitability issues. Interactions with external
stakeholders during the development of the proposed rule did not
identify significant interest in or need for including the process for
early review of site suitability issues in part 53.
Much of the subpart H regulatory text is identical to the
corresponding language in parts 50 and 52, with minor changes to
account for cross-references in part 53, to make language technology
neutral, or to reflect the unique analytical approach in part 53. In
these instances, this preamble discussion will describe the language as
``equivalent'' to the existing corresponding requirement in part 50 or
part 52 and will describe any deviations, where applicable.
Because part 53 carries over the majority of the licensing options
from parts 50 and 52, there are several sections in subpart H that are
similar to existing regulations in parts 50 and 52. Section 53.1100
addresses filing of applications for licenses, certifications, or
approvals under oath or affirmation and is equivalent to Sec. 50.30.
Section 53.1100 does not include the current requirement in Sec.
50.30(a)(2) that the applicant maintain the capability to generate
additional copies, because it is unnecessary in the age of electronic
submissions. In addition, the existing requirement on applications for
OLs in Sec. 50.30(d) is included in Sec. 53.1124(g)(2),
``Relationship between sections,'' covering OLs, rather than in Sec.
53.1100. Section 53.1100(a)(1) also includes filing requirements
equivalent to those in Sec. Sec. 52.15, 52.45, 52.135, and 52.151.
Section 53.1101 lays out activities requiring an NRC license and is
equivalent to Sec. 50.10(b). Section 53.1103 addresses combining
applications and is equivalent to Sec. Sec. 50.31, 50.52, and 52.8.
Section 53.1103(b) continues the Commission's practice of combining
multiple authorizations for a facility under parts 30, 40, 50, 52, and
70 into one license based on the Commission's authority under section
161h of the AEA to combine NRC licenses. Section 53.1106 addresses
elimination of repetition and is equivalent to Sec. 50.32.
Section 53.1109 provides general information requirements for the
content of applications submitted to the NRC under part 53 and is
equivalent to Sec. 50.33, with the exception of Sec. 50.33(f) on
financial qualifications, which is covered in subpart J, and Sec.
50.33(h) on earliest and latest dates for completion of construction,
which is covered in Sec. 53.1306 of subpart H. Each application must
include information to address the items in Sec. 53.1109 as cited in
the appropriate section of subpart H for the application type.
One change from current requirements can be found in Sec.
53.1109(i), which is not limited to electricity generation as it is
currently in part 50. Some prospective NRC applicants are considering
development of nuclear plants for other commercial ventures, such as
process heat generation or hydrogen production. In addition, Sec.
53.1109(j), which requires applications containing classified
information to separate that information from the unclassified
information in the application, refers to ``Restricted Data or
classified National Security Information'' instead of the term used in
the corresponding provision in Sec. 50.33(j), ``Restricted Data or
other defense information.'' This change was made to use the defined
term in 10 CFR part 95, ``Facility Security Clearance and Safeguarding
of National Security Information and Restricted Data,'' rather than
``defense information'' as used in Sec. 50.33(j). The usage in Sec.
50.33(j) dates back to the Atomic Energy Commission amendment of that
section on January 19, 1956 (21 FR 355, 357), and was not changed with
the issuance of part 95 (45 FR 14476; March 5, 1980) after the
establishment of the NRC and the 1975 reissuance of the former Atomic
Energy Commission regulations. The revised terminology also aligns with
its usage in Sec. 53.1115.
Section 53.1112 addresses environmental conditions and is
equivalent to Sec. 50.36b. Section 53.1115 addresses requirements for
agreements limiting access to classified information and is equivalent
to Sec. 50.37.
Section 53.1118 addresses ineligibility of certain applicants and
is similar to Sec. 50.38 but has been revised consistent with section
301 of the ADVANCE Act. That section of the ADVANCE Act designates
certain exceptions from the foreign ownership, control, or domination
(FOCD) provision set forth in the AEA. Specifically, section 301 states
that if the Commission determines that issuance of the applicable
license to that entity is not inimical to the common defense and
security or public health and safety, then the FOCD restriction shall
not apply to an entity that is owned, controlled, or dominated by: (1)
the government of a country that is a member of the Organisation for
Economic Co-operation and Development or the Republic of India; (2) a
corporation that is incorporated in one of those countries; or (3) a
citizen or national of one of those countries, subject to some
additional exclusions. Those additional exclusions are based on whether
any members of the excepted countries were on certain sanctions lists
on the ADVANCE Act's date of enactment.
Section 53.1120 addresses exceptions and exemptions from licensing
requirements for Department of Defense and DOE facilities and is
equivalent to Sec. 50.11. Section 53.1121 addresses public inspection
of applications and is equivalent to Sec. 50.39.
Section 53.1124 addresses the relationship between the various
licenses, certifications, and approvals provided in this subpart, and
the requirements are equivalent to a number of similar provisions in
parts 50 and 52, including Sec. Sec. 50.10, 52.13, 52.43, 52.73,
52.133, and 52.153. New provisions are provided in Sec. 53.1124(c) and
(d) that allow an application for either a
[[Page 15723]]
standard design approval or a standard DC under part 53 to reference
applicable licensing-basis information that supported issuance of an OL
or COL under part 53. These provisions will offer additional
flexibility beyond what is currently allowed under parts 50 or 52 for
an applicant who may wish to license a first-of-a-kind reactor for
operation prior to seeking generic approval or certification of the
standard design.
Section 53.1124(e) addresses the limitation that a manufactured
reactor may only be transported domestically to a site with a COL or CP
and is generally equivalent to Sec. 52.153. The NRC has not included
specific requirements within part 53 directing how a CP application
referencing an ML must be structured, including how the ITAAC required
for an ML under Sec. 53.1282 is to be addressed by a CP applicant.
Instead, part 53 leaves the matter open to possible approaches to be
addressed in future regulatory guidance or proposed by future
applicants. Section 53.1124(e) includes an additional statement to make
it clear that a manufactured reactor may be exported in accordance with
part 110.
Section 53.1130 addresses LWAs and is equivalent to Sec. 50.10.
Sections 53.1140 through 53.1188 address applications for, issuance
of, and other provisions related to ESPs under part 53. Section 53.1140
describes how the contents of Sec. Sec. 53.1140 through 53.1188
address ESPs and is equivalent to Sec. 52.12. Section 53.1144
addresses general information requirements for the content of
applications and is equivalent to Sec. 52.16.
Section 53.1146 specifies requirements for the technical contents
of applications and is equivalent to Sec. 52.17. Section 53.1146(b)(2)
provides applicants for ESPs a regulatory option to propose major
features of the emergency plans or complete integrated emergency plans
in accordance with either the requirements in Sec. 50.160 of this
chapter, or the requirements in appendix E to part 50 of this chapter
and Sec. 50.47(b) of this chapter, as applicable.
Section 53.1149 addresses standards for review of ESP applications
and administrative review of applications, including hearings, and is
equivalent to Sec. Sec. 52.18 and 52.21. Section 53.1155 addresses
referral to the ACRS and is equivalent to Sec. 52.23. Section 53.1158
addresses issuance of ESPs and is equivalent to Sec. 52.24. Section
53.1161 addresses the extent of activities permitted and is equivalent
to Sec. 52.25. Section 53.1164 addresses the duration of an ESP and is
equivalent to Sec. 52.26. Section 53.1167 addresses provisions for
requesting an LWA after issuance of an ESP and is equivalent to Sec.
52.27. Section 53.1170 addresses transfers of ESPs and is equivalent to
Sec. 52.28. Section 53.1173 addresses applications for ESP renewals
and is equivalent to Sec. 52.29, although the final rule removes the
requirement to refer the renewal to the ACRS consistent with current
agency practice. Section 53.1176 addresses criteria for renewal of an
ESP and is equivalent to Sec. 52.31. Section 53.1179 addresses the
duration of an ESP renewal and is equivalent to Sec. 52.33. Section
53.1182 addresses the use of a site for purposes other than those
described in the permit and is equivalent to Sec. 52.35. Section
53.1188 addresses finality of ESP determinations and is equivalent to
Sec. 52.39.
Sections 53.1200 through 53.1221 address applications for, issuance
of, and other provisions related to standard design approvals under
part 53. Section 53.1200 describes how the contents of Sec. Sec.
53.1200 through 53.1221 address standard design approvals and is
equivalent to Sec. 52.131. Section 53.1206 addresses general
information requirements for the content of applications and is
equivalent to Sec. 52.136.
Section 53.1209 addresses requirements for the technical content of
applications and is largely equivalent to Sec. 52.137. In Sec.
53.1209(a), the NRC includes text that expands the discussion of a
``major portion'' of standard design approvals. 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.'' Section 53.1209(b) outlines the required
content of the Final Safety Analysis Report (FSAR). Requirements in
Sec. 53.1209(b)(2) for portions of the application addressing design
information state that the application must include design information
equivalent to that required for a standard DC. This reference to the
pertinent DC requirements (specifically, those in Sec. 53.1239(a)(2)
through (27)) is an efficiency that prevents the need to repeat many of
the same requirements for the content of a standard design approval
application.
Section 53.1210 addresses requirements for the content of a
standard design approval application other than the FSAR. Section
53.1210(a) requires the inclusion of a description of availability
controls that are not included in the FSAR.
Section 53.1212 addresses standards for review of applications and
is equivalent to Sec. 52.139. Section 53.1215 addresses referral to
the ACRS and is equivalent to Sec. 52.141. Section 53.1218 addresses
staff approval of designs and duration of design approvals and is
equivalent to Sec. Sec. 52.143 and 52.147. Section 53.1221 addresses
finality of standard design approvals and information requests and is
equivalent to Sec. 52.145 with the exception that it extends such
finality to a standard approval referenced in a DC application.
Standard design approvals issued to date under part 52 have been issued
during the NRC's review of the standard DC application and have relied
on the same application content. However, a future scenario could arise
where the DC application is not submitted until after a design approval
has been granted. The NRC would apply the same finality provisions in
this situation as in the situation where a standard design approval is
referenced in a COL application.
There is no equivalent to Sec. 53.1221(d) in part 52 for standard
design approvals. This provision states that the Commission will
require, before granting a CP, COL, 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 part 52 in relation to a standard DC; and the
NRC would require that design and analysis information needed for the
Commission to make its safety determination be complete and available
for any application the NRC is reviewing. Making this explicit provides
increased clarity to future standard design approval applicants under
part 53.
Sections 53.1230 through 53.1263 address applications for, issuance
of, and other provisions related to standard DCs under part 53. Section
53.1230 addresses general provisions for standard DCs and is equivalent
to Sec. 52.41. Section 53.1236 addresses general information
requirements for the content of applications and is equivalent to Sec.
52.46. Section 53.1239 addresses requirements for the technical content
of applications and is equivalent to Sec. 52.47(a). The requirements
in Sec. 53.1239 have been modified from the analogous requirements in
Sec. 52.47(a) to align with the technical requirements in part 53.
Section 53.1241 addresses requirements for the content of a
standard DC application other than the
[[Page 15724]]
FSAR and is equivalent to Sec. 52.47(b) and (d).
Section 53.1242 addresses review of applications and is equivalent
to Sec. Sec. 52.48 and 52.51. Section 53.1242(c) includes a provision
that allows a DC applicant to reference applicable licensing-basis
information for an OL or COL issued under part 53. As explained
previously, this provision explicitly allows flexibility for an
applicant who may wish to license a first-of-a-kind reactor for
operation prior to seeking certification of the generic reactor design.
For NRC findings on a reactor design in an OL or COL proceeding, this
provision provides finality in a subsequent DC application that
references information on the OL or COL proceeding's docket. This
finality accorded to the OL or COL findings would bind the NRC staff
and the ACRS but would not bind members of the public or the
Commission. (To the extent an Atomic Safety and Licensing Board (ASLB)
might have a role in a DC rulemaking, the OL or COL findings would not
bind the ASLB either.) Specifically, members of the public would have
the opportunity to comment on a proposed DC rule under well-established
NRC practice. The rationale for binding the NRC staff and ACRS is
similar to the rationale for a COL applicant referencing a standard
design approval under part 52.
Section 53.1245 addresses referral to the ACRS and is equivalent to
Sec. 52.53. Section 53.1248 addresses issuance of standard DCs and is
equivalent to Sec. 52.54. Section 53.1251 addresses duration of
certifications and is equivalent to Sec. 52.55. Section 53.1254
addresses application for renewal and is equivalent to Sec. 52.57,
although the final rule removes the requirement to refer the renewal to
the ACRS consistent with current agency practice. Section 53.1257
addresses criteria for renewal and is equivalent to Sec. 52.59.
Section 53.1260 addresses duration of renewals and is equivalent to
Sec. 52.61. Section 53.1263 addresses finality of standard DCs and is
equivalent to Sec. 52.63.
Sections 53.1270 through 53.1291 address applications for, issuance
of, and other provisions related to MLs covering manufacturing
activities at one or more licensee facilities under part 53. Section
53.1270 addresses the scope of these sections and is equivalent to
Sec. 52.151.
Section 53.1276 addresses general information requirements for the
content of ML applications and is equivalent to Sec. 52.156, with one
exception. Section 53.1276 requires each application for an ML to also
include the information required by Sec. 53.1109(e). This information
includes 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 part 53.
Section 53.1279 addresses requirements for the technical content of
applications for MLs to be included in the FSAR and is equivalent to
Sec. 52.157. In addition, the requirements in Sec. 53.1279(a) and (b)
have been modified from the analogous requirements in Sec. 52.157 to
align with the technical requirements in part 53. Section 53.1279(a)(2)
outlines the required content of the application addressing design
information and states that the application must include design
information equivalent to that required for a standard DC. This
reference to the pertinent DC requirements is an efficiency that
prevents the need to repeat the same requirements for the content of an
ML application.
Section 53.1279(c) provides application requirements related to the
deployment of the completed manufactured reactor. Section 53.1279(c)(1)
requires 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 shipped items upon receipt at the
site. Section 53.1279(c)(2) requires 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 will ensure close
integration between the designer, contractors, and any licensee of a
facility in which the manufactured reactor is to be installed. Finally,
Sec. 53.1279(c)(3) requires that the application include a description
of the measures used for the control of interfaces between the holder
of the ML and the holder of the COL or CP for the commercial nuclear
plant at which the manufactured reactor is to be installed. This
information is 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.
Section 53.1279(d) includes additional requirements for application
content for applicants seeking an ML for manufactured reactors that
will be fueled at the factory under a part 70 license, consistent with
the requirements in Sec. 53.620(d). These provisions 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 can be successfully
transported, installed, and operated at a site for which the Commission
has issued a COL or CP and OL that authorizes construction and
operation of a commercial nuclear plant using the manufactured reactor.
Section 53.1282 provides requirements for other application content
for MLs and is equivalent to Sec. 52.158. Section 53.1282(a)(1)
provides requirements to include in the ML application the ITAAC within
the scope of the ML that the COL or CP holder referencing the ML must
satisfy. Section 53.1282(a)(2) requires that the ITAAC from a
referenced standard design apply to the portions of the ML design
within the scope of the referenced standard design. Section
53.1282(a)(3) states that a COL application may include a notification
that required referenced standard DC ITAAC have been satisfied at the
manufacturing facility.
Section 53.1282(b) requires an ML application to include an
environmental report and, consistent with existing requirements, Sec.
53.1282(b)(2) notes that if the ML application references a standard
DC, the environmental report need not contain a discussion of severe
accident mitigation design alternatives for the manufactured reactor as
used in a commercial nuclear plant.
Section 53.1285 provides standards for review of applications and
administrative review of applications for MLs, including hearings, and
is equivalent to Sec. Sec. 52.159 and 52.163.
Section 53.1286 addresses referral of applications to the ACRS and
is equivalent to Sec. 52.165. Section 53.1287 addresses issuance of an
ML and is equivalent to Sec. 52.167.
Section 53.1288 addresses finality of MLs and is equivalent to
Sec. 52.171, except that part 53 does not include the constraint that
the Commission may only grant a request for a departure from an ML for
an applicant who references or uses a manufactured reactor if special
circumstances outweigh any decrease in safety that may result from the
reduction in standardization caused by the departure. This is
consistent with the differences in the allowance for changes to a
manufactured reactor in part 53 (as noted in the discussion of
Sec. Sec. 53.1530 and 53.1550 in this
[[Page 15725]]
document) as compared to part 52. Section 53.1291 provides for a 40-
year duration for an ML, consistent with the duration provided for a DC
under Sec. 53.1251. Section 53.1293 addresses the transfer of MLs and
is equivalent to Sec. 52.175. Section 53.1295 addresses the renewal of
MLs and is equivalent to Sec. Sec. 52.177, 52.179 and 52.181, with
minor exceptions.
Section 53.1295(a)(3) states that an ML 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.
However, this provision omits a limitation from the equivalent
provision in Sec. 52.177 which prohibits the holder of an ML from
beginning the manufacture of a manufactured reactor less than 3 years
before the expiration of the license. This limitation was omitted in
part 53 because future reactor applicants may present smaller, simpler
designs, to include microreactor designs, in ML applications than those
that were envisioned when the existing requirements were written.
Eliminating the 3-year constraint in this provision will provide
greater flexibility for ML holders related to manufactured reactors
being produced close to the time when the ML expires. Additionally,
Sec. 53.1295(c) provides for a 40-year term for a renewed ML,
consistent with the term for an initial ML under Sec. 53.1291.
Finally, the final rule removes the requirement to refer the renewal to
the ACRS consistent with current agency practice.
Sections 53.1300 through 53.1348 address applications for, issuance
of, and other provisions related to CPs under part 53. Section 53.1300
sets out general requirements for CPs and is equivalent to Sec. 50.23.
Section 53.1306 addresses the general information requirements for the
content of applications for CPs and is similar to Sec. 50.33(f) and
(h). However, the requirements for demonstrating financial
qualification are different for part 53 applicants than the existing
requirements for applicants under part 50 or 52. The part 53
requirements do not include the existing requirements under part 50 or
52 for an applicant to provide information to demonstrate that it
``possesses or has reasonable assurance of obtaining'' the funds
necessary for construction and operation along with associated
financing details. Instead, part 53 replaces that requirement with a
requirement to provide information that demonstrates that the applicant
``appears to be financially qualified,'' similar to the standard used
in Sec. 70.23(a)(5).
Section 53.1309 addresses requirements for the technical content of
applications for CPs and includes the requirement to submit a
Preliminary Safety Analysis Report (PSAR) that describes the facility
and presents a preliminary safety analysis of the facility as a whole.
This is in contrast to an OL application, which is required to include
an FSAR that describes the facility and presents a final safety
analysis of the facility as a whole. Section 53.1309 is equivalent to
Sec. 52.17(a)(1)(iv) through (a)(1)(x) and 52.17(b), with two
exceptions. First, Sec. 53.1309 replaces the analysis of the dose
criteria required by Sec. 52.17(a)(1)(ix) with analysis to demonstrate
compliance with the safety criteria defined in Sec. Sec. 53.210 and
53.220. Second, Sec. 53.1309(a)(2) adds a requirement for a CP
application to include several categories of detailed design
information, although Sec. 53.1309(a)(2)(ii) allows certain
relaxations of this requirement in view of aspects of a design that may
not yet be fully developed. Section 53.1309 references the requirements
for the content of an ESP application to address application
requirements related to siting and references the requirements for the
content of a DC application to address application requirements related
to design of the commercial nuclear plant. Section 53.1309(a)(2)(ii)
addresses the treatment of preliminary design information and notes
that information provided in the application may include some aspects
of the design that are not fully developed. This provision requires
that the completed design, including any changes during construction,
be described in the FSAR in an application for an OL. This includes the
requirement for a description of the PRA, other SREs, or a combination
thereof required by Sec. 53.450(a) and its results. Probabilistic risk
assessments, other SREs, or a combination thereof developed for
commercial nuclear plants prior to construction are based on the design
and other information available at the time of the CP application. PRAs
performed in early design stages or prior to construction may be
inherently less detailed and may include projected information that
will be subsequently verified or revised when the plant is built.
Section 53.1309(a)(4) addresses preliminary description of the plans
for coping with emergencies.
Section 53.1312 addresses other application content for CPs.
Section 53.1312(a)(1) is equivalent to Sec. 52.80(b) but is adapted
for a CP application. Section 53.1312(a)(2) is equivalent to Sec.
52.80(c) but is adapted for a CP application. Section 53.1312(b)(1) is
equivalent to Sec. 52.79(b), (c), and (d) but is adapted for a CP
application. Section 53.1312(b)(2) is equivalent to portions of
Sec. Sec. 52.63(b)(1), 52.79(b)(1) through (b)(3), (c), and (d)(1) and
(d)(3), 52.80, and 52.93(b) but is adapted for a CP application.
Guidance for equivalent requirements in parts 50 and 52 is also
addressed in RG 1.206, ``Applications for Nuclear Power Plants,''
Revision 1, section C.1.7.
Section 53.1315 addresses standards for review of applications and
administrative review of applications, including hearings, and is
equivalent to Sec. Sec. 52.81 and 52.85 but is adapted for a CP
application.
Section 53.1318 addresses finality of NRC approvals, licenses, and
certifications referenced in a CP application and is equivalent to
Sec. 52.83(a) but is adapted for a CP application.
Section 53.1324 addresses referral to the ACRS and is equivalent to
Sec. 50.58(a) and to Sec. 52.87 but is adapted for a CP application.
Section 53.1327 addresses authorization to conduct LWA activities
and is equivalent to Sec. 52.91 but is adapted for a CP application.
Section 53.1327(a) is equivalent to Sec. 52.91(a) but is adapted for a
CP application. Section 53.1327(b) is equivalent to Sec. 52.91(b) but
is adapted for a CP application. Section 53.1330 addresses exemptions,
departures, and variances for CP applicants.
Section 53.1333 addresses issuance of CPs. Section 53.1333(a) is
equivalent to Sec. 50.35(a). Section 53.1333(b) is equivalent to Sec.
50.35(b) and to Sec. 52.97(c) but is adapted for a CP application.
Section 53.1336 addresses the effect of CPs and is equivalent to Sec.
50.35(b). Section 53.1342 addresses the duration of CPs. Section
53.1342(a) is equivalent to Sec. 50.55(a). Section 53.1342(b) is
equivalent to Sec. 50.55(b). Section 53.1345 addresses the transfer,
assignment, and disposal of CPs and is equivalent to Sec. 50.80.
Section 53.1348 addresses the termination of CPs and is equivalent to
Sec. Sec. 52.3(b)(8) and 52.110(a)(1) but is adapted for a CP
application.
Sections 53.1360 through 53.1405 address applications for, issuance
of and other provisions related to OLs under part 53.
Section 53.1366 addresses requirements for the general content of
applications for OLs. It refers to general content requirements in
Sec. 53.1109 and requires supplemental information. Section 53.1366 is
similar to Sec. 50.33(f). However, the requirements for demonstrating
financial qualification
[[Page 15726]]
are different for part 53 applicants than the existing requirements for
applicants under part 50 or 52. The part 53 requirements do not include
the existing requirements for an applicant to provide information to
demonstrate that it ``possesses or has reasonable assurance of
obtaining'' the funds necessary for construction and operation along
with associated financing details. Instead, part 53 replaces that
requirement with a requirement to provide information that demonstrates
that the applicant ``appears to be financially qualified,'' similar to
the standard used in Sec. 70.23(a)(5).
Section 53.1369 provides requirements for the technical content of
applications for OLs to be included in the FSAR and is equivalent to
Sec. 50.34(b) but has been modified to align with the technical
requirements in part 53. It requires that the FSAR include and, as
needed, update information provided in the PSAR that was submitted and
reviewed to support the associated CP application.
Similar to the requirements for the content of CP applications,
Sec. 53.1369(a) references the requirements for the content of an ESP
application to address application requirements related to the site.
Section 53.1369(b) references the requirements for the content of a DC
application to address some of the application requirements related to
design of the commercial nuclear plant.
Section 53.1369(d) requires a description of the Integrity
Assessment Program that is required by Sec. 53.870. Section 53.1369(e)
is equivalent to Sec. 50.34(e). Section 53.1369(g) provides
requirements for OL application content to support Sec. 53.730 related
to the role of personnel in the operation of the commercial nuclear
plant and is adapted from requirements in part 55 and Sec. 50.34(f).
Likewise, Sec. 53.1369(h) provides requirements for OL application
content related to training programs to support Sec. Sec. 53.730(g)
and 53.830 and includes requirements equivalent to Sec. 50.34(b)(8),
Sec. 52.79(a)(33), and part 55. Section 53.1369(i) provides
requirements for OL application content related to emergency plans to
support Sec. 53.855 and is equivalent to Sec. 50.34(b)(6)(v).
Section 53.1369(j) provides requirements for OL application content
related to the applicant's organizational structure and is equivalent
to Sec. 50.34(b)(6)(i). Section 53.1369(k) provides requirements for
OL application content related to the applicant's proposed maintenance
program to support Sec. 53.715 and is equivalent to Sec.
50.34(b)(6)(iv). Section 53.1369(l) provides requirements for OL
application content related to the applicant's quality assurance
program to support Sec. 53.865 and is equivalent to Sec.
50.34(b)(6)(ii). Section 53.1369(m) provides requirements for OL
application content related to the applicant's proposed radiation
protection program to support Sec. 53.850 and is equivalent to Sec.
50.34(b)(3).
Sections 53.1369(n) through (p) provide requirements for OL
application content related to the applicant's proposed physical
security program to support Sec. 53.860(a) and are equivalent to Sec.
50.34(c) and (d). Section 53.1369(q) provides requirements for OL
application content related to the applicant's proposed cybersecurity
plan to support Sec. 53.860(d) and is equivalent to Sec. Sec.
52.79(a)(36)(iv) and 73.54. Section 53.1369(r) provides requirements
for OL application content related to the implementation of proposed
security, safeguards, and cybersecurity plans to support Sec. 53.860
and is equivalent to Sec. 52.79(a)(35)(ii) and 52.79(a)(36)(iv) and
(v).
Section 53.1369(s) provides requirements for OL application content
related to the applicant's proposed fire protection program to support
Sec. 53.875 and is equivalent to Sec. 52.79(a)(40). Section
53.1369(t) provides requirements for OL application content related to
the applicant's proposed ISI and IST program to support Sec. 53.880
and is equivalent to part of Sec. 52.79(a)(11). Section 53.1369(w)
provides requirements for OL application content related to the
applicant's general employee training program to support Sec. 53.830
and is equivalent to Sec. 52.79(a)(33). Section 53.1369(x) provides
requirements for OL application content related to the applicant's FFD
program to support part 26 and is equivalent to Sec. 52.79(a)(44).
Section 53.1369(y) provides requirements for OL applicants' programs to
demonstrate that any safety questions identified at the CP stage have
been resolved and is equivalent to Sec. 50.34(b)(5). Section
53.1369(z) provides requirements for OL applicants to describe how the
performance of each safety design feature has been demonstrated capable
of fulfilling functional design criteria considering interdependent
effects through either analysis, appropriate test programs, prototype
testing, operating experience, or a combination thereof to support
Sec. 53.440(a). It is largely equivalent to Sec. Sec. 50.34(b)(5) and
50.43(e). Section 53.1369(aa) provides requirements for OL application
content related to the applicant's proposed TS to support Sec.
53.710(a) and is equivalent to Sec. 50.34(b)(6)(vi).
Section 53.1372 addresses requirements for the content of OL
applications other than the FSAR. Section 53.1372(a) requires
submission of an environmental report and is equivalent to Sec.
50.30(f) and Sec. 51.53(b). Section 53.1372(b) does not have a direct
parallel in parts 50 and 52 and requires the inclusion of a description
of availability controls that are not included in the FSAR to support
Sec. 53.710(b).
Section 53.1375 addresses standards for review of OL applications
and the administrative review of applications, including hearings, and
is equivalent to Sec. Sec. 52.81 and 52.85, except that the NRC has
omitted 10 CFR part 54, ``Requirements for Renewal of Operating
Licenses for Nuclear Power Plants,'' from the list of standards in
Sec. 53.1375(a). Part 53 does not include detailed requirements
related to renewal of licenses, although a general provision and
possible placeholder for future requirements has been included as Sec.
53.1595. The NRC will decide after the part 53 final rule is published
whether this section will be retained in part 53 to address license
renewal or whether the agency will take another approach to address
license renewal for part 53 licensees, such as amending part 54 to
address part 53 licensees.
Section 53.1381 addresses referral to the ACRS and is equivalent to
Sec. Sec. 50.58 and 52.87. Section 53.1384 addresses exemptions,
departures, and variances for OL applicants. Section 53.1384(a) is
equivalent to Sec. 52.93 but is adapted for OLs. Section 53.1384(b) is
equivalent to Sec. Sec. 52.39(d) (with respect to ESPs) and 52.93 but
is adapted for OLs.
Section 53.1387 addresses issuance of OLs. The introductory
paragraph is equivalent to Sec. 50.56. Section 53.1387(a)(1)(i) is
equivalent to Sec. Sec. 50.50 and 50.57(a)(1). Section
53.1387(a)(1)(ii) is equivalent to Sec. 50.50. Section
53.1387(a)(1)(iii) is equivalent to Sec. 50.57(a)(2). Section
53.1387(a)(1)(iv) is equivalent to Sec. 50.57(a)(3). Section
53.1387(a)(1)(v) is equivalent to Sec. 50.57(a)(4). Section
53.1387(a)(1)(vi) is equivalent to Sec. 50.57(a)(6). Section
53.1387(a)(1)(vii) is equivalent to Sec. 50.57(a)(5). Section
53.1387(a)(1)(viii) is equivalent to Sec. 52.97(a)(1)(vi) but is
adapted for OLs. Section 53.1387(c) is equivalent to Sec. 50.57(b).
Section 53.1387(d) is equivalent to Sec. Sec. 50.36(b) and 50.50.
Section 53.1390 addresses backfitting of OLs and is equivalent to
Sec. 52.98(a) but adapted for an OL application. Section 53.1396
addresses duration of an OL and is equivalent to Sec. 50.51(a) and
Sec. 52.104. Section 53.1399 addresses transfer, assignment, and other
[[Page 15727]]
disposition of an OL and is equivalent to Sec. 50.80. Section 53.1402
addresses applications for renewal of an OL and refers to Sec.
53.1595. Section 53.1405 addresses continuation of an OL and is
equivalent to Sec. 52.109 but is adapted to address an OL.
Sections 53.1410 through 53.1461 address applications for, issuance
of, and other provisions related to COLs under part 53. Section 53.1410
describes the contents of Sec. Sec. 53.1410 through 53.1461 and is
equivalent to Sec. 52.71. Section 53.1413 addresses general
information requirements for the content of applications for COLs and
is equivalent to Sec. 52.77, which references Sec. 50.33. Most of the
provisions from Sec. 50.33 applicable to COLs are restated in Sec.
53.1109. However, the requirements for demonstrating financial
qualification are different for part 53 applicants than the existing
requirements for applicants under part 50 or 52. The part 53
requirements do not include the existing requirements under part 50 or
52 for an applicant to provide information to demonstrate that it
``possesses or has reasonable assurance of obtaining'' the funds
necessary for construction and operation along with associated
financing details. Instead, part 53 replaces that requirement with a
requirement to provide information that demonstrates that the applicant
``appears to be financially qualified,'' similar to the standard used
in Sec. 70.23(a)(5).
Section 53.1416 addresses the technical content to be included in
an FSAR for an application for a COL and is equivalent to Sec. 52.79
except as modified to reflect the technical requirements in part 53 and
with one addition. Section 53.1416 includes the statement that the
Commission will require, before issuance of a COL, that information
normally contained in 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. This
statement is equivalent to DC application requirements in Sec. 52.47
and is included in Sec. 53.1416 for clarity.
Similar to the requirements for the content of OL applications,
Sec. 53.1416(a)(1) references the requirements for the content of an
ESP application to address application requirements related to siting.
Section 53.1416(a)(2) references the requirements for the content of a
DC application to address some of the application requirements related
to design of the commercial nuclear plant. The remaining items under
Sec. 53.1416(a) are likewise similar to the required content for OL
applications under Sec. 53.1369(a). Section 53.1416(b) requires COL
applicants to provide a report documenting the resolution of any safety
questions for SSCs for which research and development was necessary to
confirm the adequacy of their design and is equivalent to Sec.
50.34(b)(5). Section 53.1416(c) provides requirements for COL
applicants to describe how the performance of each safety design
feature has been demonstrated capable of fulfilling functional design
criteria considering interdependent effects through either analysis,
appropriate test programs, prototype testing, operating experience, or
a combination thereof to support Sec. 53.440(a). It is largely
equivalent to Sec. Sec. 52.79(a)(24) and 50.43(e). Section 53.1416(d)
addresses the content of COL applications referencing an ESP. Section
53.1416(e) addresses the content of COL applications referencing a
standard design approval. Section 53.1416(f) addresses the content of
COL applications referencing a standard DC. Section 53.1416(g)
addresses the content of COL applications referencing an ML.
Section 53.1419 addresses other application content for COLs and is
equivalent to Sec. 52.80. Section 53.1419(a)(2) is new and requires
the inclusion of a description of availability controls that are not
required to be included in the FSAR.
Section 53.1422 addresses standards for review of applications and
the administrative review of applications, including hearings, and is
equivalent to Sec. Sec. 52.81 and 52.85. The NRC has removed part 54
from the list of standards in Sec. 53.1422(a). Part 53 does not
include requirements related to renewal of licenses, in relation to
Sec. Sec. 53.1422 and 53.1595.
Section 53.1425 addresses the finality of NRC approvals referenced
in a COL application and is equivalent to Sec. 52.83(a). Section
53.1431 addresses the referral of COL applications to the ACRS for
review and is equivalent to Sec. 52.87. Section 53.1434 addresses the
authorization to conduct LWA activities and is equivalent to Sec.
52.91. Section 53.1437 addresses exemptions, departures, and variances
and is equivalent to Sec. 52.93. Section 53.1440 addresses issuance of
COLs and is equivalent to Sec. 52.97. Section 53.1443 addresses
finality of COLs and is equivalent to Sec. 52.98.
Section 53.1449 addresses inspection during construction and is
equivalent to Sec. 52.99. Section 53.1452 addresses operation under a
COL and is equivalent to Sec. 52.103. Paragraph Sec. 53.1452(a)
includes footnotes to provide that, for licensees installing fueled
manufactured reactors under a COL, (1) the COL holder will notify the
NRC of its scheduled date for initiating the removal of the features to
prevent criticality required under Sec. 53.620(d)(1) rather than its
scheduled date for the initial loading of fuel, and (2) the NRC will
time its publication of the notice of intended operation based on the
COL holder's schedule for initiating the removal of the features to
prevent criticality required under Sec. 53.620(d)(1) rather than the
COL holder's scheduled date for the initial loading of fuel. These
footnotes are consistent with the provisions of Sec. 53.620(d)(1)(iv),
which state that, upon initiating the removal of the features to
prevent criticality in the manufactured reactor's place of operation,
the fueled manufactured reactor has commenced operation. For reactors
without the features to prevent criticality under Sec. 53.620(d)(1),
operation begins with initial fuel load. In both cases, removal of the
physical features to prevent criticality (for reactors with such
features) and initial fuel load (for reactors without such features)
put a fully constructed utilization facility in a position to sustain a
nuclear chain reaction, and in both cases, the utilization facility
cannot sustain a nuclear chain reaction (for lack of sufficient
reactivity) until the action takes place. Therefore, the NRC believes
that initiating the removal of the features to prevent criticality is
the best analogue to initial loading of fuel for reactors without such
features.
The footnote in Sec. 53.1452(a) regarding timing of the notice of
intended operation for fueled manufactured reactors with features to
prevent criticality also addresses the requirements of section
189a.(1)(B)(i) of the AEA. This section requires, in part, that ``[n]ot
less than 180 days before the date scheduled for initial loading of
fuel into a plant by a licensee that has been issued a combined
construction permit and operating license under section 185b., the
Commission shall publish in the Federal Register notice of intended
operation.'' That section further requires that this notice provide a
60-day period in which to request a hearing ``on whether the facility
as constructed complies, or on completion will comply, with the
acceptance criteria of the license.'' In the case where a fueled
manufactured reactor arrives at the site where it is to be operated by
a COL holder, the manufacturer would have
[[Page 15728]]
loaded fuel at the factory under its part 70 license. Therefore, at the
site of operation, there would not be ``initial loading of fuel into a
plant by a licensee that has been issued a combined construction permit
and operating license'' (emphasis added). Under a literal reading of
the entry condition in Act section 189a.(1)(B)(i), this situation would
not trigger its requirements. However, the purpose of the provision is
to offer the hearing opportunity at least 180 days prior to when the
fuel is loaded and ready for use at its authorized location. It would
be contrary to that purpose if, in this situation, the Commission did
not publish the notice of intended operation and opportunity for the
public to request a hearing on conformance with the acceptance criteria
in the COL for the site of operation. To fulfill the underlying purpose
of the law, the NRC is timing the notice of intended operation based on
the COL holder's schedule for initiating the removal of the features to
prevent criticality required under Sec. 53.620(d)(1). This action by
the COL holder is the best analogue to initial fuel load by the COL
holder for the reasons stated previously. This analogue is adopted in
other sections of part 53 and related sections in parts 50 and 73 that
use initial fuel loading to identify a transition point for the
applicability of regulatory requirements. To address the possible
loading of fuel into a manufactured reactor for subsequent transport to
and use at a commercial nuclear plant, multiple sections that determine
the applicability of regulations have been developed or revised to
allow for either initial fuel load or initiating the removal of the
features to prevent criticality required under Sec. 53.620(d)(1) for a
fueled manufactured reactor to determine the applicability of the
requirement, as appropriate.
Section 53.1455 addresses duration of COL and is equivalent to
Sec. 52.104. Section 53.1456 addresses the transfer of a COL and is
equivalent to Sec. 52.105. Section 53.1458 addresses application for
renewal and is equivalent to Sec. 52.107. Section 53.1461 addresses
continuation of COL and is equivalent to Sec. 52.109.
Section 53.1470 addresses standardization of commercial nuclear
plant designs and licenses to construct and operate commercial power
reactors of identical design at multiple sites and is equivalent to
appendix N to part 52, with one exception. Paragraph Sec. 53.1470(b)
provides flexibility regarding the timing of the applications to be
treated together under Sec. 53.1470. Each application can either list
the other applications or specify that such other applications will be
submitted to the NRC within 12 months of submittal of the first
application. This section sets out the particular requirements and
provisions applicable to situations in which applications for CPs and
subsequent OLs, or COLs, under this part are filed by one or more
applicants for licenses to construct and operate nuclear power reactors
of identical design (``common design'') to be located at multiple
sites. Additional information related to this section is provided in
the final rule to revise part 52 (72 FR 49352; August 28, 2007).
Subpart I--Maintaining and Revising Licensing-Basis Information
Part 53 establishes requirements for the maintenance of licensing-
basis information in subpart I.
Section 53.1500 describes the purpose of the subpart in terms of
the definition of licensing-basis information in subpart A. Subpart I
is closely tied to the requirements in subpart H, which provides the
requirements for contents of applications for the various types of
licenses issued under part 53. Subpart I is generally organized into
sections dealing with: (1) licensing-basis information that licensees
are not authorized to change without NRC approval (e.g., licenses,
regulations); and (2) licensing-basis documents that licensees may
change provided specified criteria are satisfied (e.g., FSAR, program
descriptions). The subpart also captures certain general conditions on
licenses and changes to the licenses related to the transfer and
termination of licenses.
Section 53.1502 defines specific terms and conditions of licenses.
These terms and conditions are equivalent to the regulations in: (1)
Sec. 50.54(h) stating that each license is subject to the provisions
of the AEA and requirements issued by the Commission; (2) Sec.
50.54(s) stating the actions the Commission will take if it makes a
finding that there is not reasonable assurance that adequate protective
measures can and will be taken in the event of a radiological
emergency; (3) Sec. 50.54(aa) stating that each license is subject to
the specified sections of the Federal Water Pollution Control Act; and
(4) Sec. 50.54(dd) stating that a holder of an OL or COL may take
reasonable actions that depart from the license in a national security
emergency.
Section 53.1505(a) serves as an introduction to and overview of the
sections that follow on changes to licensing-basis information
requiring prior NRC approval, namely the elements of licensing-basis
information defined by licenses, orders, and regulations. The related
sections within these subparts primarily deal with the process of how a
licensee requests and the NRC issues an amendment to a license or
issues an order that modifies a license. Another important element of
licensing-basis information that a part 53 licensee is not able to
change or deviate from without NRC approval is the NRC regulations
themselves. Section 53.1505(b) refers to Sec. 53.080 in subpart A that
provides the criteria for a licensee or other party to satisfy when
requesting an exemption from NRC regulations.
Section 53.1510 is equivalent to Sec. 50.90 and requires that a
licensee submit an application to request an amendment to a license.
The required assessments that are included within an application to
amend a license under part 53 must address the safety criteria and
analysis requirements of subparts B and C. As with parts 50 and 52,
licensees must include in their applications an analysis of whether the
amendment involves no significant hazards consideration using the
standards in Sec. 53.1520, which are equivalent to the standards in
Sec. 50.92. Although this rulemaking provided an opportunity to revise
the terminology related to no significant hazards consideration
determinations, which dates to the early 1960s when applications were
supported by final hazard summary reports, the NRC is maintaining the
same terminology used in part 50 to minimize the need for associated
changes in other regulations, guidance, and public notices.
Section 53.1515 establishes requirements for public notices and
state consultations associated with the NRC's processing of a license
amendment request. This section is equivalent to Sec. 50.91 for the
NRC's processes related to applications to amend an OL or COL. Section
50.91(b) stipulates that the Commission will make available to the
licensee the name of the appropriate State official designated to
receive such amendments. While the Commission intends to continue
following this practice, the Commission has not included this
administrative matter in part 53. Section 53.1515(b)(3) contains some
modifications compared to Sec. 50.91(b)(3) for clarity; these
revisions are not intended to revise the substance of the provisions in
part 53 compared to part 50.
Section 53.1520 is based on Sec. 50.92. The section continues to
use the criteria in Sec. 50.92 for determining that a proposed
amendment involves no significant hazards consideration.
[[Page 15729]]
Although more specific terms such as event sequence are used throughout
part 53, Sec. 53.1520 uses the term ``accident'' to maintain
consistency with the long history of making no significant hazards
consideration determinations under part 50.
Section 53.1525 provides requirements for holders of an OL or COL
requesting to revise information from a DC rule that was referenced in
the initial license application and included in or incorporated by
reference into the facility FSAR. In keeping with the current
requirements in part 52, the portion of the part 53 facility licensing-
basis information obtained from the certified design is divided into
two categories. The most significant design information and the ITAAC
are certified by rule and designated as ``certification information.''
The remaining information, which makes up the majority of the design
information approved as part of the DC, is not certified by rule and is
not considered ``certification information.'' Part 52 refers to these
categories of information as Tier 1 and Tier 2 information,
respectively, and refers to a change made to that information on a
plant-specific basis as a departure. Under part 52, a departure from
Tier 1 information requires an exemption and, for information
incorporated into the license, a license amendment.
Part 53 dispenses with the Tier 1 and Tier 2 terminology. Rather,
Sec. 53.1525 uses the term ``certification information'' in place of
Tier 1, and a plant-specific departure from the certification
information requires both a request for an exemption from the
associated DC rule and, for information such as ITAAC incorporated into
the license, a license amendment. However, as provided in Sec.
53.1525(c), a plant-specific departure from the information approved by
the NRC as part of the DC rule but that is not certification
information (i.e., Tier 2 information under part 52) will be assessed
using the process and criteria defined in Sec. 53.1550 for changes to
a FSAR. An applicant or licensee must identify such a change as a
departure from the referenced standard design in the updated FSAR. The
process for making a generic change to a certified design is described
in the associated section in subpart H.
Section 53.1530 allows the holder of an ML to make changes to the
design of the manufactured reactor or procedures as described in a FSAR
associated with the ML without requesting a license amendment from the
NRC if the change satisfies the criteria in Sec. 53.1550(a)(1) and
(2). This section is different than the provisions in Sec. Sec. 52.98
and 52.171 that do not allow any changes to the design of a
manufactured reactor without requesting a license amendment. A COL or
OL holder who references or uses a manufactured reactor may make
changes to the facility or procedures described in an FSAR, including
those portions incorporated by reference from the FSAR associated with
the ML, using Sec. 53.1550 to determine if a license amendment is
required.
Section 53.1535 establishes requirements for license amendments
during construction. The section provides the equivalent options and
requirements for the holders of a CP as those in Sec. 50.35(b). The
regulations allow but do not require the holder of a CP or LWA to
request an amendment under Sec. 53.1510 if the licensee desires to
obtain NRC approval of a specific design feature or specification. The
requirements for obtaining an amendment to a COL to address changes
during construction are also provided in Sec. 53.1535. The process
differs from the current requirements in part 52 by adopting a
requirement that explicitly supports a change process like that
described in RG 1.237, ``Guidance for Changes During Construction for
New Nuclear Power Plants Being Constructed Under a Combined License
Referencing a Certified Design Under 10 CFR part 52.'' Section 53.1535
allows the holder of a COL to proceed at its own risk in making a
change during the construction process and requires that licensee to
submit a license amendment request no later than 45 days from the date
the licensee begins to implement the change or departure requiring NRC
approval.
Section 53.1540 serves as an introduction to the sections that
follow on changes to licensing-basis information that are primarily
under the control of a licensee but for which evaluations are made to
determine if a submittal to the NRC requesting approval is required.
The section also includes definitions that are applicable when using
the processes in Sec. Sec. 53.1545 through 53.1565. The definitions
are largely equivalent to those in Sec. 50.59(a) but include some
revision to reflect the structure and terminology in other subparts in
part 53. For example, the definition of ``Change'' in Sec. 53.1540(b)
addresses a ``design feature or related functional design criteria''
rather than a ``design function,'' because the former are defined terms
in part 53. Similarly, in Sec. 53.1540(b), the phrase ``design bases''
from Sec. 50.59(a)(2) is replaced with functional design criteria for
SR SSCs.
Section 53.1545 provides the requirements for updating of FSARs.
While the process-related requirements under Sec. 53.1545 are largely
the same as those in Sec. 50.71, the specifics of information to be
updated differ due to the role of PRA, other SREs, or a combination
thereof in satisfying the requirements in subparts B and C.
Additionally, the use of the risk-informed approach in subpart C will
result in some but not all information related to PRA, other SREs, or a
combination thereof being in the FSAR or another licensing-basis
document and therefore a separate update requirement for PRA, other
SREs, or a combination thereof similar to Sec. 50.71(h) is not
included in subpart I. Section 53.1545(e) addresses updating of FSARs
associated with MLs and includes periodic updates to reflect the added
flexibility for ML holders to revise the FSAR associated with the ML
under Sec. 53.1530.
Section 53.1239(a)(18) in subpart H and the related references to
this requirement for the holders of OLs and COLs requires a description
of the PRA, other SREs, or a combination thereof required by Sec.
53.450(a) and its results to be included in FSARs. However, guidance
documents are planned to clarify the division of information related to
PRA, other SREs, or a combination thereof that must be in the FSAR, in
other possible licensing-basis documents, and controlled as plant
records subject to inspections and audits. At a minimum, the
information from the PRA, other SREs, or a combination thereof that is
needed to show compliance with subpart C will be included in the FSAR
(e.g., summary of PRA, other SREs, or a combination thereof and
analytical results for LBEs). The submittal of voluminous PRA
information was initially required under part 52, but that proved to be
impractical and was revised in the 2007 revision of part 52. Guidance
is being developed to ensure sufficient information is submitted to the
NRC to support the licensing process and the NRC's regulatory findings
under part 53 or similar applications using the LMP methodology under
parts 50 or 52.
Section 53.1545(a)(3) and (4) are based on the inclusion of at
least a summary of the results of the PRA, other SREs, or a combination
thereof and the related margins to safety criteria in the FSAR and
require updates to that information. The routine reporting of these
margins also informs application of the criteria for allowing changes
without an amendment in the following section (Sec. 53.1550) in
subpart I.
[[Page 15730]]
Section 53.1550 establishes requirements for evaluating changes to
a facility as described in its FSAR. This section provides the
equivalent of the requirements in Sec. 50.59 for evaluating changes to
an FSAR (as updated) and determining if a license amendment is required
to implement a change to a facility or procedures. The evaluation
criteria in Sec. 53.1550 reflect the role of the PRA, other SREs, or a
combination thereof in the safety analyses under part 53 and include
several measures related to the changes in plant risk resulting from a
change in the plant design or plant procedures. Examples include
criteria that rely on the identification of risk-significant event
sequences in accordance with the analysis requirements of Sec. 53.450;
exceeding the LBE evaluation criteria as defined in Sec. 53.450; the
consideration of potential changes in estimated comprehensive risk
metrics that exceed the associated risk performance objectives in the
safety criteria in Sec. 53.220; changes to the safety classification
of SSCs; and consideration of reductions in defense in depth.
Section 53.1550 includes certain concepts taken from existing
guidance for Sec. 50.59 in the criteria related to DBAs and aligns
with recently developed industry guidance in NEI 22-05, Revision 0,
``Technology Inclusive Risk Informed Change Evaluation (TIRICE).''
Specifically, criterion (iv) for changes made to a method of evaluation
of DBAs under Sec. 53.450(f) is equivalent to a change in a method of
evaluation under Sec. 50.59, and criterion (viii) on assessing if a
change creates a possibility for an accident of a different type than
previously analyzed in the FSAR is similar to the Sec. 50.59 criterion
(v). Criterion (v) in Sec. 53.1550 differs from the corresponding
criterion in NEI 22-05 in that it does not include changes to safety
classification of SSCs from non-safety related or NSRSS to SR because
plant changes introducing new SR SSCs would require a change to
technical specifications under Sec. 53.710. Guidance documents will be
prepared to address the content of applications for information related
to PRA, other SREs, or a combination thereof under part 53, and this
guidance will also influence how potential changes in the evaluation of
LBEs other than DBAs analyzed under Sec. 53.450(e) are evaluated and
reported under criterion (iv).
Sections 53.1560 through 53.1565 in subpart I define the processes
for a licensee to evaluate changes to the program documents included in
the licensing-basis information submitted to the NRC and to modify such
programs without NRC prior approval.
Section 53.1560 includes the requirements for updating program
documents included in licensing-basis information and provides the
equivalent of FSAR updates for key program documents. The requirements
in these sections provide a uniform approach for updating program
documents, which correspond to the programs required under subpart F.
Section 53.1565 provides a process for licensees to make changes to
program documents included in licensing-basis information without
obtaining prior NRC approval. The requirements include several generic
criteria that, if not satisfied, will prompt the need for NRC approval
of a change to a program document. These generic criteria include
whether a change will comply with TS and NRC regulations. Another
criterion for evaluating changes to program documents is conforming
with program-specific requirements, including NRC-approved program
documents with more specific criteria for a particular program,
regulations, administrative controls sections of TS, and NRC-approved
program documents.
Section 53.1565(d) includes specific criteria for evaluating
changes to several program documents that have well established change
processes and guidance for licensees under parts 50 and 52. The program
documents specifically addressed in the section include quality
assurance programs that are equivalent to Sec. 50.54(a), an emergency
preparedness program that is equivalent to Sec. 50.54(q), and the
security program that is equivalent to Sec. 50.54(p).
Section 53.1570 establishes requirements for the transfer of
commercial nuclear plant licenses by providing the equivalent
requirements of Sec. 50.80 for the possible transfer of an ESP, CP,
OL, or COL. Likewise, Sec. 53.1575 establishes requirements for the
termination of an OL or COL by providing the equivalent requirements of
Sec. 50.82. Other requirements related to decommissioning and license
termination are included in subpart G.
Section 53.1580 establishes requirements for information requests
the NRC may send to the various types of licensees and provides
requirements that are equivalent to requirements in Sec. 50.54(f).
Section 53.1585 provides the requirements that are equivalent to
requirements in Sec. 50.100 to address revocation, suspension,
modification of licenses, and approvals for cause. Section 53.1590
addresses backfitting requirements by providing requirements that are
similar to those in Sec. 50.109.
Section 53.1595 addresses license renewals under part 53 with a
simple statement that licenses may be renewed. This section may be
expanded through future rulemakings to more fully describe or reference
the processes related to requesting and processing applications to
renew ESPs, OLs, and COLs issued under part 53.
Subpart J--Reporting and Other Administrative Requirements
Part 53 addresses various reporting and administrative requirements
in subpart J.
Section 53.1600 explains the organization of the various sections
within the subpart related to providing unfettered access to NRC
inspectors; maintaining certain records and reporting specified events
or conditions; demonstrating compliance with financial qualification
requirements and providing specified financial reports; and maintaining
financial protections to address potential accidents.
Section 53.1610 establishes requirements for the provision of
facilities and unfettered access for inspections. These requirements
are equivalent to Sec. 50.70 with only minor changes to provide
additional flexibilities and address possible differences related to
reactors licensed under part 53 and the possibility that some
commercial nuclear plants may not be assigned resident inspectors.
Section 53.1620 provides for maintenance of records and the making
of various reports to the NRC. These requirements are largely
equivalent to Sec. 50.71. This section is not intended to reflect all
provisions in Sec. 50.71; several important requirements in Sec.
50.71 are captured in other sections of part 53. For example, Sec.
53.1545 within subpart I provides requirements that are equivalent to
Sec. 50.71(e) for updating FSARs. A reporting requirement related to
completion of power ascension testing is added to Sec. 53.1620 to
support the assessment of annual fees under 10 CFR part 171, ``Annual
Fees for Reactor Licenses and Materials Licenses, Including Holders of
Certificates of Compliance, Registrations, and Quality Assurance
Program Approvals and Government Agencies Licensed by the NRC,'' which
normally commence upon completion of those testing activities.
Section 53.1630 establishes requirements for immediate notification
requirements for operating commercial nuclear plants. These
requirements are equivalent to Sec. 50.72 with minor changes to make
the reporting criteria technology-inclusive. In addition, a new version
of NRC Form 361 (NRC Form 361S) has been created for use by part 53
licensees, but without LWR-specific terminology to ensure technology-
inclusiveness. The requirements in
[[Page 15731]]
Sec. 53.1630 and the new NRC Form 361S are consistent with changes to
Sec. 50.72 proposed to the Commission in SECY-24-0049, ``Proposed
Rule: Reporting Requirements for Nonemergency Events at Nuclear Power
Plants (RIN 3150-AK71; NRC-2020-0036),'' to eliminate those
nonemergency event reporting criteria that are not important to safety,
do not require prompt action from the NRC, or can be tracked using
other existing agency processes.
Section 53.1640 addresses the licensee event report system. These
requirements are equivalent to Sec. 50.73 with minor changes to make
the requirements inclusive of various reactor technologies and to
reflect appropriate internal references to other sections in part 53.
In addition, NRC Forms 366, 366A, and 366B are revised to include
corresponding check boxes for part 53 licensees.
Section 53.1645 requires periodic reporting of the quantity of
radionuclides released to unrestricted areas in liquid and gaseous
effluents, doses to members of the public, and the results of
environmental monitoring. These reporting requirements in part 53 are
largely equivalent to those in the TSs required by Sec. 50.36a,
``Technical specifications on effluents from nuclear power reactors.''
The section also includes an equivalent to the reporting requirement in
section IV of appendix I to part 50 if the radiation exposure to a
member of the public in any calendar quarter exceeds one-half of the
annual design objective.
Section 53.1650 includes a reporting requirement to support
safeguards agreements between the United States and the International
Atomic Energy Agency (IAEA) and is equivalent to Sec. 50.78.
Sections 53.1660 through 53.1700 address financial requirements and
are somewhat different than existing regulations in parts 50 and 52.
The part 53 requirements do not include the existing part 50 and 52
requirement for an applicant to demonstrate that it ``possesses or has
reasonable assurance of obtaining'' the funds necessary for
construction and operation. Instead, part 53 replaces that requirement
with an ``appears to be financially qualified'' standard similar to the
standard in Sec. 70.23(a)(5). Section 53.1670 is entitled ``Financial
qualifications'' and requires applicants other than electric utilities
to appear to be financially qualified for the activities for which the
license is being sought. The remaining financial reports in part 53 are
equivalent to Sec. 50.76 for a change of status, Sec. 50.54(cc) for
the filing of a petition for bankruptcy, and Sec. 50.81 for creditor
regulations. Part 53 does not contain a requirement for annual
financial reports equivalent to Sec. 50.71(b) because these reports
are not actively used by the NRC to assess a licensee but could be
accessed in the event that the NRC deems it necessary to look into a
licensee's financial situation due to events such as a declaration of
bankruptcy.
Sections 53.1710 through 53.1730 address financial protection
requirements. Section 53.1720 requires insurance to stabilize and
decontaminate a plant following an accident. These requirements are
taken from Sec. 50.54(w), with the only notable change being the
addition of a provision allowing plant-specific estimates of costs to
stabilize and decontaminate a plant as an alternative to the $1.06
billion minimum coverage in Sec. 50.54(w). An example of cost
estimations that, in part, provided a basis for the requirements in
Sec. 50.54(w) is provided in NUREG/CR-2601, ``Technology, Safety and
Costs of Decommissioning Reference Light Water Reactors Following
Postulated Accidents.'' Section 53.1730 is equivalent to Sec.
50.57(a)(5) and refers to the requirements in 10 CFR part 140,
``Financial Protection Requirements and Indemnity,'' related to
financial protection requirements and indemnity agreements, including
the financial protection requirements of the Price-Anderson Act.
Subpart M--Enforcement
Subpart M contains two provisions, Sec. 53.9000 and Sec. 53.9010,
which are analogous to provisions contained in other parts of 10 CFR
chapter I imposing requirements on regulated entities. Section 53.9000
provides notice of the Commission's authority under the AEA to obtain
injunctions or other court orders for the enumerated violations.
Paragraph Sec. 53.9010(a) provides notice to all persons and entities
subject to part 53 that they are subject to criminal sanctions for
willful violations, attempted violations, or conspiracy to violate
certain regulations under part 53. Criminal sanctions do not apply to
the regulations listed in paragraph (b). The regulations for which
criminal penalties do apply are limited to those that establish either
a regulatory obligation or prohibition.
IV. Changes to Other Parts of 10 CFR Chapter I
10 CFR Part 26
A. Introduction
Through this final rule, the NRC is establishing 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 part 53. The requirements
applicable to these applicants, licensees, and other entities are
commensurate with the radiological consequences presented by the
applicants' facilities and the operation of these facilities.\2\ The
FFD framework consists of a two-tiered graded approach similar to that
currently in part 26. This new FFD framework is established in subpart
M, ``Fitness-for-Duty Programs for Facilities Licensed Under Part 53,''
of part 26.
---------------------------------------------------------------------------
\2\ The NRC uses the term ``operation'' in its part 26
discussion to focus on human performance, namely the necessity of
individuals to operate, maintain, surveil, and protect the facility
and respond to operational transients and unlikely event sequences.
---------------------------------------------------------------------------
The NRC is using operating experience to provide regulatory
flexibility in the subpart M of 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 also helps in FFD program implementation because of the
wide variety of staff sizes anticipated at commercial nuclear plants
licensed under part 53 and the geographically remote locations in which
commercial nuclear plants may be sited.
The first-tier FFD program requirements apply to part 53 licensees
and other entities of commercial nuclear plants that demonstrate
compliance with Sec. 73.100(a)(1)(i), at their discretion, no later
than the start of construction activities; licensees and other entities
of facilities that do not demonstrate compliance with Sec.
73.100(a)(1)(i) no later than the start of construction activities; and
holders of MLs who are assembling or performing non-operational testing
of manufactured reactors. These requirements are provided in Sec.
26.605(a) and are essentially equivalent to those requirements in
subpart K, ``FFD Program for Construction,'' of part 26 but have been
supplemented by select requirements from subparts E, ``Collecting
Specimens for Testing,'' and I, ``Managing Fatigue,'' of part 26, and
the requirements in subparts A, ``Administrative Provisions,'' and O,
``Inspection, Violations, and Penalties,'' of part 26. The first-tier
requirements involve policies, procedures, behavioral observation,
fatigue management, drug and alcohol testing, determinations of
fitness, appeals, training, sanctions, auditing, change control,
evaluating FFD program performance, recordkeeping, and reporting. These
[[Page 15732]]
Sec. 26.605(a) FFD program requirements help deter individuals subject
to this section from impairment from any cause, including drug use,
alcohol misuse, and fatigue. These requirements also help licensees and
other entities identify individuals using impairing substances and
demonstrate compliance with Sec. 26.23, ``Performance objectives.''
The second tier includes all the first-tier requirements, plus the
more comprehensive set of FFD program requirements in current subparts
C, ``Granting and Maintaining Authorization,'' D, ``Management Actions
and Sanctions to be Imposed,'' H, ``Determining Fitness-for-Duty Policy
Violations and Determining Fitness,'' and N, ``Recordkeeping and
Reporting Requirements,'' of part 26. These requirements are provided
in Sec. 26.605(b) and are applicable to licensees and other entities
that demonstrate compliance with Sec. 73.100(a)(1)(i), if they do not
choose to comply with Sec. 26.605(a). These licensees and other
entities need to implement the FFD program before they begin
construction. Section 26.605(b) also applies to holders of
manufacturing licenses if they possess a separate license to load fuel
into a manufactured reactor. These licensees must implement their FFD
program no later than when they begin loading fuel into the reactor.
These requirements also apply to licensees or other entities that do
not demonstrate compliance with Sec. 73.100(a)(1)(i) that implement an
FFD program under subpart M of part 26 before the loading of fuel
onsite into a reactor vessel; before receiving a fueled manufactured
reactor; or before operating, testing, performing maintenance of, or
directing 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.
The second-tier requirements are based on the additional risk
presented by nuclear reactor assembly, testing, fueling, and operation
and the necessity for human actions in certain event sequences. The
inclusion of the current part 26 requirements aligns part 53 FFD and AA
program requirements with the current FFD and AA programs required for
facilities licensed under parts 50 and 52. This approach ensures
effective and consistent AA and FFD program implementation across the
commercial nuclear power industry, thereby ensuring uniform
requirements for individuals who may perform roles and responsibilities
for multiple facilities regardless of facility licensure.
Regarding fatigue management requirements, work hour controls are
required for personnel at utilization and manufacturing facilities in
accordance with the existing scoping criteria in Sec. 26.4, ``FFD
program applicability to categories of individuals,'' as revised in
this final rule. The amended Sec. 26.4 also will be used to determine
whether an individual is subject to drug and alcohol testing. The
applicability of these scoping criteria for certain individuals (such
as operators and maintenance personnel) will be determined by the
licensee or other entity through its risk-informed evaluation process
performed to assess the risk significance of the SSC upon which work is
being performed or directed by the individual. These requirements also
will be scaled based on the potential radiological consequences
presented by the facility, as determined by whether the facility
demonstrates compliance with Sec. 73.100(a)(1)(i). However, fatigue
management will be applied to all individuals subject to the FFD
program, similar to FFD program implementation by the current fleet of
commercial nuclear plants because fatigue management is a proactive
requirement designed to help prevent on-shift impairment through work
hour scheduling and time off. The behavioral observation program (BOP)
is the principal requirement to provide reasonable assurance that
individuals on shift are not mentally or physically impaired due to
fatigue, which in any way could adversely affect their ability to
safely and competently perform their duties.
This final rule establishes subpart M of part 26 for facilities
licensed under part 53, in lieu of subjecting all part 53 licensees to
the same part 26 requirements that apply to facilities licensed under
part 50 or 52, for four principal reasons. First, subpart M of part 26
applies FFD requirements in a risk-informed manner commensurate with
the radiological consequences presented by facilities licensed under
part 53 (i.e., whether a facility demonstrates compliance with Sec.
73.100(a)(1)(i)). This regulatory strategy is consistent with the
current part 26, which provides a comprehensive set of deterministic
requirements for licensees and other entities at facilities that are
operating. This approach is also consistent with the current subpart K
of part 26, which provides a more flexible framework for nuclear power
reactors under construction.
Second, subpart M of part 26 enables a part 53 licensee or other
entity 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 include drug testing of
oral fluid, urine, and hair specimens and non-invasive portal area
screening instruments that 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 53 commercial
nuclear plant be small and challenge the effective implementation of
the behavioral observation and drug and alcohol testing programs. Also,
a commercial nuclear plant that is sited at a geographically remote
location may 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 part 53 commercial 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 to
impact FFD program effectiveness. Particularly in isolated
environments, psychological phenomena known as ``groupthink'' may take
effect and could impact BOP effectiveness. For example, in
circumstances where small staffs are drawn from the same small town and
thereby have a potentially narrow experience base, it could be
challenging to maintain a work environment in which personnel feel free
to raise concerns without fear of retaliation, intimidation,
harassment, or discrimination, and organizations may resultingly
experience groupthink-like effects. Groupthink is particularly
prevalent among cohesive and insulated
[[Page 15733]]
groups that experience high levels of decisional stress.\3\ Small
staffs at part 53 commercial nuclear plants may therefore be more
susceptible to groupthink if they are working in an isolated
environment where decision-making pressures may be high.
---------------------------------------------------------------------------
\3\ See e.g., Irene W[aelig]r[oslash], Ragnar Rosness, and Stine
Skaufel Kilska, ``Human performance and safety in Arctic
environments,'' SINTEF (2018).
---------------------------------------------------------------------------
In small group dynamics, groupthink could have adverse effects on
team decision-making, as studies show that individuals will be more
hesitant to speak out against practices they deem unsafe for fear of
deviating from group norms.\4\ Individuals may also be unaware of
systematic biases in the group decision-making process and may then be
less likely to scrutinize the potential risks of the group's decision
or sufficiently contemplate alternative paths of action.\5\
Furthermore, the literature indicates that groups make riskier
decisions than individuals acting alone due to the diffusion of
responsibility among group members.\6\ This phenomenon is known as
``the risky shift.'' ``Groupthink'' and ``the risky shift'' may lead to
group behaviors that render behavioral observation less effective. As
such, alternative approaches to BOPs, such as the utilization of video-
based surveillance by individuals separate from the onsite work unit,
could serve to mitigate potential issues associated with groupthink.
The incorporation of remote observation, performed by individuals
physically separate from the site, could help to bring in independent
and objective perspectives and help to break patterns of thought and
communication that may result in groupthink.
---------------------------------------------------------------------------
\4\ See e.g., Russell Mannion and Carl 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 (arguing that small group dynamics in healthcare
teams produce systematic biases in group decision-making because
healthcare professionals may be reticent to vocalize concerns they
have about quality of care).
\5\ See e.g., W[aelig]r[oslash], Rosness, and Kilska (arguing
that groupthink leads teams to ``develop shared rationalizations
that bolster a proposed choice, rather than examining alternative
options and identifying the risks associated with the proposed
choice''). See also David Hofmann and Adam Stetzer, ``A Cross-Level
Investigation of Factors Influencing Unsafe Behaviors and
Accidents,'' Personnel Psychology, Vol. 49 (1996) (finding that in a
study of fatal accidents involving offshore oil rigs, in the absence
of standard operating procedures, workers ``equated normal work
methods (i.e., what everyone else does) with safe and/or ideal work
methods,'' revealing that the groupthink phenomena will further
cement modes of work that do not reflect safety protocols in small
groups that lack strong norms around workplace safety and tacitly
reward short-cuts that prioritize efficiency over safety).
\6\ 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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Even without the influence of small group dynamics, there are other
practical constraints to implementing FFD requirements, such as random
drug and alcohol testing, among small staffs. Random testing is 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 necessitates additional flexibilities be provided to
implement various FFD program elements. The NRC is requiring that
facilities with small staff sizes that cannot implement random drug and
alcohol testing without predictability to 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.
26.607(b)(2)(vi). Use of a C/TPA significantly improves the
effectiveness of the random testing programs of sites with small worker
populations and ensures 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 enable 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 is in Sec. 26.607(g)(2), where the NRC is
enabling 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. 26.607(b)(2)(vi). These sites 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 a virtual oral fluid collection are directly observed by
the specimen collector, a video teleconference could accomplish many
key elements of the collection process. The use of video teleconference
technology also is not new to the NRC, as some clinicians complete
other required evaluations using video teleconferencing technology,
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 this part. 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 is establishing a change control requirement to allow
a licensee or other entity to change its subpart M of part 26 FFD
program while ensuring that FFD program effectiveness is maintained.
Lastly, subpart M of part 26 consolidates the applicable FFD
requirements by placing in one subpart all part 26 requirements (either
new requirements or cross-references to existing part 26 requirements)
for part 53 licensees and other entities. This should help licensees
and other entities implement the requirements because it enables easy
cross-reference to similar requirements in other subparts that are
being implemented by non-part 53 licensees and entities subject to part
26. Understanding how other licensees or other entities implement
similar FFD requirements may facilitate the sharing of operating
experience in program implementation.
The use of innovative technologies and a risk-informed performance-
based framework parallels the considerations presented in the Advanced
Reactor Policy Statement. As stated in the policy statement,
``[S]implified systems should facilitate operator comprehension,
reliable system function, and more
[[Page 15734]]
straightforward engineering analysis.'' Furthermore, these same
attributes may reduce potential radiation exposures, help prevent the
theft of nuclear materials, and use technology and design innovations.
Should these components and systems be designed, implemented, and
maintained to minimize reliance on human actions and leverage
technology and innovation, then the robust and prescriptive FFD
requirements in, for example, subparts B, ``Program Elements,'' and E
of part 26 could be scaled to the part 53-licensed facility and its
operation. This strategy is implemented in the subpart M of part 26
framework.
Even though current subpart K of part 26 provides a more flexible
FFD program framework than the framework comprising all the subparts of
part 26 except subparts I and K, subpart M of part 26 does not allow
part 53 licensees and other entities to implement the requirements in
subpart K. The principal reasons are that (without significant changes
to subpart K that are outside the scope of this rulemaking): (1)
subpart K does not apply to holders of MLs who assemble or test a
reactor; (2) subpart K only applies during construction (and prior to
the receipt of special nuclear material in the form of fuel
assemblies), whereas subpart M applies during construction, operation,
and decommissioning through implementation of the insider mitigation
program (IMP) required by Sec. 73.55 or Sec. 73.100; (3) subpart K
does not address training, authorization as defined in Sec. 26.5, and
MRO performance; (4) subpart K does not expressly authorize the use of
innovative drug and alcohol testing technologies; (5) subpart K does
not describe the use of time-dependent alcohol limits or special
analysis testing of urine specimens; and (6) subpart K has less rigor
in the protection of worker rights and sensitive information than that
required in subpart M.
Despite the differences between subparts K and M of part 26, the
requirements in subpart M are essentially equivalent to many in subpart
K that were implemented by the licensees of Vogtle Nuclear Station and
V.C. Summer Nuclear Station when they were constructing four commercial
nuclear power reactors and NRC inspection and operating experience
evaluation determined that the use of subpart K contributed to
adequately protecting the public health and safety and the common
defense and security. Further, given the risk profile posed by
facilities licensed under part 53 and the additional requirements in
subpart M of part 26 that were developed from operating experience and
other part 26 subparts (but are not included in subpart K of part 26),
the NRC concludes that if licensees and other entities effectively
implement the requirements in subpart M of part 26, then their FFD
programs would provide reasonable assurance that individuals subject to
this final rule are fit for duty and trustworthy and reliable.
B. Changes to Part 26, Subparts A through E and I
Section 26.3(d) is the applicability paragraph for contractor/
vendors (C/Vs) who 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. This final rule amends Sec.
26.3(d) to address part 53 licensees and other entities in Sec.
26.3(f).
Section 26.3(f) places part 53 licensees or other entities within
the scope of part 26. For licensees and other entities of a part 53
commercial nuclear plant, except a holder of an ML, the FFD program is
required to be implemented no later than the start of construction
activities. The holder of an ML needs to implement its FFD program
before commencing activities that assemble a reactor.
Current Sec. 26.4 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 is amending Sec. 26.4 to include licensees and other entities
described in Sec. 26.3(f). The NRC expects that not all categories of
individuals described in current Sec. 26.4 are applicable to all part
53 facilities. The NRC is establishing regulatory guidance in RG 5.99,
``Fatigue Management for Nuclear Power Plant Personnel at Commercial
Nuclear Plants Licensed Under 10 CFR part 53,'' to help address program
applicability to certain individuals.
This final rule amends Sec. 26.4(a)(1) and (a)(4) 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 part 53.
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 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 (73 FR 16966, March 31, 2008), 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 this final rule's amendments to part 26, the limitation of
``onsite'' activities to those performed within the owner-controlled
area still applies to facilities licensed under part 50 or 52. However,
for licensees and other entities described in Sec. 26.3(f), the NRC is
removing 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 commercial 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 this final
rule's amendments to part 26, the exclusion of incidental duties while
calculating work hours is still applicable for licensees and other
entities licensed under part 53. 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, is considered in
an equivalent fashion as direction performed at non-remote facilities,
for the purposes of administering work hour controls.
Section 26.4(b) includes in an FFD program individuals who are
granted unescorted access to the protected area of a facility licensed
under part 53 and do not perform or direct the performance of the
duties described in Sec. 26.4(a). This requirement contributes to the
defense-in-depth regulatory framework that helps provide reasonable
assurance that individuals who have unescorted access are fit for duty,
trustworthy, and reliable. For example, through this final rule, the
NRC is amending part 73 to require a part 53 licensee to subject
individuals to
[[Page 15735]]
a series of reviews to help determine whether those individuals are
trustworthy and reliable before granting them unescorted access to the
facility's protected area.
Through this final rule, the NRC is amending Sec. 26.4(c) to
include in an FFD program individuals who are assigned to physically
report to the part 53 licensee's emergency response facility (or
facilities) or participate remotely in emergency response activities,
and individuals without unescorted access to the part 53 facility who,
remotely or otherwise, make decisions and/or direct actions regarding
plant safety or security. Part 53 commercial nuclear plants may be
licensed for and rely upon offsite facilities to fulfill the role of a
Technical Support Center or Emergency Operations Facility. Therefore,
this final rule accounts 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 commercial 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. This final rule amends Sec. 26.4(e) to except
subpart M as well as subparts I and K. This final rule also amends
Sec. 26.4(e) to include in an FFD program the individuals whose duties
for the 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.
This final rule revises Sec. 26.4(e)(4) to include in an FFD
program individuals who witness or determine inspections, tests, and
analyses certifications required under part 53 because current Sec.
26.4(e)(4) includes the individuals who perform the same duties under
part 52.
This final rule amends Sec. 26.4(f) to require individuals who
construct or direct the construction of safety- or security-related
SSCs at facilities licensed under part 53 to be subject to an FFD
program under subpart M of part 26 or an FFD program that demonstrates
compliance with all of the requirements of part 26 except for subparts
I, K, and M of part 26.
Section 26.4(g) is the applicability paragraph for FFD program
personnel (e.g., the FFD manager, MRO, and technicians) and persons who
perform AA determinations (e.g., the licensee- or other entity-
designated Reviewing Official). This final rule amends this section to
address part 53 licensed facilities. Specifically, a part 53 licensee
or other entity will 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 will be held accountable for program
implementation, including consistent implementation of protections
afforded to all individuals subject to the FFD program.
This final rule amends Sec. 26.4(h) to include subpart M of part
26.
Through this final rule, the NRC includes several new definitions
in Sec. 26.5, ``Definitions,'' and amends some existing definitions.
The NRC is adding a definition for ``Biological marker.'' The
definition is 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 definition for Sec. 26.5
adds that the endogenous substance used to validate that the biological
specimen ``was produced by the donor'' because subpart M of part 26
requires the MRO to evaluate any discrepant biological marker
identified in a biological specimen collected from a donor.
The NRC is including a definition for the word ``Change'' as used
in the Sec. 26.603(e), ``FFD program change control,'' process. The
definition is 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 including a definition for ``Consortium/Third-party
administrator (C/TPA),'' which is used in Sec. 26.607(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. This definition is based, in part, on the Federal
Motor Carrier Safety Administration regulations in 49 CFR part 382 and
the U.S. Department of Transportation regulations in 49 CFR part 40.
The NRC is revising the definition of ``Constructing or
construction activities'' to clarify that for licensees or other
entities in Sec. 26.3(f), the definition of ``Construction'' is
consistent with the definition in Sec. 53.020.
This final rule revises the definitions of ``Contractor/vendor''
(C/V) and ``Other entity'' to make them applicable to part 53
licensees. A holder of an ML under part 53 could be a C/V under the new
C/V definition.
The NRC is including a definition for ``Illicit substance'' because
this phrase is used in subpart M of part 26 and addresses substances
that cause impairment and possible addiction but are not an ``illegal
drug'' as defined in Sec. 26.5. This 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 including a definition for ``Reduction in FFD program
effectiveness'' because this phrase, similar to the definition for
``Change,'' is used in Sec. 26.603(e). The definition is generally
consistent with the definition of ``Reduction in effectiveness''
provided for emergency plans in Sec. 50.54(q)(1)(iv).
This final rule makes the current definition of ``Reviewing
official'' applicable to those licenses and other entities in Sec.
26.3(f).
This final rule amends the current part 26 definition of ``Safety-
related structures, systems, and components'' to use the NRC's
definition in Sec. 53.020 for the part 53 licensees and other entities
described in Sec. 26.3(d) and (f).
This final rule amends the definition of ``Security-related SSCs''
in Sec. 26.5 to make it applicable to a licensee or other entity
described in Sec. 26.3(d) and (f).
The NRC is including a definition for ``Special nuclear material''
that refers to the definition in Sec. 70.4, ``Definitions,'' of part
70 to ensure consistency.
This final rule revises the definition of ``Unit outage'' to
account for the potential use of commercial nuclear plants for purposes
other than electricity generation.
[[Page 15736]]
This final rule amends Sec. 26.21, an applicability statement for
part 26 FFD programs, to include licensees and other entities described
in Sec. 26.3(f) that choose to implement an FFD program that
implements all part 26 requirements, except those in subparts K and M
of part 26.
This final rule amends Sec. 26.35(c)(3) to include a reference to
Sec. 26.606(b)(2)(vii), which ensures that licensees and other
entities take immediate action upon receiving notice from the employee
assistance program (EAP) that an individual's condition or actions pose
or have posed an immediate hazard to themselves or others.
This final rule amends Sec. 26.51, ``Applicability,'' to apply to
licensees and other entities described in Sec. 26.3(f) that elect not
to implement the requirements in subpart M of part 26 for the
categories of individuals in Sec. 26.4 and those licensees and other
entities that elect to implement the requirements in Sec. 26.605.
This final rule amends Sec. 26.53(e), (e)(1) and (3), and (g)
through (i), which are general provisions for granting and maintaining
authorization, to apply to licensees and other entities described in
Sec. 26.3(f).
This final rule amends Sec. 26.63(d), a suitable inquiry
requirement, to apply to licensees and other entities described in
Sec. 26.3(f).
This final rule amends Sec. 26.73, the applicability statement for
subpart D of part 26, to apply to licensees and other entities
described in Sec. 26.3(f) that elect not to implement the requirements
in subpart M of part 26 for the categories of individuals in Sec. 26.4
and those licensees and other entities that elect to implement the
requirements in Sec. 26.605(b).
This final rule amends Sec. 26.81, the purpose and applicability
statement for subpart E of part 26, to apply to licensees and other
entities described in Sec. 26.3(f) that elect not to implement the
requirements in subpart M of part 26 for the categories of individuals
in Sec. 26.4 and those licensees and other entities that implement
Sec. 26.605(a) or (b). The subpart E requirements to be implemented
are listed in Sec. 26.607(c)(2)(i) and (ii), and (c)(3).
This final rule revises Sec. 26.97(a) and (b) to enable the
virtual collection of oral fluid specimens for drug and alcohol
testing, as permitted under Sec. 26.607(g)(2).
This final rule amends Sec. 26.201, the applicability statement
for subpart I of part 26, to apply to licensees and other entities
described in Sec. 26.3(f). Also, the applicability statement is
divided into two paragraphs for clarity.
The NRC is adding Sec. 26.202, ``General provisions for facilities
licensed under part 53,'' for licensees or other entities described in
Sec. 26.3(f) that elect to implement the requirements in subpart I of
part 26 in accordance with Sec. 26.605. Section 26.202 establishes
requirements equivalent to those in current Sec. 26.203, ``General
provisions,'' which is applicable to parts 50 and 52 licensees. The NRC
is adding the separate Sec. 26.202 because Sec. 26.203 refers to
various requirements under subpart B of part 26, which are not
applicable to facilities licensed under part 53 that implement subpart
M of part 26.
Additionally, Sec. 26.202(c), ``Training and assessments,'' unlike
Sec. 26.203(c), ``Training and examinations,'' does not include a
comprehensive examination requirement because trainee assessment is
conducted as part of a SAT that would be required under the FFD program
training requirements in Sec. 26.608.
Changes in Sec. Sec. 26.205, 26.207, and 26.211 add references to
new requirements in subparts I and M of part 26 that are applicable
specifically to licensees and other entities in Sec. 26.3(f). The NRC
is not changing the specific provisions for work hour requirements in
current Sec. 26.205(d). However, as addressed in the discussion of
changes to Sec. 26.4(a), whether a licensee or other entity under part
26 needs to implement work hour controls for certain individuals or
groups is dependent, in part, on determinations reached by that
licensee's risk-informed evaluation process.
Changes to Sec. Sec. 26.207(a)(1)(ii) and 26.211(b) allow
licensees and other entities in 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 changes allow supervisors to conduct such
assessments from a remote location under appropriate circumstances.
Such remotely conducted assessments 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. Sec. 26.29
and 26.203(c), or Sec. Sec. 26.608 and 26.202(c). The reasoning for
these changes and the associated need for in-person support to augment
electronic communications is addressed further in the preamble
discussion of Sec. 26.619.
C. Requirements for Part 26, Subpart M
This final rule adds a new subpart M to part 26 that provides
alternative FFD requirements for part 53 licensees and other entities.
Section 26.601 describes which entities can implement subpart M.
Section 26.601(a) makes subpart M of part 26 applicable to part 53
licensees and other entities, at their discretion. If a licensee or
other entity in Sec. 26.3(f) does not elect to implement an FFD
program that demonstrates compliance with the requirements of subpart
M, then the individuals specified in Sec. 26.4 will be subject to an
FFD program that demonstrates compliance with all part 26 requirements,
except for those requirements in subparts K and M.
For a licensee or other entity in Sec. 26.3(f) that elects to
implement an FFD program that satisfies the requirements of subpart M,
Sec. 26.601(b) and (c) describes which provisions of subpart M apply.
Under Sec. 26.601(b), a licensee or other entity that demonstrates
compliance with Sec. 73.100(a)(1)(i) has the option to implement an
FFD program under either Sec. 26.605(a) or (b). Under Sec. 26.601(c),
if a licensee or other entity elects to implement an FFD program under
subpart M but does not demonstrate compliance with Sec.
73.100(a)(1)(i), then that licensee or other entity must implement an
FFD program under both Sec. 26.605(a) and (b).
Section 26.603(a) requires an applicant to provide a description of
its FFD program and its implementation within its application for a
license. This requirement is equivalent to the existing requirements in
Sec. Sec. 26.401(b) and 52.79(a)(44). The entities required to submit
these FFD program descriptions are certain applicants that comply with
the part 53 application requirements in subpart H. In subpart H, Sec.
53.1309(a)(6) requires an applicant for a CP to provide a description
of its FFD program in its PSAR. Under Sec. Sec. 53.1279(b)(4),
53.1369(x), and 53.1416(a)(24), this final rule requires an applicant
for an ML, OL, and COL, respectively, to provide a description of its
FFD program in its FSAR.
Unlike an application for a license, a description of an FFD
program does not receive NRC review for possible approval. The
applicant provides 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 will be effectively implemented before
a licensee or other entity commences any activity making individuals at
the NRC-licensed facility subject to the FFD program.
Section 26.603(a)(1) requires the applicant to state whether it
demonstrates compliance with Sec. 73.100(a)(1)(i), which is necessary
to
[[Page 15737]]
understand FFD program applicability under Sec. 26.605(a) and (b).
Section 26.603(a)(2) requires the applicant to state what FFD
program it plans to implement under subpart M (i.e., Sec. 26.605(a),
Sec. 26.605(b), or Sec. 26.605(a) and (b)), or the existing FFD
program (i.e., all parts of part 26, except for subpart K and M).
Section 26.603(a)(3) requires a discussion that informs the NRC of
the applicability of the applicant's FFD program to individuals who
perform safety- or security-significant activities. 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 is to inform the
NRC of any substantial differences in the applicability of the FFD
program to the categories of individuals in Sec. 26.4.
Section 26.603(a)(4) requires 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 includes how individuals who test positive for a drug or
alcohol will 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.
Section 26.603(b) establishes the longevity of a license's or other
entity's FFD program. Unlike the current part 26 regulations, Sec.
26.603(b) expressly states that an FFD program is not applicable during
decommissioning of a part 53 facility for licensees and other entities
specified in Sec. 26.3(f). However, holders of an operating or
combined license should be aware that the physical protection program
regulations in Sec. 73.55, ``Requirements for physical protection of
licensed activities in nuclear power reactors against radiological
sabotage,'' and Sec. 73.100 include a requirement for the
implementation of an IMP, even during decommissioning. Section
26.603(b) also requires the holder of an ML under part 53 to maintain
its FFD program until expiration of the ML.
In Sec. 26.603(e), this final rule implements a change control
requirement for subpart M of part 26 FFD programs. Requiring licensees
and other entities to demonstrate compliance with certain requirements
before implementing changes to their FFD programs is necessary for two
primary reasons. First, compliance with Sec. 73.100(a)(1)(i)
determines which FFD program requirements may be implemented. If a
licensee makes changes to its facility that impact the licensee's
ability to comply with Sec. 73.100(a)(1)(i), then the set of
applicable FFD program requirements under Sec. 26.605 may change and
would need to be documented under Sec. 26.603(e). Second, FFD program
implementation may change periodically in response to societal changes
in substance abuse. Change control therefore relies 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.
Section 26.603(e)(1) permits the licensee or other entity to
implement changes to its FFD program if it performs and retains an
analysis demonstrating that the change does not reduce the
effectiveness of the FFD program or the change was necessitated or
justified by a change to part 26, laboratory processes, or guidance
issued by the HHS or NRC. The change control requirement enables
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.
The change control requirement was developed from the change
control requirements in Sec. 50.54(p) and (q)--the change control
requirements for security and emergency plans, respectively. However,
unlike these two requirements, the NRC does not review and approve a
licensee's or other entity's FFD program or its implementing
procedures, and the FFD program is not licensing-basis information as
described in Sec. 53.1300.
Section 26.603(e)(2) requires that if a change reduces FFD program
effectiveness, then the licensee must implement a mitigating strategy
so the FFD program, as revised, will continue to demonstrate compliance
with the performance objectives in Sec. 26.23 and not result in a
reduction in FFD program effectiveness.
Section 26.603(e)(3) prohibits, with one exception, the use of the
change control process to reduce the minimum panel of drugs to be
tested and references the drugs listed in Sec. 26.607(c)(1). Section
26.607(c)(1) references 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, except phencyclidine, and alcohol is necessary for the FFD
program to remain effective. Also, there is no subpart M of part 26
requirement stating that this panel of drugs and drug metabolites needs
to consist of only scheduled drugs.\7\ This flexibility accounts 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.
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\7\ The Drug Enforcement Administration classifies drugs,
substances, and certain chemicals used to make drugs into five (5)
distinct categories, depending upon the drug's acceptable medical
use and the drug's abuse or dependency potential. These categories
appear as Schedules I through V of section 202 of the Controlled
Substances Act (21 U.S.C. 812). Schedule I drugs have a high
potential for abuse, have no currently accepted medical uses in
treatment in the United States, and lack accepted safety for use
under medical supervision. At the other end of the classification
scheme, Schedule V drugs have the least potential for abuse among
the five categories of drugs, have a currently accepted medical use
in treatment in the United States, and abuse of the drug may lead to
limited physical dependence or psychological dependence. For more
information, see https://www.dea.gov/drug-information/drug-scheduling.
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The exception in Sec. 26.603(e)(3) is that, should HHS elect to
remove phencyclidine from the panel of drugs and drug metabolites to be
tested, a licensee or other entity could make this change in its FFD
program without resulting in a reduction in FFD program effectiveness.
This outcome is justified based on the very infrequent occurrence rate
of FFD policy violations due to phencyclidine use since 2010. However,
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), except for
phencyclidine, 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 is
not expected to result in a reduction in FFD program effectiveness
because HHS would be
[[Page 15738]]
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. Departments of
Transportation, Energy, and Defense.
Section 26.603(e)(4) requires 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 affords 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 are also required to summarize each change made under
Sec. 26.603(e) in their annual FFD performance reports required by
Sec. 26.617(b)(2) or Sec. 26.717, as applicable.
Section 26.605 establishes requirements in a graded manner similar
to the regulatory framework established by the requirements in subparts
A through I, N, O, and K of part 26. This graded approach consists of
less prescriptive FFD program requirements in Sec. 26.605(a) and more
robust program requirements in Sec. 26.605(b).
The FFD programs under Sec. 26.605(a) and (b) include FFD program
elements similar to those in subpart B of part 26, but the new
requirements are 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 Sec.
26.605. Section 26.605 requires drug and alcohol testing based on
either the requirements in part 26 or the HHS Guidelines. The principal
benefits of the Sec. 26.605 FFD program are that it provides a
regulatory framework that is consistent with the radiological
consequences for a facility that demonstrates compliance with Sec.
73.100(a)(1)(i), and affords flexibilities in the conduct of drug and
alcohol testing.
Section 26.605(a) applies to part 53 licensees and other entities
of commercial nuclear plants that demonstrate compliance with Sec.
73.100(a)(1)(i), at their discretion (no later than the start of
construction activities as defined in Sec. 26.5); licensees and other
entities of commercial nuclear plants that do not demonstrate
compliance with Sec. 73.100(a)(1)(i) (no later than the start of
construction activities as defined in Sec. 26.5); and holders of MLs
before the start of activities performed under an ML that allows the
assembly, non-operational testing, or both, of a manufactured reactor.
The timing element of the applicability statement of Sec. 26.605(a) is
equivalent to that for an LWR licensee or other entity who is
performing those same activities at a facility licensed under part 50
or 52 and helps provide assurance that those individuals who assemble,
test, 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
quality assurance 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 addresses 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 are relied upon for safe operation
but cannot be inspected or tested for proper installation,
configuration, or operation after installation. A Sec. 26.605(a) FFD
program for these types of activities is 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.
Section 26.605(a) requires the holder of an ML to implement its FFD
program no later than the start of activities that assemble a reactor,
non-operational testing of a manufactured reactor, or both. The holder
of the ML should establish in its procedures when reactor assembly
commences and what constitutes assembly. For example, the FFD program
does 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 commercial power reactor, and this regulatory
position also applies to a manufacturing facility.
Section 26.605(b) also contains timing requirements for
implementing FFD programs under that paragraph. Licensees and other
entities that demonstrate compliance with Sec. 73.100(a)(1)(i) and
elect to implement FFD programs that satisfy the requirements of Sec.
26.605(b) must establish, implement, and maintain the program no later
than the start of construction activities. Holders of MLs that also
possess a separate license to load fuel into a manufactured reactor
must establish, implement, and maintain the program no later than the
start of reactor fuel load. For all other licensees and other entities
implementing an FFD program under Sec. 26.605(b), they must establish,
implement, and maintain the program before the earliest of the loading
of fuel onsite into a reactor vessel; receiving a fueled manufactured
reactor; or 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.
These entities must establish, implement, and maintain an FFD
program that implements all the requirements in Sec. 26.605(a), except
Sec. Sec. 26.610, ``Sanctions''; 26.617, ``Recordkeeping, reporting,
and FFD program performance''; and 26.619, ``Suitability and fitness
determinations''; plus additional requirements due to the increased
radiological consequences presented by a part 53 commercial nuclear
plant as the licensee readies it for operation. These additional
requirements include those in subparts C, D, H, and N of part 26, some
of which replace Sec. Sec. 26.610, 26.617, and 26.619.
Section 26.605(b) also enables the licensee or other entity to
better integrate its facility with the LWR fleet and Category I fuel
cycle facilities because subparts C, D, and H of part 26 are required.
These subparts are required, in part, because it is expected that: (1)
individuals will be able to work at any part 50, 52, or 53 commercial
nuclear plant and will possess a nuclear safety culture and desirable
[[Page 15739]]
qualifications, skills, expertise, or services; and (2) licensees and
other entities of facilities licensed under parts 50, 52, and 70 may
venture to construct or operate a facility licensed under part 53.
Therefore, the implementation of these subparts helps ensure that all
individuals subject to part 26, whether under subpart M, subpart K, or
all subparts except M and K, are subject to FFD programs that provide
reasonable assurance that the individuals are fit for duty,
trustworthy, and reliable.
Section 26.606, ``Written policy and procedures,'' requires
licensees and other entities to implement and maintain an FFD policy
and procedures for their FFD programs. This section establishes
requirements equivalent to those in current Sec. 26.403, ``Written
policy and procedures,'' of subpart K. However, a principal difference
is that Sec. 26.606 is written to enable the drug testing of urine,
oral fluid, and hair specimens.
Section 26.606(a)(1) requires 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 is 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 also contributes 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.
Section 26.606(a)(2) requires 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 parts 50, 52, or 70. Having a standard performance outcome based
on a licensee or other entity satisfying the Sec. 26.23 performance
objectives enhances consistency in FFD program implementation across
all entities subject to part 26. It also generates 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.
Section 26.606(a)(3) requires 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).
Section 26.606(a)(4) requires 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
Sec. 26.603(b), part 26-required sanctions, and required medical/
clinical treatment and follow-up testing for FFD policy violations.
This requirement is equivalent to Sec. 26.403(a) of subpart K but
includes an additional description of what the policy statement must
include. For example, the policy describes the NRC-required sanctions
to help deter substance abuse and required medical/clinical treatment
and follow-up testing for FFD policy violations. This provision
provides a protection measure by helping the individual get the
assistance they need and help ensure that the individual refrains from
substance abuse.
Section 26.606(a)(5) requires that the FFD policy statement
describe 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.
Section 26.606(a)(6) requires that 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.
Section 26.606(a)(7) requires that 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.
Section 26.606(b) requires licensees and other entities
implementing an FFD program in accordance with subpart M of part 26 to
establish, implement, and maintain written procedures for their FFD
programs. This requirement is equivalent to that in Sec. 26.403(b) of
subpart K.
Section 26.606(b)(1) establishes requirements for the licensee or
other entity to develop and maintain written procedures for its drug
and alcohol testing program. This provision is equivalent to the
requirements in current Sec. 26.403(b)(1) of subpart K, but Sec.
26.606(b)(1)(i) through (iv) requires additional clarity and
specificity that licensees and other entities must detail in their
procedures to address new testing methods in subpart M of part 26 that
are not permitted under the current part 26 framework. Clarity and
specificity in procedural instructions support consistent program
implementation, which protects all individuals subject to the program.
Section 26.606(b)(1)(iv) requires 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 helps ensure the following: the validity and accuracy
of drug testing because the specimens are 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 are 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
will enhance MRO and FFD program personnel reviews of individual test
results because instructions will 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 benefits FFD program effectiveness
and helps prevent misunderstanding of program requirements and
processes.
Section 26.606(b)(2) requires licensees and other entities to
include in their written procedures the immediate and follow-up actions
that will be taken, and the procedures that will be used, in certain
situations specified in Sec. 26.606(b)(2)(i) through (vi). Section
26.606(b)(2) is 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
helps ensure the effectiveness of the FFD program and its consistent
implementation, because part 53 licensed facilities will be
[[Page 15740]]
implementing procedures to address the same requirements and with
individuals who understand what is expected of them no matter what part
53 facility they were assigned.
The situation specified in Sec. 26.606(b)(2)(i) arises 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 is equivalent to current Sec.
26.403(b)(2)(i), except that the phrase ``illegal drugs'' is replaced
with ``illegal substances, illegal drugs, or illicit substances.''
Illegal substances include legal substances used in a manner
inconsistent with Federal or State law.
The situation specified in Sec. 26.606(b)(2)(ii) arises when
individuals who are subject to the FFD program 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 is
equivalent to current Sec. 26.403(b)(2)(ii) of subpart K. The NRC does
not include the phrases ``to excess'' and ``accurately'' in Sec.
26.606(b)(2)(ii). Subpart M of part 26 is 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 evidential breath testing device (EBT) for alcohol
(not whether an individual drank ``to excess'').
The NRC is including the phrase ``illegal substances, illegal
drugs, and illicit substances'' in Sec. 26.606(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 the 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
Sec. 26.606(b)(2)(ii) because Sec. 26.606 applies during
construction, operation, and decommissioning, if applicable. The NRC is
including the term ``behavioral observation'' in Sec. 26.606(b)(2)(ii)
because impairment can be visibly or audibly observed in an individual,
and individuals subject to subpart M of part 26 will be trained in
behavioral observation under Sec. 26.608.
The situation specified in Sec. 26.606(b)(2)(iii) arises 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. This provision is equivalent to current Sec.
26.403(b)(2)(iii). The purpose underlying this requirement has
increased in significance since 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 Sec. 26.606(b)(2)(iv) arises 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 is equivalent to current Sec.
26.403(b)(2)(iv). The NRC is including 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 Sec. 26.606(b)(2)(v) arises when
individuals who are subject to an FFD program had legal action taken
relating to drug or alcohol use. This requirement is equivalent to
current Sec. 26.403(b)(2)(v).
The situation specified in Sec. 26.606(b)(2)(vi) is when
individuals subject to an FFD program demonstrate 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
requirement helps 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.
This requirement also helps to align the subpart M of part 26 BOP
with the BOP implemented under Sec. 73.56(f) and Sec. 73.120 and the
purpose of the IMP as described in Sec. 73.55(b)(9) and Sec.
73.100(b)(10).\8\ The demonstrated character and actions of an
individual can indicate whether the individual can be trusted and
relied upon to safely and competently perform assigned duties and
responsibilities or be afforded those types of access making the
individual subject to the FFD program. This holds true for any
demonstrated adverse character indication or action on- or offsite.
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\8\ The IMP must monitor the initial and continuing
trustworthiness and reliability of individuals granted or retaining
unescorted AA to a protected or vital area and implement defense-in-
depth methodologies to minimize the potential for an insider to
adversely affect, either directly or indirectly, the licensee's
capability to protect against radiological sabotage.
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The phrase ``character or actions'' is used in Sec.
26.606(b)(2)(vi) to focus on observed examples that indicate an
individual subject to subpart M of part 26 may not be fit for duty or
trustworthy and reliable. Character traits include but are not 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 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) and
Sec. 73.120, 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 Sec. 26.606(b)(2)(vi) is 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 is received from an individual subject to the FFD program
or someone who is not subject to the FFD program, the licensee or other
entity will need to review this information (i.e., determine if the
information and its source are credible) to determine whether the
individual should maintain authorization.
Section 26.606(b)(3) requires licensees and other entities to
address in their
[[Page 15741]]
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 also requires 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, Sec. 26.606(b)(3) is 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 uses ``on the NRC-licensed
facility'' in Sec. 26.606(b)(3), because this provision applies during
commercial nuclear plant construction, operation, and decommissioning,
if applicable, in addition to holders of an ML as described in 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.
The requirement also includes two conditions not found in current
Sec. 26.403(b) that 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,'' reflects 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,'' reflects the Sec. 26.23(a)
performance objective. In either case, as required by Sec. 26.23(c),
the FFD program 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.
Section 26.606(b)(4) requires licensees and other entities to have
written procedures that address the operation and oversight of onsite
and offsite collection facilities. This requirement is 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 enhances consistency in
program implementation, protects individuals subject to testing, and
accounts for the flexibilities afforded in the types of biological
specimens that may be collected under an FFD program subject to subpart
M of part 26. Section 26.606(b)(4), when used with the audit
requirement in Sec. 26.615, helps 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.
Section 26.606(b)(5) requires licensees and other entities to have
written procedures that address the fatigue management requirements in
Sec. 26.202(b), ``Procedures,'' and either Sec. 26.205(d)(3) or
(d)(7).
Section 26.606(b)(6) requires licensees and other entities to have
written procedures that provide measures to prevent subversion of drug
and alcohol tests conducted onsite and offsite. This requirement was
developed from Sec. 26.27(c)(1).
Section 26.607, ``Drug and alcohol testing,'' establishes drug and
alcohol testing requirements for licensees and other entities
implementing Sec. 26.605. Except for a few differences, Sec. 26.607
is equivalent to current Sec. 26.405, which requires licenses 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 53-
licensed facility as compared to a part 50 or 52 nuclear power plant.
These requirements improve flexibility in the conduct of drug and
alcohol testing while maintaining protections afforded to individuals
subject to the FFD program.
Section 26.607(a) requires licensees and other entities to obtain a
split specimen for all drug tests using oral fluid or urine for all
test conditions in Sec. 26.607(b) and (j). Neither current subpart K
nor current subparts B or E of part 26 require a split specimen.
However, many of the LWR fleet use 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 protects 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 enables 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 will be conducted by HHS-certified laboratories.
If a specimen is determined to be invalid, the occurrence would warrant
further investigation by the MRO and laboratory to identify the cause.
This protocol is 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 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.
Section 26.607(a) states that split specimen collections of oral
fluid or urine must be used for the test conditions described in Sec.
26.607(b). In addition, testing of the split specimen (specimen B)
requires the donor's permission unless ordered by the MRO
[[Page 15742]]
to resolve an invalid test result obtained for specimen A.
Section 26.607(b) requires the licensee or other entity to subject
individuals identified in Sec. 26.202 to drug and alcohol testing
under the five conditions listed in Sec. 26.607(b)(1) through (5).
Section 26.607(b) is equivalent to current Sec. 26.405(c).
Section 26.607(b)(1) requires 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 requirement does not
include the phrase, ``construct or direct the construction of safety-
or security-related SSCs,'' because, for licensees or other entities
under part 53, the pre-access test condition applies to construction,
operation, and decommissioning, if applicable, to help inform a
licensee's or other entity's authorization determination. The
requirement also uses ``pre-access'' instead of ``pre-assignment,''
which is used in current Sec. 26.405(c)(1).
A pre-access test requires 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
requirement is based on two lessons learned from operating experience.
First, the 14-day period is 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 is 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 part 53 to have the large and periodic influxes of
individuals (either licensee employees or C/Vs) that LWRs have to
support facility operation, maintenance, engineering design changes, or
nuclear refueling. Therefore, these licensees or other entities will
not be periodically challenged to in-take a large workforce within the
14-day pre-access testing window.
Section 26.607(b)(2) requires the licensee or other entity to
conduct random drug and alcohol testing of all individuals subject to
the FFD program. With notable exceptions, this requirement is
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. Section 26.607(b)(2) does not offer that option because
subpart M of part 26, unlike subpart K, does not allow a licensee or
other entity to implement a fitness monitoring program under current
Sec. 26.406 instead of a random testing program. The principal reasons
for not allowing this flexibility are 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 subpart M framework already
uses behavioral observation to help ensure FFD program effectiveness.
Supplementing the Sec. 26.609 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 will be
obtained through behavioral observation and random drug and alcohol
testing.
Section 26.607(b)(2)(i) through (v) provides specific requirements
for the conduct of a random testing program. These paragraphs are
equivalent to Sec. 26.405(b)(1) through (4), although with a few
differences. The similar provisions are in Sec. 26.607(b)(2)(i),
(b)(2)(iii), and (b)(2)(iv).
The differing provisions include Sec. 26.607(b)(2)(ii), which
refers 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. Section 26.607(b)(2)(ii)
also requires 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 being implemented because the NRC
expects that part 53 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 Sec. 26.607(b)(2)(v),
which establishes 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 new requirement is 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.
Section 26.607(b)(3) requires for-cause testing equivalent to that
used in current FFD programs implementing Sec. 26.405(c)(2). The NRC
is requiring 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.
Section 26.607(b)(4) requires post-event testing in a manner
equivalent to current Sec. 26.405(c)(3), with a few adjustments. For
part 53 licensees or other entities, the NRC is requiring post-event
testing under two conditions: events involving human errors that may
have caused or contributed to the events (Sec. 26.607(b)(4)(i)), and
events not involving human error that result in adverse health
consequences or damage to any safety- or security-related SSC (Sec.
26.607(b)(4)(ii)). The word ``significant'' is not used in Sec.
26.607(b)(4)(ii)(A) to describe the ``illness or personal injury'' as
used in Sec. 26.405(c)(3)(i) because Sec. 26.607(b)(4)(ii)(A)
describes which illnesses or injuries are covered. Section
26.607(b)(4)(ii)(B), unlike Sec. 26.405(c)(3)(ii), does not use the
word ``significant'' to describe the damage to safety- or security-
related SSCs because any damage to safety- or security-related SSCs
requires 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. Section
26.607(b)(4)(ii)(B) also does not use the word ``construction'' as in
Sec. 26.405(c)(3)(ii) because Sec. 26.607(b)(4) applies to
construction, operation, and decommissioning, if applicable.
Section 26.607(b)(4)(i) requires the licensee or other entity to
define in its procedures the term ``human error.'' This term may take
on various meanings
[[Page 15743]]
and it is not defined in the current or final rule, so the licensee or
other entity is required to describe or define this term to help ensure
consistent implementation of subpart M of part 26 and that the post-
event test condition is 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,'' is not carried
over to Sec. 26.607(b)(4). Instead, the NRC is prescribing the post-
event test conditions in Sec. 26.607(b)(4), in part so they will not
change unless the NRC amends the requirement.
Section 26.607(b)(5) requires follow-up testing. This requirement
is equivalent to current Sec. 26.405(c)(4), although Sec.
26.607(b)(5) further describes follow-up testing. This final rule
describes 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 will be
used to verify an individual's continued abstinence from substance
abuse. This final rule does 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.
Section 26.607(c) provides additional testing requirements. This
requirement is equivalent to Sec. 26.405(d) and requires
implementation of select requirements from current subpart E of part
26. The requirements govern directly observed collections, shy bladder
situations, special analysis testing, and alcohol testing. These
requirements are necessary to maintain FFD program effectiveness
equivalent to that currently implemented by the LWR fleet.
Section 26.607(c)(1) requires validity testing and establishes the
minimum panel of drugs and drug metabolites to be tested. This panel is
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), Sec.
26.607(c)(1) requires validity testing of urine specimens. Oral fluid
specimens could also be subject to validity testing, including a
biological marker, as specified in either part 26 or the HHS
Guidelines.
Section 26.607(c)(2) includes 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.
Section 26.607(c)(3) requires 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, ``Determining a confirmed positive test
result for alcohol.'' Using the same alcohol testing framework for
parts 50, 52, 70, and 53 licensees and other entities provides 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 are required
for FFD programs under subpart M.
Section 26.607(c)(4) establishes additional testing requirements.
This is equivalent to current Sec. 26.405(f) for facilities licensed
under part 53 for the conduct of drug testing. Unlike Sec. 26.405(f),
Sec. 26.607(c)(4) does not reference validity screening and initial
drug and validity tests at licensee testing facilities. Another minor
difference between Sec. 26.405(f) and Sec. 26.607(c)(4) reflects the
requirement in subpart M of part 26 to use an HHS-certified laboratory
for all biological specimens collected and not just for urine
specimens.
Consistent with Sec. 26.405(f), Sec. 26.607(c)(4) requires the
use of an HHS-certified laboratory for all test conditions listed in
Sec. 26.607(b), MRO-directed tests, and the testing of a split
specimen. Further, HHS-certified laboratory test results using urine or
oral fluid are required for the issuance of an FFD policy violation and
part 26-required sanction.
All drug testing needs 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 will 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 provides
assurance that the drug testing results are accurate and attributed to
the donor. Hair specimens may also be pre-access tested for drugs as
described in Sec. 26.607(i) and positive test results may 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 may not be used for the administration of an FFD policy
violation and sanction, except as provided for in Sec. Sec.
26.607(i)(3) and 26.610(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 is not using in Sec. 26.607(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 is not
used because, under subpart M of part 26, licensees or other entities
are required to use an HHS-certified laboratory. For a laboratory to be
HHS-certified, it must follow the HHS Guidelines and include
[[Page 15744]]
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, is not used because subpart M of part 26 requires the use of
an HHS-certified laboratory.
Section 26.607(c)(4) requires the licensee or other entity to
contract with a primary and backup HHS-certified laboratory. This
provision helps ensure that specimens are processed and tested to
maintain FFD program effectiveness should the primary laboratory be
unable to perform specimen testing. This helps maintain protections
afforded to individuals subject to the FFD program (e.g., should the
donor or MRO request testing of the split specimen, a different
laboratory could be used). This requirement also states that the
primary and backup laboratories must have a different certifying
scientist. Having a back-up HHS-certified laboratory and a different
certifying scientist benefits the program and donor because the drug
testing instruments, technicians, and certifying scientist are
independent of the primary laboratory testing and review process. The
back-up HHS-certified laboratory may be of the same corporate entity as
the primary laboratory.
Section 26.607(c)(4) also states that the laboratory is subject to
inspection or audit by the licensee or other entity and 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 is equivalent to current Sec. 26.41(d), except that
laboratories are not able to limit the use and dissemination of
documents copied or taken from the laboratory by a licensee or other
entity. This is 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.
Section 26.607(d) helps protect the donor from mistakes made during
the drug and alcohol testing processes and helps ensure FFD program
effectiveness. This final rule requires 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 requirement is equivalent
to the first sentence of current Sec. 26.405(e), except that the word
``stringent'' was removed from the phrase ``stringent quality
controls,'' because the word ``stringent'' is not defined.
Section 26.607(e) describes the requirements for licensees and
other entities that use offsite collection facilities. Consistent with
current Sec. 26.405(e), a licensee or other entity will 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 Sec. 26.607(b)(3) and (4). Unlike Sec. 26.405(e), Sec.
26.607(e) does 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 subpart M 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 Sec. 26.606(b)(1) as long as the procedures do not conflict with
the requirements in part 26 or the HHS Guidelines.
Section 26.607(e) also requires 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 requirement is based on current Sec. 26.41(a) and (b). The Sec.
26.607(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 Sec.
26.607(e) must be licensed to conduct specimen collections and perform
alcohol testing, and be audited, by the State or a State-designated
entity. This requirement helps 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).
Section 26.607(f) provides the requirements for initial drug
testing. This provision is equivalent to Sec. 26.405(f) except to
account for the testing of urine and oral fluid specimens under subpart
M of part 26. The 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. Examples of
alternative technologies include liquid or gas chromatography and mass
spectrometry. Another difference from Sec. 26.405(f) is changing the
word ``urine'' in Sec. 26.405(f) to ``biological specimens'' in Sec.
26.607(f). Lastly, Sec. 26.607(f) includes the phrase ``discrepant
biological marker'' as a drug screening result that must be analyzed by
an HHS-certified laboratory and evaluated by the MRO to help inform the
MRO's determination of a subversion attempt.
Section 26.607(g) enables a part 53 licensee to use oral fluid as a
biological specimen for testing. This requirement is 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, Sec.
26.607(g) requires that the licensee's or other entity's procedures
incorporate the HHS Guidelines or the requirements in part 26 for the
conduct of urine or oral fluid testing.
Section 26.607(g) requires 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, must be those established by the HHS
Guidelines for oral fluid drug testing and the alcohol cutoffs in part
26. If they are not established by the HHS Guidelines or this part for
the paneled drugs and drug metabolites, then they will be determined
and documented by a forensic toxicologist review under Sec.
26.31(d)(1)(i)(D).
Section 26.607(g)(2) permits the virtual collection of oral fluid
specimens for drug and alcohol testing but only at facilities that must
use a consortium/third-party administrator to implement random testing
under Sec. 26.607(b)(2)(vi). A virtual collection monitor is permitted
in the location where the specimen collection is to be performed to
assist the virtual collector, such as by completing Federal custody and
control form (Federal CCF) paperwork; observing activities outside the
viewable area of the video
[[Page 15745]]
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.
Section 26.607(i) enables the collection of hair specimens for drug
testing to supplement pre-access testing of urine or oral fluid
specimens. Hair testing is a new feature in the part 26 framework. The
NRC is permitting 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
reports a positive test result, an FFD policy violation may not be
administered. This laboratory information must be treated as
potentially disqualifying FFD information, unless the individual is
determined to have attempted to subvert the testing process, in which
case a permanent denial of authorization must be issued under Sec.
26.610(b)(4). To provide assurance of testing effectiveness and
protections afforded to individuals subject to the FFD program, Sec.
26.607(i) requires that an HHS-certified laboratory must be used to
test the hair specimen. The forensic toxicologist review is necessary
if the panel of drug or drug metabolites to be tested and their cutoffs
are not established by HHS or part 26 for hair.
Section 26.607(j) enables 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 may 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.
Section 26.607(j) also provides that if the portal area screening
instrument detects a substance that exceeds the instrument's
established setpoint, the individual then must be for-cause tested
under Sec. 26.607(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 strengthens the
effectiveness of a licensee's or other entity's BOP. The requirements
do 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
must be used to verify the continuing accuracy of the testing for each
substance tested by the instrument. A portal area screening test can be
used so long as the accuracy of the test result for a specific
substance is 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 is 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
can continue to be used. This performance-based measure balances 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 may be very practicable at facilities that operate with
small staff sizes.
Section 26.607(k) enables 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 is
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, is
important because the MRO must be familiar with the subpart M of part
26 requirements. To help ensure testing effectiveness and protect the
worker, the blood test needs 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.
Section 26.607(l) requires 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 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).
Section 26.607(m) establishes requirements for the licensee- or
other entity-designated MRO. Section 26.607(m)(1) is equivalent to
Sec. 26.405(g), however, the word ``designated'' is added to the first
sentence to clarify that the MRO is designated by the licensee or other
entity, and not by a third party. As stated with regard to Sec.
26.607(k), this change clarifies 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 is consistent with the description of FFD program personnel in
current Sec. 26.31(b) and helps 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 for a part 53
licensee or other entity that implements the FFD program described in
Sec. 26.605(b).
Section 26.607(m)(2) helps 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 is also requiring that if an MRO performed the duties and
responsibilities in Sec. Sec. 26.185 and 26.187 for at least three
continuous years in the last 10 years prior to being hired or
contracted by the licensee or other
[[Page 15746]]
entity, then the MRO would not need to repeat the initial training and
examination requirements. The basis for 3 years is that the MRO has
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.
Section 26.607(m)(3) requires that the MRO attend a medical- or
clinical-based training session every 5 years. This requirement 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 did 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
new requirement is 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 subpart M of part
26.
Section 26.607(m)(4) requires 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 helps ensure FFD program effectiveness and enhances
consistency across the commercial nuclear industry for the evaluation
of drug testing results. This also helps protect individuals because
they are 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 ensures 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 M of part 26.
The Sec. 26.607(m)(4)(ii) requirement also states that an MRO need
not review alcohol test results, including positive confirmatory
alcohol test results determined by an EBT under Sec. 26.607(c)(3)(vi)
and (vii), which are equivalent to the current requirements in
Sec. Sec. 26.101 and 26.103, respectively. Section 26.607(c)(3)(i)
requires the use of an EBT under Sec. 26.91, which ensures that
confirmatory alcohol test results are precise and accurate to issue FFD
policy violations.
Section 26.607(m)(5) requires 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 requirement is 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.
Section 26.607(m)(6) requires that the MRO review all specimen test
results associated with a drug-related FFD policy violation. This
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. The
broad scope of this MRO evaluation is 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 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.
Section 26.607(n) is equivalent to current Sec. 26.31(d)(6) and
establishes limits on the screening and testing of biological
specimens. This is a protection consideration afforded to individuals
subject to the FFD program and was not provided in subpart K of part
26. This requirement states that 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 and tests that may not be conducted include, but are not
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 is requiring that no biological specimens may be passively
sampled and analyzed in a manner different than described in subpart M
of part 26 to ensure workers are protected from non-consensual passive
screening. The subpart M framework enables passive detection of drugs
and alcohol, whereas passive detection is not afforded in subparts A
through I, N, and O of part 26.
Section 26.607(o) is equivalent to current Sec. Sec.
26.31(b)(1)(iii)(A) and 26.89 and requires that all specimen
collections be conducted by a licensee- or other entity-designated and
-trained individual. For subpart M of part 26, this includes onsite
specimen collections, except a collection by a portal area screening
instrument in Sec. 26.607(j).
Section 26.608 requires licensees and other entities to provide FFD
program training to individuals subject to the FFD program. The
performance-based Sec. 26.608 requirement was developed from the
prescriptive training requirements in current Sec. 26.29 and modeled
on current Sec. 50.120 and the requirements in Sec. Sec. 53.725 and
53.830 because there is no training requirement in subpart K of part
26.
Section 26.608(a)(1) requires 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 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 individuals specified
in Sec. 26.4, a licensee or other entity of a commercial nuclear plant
is required to use a SAT, as defined in Sec. 53.725(c)(13). These
requirements are based on requirements in Sec. 26.29(a)(2), (3), (9),
and (10).
Section 26.608(a)(2) requires training on the BOP. This requirement
is based on Sec. Sec. 26.29(a)(8), (9), and (10), and 26.33. The
provision requires 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 also the use of
illicit drugs and substance abuse onsite and offsite. Also, in
reference to impairment from fatigue or any cause if left
[[Page 15747]]
unattended, the phrase in Sec. 26.33, ``may constitute a risk to
public health and safety or the common defense and security,'' is
replaced in Sec. 26.608(a)(2)(iii) with ``could result in
inattentiveness or human errors,'' because subpart M of part 26 is
focused, in part, on ensuring individuals are fit for duty to perform
or direct the performance of assigned duties and responsibilities
safely and competently.
Section 26.608(a)(2)(iv) focuses on training to inform individuals
that they are responsible for their own conduct, as well as observing
others. Specifically, individuals will 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 Sec. 26.606(a)(5) are in the
interest of safety and security because the individual is proactively
announcing that assistance may be necessary. This is 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.
Section 26.608(a)(3) helps ensure that individuals subject to the
FFD program understand that FFD policy violations result in an FFD
program sanction and that program information learned or generated by
FFD program implementation will 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
requirement is equivalent to Sec. 26.29(a)(1). Section 26.608(a)(3) is
a protection measure afforded to individuals subject to the FFD program
because they will understand that licensees and other entities subject
to parts 26 and 73 will 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. FFD-related
information includes 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.
Section 26.608(b) requires individuals to be trained on the FFD
program and to receive a trainee assessment before pre-access testing.
Section 26.608(b) also requires 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 are
similar to Sec. 26.29(c)(1). However, Sec. 26.608(b) was developed
from the SAT-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 is the same as in currently
required SAT-based training programs.
Section 26.608(c) requires 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 addresses the flexibilities afforded in subpart M
of part 26. This section is equivalent to Sec. 50.120(b)(3).
Section 26.609 requires the implementation of a BOP. The
requirement is equivalent to that in Sec. Sec. 26.33 and 26.407,
``Behavioral observation,'' and applies during construction, operation,
and decommissioning, if applicable. Because subpart M of part 26
applies during decommissioning through a licensee's IMP, Sec.
26.609(a) and (b) were developed, in part, from new Sec. 73.100(b)(9)
and current Sec. Sec. 73.55(b)(9) and 73.56(f) to help ensure
consistency in the conduct of behavioral observation whether conducted
for FFD or security purposes.
Under the FFD program, the purpose of the BOP is 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 is accomplished
by requiring each individual subject to subpart M of part 26 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 is not 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 must
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 is not 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 indicate
that the individual is not trustworthy and reliable.
Section 26.609(a) differs from Sec. Sec. 26.33 and 26.407 in that
it places 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.
Section 26.609(b) requires 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. The NRC is requiring 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. Section 26.609(b)
includes 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 includes 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. This better aligns with the Sec. 73.120
BOP requirement, which states that each person subject to behavioral
observation must communicate to the licensee or applicant observed
behaviors or activities of individuals that may constitute an
unreasonable risk to the health and safety of the public and common
defense and security. Section 26.609(a) and (b) were written broadly
[[Page 15748]]
to include offsite conduct that the reporting individual considers
serious enough to call into question the character or reputation of the
subject individual.
Section 26.609(c) requires 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 is the preferred method to ascertain whether the observed
individual can safely and competently perform assigned duties and
responsibilities. When in-person observations are not feasible (e.g.,
during solo operations), the requirement enables the use of video
monitoring. This is addressed, for example, in Sec. 26.609(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 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. Enabling video and audible streaming and capture to enhance the
BOP is consistent with the security-related behavioral observation
requirement in Sec. 73.120(c)(2)(ii), which also enables video
conferencing or other acceptable electronic means promoting face-to-
face interaction for those individuals working remotely.
Section 26.609(d) requires that licensees or other entities perform
behavioral observation of NRC-licensed operators who manipulate the
controls of any commercial nuclear plant licensed under part 53,
remotely by live video and audible streaming capture for those part 53
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 is similar to that of
Sec. 26.609(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.
Section 26.610(a) is similar to Sec. 26.409, ``Sanctions,'' and
requires the licensee or other entity to 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 subpart M of part 26. To be consistent with Sec. 26.75,
``Sanctions,'' the severity of the sanction as described in Sec.
26.610(b) escalates with the number of occurrences and severity of the
FFD policy violation. The sanction is 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.
Equivalent to Sec. 26.75(c), Sec. 26.610(b)(3) also requires a
minimum 5-year denial of access to the NRC-licensed facility for
certain violations of the FFD policy within the protected area of a
commercial nuclear plant and by an individual or individuals who are
the operators of the conveyance to transport or use formula quantities
of strategic SNM. Equivalent to Sec. 26.75(b), Sec. 26.610(b)(4)
requires a permanent denial of authorization be issued for any
subversion attempt.
Section 26.611 protects information collected from FFD program
implementation and is equivalent to current Sec. 26.411, ``Protection
of information.'' The protected information includes, but is not
limited to, privacy and medical information. Section 26.611 does 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 is unnecessary in light of the Sec.
26.611(a) requirement that licensees and other entities must establish
and maintain a system of files and procedures to prevent unauthorized
disclosure.
Section 26.611(b), although equivalent to Sec. 26.411(b), requires
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 Sec.
26.607(b)(1), unlike Sec. 26.411(b)). The purpose of this requirement
is 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 collection,
MRO reviews, and other FFD program elements. Like the consent required
by Sec. 26.411(b), the consent authorizes disclosure of the collected
information. Consent is not 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 Sec. 26.613, ``Appeals
process.''
Section 26.613 is equivalent to Sec. 26.413, ``Review process.''
The title was changed to an appeal process to clarify that Sec. 26.613
is the process implemented when an individual elects to appeal a
licensee or other entity determination that the individual had violated
the FFD policy. The provision also requires 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 is establishing this
requirement because operating experience demonstrates that workers may
not be protected from a continuous review process that does not result
in an outcome.
Section 26.615 requires licensees and other entities to perform
audits of the FFD program. The section is similar to Sec. 26.415,
``Audits.'' Under Sec. 26.615(a), audits are performed at a frequency
that ensures the FFD program's continuing effectiveness. Corrective
actions will be taken as soon as reasonably practicable to resolve any
problems identified and preclude recurrence. Section 26.615(b) requires
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 Sec.
26.617(d) and unsatisfactory performance or programmatic weaknesses
identified under Sec. 26.617(b)(3) and (e).
Section 26.615(c) is equivalent to Sec. 26.415(b) and enables
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.
Section 26.615(d) is equivalent to Sec. 26.415(c) by establishing
requirements for the auditing of HHS-certified laboratories. Unlike
Sec. 26.415(c), the new requirement does not contain a reference to
the U.S. Department of Transportation drug and alcohol testing
requirements. This broadens the regulatory flexibility afforded to a
[[Page 15749]]
licensee or other entity in that they may use an offsite collection or
testing facility that does not meet the Department of Transportation
requirements.
Section 26.615(d) states 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 affords 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 subpart M of part 26,
a licensee or other entity may still use (and not audit) the HHS-
certified laboratory because the Sec. 26.603(e) change control process
maintains FFD program effectiveness.
To help ensure FFD program effectiveness, Sec. 26.615(d) also
requires that collection facility procedures are comparable to those
required in subpart E of part 26, including a 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 is equivalent to that required in Sec.
26.41(b) and helps ensure that the specimen collection process at the
facility remains effective.
Section 26.617 establishes recordkeeping, reporting, and FFD
program performance requirements similar to those in current Sec.
26.417. However, Sec. 26.617 requires 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.
Section 26.617(b)(1) is identical to the reporting requirements in
Sec. 26.417(b)(1) regarding the licensee's or other entity's FFD
program.
Section 26.617(b)(2) requires the reporting of annual (i.e.,
January through December) FFD program performance data for each FFD
program subject to subpart M. Licensees and other entities must submit
the program performance data to the NRC before March 1 of the following
year. This reporting is 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
are 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 M FFD Program'' and 894, ``Annual Reporting Form, 10
CFR part 26, subpart M FFD Program'').
Section 26.617(b)(3) requires 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
Sec. 26.613, or matters that could adversely reflect on the integrity
of the random selection or random testing process. Licensees and other
entities must describe in the reports the incident and any corrective
actions taken or planned.
Section 26.617(c) requires that FFD-related information be shared
within the commercial nuclear industry when requested to support
authorization determinations. This requirement helps 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 must be shared
when requested.
Section 26.617(d) requires licensees and other entities to analyze
FFD program performance data at least annually and take appropriate
actions to correct any identified program weakness.
Section 26.617(e) requires licensees and other entities to
document, trend, and correct non-reportable indicators of FFD
programmatic weaknesses under the licensee or other entity's corrective
action program. However, to protect individual privacy, drug and
alcohol test results may not be tracked in a manner that would permit
the identification of any individuals.
Section 26.619 requires licensees or other entities to establish a
process to evaluate individuals when their fitness or trustworthiness
and reliability are in question. Section 26.619 is equivalent to Sec.
26.419, ``Suitability and fitness determinations,'' but, unlike Sec.
26.419, applies during the construction and operation phases. Also,
Sec. 26.619 requires that a suitability or fitness determination
conducted for cause be conducted face-to-face. This requirement is
based on current Sec. 26.189(c); however, unlike Sec. 26.189(c),
Sec. 26.619 does not prohibit augmenting determinations via electronic
means of communication (i.e., provides sufficient visual and aural
clarity to complete the process). Instead, Sec. 26.619 explicitly
permits determinations to be performed via electronic means and
explains 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 (PNNL), DOE, 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.\9\ 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 part
53 (i.e., those determinations and assessments performed in accordance
with Sec. 26.619, Sec. 26.207, or Sec. 26.211) may be augmented via
electronic communications. Such remotely conducted determinations and
assessments are 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 Sec. 26.608 and Sec. 26.202(c). Permitting the use of
electronic communications helps ensure FFD program effectiveness,
especially in instances where the part 53 commercial nuclear plant is
sited in a geographically remote location, when the facility has a
small staff size, and when an urgent determination is required.
---------------------------------------------------------------------------
\9\ PNNL, Technical Letter Report, ``The Use of Electronic
Communications to Perform Determinations of Fitness,'' dated August
2017.
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[[Page 15750]]
D. Changes to Part 26, Subpart N
Section 26.709 makes the recordkeeping and reporting requirements
in subpart N of part 26 applicable to licensees and other entities of
facilities licensed under part 53 that elect not to implement the
requirements in subpart M of part 26 or elect to implement the
requirements in Sec. 26.605(b).
This final rule amends Sec. 26.711(c) and (d) to make these
requirements applicable to licensees or other entities described in
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.
E. Changes to Part 26, Subpart O
Most of the changes to part 26 are new or revised substantive
provisions that establish a regulatory obligation or prohibition or are
conforming edits to reflect the addition of part 53. The only new
provision that is not substantive, such that violation of it would not
result in a criminal penalty, is Sec. 26.601. Therefore, the NRC is
adding Sec. 26.601 to the list of regulations in Sec. 26.825(b) to
which criminal sanctions do not apply.
10 CFR Part 50
A. Section 50.160: Emergency Preparedness for Small Modular Reactors,
Non-Light-Water Reactors, and Non-Power Production or Utilization
Facilities
This final rule revises Sec. 50.160(b)(3) and (c)(2) to make that
section applicable to applicants and licensees under part 53. Section
50.160 provides an alternative to other part 50 emergency preparedness
requirements focused on large light-water reactors to provide an
optional emergency preparedness framework specifically for small
modular reactors (SMRs) and other new technologies. These alternative
emergency preparedness requirements adopt a performance-based,
technology-inclusive, risk-informed, and consequence-oriented approach.
Commercial nuclear reactor applicants complying with Sec. 50.160 must
submit as part of the application the analysis used to determine
whether the criteria in Sec. 53.1109(g)(2)(i)(A) and (B) are met and,
if they are met, the size of the plume exposure pathway emergency
planning zone (EPZ). An EPZ bounds the area surrounding a facility
within which detailed planning is needed to implement predetermined,
prompt protective actions. The criterion in Sec. 53.1109(g)(2)(i)(A)
is that public dose, as defined in Sec. 20.1003, is projected to
exceed 10 mSv (1 rem) TEDE over 96 hours from the release of
radioactive materials from the facility considering accident likelihood
and source term, timing of the accident sequence, and meteorology. The
criterion in Sec. 53.1109(g)(2)(i)(B) is that pre-determined, prompt
protective measures are necessary. These are the same criteria that are
in Sec. 50.33(g)(2)(i)(A) and (B) and are used to assess the need for
and size of an EPZ in applications under parts 50 and 52.
Applicants choosing to comply with Sec. 50.160 must determine the
radiological releases from the facility that are evaluated in the
determination of the plume exposure pathway EPZ. Applicants should
consider quantitative and qualitative information on the potential
radiological releases that make up the spectrum of accidents used to
develop the basis for the applicant's site-specific EPZ. This
information is derived from the licensing basis. The NRC plans to
update the risk-informed approach in RG 1.242 for part 53 while
maintaining its flexibility for using information already developed and
available in licensing-basis documents, including PRA results,
deterministic dose quantities, accident timing, target set analyses,
mitigation capabilities, and site-specific factors such as meteorology.
Applicants choosing to comply with Sec. 50.160 must determine the
radiological releases from the facility that are evaluated in the
radiological dose assessment to inform the determination of the plume
exposure pathway EPZ size. In its Safety Analysis Report, the applicant
will describe the LBEs relevant to the facility and consider these LBEs
as candidates for the spectrum of accidents used to develop the site-
specific EPZ. The LBEs assessed include a wide range of events that are
appropriate for considering in the facility's emergency preparedness
and response planning. In addition, Sec. 50.160(b)(1)(iv)(A)(2)
requires licensees to be capable of implementing their approved
emergency response plan in conjunction with their safeguards
contingency plan.
An appropriate EPZ and pre-determined, prompt protective measures
are elements of an effective emergency plan. The EPZ size is primarily
informed by the consequences and release characteristics of the LBEs
derived from the safety case. Each licensee should ensure that its
emergency plan documents the onsite protection strategies and, as
warranted, off-site preparedness capabilities to reasonably respond to,
monitor, and protect against the potential events associated with the
facility. The characteristics of seismic and security events should be
provided in the analysis required by Sec. 53.1109(g)(2), but the
calculated dose consequences may be, but do not need to be, explicitly
considered in the EPZ size determination in the same manner as LBE
consequences, if the consequences from these events are less than the
consequences from the LBEs. Rather, the characteristics of these events
(consequence, timing, radionuclides of release) may be discussed and
used to justify that the EPZ size and pre-determined, prompt protective
measures to address the LBEs are sufficient to ensure that capabilities
exist to reduce consequences of those events.
Part 53 applicants and licensees should consider security events in
their EPZ-sizing analysis under Sec. 53.1109(g)(2). If any such events
lead to consequences greater than licensing-basis events already being
considered in the EPZ size justification and would warrant preplanned
prompt protective measures, then the applicant or licensee should
include the security event(s) in its EPZ-sizing analysis or provide an
adequate alternate method(s) for addressing them. If any of the events
do not lead to consequences greater than licensing-basis events already
being considered in the EPZ size justification or would not warrant
preplanned prompt protective measures, then the applicant or licensee
would not need to include those events in the EPZ-sizing analysis.
B. Appendix B to Part 50: Quality Assurance Criteria for Nuclear Power
Plants and Fuel Reprocessing Plants
This final rule amends appendix B to part 50 to make it applicable
to applicants and licensees under part 53. This results in the need for
some revisions to recognize differences in terminology between parts 50
and 53. Namely, the term ``design bases,'' which is defined in Sec.
50.2, is not used in part 53. For this reason, this final rule adds
text in both section III, ``Design Control,'' and section IV,
``Procurement Document Control,'' to refer to ``functional design
criteria, as defined in Sec. 53.020,'' as the part 53 equivalent of
the term ``design bases.''
C. Appendix E to Part 50: Emergency Planning and Preparedness for
Production and Utilization Facilities
This final rule amends appendix E to part 50 to make it applicable
to
[[Page 15751]]
applicants and licensees, under part 53, that choose to comply with the
requirements in appendix E to part 50 and the planning standards of
Sec. 50.47(b) in accordance with Sec. 53.855. Because the regulations
contained in Sec. 50.160 are not applicable to large LWR designs, NRC
has revised appendix E to recognize the applicability of the appendix
to part 53 applicants and licensees. The conforming changes made to
appendix E allow its use in conjunction with Sec. 50.47 to provide
emergency preparedness framework for large LWR or other designs under
part 53.
10 CFR Part 73
A. Section 73.100: Technology-Inclusive Requirements for Physical
Protection of Licensed Activities at Commercial Nuclear Plants Against
Radiological Sabotage
Section 73.100 provides a performance-based regulatory framework
for the design, implementation, and maintenance of a physical
protection program and security organization for certain commercial
nuclear plants licensed under part 53. The current Sec. 73.55 physical
security requirements for nuclear power reactors licensed under part 50
and part 52 use a combination of performance criteria (e.g., Sec.
73.55(b)(1) through (3)) and numerous prescriptive requirements
developed to achieve performance objectives (e.g., Sec.
73.55(k)(5)(ii)). By contrast, in the performance-based approach to
physical security for part 53, performance objectives and requirements
are the primary bases for regulatory decision-making, giving the
licensee the flexibility to determine how to demonstrate compliance
with the established performance criteria for an effective physical
protection program. This physical protection program provides
reasonable assurance that activities involving SNM are not inimical to
the common defense and security and do not constitute an unreasonable
risk to the public health and safety.
Section 73.100(a) requires each part 53 licensee that elects to
demonstrate compliance with this section rather than Sec. 73.55 to
implement the requirements therein through a graded approach based on
achievability of target sets. For licensees that identify achievable
target sets, all of the requirements of Sec. 73.100 apply, and must be
implemented prior to initial fuel load into the reactor (or, for a
fueled manufactured reactor, before initiating the removal of the
features to prevent criticality required under Sec. 53.620(d)(1)). For
licensees that demonstrate they have no achievable target sets, the
requirements are further graded based on whether that demonstration
relies on the implementation of active measures, such as operator
action, mitigative action, detection, assessment, or armed response.
Licensees that do not rely on active measures are exempt from the
remaining requirements of Sec. 73.100 (but must still meet physical
protection requirements for SNM or radioactive material, as
applicable). Licensees that rely on active measures may limit the scope
of their physical protection program by ensuring that the credited
active measures will be implemented when needed in response to threats.
Based on experience from recent new reactor licensing reviews, the
NRC recognizes that licensees may seek to receive unirradiated fuel
onsite before carrying out the security requirements in Sec. 73.100.
However, these security requirements must be implemented at some point
before reactor operation to address the increased risk arising from
irradiated fuel onsite. This final rule makes clear that part 53
applicants and licensees using Sec. 73.100 may bring unirradiated
nuclear fuel onsite and protect it in accordance with the NRC's
requirements for physical protection of SNM of moderate and low
strategic significance under Sec. 73.67 until initial fuel load into
the reactor (or, for a fueled manufactured reactor, until initiating
the removal of the features to prevent criticality required under Sec.
53.620(d)(1)).
Section 73.100(b) outlines the general performance objective and
design requirements of the licensee physical protection program. A
licensee's program is required to provide protection against any
deliberate act within the design-basis threat (DBT) of radiological
sabotage, including spent fuel sabotage, that could directly or
indirectly endanger the public health and safety by exposure to
radiation. The physical protection program is supported by the AA
program, cybersecurity program, and IMP to demonstrate compliance with
the general performance objective of Sec. 73.100(b).
Section 73.100(b)(2) was developed, in part, from Sec.
73.55(b)(3). To satisfy the general performance objective of Sec.
73.100(b)(1), the physical protection program must protect against the
DBT of radiological sabotage. The existing fleet of LWR satisfies this
objective by preventing significant core damage and spent fuel
sabotage. Some non-LWR reactor licensees' physical protection programs
may be designed to prevent a significant release of radionuclides from
any source. Therefore, the performance objective focuses on
radiological sabotage in general, rather than a specific focus on core
damage or spent fuel sabotage, to be technology-inclusive and allow for
flexibility for different reactor technologies.
Under Sec. 73.100(b)(2)(ii), licensees must provide defense in
depth in achieving performance requirements through the integration of
engineered systems, administrative controls, and management measures.
This requirement applies defense-in-depth concepts as part of the
physical protection program to ensure the capability to demonstrate
compliance with the performance objective of Sec. 73.100(b)(1) is
maintained in the changing threat environment. The defense-in-depth
philosophy applies to measures against intentional acts as required by
Sec. 73.100(b), and the designs of physical security systems should
employ defense in depth through systems diversity, independence, and
separation under Sec. 73.100(b)(2). The most common defense-in-depth
measures apply concepts of redundancy, diversity, independence, and
safety margin to ensure systems reliability and availability. The
defense-in-depth philosophy applies to the design of a physical
protection program, which integrates engineered controls and
administrative controls, to provide protection against the DBT for
radiological sabotage.
Section 73.100(b)(3) requires a physical protection program that
prevents the release of radionuclides from any source from exceeding
the dose reference values defined in Sec. 53.210 of this chapter. Dose
reference values are intended to assess the performance of systems for
design basis scenarios. These values were not originally designed for
application to security events. However, because of the analogous
nature of the design basis accident and design basis threat concepts,
the application of dose reference values to design basis security
events is a logical extension of this well-established NRC licensing
tool. There are two dose reference values, but typically the 2-hour 25
rem TEDE value is the most limiting and will, therefore, be the focus
of an applicant's assessment. Although this provides less prescriptive
defense in depth in achieving performance requirements, the 2-hour 25
rem dose reference value remains protective of public health and
safety.
A Part 53 applicant or licensee could voluntarily choose to
establish or maintain a physical protection program
[[Page 15752]]
that prevents significant core damage or spent fuel sabotage (in other
words prevent consequences in excess of the DBA consistent with 10 CFR
73.55), and this program will meet the new performance metric without
additional analyses.
The NRC notes that the 25 rem TEDE reference dose was originally
introduced as a screening criterion in proposed Sec. 53.860 (89 FR
86918), such that below 25 rem TEDE licensees would not be required to
meet the provisions of Sec. Sec. 73.55 or 73.100. For the final rule,
the NRC relocated the 25 rem TEDE reference dose to the requirements of
Sec. 73.100(b)(3), in response to public comments. The Commission's
use of the 25 rem TEDE reference value for this assessment does not
imply that the Commission considers it to be an acceptable limit for a
security event, but only that it represents a reference value to be
used for evaluating plant features and site characteristics.
Section 73.100(b)(4) requires the physical protection program to be
designed and implemented to achieve and maintain the reliability and
availability of SSCs required for demonstrating compliance with
specified performance requirements. These physical protection
performance requirements were informed by Sec. 73.55(b) and the
Commission's Advanced Reactor Policy Statement.
The performance objective of protecting against the DBT of
radiological sabotage is achieved by the design and implementation of
the physical protection program, maintained at all times, with the
following required performance capabilities in the provisions in Sec.
73.100(b)(4): intrusion detection, intrusion assessment, security
communication, security response, protecting against land and
waterborne vehicle bomb assaults, and access control portals. The
physical protection program must maintain the reliability and
availability of SSCs relied upon for demonstrating compliance with the
performance requirements. The terms ``reliability and availability''
are intended to describe defense in depth in a performance-based manner
and are critical elements for demonstrating compliance with the
requirement for protection against the DBT of radiological sabotage as
described in Sec. 73.100(b)(2).
The first element, ``intrusion detection,'' is provided through the
use of detection equipment, patrols, access controls, and other program
elements and provides notification to the licensee that a potential
threat is present and where the threat is located.
The second element, ``intrusion assessment,'' provides a mechanism
through which the licensee identifies the nature of the threat
detected. This is accomplished through the use of video equipment,
patrols, and other program elements that provide the licensee with
timely information about the threat for use in determining how to
respond.
The third element, ``security communication,'' provides a mechanism
through which the licensee communicates the necessary information to
the response force to ensure effectiveness of the physical protection
program. This is accomplished through the redundant, independent, and
diverse design of physical security and/or plant SSCs relied on for
onsite and offsite security communications. The continuity and
integrity of communications should account for the DBT's ability to
affect the reliability and availability of communications.
The fourth element, ``security response,'' provides a mechanism
through which the licensee is capable of timely security response to
interdict and neutralize threats up to and including the DBT of
radiological sabotage. The security response may include the use of
onsite armed responders, law enforcement responders (local, State, or
Federal), or other offsite armed responders (e.g., licensee proprietary
or contract security personnel who are positioned offsite), or a
combination thereof, as appropriate.\10\ The licensee must provide
protection against any element of the DBT, to include those that do not
rise to the full capability of the DBT. Structures, systems, and
components relied on to provide delay functions must be designed to
provide for timely response to adversary attacks with adequate defense
in depth. Delay allows the licensee to take necessary actions to
counter any attempt by the threat to advance toward the protected
target or target set element. The overall response objective is to
place the threat in a condition from which the threat no longer has the
potential for, or capability of, doing harm to the protected target.
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\10\ The NRC's security regulations for commercial nuclear power
reactors have historically considered onsite armed responders to be
the only acceptable method for interdicting and neutralizing threats
up to and including the DBT of radiological sabotage. This final
rule permits advanced power reactor licensees to use any
interdiction and neutralization method, which is an extension of the
Commission's position in SRM-SECY-17-0100, ``Security Baseline
Inspection Program Assessment Results and Recommendations for
Program Efficiencies,'' dated October 8, 2018, and SRM-SECY-20-0070,
``Technical Evaluation of the Security Bounding Time Concept for
Operating Nuclear Power Plants,'' dated June 6, 2024. Under this
final rule, a licensee retains the responsibility to detect, assess,
interdict, and neutralize threats up to and including the DBT of
radiological sabotage, but may rely on law enforcement or other
offsite armed responders as a method for fulfilling the required
interdiction and neutralization capabilities. For licensees that
choose to rely on law enforcement to fulfill these capabilities,
this final rule does not create any NRC regulatory jurisdiction
over, or requirements for, law enforcement.
---------------------------------------------------------------------------
The fifth element, ``protecting against land and waterborne vehicle
bomb assaults,'' provides a mechanism through which the licensee is
capable of protecting the plant against the DBT vehicle bomb assault.
The methods that are relied on to protect against a DBT land vehicle
and waterborne vehicle bomb assault must be designed to protect the
reactor building, structures containing safety or security-related
systems, and components from explosive effects.
The sixth element, ``access control portals,'' provides a mechanism
through which the licensee is capable of detecting and denying
unauthorized access to persons and pass-through of contraband materials
(e.g., weapons, incendiary devices, explosives) to protected areas.
Integrity of the access control system is maintained through licensee
oversight and ensures that attempts to circumvent or bypass the
established process will be detected and access denied.
The performance requirements permit the applicant or licensee to
determine how to design the physical protection program to protect the
plant against the DBT of radiological sabotage without prescriptive
requirements such as those currently found in Sec. 73.55. RG 5.97,
``Guidance for Technology-Inclusive Requirements for Physical
Protection of Licensed Activities at Commercial Nuclear Plants,'' has
been developed by the NRC to describe one acceptable approach to
demonstrate compliance with requirements in Sec. 73.100.
Section 73.100(b)(5) requires the licensee to identify target sets.
Target sets are defined in RG 5.81, ``Target Set Identification and
Development for Nuclear Power Plants,'' Revision 2, as the minimum
combination of equipment, operator actions, and/or structures that, if
all are prevented from performing their intended safety function or
prevented from being accomplished, barring extraordinary actions by
plant operations, would likely result in a release of radionuclides
from any source that would exceed the dose reference values defined in
Sec. 53.210. The licensee must further identify which of its target
sets are ``achievable'' (i.e., those that are within the capabilities
of the DBT and,
[[Page 15753]]
if destroyed or disabled, can lead to a significant offsite release of
radionuclides that cannot be mitigated).
Section 73.100(b)(6) requires that each licensee perform a site-
specific analysis for the purpose of identifying and analyzing site-
specific conditions that affect the design of the onsite physical
protection program.
Section 73.100(b)(7) requires licensees to implement a performance
evaluation program, which ensures that a licensee will periodically
test and evaluate the effectiveness of the physical protection program
to protect against the DBT. This program will ensure that licensees are
able to demonstrate that the physical protection program satisfies the
response requirements of Sec. 73.100 and that the site's protective
strategy effectively protects against the DBT. Licensee performance
evaluations will include methods to assess, test, and challenge the
integration of the physical protection programs functions and
demonstrate the effectiveness of security plans, licensee protective
strategy, and implementing procedures in accordance with Sec.
73.100(g).
Section 73.100(b)(8) requires licensees to implement an AA program
in accordance with Sec. 73.56. Section 73.100(b)(9) requires licensees
to establish, maintain, and implement protection against a cyberattack
based on either the cybersecurity program described in Sec. 73.110 or
the program described in existing Sec. 73.54.
Section 73.100(b)(10) requires an IMP that monitors the initial and
continuing trustworthiness and reliability of individuals granted or
retaining unescorted access or unescorted AA to a protected or vital
area. The IMP must also implement defense-in-depth methodologies to
minimize the potential for an insider (active, passive, or both) to
adversely affect the licensee's capability to protect against
radiological sabotage. Because no one element of the AA program, FFD
program, cybersecurity program, or physical protection program would,
by itself, provide the level of protection against the insider
necessary to demonstrate compliance with the performance objective of
Sec. 73.100(b), the effective integration of these programs is a
necessary requirement to achieve defense in depth against the potential
insider.
Section 73.100(b)(11) requires that the licensee have the
capability to track, trend, correct, and prevent recurrence of failures
and deficiencies in the implementation of the requirements in Sec.
73.100.
Section 73.100(b)(12) requires the coordination of the security
plans and associated procedures with other onsite plans to manage the
safety and security interface during normal or emergency operations.
Section 73.100(b)(13) requires firearms background check
requirements for all members of the security organization whose
official duties require access to covered weapons or who inventory
enhanced weapons.
Section 73.100(c) was developed from Sec. 73.55(c)(7), ``Security
implementing procedures,'' and Sec. 73.55(d), ``Security
organization,'' and outlines the requirements for the composition,
equipping, and training of the security organization. The purpose of
the security organization is to effectively implement the physical
protection program. Individuals assigned to perform physical protection
or contingency response duties must be trained, equipped, and qualified
to perform assigned duties and responsibilities.
Section 73.100(d) establishes a performance requirement for
searches of personnel, vehicles, and materials for the protection
against radiological sabotage. The requirement describes broad
categories of material (explosives, firearms, incendiary devices, etc.)
to be detected and prevented from entry into the protected area;
specific items that will be prohibited are not prescribed in the
regulation but will be stated in the licensee security plans with
detailed descriptions being identified in implementation procedures.
Section 73.100(e) requires a training and qualification program,
described in the training and qualification plan, that ensures
personnel are able to effectively perform their assigned security-
related job duties. This high-level requirement allows flexibility in
how the licensee chooses to train its security personnel. One method
for accomplishing this requirement would be to provide a training and
qualification program that is equivalent to appendix B to part 73.
Section 73.100(f) requires periodic security reviews of the
physical protection program to ensure effective implementation of the
program by independent individuals. The evaluation process provides a
systematized approach for assessing the physical protection program as
a basis for further development and improvement. Program reviews should
be designed to ensure that the physical protection program maintains
effectiveness and demonstrates compliance with NRC requirements.
Section 73.100(f)(1) was developed from Sec. 73.55(m) and requires
review of each element of the physical protection program. Section
73.100(f)(2) requires licensees to perform self-assessments of physical
protection program functions to ensure that the capability to detect,
assess, interdict, and neutralize the DBT of radiological sabotage is
maintained. Section 73.100(f)(3) requires an audit of the effectiveness
of the physical protection program; security plans; implementing
procedures; cybersecurity programs; management of the safety/security
interface activities; the testing, maintenance, and calibration
program; and response commitments by local, State, and Federal law
enforcement authorities. Section 73.100(f)(4) requires that results and
recommendations, management findings, and any actions taken be
documented and maintained to be available for inspection by the NRC.
These reviews are independent of the ongoing performance evaluations
described in Sec. 73.100(b)(7) and (g).
Section 73.100(g) requires that licensee performance evaluations,
described in Sec. 73.100(b)(7), include methods appropriate and
necessary to assess, test, and challenge the integration of the
physical protection program's functions to protect against the DBT. The
performance evaluations must also address the licensee's measures to
protect against cyberattacks, in accordance with the required
cybersecurity plan, and engineered systems designed to protect against
the DBT standalone ground vehicle bomb attack.
Section 73.100(h) establishes performance requirements for
maintaining security SSCs relied on to perform security functions to
protect against the DBT. It requires that corrective actions and
compensatory measures be taken by a licensee in response to a
degradation of security equipment or failure of the equipment to
perform its intended functions. The licensee must maintain the SSCs
described in its design and licensing basis to ensure that they are
reliable and available.
Section 73.100(i) establishes requirements for the suspension of
security measures in response to emergency and extraordinary
conditions. The requirements of this paragraph, which were developed
from Sec. 73.55(p), are intended to provide flexibility to a licensee
for taking reasonable actions that depart from a security plan in an
emergency when such actions are immediately needed to protect the
public health and safety and no action consistent with license
conditions and TS that can provide adequate or equivalent protection is
immediately apparent in accordance with Sec. 53.740(h).
[[Page 15754]]
Section 73.100(j) establishes requirements regarding the
inspection, retention and maintenance of records required to be kept by
the NRC regulations, orders, or license conditions. These requirements
are developed from Sec. 73.55(q).
B. Section 73.110: Technology-Inclusive Requirements for Protection of
Digital Computer and Communication Systems and Networks
Section 53.860(d) requires that a licensee establish, implement,
and maintain a cybersecurity program in accordance with Sec. 73.54 or
Sec. 73.110. Part 53 applicants and licensees may demonstrate
compliance with either of these sections, regardless of whether they
elect to comply with the physical security requirements in Sec. 73.55
or Sec. 73.100.
Section 73.110 establishes requirements for the development and
maintenance of a cybersecurity program for commercial nuclear plants
licensed under part 53. This section implements a graded approach to
determine the level of cybersecurity protection required for digital
computers, communication systems, and networks. The 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 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 to be assessed by the NRC. A graded
approach based on consequences is intended to account for the differing
risk levels among reactor technologies. Specifically, the section
requires licensees to demonstrate protection against cyberattacks in a
manner that is commensurate with the potential consequences from those
attacks.
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, as stated in
10 CFR 53.440(f). This approach ensures considerations are given for
safety and security together throughout the plant's lifetime, including
the design process and prior to implementing changes to plant
configurations, to ensure risks are effectively managed. The
requirements in Sec. 73.110 align with this approach by requiring
licensees to evaluate whether a cyberattack could lead to the
consequences outlined in the rule. This evaluation helps determine
whether enhancements to the design basis or physical protection system
are warranted. Incorporating cybersecurity strategies and design
features early in the design process can be significantly more
efficient and cost-effective than retrofitting these measures after the
plant has been designed or constructed.
Under Sec. 73.110(a), licensees need to ensure that digital
computer and communications systems and networks associated with
safety, security, and emergency preparedness functions are adequately
protected against a potential cyberattack that would result in: (1)
offsite radiation doses that would endanger public health and safety
(i.e., the resulting consequence exceeds the reference dose values in
Sec. 53.210); or (2) adversely impacting \11\ the security functions
necessary to prevent unauthorized removal of material or radiological
sabotage. Security digital assets include those used for nuclear MC&A.
A cyberattack that results in the consequence defined in Sec.
73.110(a)(1) requires the protection of digital assets associated with
safety, security, and emergency preparedness functions. Emergency
preparedness functions are included within the scope of this final rule
because they are essential for recovering from and mitigating the
consequences of radiological sabotage that may result from a successful
cyberattack, as required by Sec. 73.110(d)(2) and (d)(3) (i.e., they
are part of the defense-in-depth strategy). Digital assets associated
with safety-related and non-safety-related but safety-significant
systems that perform or support safety functions are within the scope
of this final rule as these systems are needed to satisfy the safety
criteria in Sec. 53.210 and Sec. 53.220 per Sec. 53.460.
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\11\ As defined in Regulatory Guide 5.71, ``Cybersecurity
Programs for Nuclear Power Reactors,'' Revision 1, adverse impact
means a direct deleterious effect on safety-related, important-to-
safety, security, or emergency preparedness functions; or the
operation of systems, networks, and associated equipment; or the
integrity and confidentiality of data and software. Examples include
loss or impairment of function; reduction in reliability; reduction
in ability to detect, delay, assess or respond to malevolent
activities; reduction of ability to call for or communicate with
offsite assistance; or the reduction in emergency response ability
to implement appropriate protective measures in the event of a
radiological emergency. If the direct or indirect compromise of a
support system causes a safety-related, important-to-safety,
security, or emergency preparedness system or support system to
actuate or ``fail safe'' and not result in radiological sabotage
(i.e., causes the system to actuate properly in response to
established parameters and thresholds), this is not considered to be
an adverse impact.
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Section 73.110(b) requires licensees to protect the communication
system and networks associated with the functions described in Sec.
73.110(a)(1) and (a)(2) from cyberattacks. To accomplish this, the
licensee will establish, implement, and maintain a cybersecurity
program for protecting digital assets within the scope of Sec. 73.110
that makes use of risk insights, including threat information, and
considers the resulting level of consequences of the threats. If the
outcome of the assessment by the licensee under Sec. 73.110(b)(1)
revealed that a potential cyberattack would not compromise any digital
assets that support safety, security, and emergency preparedness
functions and therefore would not result in the consequences listed in
Sec. 73.110(a) (e.g., would not exceed the reference dose values),
then only a narrow set of the cybersecurity program requirements in
Sec. 73.110(d) and (e) would apply. For example, the licensee would
only need to develop a cybersecurity program that implements the
following requirements:
Analyze modifications of any asset before implementation
to demonstrate compliance with the potential consequences in Sec.
73.110(a);
Ensure employees and contractors are aware of
cybersecurity requirements and have some level of cybersecurity
training;
Evaluate and manage cybersecurity risks to the plant;
Review the cybersecurity plan for any required changes;
and,
Retain records of the cybersecurity plan along with any
plan changes.
Section 73.110(c) through (e) were developed from Sec.
73.54(a)(2), and (c) through (h), respectively.
The requirements address the need for the licensee to develop a
cybersecurity program that implements a defense-in-depth protective
strategy as required by Sec. 73.110(d)(2). A defense-in-depth
protective strategy for cybersecurity is represented by collections of
complementary and redundant security controls that establish multiple
layers of protection to safeguard critical digital assets. Under a
defense-in-depth protective strategy, the failure of a single
protective strategy or security control should not result in the
compromise of safety and security functions.
C. Section 73.120: Access Authorization Program for Commercial Nuclear
Plants
Section 73.120 addresses AA for certain commercial nuclear plants
licensed under part 53. The language in Sec. 73.120 provides an
alternate approach to the existing framework for AA under
[[Page 15755]]
Sec. Sec. 73.56 and 73.57, commensurate with risk and consequences to
public health and safety. It is available to part 53 applicants and
licensees who demonstrate compliance with Sec. 73.100(a)(1)(i). The
requirements in Sec. 73.120 are similar to the existing AA program
elements for those NRC-licensed facilities issued additional security
measures (ASMs) orders and for materials licensees under Sec. 37.21.
Applicants not meeting Sec. 73.100(a)(1)(i) will need to establish,
implement, and maintain a full AA program, including an IMP, in
accordance with Sec. 73.56.
Section 73.120(a) is based on an applicant demonstrating that they
do not exceed the dose reference values defined in Sec. 53.210, as
demonstrated through compliance with Sec. 73.100(a)(1)(i). Section
73.120(b) identifies the categories of individuals who are subject to
an AA program in accordance with this section. The applicability
statement in Sec. 73.120(b)(1)(i) encompasses individuals whom the
licensee intends to grant unescorted access to the facilities' most
sensitive areas, consistent with Sec. 73.56(b)(1)(i) for power
reactors and the ASM orders and license conditions issued to any NRC-
licensed facility or material licensee. Sections 73.120(b)(1)(ii)
through (iv) are consistent with Sec. 73.56(b)(1)(ii) through (iv),
respectively. The program will include individuals who may be onsite or
offsite (e.g., remote operators or information technology staff) and
have virtual access to important plant operational and communication
systems based upon assigned duties and responsibilities. An individual
who has remote access to plant equipment and communication systems may
have trusted privileges greater than the personnel at the plant site.
Section 73.120(b)(1)(iii) states that offsite law enforcement personnel
on official duty are not subject to the licensee AA program.
Section 73.120(c) provides general performance objectives and
requirements largely consistent with the AA program requirements for
nuclear power reactors under Sec. 73.56 and provides licensees and
applicants the flexibility in establishing their AA program to
demonstrate compliance with various performance objectives.
Section 73.120(c)(1) includes background investigation requirements
consistent with Sec. 37.25, as well as ASMs and license conditions
that are applied to non-power reactor licensees. Background
investigations include important elements to establish the
trustworthiness and reliability of an individual, such that they do not
constitute an unreasonable risk to public health and safety or the
common defense and security. These include the following: (1) personal
history disclosure, (2) verification of true identity, (3) employment
history evaluation, (4) unemployment/military service/education, (5)
credit history evaluation, (6) character and reputation evaluation, and
(7) Federal Bureau of Investigation criminal history record check.
Section 73.120(c)(2) establishes behavioral observation
requirements, which are an awareness initiative for recognizing
behaviors adverse to the safe operation and security of the facility
through observing the behavior of others in the workplace and reporting
aberrant behavior or changes in behavior that might reflect negatively
on an individual's trustworthiness or reliability. Maintaining
behavioral observation will assist and/or improve worker safety and
reduce the risk of an insider threat. This requirement in Sec.
73.120(c)(2) is a scaled version of the full BOP required under Sec.
73.56(f).
Section 73.120(c)(2) provides licensees greater flexibility to
implement behavioral observation options for individuals granted
unescorted access to the commercial nuclear plant's protected area.
Such options on reporting questionable behavior may include a program
similar to the Department of Homeland Security's program, ``If you see
something, say something,'' or to a corporate behavioral awareness
program. Commensurate with the potential lower safety and security
risks of a commercial nuclear plant that does not exceed the dose
reference values defined in Sec. 53.210, as demonstrated through
compliance with Sec. 73.100(a)(1)(i), Sec. 73.120(c)(2) does not
require the establishment of a comprehensive training program for
behavioral observation (i.e., initial and refresher training including
knowledge checks) as required for power reactors under Sec. 73.56 and
part 26. Under Sec. 73.120(c)(2)(ii), behavioral observation can be
performed in-person or remotely by video, and identified behavior of
concern must be reported to plant supervision. The remote access
alternative to face-to-face interactions provides substantial
flexibility for licensees and applicants. Any video conferencing or
other acceptable electronic means promoting face-to-face interaction
for those individuals working remotely will demonstrate compliance with
this regulation.
Section 73.120(c)(3) captures and maintains the self-reporting of
legal actions as an essential performance element to enhance the
licensee's behavioral observation initiative similar to the current
requirements under Sec. 73.56(g), assuring that personnel who are
granted and who maintain unescorted access are trustworthy and
reliable.
Section 73.120(c)(4) provides a scalable approach for granting and
maintaining unescorted access. One component not included from Sec.
73.56 is the need for a psychological assessment and reassessment under
Sec. 73.56(e) for granting unescorted access and Sec. 73.56(i)(v)(B)
for individuals who perform one or more of the job functions described
in Sec. 73.120(b)(1)(ii) for maintaining unescorted access. Moreover,
the requirement permits criminal history updates to be completed within
10 years of the last review, compared to the 3- or 5-year
reinvestigation periodicity for personnel at an operating commercial
nuclear plant. In addition, no credit check re-evaluation is required
for these individuals.
The continued need to maintain unescorted access will be evaluated
on an annual basis by the reviewing official. Guidance in RG 5.95,
``Access Authorization Program for Commercial Nuclear Plants,''
specifies that this evaluation should be based on a compilation of
personnel interactions as described in the licensee's or applicant's
policy and procedures for behavioral observation and the maintenance of
an approved AA list.
Section 73.120(c)(5) requires licensees and applicants to determine
when a person no longer requires the need for unescorted access or no
longer satisfies the AA requirement found within this section. Guidance
in RG 5.95 further explains that licensees have the flexibility to
terminate unescorted access to specific areas of the site if
individuals lack the continued need for that access to perform their
duties and responsibilities.
Section 73.120(c)(6) is consistent with the purpose of Sec.
37.23(e) and includes the individual's right to correct and complete
information as required under Sec. 37.23(g). The section includes a
requirement for designating a reviewing official. The language provides
clarity regarding the roles and responsibility of a reviewing official,
who is the only individual authorized to make unescorted access
determinations.
Section 73.120(c)(7) aligns with the corresponding requirements
under Sec. 37.23(f), and Sec. 73.120(c)(8) aligns with the
corresponding requirements under Sec. 37.31. These requirements
encompass the roles and responsibilities for licensees, applicants,
and, if applicable,
[[Page 15756]]
the contractor/vendors to establish, implement, and maintain a system
of files and records to ensure personal information is not disclosed to
unauthorized persons.
Section 73.120(c)(9) aligns with the requirements of Sec. 37.33.
Section 73.120(c)(10) requires licensees, applicants, and
contractors or vendors to maintain the records that are required by the
regulations in this section and retain them for a period of 3 years
after the record is superseded or no longer needed. The record
retention period of 3 years is consistent with Sec. 37.23(h),
contrasting with the 5-year retention period under Sec. 73.56(o).
Records maintained in any database(s) must be available for NRC review,
consistent with the requirements found under Sec. 73.56(o)(6)(ii).
V. Opportunities for Public Participation
The NRC published the proposed rule on October 31, 2024 (89 FR
86918), and the comment period was open until December 30, 2024. On
November 22, 2024 (89 FR 92609), the NRC extended the public comment
period by an additional 60 days to February 28, 2025, to allow more
time for members of the public and other stakeholders to develop and
submit their comments.
The NRC hosted two public meetings to engage with external
stakeholders on the proposed rule and associated draft guidance
documents during the public comment period. The first public meeting
was held on November 19, 20, and 21, 2024. The second public meeting
was held on January 8, 2025. A summary of both public meetings is
available in ADAMS, as provided in the ``Availability of Documents''
section. The feedback from these public meetings informed the
development of this final rule.
VI. Public Comment Analysis
The NRC prepared a summary and analysis of public comments
(``Comment Response Document for the Final Rule: Risk-Informed,
Technology-Inclusive Regulatory Framework for Advanced Reactors,''
Volumes I and II) received on the proposed rule and draft RGs, as
referenced in the ``Availability of Documents'' section. In response to
the proposed rule and draft RGs, the NRC received 152 unique comment
submissions. They can be generally separated into the following classes
of stakeholders:
Industry Groups and Licensees--68 comment submissions
Non-Government Organizations--14 comment submissions
States, Tribes, and Local Governments--3 comment submissions
General Public/Individuals--67 comment submissions
The public comment submissions are available from the Federal e-
Rulemaking website at https://www.regulations.gov under Docket ID NRC-
2019-0062. Responses to the public comments, including a summary of how
this final rule and the guidance changed as a result of the public
comments, can be found in the public comment response documents as
indicated in the ``Availability of Documents'' section of this
document.
For more information about the associated guidance documents, see
the ``Availability of Guidance'' section of this document.
VII. Regulatory Flexibility Certification
The Regulatory Flexibility Act of 1980, as amended at 5 U.S.C. 601
et seq, requires that agencies consider the impact of their rulemakings
on small entities and, consistent with applicable statutes, consider
alternatives to minimize these impacts on the businesses,
organizations, and government jurisdictions to which they apply.
In accordance with the Small Business Administration's (SBA's)
regulation at 13 CFR 121.903(c), the NRC has developed its own size
standards for performing an RFA analysis and has verified with the SBA
Office of Advocacy that its size standards are appropriate for NRC
analyses. The NRC size standards at Sec. 2.810, ``NRC size
standards,'' are used to determine whether an applicant or licensee
qualifies as a small entity in the NRC's regulatory programs.
Number of Small Entities Affected
The NRC is currently not aware of any known small entities as
defined in Sec. 2.810 that are planning to apply for a commercial
nuclear plant ESP, CP, OL, ML, or COL under part 53 that would be
impacted by this final rule. Based on this finding, the NRC has
determined that the final rule does not have a significant economic
impact on a substantial number of small entities.
Economic Impact on Small Entities
Although the NRC is not aware of any small entities that are
affected by the final rule, there is a possibility that future
applications for a commercial nuclear plant permit or license could be
submitted by small entities. Commercial nuclear plants of a size
operated by a small entity would most likely be used to support
electrical demand for military bases or small remote towns and would
provide process heat, so they would not directly compete with a larger
commercial nuclear plant that would typically produce electricity for
the grid. As a result of these differing purposes, the NRC would expect
that small and large entities would not be in direct competition with
each other.
Therefore, the NRC concludes that this final rule will not have a
significant economic impact on a substantial number of small entities.
VIII. Regulatory Analysis
The NRC has prepared a final regulatory analysis for this rule. The
analysis examines the costs and benefits of the alternatives considered
by the NRC. The regulatory analysis is available as indicated in the
``Availability of Documents'' section of this document. The conclusion
from the analysis is that this final rule and associated guidance will
result in net averted costs to the industry and the NRC of $152 million
using a 7-percent discount rate and $203 million using a 3-percent
discount rate, using a 66-year analysis period. Detailed information on
the costs and cost savings is presented in Table 1.
Table 1--Total Costs and Cost Savings of Final Rule
[In 2024 dollars]
----------------------------------------------------------------------------------------------------------------
Undiscounted Discounted (7%) Discounted (3%)
----------------------------------------------------------------------------------------------------------------
Attribute Costs
----------------------------------------------------------------------------------------------------------------
Industry Total......................................... $63,823,000 $11,078,000 $25,492,000
NRC Total.............................................. 35,942,000 5,499,000 13,630,000
Net.................................................... 99,765,000 16,577,000 39,122,000
[[Page 15757]]
Annualized............................................. ................. 1,174,000 1,368,000
----------------------------------------------------------------------------------------------------------------
Attribute Cost Savings
----------------------------------------------------------------------------------------------------------------
Industry Total......................................... (346,524,000) (139,576,000) (203,353,000)
NRC Total.............................................. (55,609,000) (28,685,000) (38,582,000)
Net.................................................... (402,133,000) (168,261,000) (241,935,000)
Annualized............................................. ................. (11,915,000) (8,461,000)
----------------------------------------------------------------------------------------------------------------
Attribute Net Cost Savings
----------------------------------------------------------------------------------------------------------------
Industry Net........................................... (282,700,000) (128,500,000) (177,860,000)
NRC Net................................................ (19,670,000) (23,190,000) (24,950,000)
Net.................................................... (302,370,000) (151,690,000) (202,810,000)
Annualized............................................. ................. (10,741,000) (7,093,000)
----------------------------------------------------------------------------------------------------------------
Qualitative Benefits................................... Improvements in Knowledge, Regulatory Efficiency, and
Increased Public Confidence.
----------------------------------------------------------------------------------------------------------------
IX. Backfitting and Issue Finality
This section describes the backfitting and issue finality
implications of this final rule and the final guidance documents
described in section XVIII, ``Availability of Guidance,'' in this
document, as applied to pertinent NRC approvals and certain applicants
that reference NRC approvals in their applications. The NRC's current
backfitting provisions associated with nuclear power plants appear in
Sec. 50.109, ``Backfitting,'' and apply to CPs and OLs under part 50.
Issue finality provisions (analogous to the backfitting provisions in
Sec. 50.109) for approvals under part 52 are located 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 final rule provides a regulatory scheme for entities to apply
for approvals under part 53. The part 50 backfitting provisions and
part 52 issue finality provisions apply to actions taken by the NRC
under part 50 or part 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 or part 52, inextricably affect their
activities regulated under part 50 or part 52. Issuance and
implementation of part 53 will not constitute actions taken under part
50 or part 52. Also, part 53 does not allow an applicant to reference
approvals issued under part 50 or part 52. Therefore, the issuance and
implementation of part 53 will not affect part 50 or part 52 entities'
activities regulated under part 50 or part 52. Therefore, the addition
of part 53 through this final rule is not within the scope of the part
50 backfitting and part 52 issue finality provisions.
The NRC is also making conforming changes to parts 1, 2, 10, 11,
19, 20, 21, 25, 26, 30, 40, 50, 51, 70, 72, 73, 74, 75, 95, 140, 150,
170, and 171 to reflect the addition of part 53. These changes do not
meet the definition of ``backfitting'' in Sec. 50.109 or Sec. 70.76,
``Backfitting,'' because the changes do 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 do not meet the definition of ``backfitting'' in
Sec. 72.62, ``Backfitting,'' because the changes do not add,
eliminate, or modify the SSCs of an independent spent fuel storage
installation (ISFSI) or the procedures or organization required to
operate an ISFSI. These changes do not inextricably affect activities
regulated under parts 50, 52, 70, or 72. Therefore, the changes to
parts 1, 2, 10, 11, 19, 20, 21, 25, 26, 30, 40, 50, 51, 70, 72, 73, 74,
75, 95, 140, 150, 170, and 171 do not constitute backfitting under
parts 50, 70, or 72 or affect the issue finality of an approval under
part 52.
The NRC is issuing nine final guidance documents that provide
guidance on the methods acceptable to the NRC for complying with
aspects of this final rule. Further, as discussed in the guidance
documents, applicants and licensees are not required to comply with the
positions set forth in the guidance. Therefore, the final guidance
documents do not constitute backfitting under part 50 or affect the
issue finality of any approval issued under part 52.
X. Cumulative Effects of Regulation
The NRC seeks to minimize any potential negative consequences
resulting from the cumulative effects of regulation (CER). The CER
describes the challenges that licensees, or other impacted entities
such as State partners, may face while implementing new regulatory
positions, programs, or requirements (e.g., rules, generic letters,
backfits, inspections). The CER is an organizational effectiveness
challenge that may result from a licensee or impacted entity
implementing a number of complex regulatory actions, programs, or
requirements within limited available resources.
The goals of the NRC's CER effort were met throughout the
development of this final rule. The NRC engaged with external
stakeholders at public meetings and solicited public comments on the
proposed rule and associated draft guidance documents. The NRC also
held numerous public meetings prior to publication of the proposed rule
and published numerous versions of preliminary proposed rule language.
Although the use of part 53 is voluntary, the NRC included in the
proposed rule a request for feedback related to CER. Specifically, the
NRC requested feedback on the implementation and potential unintended
consequences of the proposed rule. The NRC received two comment
submissions in response to these CER questions, but no comments
required a change to the rule.
XI. 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
[[Page 15758]]
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).
XII. Environmental Assessment and Final Finding of No Significant
Environmental Impact
The Commission has determined under the National Environmental
Policy Act of 1969, as amended, and the Commission's regulations in
subpart A of part 51, that this rule is not a major Federal action
significantly affecting the quality of the human environment, and,
therefore, an EIS is not required. The basis of this determination
reads as follows: the implementation of the final rule will not have a
significant impact on the environment. The final rulemaking has
requirements that are administrative in application, matters of
procedure, or provide an equivalent level of safety as existing
requirements; therefore, there will be similar environmental impacts
from the implementation of the part 53 regulations as there are for
existing requirements.
The NRC requested the views of States on the draft environmental
assessment on the proposed rule. The NRC received three comment
submissions from States (two comment submissions from the State of New
York and one comment submission from the State of Utah), one of which
commented on the draft environmental assessment. The NRC received three
additional comment submissions related to the draft environmental
assessment in the proposed rule. The NRC addressed the comments from
the States, along with the other comments on the proposed rule, as
discussed in Section VI, ``Public Comment Analysis.'' None of these
comments resulted in changes to the environmental assessment.
The determination of this environmental assessment is that there
will be no significant environmental impacts to the public from this
action. The environmental assessment and finding of no significant
impact are available as indicated under the ``Availability of
Documents'' section.
XIII. Paperwork Reduction Act
This final rule contains new and amended collections of information
that are subject to the Paperwork Reduction Act of 1995 (44 U.S.C. 3501
et seq.). The collections of information were approved by the Office of
Management and Budget, approval numbers 3150-0274 (part 53), 3150-0146
(part 26), 3150-0271 (part 50), 3150-0002 (part 73), 3150-0278 (NRC
Forms 361, 361A, 361N, and 361S), 3150-0104 (NRC Forms 366, 366A, and
366B), 3150-0277 (NRC Form 396), 3150-0276 (NRC Form 398), and 3150-
0272 (NRC Forms 893 and 894). The changes to parts 2, 10, 11, 19, 20,
21, 25, 30, 40, 51, 70, 72, 74, 75, 95, 140, 150, 170, and 171 do not
contain any new or amended collections of information subject to the
Paperwork Reduction Act of 1995. Existing collections of information
were approved by the Office of Management and Budget, approval numbers
3150-0062 (part 11), 3150-0044 (part 19), 3150-0014 (part 20), 3150-
0035 (part 21), 3150-0046 (part 25), 3150-0017 (part 30), 3150-0020
(part 40), 3150-0021 (part 51), 3150-0009 (part 70), 3150-0132 (part
72), 3150-0123 (part 74), 3150-0055 (part 75), 3150-0047 (part 95),
3150-0039 (part 140), and 3150-0032 (part 150).
The burden to the public for these information collections is
estimated to average 2,257 hours per response for part 53, 9 hours per
response for part 26, 4,383 hours per response for part 50, 1,502 hours
per response for part 73, and 2 hours per response for NRC Forms 893
and 894, including the time for reviewing instructions, searching
existing data sources, gathering and maintaining the data needed, and
completing and reviewing the information collection. Other identified
information collections (NRC Forms 361, 366, 396, and 398) are not
estimated to impose burden during the next 3 years.
The information collection is being conducted to evaluate
applications for, issue, and regulate operations under part 53 licenses
and exercise its oversight functions in an effective and efficient
manner to ensure protection of public health and safety, the promotion
of the common defense and security, and the protection of the
environment. Information will be used by the NRC to make decisions
regarding applications and license amendments, assess licensee
compliance with part 53, and take corrective actions as needed.
Responses to this collection of information are mandatory for licensees
choosing to comply with part 53. Confidential and proprietary
information submitted to the NRC is protected in accordance with NRC
regulations at 10 CFR 9.17(a) and 10 CFR 2.390(b).
You may submit comments on any aspect of these information
collections, including suggestions for reducing the burden, by the
following methods:
Federal Rulemaking Website: Go to https://www.regulations.gov and search for Docket ID NRC-2019-0062.
Mail comments to: FOIA, Library, and Information
Collections Branch, Office of the Chief Information Officer, Mail Stop:
T-6 A10M, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001
or to the OMB reviewer at OMB Office of Information and Regulatory
Affairs (3150-0274), Attention: Desk Officer for the Nuclear Regulatory
Commission, 725 17th Street NW, Washington, DC 20503.
Public Protection Notification
The NRC may not conduct or sponsor, and a person is not required to
respond to, a collection of information unless the document requesting
or requiring the collection displays a currently valid OMB control
number.
XIV. Executive Orders
The following are E.O.s that are related to this final rulemaking.
A. Executive Order 12866: Regulatory Planning and Review (as Amended by
Executive Order 14215: Ensuring Accountability for All Agencies)
This action is a significant regulatory action under section 3(f)
of E.O. 12866 and therefore was submitted to OMB for review.
B. Executive Order 14154: Unleashing American Energy
NRC has examined this final 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 a deregulatory action as defined by E.O. 14192. An
E.O. 14192 deregulatory action is defined as ``an action that has been
finalized and has total costs less than zero.'' The final rule and
associated guidance will result in net cost savings to the industry and
the NRC of $152 million using a 7-percent discount rate and $203
million using a 3-percent discount rate, over the 66-year analysis
period. The annualized costs are approximately $1.17 million per year
at a 7 percent discount rate, and $1.37 million per year at a 3 percent
discount rate. The annualized cost savings are approximately $11.9
million per year at a 7 percent discount rate, and $8.46 million per
year at a 3 percent discount rate. Therefore, the net cost savings are
estimated at $10.7 million per year at a 7 percent discount rate and
$7.09 million per year at a 3 percent discount rate. Accordingly, this
final rule has total costs less than zero, and therefore is an E.O.
14192 deregulatory action. Details on the estimated costs of this final
rule can be found in Section
[[Page 15759]]
VIII 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 AEA, the Energy Reorganization Act of 1974, as
amended, or the NWPA; (2) not statutorily required; and (3) not part of
the NRC's permitting regime. The NRC determined that the regulatory
changes in this rule are statutorily required to comply with NEIMA,
necessary for the reasonable assurance of adequate protection of public
health and safety, and part of the NRC's regulatory permitting scheme
authorized by the AEA. Therefore, the NRC views this rulemaking to be
outside the scope of E.O. 14270 and did not insert conditional sunset
dates for the regulatory changes in this final rule.
XV. Congressional Review Act
This final rule is a rule as defined in the Congressional Review
Act (5 U.S.C. 801-808). However, the Office of Information and
Regulatory Affairs in the Office of Management and Budget has found
that it does not meet the criteria at 5 U.S.C. 804(2).
XVI. Criminal Penalties
This final rule includes federal regulations that will be
enforceable by criminal penalty, as authorized by Section 223 of the
AEA. Therefore, per E.O. 14294, these regulations constitute ``criminal
regulatory offenses.''
For the purposes of Section 223 of the AEA, the NRC is issuing this
final rule that will add a new 10 CFR part 53 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 161b, 161i, or 161o of the AEA. Willful violations
of the regulations in these parts will be subject to criminal
enforcement, other than those listed in Sec. 19.40(b), Sec.
20.2402(b), Sec. 21.62(b), Sec. 25.39(b), Sec. 26.825(b), Sec.
30.64(b), Sec. 40.82(b), Sec. 50.111(b), Sec. 53.9010(b), Sec.
70.92(b), Sec. 72.86(b), Sec. 73.81(b), Sec. 74.84(b), Sec.
95.63(b), or Sec. 140.89(b). Criminal penalties as they apply to
regulations in part 53 are discussed in Sec. 53.9010.
XVII. Voluntary Consensus Standards
The National Technology Transfer and Advancement Act of 1995,
Public Law 104-113, requires that Federal agencies use technical
standards that are developed or adopted by voluntary consensus
standards bodies unless the use of such a standard is inconsistent with
applicable law or otherwise impractical. In this final rule, the NRC
will revise its regulations by adding a risk-informed, technology-
inclusive regulatory framework for commercial advanced nuclear
reactors. This action does not constitute the establishment of a
standard that contains generally applicable requirements.
XVIII. Availability of Guidance
As discussed in section II, Background, of this document, the NRC's
development of part 53 built upon activities such as those described in
SECY-19-0117. Because a number of those activities are ongoing to
support new reactor applications under the existing regulatory
framework of 10 CFR parts 50 and 52, the NRC staff identified in its
response to SRM-SECY-20-0032 that the timing of guidance document
development to support the part 53 rulemaking was a key risk and
uncertainty to publishing the final part 53 rule. To mitigate this
risk, the NRC engaged external stakeholders to ensure a common
prioritization of the development of these guidance documents and to
work diligently on those that would be needed to support this
rulemaking, forthcoming applications, or broader efforts such as the
Advanced Reactor Demonstration Program being sponsored by the DOE. The
NRC also recognizes that guidance development to support part 53 and
advanced reactors will continue as the industry and NRC learn lessons
from licensing reviews and operating experience.
The NRC is issuing nine guidance documents for the implementation
of the requirements in this rulemaking. The guidance is available in
ADAMS under the Accession Numbers as indicated under the ``Availability
of Documents'' section in this document.
RG 5.81, Revision 2, ``Target Set Identification and
Development for Nuclear Power Reactors'' (nonpublic)
This regulatory guide (RG) was issued in draft form as Draft
Regulatory Guide (DG)-5071 with the proposed rulemaking on Alternative
Physical Security Requirements for Advanced Reactors (RIN 3150-AK19;
Docket ID NRC-2017-0227) on August 9, 2024. (89 FR 65226). In addition,
some sections from DG-5072, ``Guidance for Alternative Physical
Security Requirements for Small Modular Reactors and Non-Light-Water
Reactors,'' which was also issued with the same rulemaking, have been
incorporated into RG 5.97, ``Guidance for Technology-Inclusive
Requirements for Physical Protection of Licensed Activities at
Commercial Nuclear Plants.'' The changes to these guidance documents
are a result of the NRC's resolution of public comments on the 10 CFR
part 53 proposed rule that requested the NRC address comments made on
the proposed Alternative Physical Security Requirements for Advanced
Reactors rule. As a result, the NRC addressed the public comments
received on those draft guidance documents that were incorporated into
the guidance documents for this final rulemaking. Those comment
responses can be found in ``Comment Response Document for the Final
Rule: Risk-Informed, Technology-Inclusive Regulatory Framework for
Advanced Reactors,'' Volume II.
RG 1.254, Revision 0, ``Technology-Inclusive Identification of
Licensing Events for Commercial Nuclear Plants''
This RG describes an acceptable approach for identifying licensing
events that can be used to inform the design basis, licensing basis,
and content of applications for commercial nuclear plants, including
large LWRs and non-LWRs. It applies to nuclear power reactor designers,
applicants, and licensees of commercial nuclear plants applying for
permits, licenses, certifications, and approvals under parts 50, 52,
and 53. In this RG, the term ``licensing events'' is used in a generic
sense to refer to collections of designated event categories such as,
but not limited to AOOs, DBAs, DBEs, and postulated accidents.
Specifically, this RG provides an acceptable approach for: (1)
conducting a comprehensive and systematic search for initiating events;
(2) using a systematic process to delineate a comprehensive set of
event sequences; (3) grouping initiating events and event sequences
into designated licensing event categories; and (4) providing assurance
that the set of licensing events is complete.
RG 5.95, Revision 0, ``Access Authorization Program for
Commercial Nuclear Plants''
This RG describes a method that the staff considers acceptable to
comply with requirements in Sec. 73.120, ``Access authorization
program for commercial nuclear plants,'' related to an AA program. This
document provides guidance and is one NRC-approved method (not the only
method) for meeting regulatory requirements for part 53. The language
in Sec. 73.120 provides flexibility through availability of the use of
an alternate approach, commensurate
[[Page 15760]]
with risk and consequence to public health and safety, for part 53
applicants who demonstrate compliance with Sec. 73.100(a)(1)(i).
RG 5.96, Revision 0, ``Establishing Cybersecurity Programs for
Commercial Nuclear Plants Licensed Under 10 CFR part 53''
This RG describes an approach the NRC staff deems acceptable for
complying with the Commission's regulations for establishing,
implementing, and maintaining a cybersecurity program at commercial
nuclear plants licensed under part 53. This guidance provides an
approach for meeting the requirements of Sec. 73.110, ``Technology-
inclusive requirements for protection of digital computer and
communication systems and networks.''
RG 5.97, Revision 0, ``Guidance for Technology-Inclusive
Requirements for Physical Protection of Licensed Activities at
Commercial Nuclear Plants''
This RG describes methods and approaches that the NRC staff
considers acceptable for meeting the physical security requirements of
10 CFR part 53 and 10 CFR 73.100.
RG 5.99, Revision 0, ``Fatigue Management for Nuclear
Power Plant Personnel at Commercial Nuclear Plants Licensed Under 10
CFR part 53''
This RG describes methods that the NRC staff considers acceptable
for addressing certain aspects of FFD programs established at
commercial nuclear facilities licensed under part 53. This guidance, in
conjunction with the existing RG 5.73, ``Fatigue Management for Nuclear
Plant Personnel,'' provides comprehensive guidance regarding acceptable
methods for the development and implementation of licensee fatigue
management programs.
The NRC is issuing the following interim staff guidance (ISG)
documents for the implementation of NRC staff review of applications
under the requirements in this rulemaking:
DRO-ISG-2023-01, ``Operator Licensing Programs''
This ISG provides guidance for the review of tailored operator
licensing programs that are submitted for review consistent with the
technical requirements of Sec. 53.730(g). This guidance primarily
addresses the review of operator licensing examination processes to
facilitate the ability of reviewers to assess whether a proposed
approach to the testing of licensed operators and trainees reflects
sound assessment testing practices that are suitable for the screening
of competent licensed operators. Additionally, this ISG provides
further review guidance in other areas such as licensed operator
continuing training and proficiency programs.
DRO-ISG-2023-02, ``Interim Staff Guidance Augmenting
NUREG-1791, `Guidance for Assessing Exemption Requests from the Nuclear
Power Plant Licensed Operator Staffing Requirements Specified in 10 CFR
50.54(m),' for Licensing Commercial Nuclear Plants under 10 CFR part
53''
This ISG provides guidance for the review of customized facility
operator staffing plans that are submitted for review consistent with
the technical requirements of Sec. 53.730(f). This ISG is structured
as a companion document to the existing NUREG-1791 and adapts the
existing HFE-based methodologies of that document for use in the
evaluation of staffing plans that will be submitted within the context
of part 53 facilities. Additionally, this ISG provides further guidance
to address other staffing-related considerations, such as provisions
for engineering expertise.
DRO-ISG-2023-03, ``Development of Scalable Human Factors
Engineering Review Plans''
This ISG applies to the HFE review of applications for OLs, COLs,
DCs, and standard design approvals for commercial nuclear plants
submitted under part 53. The purpose of this ISG is to facilitate NRC
understanding of an acceptable method for developing a scalable (i.e.,
application-specific) plan for the review of these applications for
compliance with applicable HFE requirements. The ISG describes a
process and provides implementation guidance for the NRC to tailor HFE
review plans to each application to achieve an effective and efficient
review.
The NRC has identified future guidance activities that need to be
completed after this final rule is published to support advanced
reactor applications and NRC reviews. This includes issuance of
revisions or part 53-related companions to already available guidance
documents after the final part 53 rule is published.
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.
------------------------------------------------------------------------
ADAMS accession No./web
Document link/Federal Register
citation
------------------------------------------------------------------------
Final Rule Documents
------------------------------------------------------------------------
Federal Register Notice, ``Final Rule: Risk- ML26042A232.
Informed, Technology-Inclusive Regulatory
Framework for Advanced Reactors,'' dated
March, 2026.
``Environmental Assessment for the Final ML26042A231.
Rule--Risk-Informed, Technology-Inclusive
Regulatory Framework for Advanced
Reactors,'' dated March, 2026.
``Regulatory Analysis for the Final Rule: ML26042A230.
Risk-Informed, Technology-Inclusive
Regulatory Framework for Advanced
Reactors,'' March, 2026.
``Comment Response Document for the Final ML26042A229.
Rule: Risk-Informed, Technology-Inclusive
Regulatory Framework for Advanced
Reactors,'' Volume I, dated March, 2026.
``Comment Response Document for the Final ML26042A228.
Rule: Risk-Informed, Technology-Inclusive
Regulatory Framework for Advanced
Reactors,'' Volume II, dated March, 2026.
------------------------------------------------------------------------
Information Collection Documents
------------------------------------------------------------------------
Supporting Statement for Information ML25230A038.
Collection Analysis--10 CFR Part 53.
Supporting Statement for Information ML25230A037.
Collection Analysis--10 CFR Part 26.
Supporting Statement for Information ML25232A004.
Collection Analysis--10 CFR Part 50.
Supporting Statement for Information ML25230A039.
Collection Analysis--10 CFR Part 73.
Supporting Statement for Information ML25230A034.
Collection Analysis--NRC Form 361S.
Supporting Statement for Information ML25230A035.
Collection Analysis--NRC Form 366.
Supporting Statement for Information ML25245A175.
Collection Analysis--NRC Form 396.
Supporting Statement for Information ML25245A176.
Collection Analysis--NRC Form 398.
[[Page 15761]]
Supporting Statement for Information ML25230A036.
Collection Analysis--NRC Form 893 and 894.
Final Rule--Part 26 Burden Tables for Risk- ML25282A045.
Informed, Technology-Inclusive Regulatory
Framework for Advanced Reactors.
Final Rule--Part 50 Burden Tables for Risk- ML25282A044.
Informed, Technology-Inclusive Regulatory
Framework for Advanced Reactors.
Final Rule--Part 53 Burden Tables for Risk- ML25282A046.
Informed, Technology-Inclusive Regulatory
Framework for Advanced Reactors.
Final Rule--Part 73 Burden Tables for Risk- ML25282A043.
Informed, Technology-Inclusive Regulatory
Framework for Advanced Reactors.
NRC Form 361S, ``Part 53 Plant Event ML25230A030.
Notification Worksheet''.
NRC Form 366, ``Licensee Event Report ML25230A031.
(LER)''.
NRC Form 366A, ``Licensee Event Report ML25230A032.
(LER) Continuation Sheet''.
NRC Form 366B, ``Licensee Event Report ML25231A040.
(LER) (Failure Continuation)''.
NRC Form 396, ``Certification of Medical ML25245A149.
Examination by Facility Licensee''.
NRC Form 398, ``Personnel Qualification ML25245A173.
Statement--Licensee''.
NRC Form 893, ``Single Positive Test Form, ML25230A033.
10 CFR Part 26, Subpart M FFD Program''.
NRC Form 894, ``Annual Reporting Form, 10 ML25231A039.
CFR Part 26, Subpart M FFD Program''.
------------------------------------------------------------------------
Regulatory Guidance Documents
------------------------------------------------------------------------
RG 1.254, ``Technology-Inclusive ML25232A005.
Identification Of Licensing Events For
Commercial Nuclear Plants,'' dated March,
2026.
RG 5.81, ``Target Set Identification and ML24229A186.
Development for Nuclear Power Reactors,''
Revision 2, (non-public) dated March, 2026.
RG 5.95, ``Access Authorization Program for ML25232A007.
Commercial Nuclear Plants,'' dated March,
2026.
RG 5.96, ``Establishing Cybersecurity ML25232A008.
Programs For Commercial Nuclear Plants
Licensed Under 10 CFR Part 53,'' dated
March, 2026.
RG 5.97, ``Guidance for Technology- ML25232A009.
Inclusive Requirements for Physical
Protection of Licensed Activities at
Commercial Nuclear Plants,'' dated March,
2026.
RG 5.99, ``Fatigue Management For Nuclear ML25232A010.
Power Plant Personnel At Commercial
Nuclear Plants Licensed Under 10 CFR Part
53,'' dated March, 2026.
------------------------------------------------------------------------
ISG Documents
------------------------------------------------------------------------
DRO-ISG-2023-01, ``Operator Licensing ML25232A011.
Programs,'' dated March, 2026.
DRO-ISG-2023-02, ``Interim Staff Guidance ML25232A023.
Augmenting NUREG-1791, `Guidance for
Assessing Exemption Requests from the
Nuclear Power Plant Licensed Operator
Staffing Requirements Specified in 10 CFR
50.54(m),' for Licensing Commercial
Nuclear Plants under 10 CFR Part 53,''
dated March, 2026.
DRO-ISG-2023-03, ``Development of Scalable ML25232A022.
Human Factors Engineering Review Plans,''
dated March, 2026.
------------------------------------------------------------------------
Other References
------------------------------------------------------------------------
American National Standards Institute https://webstore.ansi.org/
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dated April 15, 2025.
EO 14300, ``Ordering the Reform of the 90 FR 22587.
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29, 2025.
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dated August 16, 1995.
Federal Register notice--Final rule, 74 FR 28112.
``Consideration of Aircraft Impacts for
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12, 2009.
Federal Register notice--Final rule, 73 FR 16966.
``Fitness for Duty Programs,'' dated March
31, 2008.
Federal Register notice--Final rule, 72 FR 49352.
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for Nuclear Power Plants,'' dated August
28, 2007.
Federal Register notice--Final rule, 53 FR 23203.
``Station Blackout,'' dated June 21, 1988.
Federal Register notice--Final rule, 60 FR 36953.
``Technical Specifications,'' dated July
19,1995.
Federal Register notice--Guidance, 84 FR 57554.
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Fluid,'' dated October 25, 2019.
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[[Page 15762]]
Federal Register notice--Policy Statement, 51 FR 30028.
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``Tribal Policy Statement,'' dated January
9, 2017.
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2008.
Federal Register notice--Policy Statement, 76 FR 34773.
``Final Safety Culture Policy Statement,''
dated June 14, 2011.
Federal Register notice--Proposed rule, 87 FR 12254.
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to Decommissioning,'' dated March 3, 2022.
Federal Register notice--Proposed rule, 89 FR 86918.
``Risk-Informed, Technology-Inclusive
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Reactors,'' dated October 31, 2024.
Federal Register notice--Proposed rule; 89 FR 92609.
extension of comment period, ``Risk-
Informed, Technology-Inclusive Regulatory
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Federal Register notice--Public meeting, 86 FR 67669.
``Reporting Requirements for Nonemergency
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(Index).
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Letter to Chairman Hanson, NRC, ``Fourth ML22196A292.
Interim Letter on 10 CFR Part 53
Rulemaking Language,'' dated August 2,
2022.
Letter to Chairman Hanson, NRC, ML21140A354.
``Preliminary Proposed Rule Language For
10 CFR Part 53, Regulation of Advanced
Nuclear Reactors, Interim Report,'' dated
May 30, 2021.
Letter to Chairman Hanson, NRC, ML22040A361.
``Preliminary Rule Language For 10 CFR
Part 53, Subpart F, `Requirements for
Operations,' Interim Report,'' dated
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to the Advisory Committee on Reactor
Safeguards, `Fourth Interim Letter on 10
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Letter to Chairman Rempe, ACRS, ``Response ML22063A012.
to the Advisory Committee on Reactor
Safeguards Letter on Preliminary Rule
Language for 10 CFR Part 53, Subpart F,
`Requirements for Operations,' Interim
Report,'' dated March 30, 2022.
Letter to Chairman Sunseri, ACRS, ``Part ML20311A006.
53, Licensing and Regulation of Advanced
Nuclear Reactors,'' dated November 24,
2020.
Letter to Chairman Svinicki, NRC, ``10 CFR ML20295A647.
Part 53, Licensing and Regulation of
Advanced Nuclear Reactors,'' dated October
21, 2020.
National Library of Medicine, National https://
Institutes of Health, Workshop Summary, www.ncbi.nlm.nih.gov/books/
``The Evolution of Telehealth: Where Have NBK207141/.
We Been and Where Are We Going?,'' dated
November 2012.
NEI 18-04, Rev. 1, ``Risk-Informed ML19241A472.
Performance-Based Technology-Inclusive
Guidance for Non-Light Water Reactors,''
dated August 2019.
NEI 22-05, Rev. 0, ``Technology Inclusive ML24032A237.
Risk Informed Change Evaluation
(TIRICE),'' dated January 2024.
Nuclear Innovation Alliance (NIA), https://
``Clarifying `Major Portions' of a Reactor www.nuclearinnovationallia
Design in Support of a Standard Design nce.org/clarifying-major-
Approval,'' dated April 2017. portions-reactor-design-
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NRC, ``Advanced Reactor Vision and ML16356A670.
Strategy: Safely Achieving Effective and
Efficient Non-Light Water Reactor Mission
Readiness,'' dated December 2016.
NRC, ``A Regulatory Review Roadmap for Non- ML17312B567.
Light Water Reactors,'' dated December
2017.
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Production of Up to Eight Floating Nuclear
Plants,'' dated September 30, 1982.
NRC, ``Report to Congress: Advanced Reactor ML12158A398 (cover letter)
Licensing,'' dated August 2012. ML12153A014 (report).
NRC, ``Risk-Informed and Performance-Based ML21069A003.
Human-System Considerations for Advanced
Reactors,'' dated March 2021.
NRC Form 890, ``Single Positive Test Form'' ML25044A086.
NRC Form 891, ``Annual Reporting for Drug ML26016A656.
and Alcohol Tests''.
NRC Form 892, ``Annual Fatigue Reporting ML22013B250.
Form''.
NRC Public Meeting Summary, ``Public ML25014A024.
Meeting to Discuss the Part 53 Risk-
Informed, Technology-Inclusive Regulatory
Framework for Advanced Reactors
Rulemaking--Proposed Rule'' (November 19,
20, and 21, 2024), dated January 14, 2025.
NRC Public Meeting Summary, ``Public ML25042A010.
Meeting to Discuss the Part 53 Risk-
Informed, Technology-Inclusive Regulatory
Framework for Advanced Reactors
Rulemaking--Proposed Rule'' (January 8,
2025), dated January 24, 2025.
NUREG-0880, ``Safety Goals for Nuclear ML071770230.
Power Plant Operation,'' dated May 1983.
NUREG/CR-2601, ``Technology, Safety and ML14023A046 (package).
Costs of Decommissioning Reference Light
Water Reactors Following Postulated
Accidents'' dated November 1982.
NUREG-1530, Revision 1, ``Reassessment of ML22053A025.
NRC's Dollar Per Person-Rem Conversion
Factor Policy, Final Report,'' dated
February 2022.
NUREG-1860, Volumes 1 and 2, ``Feasibility ML080440170
Study for a Risk-Informed and Performance- ML080440215.
Based Regulatory Structure for Future
Plant Licensing,'' dated December 2007.
[[Page 15763]]
NUREG/BR-0058, Revision 5, ``Regulatory ML17100A480.
Analysis Guidelines of the U.S. Nuclear
Regulatory Commission,'' dated April 2017.
NUREG/CR-5884, ``Revised Analyses of ML14008A187.
Decommissioning for the Reference
Pressurized Water Reactor Power Station,''
dated November 1995.
NUREG/CR-6187, Volume 1, ``Revised Analyses ML14008A186.
of Decommissioning for the Reference
Boiling Water Reactor Power Station,''
dated July 1996.
PNNL, Technical Letter Report, ``The Use of ML18081A607.
Electronic Communications to Perform
Determinations of Fitness,'' dated August
2017.
Pre-decisional DG, ML22276A149.
``Technology[dash]Inclusive,
Risk[dash]Informed, and
Performance[dash]Based Methodology for
Seismic Design of Commercial Nuclear
Plants,'' dated October 3, 2022.
Research Information Letter 2021-04, ML21113A066.
``Feasibility Study on a Potential
Consequence-Based Seismic Design Approach
for Nuclear Facilities,'' dated April 2021.
RG 1.110, Revision 1, ``Cost-Benefit ML13241A052.
Analysis for Radwaste Systems for
Light[dash]Water-Cooled Nuclear Power
Reactors,'' dated October 2013.
RG 1.134, Revision 4, ``Medical Assessment ML14189A385.
Of Licensed Operators Or Applicants For
Operator Licenses At Nuclear Power
Plants,'' dated September 2014.
RG 1.174, ``An Approach for Using ML17317A256.
Probabilistic Risk Assessment in Risk-
Informed Decisions on Plant-Specific
Changes to the Licensing Basis,'' Revision
3, dated January 2018.
RG 1.208, ``A Performance-Based Approach to ML070310619.
Define the Site-Specific Earthquake Ground
Motion,'' dated March 2007.
RG 1.232, ``Guidance for Developing ML17325A611.
Principal Design Criteria for Non-Light-
Water Reactors,'' Revision 0, dated April
2018.
RG 1.233, Revision 0, ``Guidance for a ML20091L698.
Technology-Inclusive, Risk-Informed, and
Performance-Based Methodology to Inform
the Licensing Basis and Content of
Applications for Licenses, Certifications,
and Approvals for Non-Light-Water
Reactors,'' dated June 2020.
RG 1.247, ``Acceptability of Probabilistic ML21235A008.
Risk Assessment Results for Non-Light-
Water Reactor Risk-Informed Activities,''
issued March 2022 for trial use.
RG 5.71, ``Cybersecurity Programs for ML22258A204.
Nuclear Power Reactors,'' Revision 1,
dated February 3, 2023.
RG 5.73, ``Fatigue Management for Nuclear ML083450028.
Power Plant Personnel,'' dated March 20,
2009.
SECY-18-0096, ``Functional Containment ML18115A157.
Performance Criteria For Non-Light-Water-
Reactors,'' dated September 28, 2018.
SECY-19-0117, ``Technology-Inclusive, Risk- ML18311A264 (package).
Informed, and Performance-Based
Methodology to Inform the Licensing Basis
and Content of Applications for Licenses,
Certifications, and Approvals for Non-
Light-Water Reactors,'' dated December
2019.
SECY-20-0032, ``Rulemaking Plan on `Risk- ML19340A056.
Informed, Technology-Inclusive Regulatory
Framework for Advanced Reactors (RIN-3150-
AK31; NRC-2019-0062,' '' dated April 13,
2020.
SECY-20-0070, ``(Redacted) Technical ML20126G265 (package).
Evaluation of the Security Bounding Time
Concept for Operating Nuclear Power
Plants,'' dated November 8, 2021.
SECY-24-0049, ``Proposed Rule: Reporting ML23318A479.
Requirements for Nonemergency Events at
Nuclear Power Plants (RIN 3150-AK71; NRC-
2020-0036),'' dated June 10, 2024.
SECY-93-092, ``Issues Pertaining to the ML040210725.
Advanced Reactor (PRISM, MHTGR, and PIUS)
and CANDU 3 Designs and their Relationship
to Current Regulatory Requirements,''
dated April 8, 1993.
SRM-SECY-10-0121, ``Modifying the Risk- ML110610166.
Informed Regulatory Guidance for New
Reactors,'' dated March 2, 2011.
SRM-SECY-17-0100, ``Security Baseline ML18283A072.
Inspection Program Assessment Results and
Recommendations for Program
Efficiencies,'' dated October 8, 2018.
SRM-SECY-20-0032, ``Rulemaking Plan on ML20276A293.
`Risk-Informed, Technology-Inclusive
Regulatory Framework for Advanced Reactors
(RIN-3150-AK31; NRC-2019-0062),''' dated
October 2, 2020.
SRM-SECY-20-0045, ``Population Related ML22194A885.
Siting Considerations for Advanced
Reactors,'' dated July 30, 2022.
SRM-SECY-98-144, ``Staff Requirements--SECY- ML003753593.
98-144--White Paper on Risk-Informed and
Performance-Based Regulations,'' dated
February 24, 1999.
SECY-23-0021, ``Proposed Rule: Risk- ML21162A095.
Informed, Technology-Inclusive Regulatory
Framework for Advanced Reactors (RIN 3150-
AK31),'' March 1, 2023.
SECY-23-0021, Enclosure 1, ``Draft Federal ML21162A102.
Register Notification''.
SECY-23-0021, Enclosure 2, ``Draft ML21162A104.
Environmental Assessment for the Proposed
Rule--Risk-Informed, Technology-Inclusive
Regulatory Framework for Advanced
Reactors''.
SECY-23-0021, Enclosure 3, ``Draft ML21165A112.
Regulatory Analysis for the Proposed Rule:
Risk-Informed, Technology[dash]Inclusive
Regulatory Framework for Advanced
Reactors''.
SECY-23-0021, Enclosure 4, ``Alternative ML22244A001.
Approaches Considered for Selected Topics
During the Development of 10 CFR Part 53''.
SECY-23-0021, Enclosure 5, ``Estimated ML22304A099 (non-public).
Resources for The Risk-Informed,
Technology-Inclusive Regulatory Framework
For Advanced Reactors Rulemaking''.
Staff Requirements--SECY-23-0021, ML24064A047 (package).
``Proposed Rule: Risk-Informed, Technology-
Inclusive Regulatory Framework for
Advanced Reactors (RIN 3150-AK31),'' March
4, 2024.
------------------------------------------------------------------------
[[Page 15764]]
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-2019-0062. 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-2019-0062); (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.
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, 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 53
Administrative practice and procedure, Antitrust, Backfitting,
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 power plants
and reactors, Operating license, Penalties, Prototype, Radiation
protection, 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,
[[Page 15765]]
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.
10 CFR Part 170
Byproduct material, Import and export licenses, Intergovernmental
relations, Non-payment penalties, Nuclear energy, Nuclear materials,
Nuclear power plants and reactors, Source material, Special nuclear
material.
10 CFR Part 171
Annual charges, Approvals, Byproduct material, Holders of
certificates, Intergovernmental relations, Nonpayment penalties,
Nuclear materials, Nuclear power plants and reactors, Registrations,
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 adopting
the following amendments to 10 CFR chapter I:
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), remove ``50, 52, and 54'' add in
its place ``50, 52, 53, and 54''.
PART 2--AGENCY RULES OF PRACTICE AND PROCEDURE
0
3. The authority citation for part 2 continues to read as follows:
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 Sec. 31001(s), Pub. L. 104-134, 110 Stat. 1321-373
(28 U.S.C. 2461 note).
Sec. 2.1 [Amended]
0
4. In Sec. 2.1, in paragraph (e), remove the phrase ``part 52'' and
add in its place the phrase ``part 52 or 53''.
0
5. In Sec. 2.4, revise the definitions for ``Contested proceeding''
and ``Facility'' to read as follows:
Sec. 2.4 Definitions.
* * * * *
Contested proceeding means--
(1) A proceeding in which there is a controversy between the NRC
staff and the applicant for a license or permit concerning the issuance
of the license or permit or any of the terms or conditions thereof;
(2) A proceeding in which the NRC is imposing a civil penalty or
other enforcement action, and the subject of the civil penalty or
enforcement action is an applicant for or holder of a license or
permit, or is or was an applicant for or holder of a license or permit,
or is or was an applicant for a standard design certification under
part 52 or 53 of this chapter; and
(3) A proceeding in which a petition for leave to intervene in
opposition to an application for a license or permit has been granted
or is pending before the Commission.
* * * * *
Facility means a production facility or a utilization facility as
defined in Sec. Sec. 50.2 and 53.020 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 H of part
53''.
0
7. In Sec. 2.101, revise paragraphs (a)(3)(i), (a)(5), (a)(9)
introductory text, and (a)(9)(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,
subpart A of part 51, and part 53 of this chapter require;
* * * * *
(5) An applicant for a construction permit under part 50 or 53 of
this chapter or a combined license under part 52 or 53 of this chapter
for a production or utilization facility which is subject to Sec.
51.20(b) of this chapter, and is of the type specified in Sec.
50.21(b)(2) or (3); or Sec. 50.22; or part 53, as applicable, of this
chapter, or is a testing facility, may submit the information required
of applicants by part 50, 52, or 53 of this chapter in two parts. One
part shall be accompanied by the information required by Sec.
50.30(f), Sec. 52.80(b), or Sec. 53.1100(f) of this chapter, as
applicable. The other part shall include any information required by
Sec. 50.34(a) and, if applicable, Sec. 50.34a of this chapter; or
Sec. Sec. 52.79 and 52.80(a) of this chapter; or Sec. Sec. 53.1109,
53.1306, 53.1309, and 53.1312 of this chapter; or Sec. Sec. 53.1109,
53.1413, 53.1416, and 53.1419 of this chapter, as applicable. One part
may precede or follow other parts by no longer than 6 months. If it is
determined that either of the parts as described above is incomplete
and not acceptable for processing, the Director, Office of Nuclear
Reactor Regulation, or Director, Office of Nuclear Material Safety and
Safeguards, as appropriate, will inform the applicant of this
determination and the respects in which the document is deficient. Such
a determination of completeness will generally be made within a period
of 30 days. Whichever part is filed first shall also include the fee
required by Sec. 50.30(e) or Sec. 53.1100(e) and Sec. 170.21 of this
chapter and the information required by Sec. Sec. 50.33, 50.34(a)(1),
and 52.79(a)(1) of this chapter; or Sec. Sec. 53.1109, 53.1309, and
53.1416 of this chapter, as applicable, and Sec. 50.37 or Sec.
53.1115, as applicable, of this chapter. The Director, Office of
Nuclear Reactor Regulation, or Director, Office of Nuclear Material
Safety and Safeguards, as appropriate, will accept for docketing an
application for a construction permit under part 50 or 53 of this
chapter or a combined license under part 52 or 53 of this chapter for a
production or utilization facility which is subject to Sec. 51.20(b)
of this chapter, and is of the type specified in Sec. 50.21(b)(2) or
(3), or Sec. 50.22, or part 53, as applicable, of this chapter or is a
testing facility where one part of the application as described above
is complete and conforms to the requirements of part 50 of this
chapter. The additional parts will be docketed upon a determination by
the Director, Office of Nuclear Reactor Regulation, or Director, Office
of Nuclear Material Safety and Safeguards, as appropriate, that it is
complete.
* * * * *
[[Page 15766]]
(9) An applicant for a construction permit for a utilization
facility which is subject to Sec. 51.20(b) of this chapter and is of
the type specified in Sec. 50.21(b)(2) or (3), or Sec. 50.22, or part
53 of this chapter, an applicant for or holder of an early site permit
under part 52 or 53 of this chapter, or an applicant for a combined
license under part 52 or 53 of this chapter, who seeks to conduct the
activities authorized under Sec. 50.10(d) or Sec. 53.1130 of this
chapter may submit a complete application under paragraphs (a)(1)
through (4) of this section, which includes the information required by
Sec. 50.10(d) or Sec. 53.1130 of this chapter. Alternatively, the
applicant (other than an applicant for or holder of an early site
permit) may submit its application in two parts:
(i) Part one must include the information required by Sec.
50.33(a) through (f) or Sec. Sec. 53.1109(a) through (e) and 53.1306
of this chapter, and the information required by Sec. 50.10(d)(2) and
(3) or Sec. 53.1130(a)(2) and (3) of this chapter, as applicable.
* * * * *
0
8. In Sec. 2.104, revise paragraph (a) to read as follows:
Sec. 2.104 Notice of hearing.
(a) In the case of an application on which a hearing is required by
the Act or this chapter, or in which the Commission finds that a
hearing is required in the public interest, the Secretary will issue a
notice of hearing to be published in the Federal Register. The notice
must be published at least 15 days, and in the case of an application
concerning a limited work authorization, construction permit, early
site permit, or combined license for a facility of the type described
in Sec. 50.21(b) or Sec. 50.22 or subpart H of part 53 of this
chapter, as applicable, or a testing facility, at least 30 days, before
the date set for hearing in the notice.\1\ In addition, in the case of
an application for a limited work authorization, construction permit,
early site permit, or combined license for a facility of the type
described in Sec. 50.22 or subpart H of part 53 of this chapter, as
applicable, or a testing facility, the notice must be issued as soon as
practicable after the NRC has docketed the application. If the
Commission decides, under Sec. 2.101(a)(2), to determine the
acceptability of the application based on its technical adequacy as
well as completeness, the notice must be issued as soon as practicable
after the application has been tendered.
* * * * *
\1\ If the notice of hearing concerning an application for a
limited work authorization, construction permit, early site permit,
or combined license for a facility of the type described in Sec.
50.21(b) or Sec. 50.22 or subpart H of part 53 of this chapter, as
applicable, or a testing facility, does not specify the time and
place of initial hearing, a subsequent notice will be published in
the Federal Register which will provide at least 30-day notice of
the time and place of that hearing. After this notice is given, the
presiding officer may reschedule the commencement of the initial
hearing for a later date or reconvene a recessed hearing without
again providing at least 30-day notice.
0
9. In Sec. 2.105, revise paragraph (a) introductory text and
paragraphs (a)(4), (10), (12), and (13), (b)(3) introductory text, and
(b)(3)(i), (ii), and (iv) to read as follows:
Sec. 2.105 Notice of proposed action.
(a) If a hearing is not required by the Act or this chapter, and if
the Commission has not found that a hearing is in the public interest,
it will, before acting thereon, publish in the Federal Register, as
applicable, or on the NRC's website, https://www.nrc.gov, or both, at
the Commission's discretion, either a notice of intended operation
under Sec. 52.103(a) or Sec. 53.1452(a) of this chapter, as
applicable, and a proposed finding that inspections, tests, analyses,
and acceptance criteria for a combined license under subpart C of part
52 or under subpart H of part 53 of this chapter, have been or will be
met, or a notice of proposed action with respect to an application for:
* * * * *
(4) An amendment to an operating license, combined license, or
manufacturing license for a facility licensed under Sec. 50.21(b) or
Sec. 50.22 or under subpart H of part 53 of this chapter, as
applicable, or for a testing facility, as follows:
(i) If the Commission determines under Sec. 50.58 or Sec. 53.1515
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. 53.1515 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);
* * * * *
(10) In the case of an application for an operating license for a
facility of a type described in Sec. 50.21(b) or Sec. 50.22 or part
53 of this chapter, or a testing facility, a notice of opportunity for
hearing shall be issued as soon as practicable after the application
has been docketed; or
* * * * *
(12) An amendment to an early site permit issued under subpart A of
part 52, or under subpart H of part 53 of this chapter, as follows:
(i) If the early site permit does not provide authority to conduct
the activities allowed under Sec. 50.10(e)(1) or Sec. 53.1130(b)(1)
of this chapter, the amendment will involve no significant hazards
consideration, and though the NRC will provide notice of opportunity
for a hearing under this section, it may make the amendment immediately
effective and grant a hearing thereafter; and
(ii) If the early site permit provides authority to conduct the
activities allowed under Sec. 50.10(e)(1) or Sec. 53.1130(b)(1) of
this chapter and the Commission determines under Sec. Sec. 50.58 and
50.91 or Sec. 53.1515 of this chapter 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 under Sec. 2.106 (if a hearing is
requested, which will be held after issuance of the amendment).
(13) A manufacturing license under subpart F of part 52 or subpart
H of part 53 of this chapter.
(b) * * *
(3) For a notice of intended operation under Sec. 52.103(a) or
Sec. 53.1452(a) of this chapter, the following information:
(i) The identification of the NRC action as making the finding
required under Sec. 52.103(g) or Sec. 53.1452(g) of this chapter;
(ii) The manner in which the licensee notifications under Sec.
52.99(c) or Sec. 53.1449(c) of this chapter which are required to be
made available by Sec. 52.99(e)(2) or Sec. 53.1449(e)(2) of this
chapter may be obtained and examined;
* * * * *
(iv) Any conditions, limitations, or restrictions to be placed on
the license in connection with the finding under Sec. 52.103(g) or
Sec. 53.1452(g) of this chapter, and the expiration date or
circumstances (if any) under which the conditions, limitations or
restrictions will no longer apply.
* * * * *
0
10. In Sec. 2.106, revise paragraphs (a)(2) and (3) and (b)(2)
introductory text to read as follows:
Sec. 2.106 Notice of issuance.
(a) * * *
[[Page 15767]]
(2) An amendment of a license for a facility of the type described
in Sec. 50.21(b) or Sec. 50.22 or part 53 of this chapter, as
applicable, or a testing facility, whether or not a notice of proposed
action has been previously published; and
(3) The finding under Sec. 52.103(g) or Sec. 53.1452(g) of this
chapter.
(b) * * *
(2) In the case of a finding under Sec. 52.103(g) or Sec.
53.1452(g) of this chapter:
* * * * *
0
11. In Sec. 2.109, revise paragraphs (b), (c), and (d) to read as
follows:
Sec. 2.109 Effect of timely renewal application.
* * * * *
(b) If the licensee of a nuclear power plant licensed under Sec.
50.21(b) or Sec. 50.22 or under subpart H of part 53 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.
(c) If the holder of an early site permit licensed under subpart A
of part 52 or under subpart H of part 53 of this chapter, as
applicable, files a sufficient application for renewal under Sec.
52.29 or Sec. 53.1173 of this chapter, as applicable, at least 12
months before the expiration of the existing early site permit, the
existing permit 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 H of part 53 of this chapter files a
sufficient application for renewal under Sec. 52.177 or Sec. 53.1295
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
12. 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 H of part 53 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 H of part 53 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
13. In Sec. 2.202, revise paragraph (e) to read as follows:
Sec. 2.202 Orders.
* * * * *
(e)(1) If the order involves the modification of a part 50 or a
part 53 license and is a backfit, the requirements of Sec. 50.109 or
Sec. 53.1590 of this chapter, as applicable, shall be followed, unless
the licensee has consented to the action required.
(2) If the order involves the modification of combined license
under subpart C of part 52 or subpart H of part 53 of this chapter, the
requirements of Sec. 52.98 or Sec. 53.1443 of this chapter, as
applicable, shall be followed unless the licensee has consented to the
action required.
(3) If the order involves a change to an early site permit under
subpart A of part 52 or under subpart H of part 53 of this chapter, the
requirements of Sec. 52.39 or Sec. 53.1188 of this chapter, as
applicable, must be followed, unless the applicant or licensee has
consented to the action required.
(4) If the order involves a change to a standard design
certification rule referenced by that plant's application, the
requirements, if any, in the referenced design certification rule with
respect to changes must be followed, or, in the absence of these
requirements, the requirements of Sec. 52.63 or Sec. 53.1263 of this
chapter, as applicable, must be followed, unless the applicant or
licensee has consented to follow 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. 53.1221 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 H of part 53 of this
chapter, the requirements of Sec. 52.171 or Sec. 53.1288 of this
chapter, as applicable, must be followed, unless the applicant or
licensee has consented to the action required.
0
14. In Sec. 2.309, revise paragraphs (a), (f)(1)(i), (vi), and (vii),
(g), (h)(2), (i)(2), and (j) to read as follows:
Sec. 2.309 Hearing requests, petitions to intervene, requirements for
standing, and contentions.
(a) General requirements. Any person whose interest may be affected
by a proceeding and who desires to participate as a party must file a
written request for hearing and a specification of the contentions
which the person seeks to have litigated in the hearing. In a
proceeding under Sec. 52.103 or Sec. 53.1452 of this chapter, as
applicable, the Commission, acting as the presiding officer, will grant
the request if it determines that the requestor has standing under the
provisions of paragraph (d) of this section and has proposed at least
one admissible contention that meets the requirements of paragraph (f)
of this section. For all other proceedings, except as provided in
paragraph (e) of this section, the Commission, presiding officer, or
the Atomic Safety and Licensing Board designated to rule on the request
for hearing and/or petition for leave to intervene, will grant the
request/petition if it determines that the requestor/petitioner has
standing under the provisions of paragraph (d) of this section and has
proposed at least one admissible contention that meets the requirements
of paragraph (f) of this section. In ruling on the request for hearing/
petition to intervene submitted by petitioners seeking to intervene in
the proceeding on the HLW repository, the Commission, the presiding
officer, or the Atomic Safety and Licensing Board shall also consider
any failure of the petitioner to participate as a potential party in
the pre-license application phase under subpart J of this part in
addition to the factors in paragraph (d) of this section. If a request
for hearing or petition to intervene is filed in response to any notice
of hearing or opportunity for hearing, the applicant/licensee shall be
deemed to be a party.
* * * * *
(f) * * *
(1) * * *
(i) Provide a specific statement of the issue of law or fact to be
raised or controverted, provided further, that the issue of law or fact
to be raised in a request for hearing under Sec. 52.103(b) or Sec.
53.1452(b) of this chapter, as applicable, must be directed at
demonstrating that one or more of the acceptance criteria in the
combined license have not been, or will not be
[[Page 15768]]
met, and that the specific operational consequences of nonconformance
would be contrary to providing reasonable assurance of adequate
protection of the public health and safety;
* * * * *
(vi) In a proceeding other than one under Sec. 52.103 or Sec.
53.1452 of this chapter provide sufficient information to show that a
genuine dispute exists with the applicant/licensee on a material issue
of law or fact. This information must include references to specific
portions of the application (including the applicant's environmental
report and safety report) that the petitioner disputes and the
supporting reasons for each dispute, or, if the petitioner believes
that the application fails to contain information on a relevant matter
as required by law, the identification of each failure and the
supporting reasons for the petitioner's belief; and
(vii) In a proceeding under Sec. 52.103(b) or Sec. 53.1452(b) of
this chapter, as applicable, the information must be sufficient, and
include supporting information showing, prima facie, that one or more
of the acceptance criteria in the combined license have not been, or
will not be met, and that the specific operational consequences of
nonconformance would be contrary to providing reasonable assurance of
adequate protection of the public health and safety. This information
must include the specific portion of the report required by Sec.
52.99(c) or Sec. 53.1449(c) of this chapter, as applicable, which the
requestor believes is inaccurate, incorrect, and/or incomplete (i.e.,
fails to contain the necessary information required by Sec. 52.99(c)
or Sec. 53.1449(c) of this chapter, as applicable). If the requestor
identifies a specific portion of the report under Sec. 52.99(c) or
Sec. 53.1449(c) of this chapter, as applicable, as incomplete and the
requestor contends that the incomplete portion prevents the requestor
from making the necessary prima facie showing, then the requestor must
explain why this deficiency prevents the requestor from making the
prima facie showing.
* * * * *
(g) Selection of hearing procedures. A request for hearing and/or
petition for leave to intervene may, except in a proceeding under Sec.
52.103 or Sec. 53.1452 of this chapter, as applicable, also address
the selection of hearing procedures, taking into account the provisions
of Sec. 2.310. If a request/petition relies upon Sec. 2.310(d), the
request/petition must demonstrate, by reference to the contention and
the bases provided and the specific procedures in subpart G of this
part, that resolution of the contention necessitates resolution of
material issues of fact which may be best determined through the use of
the identified procedures.
(h) * * *
(2) If the proceeding pertains to a production or utilization
facility (as defined in Sec. 50.2 or Sec. 53.020 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.
* * * * *
(i) * * *
(2) Except in a proceeding under Sec. 52.103 or Sec. 53.1452 of
this chapter, as applicable, the participant who filed the hearing
request, intervention petition, or motion for leave to file new or
amended contentions after the deadline may file a reply to any answer.
The reply must be filed within 7 days after service of that answer.
* * * * *
(j) Decision on request/petition. (1) In all proceedings other than
a proceeding under Sec. 52.103 or Sec. 53.1452 of this chapter, as
applicable, the presiding officer shall issue a decision on each
request for hearing or petition to intervene within 45 days of the
conclusion of the initial pre-hearing conference or, if no pre-hearing
conference is conducted, within 45 days after the filing of answers and
replies under paragraph (i) of this section. With respect to a request
to admit amended or new contentions, the presiding officer shall issue
a decision on each such request within 45 days of the conclusion of any
pre-hearing conference that may be conducted regarding the proposed
amended or new contentions or, if no pre-hearing conference is
conducted, within 45 days after the filing of answers and replies, if
any. In the event the presiding officer cannot issue a decision within
45 days, the presiding officer shall issue a notice advising the
Commission and the parties, and the notice shall include the expected
date of when the decision will issue.
(2) The Commission, acting as the presiding officer, shall
expeditiously grant or deny the request for hearing in a proceeding
under Sec. 52.103 or Sec. 53.1452 of this chapter, as applicable. The
Commission's decision may not be the subject of any appeal under Sec.
2.311.
0
15. Amend Sec. 2.310 by:
0
a. In paragraphs (a) and (h) introductory text, removing the cross-
reference ``parts 30, 32 through 36, 39, 40, 50, 52, 54, 55, 61, 70 and
72 of this chapter'' and adding, in its place, the cross reference
``parts 30, 32 through 36, 39, 40, 50, 52, 53, 54, 55, 61, 70, and 72
of this chapter''; and
0
b. Revising paragraphs (i) and (j).
The revisions read as follows:
Sec. 2.310 Selection of hearing procedures.
* * * * *
(i) In design certification rulemaking proceedings under part 52 or
part 53 of this chapter, any informal hearing held under Sec. 52.51 or
Sec. 53.1242 of this chapter, as applicable, must be conducted under
the procedures of subpart O of this part.
(j) Proceedings on a Commission finding under Sec. 52.103(c) and
(g) or Sec. 53.1452(c) and (g) of this chapter, as applicable, shall
be conducted in accordance with the procedures designated by the
Commission in each proceeding.
* * * * *
0
16. In Sec. 2.329, revise paragraph (a) to read as follows:
Sec. 2.329 Prehearing conference.
(a) Necessity for prehearing conference; timing. The Commission or
the presiding officer may, and in the case of a proceeding on an
application for a construction permit or an operating license for a
facility of a type described in Sec. 50.21(b) or Sec. 50.22 or part
53 of this chapter, or a testing facility, must direct the parties or
their counsel to appear at a specified time and place for a conference
or conferences before trial. A prehearing conference in a proceeding
involving a construction permit or operating license for a facility of
a type described in Sec. 50.21(b) or Sec. 50.22 or part 53 of this
chapter must be held within sixty (60) days after discovery has been
completed or any other time specified by the Commission or the
presiding officer.
* * * * *
0
17. In Sec. 2.339, revise paragraph (d) to read as follows:
Sec. 2.339 Expedited decision-making procedure.
* * * * *
(d) The provisions of this section do not apply to an initial
decision directing the issuance of a limited work authorization under
Sec. 50.10 or Sec. 53.1130
[[Page 15769]]
of this chapter; an early site permit under subpart A of part 52 or
under subpart H of part 53 of this chapter; a construction permit or
construction authorization under part 50 or 53 of this chapter; a
combined license under subpart C of part 52 or under subpart H of part
53 of this chapter; or a manufacturing license under subpart F of part
52 or under subpart H of part 53.
0
18. In Sec. 2.340, revise paragraphs (b), (c), (d), (f), (i), and (j)
to read as follows:
Sec. 2.340 Initial decision in certain contested proceedings;
immediate effectiveness of initial decisions; issuance of
authorizations, permits and licenses.
* * * * *
(b) Initial decision--combined license under part 52 or 53 of this
chapter--(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 combined license under
part 52 or 53 of this chapter (including an amendment to or renewal of
combined 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 shall also 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 combined license under part 52 or 53 of this chapter, or
the amendment of a combined license where the NRC has not made a
determination of no significant hazards consideration, 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 amendment of a combined
license under part 52 or 53 of this chapter where the NRC has made a
determination of no significant hazards consideration, the Commission
or the Director, Office of Nuclear Reactor Regulation, as appropriate
(appropriate official), after making the requisite findings and
complying with any applicable provisions of Sec. 2.1202(a) or Sec.
2.1403(a), may issue the amendment before the presiding officer's
initial decision becomes effective. Once the presiding officer's
initial decision becomes effective, the appropriate official shall take
action with respect to that amendment in accordance with the initial
decision. If the presiding officer's initial decision becomes effective
before the appropriate official issues the amendment, then the
appropriate official, after making the requisite findings, shall issue,
deny, or appropriately condition the amendment in accordance with the
presiding officer's initial decision.
(c) Initial decision on findings under Sec. 52.103 or Sec.
53.1452 of this chapter with respect to acceptance criteria in nuclear
power reactor combined licenses. In any initial decision under Sec.
52.103(g) or Sec. 53.1452(g) of this chapter with respect to whether
acceptance criteria have been or will be met, 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. Matters not put into
controversy by the parties but identified by the presiding officer as
matters requiring further examination, shall be referred to the
Commission for its determination; the Commission may, in its
discretion, treat any of these referred matters as a request for action
under Sec. 2.206 and process the matter in accordance with Sec.
52.103(f) or Sec. 53.1452(f) of this chapter.
(d) Initial decision--manufacturing license under part 52 or 53 of
this chapter--(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 C of part 52 or subpart H of part
53 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 C of part 52 or
subpart H of part 53 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 C of part 52 or subpart H of
part 53 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 or Sec.
53.1130 of this chapter; an early site permit under subpart A of part
52 or under subpart H of part 53 of this chapter; a construction permit
or construction authorization under part 50 or 53 of this chapter; an
operating license under part 50 or 53 of this chapter; a combined
license under subpart C of part 52 or subpart H or part 53 of this
chapter; a manufacturing license under subpart F of part 52 or subpart
H of part 53 of this chapter; a renewed license under part 53 or 54 of
this chapter; or a license under part 72 of this chapter to store spent
fuel in an independent spent fuel storage facility (ISFSI) or a
monitored retrievable storage installation (MRS); an initial decision
directing issuance of a license
[[Page 15770]]
under part 61 of this chapter; or an initial decision under Sec.
52.103(g) or Sec. 53.1452(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 or Sec. 53.1130 of this chapter; an
early site permit under subpart A of part 52 or subpart H of part 53 of
this chapter; a construction permit or construction authorization under
part 50 or 53 of this chapter; an operating license under part 50 or 53
of this chapter; a combined license under subpart C of part 52 or part
53 of this chapter; or a manufacturing license under subpart F of part
52 or part 53 of this chapter within 10 days from the date of issuance
of the initial decision:
(1) If the Commission or the Director has made all findings
necessary for issuance of the authorization, permit or license, not
within the scope of the initial decision of the presiding officer; and
(2) Notwithstanding the pendency of a petition for reconsideration
under Sec. 2.345, a petition for review under Sec. 2.341, or a motion
for stay under Sec. 2.342, or the filing of a petition under Sec.
2.206.
(j) Issuance of finding on acceptance criteria under Sec. 52.103
or Sec. 53.1452 of this chapter. The Commission or the Director,
Office of Nuclear Reactor Regulation, as appropriate, shall make the
finding under Sec. 52.103(g) or Sec. 53.1452(g) of this chapter, that
acceptance criteria in a combined license are met within 10 days from
the date of the presiding officer's initial decision:
(1) If the Commission or the Director is otherwise able to make the
finding under Sec. 52.103(g) or Sec. 53.1452(g) of this chapter, that
the prescribed acceptance criteria are met for those acceptance
criteria not within the scope of the initial decision of the presiding
officer;
(2) If the presiding officer's initial decision--with respect to
contentions that the prescribed acceptance criteria have not been met--
finds that those acceptance criteria have been met, and the Commission
or the Director thereafter is able to make the finding that those
acceptance criteria are met;
(3) If the presiding officer's initial decision--with respect to
contentions that the prescribed acceptance criteria will not be met--
finds that those acceptance criteria will be met, and the Commission or
the Director thereafter is able to make the finding that those
acceptance criteria are met; and
(4) Notwithstanding the pendency of a petition for reconsideration
under Sec. 2.345, a petition for review under Sec. 2.341, or a motion
for stay under Sec. 2.342, or the filing of a petition under Sec.
2.206.
* * * * *
Sec. 2.341 [Amended]
0
19. In Sec. 2.341, in paragraph (a)(1), remove the phrase ``Sec.
52.103(c)'' and add in its place the phrase ``Sec. 52.103(c) or Sec.
53.1452(c)''.
Sec. 2.400 [Amended]
0
20. In Sec. 2.400, remove the phrase ``parts 50 or 52'' and add in its
place the phrase ``part 50 or part 52 or Sec. 53.1470''.
0
21. In Sec. 2.401, revise the section heading and paragraph (a) to
read as follows:
Sec. 2.401 Notice of hearing on construction permit or combined
license applications pursuant to appendix N of 10 CFR part 50, 52, or
53.
(a) In the case of applications under appendix N of part 50 or
Sec. 53.1470 of this chapter for construction permits for nuclear
power reactors of the type described in Sec. 50.22 or part 53 of this
chapter, or applications under appendix N of part 52 or Sec. 53.1470
of this chapter for combined licenses, the Secretary will issue notices
of hearing pursuant to Sec. 2.104.
* * * * *
0
22. 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 or
Sec. 53.1470 of this chapter for construction permits for nuclear
power reactors of a type described in Sec. 50.22 or part 53 of this
chapter, or applications pursuant to appendix N of part 52 or Sec.
53.1470 of this chapter for combined 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.
* * * * *
Sec. 2.403 [Amended]
0
23. In Sec. 2.403, remove the phrase ``appendix N of part 50'' and add
in its place the phrase ``appendix N to part 50 or Sec. 53.1470''.
Sec. 2.404 [Amended]
0
24. In Sec. 2.404, remove the phrase ``appendix N of part 50'' and add
in its place the phrase ``appendix N to part 50 or Sec. 53.1470''.
Sec. 2.405 [Amended]
0
25. In Sec. 2.405, remove the phrase ``part 52'' and add in its place
the phrase ``part 52 or 53''.
Sec. 2.406 [Amended]
0
26. In Sec. 2.406, remove the phrase ``appendices N of parts 50 or
52'' and add in its place the phrase ``appendix N to part 50 or part 52
or Sec. 53.1470''.
Sec. 2.500 [Amended]
0
27. In Sec. 2.500, remove the phrase ``subpart F of part 52'' and add
in its place the phrase ``subpart F of part 52 or subpart H of part
53''.
0
28. In Sec. 2.501, revise the section heading and paragraph (a)
introductory text to read as follows:
Sec. 2.501 Notice of hearing on application under 10 CFR part 52 or
53 for a license to manufacture nuclear power reactors.
(a) In the case of an application under subpart F of part 52 or
subpart H of part 53 of this chapter for a license to manufacture
nuclear power reactors of the type described in Sec. 50.22 or part 53
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 (a) 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.
0
29. In Sec. 2.643, revise paragraph (b) to read as follows:
Sec. 2.643 Acceptance and docketing of application for limited work
authorization.
* * * * *
(b) The Director will accept for docketing part one of an
application for a construction permit for a utilization facility which
is subject to Sec. 51.20(b) of this chapter and is of the type
specified in Sec. 50.21(b)(2) or (3) or Sec. 50.22 or part 53 of this
chapter or an application for a combined license where part one of
[[Page 15771]]
the application as described in Sec. 2.101(a)(9) is complete. Part one
will not be considered complete unless it contains the information
required by Sec. 50.10(d)(3) or Sec. 53.1130(a)(3) of this chapter.
Upon assignment of a docket number, the procedures in Sec. 2.101(a)(3)
and (4) relating to formal docketing and the submission and
distribution of additional copies of the application must be followed.
* * * * *
Sec. 2.645 [Amended]
0
30. In Sec. 2.645, in paragraph (a), remove the phrase ``Sec.
50.33(a) through (f) of this chapter'' and add in its place the phrase
``Sec. Sec. 50.33(a) through (f), 53.1109, and 53.1306(a) or Sec.
53.1413 of this chapter, as applicable''.
Sec. 2.649 [Amended]
0
31. In Sec. 2.649, remove the phrase ``10 CFR 50.10(d)'' and add in
its place the phrase ``Sec. 50.10(d) or Sec. 53.1130(a) of this
chapter''.
Sec. 2.800 [Amended]
0
32. In Sec. 2.800:
0
a. In paragraph (c), remove the phrase ``subpart B of part 52'' and add
in its place the phrase ``subpart B of part 52 or subpart H of part
53''; and
0
b. In paragraph (d), remove the phrase ``subpart B of part 52'' and add
in its place the phrase ``subpart B of part 52 or subpart H of part 53
of this chapter''.
Sec. 2.801 [Amended]
0
33. In Sec. 2.801, remove the phrase ``subpart B of part 52'' and add
in its place the phrase ``subpart B of part 52 or subpart H of part
53''.
Sec. 2.813 [Amended]
0
34. In Sec. 2.813, in paragraph (a), remove the phrase ``parts 50, 52,
and 100'' and add in its place the phrase ``parts 50, 52, 53, and
100''.
Sec. 2.1103 [Amended]
0
35. In Sec. 2.1103, remove the phrase ``part 50 of this chapter'' and
add in its place the phrase ``part 50 or 53 of this chapter''.
Sec. 2.1200 [Amended]
0
36. In Sec. 2.1200, remove the phrase ``parts 30, 32 through 36, 39,
40, 50, 52, 54, 55, 61, 70, and 72 of this chapter,'' and add in its
place ``parts 30, 32 through 36, 39, 40, 50, 52, 53, 54, 55, 61, 70,
and 72 of this chapter,''.
0
37. In Sec. 2.1202, revise paragraphs (a)(1) through (3) and (6) to
read as follows:
Sec. 2.1202 Authority and role of NRC staff.
(a) * * *
(1) An application to construct and/or operate a production or
utilization facility (including an application for a limited work
authorization under 10 CFR 50.10 or 53.1130, or an application for a
combined license under subpart C of 10 CFR part 52, or under subpart H
of 10 CFR part 53;
(2) An application for an early site permit under subpart A of 10
CFR part 52 or under subpart H of 10 CFR part 53;
(3) An application for a manufacturing license under subpart F of
10 CFR part 52 or under subpart H of 10 CFR part 53;
* * * * *
(6) Production or utilization facility licensing actions that
involve significant hazards considerations as defined in 10 CFR 50.92
or 53.1520.
* * * * *
Sec. 2.1301 [Amended]
0
38. In Sec. 2.1301, in paragraph (b), remove ``part 50 and part 52''
and add in its place ``parts 50, 52, and 53''.
Sec. 2.1403 [Amended]
0
39. In Sec. 2.1403, in paragraph (a)(3), remove the phrase ``10 CFR
50.92'' and add in its place the phrase ``10 CFR 50.92 or 53.1520''.
Sec. 2.1500 [Amended]
0
40. In Sec. 2.1500, remove the phrase ``subpart B of part 52'' and add
in its place the phrase ``subpart B of part 52 or under subpart H of
part 53''.
Sec. 2.1502 [Amended]
0
41. In Sec. 2.1502:
0
a. In paragraph (a), remove the phrase ``Sec. 52.51(b)'' and add in
its place the phrase ``Sec. 52.51(b) or Sec. 53.1242(b)(2)''; and
0
b. In paragraph (b)(1), wherever it may appear, remove the phrase
``Sec. 52.51(a)'' and add in its place the phrase ``Sec. 52.51(a) or
Sec. 53.1242(b)''.
PART 10--CRITERIA AND PROCEDURES FOR DETERMINING ELIGIBILITY FOR
ACCESS TO RESTRICTED DATA OR NATIONAL SECURITY INFORMATION OR AN
EMPLOYMENT CLEARANCE
0
42. 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
43. In Sec. 10.1, in paragraph (a)(3), remove the phrase ``under part
52'' and add in its place the phrase ``under part 52 or 53''.
Sec. 10.2 [Amended]
0
44. In Sec. 10.2, in paragraph (b), wherever it may appear, remove the
phrase ``under part 52'' and add in its place the phrase ``under part
52 or 53''.
PART 11--CRITERIA AND PROCEDURES FOR DETERMINING ELIGIBILITY FOR
ACCESS TO OR CONTROL OVER SPECIAL NUCLEAR MATERIAL
0
45. 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
46. In Sec. 11.7:
0
a. Revise the introductory text; and
0
b. Remove the first undesignated paragraph.
The revision reads as follows:
Sec. 11.7 Definitions.
Terms defined in parts 10, 25, 50, 53, 70, 72, 73, and 95 of this
chapter have the same meaning when used in this part. Also, as used in
this part:
* * * * *
PART 19--NOTICES, INSTRUCTIONS AND REPORTS TO WORKERS: INSPECTION
AND INVESTIGATIONS
0
47. 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
48. In Sec. 19.2, revise paragraph (a) to read as follows:
Sec. 19.2 Scope.
(a) The regulations in this part apply to:
(1) All persons who receive, possess, use, or transfer material
licensed by the NRC under the regulations in parts 30 through 36 or
part 39, 40, 60, 61, 63, 70, or 72 of this chapter, including persons
licensed to operate a production or utilization facility under part 50,
52, or 53 of this chapter, persons licensed to possess power reactor
spent fuel in an
[[Page 15772]]
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 and early site permits) under parts 50, 52, 53,
and 54 of this chapter;
(3) All applicants for and holders of a standard design approval
under subpart E of part 52 or under subpart H of part 53 of this
chapter; and
(4) All applicants for a standard design certification under
subpart B of part 52 or under subpart H of part 53 of this chapter, and
those (former) applicants whose designs have been certified under that
subpart.
* * * * *
0
49. 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 parts 30
through 36 or part 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, 53, or 54 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 H of part
53 of this chapter or a standard design certification under subpart B
of part 52 or under subpart H of part 53 of this chapter.
* * * * *
Sec. 19.11 [Amended]
0
50. In Sec. 19.11, in paragraph (a) introductory text, paragraph (b)
introductory text, and paragraph (e)(1), wherever it may appear, remove
the phrase ``of part 52'' and add in its place the phrase ``of part 52
or under subpart H of part 53''.
Sec. 19.14 [Amended]
0
51. In Sec. 19.14, in paragraph (a), wherever it may appear, remove
the phrase ``of part 52'' and add in its place the phrase ``of part 52
or under subpart H of part 53''.
Sec. 19.20 [Amended]
0
52. In Sec. 19.20, remove ``parts 30, 40, 50, 52, 54, 60, 61, 63, 70,
72, 76, or 150'' and add in its place ``part 30, 40, 50, 52, 53, 54,
60, 61, 63, 70, 72, 76, or 150''.
PART 20--STANDARDS FOR PROTECTION AGAINST RADIATION
0
53. 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 [Amended]
0
54. In Sec. 20.1002, remove ``parts 30 through 36, 39, 40, 50, 52, 60,
61, 63, 70, or 72'' and add in its place ``under parts 30 through 36 or
part 39, 40, 50, 52, 53, 60, 61, 63, 70, or 72''.
0
55. 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 parts 30
through 36 or part 39, 40, 50, 53, 60, 61, 63, 70, or 72 of this
chapter.
* * * * *
Sec. 20.1101 [Amended]
0
56. In Sec. 20.1101, in paragraph (d):
0
a. Remove ``Sec. 20.1101 (b)'' and add in its place ``paragraph (b) of
this section'';
0
b. Remove ``of this part''; and
0
c. Remove the phrase ``subject to Sec. 50.34a'' and add in its place
the phrase ``subject to Sec. 50.34a or Sec. 53.260 of this chapter''.
Sec. 20.1401 [Amended]
0
57. In Sec. 20.1401:
0
a. In paragraph (a), remove ``parts 30, 40, 50, 52, 60, 61, 63, 70, and
72'' and add in its place ``under parts 30, 40, 50, 52, 53, 60, 61, 63,
70, and 72''; and
0
b. In paragraphs (a) and (c), remove the phrase ``in accordance with
Sec. 50.83'' and add in its place the phrase ``in accordance with
Sec. 50.83 or Sec. 53.1080''.
0
58. In Sec. 20.1403, revise paragraph (d) introductory text to read as
follows:
Sec. 20.1403 Criteria for license termination under restricted
conditions.
* * * * *
(d) The licensee has submitted a decommissioning plan or License
Termination Plan (LTP) to the Commission indicating the licensee's
intent to decommission in accordance with Sec. 30.36(d), Sec.
40.42(d), Sec. 50.82 (a) and (b), subpart G of part 53, Sec.
70.38(d), or Sec. 72.54 of this chapter, and specifying that the
licensee intends to decommission by restricting use of the site. The
licensee shall document in the LTP or decommissioning plan how the
advice of individuals and institutions in the community who may be
affected by the decommissioning has been sought and incorporated, as
appropriate, following analysis of that advice.
* * * * *
0
59. In Sec. 20.1404, revise paragraph (a)(4) introductory text to read
as follows:
Sec. 20.1404 Alternate criteria for license termination.
(a) * * *
(4) Has submitted a decommissioning plan or License Termination
Plan (LTP) to the Commission indicating the licensee's intent to
decommission in accordance with Sec. 30.36(d), Sec. 40.42(d), Sec.
50.82 (a) and (b), subpart G of part 53, Sec. 70.38(d), or Sec. 72.54
of this chapter, and specifying that the licensee proposes to
decommission by use of alternate criteria. The licensee shall document
in the decommissioning plan or LTP how the advice of individuals and
institutions in the community who may be affected by the
decommissioning has been sought and addressed, as appropriate,
following analysis of that advice. In seeking such advice, the licensee
shall provide for:
* * * * *
Sec. 20.1406 [Amended]
0
60. In Sec. 20.1406, in paragraphs (a) and (b), remove the phrase
``under part 52'' and add in its place the phrase ``under part 52 or
53''.
0
61. In Sec. 20.1501, revise paragraph (b) to read as follows:
Sec. 20.1501 General.
* * * * *
(b) Notwithstanding Sec. 20.2103(a), records from surveys
describing the location and amount of subsurface residual radioactivity
identified at the site must be kept with records important for
decommissioning, and such records must be retained in accordance with
Sec. 30.35(g), Sec. 40.36(f), Sec. 50.75(g), subpart G of part 53,
Sec. 70.25(g), or Sec. 72.30(d) of this chapter, as applicable.
* * * * *
Sec. 20.1905 [Amended]
0
62. In Sec. 20.1905, in paragraph (g) introductory text, remove the
phrase ``Parts 50 or 52'' and add in its place the phrase ``part 50,
52, or 53''.
[[Page 15773]]
0
63. 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 53 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 or Sec. 53.425(c) 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 or Sec. 50.34a
or under subpart H of part 53 of this chapter associated with this
incineration pursuant to Sec. 50.71 or Sec. 53.1620 of this chapter,
as appropriate. The licensee shall also follow the procedures of Sec.
50.59 or Sec. 53.1565 of this chapter with respect to such changes to
the facility or procedures.
* * * * *
0
64. In Sec. 20.2201, revise paragraphs (a)(2)(i), (b)(2)(i), and (c)
to read as follows:
Sec. 20.2201 Reports of theft or loss of licensed material.
(a) * * *
(2) * * *
(i) Licensees having an installed Emergency Notification System
shall make the reports to the NRC Operations Center under Sec. 50.72
or Sec. 53.1630 of this chapter, and
* * * * *
(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) through (e) and (g) or
Sec. 53.1640(b) through (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, Sec. 53.1630, Sec. 53.1640, Sec. 70.52, Sec. 73.27(b),
Sec. 73.67(e)(3)(vii) or (g)(3)(iii), Sec. 73.1205, or Sec.
150.19(c) of this chapter.
* * * * *
Sec. 20.2202 [Amended]
0
65. In Sec. 20.2202, in paragraph (d)(1), remove the phrase ``10 CFR
50.72'' and add in its place the phrase ``Sec. 50.72 or Sec. 53.1630
of this chapter;''.
0
66. 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) through (e) and (g) or Sec. 53.1640(b) through (e) of this
chapter, and must include the information required by paragraph (b) of
this section. Occurrences reported under Sec. 50.73 or Sec. 53.1640
of this chapter need not be reported by a duplicate report under
paragraph (a) of this section.
* * * * *
Sec. 20.2206 [Amended]
0
67. In Sec. 20.2206, in paragraph (a)(1), remove the phrase ``or Sec.
50.22'' and add in its place the phrase ``or Sec. 50.22 or part 53''.
PART 21--REPORTING OF DEFECTS AND NONCOMPLIANCE
0
68. 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
69. In Sec. 21.2, revise paragraphs (a)(2) through (4), (b), and (c)
to read as follows:
Sec. 21.2 Scope.
(a) * * *
(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, 52, or 53 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 63 of this
chapter; or supplies basic components for a facility or activity
licensed, other than for export, under part 30, 40, 50, 52, 53, 60, 61,
63, 70, 71, or 72 of this chapter;
(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 a design
certification rule under part 52 or 53 of this chapter; or supplying
basic components with respect to that design certification, and each
individual, corporation, partnership, or other entity doing business
within the United States, and each director and responsible officer of
such an organization, whose application for design certification has
been granted under part 52 or 53 of this chapter, or who has supplied
or is supplying basic components with respect to that design
certification;
(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 53 of this chapter; or
supplying basic components with respect to a standard design approval
under part 52 or 53 of this chapter;
(b) For persons licensed to construct a facility under either a
construction permit issued under Sec. 50.23 or Sec. 53.1333 of this
chapter or a combined license under part 52 or 53 of this chapter (for
the period of construction until the date that the Commission makes the
finding under Sec. 52.103(g) or Sec. 53.1452(g) of this chapter), or
to manufacture a facility under part 52 or 53 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. 53.605 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, 52, or 53 of this chapter, evaluation of potential defects and
appropriate reporting of defects under Sec. 50.72, Sec. 50.73, Sec.
53.1630, Sec. 53.1640, 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
[[Page 15774]]
and responsible officers of these licensees to report defects under
Section 206 of the Energy Reorganization Act of 1974.
* * * * *
0
70. In Sec. 21.3, revise the definitions for ``Basic component'',
``Commercial grade item'', ``Critical characteristics'', ``Dedicating
entity'', ``Dedication'', ``Defect'', and ``Substantial safety hazard''
to read as follows:
Sec. 21.3 Definitions.
* * * * *
Basic component. (1)(i) When applied to nuclear power plants
licensed under part 53 of this chapter, basic component means a safety-
related structure, system, or component (SSC), or part thereof, and
when applied to nuclear power plants licensed under part 50 or 52 of
this chapter, basic component means an SSC, or part thereof, that
affects its safety function necessary to assure:
(A) The integrity of the reactor coolant pressure boundary;
(B) The capability to shut down the reactor and maintain it in a
safe-shutdown condition; or
(C) The capability to prevent or mitigate the consequences of
accidents which could result in potential offsite exposures comparable
to those referred to in Sec. 50.34(a)(1), Sec. 50.67(b)(2), or Sec.
100.11 of this chapter, as applicable.
(ii) Basic components are items designed and manufactured under a
quality assurance program complying with appendix B to part 50 of this
chapter, or commercial grade items which have successfully completed
the dedication process.
(2) When applied to standard design certifications and approvals
under part 53 of this chapter, basic component means the design or
procurement information approved or to be approved within the scope of
the design certification or approval for a safety-related SSC, or part
thereof. When applied to standard design certifications under subpart B
of part 52 of this chapter and standard design approvals under part 52
of this chapter, basic component means the design or procurement
information approved or to be approved within the scope of the design
certification or approval for an SSC, or part thereof, that affects its
safety function necessary to assure:
(i) The integrity of the reactor coolant pressure boundary;
(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 which could result in potential offsite exposures comparable
to those referred to in Sec. 50.34(a)(1), Sec. 50.67(b)(2), or Sec.
100.11 of this chapter, as applicable.
(3) When applied to other facilities and other activities licensed
under part 30, 40, 50 (other than nuclear power plants), 60, 61, 63,
70, 71, or 72 of this chapter, basic component means a structure,
system, or component, or part thereof, that affects their safety
function, that is directly procured by the licensee of a facility or
activity subject to the regulations in this part and in which a defect
or failure to comply with any applicable regulation in this chapter,
order, or license issued by the Commission could create a substantial
safety hazard.
(4) In all cases, basic component includes safety-related design,
analysis, inspection, testing, fabrication, replacement of parts, or
consulting services that are associated with the component hardware,
design certification, design approval, or information in support of an
early site permit application under part 52 or 53 of this chapter,
whether these services are performed by the component supplier or
others.
Commercial grade item. (1) When applied to nuclear power plants
licensed under part 50 or 53 of this chapter, commercial grade item
means an SSC, 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
part 30, 40, 50 (other than nuclear power plants), 60, 61, 63, 70, 71,
or 72 of this chapter, 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, 52, or 53 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, 52, or 53 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), 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.
Dedication. (1) When applied to nuclear power plants licensed
pursuant to part 30, 40, 50, 53, or 60 of this chapter, dedication is
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 quality assurance program
under appendix B to part 50 of this chapter. 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 holdpoints at the manufacturer's
facility, and analysis of historical records for acceptable
performance. In all cases, the dedication process must be conducted
under the applicable provisions of appendix B to part 50. The process
is considered complete when the item is designated for use as a basic
component.
(2) When applied to facilities and activities licensed pursuant to
part 30, 40, 50 (other than nuclear power plants), 60, 61, 63, 70, 71,
or 72 of this chapter, dedication occurs after receipt when that 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 section;
(3) A deviation in a portion of a facility subject to the early
site permit, standard design certification, standard design approval,
construction permit,
[[Page 15775]]
combined license or manufacturing licensing requirements of part 50,
52, or 53 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, 52, or 53 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, 53, 60, 61, 63, 70, 71, or 72 of
this chapter.
* * * * *
Sec. 21.21 [Amended]
0
71. In Sec. 21.21:
0
a. In paragraphs (a)(3) introductory text and (a)(3)(i), remove the
phrase ``under part 52'' and add in its place the phrase ``under part
52 or 53''; and
0
b. In paragraphs (d)(1)(i) and (ii), remove ``parts 30, 40, 50, 52, 60,
61, 63, 70, 71, or 72'' and add ``part 30, 40, 50, 52, 53, 60, 61, 63,
70, 71, or 72'' in its place.
Sec. 21.51 [Amended]
0
72. In Sec. 21.51, in paragraphs (a)(4) and (5), remove the phrase
``of part 52'' and add in its place the phrase ``of part 52 or under
subpart H of part 53''.
Sec. 21.61 [Amended]
0
73. In Sec. 21.61, in paragraph (b), remove the phrase ``under part
52'' wherever it may appear and add in its place the phrase ``under
part 52 or 53''.
PART 25--ACCESS AUTHORIZATION
0
74. 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
75. In Sec. 25.5, revise the definition for ``License'' to read as
follows:
Sec. 25.5 Definitions.
* * * * *
License means a license issued pursuant to part 50, 52, 53, 60, 63,
70, or 72 of this chapter.
* * * * *
Sec. 25.17 [Amended]
0
76. In Sec. 25.17, in paragraph (a), remove ``10 CFR parts 50, 52, 54,
60, 63, 70, 72, or 76'' and add in its place ``part 50, 52, 53, 54, 60,
63, 70, 72, or 76 of this chapter''.
Sec. 25.35 [Amended]
0
77. In Sec. 25.35, in paragraph (a), wherever it may appear, remove
the phrase ``under part 52'' and add in its place the phrase ``under
part 52 or 53''.
PART 26--FITNESS FOR DUTY PROGRAMS
0
78. 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
79. In Sec. 26.3, revise paragraph (d) and add paragraph (f) to read
as follows:
Sec. 26.3 Scope.
* * * * *
(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) No later than the start of construction activities, licensees
and other entities that have applied for or have been issued a license
under part 53 of this chapter, other than a manufacturing license (ML),
must implement the requirements in subpart M of this part or all the
requirements of this part except subparts K and M. Holders of an ML
under part 53 of this chapter must implement the requirements in
subpart M or all the requirements of this part except subparts K and M,
before commencing activities that assemble a manufactured reactor.
0
80. In Sec. 26.4, revise paragraphs (a) introductory text, (a)(1) and
(4), (b), (c), (e) introductory, (e)(4), (f), (g) introductory text,
and (h) 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 M 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 licensed commercial 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
the requirements in subpart M of this part, unless the licensee or
other entity subjects these individuals to an FFD program that
satisfies all of the requirements of this part except for those
requirements in subparts K and M:
(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
[[Page 15776]]
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 M of this part. All
persons who are granted unescorted access to a facility licensed under
part 53 of this chapter, and who do not perform or direct the
performance of the duties described in paragraph (a) of this section,
must be subject to the requirements in subpart M of this part, unless
the licensee or other entity implements an FFD program that satisfies
all of the requirements of this part, except Sec. Sec. 26.205 through
26.209 and subparts K and M.
(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 M of this part. Also, 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
all of the requirements described in subpart M of this part, unless the
licensee or other entity implements an FFD program that satisfies all
of the requirements of this part, except Sec. Sec. 26.205 through
26.209 and subparts K and M.
* * * * *
(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 M 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 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 M
or an FFD program that satisfies all of the requirements of this part,
except subparts I, K, and M:
* * * * *
(4) Witnesses or determines inspections, tests, and analyses
certification required under part 52 or 53 of this chapter;
* * * * *
(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 M of this part, unless the licensee or other entity
subjects these individuals to an FFD program that meets all of the
requirements of this part, except for subparts I, K, and M of this
part.
(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 M 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 53 of this
chapter must be subject to the requirements in subpart M or an FFD
program that satisfies all of the requirements of this part, except
subparts I, K, and M, and, at the licensee's or other entity's
discretion, subpart C of this part:
* * * * *
(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, and 26.39, and the applicable requirements of subparts
C, E through H, and M of this part.
* * * * *
0
81. Amend Sec. 26.5 by:
0
a. Adding the definitions for ``Biological marker'', ``Change'', and
``Consortium/Third party administrator (C/TPA)'' in alphabetical order;
0
b. Revising the definition for ``Constructing or construction
activities'' and ``Contractor/vendor (C/V)'';
0
c. Adding the definition of ``Illicit substance'' in alphabetical
order;
0
d. Revising the definition of ``Other entity'';
0
e. Adding the definition of ``Reduction in FFD program effectiveness''
in alphabetical order;
0
f. Revising the definitions of ``Reviewing official'', ``Safety-related
structures, systems, and components (SSCs)'', and ``Security-related
SSCs'';
0
g. Adding the definition of ``Special nuclear material (SNM)'' in
alphabetical order; and
0
h. Revising the definition of ``Unit outage''.
The additions and revisions read as follows:
Sec. 26.5 Definitions.
* * * * *
Biological marker means, for a part 53 licensee implementing
subpart M 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. 26.603(e) means an action that results in a
modification of, addition to, or removal from the licensee's or other
entity's FFD program.
* * * * *
Consortium/Third-party administrator (C/TPA) 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. 26.607(b)(2)(vi), that otherwise could not be independently
implemented by those licensees or other entities. A C/TPA 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.
Constructing or construction activities means, 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. 53.020 of this chapter.
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.
* * * * *
[[Page 15777]]
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 53
licensee or other entity implementing subpart M 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 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 53 licensees and other entities
described in Sec. 26.3(d) and (f), safety-related has the same meaning
as that in Sec. 53.020 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 permit applicant or holder or an early site
permit holder, as described in Sec. 26.3(c)(3) through (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 (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
82. 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, 26.603, 26.605,
26.606, 26.607, 26.608, 26.609, 26.611, 26.613, 26.617, 26.619, 26.711,
26.713, 26.715, 26.717, 26.719, and 26.821.
0
83. 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 M and K of this part) 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
84. 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.
26.606(b)(2)(vii) are implemented, as applicable.
0
85. 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 M of this part for the categories of
individuals in Sec. 26.4 and those licensees and other entities that
elect to implement the requirements in Sec. 26.605.
Sec. 26.53 [Amended]
0
86. In Sec. 26.53:
0
a. In paragraph (e), wherever it may appear, remove the phrase ``Sec.
26.3(a) through (c)'' and add in its place the phrase ``Sec. 26.3(a)
through (c) and (f)'';
0
b. In paragraph (g) and paragraph (h) introductory text, remove the
phrase ``(c) and (d)'' and add in its place the phrase ``(c), (d), and
(f)''; and
0
c. In paragraph (i) introductory text, remove the phrase ``(c) and(d)''
and add in its place the phrase ``(c), (d), and (f)''.
Sec. 26.63 [Amended]
0
87. In Sec. 26.63, in paragraph (d), remove the phrase ``Sec. 26.3(a)
through
[[Page 15778]]
(d)'' and add in its place the phrase ``Sec. 26.3(a) through (d) and
(f)''.
0
88. 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 M of this part for the categories of
individuals in Sec. 26.4 and those licensees and other entities that
elect to implement the requirements in Sec. 26.605(b).
0
89. 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 section and under Sec. Sec. 26.4(j) and 26.31(b)(2)
and subpart K of this part. 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 M of
this part for the categories of individuals in Sec. 26.4 and those
licensees and other entities that elect to implement the requirements
in Sec. 26.605.
0
90. In Sec. 26.97, revise paragraph (a) introductory text and
paragraph (b) introductory text 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. 26.607(g)(2) if applicable, shall
perform the oral fluid specimen collection consistent with the device
manufacturer's instructions. The collector shall:
* * * * *
(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.
26.607(g)(2) if applicable, shall:
* * * * *
0
91. 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 M. 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. 26.605. 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 53; and the
requirements in Sec. Sec. 26.205 through 26.209 apply to the
individuals identified in Sec. 26.4(a).
0
92. Add Sec. 26.202 to read as follows:
Sec. 26.202 General provisions for facilities licensed under 10 CFR
part 53.
(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.
26.606(a).
(b) Procedures. In addition to the procedures required in Sec.
26.606(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
he or she 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--
(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
KAs in the content of the training and trainee assessments required in
Sec. 26.608:
(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
[[Page 15779]]
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. 26.617(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), (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. 26.615.
0
93. 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. Sec. 26.27 and 26.203(a) and (b) or Sec. Sec.
26.202(a) and (b) and 26.606, the work hour counting system in
paragraph (d)(7)(ii) of this section the licensee is using.
(8) Each licensee shall state, in its FFD policy and procedures
required by either Sec. Sec. 26.27 and 26.203(a) and (b) or Sec. Sec.
26.202(a) and (b) and 26.606, the requirements with which the licensee
is complying: the minimum days off requirements in paragraph (d)(3) of
this section or maximum average work hours requirements in paragraph
(d)(7) of this section.
* * * * *
0
94. 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. Sec. 26.29 and 26.203(c) or Sec. Sec. 26.202(c) and
26.608 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. Sec. 26.29 and 26.203(c)
or Sec. Sec. 26.202(c) and 26.608.
* * * * *
0
95. In Sec. 26.211, revise paragraphs (a)(1) and (3) and paragraph (b)
introductory text 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,
26.607(b), and 26.619, a fatigue 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. 26.607(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
26.608 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 26.608.
* * * * *
0
96. Add subpart M, consisting of Sec. Sec. 26.601 through 26.619, to
read as follows:
[[Page 15780]]
Subpart M--Fitness-for-Duty Programs for Facilities Licensed Under
10 CFR Part 53
Sec.
26.601 Applicability.
26.603 General provisions.
26.605 FFD program requirements.
26.606 Written policy and procedures.
26.607 Drug and alcohol testing.
26.608 FFD program training.
26.609 Behavioral observation.
26.610 Sanctions.
26.611 Protection of information.
26.613 Appeals process.
26.615 Audits.
26.617 Recordkeeping, reporting, and FFD program performance.
26.619 Suitability and fitness determinations.
Sec. 26.601 Applicability.
(a) A licensee or other entity in Sec. 26.3(f), at its discretion,
may 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 part 53 of this chapter. If a licensee or other entity in
Sec. 26.3(f) does not elect to implement an FFD program that satisfies
the requirements of this subpart, then those categories of individuals
in Sec. 26.4, as applicable, and any person licensed to operate under
part 53 of this chapter must be subject to an FFD program that
satisfies all requirements under this part, except for those
requirements in subpart K and this subpart.
(b) A licensee or other entity in Sec. 26.3(f) that elects to
implement an FFD program that satisfies the requirements of this
subpart, and that demonstrates compliance with Sec. 73.100(a)(1)(i) of
this chapter, must establish, implement, and maintain an FFD program
under Sec. 26.605(a) or (b).
(c) A licensee or other entity in Sec. 26.3(f) that elects to
implement an FFD program that satisfies the requirements of this
subpart and does not demonstrate compliance with Sec. 73.100(a)(1)(i)
of this chapter must establish, implement, and maintain an FFD program
under Sec. 26.605(a) and (b).
Sec. 26.603 General provisions.
(a) FFD program description. An applicant's description of the FFD
program in its Final Safety Analysis Report, required by subpart H of
part 53 of this chapter, must include--
(1) Whether the applicant demonstrates compliance with Sec.
73.100(a)(1)(i) of this chapter;
(2) A statement whether the FFD program will be implemented
pursuant to Sec. 26.605(a) or (b) or Sec. 26.605(a) and (b), or will
satisfy all requirements under this part, except for the requirements
in subpart K and this subpart;
(3) A discussion of the applicability of the FFD program to those
individuals described in Sec. 26.4 and how the program will be
implemented offsite at a U.S. Nuclear Regulatory Commission (NRC)-
licensed facility authorized to assemble or test a manufactured
reactor, if applicable; and
(4) 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 in Sec. 26.3(f), other than the holder of a
manufacturing license (ML), the FFD program must be maintained until
the NRC's docketing of the license holder's certifications described in
Sec. 53.1070 of this chapter. For holders of an ML, the FFD program
must be maintained until expiration of the ML.
(c)-(d) [Reserved]
(e) 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 this
part, laboratory processes or procedures, or guidance issued by the
U.S. Department of Health and Human Services (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 not result in a reduction in
FFD program effectiveness.
(3) Except for phencyclidine, and notwithstanding paragraph
(e)(1)(ii) of this section, the change control process may not be used
to reduce the minimum panel of drugs to be tested in Sec.
26.607(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. 26.617(b)(2) or Sec. 26.717, as applicable.
Sec. 26.605 FFD program requirements.
(a)(1) The licensee or other entity must establish, implement, and
maintain an FFD program under this paragraph (a) before the start of--
(i) Construction activities as defined in Sec. 26.5;
(ii) Activities performed under an ML that authorizes the assembly,
non-operational testing, or both of a manufactured reactor; or
(iii) Activities described in paragraphs (a)(1)(i) and (ii) of this
section.
(2) An FFD program under this paragraph (a) must--
(i) Apply to those individuals described in Sec. 26.4, as
applicable; and,
(ii) Implement the following requirements and subparts:
(A) Section 26.23, for performance objectives;
(B) Section 26.35, for employee assistance programs:
(1) For licensees and other entities who demonstrate compliance
with Sec. 73.100(a)(1)(i) of this chapter, Sec. 26.35 must be
implemented before the earliest occurrence of the following, as
applicable: the loading of fuel onsite into a reactor vessel; receiving
a fueled manufactured reactor; or individuals subject to this part
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
(2) For licensees and other entities that do not demonstrate
compliance with Sec. 73.100(a)(1)(i) of this chapter, Sec. 26.35 must
be implemented before the start of activities performed under an ML
that allows the assembly, non-operational testing, or both, of a
manufactured reactor;
(C) Section 26.603, for general provisions;
(D) Section 26.606, for written policy and procedures;
(E) Section 26.607, for drug and alcohol testing;
(F) Section 26.608, for FFD program training;
(G) Section 26.609, for behavioral observation;
(H) Section 26.610, for sanctions;
(I) Section 26.611, for protection of information;
(J) Section 26.613, for the appeals process;
(K) Section 26.615, for audits;
(L) Section 26.617, for recordkeeping, reporting, and FFD program
performance;
(M) Section 26.619, for suitability and fitness determinations;
(N) Subpart A, for administrative provisions;
[[Page 15781]]
(O) Subpart I, for managing fatigue, in the case of holders of an
ML that allows the assembly, non-operational testing, or both of a
manufactured reactor; and
(P) Subpart O, for inspections, violations, and penalties.
(b)(1) Except for a licensee or other entity that demonstrates
compliance with Sec. 73.100(a)(1)(i) of this chapter and elects to
implement an FFD program that satisfies the requirements of this
paragraph (b), for which its FFD program must be established,
implemented, and maintained no later than the start of construction
activities, as defined in Sec. 26.5, and except for holders of
manufacturing licenses if they possess a separate license to load fuel
into a manufactured reactor, for which its FFD program under this
paragraph (b) must be established, implemented, and maintained no later
than the start of the loading of fuel into the reactor, the licensee or
other entity must establish, implement, and maintain an FFD program
under this paragraph (b) before the earliest occurrence of the
following, as applicable:
(i) The loading of fuel onsite into a reactor vessel;
(ii) Receiving a fueled manufactured reactor; and
(iii) Individuals subject to this part 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.
(2) An FFD program under this paragraph (b) must--
(i) Apply to those individuals described in Sec. 26.4, as
applicable; and
(ii) Implements the following requirements and subparts:
(A) Section 26.23, for performance objectives;
(B) Section 26.35, for employee assistance programs;
(C) Section 26.603, for general provisions;
(D) Section 26.606, for written policy and procedures;
(E) Section 26.607, for drug and alcohol testing;
(F) Section 26.608, for FFD program training;
(G) Section 26.609, for behavioral observation;
(H) Section 26.611, for protection of information;
(I) Section 26.613, for the appeals process;
(J) Section 26.615, for audits;
(K) Subpart A, for administrative provisions;
(L) Subpart C, for granting and maintaining authorization;
(M) Subpart D, for management actions and sanctions to be imposed;
(N) Subpart H, for determining fitness-for-duty Policy violations
and determining fitness, unless using the Mandatory Guidelines for
Federal Workplace Drug Testing Programs (the HHS Guidelines) for
Medical Review Officer (MRO) evaluation of drug test results, and
determining fitness;
(O) Subpart I, for managing fatigue;
(P) Subpart N, for recordkeeping and reporting requirements; and
(Q) Subpart O, for inspections, violations, and penalties.
Sec. 26.606 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 paragraph (b) of this
section, sanctions required under this part, 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 15782]]
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. 26.202(b) and
either Sec. 26.205(d)(3) or (7).
(6) Measures to prevent subversion of drug and alcohol tests
conducted onsite and offsite.
Sec. 26.607 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.
(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 (j) 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 this part 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. 26.603(e)(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, for collecting a urine specimen under direct
observation;
(ii) Section 26.119, for determining ``shy'' bladder; and
(iii) Section 26.163, for cutoff levels for drugs and drug
metabolites.
(3) For alcohol testing onsite, the following requirements must be
implemented--
(i) Section 26.91, for acceptable devices for conducting initial
and confirmatory tests for alcohol and methods of use;
(ii) Section 26.93, for preparing for alcohol testing;
(iii) Section 26.95, for conducting an initial test for alcohol
using a breath specimen;
(iv) Section 26.97, for collecting oral fluid specimens for alcohol
and drug testing;
(v) Section 26.99, for determining the need for a confirmatory test
for alcohol;
(vi) Section 26.101, for conducting a confirmatory test for
alcohol; and,
(vii) Section 26.103, for determining a confirmed positive test
result for alcohol.
(4) For all test conditions in paragraph (b) of this section and
for MRO-directed tests under Sec. 26.185, drug testing must be
performed at an HHS-certified laboratory for the specific biological
specimen to be tested. Only HHS-certified laboratory test results from
urine and oral fluid specimens may be used for the issuance of a
sanction
[[Page 15783]]
required under this part. The licensee or other entity must establish
and maintain a contract with a primary and a back-up HHS-certified
laboratory (with a different Certifying Scientist) for the specimen(s)
to be tested. These contracts must stipulate that the laboratories are
subject to inspection or audit by the licensee or other entity and that
records and documents must be provided and/or able to be photocopied
and removed from the premises to support the inspection or audit.
(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 paragraphs (b)(3) and (4) of this section, 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. 26.606(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 C/TPA to
implement random testing under paragraph (b)(2)(vi) of this section.
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 Sec. 26.85.
(ii) The oral fluid specimen collection process must be completed
as described under Sec. Sec. 26.97 and 26.99.
(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 custody and control form (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 evidential breath testing device (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) [Reserved]
(i) 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
paragraph (b)(1) of this section to 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. 26.613. 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.
[[Page 15784]]
(j) 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,
the individual screened by the instrument must be subject to for-cause
testing under paragraph (b)(3) of this section.
(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 (j)(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 sanction under this part 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.
(k) 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.
(l) 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.
(m) 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. 26.619, 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 (m)(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
paragraphs (c)(3)(vi) and (vii) of this section.
(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 re-test. To resolve a
discrepant condition, the MRO is authorized to test a specimen for a
biological marker, adulterants, or additional drugs.
(n) 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.
(o) Specimen collectors. All onsite specimen collections, except a
collection by a portal area screening instrument in paragraph (j) of
this section, must be conducted by licensee- or other entity-designated
and -trained personnel.
Sec. 26.608 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 commercial nuclear plants, the FFD
program training program must use a systems approach to training as
defined in Sec. 53.725 of this chapter and described in Sec. 53.830
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.
[[Page 15785]]
(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 and access authorization programs under Sec. 73.55, Sec.
73.56, Sec. 73.100, or Sec. 73.120 of this chapter; and
(iv) Be used to inform other NRC licensees and other entities
subject to this part when FFD program information is requested to
support authorization determinations under subpart C of this part or
Sec. 73.56 or Sec. 73.120 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. 26.609 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 this part.
(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 paragraph (c) of this section, 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 part 53 of this chapter who
manipulate the controls of any commercial nuclear plant licensed under
part 53.
Sec. 26.610 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 53 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. 26.611 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) through (6)
and (8), and persons deciding matters under review in Sec. 26.613.
This signed and dated consent must be obtained before making the
individual subject to the FFD program.
Sec. 26.613 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. 26.615 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.
26.617(d) and unsatisfactory performance or programmatic weaknesses
identified under Sec. 26.617(b)(3) and (e).
[[Page 15786]]
(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. 26.617 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 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. 26.613, 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. 26.611) and FFD program weakness to
NRC, licensees, and other entities subject to this part when requested
to support authorization determinations under subpart C of this part or
Sec. 73.120 of this chapter, as applicable, 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. 26.619 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. 26.607(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. 26.608.
0
97. 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 M or elect to implement the
requirements in Sec. 26.605(b).
Sec. 26.711 [Amended]
0
98. In Sec. 26.711, in paragraphs (c) and (d), remove the phrase ``(c)
and (d),'' and add in its place the phrase ``(c), (d), and (f),''.
0
99. In Sec. 26.825, revise paragraph (b) to read as follows:
Sec. 26.825 Criminal penalties.
* * * * *
(b) The regulations in this part that are not issued under sections
161b, 161i, or 161o for the purposes of section 223 are as follows:
Sec. Sec. 26.1, 26.3, 26.5, 26.7, 26.8, 26.9, 26.11, 26.51, 26.81,
26.121, 26.151, 26.181, 26.201, 26.601, 26.823, and 26.825.
PART 30--RULES OF GENERAL APPLICABILITY TO DOMESTIC LICENSING OF
BYPRODUCT MATERIAL
0
100. 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
101. 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 53 of this chapter.
0
102. In Sec. 30.50, revise paragraph (c)(3) to read as follows:
Sec. 30.50 Reporting requirements.
* * * * *
(c) * * *
(3) The provisions of this section do not apply to licensees
subject to the notification requirements in Sec. 50.72 or Sec.
53.1630 of this chapter. They do apply to those licensees under part 50
of this chapter possessing material licensed under this part, who are
not subject to
[[Page 15787]]
the notification requirements in Sec. 50.72 of this chapter.
PART 40--DOMESTIC LICENSING OF SOURCE MATERIAL
0
103. 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
104. In Sec. 40.60, revise paragraph (c)(3) to read as follows:
Sec. 40.60 Reporting requirements.
* * * * *
(c) * * *
(3) The provisions of this section do not apply to licensees
subject to the notification requirements in Sec. 50.72 or Sec.
53.1630 of this chapter. They do apply to those licensees under part 50
of this chapter possessing material licensed under this part who are
not subject to the notification requirements in Sec. 50.72 of this
chapter.
PART 50--DOMESTIC LICENSING OF PRODUCTION AND UTILIZATION
FACILITIES
0
105. 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
106. In Sec. 50.47, revise paragraphs (a)(1) and (e) to read as
follows:
Sec. 50.47 Emergency plans.
(a)(1)(i) Except as provided in paragraph (d) of this section, no
initial operating license for a nuclear power reactor will be issued
under this part or under part 53 of this chapter unless a finding is
made by the NRC that there is reasonable assurance that adequate
protective measures can and will be taken in the event of a
radiological emergency. No finding under this section is necessary for
issuance of a renewed nuclear power reactor operating license.
(ii) No initial combined license under part 52 or 53 of this
chapter will be issued unless a finding is made by the NRC that there
is reasonable assurance that adequate protective measures can and will
be taken in the event of a radiological emergency. No finding under
this section is necessary for issuance of a renewed combined license.
(iii) If an application for an early site permit under subpart A of
part 52 of this chapter includes complete and integrated emergency
plans under Sec. 52.17(b)(2)(ii) of this chapter or an application for
an early site permit under subpart H of part 53 of this chapter
includes complete and integrated emergency plans under Sec.
53.1146(b)(2)(ii) of this chapter, no early site permit will be issued
unless a finding is made by the NRC that the emergency plans provide
reasonable assurance that adequate protective measures can and will be
taken in the event of a radiological emergency.
(iv) If an application for an early site permit proposes major
features of the emergency plans under Sec. 52.17(b)(2)(i) or Sec.
53.1146(b)(2)(i) of this chapter, no early site permit will be issued
unless a finding is made by the NRC that the major features are
acceptable in accordance with the applicable standards of either this
section and appendix E to this part or the applicable requirements of
Sec. 50.160, within the scope of emergency preparedness matters
addressed in the major features.
* * * * *
(e) Notwithstanding the requirements of paragraph (b) of this
section and the provisions of Sec. 52.103 or Sec. 53.1452 of this
chapter, a holder of a combined license under part 52 or 53 of this
chapter, as applicable, that is complying with the requirements of
paragraph (b) of this section and appendix E to this part may not load
fuel or operate except as provided in accordance with appendix E to
this part and Sec. 50.54(gg), and a holder of a combined license under
part 52 or 53 of this chapter that is complying with the requirements
of Sec. 50.160 may not load fuel or operate except as provided in
accordance with Sec. Sec. 50.160(c)(2) and 50.54(gg).
* * * * *
0
107. In Sec. 50.54, revise paragraph (gg)(1) introductory text to read
as follows:
Sec. 50.54 Conditions of licenses.
* * * * *
(gg)(1) Notwithstanding Sec. 52.103 or Sec. 53.1452 of this
chapter, if following the conduct of the exercise required by paragraph
IV.f.2.a of appendix E to this part or Sec. 50.160(c)(2), as
applicable, FEMA identifies one or more deficiencies in the state of
offsite emergency preparedness, the holder of a combined license under
part 52 or 53 of this chapter, as applicable, may operate at up to 5
percent of rated thermal power only if the Commission finds that the
state of onsite emergency preparedness provides reasonable assurance
that adequate protective measures can and will be taken in the event of
a radiological emergency. The NRC will base this finding on its
assessment of the applicant's onsite emergency plans against the
pertinent standards in either Sec. 50.47 and appendix E to this part,
or Sec. 50.160, as applicable. Review of the applicant's emergency
plans will include the following standards with offsite aspects:
* * * * *
0
108. In Sec. 50.160, revise paragraphs (b)(3) and (c)(2) to read as
follows:
Sec. 50.160 Emergency preparedness for small modular reactors, non-
light-water reactors, and non-power production or utilization
facilities.
* * * * *
(b) * * *
(3) Emergency planning zone. For an applicant whose analysis
required by Sec. 50.33(g)(2) or Sec. 53.1109(g)(2) of this chapter
meets the criteria in Sec. 50.33(g)(2)(i) or Sec. 53.1109(g)(2)(i) of
this chapter, as applicable, determine and describe the boundary and
physical characteristics of the EPZ in the emergency plan.
* * * * *
(c) * * *
(2) A holder of a combined license issued under part 52 or 53 of
this chapter before the Commission has made the finding under Sec.
52.103(g) or Sec. 53.1452(g) of this chapter, as applicable, must
establish, implement, and maintain an emergency preparedness program
that meets the requirements of paragraph (b) of this section, as
described in the approved emergency plan and license, and conduct an
initial exercise to demonstrate this compliance within 2 years before
the scheduled date for initial loading of fuel (or, for a fueled
manufactured reactor, within 2 years before the scheduled date for
initiating the removal of the features to prevent criticality required
under Sec. 53.620(d)(1) of this chapter).
0
109. In appendix B to part 50, revise the first paragraph in the
Introduction section, the first paragraph of section III, and section
IV to read as follows:
[[Page 15788]]
Appendix B to Part 50--Quality Assurance Criteria for Nuclear Power
Plants and Fuel Reprocessing Plants
Introduction. Every applicant for a construction permit is
required by the provisions of Sec. 50.34 or Sec. 53.1309 of this
chapter to include in its Preliminary Safety Analysis Report a
description of the quality assurance program to be applied to the
design, fabrication, construction, and testing of the structures,
systems, and components of the facility. Every applicant for an
operating license is required by the provisions of Sec. 50.34 or
Sec. 53.1369 of this chapter to include, in its Final Safety
Analysis Report, information pertaining to the managerial and
administrative controls to be used to assure safe operation. Every
applicant for a combined license is required by the provisions of
Sec. 52.79 or Sec. 53.1416 of this chapter to include in its Final
Safety Analysis Report a description of the quality assurance
applied to the design, and to be applied to the fabrication,
construction, and testing of the structures, systems, and components
of the facility and to the managerial and administrative controls to
be used to assure safe operation. For applications submitted after
September 27, 2007, every applicant for an early site permit is
required by the provisions of Sec. 52.17 or Sec. 53.1146 of this
chapter to include in its Site Safety Analysis Report a description
of the quality assurance program applied to site activities related
to the design, fabrication, construction, and testing of the
structures, systems, and components of a facility or facilities that
may be constructed on the site. Every applicant for a design
approval is required by the provisions of Sec. 52.137 or Sec.
53.1209 of this chapter to include in its Final Safety Analysis
Report a description of the quality assurance program applied to the
design of the structures, systems, and components of the facility.
Every applicant for a design certification is required by the
provisions of Sec. 52.47 or Sec. 53.1239 of this chapter to
include in its Final Safety Analysis Report a description of the
quality assurance program applied to the design of the structures,
systems, and components of the facility. Every applicant for a
manufacturing license is required by the provisions of Sec. 52.157
or Sec. 53.1279 of this chapter to include in its Final Safety
Analysis Report a description of the quality assurance program
applied to the design, and to be applied to the manufacture of, the
structures, systems, and components of the reactor. Nuclear power
plants and fuel reprocessing plants include structures, systems, and
components that prevent or mitigate the consequences of postulated
accidents that could cause undue risk to the health and safety of
the public. This appendix establishes quality assurance requirements
for the design, manufacture, construction, and operation of those
structures, systems, and components. The pertinent requirements of
this appendix apply to all activities affecting the safety-related
functions of those structures, systems, and components; these
activities include designing, purchasing, fabricating, handling,
shipping, storing, cleaning, erecting, installing, inspecting,
testing, operating, maintaining, repairing, refueling, and
modifying.
* * * * *
III. * * *
Measures shall be established to assure that applicable
regulatory requirements and the design bases, as defined in Sec.
50.2 and as specified in the license application, or the functional
design criteria, as defined in Sec. 53.020 of this chapter and as
specified in the license application, for those structures, systems,
and components to which this appendix applies are correctly
translated into specifications, drawings, procedures, and
instructions. These measures shall include provisions to assure that
appropriate quality standards are specified and included in design
documents and that deviations from such standards are controlled.
Measures shall also be established for the selection and review for
suitability of application of materials, parts, equipment, and
processes that are essential to the safety-related functions of the
structures, systems and components.
* * * * *
IV. Procurement Document Control
Measures shall be established to assure that applicable
regulatory requirements, design bases or functional design criteria,
and other requirements which are necessary to assure adequate
quality are suitably included or referenced in the documents for
procurement of material, equipment, and services, whether purchased
by the applicant or by its contractors or subcontractors. To the
extent necessary, procurement documents shall require contractors or
subcontractors to provide a quality assurance program consistent
with the pertinent provisions of this appendix.
* * * * *
0
110. In appendix E to part 50:
0
a. Revise paragraph I.1;
0
b. Add paragraph I.7;
0
c. Revise the first paragraph of section III;
0
d. Revise and republish section IV; and
0
e. Revise section V.
The addition and revisions read as follows:
Appendix E to Part 50--Emergency Planning and Preparedness for
Production and Utilization Facilities
* * * * *
I. Introduction
1. Each applicant for a construction permit is required by Sec.
50.34(a) or Sec. 53.1309(a)(4) of this chapter to include in the
preliminary safety analysis report a discussion of preliminary plans
for coping with emergencies. Each applicant for an operating license
is required by Sec. 50.34(b) or Sec. 53.1416 of this chapter to
include in the application plans for coping with emergencies. Each
applicant for an early site permit under subpart A of part 52 or
under subpart H of part 53 of this chapter may submit plans for
coping with emergencies under Sec. 52.17 or Sec. 53.1146 of this
chapter.
* * * * *
7. For a fueled manufactured reactor licensed under part 53 of
this chapter, the date for initiating the removal of the features to
prevent criticality required under Sec. 53.620(d)(1) is equivalent
to the initial loading of fuel in this appendix.
* * * * *
III. The Final Safety Analysis Report; Site Safety Analysis Report
The final safety analysis report or the site safety analysis
report for an early site permit that includes complete and
integrated emergency plans under Sec. 52.17(b)(2)(ii) or Sec.
53.1146(b)(2)(ii) of this chapter shall contain the plans for coping
with emergencies. The plans shall be an expression of the overall
concept of operation; they shall describe the essential elements of
advance planning that have been considered and the provisions that
have been made to cope with emergency situations. The plans shall
incorporate information about the emergency response roles of
supporting organizations and offsite agencies. That information
shall be sufficient to provide assurance of coordination among the
supporting groups and with the licensee. The site safety analysis
report for an early site permit which proposes major features must
address the relevant provisions of 10 CFR 50.47 and 10 CFR part 50,
appendix E, within the scope of emergency preparedness matters
addressed in the major features.
* * * * *
IV. Content of Emergency Plans
1. The applicant's emergency plans shall contain, but not
necessarily be limited to, information needed to demonstrate
compliance with the elements in this section IV., i.e., organization
for coping with radiological emergencies, assessment actions,
activation of emergency organization, notification procedures,
emergency facilities and equipment, training, maintaining emergency
preparedness, recovery, and onsite protective actions during hostile
action. In addition, the emergency response plans submitted by an
applicant for a nuclear power reactor operating license under 10 CFR
part 50 or 10 CFR part 53, or for an early site permit (as
applicable) or combined license under 10 CFR part 52 or 10 CFR part
53, shall contain information needed to demonstrate compliance with
the standards described in Sec. 50.47(b), and they will be
evaluated against those standards.
2. This nuclear power reactor license applicant shall also
provide an analysis of the time required to evacuate various sectors
and distances within the plume exposure pathway EPZ for transient
and permanent populations, using the most recent U.S. Census Bureau
data as of the date the applicant submits its application to the
NRC.
3. Nuclear power reactor licensees shall use NRC approved
evacuation time estimates (ETEs) and updates to the ETEs in the
formulation of protective action recommendations and shall provide
the ETEs and ETE updates to State and local
[[Page 15789]]
governmental authorities for use in developing offsite protective
action strategies.
4. Within 365 days of the date of the availability of the most
recent decennial census data from the U.S. Census Bureau, nuclear
power reactor licensees shall develop an ETE analysis using this
decennial data and submit it under Sec. 50.4 or Sec. 53.040 of
this chapter to the NRC. These licensees shall submit this ETE
analysis to the NRC at least 180 days before using it to form
protective action recommendations and providing it to State and
local governmental authorities for use in developing offsite
protective action strategies.
5. During the years between decennial censuses, nuclear power
reactor licensees shall estimate EPZ permanent resident population
changes once a year, but no later than 365 days from the date of the
previous estimate, using the most recent U.S. Census Bureau annual
resident population estimate and State/local government population
data, if available. These licensees shall maintain these estimates
so that they are available for NRC inspection during the period
between decennial censuses and shall submit these estimates to the
NRC with any updated ETE analysis.
6. If at any time during the decennial period, the EPZ permanent
resident population increases such that it causes the longest ETE
value for the 2-mile zone or 5-mile zone, including all affected
Emergency Response Planning Areas, or for the entire 10-mile EPZ to
increase by 25 percent or 30 minutes, whichever is less, from the
nuclear power reactor licensee's currently NRC-approved or updated
ETE, the licensee shall update the ETE analysis to reflect the
impact of that population increase. The licensee shall submit the
updated ETE analysis to the NRC under Sec. 50.4 or Sec. 53.040 of
this chapter no later than 365 days after the licensee's
determination that the criteria for updating the ETE have been met
and at least 180 days before using it to form protective action
recommendations and providing it to State and local governmental
authorities for use in developing offsite protective action
strategies.
7. After an applicant for a combined license under part 52 or
part 53 of this chapter receives its license, the licensee shall
conduct at least one review of any changes in the population of its
EPZ at least 365 days prior to its scheduled fuel load. The licensee
shall estimate EPZ permanent resident population changes using the
most recent U.S. Census Bureau annual resident population estimate
and State/local government population data, if available. If the EPZ
permanent resident population increases such that it causes the
longest ETE value for the 2-mile zone or 5-mile zone, including all
affected Emergency Response Planning Areas, or for the entire 10-
mile EPZ, to increase by 25 percent or 30 minutes, whichever is
less, from the licensee's currently approved ETE, the licensee shall
update the ETE analysis to reflect the impact of that population
increase. The licensee shall submit the updated ETE analysis to the
NRC for review under Sec. 50.4 or Sec. 53.040 of this chapter no
later than 365 days before the licensee's scheduled fuel load.
A. Organization
The organization for coping with radiological emergencies shall
be described, including definition of authorities, responsibilities,
and duties of individuals assigned to the licensee's emergency
organization and the means for notification of such individuals in
the event of an emergency. Specifically, the following shall be
included:
1. A description of the normal plant operating organization.
2. A description of the onsite emergency response organization
(ERO) with a detailed discussion of:
a. Authorities, responsibilities, and duties of the
individual(s) who will take charge during an emergency;
b. Plant staff emergency assignments;
c. Authorities, responsibilities, and duties of an onsite
emergency coordinator who shall be in charge of the exchange of
information with offsite authorities responsible for coordinating
and implementing offsite emergency measures.
3. A description, by position and function to be performed, of
the licensee's headquarters personnel who will be sent to the plant
site to augment the onsite emergency organization.
4. Identification, by position and function to be performed, of
persons within the licensee organization who will be responsible for
making offsite dose projections, and a description of how these
projections will be made and the results transmitted to State and
local authorities, NRC, and other appropriate governmental entities.
5. Identification, by position and function to be performed, of
other employees of the licensee with special qualifications for
coping with emergency conditions that may arise. Other persons with
special qualifications, such as consultants, who are not employees
of the licensee and who may be called upon for assistance for
emergencies shall also be identified. The special qualifications of
these persons shall be described.
6. A description of the local offsite services to be provided in
support of the licensee's emergency organization.
7. Identification of, and a description of the assistance
expected from, appropriate State, local, and Federal agencies with
responsibilities for coping with emergencies, including hostile
action at the site. For purposes of this appendix, ``hostile
action'' is defined as an act directed toward a nuclear power plant
or its personnel that includes the use of violent force to destroy
equipment, take hostages, and/or intimidate the licensee to achieve
an end. This includes attack by air, land, or water using guns,
explosives, projectiles, vehicles, or other devices used to deliver
destructive force.
8. Identification of the State and/or local officials
responsible for planning for, ordering, and controlling appropriate
protective actions, including evacuations when necessary.
9. For nuclear power reactor licensees, a detailed analysis
demonstrating that on-shift personnel assigned emergency plan
implementation functions are not assigned responsibilities that
would prevent the timely performance of their assigned functions as
specified in the emergency plan.
B. Assessment Actions
1. The means to be used for determining the magnitude of, and
for continually assessing the impact of, the release of radioactive
materials shall be described, including emergency action levels that
are to be used as criteria for determining the need for notification
and participation of local and State agencies, the Commission, and
other Federal agencies, and the emergency action levels that are to
be used for determining when and what type of protective measures
should be considered within and outside the site boundary to protect
health and safety. The emergency action levels shall be based on in-
plant conditions and instrumentation in addition to onsite and
offsite monitoring. For nuclear power reactor licensees, these
action levels must include hostile action that may adversely affect
the nuclear power plant. The initial emergency action levels shall
be discussed and agreed on by the applicant or licensee and state
and local governmental authorities, and approved by the NRC.
Thereafter, emergency action levels shall be reviewed with the State
and local governmental authorities on an annual basis.
2. A licensee desiring to change its entire emergency action
level scheme shall submit an application for an amendment to its
license and receive NRC approval before implementing the change.
Licensees shall follow the change process in Sec. 50.54(q) or Sec.
53.1565(d)(3) of this chapter for all other emergency action level
changes.
C. Activation of Emergency Organization
1. The entire spectrum of emergency conditions that involve the
alerting or activating of progressively larger segments of the total
emergency organization shall be described. The communication steps
to be taken to alert or activate emergency personnel under each
class of emergency shall be described. Emergency action levels
(based not only on onsite and offsite radiation monitoring
information but also on readings from a number of sensors that
indicate a potential emergency, such as the pressure in containment
and the response of the Emergency Core Cooling System) for
notification of offsite agencies shall be described. The existence,
but not the details, of a message authentication scheme shall be
noted for such agencies. The emergency classes defined shall
include: (1) Notification of unusual events, (2) alert, (3) site
area emergency, and (4) general emergency. These classes are further
discussed in NUREG-0654/FEMA-REP-1.
2. Nuclear power reactor licensees shall establish and maintain
the capability to assess, classify, and declare an emergency
condition within 15 minutes after the availability of indications to
plant operators that an emergency action level has been exceeded and
shall promptly declare the emergency condition as soon as possible
following identification of the appropriate emergency classification
level. Licensees shall not construe these criteria as a grace period
to attempt to restore plant conditions to avoid declaring an
emergency action due
[[Page 15790]]
to an emergency action level that has been exceeded. Licensees shall
not construe these criteria as preventing implementation of response
actions deemed by the licensee to be necessary to protect public
health and safety provided that any delay in declaration does not
deny the State and local authorities the opportunity to implement
measures necessary to protect the public health and safety.
D. Notification Procedures
1. Administrative and physical means for notifying local, State,
and Federal officials and agencies and agreements reached with these
officials and agencies for the prompt notification of the public and
for public evacuation or other protective measures, should they
become necessary, shall be described. This description shall include
identification of the appropriate officials, by title and agency, of
the State and local government agencies within the EPZs.
2. Provisions shall be described for yearly dissemination to the
public within the plume exposure pathway EPZ of basic emergency
planning information, such as the methods and times required for
public notification and the protective actions planned if an
accident occurs, general information as to the nature and effects of
radiation, and a listing of local broadcast stations that will be
used for dissemination of information during an emergency. Signs or
other measures shall also be used to disseminate to any transient
population within the plume exposure pathway EPZ appropriate
information that would be helpful if an accident occurs.
3. A licensee shall have the capability to notify responsible
State and local governmental agencies within 15 minutes after
declaring an emergency. The licensee shall demonstrate that the
appropriate governmental authorities have the capability to make a
public alerting and notification decision promptly on being informed
by the licensee of an emergency condition. Prior to initial
operation greater than 5 percent of rated thermal power of the first
reactor at a site, each nuclear power reactor licensee shall
demonstrate that administrative and physical means have been
established for alerting and providing prompt instructions to the
public within the plume exposure pathway EPZ. The design objective
of the prompt public alert and notification system shall be to have
the capability to essentially complete the initial alerting and
initiate notification of the public within the plume exposure
pathway EPZ within about 15 minutes. The use of this alerting and
notification capability will range from immediate alerting and
notification of the public (within 15 minutes of the time that State
and local officials are notified that a situation exists requiring
urgent action) to the more likely events where there is substantial
time available for the appropriate governmental authorities to make
a judgment whether or not to activate the public alert and
notification system. The alerting and notification capability shall
additionally include administrative and physical means for a backup
method of public alerting and notification capable of being used in
the event the primary method of alerting and notification is
unavailable during an emergency to alert or notify all or portions
of the plume exposure pathway EPZ population. The backup method
shall have the capability to alert and notify the public within the
plume exposure pathway EPZ, but does not need to meet the 15-minute
design objective for the primary prompt public alert and
notification system. When there is a decision to activate the alert
and notification system, the appropriate governmental authorities
will determine whether to activate the entire alert and notification
system simultaneously or in a graduated or staged manner. The
responsibility for activating such a public alert and notification
system shall remain with the appropriate governmental authorities.
E. Emergency Facilities and Equipment
Adequate provisions shall be made and described for emergency
facilities and equipment, including:
1. Equipment at the site for personnel monitoring;
2. Equipment for determining the magnitude of and for
continuously assessing the impact of the release of radioactive
materials to the environment;
3. Facilities and supplies at the site for decontamination of
onsite individuals;
4. Facilities and medical supplies at the site for appropriate
emergency first aid treatment;
5. Arrangements for medical service providers qualified to
handle radiological emergencies onsite;
6. Arrangements for transportation of contaminated injured
individuals from the site to specifically identified treatment
facilities outside the site boundary;
7. Arrangements for treatment of individuals injured in support
of licensed activities on the site at treatment facilities outside
the site boundary;
8. a. (i) A licensee onsite technical support center and an
emergency operations facility from which effective direction can be
given and effective control can be exercised during an emergency;
(ii) For nuclear power reactor licensees, a licensee onsite
operational support center;
b. For a nuclear power reactor licensee's emergency operations
facility required by paragraph 8.a of this section, either a
facility located between 10 miles and 25 miles of the nuclear power
reactor site(s), or a primary facility located less than 10 miles
from the nuclear power reactor site(s) and a backup facility located
between 10 miles and 25 miles of the nuclear power reactor site(s).
An emergency operations facility may serve more than one nuclear
power reactor site. A licensee desiring to locate an emergency
operations facility more than 25 miles from a nuclear power reactor
site shall request prior Commission approval by submitting an
application for an amendment to its license. For an emergency
operations facility located more than 25 miles from a nuclear power
reactor site, provisions must be made for locating NRC and offsite
responders closer to the nuclear power reactor site so that NRC and
offsite responders can interact face-to-face with emergency response
personnel entering and leaving the nuclear power reactor site.
Provisions for locating NRC and offsite responders closer to a
nuclear power reactor site that is more than 25 miles from the
emergency operations facility must include the following:
(1) Space for members of an NRC site team and Federal, State,
and local responders;
(2) Additional space for conducting briefings with emergency
response personnel;
(3) Communication with other licensee and offsite emergency
response facilities;
(4) Access to plant data and radiological information; and
(5) Access to copying equipment and office supplies;
c. For a nuclear power reactor licensee's emergency operations
facility required by paragraph 8.a of this section, a facility
having the following capabilities:
(1) The capability for obtaining and displaying plant data and
radiological information for each reactor at a nuclear power reactor
site and for each nuclear power reactor site that the facility
serves;
(2) The capability to analyze plant technical information and
provide technical briefings on event conditions and prognosis to
licensee and offsite response organizations for each reactor at a
nuclear power reactor site and for each nuclear power reactor site
that the facility serves; and
(3) The capability to support response to events occurring
simultaneously at more than one nuclear power reactor site if the
emergency operations facility serves more than one site; and
d. For nuclear power reactor licensees, an alternative facility
(or facilities) that would be accessible even if the site is under
threat of or experiencing hostile action, to function as a staging
area for augmentation of emergency response staff and collectively
having the following characteristics: the capability for
communication with the emergency operations facility, control room,
and plant security; the capability to perform offsite notifications;
and the capability for engineering assessment activities, including
damage control team planning and preparation, for use when onsite
emergency facilities cannot be safely accessed during hostile
action. The requirements in this paragraph 8.d must be implemented
no later than December 23, 2014, with the exception of the
capability for staging emergency response organization personnel at
the alternative facility (or facilities) and the capability for
communications with the emergency operations facility, control room,
and plant security, which must be implemented no later than June 20,
2012.
e. A licensee shall not be subject to the requirements of
paragraph 8.b of this section for an existing emergency operations
facility approved as of December 23, 2011;
9. At least one onsite and one offsite communications system;
each system shall have a backup power source. All communication
plans shall have arrangements for emergencies, including titles and
alternates for those in charge at both ends of the communication
links and the primary and backup means of communication. Where
consistent with the function of the governmental agency, these
arrangements will include:
[[Page 15791]]
a. Provision for communications with contiguous State/local
governments within the plume exposure pathway EPZ. Such
communications shall be tested monthly.
b. Provision for communications with Federal emergency response
organizations. Such communications systems shall be tested annually.
c. Provision for communications among the nuclear power reactor
control room, the onsite technical support center, and the emergency
operations facility; and among the nuclear facility, the principal
State and local emergency operations centers, and the field
assessment teams. Such communications systems shall be tested
annually.
d. Provisions for communications by the licensee with NRC
Headquarters and the appropriate NRC Regional Office Operations
Center from the nuclear power reactor control room, the onsite
technical support center, and the emergency operations facility.
Such communications shall be tested monthly.
F. Training
1. The program to provide for: (a) The training of employees and
exercising, by periodic drills, of emergency plans to ensure that
employees of the licensee are familiar with their specific emergency
response duties, and (b) The participation in the training and
drills by other persons whose assistance may be needed in the event
of a radiological emergency shall be described. This shall include a
description of specialized initial training and periodic retraining
programs to be provided to each of the following categories of
emergency personnel:
i. Directors and/or coordinators of the plant emergency
organization;
ii. Personnel responsible for accident assessment, including
control room shift personnel;
iii. Radiological monitoring teams;
iv. Fire control teams (fire brigades);
v. Repair and damage control teams;
vi. First aid and rescue teams;
vii. Medical support personnel;
viii. Licensee's headquarters support personnel;
ix. Security personnel.
In addition, a radiological orientation training program shall
be made available to local services personnel; e.g., local emergency
services/Civil Defense, local law enforcement personnel, local news
media persons.
2. The plan shall describe provisions for the conduct of
emergency preparedness exercises as follows: Exercises shall test
the adequacy of timing and content of implementing procedures and
methods, test emergency equipment and communications networks, test
the public alert and notification system, and ensure that emergency
organization personnel are familiar with their duties.\3\
a. A full participation \4\ exercise which tests as much of the
licensee, State, and local emergency plans as is reasonably
achievable without mandatory public participation shall be conducted
for each site at which a power reactor is located. Nuclear power
reactor licensees shall submit exercise scenarios under Sec. 50.4
or Sec. 53.040 of this chapter at least 60 days before use in a
full participation exercise required by this paragraph 2.a.
(i) For an operating license issued under part 50 or part 53 of
this chapter, this exercise must be conducted within 2 years before
the issuance of the first operating license for full power (one
authorizing operation above 5 percent of rated thermal power) of the
first reactor and shall include participation by each State and
local government within the plume exposure pathway EPZ and each
state within the ingestion exposure pathway EPZ. If the full
participation exercise is conducted more than 1 year prior to
issuance of an operating licensee for full power, an exercise which
tests the licensee's onsite emergency plans must be conducted within
1 year before issuance of an operating license for full power. This
exercise need not have State or local government participation.
(ii) For a combined license issued under part 52 or part 53 of
this chapter, this exercise must be conducted within 2 years of the
scheduled date for initial loading of fuel. If the first full
participation exercise is conducted more than 1 year before the
scheduled date for initial loading of fuel, an exercise which tests
the licensee's onsite emergency plans must be conducted within 1
year before the scheduled date for initial loading of fuel. This
exercise need not have State or local government participation. If
FEMA identifies one or more deficiencies in the state of offsite
emergency preparedness as the result of the first full participation
exercise, or if the Commission finds that the state of emergency
preparedness does not provide reasonable assurance that adequate
protective measures can and will be taken in the event of a
radiological emergency, the provisions of Sec. 50.54(gg) apply.
(iii) For a combined license issued under part 52 or part 53 of
this chapter, if the applicant currently has an operating reactor at
the site, an exercise, either full or partial participation,\5\
shall be conducted for each subsequent reactor constructed on the
site. This exercise may be incorporated in the exercise requirements
of Sections IV.F.2.b. and c. in this appendix. If FEMA identifies
one or more deficiencies in the state of offsite emergency
preparedness as the result of this exercise for the new reactor, or
if the Commission finds that the state of emergency preparedness
does not provide reasonable assurance that adequate protective
measures can and will be taken in the event of a radiological
emergency, the provisions of Sec. 50.54(gg) apply.
b. Each licensee at each site shall conduct a subsequent
exercise of its onsite emergency plan every 2 years. Nuclear power
reactor licensees shall submit exercise scenarios under Sec. 50.4
or Sec. 53.040 of this chapter at least 60 days before use in an
exercise required by this paragraph 2.b. The exercise may be
included in the full participation biennial exercise required by
paragraph 2.c. of this section. In addition, the licensee shall take
actions necessary to ensure that adequate emergency response
capabilities are maintained during the interval between biennial
exercises by conducting drills, including at least one drill
involving a combination of some of the principal functional areas of
the licensee's onsite emergency response capabilities. The principal
functional areas of emergency response include activities such as
management and coordination of emergency response, accident
assessment, event classification, notification of offsite
authorities, assessment of the onsite and offsite impact of
radiological releases, protective action recommendation development,
protective action decision making, plant system repair and
mitigative action implementation. During these drills, activation of
all of the licensee's emergency response facilities (Technical
Support Center (TSC), Operations Support Center (OSC), and the
Emergency Operations Facility (EOF)) would not be necessary,
licensees would have the opportunity to consider accident management
strategies, supervised instruction would be permitted, operating
staff in all participating facilities would have the opportunity to
resolve problems (success paths) rather than have controllers
intervene, and the drills may focus on the onsite exercise training
objectives.
c. Offsite plans for each site shall be exercised biennially
with full participation by each offsite authority having a role
under the radiological response plan. Where the offsite authority
has a role under a radiological response plan for more than one
site, it shall fully participate in one exercise every 2 years and
shall, at least, partially participate in other offsite plan
exercises in this period. If two different licensees each have
licensed facilities located either on the same site or on adjacent,
contiguous sites, and share most of the elements defining co-located
licensees,\6\ then each licensee shall:
(1) Conduct an exercise biennially of its onsite emergency plan;
(2) Participate quadrennially in an offsite biennial full or
partial participation exercise;
(3) Conduct emergency preparedness activities and interactions
in the years between its participation in the offsite full or
partial participation exercise with offsite authorities, to test and
maintain interface among the affected State and local authorities
and the licensee. Co-located licensees shall also participate in
emergency preparedness activities and interaction with offsite
authorities for the period between exercises;
(4) Conduct a hostile action exercise of its onsite emergency
plan in each exercise cycle; and
(5) Participate in an offsite biennial full or partial
participation hostile action exercise in alternating exercise
cycles.
d. Each State with responsibility for nuclear power reactor
emergency preparedness should fully participate in the ingestion
pathway portion of exercises at least once every exercise cycle. In
States with more than one nuclear power reactor plume exposure
pathway EPZ, the State should rotate this participation from site to
site. Each State with responsibility for nuclear power reactor
emergency preparedness should fully participate in a hostile action
exercise at least once every cycle. States with more than one
nuclear power reactor plume exposure pathway EPZ should rotate this
participation from site to site.
[[Page 15792]]
e. Licensees shall enable any State or local government located
within the plume exposure pathway EPZ to participate in the
licensee's drills when requested by such State or local government.
f. Remedial exercises will be required if the emergency plan is
not satisfactorily tested during the biennial exercise, such that
NRC, in consultation with FEMA, cannot (1) find reasonable assurance
that adequate protective measures can and will be taken in the event
of a radiological emergency or (2) determine that the Emergency
Response Organization (ERO) has maintained key skills specific to
emergency response. The extent of State and local participation in
remedial exercises must be sufficient to show that appropriate
corrective measures have been taken regarding the elements of the
plan not properly tested in the previous exercises.
g. All exercises, drills, and training that provide performance
opportunities to develop, maintain, or demonstrate key skills must
provide for formal critiques in order to identify weak or deficient
areas that need correction. Any weaknesses or deficiencies that are
identified in a critique of exercises, drills, or training must be
corrected.
h. The participation of State and local governments in an
emergency exercise is not required to the extent that the applicant
has identified those governments as refusing to participate further
in emergency planning activities, pursuant to Sec. 50.47(c)(1). In
such cases, an exercise shall be held with the applicant or licensee
and such governmental entities as elect to participate in the
emergency planning process.
i. Licensees shall use drill and exercise scenarios that provide
reasonable assurance that anticipatory responses will not result
from preconditioning of participants. Such scenarios for nuclear
power reactor licensees must include a wide spectrum of radiological
releases and events, including hostile action. Exercise and drill
scenarios as appropriate must emphasize coordination among onsite
and offsite response organizations.
j. (i) The exercises conducted under paragraph 2 of this section
by nuclear power reactor licensees must provide the opportunity for
the ERO to demonstrate proficiency in the key skills necessary to
implement the principal functional areas of emergency response
identified in paragraph 2.b of this section.
(ii) Each exercise must provide the opportunity for the ERO to
demonstrate key skills specific to emergency response duties in the
control room, TSC, OSC, EOF, and joint information center.
(iii) In each 8-calendar-year exercise cycle, nuclear power
reactor licensees shall vary the content of scenarios during
exercises conducted under paragraph 2 of this section to provide the
opportunity for the ERO to demonstrate proficiency in the key skills
necessary to respond to the following scenario elements:
(1) Hostile action directed at the plant site;
(2) No radiological release or an unplanned minimal radiological
release that does not require public protective actions;
(3) An initial classification of, or rapid escalation to, a Site
Area Emergency or General Emergency;
(4) Implementation of strategies, procedures, and guidance under
Sec. 50.155(b)(2) for applicants and licensees under parts 50 and
52 of this chapter; and
(5) Integration of offsite resources with onsite response.
(iv) The licensee shall maintain a record of exercises conducted
during each 8-year exercise cycle that documents the content of
scenarios used to comply with the requirements of section IV.F.2.j
of this appendix.
(v) Each licensee shall conduct a hostile action exercise for
each of its sites no later than December 31, 2015.
(vi) The first 8-year exercise cycle for a site will begin in
the calendar year in which the first hostile action exercise is
conducted. For a site licensed under 10 CFR part 52 or 10 CFR part
53 using 10 CFR 50.47 and this appendix, the first 8-year exercise
cycle begins in the calendar year of the initial exercise required
by section IV.F.2.a of this appendix.
G. Maintaining Emergency Preparedness
Provisions to be employed to ensure that the emergency plan, its
implementing procedures, and emergency equipment and supplies are
maintained up to date shall be described.
H. Recovery
Criteria to be used to determine when, following an accident,
reentry of the facility would be appropriate or when operation could
be resumed shall be described.
I. Onsite Protective Actions During Hostile Action
For nuclear power reactor licensees, a range of protective
actions to protect onsite personnel during hostile action must be
developed to ensure the continued ability of the licensee to safely
shut down the reactor and perform the functions of the licensee's
emergency plan.
V. Implementing Procedures
No less than 180 days before the scheduled issuance of an
operating license for a nuclear power reactor or a license to
possess nuclear material, or the scheduled date for initial loading
of fuel for a combined license under part 52 or part 53 of this
chapter, the applicant's or licensee's detailed implementing
procedures for its emergency plan shall be submitted to the
Commission as specified in Sec. 50.4 or Sec. 53.040.
* * * * *
\3\ Use of site specific simulators or computers is acceptable
for any exercise.
\4\ Full participation when used in conjunction with emergency
preparedness exercises for a particular site means appropriate
offsite local and State authorities and licensee personnel
physically and actively take part in testing their integrated
capability to adequately assess and respond to an accident at a
commercial nuclear power plant. Full participation includes testing
major observable portions of the onsite and offsite emergency plans
and mobilization of State, local and licensee personnel and other
resources in sufficient numbers to verify the capability to respond
to the accident scenario.
\5\ Partial participation when used in conjunction with
emergency preparedness exercises for a particular site means
appropriate offsite authorities shall actively take part in the
exercise sufficient to test direction and control functions; i.e.,
(a) protective action decision making related to emergency action
levels, and (b) communication capabilities among affected State and
local authorities and the licensee.
\6\ Co-located licensees are two different licensees whose
licensed facilities are located either on the same site or on
adjacent, contiguous sites, and that share most of the following
emergency planning and siting elements:
a. Plume exposure and ingestion emergency planning zones;
b. Offsite governmental authorities;
c. Offsite emergency response organizations;
d. Public notification system; and/or
e. Emergency facilities.
* * * * *
PART 51--ENVIRONMENTAL PROTECTION REGULATIONS FOR DOMESTIC
LICENSING AND RELATED REGULATORY FUNCTIONS
0
111. The authority citation for part 51 is revised to read as follows:
Authority: Atomic Energy Act of 1954, secs. 161, 193 (42 U.S.C.
2201, 2243); Energy Reorganization Act of 1974, secs. 201, 202 (42
U.S.C. 5841, 5842); National Environmental Policy Act of 1969 (42
U.S.C. 4332, 4334, 4335); Nuclear Waste Policy Act of 1982, secs.
144(f), 121, 135, 141, 148 (42 U.S.C. 10134(f), 10141, 10155, 10161,
10168); 44 U.S.C. 3504 note.
0
112. In Sec. 51.4, revise the definition for ``Construction'' to read
as follows:
Sec. 51.4 Definitions.
* * * * *
Construction means:
(1)(i) For production and utilization facilities licensed under 10
CFR part 50 or 10 CFR part 52, the activities in 10 CFR 50.10(a)(1),
and does not mean the activities in 10 CFR 50.10(a)(2).
(ii) For utilization facilities licensed under 10 CFR part 53, the
activities in paragraph (1) of the definition of construction in 10 CFR
53.020, and does not mean the activities in paragraph (2) of the
definition of construction in 10 CFR 53.020.
(2) For materials licenses, the activities in paragraph (2)(i) of
this definition, and does not mean the activities in paragraph (2)(ii)
of this definition.
(i) Taking any site-preparation activity at the site of a facility
subject to the regulations in 10 CFR parts 30, 36, 40, and 70 that has
a reasonable nexus to radiological health and safety or the common
defense and security.
(ii) Construction does not include:
[[Page 15793]]
(A) The activities listed in 10 CFR 50.10(a)(2)(i)-(viii); or
(B) Taking any other action that has no reasonable nexus to
radiological health and safety or the common defense and security.
* * * * *
0
113. In Sec. 51.20, revise paragraphs (b)(1) and (2) to read as
follows:
Sec. 51.20 Criteria for and identification of licensing and
regulatory actions requiring environmental impact statements.
* * * * *
(b) * * *
(1) Issuance of a limited work authorization or a permit to
construct a nuclear power reactor, testing facility, or fuel
reprocessing plant under part 50 of this chapter, issuance of an early
site permit under part 52 of this chapter, or issuance of a limited
work authorization, construction permit, or early site permit under
part 53 of this chapter.
(2) Issuance or renewal of a full power or design capacity license
to operate a nuclear power reactor, testing facility, or fuel
reprocessing plant under part 50 or part 53 of this chapter, or a
combined license under part 52 or part 53 of this chapter.
* * * * *
0
114. In Sec. 51.22, revise paragraphs (c)(3) introductory text, (c)(9)
introductory text, (c)(12) introductory text, (c)(17), (c)(22), and
(c)(23) to read as follows:
Sec. 51.22 Criterion for categorical exclusion; identification of
licensing and regulatory actions eligible for categorical exclusion or
otherwise not requiring environmental review.
* * * * *
(c) * * *
(3) Amendments to parts 20, 30, 31, 32, 33, 34, 35, 37, 39, 40, 50,
51, 52, 53, 54, 60, 61, 63, 70, 71, 72, 73, 74, 81, and 100 of this
chapter which relate to--
* * * * *
(9) Issuance of an amendment to a permit or license for a reactor
under part 50, part 52, or part 53 of this chapter that changes a
requirement or issuance of an exemption from a requirement, with
respect to installation or use of a facility component located within
the restricted area, as defined in part 20 of this chapter; or the
issuance of an amendment to a permit or license for a reactor under
part 50, part 52, or part 53 of this chapter that changes an inspection
or a surveillance requirement; provided that:
* * * * *
(12) Issuance of an amendment to a license under parts 50, 52, 53,
60, 61, 63, 70, 72, or 75 of this chapter relating solely to safeguards
matters (i.e., protection against sabotage or loss or diversion of
special nuclear material) or issuance of an approval of a safeguards
plan submitted under parts 50, 52, 53, 70, 72, and 73 of this chapter,
provided that the amendment or approval does not involve any
significant construction impacts. These amendments and approvals are
confined to--
* * * * *
(17) Issuance of an amendment to a permit or license under part 30,
part 40, part 50, part 52, part 53, or part 70 of this chapter which
deletes any limiting condition of operation or monitoring requirement
based on or applicable to any matter subject to the provisions of the
Federal Water Pollution Control Act.
* * * * *
(22) Issuance of a standard design approval under part 52 or part
53 of this chapter.
(23) The Commission finding for a combined license under Sec.
52.103(g) or Sec. 53.1452(g) of this chapter.
* * * * *
Sec. 51.26 [Amended]
0
115. In Sec. 51.26, in paragraph (d) remove the phrase ``under part
52'' and add in its place the phrase ``under 10 CFR part 52 or part
53,''.
0
116. In Sec. 51.30, revise paragraph (a) introductory text and
paragraphs (d) and (e) to read as follows:
Sec. 51.30 Environmental assessment.
(a) An environmental assessment for proposed actions, other than
those for a standard design certification under 10 CFR part 52 or part
53, or a manufacturing license under 10 CFR part 52 or part 53, shall
identify the proposed action and include:
* * * * *
(d) An environmental assessment for a standard design certification
under subpart B of part 52, or under subpart H of part 53 of this
chapter must identify the proposed action and will be limited to the
consideration of the costs and benefits of severe accident mitigation
design alternatives and the bases for not incorporating severe accident
mitigation design alternatives in the design certification. An
environmental assessment for an amendment to a design certification
will be limited to the consideration of whether the design change which
is the subject of the proposed amendment renders a severe accident
mitigation design alternative previously rejected in the earlier
environmental assessment to become cost beneficial, or results in the
identification of new severe accident mitigation design alternatives,
in which case the costs and benefits of new severe accident mitigation
design alternatives and the bases for not incorporating new severe
accident mitigation design alternatives in the design certification
must be addressed.
(e) An environmental assessment for a manufacturing license under
subpart F of part 52 of this chapter or under subpart H of part 53 of
this chapter must identify the proposed action and will be limited to
the consideration of the costs and benefits of severe accident
mitigation design alternatives and the bases for not incorporating
severe accident mitigation design alternatives in the manufacturing
license. An environmental assessment for an amendment to a
manufacturing license will be limited to consideration of whether the
design change which is the subject of the proposed amendment either
renders a severe accident mitigation design alternative previously
rejected in an environmental assessment to become cost beneficial, or
results in the identification of new severe accident mitigation design
alternatives, in which case the costs and benefits of new severe
accident mitigation design alternatives and the bases for not
incorporating new severe accident mitigation design alternatives in the
manufacturing license must be addressed. In either case, the
environmental assessment will not address the environmental impacts
associated with manufacturing the reactor under the manufacturing
license.
Sec. 51.31 [Amended]
0
117. In Sec. 51.31, in paragraph (a), remove the phrase ``under part
52'' and add in its place the phrase ``under part 52 or part 53''.
Sec. 51.32 [Amended]
0
118. In Sec. 51.32, in paragraphs (b)(1) and (3), remove the phrase
``of part 52 of this chapter'' and add in its place the phrase ``of
part 52 or subpart H of part 53 of this chapter''.
Sec. 51.49 [Amended]
0
119. In Sec. 51.49, in paragraph (c) introductory text, remove the
phrase ``of part 52 of this chapter'' and add in its place the phrase
``of part 52 or under subpart H of part 53 of this chapter''.
Sec. 51.50 [Amended]
0
120. In Sec. 51.50, wherever it may appear, remove the phrase ``in
accordance with Sec. 50.36b of this chapter'' and add in its place the
phrase ``in accordance with Sec. 50.36b or Sec. 53.1112 of this
chapter''.
[[Page 15794]]
Sec. 51.53 [Amended]
0
121. In Sec. 51.53, in paragraph (d), remove the phrase ``under Sec.
50.82 of this chapter'' and add in its place the phrase ``under Sec.
50.82 or Sec. 53.1080 of this chapter''.
Sec. 51.54 [Amended]
0
122. In Sec. 51.54, in paragraph (a), remove the phrase ``of part 52
of this chapter'' and add in its place the phrase ``of part 52 or under
subpart H of part 53 of this chapter''.
Sec. 51.55 [Amended]
0
123. In Sec. 51.55, in paragraph (a), remove the phrase ``of part 52
of this chapter'' and add in its place the phrase ``of part 52 or under
subpart H of part 53 of this chapter''.
0
124. In Sec. 51.58, revise paragraph (b) to read as follows:
Sec. 51.58 Environmental report--number of copies; distribution.
* * * * *
(b) Each applicant for a license to manufacture a nuclear power
reactor, or for an amendment to a license to manufacture, seeking
approval of the final design of the nuclear power reactor under subpart
F of part 52 or under subpart H of part 53 of this chapter, shall
submit to the Commission an environmental report or any supplement to
an environmental report in the manner specified in Sec. 50.3 or Sec.
53.040 of this chapter. The applicant shall maintain the capability to
generate additional copies of the environmental report or any
supplement to the environmental report for subsequent distribution to
parties and Boards in the NRC proceeding; Federal, State, and local
officials; and any affected Indian Tribes, in accordance with written
instructions issued by the Director, Office of Nuclear Reactor
Regulation.
0
125. In Sec. 51.77, revise paragraph (a) introductory text to read as
follows:
Sec. 51.77 Distribution of draft environmental impact statement.
(a) In addition to the distribution authorized by Sec. 51.74, a
copy of a draft environmental statement for a licensing action for a
production or utilization facility, except an action authorizing
issuance, amendment, or renewal of a license to manufacture a nuclear
power reactor pursuant to 10 CFR part 52, subpart F or 10 CFR part 53,
subparts H or I will also be distributed to:
* * * * *
Sec. 51.92 [Amended]
0
126. In Sec. 51.92, in paragraph (b), remove the phrase ``10 CFR part
52'' and add in its place the phrase ``10 CFR part 52 or part 53''.
Sec. 51.95 [Amended]
0
127. In Sec. 51.95, in paragraph (c) introductory text remove the
phrase ``under 10 CFR parts 52 or 54'' and add in its place the phrase
``under 10 CFR part 52, part 53, or part 54''.
0
128. In Sec. 51.101, revise paragraph (a)(2) to read as follows:
Sec. 51.101 Limitations on actions.
(a) * * *
(2) Any action concerning the proposal taken by an applicant which
would--
(i) Have an adverse environmental impact; or
(ii) Limit the choice of reasonable alternatives that may be
grounds for denial of the license. In the case of an application
covered by Sec. 30.32(f), Sec. 40.31(f), Sec. 50.10(c), Sec.
53.1130, Sec. 70.21(f), or Sec. 72.16 and Sec. 72.34 of this
chapter, the provisions of this paragraph will be applied in accordance
with Sec. 30.33(a)(5), Sec. 40.32(e), Sec. 50.10(c), Sec. 53.1130,
Sec. 70.23(a)(7), or Sec. 72.40(b) of this chapter, as appropriate.
* * * * *
Sec. 51.103 [Amended]
0
129. In Sec. 51.103, in paragraph (a)(6), remove the phrase ``under 10
CFR 50.10'' and add in its place the phrase ``under Sec. 50.10 or
Sec. 53.1130 of this chapter''.
0
130. In Sec. 51.105, revise paragraph (c)(1) introductory text to read
as follows:
Sec. 51.105 Public hearings in proceedings for issuance of
construction permits or early site permits; limited work
authorizations.
* * * * *
(c)(1) In addition to complying with the applicable provisions of
Sec. 51.104, in any proceeding for the issuance of a construction
permit for a nuclear power plant or an early site permit under part 52
or part 53 of this chapter, where the applicant requests a limited work
authorization under Sec. 50.10(d) or Sec. 53.1130 of this chapter,
the presiding officer will--
* * * * *
0
131. In Sec. 51.107, revise paragraphs (a) introductory text, (b)
introductory text, and (d)(1) introductory text to read as follows:
Sec. 51.107 Public hearings in proceedings for issuance of combined
licenses; limited work authorizations.
(a) In addition to complying with the applicable requirements of
Sec. 51.104, in a proceeding for the issuance of a combined license
for a nuclear power reactor under part 52 or part 53 of this chapter,
the presiding officer will:
* * * * *
(b) If a combined license application references an early site
permit, then the presiding officer in the combined license hearing must
not admit any contention proffered by any party on environmental issues
that have been accorded finality under Sec. 52.39 or Sec. 53.1188 of
this chapter, unless the contention:
* * * * *
(d)(1) In any proceeding for the issuance of a combined license
where the applicant requests a limited work authorization under Sec.
50.10(d) or Sec. 53.1130(a) of this chapter, the presiding officer, in
addition to complying with any applicable provision of Sec. 51.104,
will:
* * * * *
0
132. Revise Sec. 51.108 to read as follows:
Sec. 51.108 Public hearings on Commission findings that inspections,
tests, analyses, and acceptance criteria of combined licenses are met.
In any public hearing requested under Sec. 52.103(b) or Sec.
53.1452(b) of this chapter, the Commission will not admit any
contentions on environmental issues, the adequacy of the environmental
impact statement for the combined license issued under subpart C of
part 52 or under subpart H of part 53 of this chapter, or the adequacy
of any other environmental impact statement or environmental assessment
referenced in the combined license application. The Commission will not
make any environmental findings in connection with the finding under
Sec. 52.103(g) or Sec. 53.1452(g) of this chapter.
0
133. Add part 53, consisting of Sec. Sec. 53.000 through 53.9010, to
read as follows:
PART 53--RISK-INFORMED, TECHNOLOGY-INCLUSIVE REGULATORY FRAMEWORK
FOR COMMERCIAL NUCLEAR PLANTS
Sec.
53.000 Purpose.
Subpart A--General Provisions
53.015 Scope.
53.020 Definitions.
53.030 [Reserved]
53.040 Written communications.
53.050 Deliberate misconduct.
53.060 Employee protection.
53.070 Completeness and accuracy of information.
53.080 Specific exemptions.
53.090 Standards for review.
[[Page 15795]]
53.100 Jurisdictional limits.
53.110 Attacks and destructive acts.
53.115 Rights related to special nuclear material.
53.117 License suspension and rights of recapture.
53.120 Information collection requirements: OMB approval.
Subpart B--Technology-Inclusive Safety Requirements
53.210 Safety criteria for design-basis accidents.
53.220 Safety criteria for licensing-basis events other than design-
basis accidents.
53.230 Safety functions.
53.240 Licensing-basis events.
53.250 Defense in depth.
53.260 Normal operations.
53.270 Protection of plant workers.
Subpart C--Design and Analysis Requirements
53.400 Design features for licensing-basis events.
53.410 Functional design criteria for design-basis accidents.
53.415 Protection against external hazards.
53.420 Functional design criteria for licensing-basis events other
than design-basis accidents.
53.425 Design features and functional design criteria for normal
operations.
53.430 Design features and functional design criteria for protection
of plant workers.
53.440 Design requirements.
53.450 Analysis requirements.
53.460 Safety categorization and treatments.
53.470 [Reserved]
53.480 Earthquake engineering.
Subpart D--Siting Requirements
53.500 General siting and siting assessment.
53.510 External hazards.
53.520 Site characteristics.
53.530 Population-related considerations.
53.540 Siting interfaces.
Subpart E--Construction and Manufacturing Requirements
53.600 Construction and manufacturing--scope and purpose.
53.605 Reporting of defects and noncompliance.
53.610 Construction.
53.620 Manufacturing.
Subpart F--Requirements for Operation
53.700 Operational objectives.
53.710 Maintaining capabilities and availability of structures,
systems, and components.
53.715 Maintenance, repair, and inspection programs.
53.720 Response to seismic events.
53.725 General staffing, training, personnel qualifications, and
human factors requirements.
53.726 Communications.
53.728 Completeness and accuracy of information.
53.730 Defining, fulfilling, and maintaining the role of personnel
in ensuring safe operations.
53.735 General exemptions.
53.740 Facility licensee requirements--general.
53.745 Operator license requirements.
53.760 Operator licensing.
53.765 Medical requirements.
53.770 Incapacitation because of disability or illness.
53.775 Applications for operators and senior operators.
53.780 Training, examination, and proficiency program.
53.785 Conditions of operator and senior operator licenses.
53.790 Issuance, modification, and revocation of operator and senior
operator licenses.
53.795 Expiration and renewal of operator and senior operator
licenses.
53.800 Facility licensees for self-reliant-mitigation facilities.
53.805 Facility licensee requirements related to generally licensed
reactor operators.
53.810 Generally licensed reactor operators.
53.815 Generally licensed reactor operator training, examination,
and proficiency programs.
53.820 Cessation of individual applicability.
53.830 Training and qualification of commercial nuclear personnel.
53.845 Programs.
53.850 Radiation protection.
53.855 Emergency preparedness.
53.860 Security programs.
53.865 Quality assurance.
53.870 Integrity assessment programs.
53.875 Fire protection.
53.880 Inservice inspection and inservice testing.
53.910 Procedures and guidelines.
Subpart G--Decommissioning Requirements
53.1000 Scope and purpose.
53.1010 Financial assurance for decommissioning.
53.1020 Cost estimates for decommissioning.
53.1030 Annual adjustments to cost estimates for decommissioning.
53.1040 Methods for providing financial assurance for
decommissioning.
53.1045 Limitations on the use of decommissioning trust funds.
53.1050 NRC oversight.
53.1060 Reporting and recordkeeping requirements.
53.1070 Termination of license.
53.1075 Program requirements during decommissioning.
53.1080 Release of part of a commercial nuclear plant or site for
unrestricted use.
Subpart H--Licenses, Certifications, and Approvals
53.1100 Filing of application for licenses, certifications, or
approvals; oath or affirmation.
53.1101 Requirement for license.
53.1103 Combining applications and licenses.
53.1106 Elimination of repetition.
53.1109 Contents of applications; general information.
53.1112 Environmental conditions.
53.1115 Agreement limiting access to classified information.
53.1118 Ineligibility of certain applicants.
53.1120 Exceptions and exemptions from licensing requirements.
53.1121 Public inspection of applications.
53.1124 Relationship between sections.
53.1130 Limited work authorizations.
53.1140 Early site permits.
53.1144 Contents of applications for early site permits; general
information.
53.1146 Contents of applications for early site permits; technical
information.
53.1149 Review of applications.
53.1155 Referral to the Advisory Committee on Reactor Safeguards.
53.1158 Issuance of early site permit.
53.1161 Extent of activities permitted.
53.1164 Duration of permit.
53.1167 Limited work authorization after issuance of early site
permit.
53.1170 Transfer of early site permit.
53.1173 Application for renewal.
53.1176 Criteria for renewal.
53.1179 Duration of renewal.
53.1182 Use of site for other purposes.
53.1188 Finality of early site permit determinations.
53.1200 Standard design approvals.
53.1206 Contents of applications for standard design approvals;
general information.
53.1209 Contents of applications for standard design approvals;
technical information.
53.1210 Contents of applications for standard design approvals;
other application content.
53.1212 Standards for review of applications.
53.1215 Referral to the Advisory Committee on Reactor Safeguards.
53.1218 Staff approval of design.
53.1221 Finality of standard design approvals; information requests.
53.1230 Standard design certifications.
53.1236 Contents of applications for standard design certifications;
general information.
53.1239 Contents of applications for standard design certifications;
technical information.
53.1241 Contents of applications for standard design certifications;
other application content.
53.1242 Review of applications.
53.1245 Referral to the Advisory Committee on Reactor Safeguards.
53.1248 Issuance of standard design certification.
53.1251 Duration of certification.
53.1254 Application for renewal.
53.1257 Criteria for renewal.
53.1260 Duration of renewal.
53.1263 Finality of standard design certifications.
53.1270 Manufacturing licenses.
53.1276 Contents of applications for manufacturing licenses; general
information.
53.1279 Contents of applications for manufacturing licenses;
technical information.
53.1282 Contents of applications for manufacturing licenses; other
application content.
[[Page 15796]]
53.1285 Review of applications.
53.1286 Referral to the Advisory Committee on Reactor Safeguards.
53.1287 Issuance of manufacturing licenses.
53.1288 Finality of manufacturing licenses.
53.1291 Duration of manufacturing licenses.
53.1293 Transfer of manufacturing licenses.
53.1295 Renewal of manufacturing licenses.
53.1300 Construction permits.
53.1306 Contents of applications for construction permits; general
information.
53.1309 Contents of applications for construction permits; technical
information.
53.1312 Contents of applications for construction permits; other
application content.
53.1315 Review of applications.
53.1318 Finality of referenced NRC approvals, permits, and
certifications.
53.1324 Referral to the Advisory Committee on Reactor Safeguards.
53.1327 Authorization to conduct limited work authorization
activities.
53.1330 Exemptions, departures, and variances.
53.1333 Issuance of construction permits.
53.1336 Finality of construction permits.
53.1342 Duration of construction permits.
53.1345 Transfer of construction permits.
53.1348 Termination of construction permits.
53.1360 Operating licenses.
53.1366 Contents of applications for operating licenses; general
information.
53.1369 Contents of applications for operating licenses; technical
information.
53.1372 Contents of applications for operating licenses; other
application content.
53.1375 Review of applications.
53.1381 Referral to the Advisory Committee on Reactor Safeguards.
53.1384 Exemptions, departures, and variances.
53.1387 Issuance of operating licenses.
53.1390 Backfitting of operating licenses.
53.1396 Duration of operating licenses.
53.1399 Transfer of an operating license.
53.1402 Application for renewal.
53.1405 Continuation of an operating license.
53.1410 Combined licenses.
53.1413 Contents of applications for combined licenses; general
information.
53.1416 Contents of applications for combined licenses; technical
information.
53.1419 Contents of applications for combined licenses; other
application content.
53.1422 Review of applications.
53.1425 Finality of referenced NRC approvals.
53.1431 Referral to the Advisory Committee on Reactor Safeguards.
53.1434 Authorization to conduct limited work authorization
activities.
53.1437 Exemptions, departures, and variances.
53.1440 Issuance of combined licenses.
53.1443 Finality of combined licenses.
53.1449 Inspection during construction.
53.1452 Operation under a combined license.
53.1455 Duration of combined license.
53.1456 Transfer of a combined license.
53.1458 Application for renewal.
53.1461 Continuation of combined license.
53.1470 Standardization of commercial nuclear plant designs:
licenses to construct and operate nuclear power reactors of
identical design at multiple sites.
Subpart I--Maintaining and Revising Licensing-Basis Information
53.1500 Licensing-basis information.
53.1502 Specific terms and conditions of licenses.
53.1505 Changes to licensing-basis information requiring prior NRC
approval.
53.1510 Application for amendment of license.
53.1515 Public notices; State consultation.
53.1520 Issuance of amendment.
53.1525 Revising certification information within a design
certification rule.
53.1530 Revising design information within a Final Safety Analysis
Report associated with a manufacturing license.
53.1535 Amendments during construction.
53.1540 Updating licensing-basis information and determining the
need for NRC approval.
53.1545 Updating Final Safety Analysis Reports.
53.1550 Evaluating changes to facility as described in Final Safety
Analysis Reports.
53.1560 Updating program documents included in licensing-basis
information.
53.1565 Evaluating changes to programs included in licensing-basis
information.
53.1570 Transfer of licenses.
53.1575 Termination of licenses.
53.1580 Information requests.
53.1585 Revocation, suspension, modification of licenses and
approvals for cause.
53.1590 Backfitting.
53.1595 Renewal.
Subpart J--Reporting and Other Administrative Requirements
53.1600 General information.
53.1610 Unfettered access for inspections.
53.1620 Maintenance of records, making of reports.
53.1630 Immediate notification requirements for operating commercial
nuclear plants.
53.1640 Licensee event report system.
53.1645 Reports of radiation exposure to members of the public.
53.1650 Facility information and verification.
53.1660 Financial requirements.
53.1670 Financial qualifications.
53.1680 [Reserved]
53.1690 Licensee's change of status; financial qualifications.
53.1700 Creditor regulations.
53.1710 Financial protection.
53.1720 Insurance required to stabilize and decontaminate plant
following an accident.
53.1730 Financial protection requirements.
Subparts K and L [Reserved]
Subpart M--Enforcement
53.9000 Violations.
53.9010 Criminal penalties.
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. 115-439, 132
Stat. 5571.
Sec. 53.000 Purpose.
This part provides an optional, technology-inclusive, performance-
based framework for the issuance, amendment, renewal, and termination
of licenses, permits, certifications, and approvals for commercial
nuclear plants licensed under section 103 of the Atomic Energy Act of
1954, as amended (the Act) (68 Stat. 919), and Title II of the Energy
Reorganization Act of 1974, as amended (88 Stat. 1242). Also, this part
gives notice to all persons who knowingly provide to any holder of or
applicant for an approval, certification, permit, or license, or to a
contractor, subcontractor, or consultant of any of them, components,
equipment, materials, or other goods or services that relate to the
activities of a holder of or applicant for an approval, certification,
permit, or license, subject to this part, that they may be individually
subject to U.S. Nuclear Regulatory Commission enforcement action for
violation of the provisions in Sec. 53.050.
Subpart A--General Provisions
Sec. 53.015 Scope.
Subpart A provides general provisions applicable to all applicants
and licensees subject to the rules of this part.
Sec. 53.020 Definitions.
As used in this part:
Anticipated event sequence means event sequences expected to occur
one or more times during the life of a commercial nuclear plant.
Anticipated event sequences take into account the expected response of
all structures, systems, and components (SSCs) within the plant,
regardless of safety classification.
Applicant means a person applying for a license, permit, or other
form of Commission permission or approval under this part.
Certified fuel handler means, for a commercial nuclear plant,
either--
[[Page 15797]]
(1) A non-licensed operator who has qualified in accordance with a
fuel handler training program approved by the Commission; or
(2) A non-licensed operator who demonstrates compliance with the
following criteria:
(i) 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. 53.830,
and
(ii) Is responsible for decisions on--
(A) Safe conduct of decommissioning activities,
(B) Safe handling and storage of spent fuel; and
(C) Appropriate response to plant emergencies.
Combined license (COL) means a combined construction permit (CP)
and operating license (OL) with conditions for a commercial nuclear
plant issued under this part.
Commercial nuclear plant means a facility consisting of one or more
commercial nuclear reactors and associated co-located support
facilities, including the collection of buildings, radionuclide
sources, and SSCs for which a license, certification, or approval is
being sought under this part, that is or will be used for producing
power for commercial electric power or other commercial purposes. For
the purposes of requirements in this part that reference requirements
in part 50 of this chapter, a commercial nuclear plant is equivalent to
a nuclear power plant.
Commercial nuclear reactor means an apparatus, other than an atomic
weapon, designed or used to sustain nuclear fission. For the purposes
of requirements in this part that reference requirements in 10 CFR part
50, a commercial nuclear reactor is equivalent to a nuclear reactor as
defined in Sec. 50.2 of this chapter.
Commission means the U.S. Nuclear Regulatory Commission (NRC) or
its duly authorized representatives.
Construction means the activities in paragraph (1) of this
definition and does not mean the activities in paragraph (2) of this
defintion.
(1) Activities constituting construction are those activities that
are conducted on-site to build the commercial nuclear plant, including
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--
(i) Safety-related (SR) SSCs and those non-safety-related but
safety-significant (NSRSS) SSCs of a facility for which special
treatment includes requirements on design or installation, including
associated quality assurance measures;
(ii) SSCs necessary to comply with 10 CFR part 73; or
(iii) Onsite emergency facilities necessary to comply with Sec.
53.855.
(2) Construction does not include--
(i) Changes for temporary use of the land for public recreational
purposes;
(ii) Site exploration, including necessary borings to determine
foundation conditions or other preconstruction monitoring to establish
background information related to the suitability of the site, the
environmental impacts of construction or operation, or the protection
of environmental values;
(iii) Preparation of a site for construction of a facility,
including clearing of the site, grading, installation of drainage,
erosion, and other environmental mitigation measures, and construction
of temporary roads and borrow areas;
(iv) Erection of fences and other access control measures;
(v) Excavation;
(vi) Erection of support buildings (such as construction equipment
storage sheds, warehouse and shop facilities, utilities, concrete
mixing plants, docking and unloading facilities, and office buildings)
for use in connection with the construction of the facility;
(vii) Building of service facilities (such as paved roads, parking
lots, railroad spurs, exterior utility and lighting systems, potable
water systems, sanitary sewage treatment facilities, and transmission
lines);
(viii) Procurement or fabrication of components or portions of the
proposed facility occurring at locations other than the final, in-place
location at the facility; or
(ix) Manufacture of a nuclear power reactor under a manufacturing
license (ML) under subpart H of this part to be installed at the
proposed site and to be part of the proposed facility.
Custom combined license (custom COL) means a COL that does not
reference a standard design approval, standard design certification, or
manufacturing license.
Decommission or decommissioning means to remove a plant or 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.
Design-basis accidents (DBAs) means postulated event sequences that
are used to set functional design criteria and performance objectives
for the design of SR SSCs through deterministic analyses. Design-basis
accidents are a type of licensing-basis event and are based on the
capabilities and reliabilities of SR SSCs needed to mitigate and
prevent event sequences, respectively.
Design-basis external hazard level means the level of severity or
intensity of an external hazard for which the SR SSCs are protected
against or designed to withstand without losing their capability to
perform their safety functions.
Design features means the active and passive SSCs and the inherent
characteristics of those SSCs that contribute to limiting the total
effective dose equivalent to individual members of the public during
normal operations and prevent or mitigate the consequences of event
sequences.
Early site permit (ESP) means a Commission approval, issued under
subpart H of this part, for a site for one or more commercial nuclear
plants. An early site permit is a partial construction permit.
Electric utility means any entity that generates or distributes
electricity and that 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.''
Event sequence means a postulated initiating event defined for a
set of initial plant conditions followed by system, safety function,
and operator successes or failures, and terminating in a specified end
state depending on the system, safety function, and operator successes
and failures (e.g., prevention of release of radioactive material or
release in one of the reactor-specific release categories). An event
sequence may include many unique variations of events that are similar
in terms of results or end states.
[[Page 15798]]
Exclusion area means that area surrounding the reactor, in which
the reactor licensee has the authority to determine all activities
including exclusion or removal of personnel and property from the area.
This area may be traversed by a highway, railroad, or waterway,
provided these are not so close to the facility as to interfere with
normal operations of the facility and provided appropriate and
effective arrangements are made to control traffic on the highway,
railroad, or waterway, in case of emergency, to protect the public
health and safety. Residence within the exclusion area must normally be
prohibited. In any event, residents must be subject to ready removal in
case of necessity. Activities unrelated to operation of the reactor may
be permitted in an exclusion area under appropriate limitations,
provided that no significant hazards to the public health and safety
will result.
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 commercial nuclear plant.
Functional design criteria means metrics for the performance of
SSCs. For SR SSCs, these criteria define performance metrics necessary
to demonstrate compliance with the safety criteria in Sec. 53.210. For
NSRSS SSCs, these criteria define performance metrics necessary to
demonstrate compliance with the safety criteria in Sec. 53.220.
License, when used in the context of a facility, means a limited
work authorization, CP, OL, early site permit, COL, or ML under this
part, or a renewed license issued by the Commission under this part.
When used in the context of a license authorizing an individual to
manipulate the controls of a facility, license means a license issued
by the Commission to perform the function of an operator, senior
operator, or generally licensed reactor operator as defined in this
part.
Licensee means a person who is authorized to conduct activities
under a license issued under this part by the Commission.
Licensing-basis events means a collection of event sequences
considered in the design and licensing of the commercial nuclear plant.
Licensing-basis events are unplanned events and include anticipated
event sequences, unlikely event sequences, very unlikely event
sequences, and DBAs.
Licensing-basis information means the information contained in
regulations, orders, licenses, certifications, or approvals issued by
the NRC for a commercial nuclear plant licensed under this part 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 this part.
Low-population zone means the area immediately surrounding the
exclusion area which contains residents, the total number and density
of which are such that there is a reasonable probability that
appropriate protective measures could be taken on their behalf in the
event of a serious accident. A permissible population density or total
population within this zone is not included in this definition because
the situation may vary from case to case. Whether a specific number of
people can, for example, be evacuated from a specific area or
instructed to take shelter on a timely basis, will depend on many
factors such as location, number and size of highways, scope and extent
of advance planning, and actual distribution of residents within the
area.
Major decommissioning activity means, for a commercial nuclear
plant, any activity that results in permanent removal of major
radioactive components, permanently modifies the structure of the
containment, if applicable, or results in dismantling components for
shipment containing greater than class C waste in accordance with Sec.
61.55 of this chapter.
Major feature of the emergency plans means an aspect of those plans
necessary to:
(1) Address in whole or part either one or more of the 16 standards
in 10 CFR 50.47(b) or the requirements of 10 CFR 50.160(b), as
applicable; or
(2) Describe the emergency planning zones as required in 10 CFR
53.1109(g).
Manufactured reactor means the essential portions of a nuclear
reactor that are manufactured under an ML and subsequently transported
and incorporated into a commercial nuclear plant under a COL or CP.
Manufacturing license means a license issued under this part that
authorizes the manufacture of manufactured reactors but not its
construction, installation, or operation.
Non-Safety-Related but Safety-Significant (NSRSS) SSCs means those
SSCs which are not SR but are relied on to achieve adequate defense in
depth or perform risk-significant functions and warrant special
treatment.
Non-Safety-Significant SSCs means those SSCs that are not SR or
NSRSS, are not relied on to achieve adequate defense in depth or to
perform risk-significant functions, and do not warrant special
treatment.
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 of Energy, except that the
Department of Energy shall 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.
Population center distance means the distance from the reactor to
the nearest boundary of a densely populated center containing more than
about 25,000 residents.
Programmatic controls means administrative measures that govern
human action in implementing programs and operating, monitoring, and
maintaining SSCs and equipment of a commercial nuclear plant.
Programmatic controls considered to be licensing basis information are
addressed by programs under Sec. 53.845 and are specified in an
application for a requested activity of the Commission.
Quality assurance (QA) means all those planned and systematic
actions necessary to ensure that a structure, system, or component will
perform satisfactorily in service. Quality assurance includes quality
control, which comprises those QA actions related to the physical
characteristics of a material, structure, component, or system which
provide a means to control the quality of the material, structure,
component, or system to predetermined requirements.
Safety criteria means performance-based metrics that establish a
level of safety provided in requirements in Sec. Sec. 53.210 and
53.220.
Safety-related structures, systems, or components means those SSCs
that are relied upon to demonstrate compliance with the safety criteria
in Sec. 53.210 and warrant special treatment.
[[Page 15799]]
Small modular reactor means a power reactor, which may be of
modular design as defined in Sec. 52.1 of this chapter, licensed under
this part to produce heat energy up to 1,000 megawatts thermal per
module.
Site characteristics means the actual physical, environmental, and
demographic features of a site. Site characteristics are specified in
an early site permit or in a Preliminary or Final Safety Analysis
Report for a limited work authorization, CP, or COL, as applicable.
Site parameters are the postulated physical, environmental, and
demographic features of an assumed site. Site parameters are specified
in a standard design approval, standard design certification, or ML.
Source material means source material as defined in subsection 11z.
of the Atomic Energy Act of 1954, as amended, (the Act) and in the
regulations contained in part 40 of this chapter.
Special nuclear material (SNM) means:
(1) Plutonium, uranium-233, uranium enriched in the isotope-233 or
in the isotope-235, and any other material which the Commission,
pursuant to the provisions of section 51 of the Act, 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.
Special treatment means those requirements, such as QA, design
criteria, and programmatic controls, that are taken beyond the
procurement, installation, and maintenance of commercial grade products
to ensure that SR and NSRSS SSCs will provide defense in depth or
perform risk-significant functions. The requirements also ensure that
the SSCs will perform under the service conditions and with the
reliability assumed in the analysis performed under Sec. 53.450 to
demonstrate compliance with the safety criteria in Sec. Sec. 53.210
for SR SSCs and 53.220 for SR and NSRSS SSCs.
Standard design means a design which is sufficiently detailed and
complete to support certification or approval in accordance with
subpart H of this part, and which is usable under of this part for a
multiple number of units or at a multiple number of sites without
reopening or repeating the review.
Standard design approval or design approval means an NRC staff
approval, issued under subpart H of this part, of a final standard
design for a commercial nuclear plant. The approval may be for either
the final design for the entire reactor facility or the final design of
major portions thereof.
Standard design certification or design certification means a
Commission approval, issued under subpart H of this part, of a final
standard design for a nuclear power facility. This design may be
referred to as a certified standard design.
Total effective dose equivalent means the sum of the effective dose
equivalent (for external exposures) and the committed effective dose
equivalent (for internal exposures).
Utilization facility means any commercial nuclear reactor other
than one designed or used primarily for the formation of plutonium or
uranium-233.
Unlikely event sequences means event sequences that are not
expected to occur in the life of a commercial nuclear plant and are
less likely than anticipated event sequences, but are infrequent rather
than rare. Unlikely event sequences take into account the expected
response of all SSCs within the plant regardless of safety
classification.
Very unlikely event sequences means event sequences that are not
expected to occur in the life of a commercial nuclear plant, are less
likely than an unlikely event sequence, and are rare. Very unlikely
event sequences take into account the expected response of all SSCs
within the plant regardless of safety classification.
Sec. 53.030 [Reserved]
Sec. 53.040 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 or individual license conditions, or the terms and conditions
of an early site permit or standard design approval, must be submitted
to the persons listed below (addresses for the NRC Regional Offices are
listed in appendix D to 10 CFR part 20).
(1) Applications for amendment of 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, permits, and 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 permits and licenses, and amendments to
applications. Applications for licenses, 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
[[Page 15800]]
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) Change to security plan, guard training and qualification
plan, safeguards contingency plan, or cybersecurity plan made without
prior Commission approval under Sec. 53.1565; and
(v) Application for amendment of physical security plan, guard
training and qualification plan, safeguards contingency plan, or
cybersecurity plan under Sec. 53.1510.
(5) Emergency plan and related submissions. Written communications
as defined in paragraphs (b)(5)(i) through (iii) 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;
(ii) Change to an emergency plan under Sec. 53.1565; and
(iii) Emergency implementing procedures under Sec. 53.855.
(6) Updated Final Safety Analysis Report. An updated Final Safety
Analysis Report or replacement pages under Sec. 53.1545 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. Paper copy
submissions may be made using replacement pages; however, if a licensee
chooses to use electronic submission, all subsequent updates or
submissions must be performed electronically on a total replacement
basis. 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.
(7) Quality assurance related submissions. (i) A change to the
Safety Analysis Report QA program description under Sec. 53.1565, or a
change to a licensee's NRC-accepted QA topical report under Sec.
53.1565, 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.
(ii) A change to an NRC-accepted QA topical report from non-
licensees (i.e., architect/engineers, nuclear steam supply system
suppliers, fuel suppliers, constructors, etc.) must be submitted to the
NRC's Document Control Desk. If the communication is on paper, the
signed original must be sent.
(8) 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.
(9) 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.
(d) Regulation governing submission. Licensees, applicants, and
holders of standard design approvals 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. 53.050 Deliberate misconduct.
(a) Any licensee or applicant for a license; holder of or applicant
for a standard design approval; applicant for a standard design
certification; employee of a licensee, holder of a standard design
approval, or applicant for a license, standard design approval, or
standard design certification; or any contractor (including a supplier
or consultant), subcontractor, employee of a contractor or
subcontractor of any licensee or applicant for a license, holder of or
applicant for a standard design approval, or applicant for a standard
design certification, 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. 53.060 Employee protection.
(a) Discrimination by a Commission licensee, holder of a standard
design approval, an applicant for a license, standard design
certification, or standard design approval, a contractor or
subcontractor of a Commission licensee, holder of a standard design
approval, applicant for a license,
[[Page 15801]]
standard design certification, 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
Act 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 has no application 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 Act.
(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, standard design certification, or a
standard design approval, or a contractor or subcontractor of a
Commission licensee, 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) Withdrawal or revocation of a proposed or final standard design
certification;
(3) Imposition of a civil penalty on the licensee, holder of a
standard design approval, or applicant (including an applicant for a
standard design certification under this part following Commission
adoption of final design certification rule) or a contractor or
subcontractor of the licensee, holder of a standard design approval, or
applicant; or
(4) 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)(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.
(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, standard design certification, 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. 53.070 Completeness and accuracy of information.
(a) Information provided to the Commission by a holder of a
license, permit, design certification, or standard design approval
under this part or an applicant for a license, permit, design
certification, 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, each holder of a standard design
approval under this part, and each applicant for a standard design
certification under this part following Commission adoption of a final
design certification regulation, 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 (b) 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
[[Page 15802]]
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. 53.080 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
section, 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 activities prior to the issuance of a CP. 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 must not be deemed to constitute
a commitment to issue a CP. 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 must be governed by the exemption
requirements of those parts.
Sec. 53.090 Standards for review.
(a) Common standards. In determining that a CP, OL, early site
permit, COL, or ML under this part will be issued to an applicant, the
Commission will be guided by the following considerations:
(1) Except for an early site permit or ML, 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 10 CFR part
20, and that the health and safety of the public will not be
endangered.
(2) The applicant for a CP, OL, COL, or ML 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 OL for a utilization facility of the type
described in paragraph (d) of this section or for an applicant for an
ML.
(3) The issuance of a CP, OL, early site permit, COL, or ML 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.
(4) Any applicable requirements of 10 CFR part 51 have been
satisfied.
(b) Additional standards for licenses. In determining whether a
license will be issued to an applicant, the Commission will, in
addition to applying the standards set forth in paragraph (a) of this
section, consider whether the proposed activities will serve a useful
purpose proportionate to the quantities of SNM or source material to be
utilized.
(c) Additional standards and provisions affecting licenses for
commercial power. In addition to applying the standards set forth in
paragraphs (a) and (b) of this section, paragraphs (c)(1) through
(c)(4) of this section apply in the case of a license 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 or production
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, must be
subject to the regulatory provisions of the Federal Power Act.
(4) Nothing will preclude any government agency, now or hereafter
authorized by law to engage in the production, marketing, or
distribution of electric energy, if otherwise qualified,
[[Page 15803]]
from obtaining a CP, OL, or COL under this part for a utilization
facility for the primary purpose of producing electric energy for
disposition for ultimate public consumption.
(d) Licenses for commercial nuclear plants. A license will be
issued, to an applicant who qualifies, for any one or more of the
following: to transfer or receive in interstate commerce, or
manufacture, produce, transfer, acquire, possess, or use a utilization
facility for industrial or commercial purposes.
Sec. 53.100 Jurisdictional limits.
No permit, license, standard design approval, or standard design
certification under this part shall be deemed to have been issued for
activities that are not under or within the jurisdiction of the United
States.
Sec. 53.110 Attacks and destructive acts.
Licensees, applicants for licenses, permits, certifications, and
design approvals, and applicants for an amendment to any license,
permit, certification, 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. 53.115 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 Act and gives its consent in writing.
Sec. 53.117 License suspension and rights of recapture.
Any license issued under this part must 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 Act in a
state of war or national emergency declared by Congress.
Sec. 53.120 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
3150-0274.
(b) The approved information collection requirements contained in
this part appear in Sec. Sec. 53.070, 53.080, 53.240, 53.410, 53.420,
53.425, 53.430, 53.440, 53.450, 53.480, 53.500, 53.540, 53.605, 53.610,
53.620, 53.700, 53.710, 53.715, 53.720, 53.730, 53.780, 53.785, 53.805,
53.810, 53.815, 53.830, 53.850, 53.855, 53.865, 53.870, 53.875, 53.880,
53.910, 53.1010, 53.1020, 53.1030, 53.1045, 53.1060, 53.1070, 53.1075,
53.1080, 53.1100, 53.1109, 53.1115, 53.1130, 53.1140, 53.1144, 53.1146,
53.1173, 53. 1182, 53.1188, 53.1200, 53.1206, 53.1209, 53.1210,
53.1221, 53.1230, 53.1236, 53.1239, 53.1241, 53.1254, 53.1257, 53,1263,
53.1270, 53.1276, 53.1279, 53.1282, 53.1288, 53.1295, 53.1300, 53.1306,
53.1309, 53.1312, 53.1327, 53.1330, 53.1333, 53.1336, 53.1348, 53.1360,
53.1366, 53.1369, 53.1372, 53.1384, 53.1410, 53.1413, 53.1416, 53.1419,
53.1437, 53.1449, 53.1452, 53.1458, 53.1470, 53.1505, 53.1510, 53.1515,
53.1525, 53.1530, 53.1535, 53.1540, 53.1545, 53.1550, 53.1560, 53.1565,
53.1570, 53.1575, 53.1580, 53.1620, 53.1630, 53.1645, 53.1690, 53.1720.
(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. 53.765, 53.770, 53.780, and 53.795, NRC Form 396
is approved under control number 3150-0024.
(2) In Sec. Sec. 53.775 and 53.795, NRC Form 398 is approved under
control number 3150-0090.
(3) In Sec. 53.1640, NRC Form 366 is approved under control number
3150-0104.
(4) In Sec. 53.1630, NRC Form 361S is approved under control
number 3150-0238.
(5) In Sec. 53.1650, International Atomic Energy Agency Design
Information Questionnaire forms are approved under control number 3150-
0056.
(6) In Sec. 53.1650, DOC/NRC Form AP-A and associated forms are
approved under control numbers 0694-0135.
Subpart B--Technology-Inclusive Safety Requirements
Sec. 53.210 Safety criteria for design-basis accidents.
Design features and programmatic controls must be provided for each
commercial nuclear plant such that identification and analyses of
design-basis accidents (DBAs) in accordance with Sec. 53.240
demonstrate the following:
(a) An individual located at any point on the boundary of the
exclusion area for any 2-hour period following the onset of the
postulated fission product release would not receive a radiation dose
in excess of 25 rem (250 millisieverts) total effective dose equivalent
(TEDE); and
(b) An individual located at any point on the outer boundary of the
low-population zone who is exposed to the radioactive cloud resulting
from the postulated fission product release (during the entire period
of its passage) would not receive a radiation dose in excess of 25 rem
(250 millisieverts) TEDE.\1\
\1\ The use of 25 rem TEDE is not intended to imply that this
number constitutes an acceptable limit for an emergency dose to the
public under accident conditions. Rather, this dose value has been
set forth in this section as a reference value, which can be used in
the evaluation of plant design features with respect to postulated
reactor accidents, to assure that these designs provide assurance of
low risk of public exposure to radiation, in the event of an
accident.
Sec. 53.220 Safety criteria for licensing-basis events other than
design-basis accidents.
Design features and programmatic controls must be provided for each
commercial nuclear plant such that identification and analysis of
licensing-basis events (LBEs) other than DBAs in accordance with Sec.
53.240 demonstrate the following:
(a) Plant structures, systems, and components (SSCs), personnel,
and programs provide the necessary capabilities and maintain the
necessary reliability to address LBEs other than DBAs in accordance
with Sec. Sec. 53.240 and 53.450(e), and provide measures for defense
in depth in accordance with Sec. 53.250; and
(b) The analysis of risks to public health and safety resulting
from LBEs other than DBAs under Sec. 53.450(e) includes comprehensive
risk metrics that satisfy associated risk performance objectives that
are acceptable to the U.S.
[[Page 15804]]
Nuclear Regulatory Commission (NRC) and provide an appropriate level of
safety.
Sec. 53.230 Safety functions.
(a) The primary safety function is limiting the release of
radioactive materials from the facility and must be maintained during
normal operation and for LBEs over the life of the plant.
(b) Additional safety functions needed to support the retention of
radioactive materials during LBEs--such as controlling reactivity, heat
generation, heat removal, and chemical interactions--must be identified
for each commercial nuclear plant.
(c) The primary and additional safety functions are required to
satisfy the safety criteria defined in Sec. Sec. 53.210 and 53.220 and
must be fulfilled by the design features, human actions, and
programmatic controls specified throughout this part.
Sec. 53.240 Licensing-basis events.
(a) Licensing-basis events must be identified for each commercial
nuclear plant and analyzed under Sec. 53.450 to demonstrate that the
safety requirements in this subpart have been satisfied.
(b) The identified LBEs, ranging from anticipated event sequences
to very unlikely event sequences, must collectively address appropriate
risk-informed combinations of malfunctions of plant SSCs, human errors,
facility hazards, and the effects of external hazards.
(c) The analysis of LBEs must--
(1) Include analysis of one or more DBAs under Sec. 53.450(f);
(2) Confirm the adequacy of design features and programmatic
controls needed to satisfy the safety criteria defined in Sec. Sec.
53.210 and 53.220, and
(3) Establish related functional requirements for plant SSCs,
personnel, and programs.
Sec. 53.250 Defense in depth.
(a) Measures must be taken for each commercial nuclear plant to
ensure appropriate defense in depth is provided to compensate for
uncertainties in the analysis of the safety criteria such that there is
reasonable assurance that the safety criteria in this subpart are met
over the life of the plant.
(b) The uncertainties that must be addressed under paragraph (a) of
this section include those related to the state of knowledge and
modeling capabilities, the ability of barriers to limit the release of
radioactive materials from the facility during LBEs other than DBAs,
the reliability and performance of plant SSCs and personnel, and the
effectiveness of programmatic controls.
(c) The safety analysis may not exclusively rely upon a single
engineered design feature, human action, or programmatic control, no
matter how robust, to address the range of LBEs other than DBAs.
Sec. 53.260 Normal operations.
Holders of licenses to operate commercial nuclear plants under this
part must control public doses and dose rates in unrestricted areas
from normal plant operations to meet the requirements in 10 CFR part
20.
Sec. 53.270 Protection of plant workers.
Holders of licenses to operate commercial nuclear plants under this
part must control occupational doses to meet the requirements in 10 CFR
part 20.
Subpart C--Design and Analysis Requirements
Sec. 53.400 Design features for licensing-basis events.
(a) Design features must be provided for each commercial nuclear
plant such that, when combined with corresponding human actions and
programmatic controls, the plant will satisfy the safety criteria
defined in Sec. Sec. 53.210 and 53.220.
(b) Design features must ensure that the safety functions
identified in Sec. 53.230 are fulfilled during licensing-basis events
(LBEs).
Sec. 53.410 Functional design criteria for design-basis accidents.
(a) Functional design criteria must be defined for each design
feature classified as safety-related (SR) in terms of its role in
demonstrating compliance with the safety criteria defined in Sec.
53.210.
(b) The identification of special treatments associated with the
design of SR structures, systems, and components (SSCs) must consider
human actions and programmatic controls identified and implemented in
accordance with this and other subparts to achieve and maintain the
reliability and capability of SSCs relied upon to satisfy the defined
functional design criteria and the safety criteria required in Sec.
53.210, and to maintain consistency with analyses required by Sec.
53.450(f).
Sec. 53.415 Protection against external hazards.
Safety-related SSCs must be protected against or must be designed
to withstand the effects of natural phenomena (e.g., earthquakes,
tornadoes, hurricanes, floods, tsunami, and seiches) and constructed
hazards (e.g., dams, transportation routes, military and industrial
facilities) considering an event severity up to the design-basis
external hazard levels as determined under Sec. 53.510 without losing
the capability to perform the safety functions identified under Sec.
53.230. Specific requirements for earthquake engineering are included
in Sec. 53.480.
Sec. 53.420 Functional design criteria for licensing-basis events
other than design-basis accidents.
(a) Functional design criteria must be defined for each design
feature classified as SR or non-safety-related but safety-significant
(NSRSS) in terms of its role in demonstrating compliance with--
(1) The safety criteria in Sec. 53.220; and
(2) The evaluation criteria in Sec. 53.450(e).
(b) The identification of special treatments associated with the
design of SR and NSRSS SSCs must consider human actions and
programmatic controls identified and implemented in accordance with
this and other subparts to achieve and maintain the reliability and
capability of SSCs relied upon to satisfy--
(1) The safety criteria in Sec. 53.220; and
(2) The evaluation criteria in Sec. 53.450(e).
Sec. 53.425 Design features and functional design criteria for
normal operations.
(a) Design features must be provided for each commercial nuclear
plant to support the Radiation Protection Program required in Sec.
53.850.
(b) Functional design criteria must be defined for each design
feature relied upon to demonstrate compliance with Sec. 53.850.
(c) Functional design criteria, including design objectives for
dose to the maximally exposed member of the public, must be defined for
design features to show that plant design features and corresponding
programmatic controls, including monitoring programs, control liquid,
gaseous, and solid wastes, as required under part 20 of this chapter.
Sec. 53.430 Design features and functional design criteria for
protection of plant workers.
(a) Design features must be provided for each commercial nuclear
plant such that, when combined with corresponding programmatic
controls, the requirements in Sec. 53.270 can be met.
(b) Functional design criteria must be defined for each design
feature relied upon to demonstrate compliance with Sec. 53.270.
[[Page 15805]]
Sec. 53.440 Design requirements.
(a)(1) Analysis, appropriate test programs, prototype testing,
operating experience, or a combination thereof must demonstrate that
each design feature required by Sec. 53.400 meets the defined
functional design criteria required by Sec. Sec. 53.410 and 53.420.
This demonstration must consider interdependent effects throughout the
commercial nuclear plant and the range of conditions under which the
design features required by Sec. 53.400 must function throughout the
plant's lifetime.
(2) The design processes for SR and NSRSS SSCs under this part must
include administrative procedures for evaluating operating, design, and
construction experience and for considering applicable important
industry experiences in the design of those SSCs.
(b) The design features classified as SR must, wherever applicable,
be designed using generally accepted consensus codes and standards that
have been endorsed or otherwise found acceptable by the U.S. Nuclear
Regulatory Commission (NRC).
(c) The materials used for each SR and NSRSS SSC must be qualified
for their service conditions over the design life of the SSC as
appropriate to satisfy the special treatments established for the SSC
under Sec. 53.460.
(d) Possible degradation mechanisms related to aging, fatigue,
chemical interactions, operating temperatures, effects of irradiation,
and other environmental factors that may affect the performance of SR
and NSRSS SSCs must be evaluated and used to inform the design and the
development of integrity assessment programs under Sec. 53.870.
(e)(1) Safety-related SSCs and, where appropriate, NSRSS SSCs must
be designed and located to minimize, consistent with other safety
requirements in this part, the probability and effect of fires and
explosions.
(2) Noncombustible and fire-resistant materials must be used
wherever practical throughout the facility, particularly in locations
with SR and NSRSS SSCs.
(3) Fire detection and fire suppression systems of appropriate
capacity and capability must be provided and designed to minimize the
adverse effects of fires on SR and NSRSS SSCs.
(4) Fire suppression systems must be designed to ensure that their
rupture or inadvertent operation does not significantly impair the
ability of SR and NSRSS SSCs to perform their safety functions to
satisfy Sec. 53.230.
(f) 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.
(g) The reactor system and waste stores for each commercial nuclear
plant must be capable of achieving and maintaining a subcritical
condition during normal operations and following any LBE identified in
accordance with Sec. 53.240.
(h) Each commercial nuclear plant must have a capability to provide
long-term cooling of the reactor fuel and waste stores during normal
operations and following any LBE identified in accordance with Sec.
53.240.
(i) The design, analysis, staffing, and programmatic controls for
each commercial nuclear plant must consider the number of reactors,
waste stores, and other significant inventories of radioactive
materials and the associated operating configurations, common systems,
system interfaces, and system interactions.
(j) [Reserved]
(k) Design features, related functional design criteria,
programmatic controls, or a combination thereof must be defined such
that analyses demonstrate a low risk of permanent injury to the public
due to the health effects of the chemical hazards of licensed material.
(l) Measures must be taken during the design of commercial nuclear
plants to minimize, to the extent practicable, contamination of the
facility and the environment, facilitate eventual decommissioning, and
minimize, to the extent practicable, the generation of radioactive
waste in accordance with Sec. 20.1406 of this chapter.
(m)(1) Each commercial nuclear plant must include criticality
monitoring capabilities meeting the requirements of either Sec. 70.24
of this chapter or paragraph (m)(2) of this section.
(2) In lieu of maintaining a monitoring system capable of detecting
criticality as described in Sec. 70.24 of this chapter, criticality
accident requirements may be satisfied by--
(i) Demonstrating the sub-criticality of special nuclear material,
except when it is inside the reactor and the reactor is being operated,
by maintaining k-effective below 0.95 at a 95 percent probability, 95
percent confidence level, under conditions that maximize reactivity for
the applicable storage and handling configurations, and
(ii) Providing radiation monitors for fuel storage and associated
handling areas when fuel is present to detect excessive radiation
levels and to support initiating appropriate safety actions.
(3) While a spent fuel transportation package approved under 10 CFR
part 71 of this chapter or spent fuel storage cask approved under 10
CFR part 72 is in the special nuclear material handing or storage area,
the requirements in 10 CFR parts 71 or 72, as applicable, and the
requirements of the certificate of compliance for that package or cask,
are the applicable requirements for the fuel within that package or
cask.
(n)(1) The design of each commercial nuclear plant must reflect
state-of-the-art human factors principles for safe and reliable
performance in all locations that human activities are expected for
performing or supporting the continued availability of plant safety or
emergency response functions.
(2) The design must provide for the capabilities described in Sec.
53.730(b) to ensure the plant staff are able to monitor plant
conditions and respond to events.
(3) The means by which the design and human actions together will
achieve the safety requirements of subpart B of this part must be
evaluated and used to inform the design and the development of the
concept of operations required by Sec. 53.730(c).
(4) A functional requirements analysis and function allocation must
be used to ensure that plant design features address how safety
functions and functional safety criteria are satisfied, and how the
safety functions will be assigned to appropriate combinations of human
action, automation, active safety features, passive safety features, or
inherent safety characteristics.
Sec. 53.450 Analysis requirements.
(a) Requirement to have a probabilistic risk assessment (PRA), or
other systematic risk evaluations (SREs), or a combination thereof. A
PRA, other SREs, or a combination thereof for each commercial nuclear
plant must be performed and used together with other generally accepted
approaches for systematically evaluating engineered systems to identify
potential failures, susceptibility to internal and external hazards,
and other contributing factors to event sequences that might challenge
the safety functions identified in Sec. 53.230 and to support
demonstrating that each commercial nuclear plant meets the safety
criteria of Sec. 53.220.
(b) Specific uses of analyses. The PRA, other SREs, or a
combination thereof, together with other generally accepted approaches
for systematically evaluating engineered systems must be used to--
(1) Inform the selection of the LBEs, as described in Sec. 53.240,
which must be
[[Page 15806]]
considered in the design to determine compliance with the safety
criteria in subpart B of this part.
(2) Inform the classification of SSCs according to their safety
significance in accordance with Sec. 53.460 and to identify the
environmental conditions under which the SSCs and operating staff must
perform their safety functions.
(3) Evaluate the adequacy of defense-in-depth measures required in
accordance with Sec. 53.250.
(4) Identify and assess all plant operating states where there is
the potential for the uncontrolled release of radioactive material to
the environment.
(5) Identify and assess events that challenge plant control and
safety systems whose failure could lead to the uncontrolled release of
radioactive material to the environment. These include internal events,
such as human errors and equipment failures, and external events
identified in accordance with subpart D of this part.
(6) Inform the establishment and updating of appropriate measures
for plant operations, including availability controls, to ensure that
the configurations and special treatments for SR SSCs and NSRSS SSCs
provide the capabilities, availability, and reliability consistent with
satisfying the safety criteria under Sec. Sec. 53.220 and the analyses
of licensing-basis events other than design-basis accidents (DBAs)
under Sec. 53.450(e).
(c) Maintenance and upgrade of analyses. The PRA, other SREs, or a
combination thereof must be maintained (e.g., updated to reflect plant
changes such as modifications, procedure changes, or plant performance
data) at least every 5 years until the permanent cessation of
operations under Sec. 53.1070 and upgraded (e.g., changed in scope or
use of new methods) in conformance with generally accepted methods,
standards, and practices that have been endorsed or otherwise found
acceptable by the NRC.
(d) Qualification of analytical codes. The analytical codes used in
modeling the physical behavior of plant systems in the analyses of
licensing-basis events (including but not limited to thermodynamics,
reactor physics, fuel performance, and mechanistic source term codes)
must be qualified for the range of conditions for which they are to be
used.
(e) Analyses of licensing-basis events other than design-basis
accidents. (1) Analyses must be performed for LBEs other than design-
basis accidents (DBAs). These LBEs must be identified using insights
from a PRA, other SREs, or a combination thereof with other generally
accepted approaches for systematically evaluating engineered systems to
identify and analyze equipment failures and human errors.
(2) The analysis of LBEs other than DBAs must include definitions
of evaluation criteria for each event or specific categories of LBEs to
determine the acceptability of the plant response to the challenges
posed by internal and external hazards to provide an appropriate level
of safety.
(3) The analyses of LBEs other than DBAs must address event
sequences from initiation to a defined end state and be used in
combination with other engineering analyses to demonstrate that the
functional design criteria required by Sec. 53.420 provide sufficient
barriers to the unplanned release of radionuclides to satisfy the
evaluation criteria defined for each LBE other than DBAs, to satisfy
the safety criteria specified in accordance with Sec. 53.220 and
provide defense in depth as required by Sec. 53.250.
(4) The methodology used to identify, categorize, and analyze LBEs
must include a means to identify event sequences deemed significant for
controlling the risks posed to public health and safety.
(f) Analysis of design-basis accidents. (1) The analysis of LBEs
required by Sec. 53.240 must include analysis of DBAs that address
possible challenges to the safety functions identified under Sec.
53.230. The events selected as DBAs must be those that, if not
terminated, have the potential for exceeding the safety criteria in
Sec. 53.210.
(2) The DBAs selected must be analyzed using deterministic methods
that address event sequences from initiation to a safe stable end state
and assume only the SR SSCs identified under Sec. 53.460 and human
actions addressed by the requirements of subpart F of this part are
available to perform the safety functions identified in accordance with
Sec. 53.230.
(3) The analysis must conservatively demonstrate compliance with
the safety criteria in Sec. 53.210.
(g) Other required analyses. Analyses must be performed to assess--
(1) Fire protection. Fire protection measures to demonstrate,
through inclusion of fires in the analysis of LBEs or by separate
analyses, that a fire or explosion in any plant area would not--
(i) Prevent equipment from fulfilling the safety functions
identified in accordance with Sec. 53.230; or
(ii) Challenge the safety criteria in Sec. Sec. 53.210 and 53.220.
(2) [Reserved]
(3) Dose to members of the public. Measures taken under Sec.
53.425, including estimating--
(i) The quantity of each of the principal radionuclides expected to
be released annually to unrestricted areas in liquid effluents produced
during normal reactor operations and the dose to the maximally exposed
member of the public in unrestricted areas.
(ii) The quantities 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 the dose to the maximally exposed member of the public
in unrestricted areas.
(iii) The annual external radiation dose in unrestricted areas and
the maximally exposed member of the public in unrestricted areas due to
direct radiation from contained radiation sources from the commercial
nuclear plant during normal reactor operations.
Sec. 53.460 Safety categorization and special treatments.
(a) Structures, systems, and components must be classified
according to their safety significance. The SSC categories must include
``Safety-Related,'' ``Non-Safety-Related but Safety-Significant,'' and
``Non-Safety-Significant,'' as defined in subpart A of this part.
(b) For SR and NSRSS SSCs, the conditions under which they must
perform their safety function in Sec. 53.230 must be identified.
Special treatments must be established in accordance with this and
other subparts to provide confidence that the SSCs will perform under
the service conditions and with reliability consistent with the
analysis performed under Sec. 53.450 to demonstrate meeting the safety
criteria in Sec. Sec. 53.210 and 53.220.
(1) The special treatments for SR SSCs must include meeting the
applicable quality assurance requirements from appendix B of part 50 of
this chapter.
(2) The special treatments for NSRSS SSCs and special treatments
for SR SSCs beyond those required under paragraph (b)(1) of this
section may include meeting selected quality assurance requirements
from appendix B of part 50 of this chapter when such treatment is
needed to address performance requirements, equipment reliability, or
uncertainties.
(c) The identification of special treatments for SR and NSRSS SSCs
must account for human actions needed to prevent or mitigate LBEs, the
need to perform such actions reliably under the postulated
environmental conditions, and the role of programs established in
accordance with subpart F of this part to provide confidence that those
actions
[[Page 15807]]
will be performed as assumed in the analysis performed in accordance
with Sec. 53.450 to demonstrate meeting the applicable criteria in
Sec. Sec. 53.210, 53.220, and 53.450(e).
Sec. 53.470 [Reserved]
Sec. 53.480 Earthquake engineering.
(a) Effects of earthquakes. Structures, systems, and components
classified as SR or NSRSS must be able to withstand the effects of
earthquakes, commensurate with the safety significance of the SSC,
without loss of capability to perform their role in fulfilling the
safety functions required by Sec. 53.230.
(b) Definitions. As used in this section--
Design-Basis Ground Motions (DBGMs) are the vibratory ground
motions for which certain SSCs must be designed to remain functional.
Operating basis earthquake (OBE) ground motion is the vibratory
ground motion for which those features of the commercial nuclear plant
necessary for continued operation without undue risk to the health and
safety of the public are designed to remain functional. The OBE ground
motion is used in Sec. 53.720.
Response spectrum is a plot of the maximum responses (acceleration,
velocity, or displacement) of idealized single-degree-of-freedom
oscillators as a function of the natural frequencies of the oscillators
for a given damping value. The response spectrum is calculated for a
specified vibratory motion input at the oscillators' supports.
Surface deformation is the distortion of geologic strata on or near
the ground surface that occurs because of tectonic forces that result
from earthquakes.
(c) Design considerations--(1) Design-Basis Ground Motions. (i) The
DBGMs must be derived from the Site Ground Motion Response Spectra
developed in accordance with Sec. 53.510(c), by taking into
consideration the functional design criteria of SSCs in accordance with
Sec. Sec. 53.410 and 53.420. The horizontal component of the DBGM(s)
in the free-field at the foundation level of the structures must be an
appropriate response spectrum that is determined based on the risk-
significance of SSCs and their safety functions. In view of the limited
data available on vibratory ground motion of strong earthquakes, it is
acceptable that the design response spectra be smoothed spectra.
(ii) The commercial nuclear plant must be designed so that, if the
DBGMs occur, the following SSCs remain functional and within applicable
stress, strain, and deformation limits:
(A) Structures, systems, and components for which functional design
criteria are established in accordance with Sec. 53.410 or Sec.
53.420; and
(B) Structures, systems, and components classified as SR or NSRSS
commensurate with safety significance in accordance with Sec. 53.460.
(iii) In addition to seismic loads, applicable concurrent normal
operating, functional, and accident-induced loads must be taken into
account in the design of the SR SSCs and, commensurate with safety
significance, NSRSS SSCs.
(iv) The design of the commercial nuclear plant must take into
account the possible effects of seismic-induced ground disruption, such
as fissuring, lateral spreads, differential settlement, liquefaction,
and landsliding, on the facility foundations.
(v) The SSCs fulfilling the safety functions required by Sec.
53.230 must be demonstrated through design, testing, or qualification
methods to be able to fulfill those safety functions during and after
the vibratory ground motion associated with the DBGMs.
(vi) The evaluation of SSCs required by this section to show they
are able to function during and after earthquake ground motion should
consider, if applicable, soil-structure interaction effects and the
expected duration of vibratory motion. It is permissible to design for
inelastic behavior in some of these SSCs during the DBGMs and under the
postulated concurrent loads, provided the necessary safety functions
are maintained.
(2) OBE Ground Motion. The OBE Ground Motion must be characterized
by response spectra. The value of the OBE Ground Motion must be set to
one-third or less of the DBGMs response spectra.
(3) [Reserved]
(4) Required seismic instrumentation. Suitable instrumentation must
be provided so that the seismic response of commercial nuclear plant SR
SSCs or NSRSS SSCs can be evaluated promptly after an earthquake.
(d) Surface deformation. (1) The potential for surface deformation
must be taken into account in the design of the commercial nuclear
plant by providing reasonable assurance that in the event of
deformation, SSCs classified as SR or NSRSS in accordance with Sec.
53.460 will remain functional.
(2) In addition to surface deformation induced loads, the design of
SSCs must take into account, commensurate with safety significance,
seismic loads and applicable concurrent functional and accident-induced
loads.
(3) The design provisions for surface deformation must be based on
its postulated occurrence in any direction and azimuth and under any
part of the commercial nuclear plant, unless evidence indicates this
assumption is not appropriate, and must take into account the estimated
rate at which the surface deformation may occur.
(e) Seismically induced floods and water waves and other design
conditions. Seismically induced floods and water waves from either
locally or distantly generated seismic activity and other design
conditions determined pursuant to subpart D of this part must be taken
into account in the design of the commercial nuclear plant so as to
prevent undue risk to the health and safety of the public.
(f) Analysis. The analyses required by Sec. 53.450 must address
seismic hazards and related SSC responses in determining that the
safety criteria defined in Sec. 53.220 will be met.
(g) Design criteria, human actions, and programmatic controls.
Functional design criteria, human actions, and programmatic controls
needed to address seismic events must be identified and implemented in
accordance with this and other subparts to achieve and maintain the
performance of SSCs relied upon to satisfy the safety criteria in Sec.
53.220 and to maintain consistency with analyses required by Sec.
53.450 when accounting for the site-specific frequencies and magnitudes
of earthquakes for a commercial nuclear plant.
Subpart D--Siting Requirements
Sec. 53.500 General siting and siting assessment.
The purpose of this subpart and the specific requirements therein
is to ensure that:
(a) The siting of each commercial nuclear plant is supported by
assessments of proposed sites such that the design, including design
features and programmatic controls corresponding to the site
characteristics, satisfies the safety criteria defined in Sec. Sec.
53.210 and 53.220. The siting assessment addresses the site
characteristics that might contribute to the initiation, progression,
or consequences of licensing-basis events (LBEs) analyzed under
Sec. Sec. 53.450 and 53.480 that are identified and mitigated by
design features or programmatic controls. The siting assessment takes
into consideration the potential adverse impacts that a commercial
nuclear plant may have on nearby populations as a result of normal
operations or LBEs.
(b) Activities performed to identify site characteristics or
otherwise needed to determine site-specific contributors to
[[Page 15808]]
functional design criteria or analysis assumptions under subpart C of
this part satisfy the applicable special treatment requirements of
Sec. 53.460, including, where applicable, the quality assurance
requirements from appendix B of part 50 of this chapter.
Sec. 53.510 External hazards.
(a) General external hazard requirements. The design-basis external
hazard level for the relevant external hazards for a site must be
identified and characterized based on site-specific assessments of
natural and constructed hazards with the potential to adversely affect
plant functions. The external hazard frequencies and magnitudes
determined from the site-specific assessments must take into account
uncertainties and variabilities in data, models, and methods relied on
to characterize the external hazards.
(b) Definitions. As used in this section, the following terms mean:
Geological Siting Factors are geological and seismic factors that
may affect the design and operation of the proposed commercial nuclear
plant.
Ground Motion Response Spectra (GMRS) are the site-specific GMRS
resulting from the geologic investigations and evaluations of the site
vicinity and region and used to determine design-basis ground motions
for structures, systems, and components under Sec. 53.480.
Probabilistic Seismic Hazard Analysis is an analytical methodology
that incorporates uncertainty into estimates of an annual frequency of
exceedance for a certain ground motion parameter (e.g., peak ground
acceleration, peak ground velocity, response spectral values) at a
site.
(c) Geological investigations. The GMRS for the site must be
determined based on the results of investigations of the geological,
seismological, and engineering characteristics of the site and its
environs and must be characterized by both horizontal and vertical
free-field GMRS at the free ground surface. The size of the region to
be investigated and the type of data pertinent to the investigations
must be determined based on the nature of the region surrounding the
site. Data on vibratory ground motion, earthquake recurrence rates,
fault geometry and slip rates, and site subsurface material properties
must be obtained by reviewing pertinent literature and carrying out
field investigations. Uncertainties are inherent in the parameters and
models used to estimate the GMRS for the site. The site assessment must
reflect these uncertainties through an appropriate analysis, such as a
probabilistic seismic hazard analysis.
(d) Geologic and seismic siting factors. The geologic and seismic
siting factors considered for design under Sec. Sec. 53.415 and 53.480
must include, but are not limited to, determination of the potential
for surface tectonic and nontectonic deformations, the size and
character of seismically induced floods and water waves that could
affect a site from either locally or distantly generated seismic
activity, soil and rock stability, liquefaction potential, and natural
and artificial slope stability.
Sec. 53.520 Site characteristics.
Site characteristics that might contribute to the initiation,
progression, or consequences of LBEs analyzed under Sec. 53.450 must
be identified, assessed, and considered in the design and analyses
required by subpart C of this part.
Sec. 53.530 Population-related considerations.
Every site must have an exclusion area, a low-population zone, and
a population center distance as defined in Sec. 53.020.
(a) The offsite radiological consequences estimated by the analyses
required by Sec. 53.450(f) must be used to confirm that--
(1) An individual located at any point on the boundary of the
exclusion area for any 2-hour period following onset of the postulated
fission product release would not receive a radiation dose in excess of
25 rem (250 millisieverts) total effective dose equivalent.
(2) An individual located at any point on the outer boundary of the
low-population zone who is exposed to the radioactive cloud resulting
from the postulated fission product release (during the entire period
of its passage) would not receive a radiation dose in excess of 25 rem
(250 millisieverts) total effective dose equivalent.
(b) The reactor site must either:
(1) Provide a population center distance of at least one and one-
third times the distance from the reactor to the outer boundary of the
low-population zone; or
(2) Be found acceptable to the U.S. Nuclear Regulatory Commission
(NRC) based on assessments of societal risks in comparison to societal
benefits for the specific site. The boundary of the population center
or the alternate area assessed considering societal risks and benefits
must be determined upon consideration of population distribution.
Political boundaries are not controlling in the calculation of
population center distance or the alternate area assessed considering
societal risks and benefits.
(c) Reactor sites should be located away from very densely
populated centers or otherwise be shown to be acceptable by assessments
of societal risks in comparison to societal benefits for the specific
site. Areas of low-population density are, generally, preferred.
However, in determining the acceptability of a particular site located
away from a very densely populated center but not in an area of low-
population density or when assessing a site considering societal risks
and benefits, consideration will be given to safety, environmental,
economic, or other factors, which may result in the site being found
acceptable.
Sec. 53.540 Siting interfaces.
Site characteristics must be addressed by the design features,
programmatic controls, and supporting analyses used to demonstrate that
the safety criteria in Sec. Sec. 53.210 and 53.220 are met for each
commercial nuclear plant. Site characteristics must be such that
adequate emergency plans and security plans can be developed and
maintained.
Subpart E--Construction and Manufacturing Requirements
Sec. 53.600 Construction and manufacturing--scope and purpose.
This subpart applies to those construction and manufacturing
activities authorized by a construction permit (CP), combined license
(COL), manufacturing license (ML), or limited work authorization (LWA)
issued under this part.
Sec. 53.605 Reporting of defects and noncompliance.
Each CP and ML issued under this part is subject to the terms and
conditions in this section, and each COL issued under this part is
subject to the terms and conditions in this section until the date that
the Commission makes the finding under Sec. 53.1452(g).
(a) Definitions. The definitions in Sec. 21.3 of this chapter
apply to this section.
(b) Posting requirements. (1) Each individual, partnership,
corporation, dedicating entity, or other entity subject to the
regulations in this section must post current copies of this section
and the regulations in 10 CFR part 21; section 206 of the Energy
Reorganization Act of 1974, as amended; and procedures adopted under
these regulations. These documents must be posted in a conspicuous
position on any premises within the United States where the
[[Page 15809]]
activities subject to the license are conducted.
(2) If posting of these regulations or the procedures adopted under
them is not practical, the licensee may, in addition to posting section
206 of the Energy Reorganization Act of 1974, as amended, post a notice
that describes the regulations/procedures, including the name of the
individual to whom reports may be made, and states where they may be
examined.
(c) Procedures. The holder of a CP, COL, or ML subject to this
section 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 (c)(2) of
this section, in all cases within 60 days of discovery, 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 the 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 paragraph (d)(5) of this section.
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 of the holder of
a CP, COL, or ML subject to this section is informed as soon as
practicable, and, in all cases, within the 5 working days after
completion of the evaluation described in paragraph (c)(1) or (c)(2) of
this section, if the construction or manufacture of a facility or
activity, or a basic component supplied for such a facility or
activity--
(i) Fails to comply with the Atomic Energy Act of 1954, as amended,
or any applicable regulation, order, or license of the Commission
relating to a substantial safety hazard;
(ii) Contains a defect; or
(iii) Underwent any significant breakdown in any portion of the
quality assurance program (QAP) conducted under the requirements of
appendix B to part 50 of this chapter that could have produced a defect
in a basic component. These breakdowns in the QAP are reportable
whether or not the breakdown actually resulted in a defect in a design
approved and released for construction, installation, or manufacture.
(d) Reporting defects and noncompliance. (1) The holder of a CP,
COL, or ML subject to this section that obtains information reasonably
indicating that the facility or manufactured reactors fails to comply
with the Atomic Energy Act of 1954, as amended, or any applicable
regulation, order, or license of the Commission relating to a
substantial safety hazard must notify the Commission of the failure to
comply through a director, responsible officer, or designated person as
discussed in paragraph (d)(5) of this section.
(2) The holder of a CP, COL, or ML subject to this section that
obtains information reasonably indicating the existence of any defect
found in the construction or manufacture, or any defect found in the
final design of a facility as approved and released for construction or
manufacture, must notify the Commission of the defect through a
director, responsible officer, or designated person as discussed in
paragraph (d)(5) of this section.
(3) The holder of a CP, COL, or ML subject to this part, who
obtains information reasonably indicating that the QAP has undergone
any significant breakdown discussed in paragraph (c)(3)(iii) of this
section must notify the Commission of the breakdown in the QAP through
a director, responsible officer, or designated person as discussed in
paragraph (d)(5) of this section.
(4) When acting as a dedicating entity, the holder of a CP, COL, or
ML subject to this section is responsible for identifying and
evaluating deviations; reporting defects and failures to comply
associated with substantial safety hazards for dedicated items; and
maintaining auditable records for the dedication process.
(5) The notification requirements of this paragraph (d) apply to
all defects and failures to comply associated with a substantial safety
hazard regardless of whether extensive evaluation, redesign, or repair
is required to conform to the criteria and bases stated in the Safety
Analysis Report, CP, COL, or ML. Evaluation of potential defects and
failures to comply and reporting of defects and failures to comply
under this section satisfies the CP holder's, COL holder's, and ML
holder's evaluation and notification obligations under 10 CFR part 21,
and satisfies the responsibility of individual directors or responsible
officers or holders of a CP, COL, or ML subject to this section to
report defects, and failures to comply associated with substantial
safety hazards under section 206 of the Energy Reorganization Act of
1974, as amended. The director or responsible officer may authorize an
individual to provide the notification required by this section.
However, this does not relieve the director or responsible officer of
his or her responsibility under this section.
(e) Notification--timing and where sent. The notification required
by paragraph (d) of this section must consist of--
(1) Initial notification by telephone, facsimile, or email
identified in appendix A to 10 CFR part 73 to the U.S. Nuclear
Regulatory Commission (NRC) Operations Center within 2 days following
receipt of information by the director or responsible corporate officer
under paragraph (c)(3) of this section, on the identification of a
defect or a failure to comply. If the CP, COL, or ML holder elects to
use facsimile, verification that the facsimile has been received should
be made by calling the NRC Operations Center. This paragraph (e)(1)
does not apply to interim reports described in paragraph (c)(2) of this
section.
(2) Written notification submitted to the NRC Document Control Desk
by an appropriate method listed in Sec. 53.040, with a copy to the
appropriate NRC Regional Administrator at the address specified in
appendix D to 10 CFR part 20 and a copy to the appropriate NRC resident
inspector, if applicable, within 30 days following receipt of
information by the director or responsible corporate officer under
paragraph (c)(3) of this section, on the identification of a defect or
failure to comply.
(f) Content of notification. The written notification required by
paragraph (e)(2) of this section must clearly indicate that the written
notification is being submitted under this section and include the
following information, to the extent known.
(1) Name and address of the individual or individuals informing the
Commission.
(2) Identification of the facility, the activity, or the basic
component supplied for the facility or the activity within the United
States which contains a defect or fails to comply.
(3) Identification of the firm constructing or manufacturing the
facility or supplying the basic component which fails to comply or
contains a defect.
(4) Nature of the defect or failure to comply and the safety hazard
which is created or could be created by the defect or failure to
comply.
[[Page 15810]]
(5) The date on which the information of a defect or failure to
comply was obtained.
(6) In the case of a basic component that contains a defect or
failure to comply, the number and location of these components in use
at the facility subject to the regulations in this part.
(7) In the case of a completed reactor manufactured under this
part, the entities to which the reactor was supplied.
(8) The corrective action which 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.
(9) 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 other entities.
(g) Procurement documents. Each holder of a CP, COL, or ML subject
to this section must ensure that each procurement document for a
facility or a basic component specifies the provisions of 10 CFR part
21 or this section that apply, as applicable.
(h) Coordination with 10 CFR part 21. The requirements of this
section are satisfied when the defect or failure to comply associated
with a substantial safety hazard has been previously reported under 10
CFR part 21, under Sec. 73.1205 of this chapter, under this section,
or under Sec. 53.1640.
(i) Records retention. The holder of a CP, COL, or ML subject to
this section must prepare and maintain records necessary to accomplish
the purposes of this section, specifically--
(1) Retain procurement documents, which define the requirements
that facilities or basic components must satisfy 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 under paragraph (c)(1) of this section for the longest of--
(i) Ten years from the date of the evaluation;
(ii) Five years from the date that an early site permit is
referenced in an application for a COL; or
(iii) Five years from the date of delivery of a manufactured
reactor.
(3) Retain records of all interim reports to the Commission made
under paragraph (c)(2) of this section, or notifications to the
Commission made under paragraph (d) of this section for the minimum
time periods stated in paragraph (i)(2) of this section;
(4) Suppliers of basic components must retain records of--
(i) All notifications sent to affected licensees or purchasers
under paragraph (d)(4) of this section for a minimum of 10 years
following the date of the notification;
(ii) The facilities or other purchasers to whom the basic
components or associated services were supplied for a minimum of 15
years from the delivery of the basic component or associated services.
(5) Maintaining reports in accordance with this section satisfies
the recordkeeping obligations under 10 CFR part 21 of the entities,
including directors or responsible officers thereof, subject to this
section.
Sec. 53.610 Construction.
(a) Management and control. Licensees must ensure that the
following plans, programs, and organizational units are developed and
implemented to manage and control the construction activities:
(1) Programs to ensure that the construction of a commercial
nuclear plant supports the eventual compliance with the design and
analysis requirements in subpart C of this part.
(2) An organization, headed by qualified personnel, responsible for
managing, controlling, and evaluating the adequacy of the construction
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) Procedures to evaluate the applicability of other national and
international construction experience to the planned and ongoing
construction activities and to ensure the applicable experience will be
provided to those constructing the plant.
(5) A fitness-for-duty program, under 10 CFR part 26.
(6)(i) A QAP meeting the requirements of appendix B of part 50 of
this chapter as required by Sec. 53.460(b).
(ii) Appropriate programmatic controls to provide special treatment
for non-safety-related but safety-significant structures, systems, and
components (SSCs).
(7) A radiation protection program, in accordance with 10 CFR part
20, that includes measures for monitoring the dose to individuals
working with radioactive materials brought onto the site, as
applicable.
(8) An information security program in accordance with Sec. Sec.
73.21, 73.22, and 73.23 of this chapter, as applicable.
(b) Construction activities. No person may begin the construction
of a commercial nuclear plant on a site on which the facility is to be
operated under this part until that person has been issued either a CP
or COL, an early site permit authorizing activities under Sec.
53.1130, or an LWA under this part.
(1) Licensees must satisfy the following requirements:
(i) As appropriate, considering the types and quantities of
radioactive materials being brought onto the site--
(A) The licensee must maintain and follow a special nuclear
material (SNM) material control and accounting program, a measurement
control program, and other material control procedures that include
corresponding record management requirements as required by the
provisions of Sec. 70.32 of this chapter. Prior to initial receipt of
SNM onsite, the licensee must implement an SNM material control and
accounting program in accordance with 10 CFR part 74.
(B) Procedures must be in place to receive, possess, use, and store
source, byproduct, and SNM in accordance with applicable portions of 10
CFR parts 30, 40, and 70.
(C) A plant staff training program associated with the receipt of
radioactive material must be approved and implemented prior to initial
receipt of byproduct, source or SNM (excluding exempt quantities as
described in Sec. 30.18 of this chapter).
(ii) For construction of a commercial nuclear plant involving
multiple reactor units, plans and procedures must be in place to
prevent or mitigate potential hazards to the SSCs of operating units
resulting from construction activities, including the managerial and
administrative controls to be used to provide assurance that the
limiting conditions for operation of the operating units are not
exceeded as a result of construction activities.
(iii) Procedures must be in place prior to the start of
construction activities that describe how construction will be
controlled so as not to impact other features important to the design,
such as dewatering, slope stability, backfill, compaction, and seepage.
(iv) For LWA holders, a plan must be developed for redress of
activities performed under the LWA should one of the following
situations arise:
(A) LWA work activities are terminated by the holder of the LWA;
(B) The LWA is revoked by the NRC; or
(C) The Commission denies the associated CP or COL application.
[[Page 15811]]
(2)(i) Onsite fresh fuel must be protected and stored in compliance
with Sec. 73.67 of this chapter.
(ii) Before initial fuel load into the reactor (or, for a fueled
manufactured reactor, before initiating the removal of the features to
prevent criticality required under Sec. 53.620(d)(1)), a cybersecurity
program that meets the requirements of Sec. 73.54 or Sec. 73.110 of
this chapter, a physical security program that meets the requirements
of Sec. 73.55 or Sec. 73.100 of this chapter, and an access
authorization program that meets the requirements of Sec. 73.56 or
Sec. 73.120 of this chapter must be established, as applicable.
(iii) Fire protection measures must be implemented for work and
storage areas (including adjacent fire areas that could affect the work
or storage area) before initial receipt of byproduct, source, or non-
fuel SNM (excluding exempt quantities as described in Sec. 30.18 of
this chapter). The fire protection measures for areas associated with
new fuel (including all fuel handling, fuel storage, and adjacent fire
areas that could affect the new fuel) must be implemented before
receipt of fuel. Prior to the receipt of fuel, a formal letter of
agreement must be in place with the local fire department specifying
the nature of arrangements in support of the fire protection program.
(c) Inspection and acceptance. (1) The licensee must have a process
for accepting individual or groups of SSCs upon completion of
construction and protecting them from damage or tampering as other
construction activities continue.
(2) The post-construction acceptance process must address the
inspections, tests, analyses, and acceptance criteria specified in the
COL under Sec. 53.1440 or the equivalent verifications needed to
support the issuance of an operating license under Sec. 53.1387.
Sec. 53.620 Manufacturing.
(a) Management and control. Holders of MLs 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 ML:
(1) Programs to ensure that the manufacturing of a manufactured
reactor or portions of a manufactured reactor complies with the design
and analysis requirements in subpart C of this part. The entity with
design authority for the manufactured reactor covered by the ML 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 program to evaluate the applicability of other national and
international design and manufacturing experience to the planned and
ongoing manufacturing activities.
(5) A fitness-for-duty program, in accordance with 10 CFR part 26.
(6)(i) A QAP meeting the requirements of appendix B to part 50 of
this chapter, to be applied to the design, fabrication, construction,
and testing of the SSCs of the manufactured reactor.
(ii) Appropriate programmatic controls to provide special treatment
measures for non-safety-related but safety-significant SSCs.
(7) 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.
(8) An information security program in accordance with Sec. Sec.
73.21, 73.22 and 73.23 of this chapter, as applicable.
(b) Manufacturing activities. Holders of MLs must satisfy the
following requirements:
(1) The manufacturing process must be conducted within facilities
for which the ML 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 ML.
(2) Manufacturing processes must be performed in accordance with
the ML and the referenced codes and standards that have been endorsed
or otherwise found acceptable by the NRC.
(3) A post-manufacturing inspection and acceptance process must be
established and implemented before transporting a manufactured reactor
or portions of a manufactured reactor for installation at a commercial
nuclear plant. The process must consider the results of inspections,
tests, and analyses that have been performed and the acceptance
criteria that are necessary and sufficient to conclude that
manufacturing activities have been completed in accordance with the ML.
(c) Control of radioactive materials. 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 SNM 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 SNM
(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 SNM (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 SNM (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)(i) An ML may authorize possession of a
manufactured reactor into which the licensee has loaded fresh
(unirradiated) fuel pursuant to a license issued under part 70 of this
chapter only if the manufactured reactor is configured during its
loading, storage, and transport with features to prevent criticality
that are specified in the ML.
(ii) The ML applicant may file a separate, subsequent application
for the 10 CFR part 70 license or combine the application for the 10
CFR part 70 license with the application for an ML.
(iii) The Commission has determined that any such fueled
manufactured reactor in which the features to prevent criticality are
in place is not in operation.
(iv) Upon installation of the fueled manufactured reactor in its
place of operation and a Commission finding that the acceptance
criteria in the COL that authorized reactor construction are met under
Sec. 53.1452(g), or that any conditions in the CP that authorized
reactor construction are met and the associated operating license (OL)
issued, the features to prevent criticality may be removed. Upon
initiating the removal of
[[Page 15812]]
the features to prevent criticality, the fueled manufactured reactor
has commenced operation.
(2) Holders of part 70 licenses authorizing the possession and
loading of fresh fuel into manufactured reactors must comply with the
requirements of part 70 for the facilities and activities related to
the storage, movement, and loading of fresh fuel in the manufactured
reactor. Holders of these part 70 licenses must comply with the
requirements of Subpart H to part 70, regardless of whether their
proposed activities meet the applicability criteria found in 10 CFR
70.60. 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.
(i) 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:
(A) 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.
(B) The physical security program must be designed to prevent
unintended and uncontrolled criticality events.
(C) The cybersecurity program must provide reasonable assurance
that a cyberattack does 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.
(D) All holders of a 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; these determinations
must be documented.
(ii) [Reserved]
(3) 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 meeting the requirements in
subpart F of this part.
(e) Transportation. (1) A holder of an ML may not transport or
allow to be removed from the places of manufacture the manufactured
reactor or portions thereof as defined in the ML except for either
transport to a site for which the Commission has issued a COL or CP
that references the subject ML or export in accordance with 10 CFR part
110.
(2) A holder of an ML must include in any contract governing the
transport of a manufactured reactor or portions thereof as defined in
the ML 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 ML 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 commercial nuclear plant.
(4) For a manufactured reactor that is to be loaded with fresh fuel
before transport to the place of operation, the ML 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 COL or CP that references the subject ML. (1)
Installation at the site for which the Commission has issued a COL or
CP that references the subject ML must follow the regulations in Sec.
53.610.
(2) Upon arrival at the site, the manufactured reactor or portions
of a manufactured reactor may not be installed in its place of
operation unless the COL or CP holder performs inspections sufficient
to verify the reactor is in compliance with the ML and has not been
damaged in transit. The COL or CP 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 commercial nuclear plant are met.
Subpart F--Requirements for Operation
Sec. 53.700 Operational objectives.
The purpose of this subpart and the specific requirements herein is
to ensure that:
(a) Each holder of an operating license (OL) or combined license
(COL) under this part develops, implements, and maintains controls for
plant structures, systems, and components (SSCs), responsibilities of
personnel, and plant programs during the operating life of each
commercial nuclear plant such that the requirements defined in subpart
B are satisfied. More specifically:
(1) Under Sec. 53.710 through Sec. 53.730, each holder of an OL
or COL under this part must maintain the capabilities, availability,
and reliability of plant SSCs to ensure that the safety functions
identified in Sec. 53.230 will be performed if called upon during
licensing-basis events (LBEs).
(2) Under Sec. 53.725 through Sec. 53.830, each holder of an OL
or COL under this part must ensure that personnel have adequate
knowledge and skills to perform their assigned duties that support the
performance of the safety functions identified in Sec. 53.230.
(3) Under Sec. 53.845 through Sec. 53.910, each holder of an OL
or COL under this part must implement plant programs sufficient to
ensure that the safety functions identified in Sec. 53.230 will be
performed if called upon during normal operations and LBEs.
(b) [Reserved]
Sec. 53.710 Maintaining capabilities and availability of structures,
systems, and components.
Measures must be provided for each commercial nuclear plant
licensed under this part such that the capabilities, availability, and
reliability of plant SSCs, when combined with corresponding
programmatic controls and human actions, provide that the safety
criteria defined in Sec. Sec. 53.210 and 53.220 will be met.
(a) Technical specifications must be developed, implemented, and
maintained that define conditions or limitations on plant operations
that are necessary to ensure that safety-related (SR) SSCs can fulfill
the safety functions identified under Sec. 53.230 and support meeting
the safety criteria of Sec. 53.210. The technical specifications must
describe the following requirements:
(1) Limits on the inventory of radioactive materials within the
reactor system and supporting systems with the potential, individually
or collectively, to cause a release exceeding the safety criteria in
Sec. 53.210 as a result of a design-basis accident analyzed in
accordance with Sec. 53.450(f).
(2) Operating limits for the facility that if exceeded could lead
to a failure to perform a required safety function necessary to
demonstrate compliance with the safety criteria in Sec. 53.210.
[[Page 15813]]
(3) For each SSC classified as SR in accordance with Sec. 53.460,
technical specifications must define--
(i) Limiting conditions for operation. Limiting conditions for
operation are the lowest functional capability or performance levels of
SR SSCs required to ensure that the design-basis accidents analyzed in
accordance with Sec. 53.450(f) satisfy the safety criteria of Sec.
53.210. When a limiting condition for operation is not met, the
licensee must shut down the plant or follow any remedial action
permitted by the technical specifications until the condition can be
met.
(ii) Surveillance requirements. Surveillance requirements are
requirements relating to test, calibration, or inspection to assure
that the necessary quality of systems and components is maintained and
that the limiting conditions for operation will be met.
(4) Design elements to be included are those elements of the plant
such as materials of construction and geometric arrangements, which, if
altered or modified, would have a significant effect on safety and are
not covered in categories described in paragraphs (a)(1) through (3) of
this section.
(5) Administrative controls are the provisions relating to
organization and management, procedures, recordkeeping, review and
audit, and reporting necessary to assure operation of the plant in a
safe manner. Each licensee must submit any reports to the Commission
pursuant to approved technical specifications under Sec. 53.040.
(b) Control measures on plant operations, including availability
controls, must be developed and implemented to ensure that the
configurations and special treatments for SR SSCs and non-safety-
related but safety-significant (NSRSS) SSCs provide the capabilities,
availability, and reliability required to demonstrate compliance with
the criteria of Sec. Sec. 53.220 and 53.450(e).\1\ The control
measures must--
(1)(i) Identify who within the licensee's organization has
authority to make configuration changes;
(ii) Establish processes to make configuration changes to NSRSS
SSCs; and
(iii) Establish processes to ensure that all organizations of the
commercial nuclear plant affected by the configuration changes are
formally notified and approve of the change.
(2) Describe how the special treatments for each NSRSS SSC and
special treatments for SR SSCs beyond those under paragraph (a) of this
section will be established and maintained over the operating life of
the commercial nuclear plant.
\1\ The comprehensive risk metrics and related risk performance
objectives established under Sec. 53.220 involve assessing and
averaging the risks over a defined period (e.g., plant year) and do
not constitute a real-time requirement that must be continuously
demonstrated by the licensee.
Sec. 53.715 Maintenance, repair, and inspection programs.
(a) A program to control maintenance activities and monitor the
performance or condition of SR and NSRSS SSCs must be developed,
implemented, and maintained.
(b) Whenever a licensee determines through activities related to
maintenance, repair, and inspection of SSCs, the activities under Sec.
53.710, or otherwise that the performance or condition of an SR or
NSRSS SSC does not demonstrate compliance with established special
treatments or performance goals related to capabilities, availability,
or reliability, the licensee must take appropriate corrective action.
(c) Performance and condition monitoring activities and associated
goals and preventive maintenance activities must be evaluated at least
every 24 months. The evaluations must take into account, where
practical, industry-wide operating experience. Adjustments must be made
where necessary to ensure that the objective of preventing failures of
SSCs through maintenance is appropriately balanced against the
objective of minimizing unavailability of SSCs due to monitoring or
preventive maintenance.
(d) Before performing maintenance activities (including but not
limited to surveillance, post-maintenance testing, and corrective and
preventive maintenance), the licensee must assess and manage the
increase in risk that may result from the proposed maintenance
activities.
Sec. 53.720 Response to seismic events.
If vibratory ground motion exceeding that of the operating basis
earthquake Ground Motion or significant plant damage due to vibratory
ground motion occurs, the licensee must shut down the commercial
nuclear plant. If structures, systems, or components necessary for the
safe shutdown of the commercial nuclear plant are not available after
the occurrence of this vibratory ground motion, the licensee must
consult with the Commission and must propose a plan for the timely,
safe shutdown of the commercial nuclear plant. Prior to resuming
operations, the licensee must demonstrate to the Commission that those
features necessary for continued operation without undue risk to the
health and safety of the public or necessary to maintain the licensing
basis of the commercial nuclear plant were either not functionally
damaged or have been repaired.
Sec. 53.725 General staffing, training, personnel qualifications, and
human factors requirements.
(a) Two classes of commercial nuclear plants. Commercial nuclear
plants licensed under this part are either of the class of self-
reliant-mitigation facilities or of interaction-dependent-mitigation
facilities, based upon the similarity of operating and technical
characteristics of the plants in the class. A commercial nuclear plant
is a self-reliant-mitigation facility if the U.S. Nuclear Regulatory
Commission (NRC) determined as part of its approval of the OL or COL
for that plant that its design demonstrates compliance with the
criteria of Sec. 53.800(a)(1) through (a)(5). Otherwise, the
commercial nuclear plant is an interaction-dependent-mitigation
facility.
(b) Purpose and applicability. The regulations in Sec. Sec. 53.725
through 53.830 address areas related to staffing, training, personnel
qualifications, and human factors engineering for applicants for or
holders of OLs or COLs under this part. These regulations are organized
as follows:
(1) Sections 53.725 through 53.745 address general requirements for
staffing, training, personnel qualifications, and human factors
engineering. The regulations within these sections are applicable to
all applicants for or holders of OLs or COLs under this part, except
where specifically stated otherwise.
(2) Sections 53.760 through 53.795 address operator and senior
operator licensing requirements. The regulations within these sections
are applicable to those applicants for or holders of OLs or COLs under
this part for interaction-dependent-mitigation facilities that have not
yet certified the permanent cessation of operations and permanent
removal of fuel from the reactor vessel as described under Sec.
53.1070.
(3) Sections 53.800 through 53.820 address generally licensed
reactor operator requirements. The regulations within these sections
are in lieu of Sec. Sec. 53.760 through 53.795 for those applicants
for or holders of OLs or COLs under this part for self-reliant-
mitigation facilities that have not yet certified the permanent
cessation of operations and permanent removal of fuel from the reactor
vessel as described under Sec. 53.1070.
[[Page 15814]]
(4) Section 53.830 provides general personnel training
requirements. The regulations within this section are applicable to all
applicants for or holders of OLs or COLs under this part.
(c) Definitions. When used in Sec. Sec. 53.725 through 53.830,
applicant refers to an applicant for an operator or senior operator
license; licensee refers to the holder of an operator, senior operator,
or generally licensed reactor operator license; and facility licensee
refers to the licensee for the commercial nuclear plant where the
applicant would be licensed or the licensee is licensed. As also used
in Sec. Sec. 53.725 through 53.830--
Automation means a device or system that accomplishes (partially or
fully) a function or task.
Auxiliary operator means any individual who operates components of
a commercial nuclear plant 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.
Controls when used with respect to a nuclear reactor means
apparatus and mechanisms, the manipulation of which directly affects
the reactivity or power level of the reactor.
Generally licensed reactor operator means any individual licensed
under the provisions of Sec. 53.810 to manipulate controls of a self-
reliant-mitigation facility and to direct the licensed activities of
generally licensed reactor operators.
Interaction-dependent-mitigation facility means a commercial
nuclear plant design other than one that demonstrates compliance with
the operating and technical characteristics defined under Sec. 53.800.
Load following means a commercial 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. 53.760 through 53.795 to manipulate controls of an
interaction-dependent-mitigation facility.
Performance testing means testing conducted to verify a simulation
facility's performance as compared to actual or predicted reference
plant performance.
Reference plant means the specific commercial nuclear plant, or
plant design for facilities which are not yet constructed, on which a
simulation facility's configuration, system control arrangement, and
design data are based.
Self-reliant-mitigation facility means a commercial nuclear plant
design that demonstrates compliance with the operating and technical
characteristics defined under Sec. 53.800.
Senior operator means any individual licensed under the provisions
of Sec. Sec. 53.760 through 53.795 to manipulate controls of an
interaction-dependent-mitigation facility and to direct the licensed
activities of operators.
Simulation facility means an interface designed to provide a
realistic imitation of the operation of a commercial nuclear plant used
for the administration of examinations, for training, and/or to
demonstrate compliance with experience requirements for applicants or
licensees. 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:
(i) Systematic analysis of the jobs to be performed.
(ii) Learning objectives derived from the analysis which describe
desired performance after training.
(iii) Training design and implementation based on the learning
objectives.
(iv) Evaluation of trainee mastery of the objectives during
training.
(v) Evaluation and revision of the training based on the
performance of trained personnel in the job setting.
Sec. 53.726 Communications.
(a) An applicant or licensee or facility licensee must submit any
communication or report required by the regulations contained within
Sec. Sec. 53.725 through 53.830 and must submit any application filed
under these regulations to the Commission.
(b) Each facility licensee that is required to comply with the
requirements of Sec. Sec. 53.760 through 53.795 (i.e., interaction-
dependent-mitigation facilities) must notify the appropriate NRC
contact within 30 days of the following in regard to a licensed
operator or senior operator:
(1) Permanent reassignment from the position for which the facility
licensee has certified the need for a licensed operator or senior
operator under Sec. 53.775(a)(1);
(2) Termination of any operator or senior operator; or
(3) Permanent disability or illness as required under Sec. 53.770.
Sec. 53.728 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 by 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. 53.730 Defining, fulfilling, and maintaining the role of
personnel in ensuring safe operations.
Each applicant for or holder of an OL or COL for a commercial
nuclear plant under this part must comply with the following:
(a) Human factors engineering design requirements. The plant design
must reflect state-of-the-art human factors engineering principles for
safe and reliable performance in all locations that human activities
are expected for performing or supporting the continued availability of
plant safety or emergency response functions.
(b) 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) for barriers important to fulfilling
safety functions with such devices, 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. Examples
include temperatures and pressures within important systems or
structures, core or fuel system conditions (including possible damage
states), temperatures and levels associated with cooling functions,
combustible gas concentrations, radiation levels in systems and within
structures, and radioactive effluent releases;
(5) Leakage control and detection in the design of systems that
pass through barriers important to fulfilling safety functions for the
release of radionuclides. An example is an SSC
[[Page 15815]]
that penetrates a containment structure that might contain radioactive
materials that could contribute to the source term during an accident;
(6) Monitoring of in-plant radiation and airborne radioactivity as
appropriate for a broad range of normal operating and accident
conditions; and
(7) For self-reliant-mitigation facilities, the plant design must
also provide the generally licensed reactor operators with the
capability 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.
(8) For both interaction-dependent and self-reliant mitigation
facilities, the plant design must provide licensed operators with the
capability of immediately initiating a reactor shutdown from their
location.
(c) Concept of operations. A concept of operations that is of
sufficient scope and detail to address the following must be provided:
(1) Plant goals;
(2) The roles and responsibilities of operating personnel and
automation (or any combination thereof) that are responsible for
completing plant functions;
(3) Staffing, qualifications, and training;
(4) The management of normal operations;
(5) The management of off-normal conditions and emergencies;
(6) The management of maintenance and modifications; and
(7) The management of tests, inspections, and surveillances.
(d) Functional requirements analysis and function allocation. A
functional requirements analysis and a function allocation must be
provided that are sufficient to demonstrate compliance with the
following:
(1) The functional requirements analysis must address how safety
functions and functional safety criteria are satisfied; and
(2) The function allocation must describe how the safety functions
will be assigned to human action, automation, active safety features,
passive safety features, and/or inherent safety characteristics.
(e) Operating experience. A program, during construction and during
operation, as applicable, for evaluating and applying operating
experience must be developed, implemented, and maintained.
(f) Staffing plan. A staffing plan must be developed and comply
with the following:
(1) The staffing plan must include a description of how engineering
expertise will be available to the on-shift operating personnel during
all plant conditions, to assist if they encounter a situation not
covered by procedures or training. Engineering expertise includes
familiarity with the operation of the plant for which the expertise is
provided and one of the following:
(i) A bachelor's degree in engineering, engineering technology, or
physical science from an institution accredited by a U.S. Government
recognized accrediting body or equivalent; or
(ii) A Professional Engineer's license from a U.S. State or
territory.
(2) Applicants for or holders of OLs or COLs for interaction-
dependent-mitigation facilities must include within their staffing
plans a description of how the proposed numbers, positions, and
qualifications of operators and senior operators across all modes of
plant operations will be sufficient to ensure that plant safety
functions will be maintained. This description must be supported by
human factors engineering analyses and assessments.
(3) Applicants for or holders of OLs or COLs for self-reliant-
mitigation facilities must include within their staffing plans a
description of how generally licensed reactor operator staffing that is
both sufficient to continually monitor the operations of fueled
reactors and to provide for a continuity of responsibility for facility
operations at all times during the operating phase will be maintained.
(4) Applicants for or holders of OLs or COLs under this part must
include within their staffing plans 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.
(5) The staffing plan must be approved by the NRC as part of its
approval of the OL or COL for the plant. The approved staffing plan is
subject to the requirements of Sec. 53.1565.
(g) Training, examination, and proficiency programs. Develop,
implement, and maintain programs that comply with the following
requirements. These programs must be approved by the NRC as part of its
approval of the OL or COL for the plant:
(1) For those applicants for or holders of OLs or COLs for
interaction-dependent-mitigation facilities:
(i) The operator licensing initial training program required under
Sec. 53.780(a);
(ii) The operator licensing initial examination program required
under Sec. 53.780(b);
(iii) The operator licensing requalification program required under
Sec. 53.780(c); and
(iv) The operator proficiency program required under Sec.
53.780(g).
(2) For those applicants for or holders of OLs or COLs for self-
reliant-mitigation facilities, the generally licensed reactor operator
training, examination, and proficiency programs required under Sec.
53.815.
(3) The operator licensing requalification programs required under
Sec. 53.780(c) or Sec. 53.815(b) must be implemented upon commencing
the administration of initial examinations under the operator licensing
examination program required under Sec. 53.780(b) or Sec. 53.815(b),
respectively.
Sec. 53.735 General exemptions.
The regulations in Sec. Sec. 53.725 through 53.830 do not require
a license for an individual who--
(a) Under the direction and in the presence of an operator or
senior operator or a generally licensed reactor operator, as
appropriate, manipulates the controls of a commercial nuclear plant as
a part of the individual's training in a facility licensee's training
program as approved by the Commission to qualify for an operator or
senior operator license or a generally licensed reactor operator
license there, as appropriate, under these regulations; or
(b) Under the direction and in the presence of a senior operator or
generally licensed reactor operator, as appropriate, manipulates the
controls of a commercial nuclear plant to load or unload the fuel into,
out of, or within the reactor vessel while the reactor is not
operating.
Sec. 53.740 Facility licensee requirements--general.
(a) Facility licensees must demonstrate compliance with the
requirements of either Sec. Sec. 53.760 through 53.795 for
interaction-dependent-mitigation facilities or Sec. Sec. 53.800
through 53.820 for self-reliant-mitigation facilities.
(b) The facility licensee must maintain the staffing complement
described under its approved facility staffing plan until such time as
the permanent cessation of operations and
[[Page 15816]]
permanent removal of fuel from the reactor vessel has been certified as
described under Sec. 53.1070. The approved staffing plan is subject to
the requirements of Sec. 53.1565.
(c) Except as provided under Sec. 53.735, the facility licensee
may not permit the manipulation of the controls of a commercial nuclear
plant by anyone who is not an operator or senior operator or generally
licensed reactor operator, as appropriate.
(d) Facility licensees for interaction-dependent-mitigation
facilities that have not yet certified the permanent cessation of
operations and permanent removal of fuel from the reactor vessel as
described under Sec. 53.1070 must designate senior operators to be
responsible for supervising the licensed activities of operators.
(e) 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 an operator or senior operator or a generally
licensed reactor operator, as appropriate.
(f)(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; or
(ii) An automated control system; or
(iii) An operator or senior operator or a generally licensed
reactor operator, as appropriate.
(2) The provisions of paragraph (e) of this section do not apply
during load following operations.
(g)(1) Facility licensees for interaction-dependent-mitigation
facilities must have present during alteration of the core (including
fuel loading or transfer) an individual holding 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.
(2) Facility licensees for self-reliant-mitigation facilities must
have present during alteration of the core (including fuel loading or
transfer) an individual holding a generally licensed reactor operator
license to directly supervise the activity and, during this time, the
facility licensee must not assign other duties to this person.
(3) The provisions of paragraphs (g)(1) and (2) 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.
(h) Facility licensees 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. Such facility licensee action must be approved, as a minimum,
by a senior operator or a generally licensed reactor operator, as
applicable, or, after certifying the permanent cessation of operations
and permanent removal of fuel from the reactor vessel as described
under Sec. 53.1070 by a certified fuel handler, senior operator, or
generally licensed reactor operator, as applicable, prior to taking the
action.
Sec. 53.745 Operator license requirements.
A person must be authorized by a license issued by the Commission
to perform the function of an operator, senior operator, or generally
licensed reactor operator as defined in this part.
Sec. 53.760 Operator licensing.
(a) Applicability. Sections 53.760 through 53.795 address operator
and senior operator licensing requirements. The regulations within
these sections are applicable to those applicants for or holders of OLs
or COLs under this part for interaction-dependent-mitigation facilities
that have not yet certified the permanent cessation of operations and
permanent removal of fuel from the reactor vessel as described under
Sec. 53.1070.
(b) [Reserved]
Sec. 53.765 Medical requirements.
(a) An applicant for an operator or senior operator license must
have a medical examination by a physician. An operator or senior
operator must have a medical examination by a physician every 2 years.
(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 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. 53.775(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. 53.770 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. 53.775(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. 53.775(b)) is requested, the
facility licensee must provide medical certification on NRC Form 396 to
the Commission (as described in Sec. 53.765(b)).
Sec. 53.775 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.
[[Page 15817]]
(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.
53.780(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.
53.765.
(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. 53.765(b).
(ii) Examination. The applicant has passed the requisite
examination in accordance with Sec. 53.780(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
additionally, 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. 53.775(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. 53.780 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.
53.725(b), 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 the public health and maintain those plant safety
functions specific to the facility design. The program must be approved
by the Commission prior to its use for training applicants, as
described under Sec. 53.730(g). The approved operator licensing
initial training program is subject to the requirements of Sec.
53.1565.
(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, as described under Sec. 53.730(g). The
approved operator licensing initial examination program is subject to
the requirements of Sec. 53.1565.
(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, as defined by Sec. 53.725(b),
must be utilized for the continuing training of operators and senior
operators.
(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 maintain those 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, as described under Sec. 53.730(g). The
approved operator licensing requalification program is subject to the
requirements of Sec. 53.1565.
[[Page 15818]]
(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, as described under
Sec. 53.730(g). The approved requalification examination program is
subject to the requirements of Sec. 53.1565.
(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 licensed operator or senior licensed
operator who either demonstrates unsatisfactory performance on, or
fails to complete, this biennial requalification examination is removed
from the performance of licensed operator and senior licensed 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 and senior
operators, and facility licensees must not engage in any activity that
compromises the integrity of any application or examination required by
Sec. Sec. 53.760 through 53.795. 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. 53.760 through 53.795.
(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. 53.780(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 commercial 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 required under Sec. 53.620(d)(1)),
replicate the intended initial fuel load for the reference commercial
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. 53.780(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 from 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. 53.780(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. 53.780(b), requalification training
as described in Sec. 53.780(c), or performing control manipulations
that affect reactivity to establish eligibility for an operator or
senior operator license as described in Sec. 53.775(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;
(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
[[Page 15819]]
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 OL or
COL for the plant. The approved proficiency program is subject to the
requirements of Sec. 53.1565.
(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. 53.785 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. 53.780(g).
(f) The licensee must be subject to the facility's Commission-
approved operator licensing requalification and requalification
examination programs required under Sec. 53.780(c).
(g) The licensee must have a biennial medical examination as
described by Sec. 53.765.
(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 commercial 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. 53.790 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 Atomic Energy Act of 1954, as amended, (the Act) 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 Act 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 Act;
(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 Act 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.
53.785(i) or the consumption of alcoholic beverages within the
protected area of commercial nuclear plants, or a determination of
unfitness for scheduled work as a result of the consumption of
alcoholic beverages.
Sec. 53.795 Expiration and renewal of operator and senior operator
licenses.
(a) Expiration. (1) Each operator license and senior operator
license expires 6 years after the date of issuance, 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.
(2) If a licensee files an application for renewal or an upgrade of
an existing license on NRC Form 398 at least 30 days before the
expiration of the existing license, it does not expire until
disposition of the application for renewal or for an upgraded license
has been finally determined by the Commission. Filing by mail will be
deemed to be complete at the time the application is postmarked
(b) Renewal. (1) The applicant for renewal of an operator license
or senior operator license must--
(i) Complete and sign NRC Form 398 and include the number of the
license for which renewal is sought.
(ii) File an original of NRC Form 398 as specified in Sec. 53.775.
(iii) Provide written evidence of the applicant's experience under
the existing license and the approximate number of hours that the
licensee has operated the facility.
(iv) Provide a statement by an authorized representative of the
facility licensee that during the effective term of the current license
the applicant has satisfactorily completed the requalification program
for the facility for which operator or senior operator license renewal
is sought.
(v) Provide evidence that the applicant has discharged the license
responsibilities competently and safely. The Commission may accept as
evidence of the applicant's having met this requirement a certificate
of an authorized representative of the facility licensee or holder of
an authorization by which the licensee has been employed.
[[Page 15820]]
(vi) Provide certification by the facility licensee of medical
condition and general health on NRC Form 396, to comply with Sec.
53.765.
(2) The license will be renewed if the Commission finds that--
(i) The medical condition and the general health of the licensee
continue to be such as not to cause operational errors that endanger
public health and safety. The Commission will base this finding upon
the certification by the facility licensee as described in Sec.
53.765(b).
(ii) The licensee--
(A) Is capable of continuing to competently and safely assume
licensed duties;
(B) Has successfully completed a requalification program that has
been approved by the Commission as required by Sec. 53.780(c); and
(C) Has passed the requalification examinations as required by
Sec. 53.780(c).
(iii) There is a continued need for an operator to operate or for a
senior operator to supervise operators at the facility designated in
the application.
(iv) The past performance of the licensee has been satisfactory to
the Commission. In making its finding, the Commission will include in
its evaluation information such as notices of violations or letters of
reprimand in the licensee's docket.
Sec. 53.800 Facility licensees for self-reliant-mitigation
facilities.
(a) A commercial nuclear plant is a self-reliant-mitigation
facility if the NRC determined as part of its approval of the OL or COL
for that plant that its design demonstrates compliance with the
criteria in paragraphs (a)(1) though (a)(5) of this section. A self-
reliant-mitigation facility is of a class, based upon the similarity of
operating and technical characteristics of the plants in the class,
such that its licensee must comply with the requirements of Sec. Sec.
53.800 through 53.820 in lieu of those in Sec. Sec. 53.760 through
53.795.
(1) The safety performance criteria of Sec. Sec. 53.210 and 53.220
must be met without reliance upon human action for credited event
mitigation.
(2) The results of the probabilistic risk assessment (PRA), other
systematic risk evaluations, or a combination thereof required by Sec.
53.450(a) must demonstrate that the evaluation criteria for the events
analyzed in accordance with Sec. 53.450 will be met without reliance
on human actions to achieve acceptable event mitigation.
(3) The functional requirements analysis and function allocation
performed under Sec. 53.730(d) must demonstrate that functions
required for safety are not reliant upon credited human action.
(4) The plant response to events analyzed under Sec. 53.450 must
rely exclusively on safety features and characteristics that will
neither be rendered unavailable by credible human errors of commission
or omission nor credibly require manual human operation in response to
equipment failures. Compliance with this paragraph (a)(4) may be
achieved through the use of SSCs that function through inherent
characteristics or that have engineered protections against human
failures.
(5) Assessments of credited human actions within the analysis of
design-basis accidents (DBAs) and across the range of LBEs other than
DBAs do not identify important human actions needed to ensure
appropriate defense in depth is provided, as required by Sec. 53.250.
(b) [Reserved]
Sec. 53.805 Facility licensee requirements related to generally
licensed reactor operators.
(a) Licensees for self-reliant-mitigation facilities that have not
yet certified the permanent cessation of operations and permanent
removal of fuel from the reactor vessel as described under Sec.
53.1070 must demonstrate compliance with the following requirements:
(1) Ensure that, in addition to being qualified to perform those
items identified by the facility-specific systems approach to training
conducted under Sec. 53.815, generally licensed reactor operators 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.
(2) Develop, implement, and maintain facility technical
specifications that provide the necessary administrative controls to
ensure the implementation of the requirements in this section.
(3) Develop, implement, and maintain the generally licensed reactor
operator training, examination, and proficiency programs required under
Sec. 53.815.
(4) Ensure that generally licensed reactor operators are subject to
the facility's generally licensed reactor operator training,
examination, and proficiency programs required under Sec. 53.815.
Ensure that generally licensed reactor operators are subject to and
comply with the applicable programmatic requirements for personnel
required under 10 CFR parts 26 and 73. 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
commercial nuclear plant.
(5) Report annually to the NRC the identity of all generally
licensed reactor operators at the commercial nuclear plant, including
all additions and deletions since the previous report.
(6) Ensure that the facility design continues to meet the criteria
of Sec. 53.800.
(b) [Reserved]
Sec. 53.810 Generally licensed reactor operators.
(a) A general license to manipulate the controls of a self-reliant-
mitigation facility 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
self-reliant-mitigation facility and who observes the restrictions of
this section.
(b) A generally licensed reactor operator must comply with the
operating procedures and other conditions specified in the license
authorizing operation of the facility.
(c) The general license is limited to the facility or facilities at
which the operator is employed.
(d) The Commission will suspend the general license on an
individual operator basis for violations of any provision of the Act or
any rule or regulation issued thereunder whenever the Commission deems
such suspension desirable, including--
(1) For willful violation of, or failure to observe, any of the
terms and conditions of the Act or the general license, or of any rule,
regulation, or order of the Commission;
(2) For any conduct determined by the Commission to be a hazard to
safe operation of the facility; or
(3) 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
[[Page 15821]]
drugs, drug metabolites, or alcohol in violation of the conditions and
cutoff levels established by Sec. 53.810(f) or the consumption of
alcoholic beverages within the protected area of commercial nuclear
plants, or a determination of unfitness for scheduled work as a result
of the consumption of alcoholic beverages.
(e) The Commission may require information from a generally
licensed reactor operator to determine whether a general license should
be revoked or suspended with respect to that operator.
(f) The generally licensed reactor operator must not consume or
ingest alcoholic beverages within the protected area of commercial
nuclear plants. The generally licensed reactor operator must not use,
possess, or sell any illegal drugs. The generally licensed reactor
operator 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 (f), with respect to alcoholic beverages and drugs, the
term ``under the influence'' means the generally licensed reactor
operator 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 generally licensed reactor operator 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
generally licensed reactor operator duties.
(g) The generally licensed reactor operator must notify the
Commission within 30 days about a conviction for a felony.
Sec. 53.815 Generally licensed reactor operator training,
examination, and proficiency programs.
(a) Applicability. The requirements of this section apply to each
licensee of a self-reliant-mitigation facility that has not yet
certified the permanent cessation of operations and permanent removal
of fuel from the reactor vessel as described under Sec. 53.1070.
(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)
through (b)(3) of this section.
(2) The training program must provide for both the initial and
continuing training of generally licensed reactor operators and be
derived from a systems approach to training as defined in this part.
(3)(i) The training program must incorporate the instructional
requirements necessary to provide qualified generally licensed reactor
operators 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 (QA) requirements.
Facility licensee management must periodically review the training
program for effectiveness.
(ii) The training program must ensure that generally licensed
reactor operators have and maintain the necessary knowledge, skills,
and abilities.
(iii) The training program must include the generally licensed
reactor operator manipulating the controls of either the facility or a
simulation facility that demonstrates compliance with the requirements
of Sec. 53.815(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 generally licensed reactor
operator 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, to include the examination methods and criteria
to be used to assess passing performance. The requalification
examination program must specify an appropriate periodicity for
administering a complete requalification examination to each generally
licensed reactor operator, 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 generally licensed
reactor operator who either demonstrates unsatisfactory performance on,
or fails to complete, the requalification examination is removed from
the performance of generally licensed reactor operator duties until
such time that any necessary remedial training has been completed and a
retake examination has been passed.
(B) [Reserved]
(vi) The training program must be approved by the Commission prior
to its use, as described under Sec. 53.730(g). The examination program
must provide for valid and reliable examinations and must be approved
by the Commission prior to their use, as described under Sec.
53.730(g). The approved programs are subject to the requirements of
Sec. 53.1565.
(c) Records. The following is required regarding the documentation
of the generally licensed reactor operator 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 generally licensed reactor operator in the
training and examination programs. The records must contain copies of
examinations administered, the answers given by the generally licensed
reactor operator, documentation of the grading of examinations, and
documentation of any additional training administered in areas in which
a generally licensed reactor operator exhibited deficiencies. The
facility licensee must retain these records while the associated
generally licensed reactor operators 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 generally licensed
reactor operator 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
[[Page 15822]]
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 generally licensed
reactor 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 commercial 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 required under Sec. 53.620(d)(1)),
replicate the intended initial fuel load for the reference commercial
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. 53.815(d).
(f) Waiver of examination requirement. The facility licensee may
waive any or all of the requirements for an examination in accordance
with the facility licensee's Commission-approved generally licensed
reactor operator training and examination programs.
(g) Proficiency. The facility licensee must develop, implement, and
maintain a proficiency program to allow generally licensed reactor
operators to maintain proficiency regarding position functions and
familiarity with plant status. This program must include those steps
that will be taken in order to re-establish proficiency when it cannot
be maintained.
Sec. 53.820 Cessation of individual applicability.
The general license ceases to be applicable on an individual basis
once a generally licensed reactor operator is no longer being employed
in a position that may involve the manipulation of the controls of the
self-reliant mitigation facility.
Sec. 53.830 Training and qualification of commercial nuclear
personnel.
(a) This section addresses personnel training requirements. The
regulations within this section are applicable to all applicants for or
holders of OLs or COLs 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 required under Sec. 53.620(d)(1)), each holder of an
operating or COL 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 this part and must provide, at a minimum, for
the training and qualification of the following categories of
commercial nuclear personnel:
(1) Supervisors (e.g., shift supervisors);
(2) Technicians (e.g., maintenance, chemistry, and radiological);
and
(3) Other appropriate operating personnel (e.g., auxiliary
operators, certified fuel handlers, and individuals who provide
engineering expertise to on-shift operating personnel).
(d) The training program must incorporate the instructional
requirements necessary to provide qualified personnel to operate
components of a commercial 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 QA
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.
Sec. 53.845 Programs.
(a) The required plant programs under this part must include but
are not necessarily limited to the programs described in the following
sections of this subpart. Licensees may combine, separate, and
otherwise organize programs and related documents as appropriate for
the technologies and organizations associated with the commercial
nuclear plant.
(b) In addition to the programs described in the following
sections, programs must be provided for each commercial nuclear plant,
if necessary, to ensure that the performance of design features and
human actions are consistent with the analyses performed under
Sec. Sec. 53.450 and 53.730 and that the plant will demonstrate
compliance with the safety criteria defined in Sec. Sec. 53.210 and
53.220.
Sec. 53.850 Radiation protection.
(a) Each holder of an OL or COL under this part must develop,
implement, and maintain a Radiation Protection Program for operations
that is commensurate with the scope and extent of licensed activities
under this part and includes measures for limiting and monitoring
radioactive plant effluents and limiting and monitoring the dose to
individuals working with radioactive materials in accordance with 10
CFR part 20.
(b) Each holder of an OL or COL under this part must develop,
implement, and maintain a program for the control of radioactive
effluents and for environmental monitoring. The program must be
contained in an Offsite Dose Calculations Manual, must be implemented
by procedures, and must include remedial actions to be taken whenever
the program limits are exceeded. The Offsite Dose Calculations Manual
must--
(1) Contain the methodology and parameters used in the calculation
of offsite doses resulting from radioactive
[[Page 15823]]
gaseous and liquid effluents, in the calculation of gaseous and liquid
effluent monitoring alarm and trip setpoints, and in the conduct of the
radiological environmental monitoring program; and
(2) Contain the radioactive effluent controls and radiological
environmental monitoring activities, and descriptions of the
information that should be included in the Annual Radiological
Environmental Operating and Radioactive Effluent Release Reports
required by Sec. 53.1645.
(c) Each holder of an OL or COL under this part must develop,
implement, and maintain a Process Control Program that identifies the
administrative and operational controls for solid radioactive waste
processing, process parameters, and surveillance requirements
sufficient to ensure compliance with the requirements of 10 CFR part
20, 10 CFR part 61, and 10 CFR part 71.
Sec. 53.855 Emergency preparedness.
(a) Each holder of an OL or COL under this part must have an
emergency response plan that must contain information needed to
demonstrate compliance with either the requirements in Sec. 50.160 of
this chapter or the requirements in appendix E to part 50 and the
planning standards of Sec. 50.47(b) of this chapter.
(b) No initial OL, initial COL, or early site permit that includes
complete and integrated emergency plans will be issued under this part
unless a finding is made by the NRC, in accordance with Sec. 50.47 of
this chapter, that there is reasonable assurance that adequate
protective measures can and will be taken in the event of a
radiological emergency.
Sec. 53.860 Security programs.
(a) Physical protection program. Each holder of an OL or COL under
this part must develop, implement, and maintain a physical protection
program under the following requirements:
(1) The licensee must implement security requirements for the
protection of special nuclear material based on the type, enrichment,
and quantity in accordance with 10 CFR part 73, as applicable, and
implement security requirements for the protection of Category 1 and
Category 2 quantities of radioactive material in accordance with 10 CFR
part 37, as applicable; and
(2) The licensee must demonstrate compliance with the provisions
set forth in either Sec. 73.55 or Sec. 73.100 of this chapter.
(b) Fitness-for-duty. Each holder of an OL or COL under this part
must develop, implement, and maintain a fitness-for-duty program under
10 CFR part 26.
(c) Access authorization. Each holder of an OL or COL under this
part must develop, implement, and maintain an access authorization
program under Sec. 73.56 or Sec. 73.120 of this chapter, as
applicable.
(d) Cybersecurity. Each holder of an OL or COL under this part must
develop, implement, and maintain a cybersecurity program under Sec.
73.54 or Sec. 73.110 of this chapter.
(e) Information security. Each holder of an OL or COL 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.
Sec. 53.865 Quality assurance.
Each holder of an OL or COL under this part must develop,
implement, and maintain a quality assurance program in accordance with
appendix B of part 50 of this chapter. A written quality assurance
program manual must be developed and used to guide the conduct of the
program.
Sec. 53.870 Integrity assessment programs.
Each holder of an OL or COL under this part must develop,
implement, and maintain an integrity assessment program to monitor,
evaluate, and manage--
(a) The effects of plant aging on SR and NSRSS SSCs. The program
may refer to surveillances, tests, and inspections conducted for
specific SSCs in accordance with other requirements in this part or
conducted in accordance with applicable consensus codes and standards
endorsed or otherwise found acceptable by the NRC;
(b) Cyclic or transient load limits to ensure that SR and NSRSS
SSCs are maintained within the applicable design limits; and
(c) Degradation mechanisms related to chemical interactions,
operating temperatures, effects of irradiation, and other environmental
factors to ensure that the capabilities, availability, and reliability
of SR and NSRSS SSCs demonstrate compliance with the functional design
criteria of Sec. Sec. 53.410 and 53.420.
Sec. 53.875 Fire protection.
(a)(1) Each holder of an OL or COL under this part must have 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.
(2) The fire protection plan must also describe specific features
necessary to implement the program described in paragraph (a)(1) of
this section such as the following: administrative controls and
personnel requirements for fire prevention and manual fire suppression
activities; automatic and manually operated fire detection and
suppression systems; and the means to limit fire damage to SSCs so that
the capability to demonstrate compliance with the requirements of Sec.
53.210 is ensured.
(b)(1) Each holder of an OL or COL under this part must develop a
performance-based or deterministic fire protection program that
demonstrates compliance with the safety criteria outlined in Sec. Sec.
53.210 and 53.220, related safety functions outlined in Sec. 53.230,
and defense in depth as outlined in Sec. 53.250 with specific fire
protection measures related to fire prevention, fire detection, and
fire suppression.
(2) The fire protection program must comply with the following:
(i) Safety-related and, where appropriate, NSRSS SSCs must be
designed, located, and maintained to minimize, consistent with other
safety requirements, the probability and effect of fires and
explosions.
(ii) Noncombustible and fire-resistant materials must be used
wherever practical throughout the facility, particularly in locations
with SR and NSRSS SSCs.
(iii) 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 SR and NSRSS SSCs.
(iv) Fire suppression systems must be designed and maintained to
ensure that their rupture or inadvertent operation does not
significantly impair the ability of SR and NSRSS SSCs to perform their
safety functions to satisfy Sec. 53.230.
Sec. 53.880 Inservice inspection and inservice testing.
(a) Each holder of an OL or COL under this part must develop,
implement, and maintain a program for inservice inspection (ISI) and
inservice testing (IST) prior to receiving an OL or COL. The ISI/IST
programs must, wherever applicable, be in accordance with generally
accepted consensus codes and standards that have been endorsed or
otherwise found acceptable by the NRC. The ISI/IST program must
[[Page 15824]]
include all inspections and tests required by the codes and standards
used in the design and be supplemented by risk insights that identify
the most important SSCs to plant safety. The types of testing and
inspections and their frequency should be informed by risk insights to
maintain the reliability and performance of SSCs consistent with the
associated design and analyses activities involving those SSCs. Risk
insights must also be used to determine when to conduct the inspections
and tests (e.g., full power, shutdown, refueling) to minimize risk to
the plant workers and the public. The ISI/IST program must be
documented in a written manual and managed by qualified personnel
reporting to the director, responsible officer, or designated person.
(b) Prior to plant operation, baseline inspections and testing must
be performed using the same techniques as will be used for future
inspections and testing. The results of these inspections and testing
must be used as benchmarks for evaluating the results of future
inspections and testing. Sufficient room and support must be provided
to accommodate the personnel, ISI/IST equipment, and shielding
necessary to perform the inspections and testing. Acceptance criteria
for determining whether corrective action is needed must be developed
(or taken from the codes and standards used in the design) for
evaluating the results of the inspections and testing. The results of
the inspections and testing must be provided to the director,
responsible officer, or designated person who is responsible for
determining what, if any, corrective action is needed and when it
should be taken. The ISI/IST results and corrective actions must be
documented and the documentation retained for the life of the plant.
Sec. 53.910 Procedures and guidelines.
(a) Each holder of an OL or COL under this part must have a program
for developing, implementing, and maintaining an integrated set of
procedures, guidelines, and related supporting activities to support
normal operations and respond to possible unplanned events.
(b) The program required by paragraph (a) of this section must
include but is not limited to development, implementation, maintenance,
and supporting activities of procedures and guidelines for the
following:
(1) Plant operations;
(2) Maintenance activities under Sec. 53.715;
(3) Program requirements under this subpart;
(4) Emergency operating procedures, if developed to address the
role of human actions in responding to LBEs;
(5) Accident management guidelines, if developed to address the
role of human actions in responding to LBEs;
(6) Procedures for each area in which licensed special nuclear
material is handled, used, or stored to protect personnel upon the
sounding of a criticality alarm required by Sec. 53.440(m); and
(7) Procedures that describe how the licensee will address the
following areas if the licensee is notified of a potential aircraft
threat:
(i) Verification of the authenticity of threat notifications;
(ii) Maintenance of continuous communication with threat
notification sources;
(iii) Contacting all onsite personnel and applicable offsite
response organizations;
(iv) Onsite actions necessary to enhance the capability of the
facility to mitigate the consequences of an aircraft impact;
(v) Measures to reduce visual discrimination of the site relative
to its surroundings or individual buildings within the protected area;
(vi) Dispersal of equipment and personnel, as well as rapid entry
into site protected areas for essential onsite personnel and offsite
responders who are necessary to mitigate the event; and
(vii) Recall of site personnel.
Subpart G--Decommissioning Requirements
Sec. 53.1000 Scope and purpose.
This subpart defines the requirements related to decommissioning
for applicants for, or holders of, an operating license (OL) or
combined license (COL). The requirements related to maintaining
financial assurance for decommissioning are in Sec. Sec. 53.1010
through 53.1060. The requirements for transitioning from operations to
decommissioning and for the release of property and termination of the
license are in Sec. Sec. 53.1070 through 53.1080.
Sec. 53.1010 Financial assurance for decommissioning.
(a) This section establishes requirements for indicating to the
U.S. Nuclear Regulatory Commission (NRC) how an applicant for or holder
of an OL or COL under this part will provide reasonable assurance that
funds will be available for the decommissioning process. Reasonable
assurance consists of a series of steps as provided in paragraph (b) of
this section and Sec. Sec. 53.1020, 53.1030 and 53.1040. Funding for
the decommissioning of commercial nuclear plants may also be subject to
the regulation of Federal or State government agencies (e.g., Federal
Energy Regulatory Commission (FERC) and State Public Utility
Commissions) that have jurisdiction over rate regulation. The
requirements of this subpart, in particular Sec. 53.1020, are in
addition to, and not a substitution for, other requirements, and are
not intended to be used by themselves or by other agencies to establish
rates.
(b) Each applicant for an OL or COL under this part must prepare a
plan and an associated decommissioning report that ensures and
documents that adequate funding will be available to decommission the
facility. Each holder of an OL or COL must implement and maintain the
plan.
(1)(i) Before the Commission issues an OL under this part, the
applicant must update the decommissioning report to certify that it has
provided financial assurance for decommissioning in the amount proposed
in the application and approved by the NRC under Sec. 53.1020.
(ii) No later than 30 days after the Commission issues the notice
of intended operation under Sec. 53.1452 for a COL under this part,
the licensee must update the decommissioning report to certify that it
has provided financial assurance for decommissioning in the amount
proposed in the application and approved by the NRC under Sec.
53.1020.
(2) The amount of financial assurance for decommissioning to be
provided must be based on a site-specific cost estimate for
decommissioning the facility under Sec. 53.1020.
(3) The amount of financial assurance for decommissioning to be
provided must be adjusted annually using a rate at least equal to that
stated in Sec. 53.1030.
(4) The amount of financial assurance for decommissioning to be
provided must be covered by one or more of the methods described in
Sec. 53.1040 as acceptable to the NRC. A copy of the financial
instrument obtained to satisfy the requirements of Sec. 53.1040 must
be submitted to the NRC as part of the application for an OL under this
part; however, an applicant for or holder of a COL need not obtain such
financial instrument or submit a copy to the Commission except as
provided in Sec. 53.1060(b).
Sec. 53.1020 Cost estimates for decommissioning.
Cost estimates for decommissioning must be site-specific. Site-
specific decommissioning cost estimates (DCEs) must account for the
engineering, labor,
[[Page 15825]]
equipment, transportation, disposal, and related charges needed to
support termination of the license. They must include the costs for
decontaminating structures, systems, and components and the site
environs; removal of contaminated components and materials from the
plant and the site environs; disposal of removed components and
materials in appropriate facilities; and any other activities
supporting the release of the property and termination of the license.
They must also address the approach to annual adjustments required by
Sec. 53.1030. Finally, site-specific DCEs must include plans for
adjusting levels of funds assured for decommissioning to demonstrate
that a reasonable level of assurance will be provided that funds will
be available when needed to cover the cost of decommissioning.
Sec. 53.1030 Annual adjustments to cost estimates for
decommissioning.
Each holder of an OL or COL under this part must annually adjust
the cost estimate for decommissioning to account for escalation in
labor, energy, and waste burial costs. Licensees may elect to use
either a site-specific adjustment factor, approved as part of the plan
and associated decommissioning report required by Sec. 53.1010, in
paragraph (a) of this section or the generic adjustment factor in
paragraph (b) of this section.
(a) A site-specific adjustment factor must address the estimated
contributions and escalation of costs for the following aspects of
decommissioning:
(1) Labor, materials, and services;
(2) Energy and waste transportation; and
(3) Radioactive waste burial or other disposition.
(b) A generic adjustment factor must be at least equal to 0.65 L +
0.13 E + 0.22 B, where L and E are escalation factors for labor and
energy, respectively, and are to be taken from regional data of U.S.
Department of Labor Bureau of Labor Statistics and B is an escalation
factor for waste burial and is to be taken from NRC report NUREG-1307,
``Report on Waste Burial Charges.''
Sec. 53.1040 Methods for providing financial assurance for
decommissioning.
Financial assurance for decommissioning is to be provided by the
following methods.
(a) Prepayment. Prepayment is the deposit made preceding the start
of operation or the transfer of a license under Sec. 53.1570 into an
account segregated from licensee assets and outside the administrative
control of the licensee and its subsidiaries or affiliates of cash or
liquid assets such that the amount of funds would be sufficient to pay
decommissioning costs. Prepayment may be in the form of a trust, escrow
account, or Government fund with payment by certificate of deposit,
deposit of government or other securities, or other method acceptable
to the NRC. This trust, escrow account, Government fund, or other type
of agreement must be established in writing and maintained at all times
in the United States with an entity that is an appropriate State or
Federal government agency, or an entity whose operations in which the
prepayment deposit is managed are regulated and examined by a Federal
or State agency. A licensee that has prepaid funds based on a site-
specific cost estimate under Sec. 53.1020 may take credit for
projected earnings on the prepaid decommissioning trust funds, using up
to a 2 percent annual real rate of return through the time of
termination of the license. A licensee may use a credit of greater than
2 percent if the licensee's rate-setting authority has specifically
authorized a higher rate. Actual earnings on existing funds may be used
to calculate future fund needs.
(b) External sinking fund. An external sinking fund is a fund
established and maintained by setting funds aside periodically in an
account segregated from licensee assets and outside the administrative
control of the licensee and its subsidiaries or affiliates in which the
total amount of funds would be sufficient to pay decommissioning costs.
An external sinking fund may be in the form of a trust, escrow account,
or Government fund, with payment by certificate of deposit, deposit of
government or other securities, or other method acceptable to the NRC.
This trust, escrow account, Government fund, or other type of agreement
must be established in writing and maintained at all times in the
United States with an entity that is an appropriate State or Federal
government agency, or an entity whose operations in which the external
sinking fund is managed are regulated and examined by a Federal or
State agency. A licensee that has collected funds based on a site-
specific cost estimate under Sec. 53.1020 may take credit for
projected earnings on the external sinking funds using up to a 2
percent annual real rate of return from the time of future funds'
collection through the time of termination of the license. A licensee
may use a credit of greater than 2 percent if the licensee's rate-
setting authority has specifically authorized a higher rate. Actual
earnings on existing funds may be used to calculate future fund needs.
A licensee whose rates for decommissioning costs cover only a portion
of these costs may make use of this method only for the portion of
these costs that are collected in one of the manners described in this
paragraph (b). This method may be used as the exclusive mechanism
relied upon for providing financial assurance for decommissioning in
the following circumstances:
(1) By a licensee that recovers, either directly or indirectly, the
estimated total cost of decommissioning through rates established by
``cost of service'' or similar ratemaking regulation. Public utility
districts, municipalities, rural electric cooperatives, and State and
Federal agencies, including associations of any of the foregoing, that
establish their own rates and are able to recover their cost of service
allocable to decommissioning, are deemed to satisfy this condition.
(2) By a licensee whose source of revenues for its external sinking
fund is a ``non-bypassable charge,'' the total amount of which will
provide funds estimated to be needed for decommissioning pursuant to
Sec. 53.1020, Sec. 53.1060, or Sec. 53.1575.
(c) A surety method, insurance, or other guarantee method. (1)
These methods guarantee that decommissioning costs will be paid. A
surety method may be in the form of a surety bond, or letter of credit.
Any surety method or insurance used to provide financial assurance for
decommissioning must contain the following conditions:
(i) The surety method or insurance must be open-ended, or, if
written for a specified term, such as 5 years, must be renewed
automatically, unless 90 days or more prior to the renewal day the
issuer notifies the NRC, the beneficiary, and the licensee of its
intention not to renew. The surety or insurance must also provide that
the full-face amount be paid to the beneficiary automatically prior to
the expiration without proof of forfeiture if the licensee fails to
provide a replacement acceptable to the NRC within 30 days after
receipt of notification of cancellation.
(ii) The surety or insurance must be payable to a trust established
for decommissioning costs. The trustee and trust must be acceptable to
the NRC. An acceptable trustee includes an appropriate State or Federal
government agency or an entity that has the authority to act as a
trustee and whose trust operations are regulated and examined by a
Federal or State agency.
[[Page 15826]]
(2) A parent company guarantee of funds for decommissioning costs
based on a financial test may be used if the guarantee and test are as
contained in appendix A to 10 CFR part 30.
(3) For commercial companies that issue bonds, a guarantee of funds
by the applicant or licensee for decommissioning costs based on a
financial test may be used if the guarantee and test are as contained
in appendix C to 10 CFR part 30. For commercial companies that do not
issue bonds, a guarantee of funds by the applicant or licensee for
decommissioning costs may be used if the guarantee and test are as
contained in appendix D to 10 CFR part 30. A guarantee by the applicant
or licensee may not be used in any situation in which the applicant or
licensee has a parent company holding majority control of voting stock
of the company.
(d) Funding method for Federal licensees. For a Federal licensee, a
statement of intent containing a cost estimate for decommissioning and
indicating that funds for decommissioning will be obtained when
necessary.
(e) Contractual funding method. Contractual obligation(s) on the
part of a licensee's customer(s), the total amount of which over the
duration of the contract(s) will provide the licensee's total share of
uncollected funds estimated to be needed for decommissioning pursuant
to Sec. 53.1020, Sec. 53.1060, or Sec. 53.1575. To be acceptable to
the NRC as a method of decommissioning funding assurance, the terms of
the contract(s) must include provisions that the buyer(s) of
electricity or other products will pay for the decommissioning
obligations specified in the contract(s), notwithstanding the
operational status either of the licensed plant to which the
contract(s) pertains or force majeure provisions. All proceeds from the
contract(s) for decommissioning funding will be deposited to the
external sinking fund. The NRC reserves the right to evaluate the terms
of any contract(s) and the financial qualifications of the contracting
entity or entities offered as assurance for decommissioning funding.
(f) Other funding mechanisms. Any other mechanism, or combination
of mechanisms, that provides, as determined by the NRC upon its
evaluation of the specific circumstances of each licensee submittal,
assurance of decommissioning funding equivalent to that provided by the
mechanisms specified in paragraphs (a) through (e) of this section.
Licensees who do not have sources of funding described in paragraph (b)
of this section may use an external sinking fund in combination with a
guarantee mechanism, as specified in paragraph (c) of this section,
provided that the total amount of funds estimated to be necessary for
decommissioning is assured.
Sec. 53.1045 Limitations on the use of decommissioning trust funds.
(a)(1) Decommissioning trust funds may be used by licensees if--
(i) The withdrawals are for expenses for decommissioning activities
consistent with the definition of decommission or decommissioning in
Sec. 53.020;
(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) Initially, 3 percent of the amount determined in accordance
with Sec. 53.1020 may be used for decommissioning planning. For
licensees that have submitted the certifications required under Sec.
53.1070 and commencing 90 days after the NRC has received the post-
shutdown decommissioning activities report (PSDAR) required by Sec.
53.1060, an additional 20 percent may be used. An updated site-specific
DCE must be submitted to the NRC prior to the licensee using any
funding in excess of these amounts.
(b) Licensees that are not ``electric utilities'' as defined in
Sec. 53.020 that use prepayment or an external sinking fund to provide
financial assurance must provide in the terms of the arrangements
governing the trust, escrow account, or Government fund, used to
segregate and manage the funds that--
(1) The trustee, manager, investment advisor, or other person
directing investment of the funds--
(i) Is prohibited from investing the funds in securities or other
obligations of the licensee or any other owner or operator of any
commercial nuclear plant or their affiliates, subsidiaries, successors
or assigns, or in a mutual fund in which at least 50 percent of the
fund is invested in the securities of a licensee or parent company
whose subsidiary is an owner or operator of a foreign or domestic
commercial nuclear plant. However, the funds may be invested in
securities tied to market indices or other non-nuclear sector
collective, commingled, or mutual funds, provided that no more than 10
percent of trust assets may be indirectly invested in securities of any
entity owning or operating one or more commercial nuclear plants.
(ii) Is obligated at all times to adhere to a standard of care set
forth in the trust, which either shall be the standard of care, whether
in investing or otherwise, required by State or Federal law or one or
more State or Federal regulatory agencies with jurisdiction over the
trust funds, or, in the absence of any such standard of care, whether
in investing or otherwise, that a prudent investor would use in the
same circumstances. The term ``prudent investor,'' shall have the same
meaning as set forth in FERC's ``Regulations Governing Nuclear Plant
Decommissioning Trust Funds'' at 18 CFR 35.32(a)(3), or any successor
regulation.
(2) The licensee, its affiliates, and its subsidiaries are
prohibited from being engaged as investment manager for the funds or
from giving day-to-day management direction of the funds' investments
or direction on individual investments by the funds, except in the case
of passive fund management of trust funds where management is limited
to investments tracking market indices.
(3) The trust, escrow account, Government fund, or other account
used to segregate and manage the funds may not be amended in any
material respect without written notification to the Director, Office
of Nuclear Reactor Regulation, or Director, Office of Nuclear Material
Safety and Safeguards, as applicable, at least 30 working days before
the proposed effective date of the amendment. The licensee must provide
the text of the proposed amendment and a statement of the reason for
the proposed amendment. The trust, escrow account, Government fund, or
other account may not be amended if the person responsible for managing
the trust, escrow account, Government fund, or other account receives
written notice of objection from the Director, Office of Nuclear
Reactor Regulation, or Director, Office of Nuclear Material Safety and
Safeguards, as applicable, within the notice period.
(4) Except for withdrawals being made under paragraph (a) of this
section or for payments of ordinary administrative costs (including
taxes) and other incidental expenses of the fund (including legal,
accounting, actuarial, and trustee expenses) in connection with the
operation of the fund, no disbursement or payment may be made from the
trust, escrow account,
[[Page 15827]]
Government fund, or other account used to segregate and manage the
funds until written notice of the intention to make a disbursement or
payment has been given to the Director, Office of Nuclear Reactor
Regulation, or Director, Office of Nuclear Material Safety and
Safeguards, as applicable, at least 30 working days before the date of
the intended disbursement or payment. The disbursement or payment from
the trust, escrow account, Government fund or other account may be made
following the 30 working day notice period if the person responsible
for managing the trust, escrow account, Government fund, or other
account does not receive written notice of objection from the Director,
Office of Nuclear Reactor Regulation, or Director, Office of Nuclear
Material Safety and Safeguards, as applicable, within the notice
period. Disbursements or payments from the trust, escrow account,
Government fund, or other account used to segregate and manage the
funds, other than for payment of ordinary administrative costs
(including taxes) and other incidental expenses of the fund (including
legal, accounting, actuarial, and trustee expenses) in connection with
the operation of the fund, are restricted to decommissioning expenses
or transfer to another financial assurance method acceptable under
Sec. 53.1040 until final decommissioning has been completed. After
decommissioning has begun and withdrawals from the decommissioning fund
are made under paragraph (a) of this section, no further notification
need be made to the NRC.
(c) Licensees that are ``electric utilities'' under Sec. 53.020
that use prepayment or an external sinking fund to provide financial
assurance must include a provision in the terms of the trust, escrow
account, Government fund, or other account used to segregate and manage
funds that except for withdrawals being made under paragraph (a) of
this section or for payments of ordinary administrative costs
(including taxes) and other incidental expenses of the fund (including
legal, accounting, actuarial, and trustee expenses) in connection with
the operation of the fund, no disbursement or payment may be made from
the trust, escrow account, Government fund, or other account used to
segregate and manage the funds until written notice of the intention to
make a disbursement or payment has been given the Director, Office of
Nuclear Reactor Regulation, or Director, Office of Nuclear Material
Safety and Safeguards, as applicable, at least 30 working days before
the date of the intended disbursement or payment. The disbursement or
payment from the trust, escrow account, Government fund or other
account may be made following the 30 working day notice period if the
person responsible for managing the trust, escrow account, Government
fund, or other account does not receive written notice of objection
from the Director, Office of Nuclear Reactor Regulation, or Director,
Office of Nuclear Material Safety and Safeguards, as applicable, within
the notice period. Disbursements or payments from the trust, escrow
account, Government fund, or other account used to segregate and manage
the funds, other than for payment of ordinary administrative costs
(including taxes) and other incidental expenses of the fund (including
legal, accounting, actuarial, and trustee expenses) in connection with
the operation of the fund, are restricted to decommissioning expenses
or transfer to another financial assurance method acceptable under
Sec. 53.1040 until final decommissioning has been completed. After
decommissioning has begun and withdrawals from the decommissioning fund
are made under paragraph (a) of this section, no further notification
need be made to the NRC.
(d) A licensee that is not an ``electric utility'' under Sec.
53.020 and using a surety method, insurance, or other guarantee method
to provide financial assurance must provide that the trust established
for decommissioning costs to which the surety or insurance is payable
contains in its terms the requirements in Sec. 53.1045(b)(1) through
(b)(4).
Sec. 53.1050 NRC oversight.
The NRC reserves the right to take the following steps in order to
ensure a licensee's adequate accumulation of decommissioning funds:
review, as needed, the rate of accumulation of decommissioning funds
and, either independently or in cooperation with FERC and the
licensee's State Public Utility Commission, take additional actions as
appropriate on a case-by-case basis, including modification of a
licensee's schedule for the accumulation of decommissioning funds.
Sec. 53.1060 Reporting and recordkeeping requirements.
(a) Each holder of an OL under this part or holder of a COL under
this part after the date that the Commission has made the finding under
Sec. 53.1452(g) must report, at least once every 2 years, by March 31,
on the status of its certification of decommissioning funding for each
commercial nuclear reactor or part of a commercial nuclear reactor that
it owns. The information in this report must include, at a minimum, the
amount of decommissioning funds estimated to be required under
Sec. Sec. 53.1020 and 53.1030; 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; 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; any contracts upon which the
licensee is relying under Sec. 53.1040(e); any modifications occurring
to a licensee's method of providing financial assurance since the last
submitted report; and any material changes to trust agreements. If any
of the preceding items is not applicable, the licensee should so state
in its report. Any licensee for a plant that is within 5 years of the
projected end of its operation, or where conditions have changed such
that it will close within 5 years (before the end of its licensed
life), or that has already closed (before the end of its licensed
life), or that is involved in a merger or an acquisition must submit
this report annually.
(b) Each holder of a COL under this part must, 2 years before and 1
year before the scheduled date for initial loading of fuel (or, for a
fueled manufactured reactor, 2 years before and 1 year before the
scheduled date for initiating the removal of the features to prevent
criticality required under Sec. 53.620(d)(1)) submit a report to the
NRC containing a certification updating the DCEs and a copy of the
financial instrument to be used to satisfy Sec. 53.1040. No later than
30 days after the Commission publishes notice in the Federal Register
under Sec. 53.1452(a), the licensee must submit an updated
decommissioning report required under Sec. 53.1010(b)(1)(ii),
including a copy of the financial instrument obtained to satisfy Sec.
53.1040.
(c) Each licensee must keep records of information important to the
safe and effective decommissioning of the facility in an identified
location until the license is terminated by the Commission. If records
of relevant information are kept for other purposes, reference to these
records and their locations may be used. Information the Commission
considers important to decommissioning consists of--
(1) Records of spills or other unusual occurrences involving the
spread of contamination in and around the facility, equipment, or site.
These
[[Page 15828]]
records may be limited to instances when significant contamination
remains after any cleanup procedures or when there is reasonable
likelihood that contaminants may have spread to inaccessible areas as
in the case of possible seepage into porous materials such as concrete.
These records must include any known information on identification of
involved nuclides, quantities, forms, and concentrations.
(2) As-built drawings and modifications of structures and equipment
in restricted areas where radioactive materials are used and/or stored
and of locations of possible inaccessible contamination such as buried
pipes that may be subject to contamination. If required drawings are
referenced, each relevant document need not be indexed individually. If
drawings are not available, the licensee must substitute appropriate
records of available information concerning these areas and locations.
(3) Records of the cost estimate performed for the decommissioning
funding plan or of the amount certified for decommissioning, and
records of the funding method used for assuring funds if either a
funding plan or certification is used.
(4) Records of--
(i) The licensed site area, as originally licensed and any
revisions, which must include a site map and any acquisition or use of
property outside the originally licensed site area for the purpose of
receiving, possessing, or using licensed materials;
(ii) The licensed activities carried out on the acquired or used
property; and
(iii) The release and final disposition of any property recorded in
paragraph (c)(4)(i) of this section, the historical site assessment
performed for the release, radiation surveys performed to support
release of the property, submittals to the NRC made under Sec.
53.1070, and the methods employed to ensure that the property met the
radiological criteria of subpart E of 10 CFR part 20 at the time the
property was released.
(d) Each holder of an OL or COL under this part must at or about 5
years prior to the projected end of operations submit a preliminary DCE
which includes an up-to-date assessment of the major factors that could
affect the cost to decommission.
(e) Prior to or within 2 years following permanent cessation of
operations, the licensee must submit a PSDAR to the NRC, and a copy to
the affected State(s). The PSDAR must contain a description of the
planned decommissioning activities along with a schedule for their
accomplishment, a discussion that provides the reasons for concluding
that the environmental impacts associated with site-specific
decommissioning activities will be bounded by appropriate previously
issued environmental impact statements, and a site-specific DCE,
including the projected cost of managing irradiated fuel.
(f) For decommissioning activities that delay completion of
decommissioning by including a period of storage or surveillance, the
licensee must provide a means of adjusting cost estimates and
associated funding levels over the storage or surveillance period.
(g) After submitting its site-specific DCE required by paragraph
(e) of this section, 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:
(1) 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;
(2) 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;
(3) Any modifications occurring to a licensee's current method of
providing financial assurance since the last submitted report; and
(4) Any material changes to trust agreements or financial assurance
contracts.
(5) 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.
(h) After submitting its site-specific DCE required by paragraph
(e) of this section, the licensee must annually submit to the NRC, by
March 31, a report on the status of its funding for managing irradiated
fuel. The report must include the following information, current
through the end of the previous calendar year:
(1) The amount of funds accumulated to cover the cost of managing
the irradiated fuel;
(2) The projected cost of managing irradiated fuel until title to
the fuel and possession of the fuel is transferred to the Secretary of
Energy; and
(3) If the funds accumulated do not cover the projected cost, a
plan to obtain additional funds to cover the cost.
Sec. 53.1070 Termination of license.
For each holder of an OL or COL under this part--
(a)(1) When the 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.
53.040(b)(8);
(2) When appropriate to support decommissioning activities and the
eventual permanent removal of fuel from the reactor vessel, the
licensee must develop defueled technical specifications by reviewing
the operational technical specifications and determining which
specifications no longer apply during decommissioning and which ones
should remain applicable. The licensee must make the appropriate
submittals to the NRC in accordance with Sec. 53.1510 to request
changes to the technical specifications; and
(3)(i) Once fuel has been permanently removed from the reactor
vessel, the licensee must submit a written certification to the NRC
that meets the requirements of Sec. 53.040(b)(9); and
(ii) The licensee must establish and maintain staffing consisting
of certified fuel handlers, as defined under Sec. 53.020, and other
non-licensed personnel with appropriate qualifications, and in
sufficient numbers, to ensure support for facility operations and
radiological control activities, as required by the facility defueled
technical specifications. These personnel must be subject to the
training requirements of Sec. 53.830.
(b) Upon docketing of the certifications for permanent cessation of
operations and permanent removal of fuel from the reactor vessel, or
when a final legally effective order to permanently cease operations
has come into effect, the license issued under this part no longer
authorizes operation of the reactor or emplacement or retention of fuel
into the reactor vessel.
(c) Decommissioning will be completed within 60 years of permanent
cessation of operations. Completion of decommissioning beyond 60 years
will be approved by the Commission only when necessary to protect
public health and safety. Factors that will be
[[Page 15829]]
considered by the Commission in evaluating an alternative that provides
for completion of decommissioning beyond 60 years of permanent
cessation of operations include unavailability of waste disposal
capacity and other site-specific factors affecting the licensee's
capability to carry out decommissioning, including presence of other
nuclear facilities at the site.
(d)(1) Prior to or within 2 years following permanent cessation of
operations, the licensee must submit a PSDAR and site-specific DCE in
accordance with Sec. 53.1060(e).
(2) The NRC must notice receipt of the PSDAR and make the PSDAR
publicly available and publish notice of its availability for public
comment in the Federal Register. The NRC must also schedule a public
meeting readily accessible to individuals in the vicinity of the
licensee's facility. The NRC must publish a notice in the Federal
Register and in a forum, such as local newspapers, that is readily
accessible to individuals in the vicinity of the site, announcing the
date, time, and location of the meeting, along with a brief description
of the purpose of the meeting.
(e) Licensees must not perform any major decommissioning
activities, as defined in Sec. 53.020, until 90 days after the NRC has
received the licensee's PSDAR submittal and until certifications of
permanent cessation of operations and permanent removal of fuel from
the reactor vessel, as required under paragraph (a) of this section,
have been submitted.
(f) Licensees must not perform any decommissioning activities, as
defined in Sec. 53.020, that--
(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.
(g) In taking actions permitted under Sec. 53.1540 following
submittal of the PSDAR, the licensee must notify the NRC in writing,
and send a copy to the affected State(s), before performing any
decommissioning activity inconsistent with, or making any significant
schedule change from, those actions and schedules described in the
PSDAR, including changes that increase the decommissioning cost by more
than 20 percent from the previously provided DCE.
(h) Licensees may use decommissioning trust funds consistent with
the limitations of Sec. 53.1045(a). Licensees must report on the
status of decommissioning trust funds consistent with the requirements
of Sec. 53.1060.
(i) Licensees must submit an application for termination of license
in accordance with Sec. 53.1070. The application for termination of
license must be accompanied or preceded by a license termination plan
to be submitted for NRC approval.
(1) The license termination plan must be a supplement to the Final
Safety Analysis Report or equivalent and must be submitted at least 2
years before termination of the license date.
(2) The license termination plan must include--
(i) A site characterization;
(ii) Identification of remaining dismantlement activities;
(iii) Plans for site remediation;
(iv) Detailed plans for the final radiation survey;
(v) A description of the end use of the site, if restricted;
(vi) An updated site-specific estimate of remaining decommissioning
costs;
(vii) A supplement to the environmental report, pursuant to Sec.
51.53 of this chapter, describing any new information or significant
environmental change associated with the licensee's proposed
termination activities; and
(viii) Identification of parts, if any, of the facility or site
that were released for use before approval of the license termination
plan.
(3) Following receipt of the license termination plan, the NRC must
make the license termination plan publicly available and publish notice
of its availability for public comment in the Federal Register. The NRC
must also schedule a public meeting readily accessible to individuals
in the vicinity of the licensee's facility upon receipt of the license
termination plan. The NRC must publish a notice in the Federal Register
and in a forum, such as local newspapers, that is readily accessible to
individuals in the vicinity of the site, announcing the date, time, and
location of the meeting, along with a brief description of the purpose
of the meeting.
(j) If the license termination plan demonstrates that the remainder
of decommissioning activities will be performed in accordance with the
regulations in this chapter, will not be inimical to the common defense
and security or to the health and safety of the public, and will not
have a significant effect on the quality of the environment and after
notice to interested persons, the Commission will approve the plan, by
license amendment, subject to such conditions and limitations as it
deems appropriate and necessary and authorize implementation of the
license termination plan.
(k) The Commission will terminate the license if it determines
that--
(1) The remaining dismantlement has been performed in accordance
with the approved license termination plan; and
(2) The final radiation survey and associated documentation,
including an assessment of dose contributions associated with parts
released for use before approval of the license termination plan,
demonstrate that the facility and site have met the criteria for
decommissioning in subpart E of 10 CFR part 20.
Sec. 53.1075 Program requirements during decommissioning.
(a) Licensees that have submitted the certifications required under
Sec. 53.1070 must maintain a decommissioning fire protection program
to address the potential for fires that could cause the release or
spread of radioactive materials.
(1) The objectives of the decommissioning fire protection program
are to
(i) Reasonably prevent these fires from occurring;
(ii) Rapidly detect, control, and extinguish those fires that do
occur and that could result in a radiological hazard; and
(iii) Ensure that the risk of fire-induced radiological hazards to
the public, environment, and plant personnel is minimized.
(2) The licensee must assess the decommissioning fire protection
program on a regular basis. The licensee must revise the
decommissioning fire protection program documentation as appropriate
throughout the various stages of facility decommissioning.
(3) The licensee may make changes to the decommissioning fire
protection program without NRC approval if these changes do not reduce
the effectiveness of fire protection for structures, systems, and
components that could result in a radiological hazard, taking into
account the decommissioning plant conditions and activities.
(b) [Reserved]
Sec. 53.1080 Release of part of a commercial nuclear plant or site
for unrestricted use.
(a) Prior written NRC approval is required to release part of a
commercial nuclear plant or site for unrestricted use at any time
before receiving approval of a license termination plan. Section
53.1060 specifies recordkeeping requirements associated with partial
release. Holders of an OL or COL under
[[Page 15830]]
this part seeking NRC review and approval must--
(1) Evaluate the effect of releasing the property to ensure that--
(i) The dose to individual members of the public does not exceed
the limits and standards of subpart D of 10 CFR part 20;
(ii) There is no reduction in the effectiveness of emergency
planning or physical security;
(iii) Effluent releases remain within license conditions;
(iv) The environmental monitoring program and offsite dose
calculation manual are revised to account for the changes;
(v) The siting criteria of subpart D of this part continue to be
met; and
(vi) All other applicable statutory and regulatory requirements
continue to be met.
(2) Perform a historical site assessment of the part of the
commercial nuclear plant or site to be released; and
(3) Perform surveys adequate to demonstrate compliance with the
radiological criteria for unrestricted use specified in Sec. 20.1402
of this chapter for impacted areas.
(b) For release of non-impacted areas, the licensee may submit a
written request for NRC review and approval of the release if a license
amendment is not otherwise required. The request submittal must
include--
(1) The results of the evaluations performed in accordance with
paragraphs (a)(1) and (a)(2) of this section;
(2) A description of the part of the commercial nuclear plant or
site to be released;
(3) The schedule for release of the property;
(4) The results of the evaluations performed in accordance with
Sec. 53.1540; and
(5) A discussion that provides the reasons for concluding that the
environmental impacts associated with the licensee's proposed release
of the property will be bounded by appropriate previously issued
environmental impact statements.
(c) After receiving a request from the licensee for NRC approval of
the release of a non-impacted area, the NRC must--
(1) Determine whether the licensee has adequately evaluated the
effect of releasing the property as required by paragraph (a)(1) of
this section;
(2) Determine whether the licensee's classification of any release
areas as non-impacted is adequately justified; and
(3) If determining that the licensee's submittal is adequate,
inform the licensee in writing that the release is approved.
(d) For release of impacted areas, the licensee must submit an
application for amendment of its license for the release of the
property. The application must include--
(1) The information specified in paragraphs (b)(1) through (b)(3)
of this section;
(2) The methods used for and results obtained from the radiation
surveys required to demonstrate compliance with the radiological
criteria for unrestricted use specified in Sec. 20.1402; and
(3) A supplement to the environmental report, under Sec. 51.53 of
this chapter.
(e) After receiving a license amendment application from the
licensee for the release of an impacted area, the NRC must--
(1) Determine whether the licensee has adequately evaluated the
effect of releasing the property as required by paragraph (a)(1) of
this section;
(2) Determine whether the licensee's classification of any release
areas as non-impacted is adequately justified;
(3) Determine whether the licensee's radiation survey for an
impacted area is adequate; and
(4) If determining that the licensee's submittal is adequate,
approve the licensee's amendment application.
(f) The NRC must publish notice receipt of the release approval
request or license amendment application in the Federal Register and
make the approval request or license amendment application available
for public comment. Before acting on an approval request or license
amendment application submitted in accordance with this section, the
NRC must conduct a public meeting readily accessible to individuals in
the vicinity of the licensee's facility for the purpose of obtaining
public comments on the proposed release of part of the commercial
nuclear plant or site. The NRC must publish a document in the Federal
Register and in a forum, such as local newspapers, which is readily
accessible to individuals in the vicinity of the site, announcing the
date, time, and location of the meeting, along with a brief description
of the purpose of the meeting.
Subpart H--Licenses, Certifications, and Approvals
Sec. 53.1100 Filing of application for licenses, certifications, or
approvals; oath or affirmation.
(a) Serving of applications. (1) Each filing of an application for
a standard design approval, standard design certification, or license
under this part, and any amendments to the applications, must be
submitted to the U.S. Nuclear Regulatory Commission (NRC) under Sec.
53.040, as applicable.
(i) Any person, except one excluded by Sec. 53.1118, may file an
application for a manufacturing license (ML), combined license (COL),
construction permit (CP), or operating license (OL) under this part
with the Director, Office of Nuclear Reactor Regulation.
(ii) Any person who may apply for a CP or for a COL under this
part, may file an application for an early site permit (ESP) with the
Director, Office of Nuclear Reactor Regulation. An application for an
early site permit may be filed notwithstanding the fact that an
application for a CP or a COL has not been filed in connection with the
site for which a permit is sought.
(iii) Any person may submit a proposed standard design for a
commercial nuclear plant to the NRC for its review. The submittal may
consist of either the final design for the entire facility or the final
design for major portions thereof.
(iv) An application for design certification may be filed
notwithstanding the fact that an application for a CP, COL, or ML for
such a facility has not been filed. The application must comply with
Sec. Sec. 2.811 through 2.819 of this chapter.
(2) Each applicant for a construction permit (CP), early site
permit, combined license (COL), or manufacturing license (ML) under
this part must, upon notification by the presiding officer designated
to conduct the public hearing required by the Atomic Energy Act of
1954, as amended, (the Act) update the application and serve the
updated copies of the application or parts of it, eliminating all
superseded information, together with an index of the updated
application, as directed by the presiding officer. Any subsequent
amendment to the application must be served on those served copies of
the application and must be submitted to the NRC as specified in Sec.
53.040, as applicable.
(3) The applicant must make a copy of the updated application
available at the public hearing for the use of any other parties to the
proceeding and must certify that the updated copies of the application
contain the current contents of the application submitted in accordance
with the requirements under this part.
(4) At the time of filing an application, the Commission will make
available at the NRC website, https://www.nrc.gov, a copy of the
application, subsequent amendments, and other records pertinent to the
matter that is
[[Page 15831]]
the subject of the application for public inspection and copying.
(5) The serving of copies required by this section must not occur
until the application has been docketed under Sec. 2.101(a) of this
chapter. Copies must be submitted to the Commission, as specified in
Sec. 53.040, as applicable, to enable the Director, Office of Nuclear
Reactor Regulation to determine whether the application is sufficiently
complete to permit docketing.
(b) Oath or affirmation. Each application for a standard design
approval, standard design certification, or license, including,
whenever appropriate, a CP or early site permit, or amendment of it,
and each amendment of each application must be executed in a signed
original by the applicant or duly authorized officer thereof under oath
or affirmation.
(c)-(d) [Reserved]
(e) Filing fees. Each application for a standard design approval,
standard design certification, or commercial nuclear plant license
under this part, including, whenever appropriate, a CP, COL, operating
license (OL), ML, or early site permit, other than a license exempted
from 10 CFR part 170, must be accompanied by the fee prescribed in 10
CFR part 170. No fee will be required to accompany an application for
renewal, amendment, or termination of a CP, OL, COL, or ML, except as
provided in Sec. 170.21 of this chapter.
(f) Environmental report. An application for a CP, OL, early site
permit, design certification, COL, or ML for a commercial nuclear plant
must be accompanied by an environmental report required under 10 CFR
part 51.
Sec. 53.1101 Requirement for license.
Except as provided in Sec. 53.1120, no person within the United
States may transfer or receive in interstate commerce, manufacture,
produce, transfer, acquire, possess, or use any utilization facility
except as authorized by a license issued by the Commission.
Sec. 53.1103 Combining applications and licenses.
(a) An applicant may combine several applications in one
application for different kinds of licenses under the regulations in
this chapter.
(b) The Commission may combine in a single license the activities
of an applicant which would otherwise be licensed separately.
Sec. 53.1106 Elimination of repetition.
An applicant may incorporate by reference in its application
information contained in previous applications, statements, or reports
filed with the Commission, provided, however, that such references are
clear and specific.
Sec. 53.1109 Contents of applications; general information.
Each application must include, unless otherwise indicated in this
subpart--
(a) Name of applicant;
(b) Address of applicant;
(c) Description of business or occupation of applicant;
(d)(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; and
(iii) Whether it is owned, controlled, or dominated by an alien, a
foreign corporation, or foreign government, and if so, give details; or
(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 (d) with respect to
such principal;
(e) The class and 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;
(f) [Reserved]
(g)(1) Except as provided in paragraph (g)(2) of this section, if
the application is for an OL or COL for a commercial nuclear plant, or
if the application is for an early site permit for a commercial nuclear
plant and contains plans for coping with emergencies under Sec.
53.1146(b)(2)(ii), the applicant must submit the radiological emergency
response plans of State, local, and participating Tribal governmental
entities in the United States that are wholly or partially within the
plume exposure pathway emergency planning zone (EPZ),\1\ and the plans
of State governments wholly or partially within the ingestion pathway
EPZ.\2\ If the application is for an early site permit that, under
Sec. 53.1146(b)(2)(i), proposes major features of the emergency plans
describing the EPZs, then the descriptions of the EPZs must meet the
requirements of this paragraph (g)(1). Generally, the plume exposure
pathway EPZ for a commercial nuclear plant must consist of an area
about 10 miles (16 km) in radius and the ingestion pathway EPZ must
consist of an area about 50 miles (80 km) in radius. The exact size and
configuration of the EPZs surrounding a particular commercial nuclear
plant must be determined in relation to the local emergency response
needs and capabilities as they are affected by such conditions as
demography, topography, land characteristics, access routes, and
jurisdictional boundaries. The size of the EPZs also may be determined
on a case-by-case basis for gas-cooled reactors and for reactors with
an authorized power level less than 250 megawatt thermal. The plans for
the ingestion pathway must focus on such actions as are appropriate to
protect the food ingestion pathway.
(2) Applicants for commercial nuclear plants consisting of either
small modular reactors or non-light-water reactors complying with Sec.
50.160 of this chapter who apply for a CP, an OL, a COL, or an early
site permit under this part must submit as part of the application the
analysis used to determine whether the criteria in Sec.
53.1109(g)(2)(i)(A) and (B) are met and, if they are met, the size of
the plume exposure pathway EPZ.
(i) The plume exposure pathway EPZ is the area within which:
(A) Public dose, as defined in Sec. 20.1003 of this chapter, is
projected to exceed 10 millisieverts (1 rem) total effective dose
equivalent over 96 hours from the release of radioactive materials from
the facility considering accident likelihood and source term, timing of
the accident sequence, and meteorology; and
(B) Pre-determined, prompt protective measures are necessary.
(ii) If the application is for an OL or COL or if the application
is for an early site permit and contains plans for coping with
emergencies under Sec. 53.1146(b)(2)(ii), and if the plume exposure
pathway EPZ extends beyond the site boundary:
(A) The applicant must submit radiological emergency response plans
of State, local, and participating Tribal governmental entities in the
United States that are wholly or partially within the plume exposure
pathway EPZ.
(B) The exact configuration of the plume exposure pathway EPZ
surrounding the facility shall be determined in relation to the local
emergency response needs and capabilities as they are affected by such
conditions as demography, topography, land characteristics, access
routes, and jurisdictional boundaries.
[[Page 15832]]
(iii) If the application is for an early site permit that, under
Sec. 53.1146(b)(2)(i), proposes major features of the emergency plans
and describes the EPZ, and if the EPZ extends beyond the site boundary,
then the exact configuration of the plume exposure pathway EPZ
surrounding the facility must be determined in relation to the local
emergency response needs and capabilities as they are affected by such
conditions as demography, topography, land characteristics, access
routes, and jurisdictional boundaries.
(h) [Reserved]
(i) 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 which
circulate in the area where the proposed activity will be conducted and
which 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; and
(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.
\1\ EPZs are discussed in NUREG-0396, U.S. Environmental
Protection Agency 520/1-78-016, ``Planning Basis for the Development
of State and Local Government Radiological Emergency Response Plans
in Support of Light-Water Nuclear Power Plants,'' December 1978.
\2\ If the State, local, and participating Tribal emergency
response plans have been previously provided to the NRC for
inclusion in the facility docket, the applicant need only provide
the appropriate reference to meet this requirement.
Sec. 53.1112 Environmental conditions.
(a) Each CP, early site permit, and COL under this part may include
conditions to address environmental issues during construction. These
conditions are to be set out in an attachment to the license, which is
incorporated in and made a part of the license. These conditions will
be derived from information contained in the environmental report
submitted pursuant to Sec. 51.50 of this chapter, as analyzed and
evaluated in the NRC record of decision and will identify the
obligations of the licensee in the environmental area, including, as
appropriate, requirements for reporting and keeping records of
environmental data, and any conditions and monitoring requirement for
the protection of the nonaquatic environment.
(b) Each license authorizing operation of a commercial nuclear
plant under this part, and each license for a commercial nuclear plant
for which the certification of permanent cessation of operations
required under Sec. 53.1070 has been submitted may include conditions
to address environmental issues during operation and decommissioning.
These conditions are to be set out in an attachment to the license,
which is incorporated in and made a part of the license. These
conditions will be derived from information contained in the
environmental report or the supplement to the environmental report
submitted under Sec. Sec. 51.50 and 51.53 of this chapter as analyzed
and evaluated in the NRC record of decision, and will identify the
obligations of the licensee in the environmental area, including, as
appropriate, requirements for reporting and keeping records of
environmental data and any conditions and monitoring requirement for
the protection of the nonaquatic environment.
Sec. 53.1115 Agreement limiting access to classified information.
As part of its application 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 under this part, or
before the Commission has adopted a final standard design certification
rule under this part, 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. 53.1118 Ineligibility of certain applicants.
Any person who is a citizen, national, or agent of a foreign
country, or any corporation, or other entity which the Commission knows
or has reason to believe is owned, controlled, or dominated by an
alien, a foreign corporation, or a foreign government, will be
ineligible to apply for and obtain a license unless--
(a) The Commission determines that issuance of the applicable
license to the entity is not inimical to the common defense and
security or the health and safety of the public; and
(b) The entity is an alien, corporation, or other entity that is
owned, controlled, or dominated by the government of, a corporation
that is incorporated in, or an alien who is a citizen or national of
Australia, Austria, Belgium, Canada, Chile, Colombia, Costa Rica,
Czechia, Denmark, Estonia, Finland, France, Germany, Greece, Hungary,
Iceland, India, Ireland, Israel, Italy, Japan, Korea, Latvia,
Lithuania, Luxembourg, Mexico, Netherlands, New Zealand, Norway,
Poland, Portugal, Slovak Republic, Slovenia, Spain, Sweden,
Switzerland, or the United Kingdom.
Sec. 53.1120 Exceptions and exemptions from licensing requirements.
Nothing in this part must be deemed to require a license for--
(a) The manufacture, production, or acquisition by the Department
of Defense of any utilization facility authorized pursuant to section
91 of the Act or the use of such facility by the Department of Defense
or by a person under contract with and for the account of the
Department of Defense;
(b) Except to the extent that the Department of Energy facilities
of the types subject to licensing pursuant to section 202 of the Energy
Reorganization Act of 1974, as amended, are involved--
(1)(i) The processing, fabrication or refining of special nuclear
material (SNM) or the separation of SNM, or the separation of SNM from
other substances by a prime contractor of the Department of Energy
under a prime contract for--
(A) The performance of work for the Department of Energy at a
United States Government-owned or controlled site;
(B) Research in, or development, manufacture, storage, testing or
transportation of, atomic weapons or components thereof; or
(C) The use or operation of a utilization facility in a United
States owned vehicle or vessel; or
(ii) The processing, fabrication or refining of SNM of the
separation of SNM, or the separation of SNM from other substances by a
prime contractor or subcontractor of the Commission or the Department
of Energy under a prime contract or subcontract when the Commission
determines that the exemption of the prime contractor or subcontractor
is authorized by law; and that, under the terms of the contract or
subcontract, there is adequate assurance that the work thereunder can
be accomplished without undue risk to the public health and safety; or
(2)(i) The construction or operation of a utilization facility for
the Department of Energy at a United States Government-owned or
controlled site, including the transportation of the utilization
facility to or from such site
[[Page 15833]]
and the performance of contract services during temporary interruptions
of such transportation; or the construction or operation of a
utilization facility for the Department of Energy in the performance of
research in, or development, manufacture, storage, testing, or
transportation of, atomic weapons or components thereof; or the use or
operation of a utilization facility for the Department of Energy in a
United States Government-owned vehicle or vessel; provided that such
activities are conducted by a prime contractor of the Department of
Energy under a prime contract with the Department of Energy; or
(ii) The construction or operation of a utilization facility by a
prime contractor or subcontractor of the Commission or the Department
of Energy under his or her prime contract or subcontract when the
Commission determines that the exemption of the prime contractor or
subcontractor is authorized by law; and that, under the terms of the
contract or subcontract, there is adequate assurance that the work
thereunder can be accomplished without undue risk to the public health
and safety; or
(c) The transportation or possession of any utilization facility by
a common or contract carrier or warehouse employee in the regular
course of carriage for another or storage incident thereto.
Sec. 53.1121 Public inspection of applications.
Applications and documents submitted to the Commission in
connection with applications may be made available for public
inspection under the provisions of part 2 of this chapter.
Sec. 53.1124 Relationship between sections.
(a) Limited work authorization. An application for a limited work
authorization (LWA) under this part may be submitted as part of an
application for an early site permit, CP, or COL under this part as
required in Sec. 53.1130(a)(2).
(b) Early site permit. (1) A holder of an early site permit may
request an LWA.
(2) An application for a CP or COL under this part may, but need
not, reference an early site permit.
(c) Standard design approval. An application for a standard design
approval under this part may, but need not, reference an OL or custom
COL under this part that is essentially the same as the information
supporting the standard design for which approval is being requested.
(d) Standard design certification. An application for a standard
design certification under this part may, but need not, reference an OL
or custom COL under this part that is essentially the same as the
standard design for which certification is being requested.
(e) Manufacturing license. (1) A manufactured reactor or portions
thereof as defined in an ML issued under this part may be either
transported to and installed at a site for which a COL or CP under this
part has been issued or exported in accordance with part 110.
(2) An ML applicant under this part may reference a standard design
certification or a standard design approval under this part in its
application.
(f) Construction permit. An application for a CP may, but need not,
reference a standard design certification, standard design approval, or
ML issued under this part, respectively, and may also reference an
early site permit issued under this part. In the absence of a
demonstration that an entity other than the one originally sponsoring a
standard design certification is qualified to supply a design, the
Commission will entertain an application for a CP that references a
standard design certification issued under this part only if the entity
that sponsored the certification supplies the design for the
applicant's use.
(g) Operating license. (1) An application for an OL under this part
may, but need not, reference an early site permit, standard design
certification, or standard design approval issued under this part. In
the absence of a demonstration that an entity other than the one
originally sponsoring a standard design certification is qualified to
supply a design, the Commission will entertain an application for an OL
that references a standard design certification issued under this part
only if the entity that sponsored the certification supplies the design
for the applicant's use.
(2) The holder of a CP must, at the time of submission of the Final
Safety Analysis Report (FSAR), file an application for an OL.
(h) Combined licenses. An application for a COL under this part
may, but need not, reference an early site permit, standard design
certification, standard design approval, or ML issued under this part.
In the absence of a demonstration that an entity other than the one
originally sponsoring and obtaining a standard design certification is
qualified to supply a design, the Commission will entertain an
application for a COL that references a standard design certification
issued under this part only if the entity that sponsored the
certification supplies the design for the applicant's use.
Sec. 53.1130 Limited work authorizations.
(a) Request for limited work authorization. (1) Any person to whom
the Commission may otherwise issue either a license or permit related
to a commercial nuclear plant may request an LWA allowing that person
to perform the driving of piles, subsurface preparation, placement of
backfill, concrete, or permanent retaining walls within an excavation,
and installation of the foundation, including placement of concrete,
any of which are for a structure, system, or component (SSC) of the
facility for which either a CP or COL is otherwise required under Sec.
53.610.
(2) An application for an LWA may be submitted as part of a
complete application for a CP or COL in accordance with Sec.
2.101(a)(1) through (a)(5) of this chapter, or as a partial application
in accordance with Sec. 2.101(a)(9) of this chapter. An application
for an LWA by the holder of an early site permit must be submitted as a
complete application in accordance with Sec. 2.101(a)(1) through
(a)(4) of this chapter.
(3) The application must include--
(i) A Safety Analysis Report required by Sec. 53.1146, Sec.
53.1309 or Sec. 53.1416, as applicable, a description of the
activities requested to be performed, and the design and construction
information otherwise required by the Commission's rules and
regulations to be submitted for a CP or COL under this part but limited
to those portions of the facility that are within the scope of the LWA.
The Safety Analysis Report must demonstrate that activities conducted
under the LWA will be conducted in compliance with the technically
relevant Commission requirements in 10 CFR chapter I applicable to the
design of those portions of the facility within the scope of the LWA;
(ii) An environmental report in accordance with Sec. 51.49 of this
chapter; and
(iii) A plan for redress of activities performed under the LWA,
should limited work activities be terminated by the holder, or the LWA
be revoked by the NRC or upon effectiveness of the Commission's final
decision denying the associated CP or COL application, as applicable.
(b) Issuance of limited work authorization. (1) The Director,
Office of Nuclear Reactor Regulation may issue an LWA only after--
(i) The NRC staff issues the final environmental impact statement
for the LWA under part 51 of this chapter;
[[Page 15834]]
(ii) The presiding officer makes the finding in Sec. 51.105(c) or
Sec. 51.107(d) of this chapter, as applicable;
(iii) The Director determines that the applicable standards and
requirements of the Act, and the Commission's regulations applicable to
the activities to be conducted under the LWA, have been met, the
applicant is technically qualified to engage in the activities
authorized, and that issuance of the LWA will provide reasonable
assurance of adequate protection to public health and safety and will
not be inimical to the common defense and security; and
(iv) The presiding officer finds that there are no unresolved
safety issues relating to the activities to be conducted under the LWA
that would constitute good cause for withholding the authorization.
(2) Each LWA will specify the activities that the holder is
authorized to perform.
(c) Effect of limited work authorization. Any activities undertaken
under an LWA are entirely at the risk of the applicant and, except as
to the matters determined under paragraph (b)(1) of this section, the
issuance of the LWA has no bearing on the issuance of a CP or COL with
respect to the requirements of the Act and rules, regulations, or
orders issued under the Act. The environmental impact statement for a
CP or COL application for which an LWA was previously issued will not
address, and the presiding officer will not consider, the sunk costs of
the holder of the LWA in determining the proposed action (i.e.,
issuance of the CP or COL).
(d) Implementation of redress plan. If construction is terminated
by the holder, the underlying application is withdrawn by the applicant
or denied by the NRC, or the LWA is revoked by the NRC, then the holder
must begin implementation of the redress plan in a reasonable time. The
holder must also complete the redress of the site no later than 18
months after termination of construction, revocation of the LWA, or
upon effectiveness of the Commission's final decision denying the
associated CP application or the associated COL application, as
applicable.
Sec. 53.1140 Early site permits.
Sections 53.1140 through 53.1188 set out the requirements and
procedures applicable to Commission issuance of an early site permit
under this part for approval of a site for a commercial nuclear plant
separate from the filing of an application for a CP or COL for the
facility.
Sec. 53.1144 Contents of applications for early site permits; general
information.
The application must contain all of the information required by
Sec. 53.1109(a) through (d) and (j).
Sec. 53.1146 Contents of applications for early site permits;
technical information.
(a) The application must contain--
(1) A Site Safety Analysis Report that must include the following:
(i) The specific number, type, and thermal power level of the
facilities, or range of possible facilities, for which the site may be
used;
(ii) The anticipated maximum levels of radiological and thermal
effluents each facility will produce;
(iii) The type of cooling systems, including intakes and outflows,
where appropriate, that may be associated with each facility;
(iv) The boundaries of the site;
(v) The proposed general location of each facility on the site;
(vi) The external hazards and site characteristics required by this
part;
(vii) The location and description of any nearby industrial,
military, or transportation facilities and routes;
(viii) The existing and projected future population profile of the
area surrounding the site;
(ix) A description and assessment of the site on which a facility
is to be located. The assessment must address the requirements of
subpart D of this part;
(x) Information demonstrating that site characteristics are such
that adequate security plans and measures can be developed; and
(xi) A description of the quality assurance program (QAP) required
by appendix B to part 50 of this chapter applied to site-related
activities for the future design, fabrication, construction, and
testing of the SSCs of a facility or facilities that may be constructed
on the site.
(2) A complete environmental report as required by Sec. 51.50(b)
of this chapter.
(b)(1) The Site Safety Analysis Report must identify physical
characteristics of the proposed site, such as egress limitations from
the area surrounding the site, that could pose a significant impediment
to the development of emergency plans. If physical characteristics are
identified that could pose a significant impediment to the development
of emergency plans, the application must identify measures that would,
when implemented, mitigate or eliminate the significant impediment.
(2) The Site Safety Analysis Report may also--
(i) Propose major features of the emergency plans, under either
Sec. 50.160 or the requirements in appendix E to part 50 and Sec.
50.47(b) of this chapter, as applicable, such as the exact size and
configuration of the EPZs, for review and approval by the NRC, in
consultation with the Federal Emergency Management Agency (FEMA), as
applicable, in the absence of complete and integrated emergency plans;
or
(ii) Propose complete and integrated emergency plans for review and
approval by the NRC, in consultation with FEMA, as applicable, in
accordance with either Sec. 50.160 or the requirements in appendix E
to part 50 and Sec. 50.47(b) of this chapter. To the extent approval
of emergency plans is sought, the application must contain the
information required by Sec. 53.1109(g).
(3) Emergency plans submitted under paragraph (b)(2)(ii) of this
section must include the proposed inspections, tests, and analyses that
the holder of a COL referencing the early site permit must perform, and
the acceptance criteria that are necessary and sufficient to provide
reasonable assurance that, if the inspections, tests, and analyses are
performed and the acceptance criteria met, the facility has been
constructed and will be operated in conformity with the emergency
plans, the provisions of the Act, and the Commission's rules and
regulations. Major features of an emergency plan submitted under
paragraph (b)(2)(i) of this section may include proposed inspections,
tests, analyses, and acceptance criteria (ITAAC).
(4) Under paragraphs (b)(1) and (b)(2)(i) of this section, the Site
Safety Analysis Report must include, where appropriate, a description
of contacts and arrangements made with Federal, State, participating
Tribal, and local governmental agencies with emergency planning
responsibilities. The Site Safety Analysis Report must contain any
certifications that have been obtained. If these certifications, where
appropriate, cannot be obtained, the Site Safety Analysis Report must
contain information, including a utility plan, sufficient to show that
the proposed plans provide reasonable assurance that adequate
protective measures can and will be taken in the event of a
radiological emergency at the site. Under the option set forth in
paragraph (b)(2)(ii) of this section, the applicant must make good
faith efforts, where appropriate, to obtain from the same governmental
agencies certifications that--
(i) The proposed emergency plans are practicable;
(ii) These agencies are committed to participating in any further
[[Page 15835]]
development of the plans, including any required field demonstrations;
and
(iii) That these agencies are committed to executing their
responsibilities under the plans in the event of an emergency.
(c) An applicant may request that an LWA under Sec. 53.1130 be
issued in conjunction with the early site permit. The application must
include the information otherwise required by Sec. 53.1130.
(d) Each applicant for an early site permit under this part must
protect safeguards information against unauthorized disclosure in
accordance with the requirements in Sec. Sec. 73.21 and 73.22 of this
chapter, as applicable.
Sec. 53.1149 Review of applications.
(a) Standards for review of applications. Applications filed under
this part will be reviewed according to the applicable standards set
out in this part. In addition, the Commission must prepare an
environmental impact statement during review of the application, under
the applicable provisions of 10 CFR part 51. The Commission must
determine, after consultation with FEMA, as applicable, whether the
information required of the applicant by Sec. 53.1146(b)(1) shows that
there is no significant impediment to the development of emergency
plans that cannot be mitigated or eliminated by measures proposed by
the applicant, whether any major features of emergency plans submitted
by the applicant under Sec. 53.1146(b)(2)(i) are acceptable under
either Sec. 50.160 or appendix E to part 50 and Sec. 50.47(b) of this
chapter, and whether any emergency plans submitted by the applicant
under Sec. 53.1146(b)(2)(ii) provide reasonable assurance that
adequate protective measures can and will be taken in the event of a
radiological emergency.
(b) Administrative review of applications; hearings. An early site
permit application is subject to all procedural requirements 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 hearing in Sec. 2.104(a) and (d) of this chapter, provided
that the designated sections may not be construed to require that the
environmental report, or draft or final environmental impact statement
includes an assessment of the benefits of construction and operation of
the reactor or reactors, or an analysis of alternative energy sources.
The presiding officer in an early site permit hearing must not admit
contentions proffered by any party concerning an assessment of the
benefits of construction and operation of the reactor or reactors, or
an analysis of alternative energy sources if those issues were not
addressed by the applicant in the early site permit application. All
hearings conducted on applications for early site permits filed under
this part are governed by the procedures contained in subparts C, G, L,
and N of 10 CFR part 2, as applicable.
Sec. 53.1155 Referral to the Advisory Committee on Reactor
Safeguards.
The Commission must refer a copy of the application for an early
site permit to the Advisory Committee on Reactor Safeguards (ACRS). The
ACRS must report on those portions of the application which concern
safety.
Sec. 53.1158 Issuance of early site permit.
(a) After conducting a hearing under Sec. 53.1149(b) and receiving
the report to be submitted by the ACRS under Sec. 53.1155, the
Commission may issue an early site permit, in the form the Commission
deems appropriate, if the Commission finds that--
(1) An application for an early site permit demonstrates compliance
with the applicable standards and requirements of the Act and the
Commission's regulations;
(2) Notifications, if any, to other agencies or bodies have been
duly made;
(3) There is reasonable assurance that the site is in conformity
with the provisions of the Act and the Commission's regulations;
(4) The applicant is technically qualified to engage in any
activities authorized;
(5) The proposed ITAAC, including any on emergency planning, are
necessary and sufficient, within the scope of the early site permit, to
provide reasonable assurance that the facility has been constructed and
will be operated in conformity with the license, the provisions of the
Act, and the Commission's regulations;
(6) Issuance of the permit will not be inimical to the common
defense and security or to the health and safety of the public;
(7) Any significant adverse environmental impact resulting from
activities requested under Sec. 53.1146(c) can be redressed; and
(8) The findings required by 10 CFR part 51 have been made.
(b) The early site permit must specify the site characteristics,
design parameters, and terms and conditions of the early site permit
the Commission deems appropriate. Before issuance of either a CP or COL
referencing an early site permit, the Commission must find that any
relevant terms and conditions of the early site permit have been met.
Any terms or conditions of the early site permit that could not be met
by the time of issuance of the CP or COL, must be set forth as terms or
conditions of the CP or COL.
(c) The early site permit must specify those Sec. 53.1130(b)
activities requested under Sec. 53.1146(c) that the permit holder is
authorized to perform.
Sec. 53.1161 Extent of activities permitted.
If the activities authorized by Sec. 53.1158(c) are performed and
the site is not referenced in an application for a CP or a COL issued
under this part while the permit remains valid, then the early site
permit remains in effect solely for the purpose of site redress, and
the holder of the permit must redress the site under the terms of the
site redress plan required by Sec. 53.1146(c). If, before redress is
complete, a use not envisaged in the redress plan is found for the site
or parts thereof, the holder of the permit must carry out the redress
plan to the greatest extent possible consistent with the alternate use.
Sec. 53.1164 Duration of permit.
(a) Except as provided in paragraph (b) of this section, an early
site permit issued under this subpart may be valid for not less than
10, nor more than 20 years from the date of issuance.
(b) An early site permit continues to be valid beyond the date of
expiration in any proceeding on a CP application or a COL application
that references the early site permit and is docketed before the date
of expiration of the early site permit, or, if a timely application for
renewal of the permit has been docketed, before the Commission has
determined whether to renew the permit.
(c) An applicant for a CP or COL may, at its own risk, reference in
its application a site for which an early site permit application has
been docketed but not granted.
(d) Upon issuance of a CP or COL, a referenced early site permit is
subsumed, to the extent referenced, into the CP or COL.
Sec. 53.1167 Limited work authorization after issuance of early site
permit.
A holder of an early site permit may request an LWA under Sec.
53.1130.
Sec. 53.1170 Transfer of early site permit.
An application to transfer an early site permit will be processed
under Sec. 53.1570.
[[Page 15836]]
Sec. 53.1173 Application for renewal.
(a) Not less than 12, nor more than 36 months before the expiration
date stated in the early site permit, or any later renewal period, the
permit holder may apply for a renewal of the permit. An application for
renewal must contain all information necessary to bring up to date the
information and data contained in the previous application.
(b) Any person whose interests may be affected by renewal of the
permit 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 under
Sec. 2.309 of this chapter.
(c) An early site permit, either original or renewed, for which a
timely application for renewal has been filed, remains in effect until
the Commission has determined whether to renew the permit. If the
permit is not renewed, it continues to be valid in certain proceedings
in accordance with the provisions of Sec. 53.1164(b).
Sec. 53.1176 Criteria for renewal.
(a) The Commission must grant the renewal if it determines that--
(1) The site complies with the Act, the Commission's regulations,
and orders applicable and in effect at the time the site permit was
originally issued; and
(2) Any new requirements the Commission may wish to impose--
(i) Are necessary for adequate protection to public health and
safety or common defense and security;
(ii) Are necessary for compliance with the Commission's
regulations, and orders applicable and in effect at the time the site
permit was originally issued; or
(iii) Would provide 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.
(b) A denial of renewal under the provisions of Sec. 53.1176(a)
does not bar the permit holder or another applicant from filing a new
application for the site which proposes changes to the site or the way
that it is used to correct the deficiencies cited in the denial of the
renewal.
Sec. 53.1179 Duration of renewal.
Each renewal of an early site permit may be for not less than 10,
nor more than 20 years, plus any remaining years on the early site
permit then in effect before renewal.
Sec. 53.1182 Use of site for other purposes.
A site for which an early site permit has been issued under this
part may be used for purposes other than those described in the permit,
including the location of other types of energy facilities. The permit
holder must inform the Director, Office of Nuclear Reactor Regulation
(Director), of any significant uses for the site which have not been
approved in the early site permit. The information about the activities
must be given to the Director at least 30 days in advance of any actual
construction or site modification for the activities. The information
provided could be the basis for imposing new requirements on the
permit, under the provisions of Sec. 53.1188. If the permit holder
informs the Director that the holder no longer intends to use the site
for a commercial nuclear plant, the Director may terminate the permit.
Sec. 53.1188 Finality of early site permit determinations.
(a) Commission finality. (1) While an early site permit is in
effect under Sec. 53.1164 or Sec. 53.1179, the Commission may not
change or impose new site characteristics, design parameters, or terms
and conditions, including emergency planning requirements, on the early
site permit unless the Commission--
(i) Determines that a modification is necessary to bring the permit
or the site into compliance with the Commission's regulations and
orders applicable and in effect at the time the permit was issued;
(ii) Determines the modification is necessary to assure adequate
protection of the public health and safety or the common defense and
security;
(iii) Determines that a modification is necessary based on an
update under paragraph (b) of this section; or
(iv) Issues a variance requested under paragraph (d) of this
section.
(2) In making the findings required for issuance of a CP, COL, or
OL, or the findings required by Sec. 53.1452(g), or in any enforcement
hearing other than one initiated by the Commission under paragraph
(a)(1) of this section, if the application for the CP, COL, or OL
references an early site permit, the Commission must treat as resolved
those matters resolved in the proceeding on the application for
issuance or renewal of the early site permit, except as provided for in
paragraphs (b), (c), and (d) of this section.
(i) If the Commission grants a CP application that references an
early site permit and an application for an OL references the CP, the
Commission must treat as resolved those matters resolved in the
proceeding for the issuance or renewal of the early site permit, except
as provided for in paragraphs (b), (c), and (d) of this section.
(ii) If the early site permit approved an emergency plan (or major
features thereof) that is in use by a licensee of a commercial nuclear
plant, the Commission must treat as resolved changes to the early site
permit emergency plan (or major features thereof) that are identical to
changes made to the licensee's emergency plans under Sec. 53.1565
occurring after issuance of the early site permit.
(iii) If the early site permit approved an emergency plan (or major
features thereof) that is not in use by a licensee of a commercial
nuclear plant, the Commission must treat as resolved changes that are
equivalent to those that could be made under Sec. 53.1565 without
prior NRC approval had the emergency plan been in use by a licensee.
(b) Updating of early site permit-emergency preparedness. An
applicant for a CP, OL, or COL who has filed an application referencing
an early site permit issued under this subpart must update the
emergency preparedness information that was provided under Sec.
53.1146(b) and discuss whether the updated information materially
changes the bases for compliance with applicable NRC requirements.
(c) Hearings and petitions. (1) In any proceeding for the issuance
of a CP, OL, or COL referencing an early site permit, contentions on
the following matters may be litigated in the same manner as other
issues material to the proceeding:
(i) The nuclear reactor proposed to be built does not fit within
one or more of the site characteristics or design parameters included
in the early site permit;
(ii) One or more of the terms and conditions of the early site
permit have not been met;
(iii) A variance requested under paragraph (d) of this section is
unwarranted or should be modified;
(iv) New or additional information is provided in the application
that substantially alters the bases for a previous NRC conclusion or
constitutes a sufficient basis for the Commission to modify or impose
new terms and conditions related to emergency preparedness; or
(v) Any significant environmental issue that was not resolved in
the early site permit proceeding, or any issue involving the impacts of
construction and operation of the facility that was resolved in the
early site permit proceeding for which significant new information has
been identified.
(2) Any person may file a petition requesting that the site
characteristics,
[[Page 15837]]
design parameters, or terms and conditions of the early site permit be
modified, or that the permit be suspended or revoked. The petition will
be considered under Sec. 2.206 of this chapter. Before construction
commences, the Commission must consider the petition and determine
whether any immediate action is required. If the petition is granted,
an appropriate order will be issued. Construction under the CP or COL
will not be affected by the granting of the petition unless the order
is made immediately effective. Any change required by the Commission in
response to the petition must demonstrate compliance with the
requirements of paragraph (a)(1) of this section.
(d) Variances. An applicant for a CP, OL, or COL referencing an
early site permit may include in its application a request for a
variance from one or more site characteristics, design parameters, or
terms and conditions of the early site permit, or from the Site Safety
Analysis Report. In determining whether to grant the variance, the
Commission must apply the same technically relevant criteria applicable
to the application for the original or renewed early site permit. Once
a CP or COL referencing an early site permit is issued, variances from
the early site permit will not be granted for that CP or COL.
(e) Early site permit amendment. The holder of an early site permit
may not make changes to the early site permit or the Site Safety
Analysis Report without prior Commission approval. The request for a
change to the early site permit must be in the form of an application
for a license amendment and must demonstrate compliance with the
requirements of Sec. Sec. 53.1510 and 53.1520.
Sec. 53.1200 Standard design approvals.
Sections 53.1200 through 53.1221 set out procedures for the filing,
NRC staff review, and referral to the ACRS of standard designs, or
major portions thereof, for a commercial nuclear plant under this part.
Sec. 53.1206 Contents of applications for standard design approvals;
general information.
The application must contain all of the information required by
Sec. 53.1109(a) through (c) and (j).
Sec. 53.1209 Contents of applications for standard design approvals;
technical information.
(a) Major portion of a standard design. 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 scope of the application for
which approval is sought must include all functional design criteria
necessary to demonstrate compliance with the safety criteria in
Sec. Sec. 53.210, 53.220 and 53.450(e), 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 CP, design
certification, ML, or COL application.
(b) Final Safety Analysis Report. The application must contain an
FSAR that describes the facility and the limits on its operation,
presents a safety analysis of the SSCs and of the facility, or major
portions thereof, for which the applicant seeks design approval, and
must include the following information:
(1) Site parameters. The site parameters postulated for the design
under this part, 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.
(2) Design information. Except as specified in this paragraph (b),
an application for a standard design approval for a commercial nuclear
plant must include the design information equivalent to that required
for a standard design certification under Sec. 53.1239(a)(2) through
(27) for those portions of a commercial nuclear plant included in the
standard design approval.
Sec. 53.1210 Contents of applications for standard design approvals;
other application content.
(a) In addition to the FSAR, the application must also include the
following:
(1) Availability controls (if not included in the FSAR). A
description of the controls on plant operations, including availability
controls, to provide reasonable confidence that the configurations and
special treatments for safety-related (SR) SSCs and non-safety-related
but safety-significant (NSRSS) SSCs provide the capabilities and
reliabilities required to demonstrate compliance with the safety
criteria of Sec. 53.220.
(2) Safeguards information. 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.
(b) If there are SSCs of the plant which required research and
development to confirm the adequacy of their design, provide a report
in the application which documents the resolution of any safety
questions associated with such SSCs.
(c) A description of how the performance of each design feature has
been demonstrated capable of fulfilling functional design criteria
considering interdependent effects through either analysis, appropriate
test programs, prototype testing, operating experience, or a
combination thereof, in accordance with Sec. 53.440(a).
Sec. 53.1212 Standards for review of applications.
Applications filed under this part will be reviewed under the
standards set out in 10 CFR parts 20, 53, and 73.
Sec. 53.1215 Referral to the Advisory Committee on Reactor
Safeguards.
The Commission must refer a copy of the application to the ACRS.
The ACRS must report on those portions of the application which concern
safety.
Sec. 53.1218 Staff approval of design.
(a) Upon completion of its review of a submittal under Sec. Sec.
53.1200 through 53.1221 and receipt of a report by the ACRS under Sec.
53.1215, the NRC staff must publish a determination in the Federal
Register as to whether or not 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.
(b) A standard design approval issued under this section is valid
for 15 years from the date of issuance and may not be renewed. A design
approval continues to be valid beyond the date of expiration in any
proceeding on an application for a CP, OL, COL, or ML under this part
that references the design approval and is docketed before the date of
expiration of the design approval.
Sec. 53.1221 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 reviews of any standard design
certification or individual facility license application
[[Page 15838]]
under this part that incorporates by reference a standard design
approved under this part 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 permit or license, or 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,
information requests to the holder of a standard design approval must
be evaluated 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
evaluation performed by the NRC staff must be in accordance with Sec.
53.1580 and must be approved by the Executive Director for Operations
or authorized designee before issuance of the request.
(d) The Commission will require, before granting a CP, COL, OL, or
ML that references a standard design approval, that information
normally contained in 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. 53.1230 Standard design certifications.
Sections 53.1230 through 53.1263 set forth the requirements and
procedures applicable to the Commission's issuance of rules granting
standard design certifications for commercial nuclear plants under this
part separate from the filing of an application for a CP or COL for
such a facility.
Sec. 53.1236 Contents of applications for standard design
certifications; general information.
The application must contain all of the information required by
Sec. 53.1109(a) through (c) and (j).
Sec. 53.1239 Contents of applications for standard design
certifications; technical information.
The application must contain a level of design information
sufficient to enable the Commission to judge the applicant's proposed
means of assuring that construction conforms to the design and to reach
a final conclusion on all safety questions associated with the design
before the certification is granted. The information submitted for a
design certification must include performance requirements and design
information sufficiently detailed to permit the preparation of
acceptance and inspection requirements by the NRC. The Commission will
require, before design certification, that information normally
contained in 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.
(a) Final Safety Analysis Report. The application must contain an
FSAR that describes the facility and the limits on its operation, and
presents a safety analysis of the SSCs, and must include the following
information:
(1) Site parameters. The site parameters postulated for the design
under this part, 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.
(2) Plant description and safety functions--(i) General plant
description. A general description of the commercial nuclear plant
including reactor type, the intended use of the reactor, nuclear design
(e.g., neutron spectrum, reactor control, multi-unit reactor control),
overall layout of the plant including significant plant features and
SSCs, maximum power level and the nature and inventory of radioactive
materials.
(ii) Safety functions. A description of the primary and additional
safety functions required under Sec. 53.230 and a summary of how each
safety function is satisfied.
(3) Design features and functional design criteria--licensing-basis
events. (i) A description of the design features required by Sec.
53.400 and the functional design criteria required by Sec. Sec. 53.410
and 53.420 that, when combined with corresponding human actions and
programmatic controls, demonstrate that the plant will demonstrate
compliance with the safety criteria defined in Sec. 53.210 and
established in accordance with Sec. 53.220 during licensing-basis
events (LBEs).
(ii) A description of how design features demonstrate compliance
with the requirements of Sec. 53.440(a) through (i) and (k) through
(m).
(4) Design features supporting normal operations. A description of
the design features required by Sec. 53.425 to support the holder of
an OL or COL complying with Sec. 53.260 during normal operations.
(5) [Reserved]
(6) Earthquake engineering. The information necessary to
demonstrate that the commercial nuclear plant complies with the
earthquake engineering criteria in Sec. 53.480.
(7) Programmatic controls and interfaces. (i) A description of the
corresponding programmatic controls and interfaces necessary to achieve
and maintain the reliability and capability of SSCs relied upon to
demonstrate compliance with the functional design criteria required by
Sec. Sec. 53.410 and 53.420 and the safety criteria in Sec. Sec.
53.210 and 53.220 and necessary to maintain consistency with analyses
required by Sec. 53.450.
(ii) For an application for a multi-unit commercial nuclear plant,
the programmatic controls and interfaces must also be described for
different modular configurations, as required by Sec. 53.440(i),
including any restrictions that will be necessary during the
construction and startup of any given unit to ensure the safe operation
of the overall commercial nuclear plant to be licensed under this part.
(8) Programmatic controls for normal operations. A description of
how programmatic controls, including monitoring programs, would provide
assurance that design features and procedures will enable the holder of
an OL or COL to comply with Sec. 53.260.
(9) Design features supporting the protection of plant workers. A
description of the design features required by Sec. 53.430 to support
the holder of an OL or COL complying with Sec. 53.270.
(10) Programmatic controls for protection of plant workers. A
description of how programmatic controls, including monitoring
programs, would provide assurance that design features and procedures
will enable the holder of an OL or COL to comply with Sec. 53.270.
(11) Codes and standards. A description of generally accepted
consensus codes and standards used to design the design features, as
required by Sec. 53.440(b).
(12) Materials. A description of the materials used for SR and
NSRSS SSCs and a description of the qualification of
[[Page 15839]]
these materials for their service conditions over the plant lifetime,
as required by Sec. 53.440(c).
(13) Integrity assessment program. A description of a design
integrity assessment program that addresses the elements described in
Sec. 53.440(d).
(14) [Reserved]
(15) Criticality. Information demonstrating how the applicant will
comply with requirements for criticality accidents in Sec. 53.440(m).
(16) Multi-unit plants. For an application for standard design
certification of a multi-unit commercial nuclear plant, the possible
operating configurations of the reactor units, including common
systems, interface requirements, and system interactions, as required
by Sec. 53.440(i).
(17) SSC classification. (i) The classification of SSCs according
to their safety significance under Sec. 53.460(a).
(ii) For SR and NSRSS SSCs, the conditions under which they must
perform the safety functions required by Sec. 53.230, including
environmental conditions.
(18) Probabilistic risk assessment or other systematic risk
evaluations (SREs). A description of the probabilistic risk assessment
(PRA), other SREs, or a combination thereof required by Sec. 53.450(a)
and associated results.
(19) Analyses. A description of the analyses performed under Sec.
53.450(b) through (g) that includes the following information:
(i) A description of the analysis of LBEs and its results, as
described in Sec. 53.240. This analysis description must--
(A) Address the elements in Sec. 53.450(e) and (f); and
(B) Under Sec. 53.460(c)--
(1) Describe any human actions that are necessary to prevent or
mitigate LBEs;
(2) Describe how those human actions are capable of being reliably
performed under the postulated environmental conditions present; and
(3) Describe how those human actions would be addressed by programs
established under subpart F of this part.
(ii)(A) A description of how SSCs relied on to meet the safety
criteria defined in Sec. 53.210 are protected against or designed to
withstand the effects of external hazards under Sec. 53.510.
(B) The information necessary to demonstrate that the commercial
nuclear plant complies with the earthquake engineering criteria in
Sec. 53.480.
(iii) A description of the defense-in-depth measures required by
Sec. 53.250.
(iv) A description of all plant operating states where there is the
potential for the uncontrolled release of radioactive material to the
environment, as required by Sec. 53.450(b)(4).
(v) A description of the events that challenge plant control and
safety systems whose failure could lead to an undesirable end state
and/or radioactive material release, as required by Sec. 53.450(b)(5).
(vi) A description of the analytical codes used in modeling plant
behavior in analyses of LBEs and how these codes are qualified for the
range of conditions for which they were used, as required by Sec.
53.450(d).
(vii) A description of the results of other analyses required by
Sec. 53.450(g).
(20) Special treatments. A description of special treatments
established as required by Sec. 53.460.
(21) [Reserved]
(22) Quality assurance. A description of the QAP applied to the
design of the SSCs of the commercial nuclear plant, as required by
Sec. 53.460(b). The description of the QAP for a commercial nuclear
plant must include a discussion of how the applicable requirements of
appendix B to part 50 of this chapter were satisfied.
(23) Design features and controls to address the minimization of
contamination. The information required by Sec. 20.1406 of this
chapter.
(24) Interface requirements. (i) A description analysis, and
evaluation of the interfaces between the standard design and the
balance of the commercial nuclear plant that may impact the ability of
the plant to demonstrate compliance with the functional design criteria
or the safety criteria of subparts B and C of this part.
(ii) Confirmation that interface requirements are verifiable
through inspections, testing, or analysis. These requirements must be
sufficiently detailed to allow for completion of the final safety
analysis by license applicants that reference the certified design
under this subpart. The method to be used for verification of interface
requirements must be included as part of the proposed ITAAC required by
Sec. 53.1241(a)(3).
(iii) A representative conceptual design for those portions of the
plant for which the application does not seek certification to aid the
NRC in its review of the FSAR and to permit assessment of the adequacy
of the interface requirements under paragraph (a)(24)(i) of this
section.
(25) Technical qualifications. A description of the technical
qualifications of the applicant to engage in the proposed activities in
accordance with the regulations in this chapter.
(26) Technical specifications. Proposed technical specifications
prepared under Sec. 53.710(a) for those areas addressed by the design
certification.
(27) Role of personnel. Information to address the following for
the role of personnel in ensuring safe operations:
(i) A description of how the human factors engineering design
requirements of Sec. 53.440(n)(1) are addressed;
(ii) A description of how the human system interface design
requirements of Sec. 53.440(n)(2) are addressed;
(iii) A concept of operations that is of sufficient scope and
detail to address the requirements of Sec. 53.440(n)(3);
(iv) A functional requirements analysis and function allocation
that is of sufficient scope and detail to address the requirements of
Sec. 53.440(n)(4).
(b) [Reserved]
Sec. 53.1241 Contents of applications for standard design
certifications; other application content.
(a) In addition to the FSAR, the application must also include the
following:
(1) Environmental report. An environmental report as required by
Sec. 51.55 of this chapter.
(2) Availability controls (if not included in the FSAR). A
description of the controls on plant operations, including availability
controls, to provide reasonable confidence that the configurations and
special treatments for SR and NSRSS SSCs provide the capabilities and
reliabilities required to demonstrate compliance with the safety
criteria of Sec. 53.220.
(3) Inspections, tests, analyses, and acceptance criteria. The
proposed ITAAC that are necessary and sufficient to provide reasonable
assurance that, if the inspections, tests, and analyses are performed
and the acceptance criteria met, a facility that incorporates the
design certification has been constructed and will be operated in
conformity with the design certification, the provisions of the Act,
and the Commission's rules and regulations.
(4) Safeguards information. 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.
(b) If there are SSCs of the plant which required research and
development to confirm the adequacy of their design, provide a report
in the application which documents the resolution of any safety
questions associated with such SSCs.
(c) A description of how the performance of each design feature has
[[Page 15840]]
been demonstrated capable of fulfilling functional design criteria
considering interdependent effects through either analysis, appropriate
test programs, prototype testing, operating experience, or a
combination thereof, in accordance with Sec. 53.440(a).
Sec. 53.1242 Review of applications.
(a) Standards for review of applications. Applications filed under
this part will be reviewed for compliance with the standards set out in
10 CFR parts 20, 51, 53, and 73.
(b) Administrative review of applications; hearings. (1) A standard
design certification is a rule that will be issued under the provisions
of subpart H of 10 CFR part 2, as supplemented by the provisions of
this section. The Commission must initiate the rulemaking after an
application has been filed under Sec. 53.1100(a)(1)(iii) and must
specify the procedures to be used for the rulemaking. The notice of
proposed rulemaking published in the Federal Register must provide an
opportunity for the submission of comments on the proposed design
certification rule. If, at the time a proposed design certification
rule is published in the Federal Register under this paragraph (b)(1),
the Commission decides that a legislative hearing should be held, the
information required by Sec. 2.1502(c) of this chapter must be
included in the Federal Register document for the proposed design
certification.
(2) Following the submission of comments on the proposed design
certification rule, the Commission may, at its discretion, hold a
legislative hearing under the procedures in subpart O of part 2 of this
chapter. The Commission must publish a document in the Federal Register
of its decision to hold a legislative hearing. The document must
contain the information specified in Sec. 2.1502(c) of this chapter
and specify whether the Commission or a presiding officer will conduct
the legislative hearing.
(3) Notwithstanding anything in Sec. 2.390 of this chapter to the
contrary, proprietary information will be protected in the same manner
and to the same extent as proprietary information submitted in
connection with applications for licenses, provided that the design
certification will be published in chapter I of this title.
(c) Reference to an issued operating license or combined license.
In those cases where a design certification application is preceded by
the issuance of an OL or custom COL for a commercial nuclear plant that
is essentially the same as the standard design for which certification
is being requested, the NRC review will follow the processes for
referencing a standard design approval in Sec. 53.1221, to the extent
practicable.
Sec. 53.1245 Referral to the Advisory Committee on Reactor
Safeguards.
The Commission must refer a copy of the application to the ACRS.
The ACRS must report on those portions of the application which concern
safety.
Sec. 53.1248 Issuance of standard design certification.
(a) After conducting a rulemaking proceeding under Sec. 53.1242 on
an application for a standard design certification and receiving the
report to be submitted by the ACRS under Sec. 53.1245, the Commission
may issue a standard design certification in the form of a rule for the
design that is the subject of the application, if the Commission
determines that--
(1) The application demonstrates compliance with the applicable
standards and requirements of the Act and the Commission's regulations;
(2) Notifications, if any, to other agencies or bodies have been
duly made;
(3) There is reasonable assurance that the standard design conforms
with the provisions of the Act and the Commission's regulations;
(4) The applicant is technically qualified;
(5) The proposed ITAAC are necessary and sufficient, within the
scope of the standard design, to provide reasonable assurance that, if
the inspections, tests, and analyses are performed and the acceptance
criteria met, the facility has been constructed and will be operated in
accordance with the design certification, the provisions of the Act,
and the Commission's regulations;
(6) Issuance of the standard design certification will not be
inimical to the common defense and security or to the health and safety
of the public;
(7) The findings required by part 51 of this chapter have been
made; and
(8) The applicant has implemented the QAP described or referenced
in the Safety Analysis Report.
(b) The design certification rule must specify the site parameters,
design characteristics, and any additional requirements and
restrictions of the design certification rule.
(c) After the Commission has adopted a final design certification
rule, the applicant must 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, as
applicable.
Sec. 53.1251 Duration of certification.
(a) Except as provided in paragraph (b) of this section, a standard
design certification issued under this subpart is valid for 40 years
from the effective date of the rule.
(b) A standard design certification continues to be valid beyond
the date of expiration in any proceeding on an application for a COL or
an OL under this part that references the standard design certification
and is docketed either before the date of expiration of the
certification, or, if a timely application for renewal of the
certification has been filed, before the Commission has determined
whether to renew the certification. A design certification also
continues to be valid beyond the date of expiration in any hearing held
under Sec. 53.1452 before operation begins under a COL that references
the design certification.
(c) An applicant for a CP, OL, COL, or ML under this part may, at
its own risk, reference in its application a design for which a design
certification application has been docketed but not granted.
Sec. 53.1254 Application for renewal.
(a) Not less than 12 nor more than 36 months before the expiration
of the initial 40-year period, or any later renewal period, any person
may apply for renewal of the certification. An application for renewal
must contain all information necessary to bring up to date the
information and data contained in the previous application. The
Commission will require, before renewal of certification, that
information normally contained in 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.
Notice and comment procedures must be used for a rulemaking proceeding
on the application for renewal. The Commission, in its discretion, may
require the use of additional procedures in individual renewal
proceedings.
(b) A design certification, either original or renewed, for which a
timely application for renewal has been filed remains in effect until
the Commission has determined whether to renew the certification. If
the certification is not
[[Page 15841]]
renewed, it continues to be valid in certain proceedings under Sec.
53.1251.
Sec. 53.1257 Criteria for renewal.
(a) The Commission must issue a rule granting the renewal if the
design, either as originally certified or as modified during the
rulemaking on the renewal, complies with the Act and the Commission's
regulations applicable and in effect at the time the certification was
issued.
(b) The Commission may impose other requirements if it determines
that--
(1) They are necessary for adequate protection to public health and
safety or common defense and security;
(2) They are necessary for compliance with the Commission's
regulations and orders applicable and in effect at the time the design
certification was issued; or
(3) There is 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
implementing those requirements are justified in view of this increased
protection.
(c) In addition, the applicant for renewal may request an amendment
to the design certification. The Commission must grant the amendment
request if it determines that the amendment will comply with the Act
and the Commission's regulations in effect at the time of renewal. If
the amendment request entails such an extensive change to the design
certification that an essentially new standard design is being
proposed, an application for a design certification must be filed in
accordance with this subpart.
(d) Denial of renewal does not bar the applicant, or another
applicant, from filing a new application for certification of the
design, which proposes design changes that correct the deficiencies
cited in the denial of the renewal.
Sec. 53.1260 Duration of renewal.
Each renewal of certification for a standard design will be for not
less than 10, nor more than 40 years.
Sec. 53.1263 Finality of standard design certifications.
(a)(1) While a standard design certification rule is in effect
under Sec. 53.1251 or Sec. 53.1260, the Commission may not modify,
rescind, or impose new requirements on the certification information,
whether on its own motion, or in response to a petition from any
person, unless the Commission determines in a rulemaking that the
change--
(i) Is necessary either to bring the certification information or
the referencing plants into compliance with the Commission's
regulations applicable and in effect at the time the certification was
issued;
(ii) Is necessary to provide adequate protection of the public
health and safety or the common defense and security;
(iii) Reduces unnecessary regulatory burden and maintains
protection to public health and safety and the common defense and
security;
(iv) Provides the detailed design information to be verified under
those ITAAC that are directed at certification information (i.e.,
design acceptance criteria);
(v) Is necessary to correct material errors in the certification
information;
(vi) Substantially increases overall safety, reliability, or
security of facility design, construction, or operation, and the direct
and indirect costs of implementation of the rule change are justified
in view of this increased safety, reliability, or security; or
(vii) Contributes to increased standardization of the certification
information.
(2)(i) In a rulemaking under Sec. 53.1263(a)(1), except for Sec.
53.1263(a)(1)(ii), the Commission will give consideration to whether
the benefits justify the costs for plants that are already licensed or
for which an application for a permit or license is under
consideration.
(ii) The rulemaking procedures for changes under Sec.
53.1263(a)(1) must provide for notice and opportunity for public
comment.
(3) Any modification the NRC imposes on a design certification rule
under paragraph (a)(1) of this section will be applied to all plants
referencing the certified design, except those to which the
modification has been rendered technically irrelevant by action taken
under paragraphs (a)(4) or (b) of this section.
(4) The Commission may not impose new requirements by plant-
specific order on any part of the design of a specific plant
referencing the design certification rule if that part was approved in
the design certification while a design certification rule is in effect
under Sec. 53.1248, unless--
(i) A modification is necessary to secure compliance with the
Commission's regulations applicable and in effect at the time the
certification was issued, or to assure adequate protection of the
public health and safety or the common defense and security; and
(ii) Special circumstances as defined in Sec. 53.080 are present.
In addition to the factors listed in Sec. 53.080, the Commission must
consider whether the special circumstances which Sec. 53.080 requires
to be present outweigh any decrease in safety that may result from the
reduction in standardization caused by the plant-specific order.
(5) Except as provided in Sec. 2.335 of this chapter, in making
the findings required for issuance of a COL, CP, OL, or ML, or for any
hearing under Sec. 53.1452, the Commission must treat as resolved
those matters resolved in connection with the issuance or renewal of a
design certification rule.
(b) An applicant who references a design certification rule may
request an exemption from one or more elements of the certification
information. The Commission may grant such a request only if it
determines that the exemption will comply with the requirements of
Sec. 53.080. In addition to the factors listed in Sec. 53.080, the
Commission must consider whether the special circumstances that Sec.
53.080 requires to be present outweigh any decrease in safety that may
result from the reduction in standardization caused by the exemption.
The granting of an exemption on request of an applicant is subject to
litigation in the same manner as other issues in the OL or COL hearing.
(c) The Commission will require, before granting a CP, COL, OL, or
ML that references a design certification rule, that information
normally contained in 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
certification information. This information may be acquired by
appropriate arrangements with the design certification applicant.
Sec. 53.1270 Manufacturing licenses.
Sections 53.1270 through 53.1295 set out the requirements and
procedures applicable to Commission issuance of a license under this
part authorizing manufacture of manufactured reactors to be installed
at sites not identified in the ML application.
[[Page 15842]]
Sec. 53.1276 Contents of applications for manufacturing licenses;
general information.
Each application for an ML must include the information contained
in Sec. 53.1109(a) through (e), and (j).
Sec. 53.1279 Contents of applications for manufacturing licenses;
technical information.
(a) Final Safety Analysis Report-siting and design. The application
must include an FSAR containing the information set forth below, with a
level of design information sufficient to enable the Commission to
judge the applicant's proposed means of ensuring that the manufacturing
conforms to the design 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:
(1) Site parameters. The site parameters postulated for the design
under this part, 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.
(2) Design information. The design information equivalent to that
required for a standard design certification as defined in Sec.
53.1239(a)(2) through (27) for those portions of a commercial nuclear
plant included in the manufactured reactor.
(3) Quality assurance program. A description of the QAP applied to
the design, and to be applied to the fabrication and testing of the
SSCs of the manufactured reactor under Sec. 53.620(a)(6), including a
discussion of how the applicable requirements of appendix B to part 50
of this chapter will be satisfied;
(4) Conceptual designs. Representative conceptual designs for one
or more commercial nuclear plants using the manufactured reactor;
(5) Operating configurations. If multiple manufactured reactors may
be installed at a commercial nuclear plant, a description of the
possible operating configurations, including common systems, interface
requirements, and system interactions. The final safety analysis must
also account for differences among the possible configurations,
including any restrictions that will be necessary during the
construction and startup of a given manufactured reactor to ensure the
safe operation of any commercial nuclear reactor already operating;
(6) Interface requirements. (i) The interface requirements between
the manufactured reactor and the remaining portions of the commercial
nuclear plant or connections to other facilities outside of the
commercial nuclear plant.
(ii) Confirmation that interface requirements are verifiable
through inspections, testing, or analysis. These requirements must be
sufficiently detailed to allow for completion of the final safety
analysis by license applicants that reference the manufactured reactor
manufactured under this subpart. Applicants for a COL under this part
will need to verify the interface requirements at the installation
site. The method to be used for verification of interface requirements
must be included as part of the proposed ITAAC required by Sec.
53.1282(a).
(iii) Information to support development of radiation monitoring
programs required under subpart F of this part by an applicant for a
COL, including potential pathways for radionuclides produced within the
manufactured reactor to enter interfacing systems.
(b) Final Safety Analysis Report--manufacturing information. The
FSAR must include the following information related to the
manufacturing processes, organization, controls, and inspections:
(1) A description, including references to generally accepted
consensus codes and standards, of the processes that will be used to
procure, fabricate, and assemble components that make up the
manufactured reactor. The description should clearly define which
activities are proposed to be within the scope of the ML 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 ML;
(2) A description of the organizational and management structure
singularly responsible for direction of 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. 53.620;
(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 such as the
molding, welding, or coating of components, and inspections and testing
of the assembled manufactured reactor or portions of the manufactured
reactor;
(4) A description of the fitness-for-duty program required by part
26 of this chapter and its implementation.
(c) Final Safety Analysis Report--deployment of the completed
manufactured reactor. 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 the manner by which the applicant will ensure close
integration between the designer, contractors, and any facility in
which the manufactured reactor is to be installed;
(3) Measures to be used for the control of interfaces, including
the consideration of key site parameters, between the holder of the ML
and the holder of the COL or CP for the commercial nuclear plant at
which the manufactured reactor is to be installed.
(d) Final Safety Analysis Report--special considerations for
factory fueling. In addition to the above paragraphs (a) through (c) of
this section, an application for an ML for a manufactured reactor that
will be fueled at the factory 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
assurance that the fueled manufactured reactor can be successfully
transported, installed, and operated at a site for which the Commission
has issued a COL or a CP and OL that authorizes construction and
operation of a commercial nuclear plant using the manufactured reactor:
(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 COL
or OL at the place of operation. The description may reference the
applicable 10 CFR part 70 application and other sections of the Safety
Analysis Report supporting the ML license application.
(i) The application must describe the measures taken for in-factory
[[Page 15843]]
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 COL or OL 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 assure 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 ML and removal under the COL
or OL 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 ML to the holder of the COL or CP and OL for the installation site.
(3) If available at the time of filing the ML application or, if
not available at the time of filing the ML application, submitted as an
amendment to the ML or ML application at the time of filing the Part 70
application, a description of the programs needed to demonstrate
compliance with the requirements of Sec. 53.620(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 COL or CP and
OL that authorizes construction and operation of a commercial nuclear
plant using the manufactured reactor, including the following.
(i) A physical security program in accordance with Sec.
53.620(d)(2)(i).
(ii) A cybersecurity program in accordance with Sec.
53.620(d)(2)(i).
Sec. 53.1282 Contents of applications for manufacturing licenses;
other application content.
(a) Inspections, tests, analyses, and acceptance criteria. (1) The
application must contain proposed inspections, tests, and analyses that
the COL or CP holder must perform, and the acceptance criteria that are
necessary and sufficient to provide reasonable assurance that, if the
inspections, tests, and analyses are performed and the acceptance
criteria met:
(i) The reactor has been manufactured in conformity with the ML,
the provisions of the Act, and the Commission's rules and regulations;
and
(ii) The manufactured reactor will be operated in conformity with
the approved design and any license authorizing operation of the
manufactured reactor.
(2) If the application references a standard design certification,
the ITAAC contained in the certified design must apply to those
portions of the facility design that are covered by the design
certification.
(3) If the application references a standard design certification,
the application may include a notification that a required inspection,
test, or analysis in the design certification ITAAC has been
successfully completed and that the corresponding acceptance criterion
has been met. The Federal Register notice required by Sec. 53.1285
must indicate that the application includes this notification.
(b) Environmental report. (1) The application must contain an
environmental report as required by Sec. 51.54 of this chapter.
(2) If the ML application references a standard design
certification, the environmental report need not contain a discussion
of severe accident mitigation design alternatives for the manufactured
reactor as used in a commercial nuclear plant.
(c) Safeguards information. 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) Performance demonstration. A description of how the performance
of each design feature has been demonstrated capable of fulfilling
functional design criteria considering interdependent effects through
either analysis, appropriate test programs, prototype testing,
operating experience, or a combination thereof, in accordance with
Sec. 53.440(a).
Sec. 53.1285 Review of applications.
(a) Standards for review of applications. Applications for MLs
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, 51, 53, 70, 71, 73, and 75.
(b) Administrative review of applications, hearings. A proceeding
on an ML 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 draft or final environmental
impact statement include an assessment of the benefits of constructing
and/or operating the manufactured reactor or an evaluation of
alternative energy sources. All hearings on MLs are governed by the
hearing procedures contained in 10 CFR part 2, subparts C, E, G, L, and
N.
Sec. 53.1286 Referral to the Advisory Committee on Reactor
Safeguards.
The Commission must refer a copy of the application to the ACRS.
The ACRS must report on those portions of the application which concern
safety.
Sec. 53.1287 Issuance of manufacturing licenses.
(a) After completing any hearing under Sec. 53.1285(b), and
receiving the report submitted by the ACRS, the Commission may issue an
ML if the Commission finds that--
(1) Applicable standards and requirements of the Act 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
commercial nuclear plant, and operated in conformity with the ML, the
provision of the Act, and the Commission's regulations;
(3) The proposed manufactured reactor can be incorporated into a
commercial nuclear plant and operated at sites having characteristics
that fall within the site parameters postulated for the design of the
manufactured reactors 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 ITAAC are necessary and sufficient, within the
scope of the ML, to provide reasonable assurance that the manufactured
reactor has been manufactured and will be operated in conformity with
the license, the provisions of the Act, and the Commission's
regulations;
(6) The issuance of a license to the applicant will not be inimical
to the common defense and security or to the health and safety of the
public; and
(7) The findings required by 10 CFR part 51 have been made.
(b) Each ML issued under this subpart must specify--
(1) Terms and conditions as the Commission deems necessary and
appropriate;
[[Page 15844]]
(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;
(4) The interface requirements to be met by the site-specific
elements of the facility, such as the energy conversions systems and
ultimate heat sink, not within the scope of the manufactured reactor;
and
(5) The entity with design authority for the manufactured reactor
covered by the license.
Sec. 53.1288 Finality of manufacturing licenses.
(a)(1) During the term of an ML issued under this part, the
Commission may not modify, rescind, or impose new requirements on the
design of the manufactured reactor, or the requirements for the
manufacture of the manufactured 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 ML was issued, or
to provide reasonable assurance of adequate protection to public health
and safety or common defense and security.
(2) Any modification to the design of a manufactured reactor that
is imposed by the Commission under paragraph (a)(1) of this section
will be applied to all manufactured reactors manufactured under the
license, including those that have already been transported and sited,
except those manufactured reactors to which the modification has been
rendered technically irrelevant or otherwise unnecessary by action
taken under Sec. 53.1530, Sec. 53.1550, or paragraph (b) of this
section.
(3) In making the findings required under this part for issuance of
a COL, CP, or OL, in any hearing under Sec. 53.1452, or in any
enforcement hearing other than one initiated by the Commission under
paragraph (a)(1) of this section, for which a manufactured reactor
manufactured under this subpart is referenced or used, the Commission
must treat as resolved those matters resolved in the proceeding on the
application for issuance or renewal of the ML, including the adequacy
of design of the manufactured reactor, the costs and benefits of severe
accident mitigation design alternatives, and the bases for not
incorporating severe accident mitigation design alternatives into the
design of the manufactured reactor to be manufactured.
(b) An applicant who references or uses a manufactured reactor
manufactured under an ML under this part may include in the application
a request for a departure from the design characteristics, site
parameters, terms and conditions, or approved design of the
manufactured reactor. The Commission may grant a request only if it
determines that the departure will comply with the requirements of
Sec. 53.080 The granting of a departure on request of an applicant is
subject to litigation in the same manner as other issues in the COL or
CP hearing.
Sec. 53.1291 Duration of manufacturing licenses.
An ML issued under this part is valid for not less than 5, nor more
than 40 years from the date of issuance. Upon expiration of the ML, the
manufacture of any uncompleted manufactured reactors must cease unless
a timely application for renewal has been docketed with the NRC.
Sec. 53.1293 Transfer of manufacturing licenses.
An ML may be transferred under Sec. 53.1570.
Sec. 53.1295 Renewal of manufacturing licenses.
(a)(1) Not less than 12 months, nor more than 5 years before the
expiration of the ML, or any later renewal period, the holder of the ML
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.
(2) The filing of an application for a renewed license must be in
accordance with subpart A of 10 CFR part 2 and Sec. 53.1100.
(3) An ML 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.
(4) 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.
(b) The Commission may grant the renewal if the Commission
determines--
(1) The ML complies with the Act and the Commission's regulations
and orders applicable and in effect at the time the ML 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 ML 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.
(c) A renewed ML may be issued for a term of not less than 5, nor
more than 40 years, plus any remaining years on the ML then in effect
before renewal. The renewed license must be subject to the requirements
of Sec. 53.1288.
Sec. 53.1300 Construction permits.
Sections 53.1300 through 53.1348 set out the requirements and
procedures applicable to Commission issuance of a CP for commercial
nuclear plants. A CP for the construction of a commercial nuclear plant
under this part will be issued before the issuance of an OL if the
application is otherwise acceptable and will be converted upon
completion of the facility and Commission action, into an OL as
provided under Sec. Sec. 53.1360 through 53.1405.
Sec. 53.1306 Contents of applications for construction permits;
general information.
An application for a CP must include the information required by
Sec. 53.1109 and the following information:
(a) Information sufficient to demonstrate to the Commission the
financial qualification of the applicant to carry out, under the
regulations in this chapter, the activities for which the permit is
sought. As applicable, the applicant should provide information that
demonstrates that the applicant appears to be financially qualified to
cover estimated construction costs and related fuel cycle costs,
including estimates of the total construction costs and related fuel
cycle costs of the facility, 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.
(b) If the applicant proposes to construct or alter a facility, the
application must state the earliest and
[[Page 15845]]
latest dates for completion of the construction or alteration.
Sec. 53.1309 Contents of applications for construction permits;
technical information.
The application must contain a Preliminary Safety Analysis Report
(PSAR) that describes the facility and the limits on its operation and
presents a preliminary safety analysis of the SSCs of the facility as a
whole. The PSAR must include the following information, at a level of
detail sufficient to enable the Commission to reach a conclusion on
safety matters that must be resolved by the Commission before issuance
of a CP:
(a)(1) Site information. An application for a CP for a commercial
nuclear reactor must include the site information equivalent to that
required for an early site permit in Sec. 53.1146(a)(1)(iv) through
(x).
(2) Design information. Except as specified in this paragraph
(a)(2), an application for a CP for a commercial nuclear plant must
include the design information equivalent to that required for a
standard design certification as defined in Sec. 53.1239(a)(2) through
(a)(21), (a)(23), and Sec. 53.1239(a)(26) through (27).
(i) Quality assurance program. A description of the QAP to be
applied to the design, fabrication, construction, and testing of the
SSCs of the facility under Sec. 53.610(a)(6), including a discussion
of how the requirements of appendix B to part 50 of this chapter will
be satisfied.
(ii) Preliminary design information. The information provided in
the application may include some aspects of the design that are not
fully developed, and the information is therefore preliminary. The
completed design, including any changes during construction, must be
described in the FSAR required in Sec. 53.1369 that supports an
application for an OL.
(iii) Planned research or testing. Descriptions of how design
features and related functional design criteria will fulfill the safety
criteria in subpart B and how that has been or will be demonstrated
through either analysis, appropriate test programs, experience, or a
combination thereof. Where any design feature has not been fully
developed or demonstrated to fulfill the functional design criteria at
the time of an application for a CP, the applicant must provide a plan
for future analysis, research and development, test programs, gathering
of experience, or a combination thereof to provide reasonable
confidence that the required demonstration will be available for an
application for an OL
(iv) Programmatic controls. Descriptions of the programmatic
controls may include those to be provided in the FSAR or other
licensing-basis documents because they are necessary to achieve and
maintain the reliability and capability of SSCs relied upon to
demonstrate compliance with the established safety criteria and
functional design criteria required in subpart B, and to maintain
consistency with analyses required by Sec. 53.450.
(3) Technical qualifications. A description of the technical
qualifications of the applicant to engage in the proposed activities
under the regulations in this chapter.
(4) Emergency preparedness. A description of the applicant's
preliminary plans for coping with emergencies based on:
(i) Except as provided in paragraph (a)(4)(ii) of this section, the
requirements in appendix E to part 50.
(ii) For a commercial nuclear plant consisting of either small
modular reactors or non-light-water reactors, the requirements in
either Sec. 50.160 or appendix E to part 50.
(5) Physical security. A report that provides a preliminary
description of how the site characteristics support the development of
adequate security plans and measures consistent with the requirements
in Sec. 53.540.
(6) Fitness-for-duty program. A description of the fitness-for-duty
(FFD) program required by 10 CFR part 26 and its implementation.
(b) 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.
Sec. 53.1312 Contents of applications for construction permits; other
application content.
(a) In addition to the PSAR, the application must include the
following:
(1) An environmental report either under Sec. 51.50(a) of this
chapter if an LWA under Sec. 53.1130 is not requested in conjunction
with the CP application, or under Sec. Sec. 51.49 and 51.50(a) of this
chapter if an LWA is requested in conjunction with the CP application;
or
(2) If the applicant wishes to request that an LWA under Sec.
53.1130 be issued before issuance of the CP, the information otherwise
required by Sec. 53.1130, in accordance with either Sec. 2.101(a)(1)
through (a)(5), or Sec. 2.101(a)(9) of this chapter.
(b) If the CP application references an early site permit, standard
design approval, standard design certification, or ML issued under this
part, then the following requirements apply:
(1) The PSAR need not contain information or analyses submitted to
the Commission in connection with the referenced NRC approval,
license,, or certification, provided, however, that the PSAR
incorporates the material by reference and confirms that the site and
design of the facility falls within parameter values postulated in the
referenced NRC approval, license, or certification.
(2) The PSAR must provide a means to demonstrate that all terms and
conditions that have been included in the referenced NRC approval,
license, or certification will be satisfied by the date of issuance of
the OL, as appropriate. If the PSAR does not demonstrate that each site
characteristic falls within the corresponding postulated site parameter
and each design characteristic of the facility falls within the
corresponding postulated design parameter, the application must justify
a departure, variance, or exemption from the referenced NRC approval,
license, or certification in regard to that particular site or design
characteristic in compliance with the requirements of this part.
(3) If a referenced early site permit approves complete and
integrated emergency plans, or major features of emergency plans, then
the PSAR must include any new or additional information that updates
and corrects the information that was provided under Sec.
53.1146(b)(2) and discuss whether the new or additional information
materially changes the bases for compliance with the applicable
requirements.
Sec. 53.1315 Review of applications.
(a) Standards for review of applications. Applications filed under
this part will be reviewed according to the standards set out in 10 CFR
parts 20, 51, 53, 73, and 140.
(b) Administrative review of applications; hearings. A proceeding
on a CP application is subject to all applicable procedural
requirements contained in 10 CFR part 2, including the requirements for
docketing (Sec. 2.101 of this chapter) and issuance of a notice of
hearing (Sec. 2.104 of this chapter). All hearings on CP applications
are governed by the procedures contained in 10 CFR part 2.
Sec. 53.1318 Finality of referenced NRC approvals, permits, and
certifications.
If the application for a CP under this part references an early
site permit, standard design approval, standard design certification,
or ML, the scope and nature of matters resolved for the application are
governed by the relevant
[[Page 15846]]
provisions addressing finality, including Sec. Sec. 53.1188, 53.1221,
53.1263, and 53.1288.
Sec. 53.1324 Referral to the Advisory Committee on Reactor
Safeguards.
The Commission must refer a copy of the application to the ACRS.
The ACRS must report on those portions of the application that concern
safety and must apply the standards referenced in Sec. 53.1315, in
accordance with the finality provisions in Sec. 53.1318.
Sec. 53.1327 Authorization to conduct limited work authorization
activities.
(a) If the application does not reference an early site permit
which authorizes the holder to perform the activities under Sec.
53.1130, the applicant may not perform those activities without
obtaining the separate authorization required by Sec. 53.1130.
Authorization may be granted only after the presiding officer in the
proceeding on the application has made the findings and determination
required by Sec. 53.1130(b)(1)(ii) and (iv), and the Director, Office
of Nuclear Reactor Regulation makes the determination required by Sec.
53.1130(b)(1)(iii).
(b) If, after an applicant has performed the activities permitted
by paragraph (a) of this section, the application for the CP is
withdrawn or denied, then the applicant must implement an approved site
redress plan.
Sec. 53.1330 Exemptions, departures, and variances.
(a) Applicants for a CP under this part, or any amendment to a CP,
may include in the application a request for an exemption from one or
more of the Commission's regulations. The Commission may grant a
request if it determines that the exemption complies with Sec. 53.080.
(b) An applicant for a CP who has filed an application referencing
an NRC approval, license, or certification issued under this part may
include in the application a request for exemptions, departures, or
variances related to the subject referenced NRC approval, license, or
certification. In determining whether to grant the departure, variance,
or exemption, the Commission must apply the same technically relevant
criteria as were applicable to the application for the original or
renewed approval, license, or certification.
Sec. 53.1333 Issuance of construction permits.
(a) After conducting a hearing in accordance with Sec. 53.1315 and
receiving the report submitted by the ACRS, the Commission may issue a
CP only if the Commission finds that--
(1) The applicant has described the proposed design of the facility
and has identified the major features or components incorporated
therein for the protection of the health and safety of the public;
(2) Such further technical or design information as may be required
to complete the safety analysis, and which can reasonably be left for
later consideration, will be supplied in the FSAR;
(3) Safety features or components, if any, that require research
and development have been described by the applicant and the applicant
has identified, and there will be conducted, a research and development
program reasonably designed to resolve any safety questions associated
with such features or components; and
(4) On the basis of the foregoing, there is reasonable assurance of
the following--
(i) Such safety questions will be satisfactorily resolved at or
before the latest date stated in the application for completion of
construction of the proposed facility; and
(ii) Taking into consideration the site criteria contained in
subpart D to this part, the proposed facility can be constructed and
operated at the proposed location without undue risk to the health and
safety of the public.
(b) A CP must contain the terms and conditions for the permit, as
the Commission deems necessary and appropriate. The Commission may, in
its discretion, incorporate in any CP provisions requiring the
applicant to furnish periodic reports of the progress and results of
research and development programs designed to resolve safety questions.
Sec. 53.1336 Finality of construction permits.
Notwithstanding any provision in Sec. 53.1590, a CP constitutes an
authorization to proceed with construction but does not constitute
Commission approval of the safety of any design feature or
specification unless the applicant specifically requests such approval
and such approval is incorporated in the permit. The applicant, at its
option, may request such approvals in the CP or by amendment to the CP.
If approved by the NRC and included in the permit, the NRC will
consider modifications to the approved design features or
specifications in accordance with Sec. 53.1590.
Sec. 53.1342 Duration of construction permits.
(a) A CP will state the earliest and latest dates for completion of
construction or alteration of the facility, not to exceed 40 years from
date of issuance.
(b) If the proposed construction or alteration of the facility is
not completed by the latest completion date, the CP shall expire, and
all rights are forfeited. However, upon good cause shown, the
Commission will extend the completion date for a reasonable period of
time. The Commission will recognize, among other things, developmental
problems attributed to the experimental nature of the facility or fire,
flood explosion, strike, sabotage, domestic violence, enemy action an
act of the elements, and other acts beyond the control of the permit
holder, as a basis for extending the completion date.
Sec. 53.1345 Transfer of construction permits.
A CP may be transferred under Sec. 53.1570.
Sec. 53.1348 Termination of construction permits.
When a permit holder has determined to permanently cease
construction, the holder must, within 30 days, submit a written
certification to the NRC.
Sec. 53.1360 Operating licenses.
Sections 53.1360 through 53.1405 set out the requirements and
procedures applicable to Commission issuance of an OL for a nuclear
power facility.
Sec. 53.1366 Contents of applications for operating licenses;
general information.
An application for an OL must include the information required by
Sec. 53.1109 and, except for an electric utility applicant,
information sufficient to demonstrate to the Commission the financial
qualification of the applicant to carry out, in accordance with the
regulations in this chapter, the activities for which the license is
sought. As applicable, 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 must 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.
[[Page 15847]]
Sec. 53.1369 Contents of applications for operating licenses;
technical information.
Final Safety Analysis Report. The application must contain an FSAR
that describes the facility and the limits on its operation and
presents a safety analysis of the SSCs of the facility as a whole. The
FSAR must include the following information, at a level of detail
sufficient to enable the Commission to reach a final conclusion on all
safety matters that must be resolved by the Commission before issuance
of an OL:
(a) Site information. An application for an OL for a commercial
nuclear reactor must include the site information equivalent to that
required for an early site permit in Sec. 53.1146(a)(1)(iv) through
(x), including all current information, such as the results of
environmental and meteorological monitoring programs, which has been
developed since issuance of the CP, relating to site evaluation factors
identified in this part.
(b) Design information. Except as specified in this paragraph (b),
an FSAR for an OL for a commercial nuclear plant must include the final
design information equivalent to that required for a standard design
certification as defined in Sec. 53.1239(a)(2) through (7), (a)(9),
(a)(11) and (12), (a)(14) through (21), (a)(23), and (a)(25).
(1) The completed design, including any changes during
construction, must be described.
(2) Where any design feature had not been fully developed or
demonstrated at the time of application for the CP, the applicant must
provide the analysis, research and development, test programs,
gathering of experience, or a combination thereof to provide the
required demonstration to fulfill the functional design criteria.
(c) [Reserved]
(d) Integrity assessment program. A description of an Integrity
Assessment Program that addresses the elements described in Sec.
53.870.
(e) Safeguards information. 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.
(f) Emergency response facility or facilities. Description of
location and capabilities to be established for command and control,
support, and coordination of onsite and offsite, as applicable,
functions during reactor accident conditions.
(g) Role of personnel. (1) A description of the completed
assessments related to the role of personnel in ensuring safe
operations considering the analyses required by Sec. 53.730. These
assessments must include the following:
(i) Human factors engineering design requirements of Sec.
53.730(a);
(ii) Human system interface design requirements of Sec. 53.730(b);
(iii) Concept of operations of Sec. 53.730(c);
(iv) Functional requirements analysis and function allocation of
Sec. 53.730(d);
(2) A description of the program to be used for evaluating and
applying operating experience as required by Sec. 53.730(e);
(3) A staffing plan and supporting analyses as required by Sec.
53.730(f).
(h) Training, examination, and proficiency programs. (1) A
description of the training, examination, and proficiency programs
required by Sec. 53.730(g);
(2) A description of the training programs required by Sec.
53.830.
(i) Emergency plan. Emergency plans complying with the requirements
of Sec. 53.855.
(1) Include all emergency plan certifications, as applicable, that
have been obtained from the State, local, and participating Tribal
governmental agencies with emergency planning responsibilities that are
wholly or partially within the EPZ plume exposure pathway. These
certifications must state that--
(i) The proposed emergency plans are practicable;
(ii) These agencies are committed to participating in any further
development of the plans, including any required field demonstrations;
and
(iii) These agencies are committed to executing their
responsibilities under the plans in the event of an emergency.
(2) If certifications cannot be obtained after sustained, good
faith efforts by the applicant, then the application must contain
information, including a utility plan, sufficient to show that the
proposed plans provide reasonable assurance that adequate protective
measures can and will be taken in the event of a radiological emergency
at the site.
(3) If complete and integrated emergency plans were approved as
part of an early site permit, or submitted, reviewed, and approved as
part of the CP application, new certifications that demonstrate
compliance with the requirements of paragraph (i)(1) of this section
are not required.
(j) Organization. A description of the applicant's organizational
structure, allocations of responsibilities and authorities, and
personnel qualifications requirements for operation.
(k) Maintenance program. A description of a maintenance program
under Sec. 53.715.
(l) Quality assurance. A description of the QAP that demonstrates
compliance with the requirements under Sec. 53.865.
(m) Radiation protection program. A radiation protection program
description under Sec. 53.850.
(n) Security program. A physical security plan that describes how
the applicant will comply with Sec. 53.860 (and 10 CFR part 11, if
applicable, including the identification and description of jobs as
required by Sec. 11.11(a) of this chapter, at the proposed facility).
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.
(o) Safeguards contingency plan. A safeguards contingency plan in
accordance with the criteria set forth in appendix C to 10 CFR part 73.
The safeguards contingency plan must include plans for dealing with
threats, thefts, and radiological sabotage, as defined in 10 CFR part
73, relating to the SNM and nuclear facilities licensed under this
chapter and in the applicant's possession and control. Each application
for this type of license must include the information contained in the
applicant's safeguards contingency plan. (Implementing procedures
required for this plan need not be submitted for approval.) \1\
(p) Security training and qualification. A training and
qualification plan that describes how the applicant will demonstrate
compliance with the criteria set forth in Sec. 73.100 of this chapter
or appendix B to 10 CFR part 73.
(q) Cybersecurity plan. A cybersecurity plan in accordance with the
criteria set forth in Sec. 73.54 or Sec. 73.110 of this chapter.
(r) Security, safeguards, and cybersecurity plan implementation. A
description of the implementation of the physical security plan,
safeguards contingency plan, training and qualification plan, and
cybersecurity plan. Each applicant who prepares a physical security
plan, a safeguards contingency plan, a training and qualification plan,
or a cybersecurity plan must protect the plans and other related
Safeguards Information against unauthorized disclosure in accordance
with the requirements of Sec. Sec. 73.21 and 73.22 of this chapter.
(s) Fire protection program. A description of the fire protection
program under Sec. 53.875.
(t) Inservice inspection/inservice testing program. A description
of the
[[Page 15848]]
inservice inspection and inservice testing programs under Sec. 53.880.
(u)-(v) [Reserved]
(w) General employee training. A description of the training
program required to demonstrate compliance with Sec. 53.830 and its
implementation.
(x) Fitness-for-duty program. A description of the FFD program
required by 10 CFR part 26 and its implementation.
(y) Other programs. A description and evaluation of the results of
the applicant's programs, including research and development, if any,
to demonstrate that any safety questions identified at the CP stage
have been resolved.
(z) Safety design feature performance. A description of how the
performance of each safety design feature has been demonstrated capable
of fulfilling functional design criteria considering interdependent
effects through either analysis, appropriate test programs, prototype
testing, operating experience, or a combination thereof, in accordance
with Sec. 53.440(a).
(aa) Technical specifications. Proposed technical specifications
prepared in accordance with the requirements of Sec. 53.710(a).
\1\ A physical security plan that contains all the information
required in both Sec. 73.55 or Sec. 73.100 of this chapter and
appendix C to 10 CFR part 73 satisfies the requirement for a
contingency plan.
Sec. 53.1372 Contents of applications for operating licenses; other
application content.
In addition to the FSAR, the application must also include the
following:
(a) Environmental report. An environmental report in accordance
with Sec. 51.53(b) of this chapter.
(b) Availability controls (if not included in the FSAR). A
description of the controls on plant operations, including availability
controls, to provide reasonable confidence of safe operation and that
the configurations and special treatments for SR and NSRSS SSCs provide
the capabilities and reliabilities required to satisfy the safety
criteria of Sec. 53.220 if not addressed by Technical Specifications
under Sec. 53.1369(aa).
Sec. 53.1375 Review of applications.
(a) Standards for review of applications. Applications filed under
this part will be reviewed according to the standards set out in 10 CFR
parts 20, 26, 51, 53, 73, and 140.
(b) Administrative review of applications; hearings. A proceeding
on an OL is subject to all applicable procedural requirements contained
in 10 CFR part 2, including the requirements for docketing (Sec. 2.101
of this chapter) and issuance of a notice of hearing (Sec. 2.104 of
this chapter). All hearings on OLs are governed by the procedures
contained in 10 CFR part 2.
Sec. 53.1381 Referral to the Advisory Committee on Reactor
Safeguards.
The Commission must refer a copy of the application to the ACRS.
The ACRS must report on those portions of the application that concern
safety and must apply the standards referenced in Sec. 53.1375.
Sec. 53.1384 Exemptions, departures, and variances.
(a) Applicants for an OL under this part, or any amendment to an
OL, may include in the application a request for an exemption from one
or more of the Commission's regulations. The Commission may grant an
exemption request if it determines that the exemption complies with
Sec. 53.080.
(b) An applicant for an OL who has filed an application referencing
an NRC approval, permit, license, or certification issued under this
part may include in the application a request for departures,
variances, or exemptions related to the subject referenced NRC
approval, permit, license, or certification. In determining whether to
grant the departure, variance, or exemption, the Commission must apply
the same technically relevant criteria as were applicable to the
application for the original or renewed approval, license, or
certification.
Sec. 53.1387 Issuance of operating licenses.
Upon completion of the construction or 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 OL or an appropriate amendment of the license,
as the case may be.
(a)(1) After receiving the report submitted by the ACRS, the
Commission may issue an OL if the Commission finds that--
(i) Construction of the facility has been substantially completed
in conformity with the CP and the application as amended, the
provisions of the Act, and the rules and regulations of the Commission;
(ii) Any required notifications to other agencies or bodies have
been duly made;
(iii) The facility will operate in conformity with the application
as amended, the provisions of the Act, and the rules and regulations of
the Commission;
(iv) There is reasonable assurance that--
(A) The activities authorized by the OL can be conducted without
endangering the health and safety of the public; and
(B) Such activities will be conducted in compliance with the
regulations in this chapter.
(v) The applicant is technically and financially qualified to
engage in the activities authorized, however, no finding of financial
qualification is necessary for an electric utility applicant for an OL;
(vi) Issuance of the license will not be inimical to the common
defense and security or to the health and safety of the public;
(vii) The applicable provisions of 10 CFR part 140 have been
satisfied; and
(viii) The findings required by 10 CFR part 51 have been made.
(2) [Reserved]
(b) [Reserved]
(c) The OL will include appropriate provisions with respect to any
uncompleted items of construction and such limitations or conditions as
are required to assure that operation during the period of the
completion of such items will not endanger public health and safety.
(d) The Commission will issue an OL in such form and containing
such conditions and limitations, including technical specifications, as
it deems necessary and appropriate.
Sec. 53.1390 Backfitting of operating licenses.
After issuance of an OL, the Commission may not modify, add, or
delete any term or condition of the OL, except in accordance with the
provisions of Sec. 53.1590.
Sec. 53.1396 Duration of operating licenses.
The Commission will issue an OL under this part for the term
requested by the applicant, not to exceed 40 years from the date of
issuance, or for the estimated useful life of the facility if the
Commission determines that the estimated useful life is less than the
term requested.
Sec. 53.1399 Transfer of an operating license.
An OL may be transferred under Sec. 53.1570.
Sec. 53.1402 Application for renewal.
The filing of an application for a renewed license must be in
accordance with Sec. 53.1595.
[[Page 15849]]
Sec. 53.1405 Continuation of an operating license.
Each OL for a facility that has permanently ceased operations
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 license is terminated. During this period
of continued effectiveness, the licensee must--
(a) Take actions necessary to decommission and decontaminate the
facility and continue to maintain the facility, including, where
applicable, the storage, control, and maintenance of the spent fuel in
a safe condition; and
(b) Conduct activities in accordance with all other restrictions
applicable to the facility in accordance with the NRC's regulations and
the provisions of the OL for the facility.
Sec. 53.1410 Combined licenses.
Sections 53.1410 through 53.1461 set out the requirements and
procedures applicable to Commission issuance of COLs for commercial
nuclear plants under this part.
Sec. 53.1413 Contents of applications for combined licenses; general
information.
An application for a COL must include the information required by
Sec. 53.1109 and, except for an electric utility applicant in regard
to financial assurance required after a Commission finding under Sec.
53.1452, the application must include information sufficient to
demonstrate to the Commission the financial qualification of the
applicant to carry out, in accordance with the regulations in this
chapter, the activities for which the permit or license is sought. As
applicable, the following should be provided:
(a) The information that demonstrates that the applicant appears to
be financially qualified to cover estimated construction costs and
related fuel cycle costs, including estimates of the total construction
costs and related fuel cycle costs of the facility, 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.
(b) 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, 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.
Sec. 53.1416 Contents of applications for combined licenses;
technical information.
(a) The application must contain an FSAR that describes the
facility and the limits on its operation and presents a safety analysis
of the SSCs of the facility as a whole. The Commission will require,
before issuance of a COL, that information normally contained in
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. The FSAR must include
the following information, at a level of detail sufficient to enable
the Commission to reach a final conclusion on all safety matters that
must be resolved by the Commission before issuance of a COL:
(1) Site information. An application for a COL for a commercial
nuclear reactor must include the site information required for an early
site permit in Sec. 53.1146(a)(1)(iv) through (x).
(2) Design information. An application for a COL for a commercial
nuclear plant must include the design information equivalent to that
required for a standard design certification as defined in Sec.
53.1239(a)(2) through (7), (a)(9), (a)(11), (a)(12), (a)(14) through
(21), and (a)(23).
(3) Technical qualifications. A description of the technical
qualifications of the applicant to engage in the proposed activities in
accordance with the regulations in this chapter.
(4) Integrity assessment program. A description of an Integrity
Assessment Program that addresses the elements described in Sec.
53.870.
(5) Safeguards information. 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.
(6) Emergency response facility or facilities. Description of the
locations and capabilities to be established for command and control,
support, and coordination of onsite and offsite, as applicable,
functions during reactor accident conditions.
(7) Role of personnel. (i) A description of the completed
assessments related to the role of personnel in ensuring safe
operations considering the analyses required by Sec. 53.730. These
assessments must include the following:
(A) Human factors engineering design requirements of Sec.
53.730(a);
(B) Human system interface design requirements of Sec. 53.730(b);
(C) Concept of operations of Sec. 53.730(c); and
(D) Functional requirements analysis and function allocation of
Sec. 53.730(d);
(ii) A description of the program to be used for evaluating and
applying operating experience as required by Sec. 53.730(e);
(iii) A staffing plan and supporting analyses as required by Sec.
53.730(f).
(8) Training, examination, and proficiency programs. (i) A
description of the training, examination, and proficiency programs
required by Sec. 53.730(g); and
(ii) A description of the training programs required by Sec.
53.830.
(9) Emergency plan. Emergency plans complying with the requirements
of Sec. 53.855.
(i) The emergency plan must include, as applicable, all emergency
plan certifications that have been obtained from the State, local, and
participating Tribal governmental agencies with emergency planning
responsibilities. The certifications must state that--
(A) The proposed emergency plans are practicable;
(B) These agencies are committed to participating in any further
development of the plans, including any required field demonstrations;
and
(C) These agencies are committed to executing their
responsibilities under the plans in the event of an emergency.
(ii) If certifications cannot be obtained after sustained, good
faith efforts by the applicant, then the application must contain
information, including a utility plan, sufficient to show that the
proposed plans provide reasonable assurance that adequate protective
measures can and will be taken in the event of a radiological emergency
at the site.
(10) Organization. A description of the applicant's organizational
structure, allocations of responsibilities and authorities, and
personnel qualifications requirements for operation.
(11) Maintenance program. A description of a maintenance program
under Sec. 53.715.
(12) Quality assurance. A description of the QAP under Sec.
53.865.
(13) Radiation protection program. A radiation protection program
description under Sec. 53.850.
[[Page 15850]]
(14) Security program. A physical security plan that describes how
the applicant will comply with Sec. 53.860 (and 10 CFR part 11, if
applicable, including the identification and description of jobs as
required by Sec. 11.11(a) of this chapter, at the proposed facility).
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.
(15) Safeguards contingency plan. A safeguards contingency plan in
accordance with the criteria set forth in appendix C to 10 CFR part 73.
The safeguards contingency plan must include plans for dealing with
threats, thefts, and radiological sabotage, as defined in 10 CFR part
73, relating to the SNM and nuclear facilities licensed under this
chapter and in the applicant's possession and control. Each application
for this type of license must include the information contained in the
applicant's safeguards contingency plan.\1\ (Implementing procedures
required for this plan need not be submitted for approval.)
(16) Security training and qualification. A training and
qualification plan that describes how the applicant will demonstrate
compliance with the criteria set forth in Sec. 73.100 of this chapter
or appendix B to 10 CFR part 73.
(17) Cybersecurity plan. A cybersecurity plan in accordance with
the criteria set forth in Sec. 73.54 or Sec. 73.110 of this chapter.
(18) Security, safeguards, and cybersecurity plan implementation. A
description of the implementation of the physical security plan,
safeguards contingency plan, training and qualification plan, and
cybersecurity plan. Each applicant who prepares a physical security
plan, a safeguards contingency plan, a training and qualification plan,
or a cybersecurity plan must protect the plans and other related
Safeguards Information against unauthorized disclosure in accordance
with the requirements of Sec. Sec. 73.21 and 73.22 of this chapter.
(19) Fire protection program. A description of the fire protection
program under Sec. 53.875.
(20) Inservice inspection/inservice testing program. Descriptions
of inservice inspection and inservice testing programs under Sec.
53.880.
(21)-(22) [Reserved]
(23) General employee training. A description of the training
program required to demonstrate compliance with Sec. 53.830 and its
implementation.
(24) Fitness-for-duty program. A description of the FFD program
under part 26 of this chapter and its implementation.
(25) Technical specifications. Proposed technical specifications
prepared in accordance with the requirements of Sec. 53.710(a).
(b) If there are SSCs of the plant for which research and
development is necessary to confirm the adequacy of their design, a
report which documents the resolution of any safety questions
associated with such SSCs.
(c) A description of how the performance of each safety design
feature has been demonstrated capable of fulfilling functional design
criteria considering interdependent effects through either analysis,
appropriate test programs, prototype testing, operating experience, or
a combination thereof, in accordance with Sec. 53.440(a).
(d) If the COL application references an early site permit, then
the following requirements apply:
(1) The FSAR need not contain information or analyses submitted to
the Commission in connection with the early site permit provided that
the FSAR must either include or incorporate by reference the early site
permit Site Safety Analysis Report and contain, in addition to the
information and analyses otherwise required, information sufficient to
demonstrate that the design of the facility falls within the site
characteristics and design parameters specified in the early site
permit.
(2) If the FSAR does not demonstrate that design of the facility
falls within the site characteristics and design parameters, the
application must include a request for a variance that complies with
the requirements of Sec. Sec. 53.1188(d) and 53.1437.
(3) The FSAR must demonstrate that all terms and conditions that
have been included in the early site permit will be satisfied by the
date of issuance of the COL. Any terms or conditions of the early site
permit that could not be met by the time of issuance of the COL must be
set forth as terms or conditions of the COL.
(4) If the early site permit approves complete and integrated
emergency plans, or major features of emergency plans, then the FSAR
must include any new or additional information that updates and
corrects the information that was provided under Sec. 53.1146(b)(2)
and discuss whether the new or additional information materially
changes the bases for compliance with the applicable requirements. The
application must identify changes to the emergency plans or major
features of emergency plans that have been incorporated into the
proposed facility emergency plans and that constitute or would
constitute a change in an emergency plan that results in reducing the
licensee's capability to perform an emergency planning function in the
event of a radiological emergency.
(5) If complete and integrated emergency plans are approved as part
of the early site permit, new certifications meeting the requirements
of paragraph (a)(9)(i) of this section are not required.
(e) If the COL application references a standard design approval,
then the following requirements apply:
(1) The FSAR need not contain information or analyses submitted to
the Commission in connection with the design approval, provided,
however, that the FSAR must either include or incorporate by reference
the standard design approval FSAR and must contain, in addition to the
information and analyses otherwise required, information sufficient to
demonstrate that the characteristics of the site fall within the site
parameters specified in the design approval. In addition, the plant-
specific information of the PRA, other SREs, or a combination thereof
must use the information of the PRA, other SREs, or a combination
thereof for the design approval and must be updated to account for
site-specific design information and any design changes or departures.
(2) The FSAR must demonstrate that all terms and conditions that
have been included in the design approval will be satisfied by the date
of issuance of the COL.
(f) If the COL application references a standard design
certification, then the following requirements apply:
(1) The FSAR need not contain information or analyses submitted to
the Commission in connection with the standard design certification,
provided, however, that the FSAR must either include or incorporate by
reference the standard design certification FSAR and must contain, in
addition to the information and analyses otherwise required,
information sufficient to demonstrate that the site characteristics
fall within the site parameters specified in the standard design
certification. In addition, the plant-specific information of the PRA,
other SREs, or a combination thereof must use the information of the
PRA, other SREs, or a combination thereof for the standard design
certification and must be updated to account for site-specific design
information and any design changes or departures.
(2) The FSAR must demonstrate that the interface requirements
established for the design under Sec. 53.1239(a)(24) have been met.
[[Page 15851]]
(3) The FSAR must demonstrate that all requirements and
restrictions set forth in the referenced standard design certification
rule must be satisfied by the date of issuance of the COL. Any
requirements and restrictions set forth in the referenced standard
design certification rule that could not be satisfied by the time of
issuance of the COL, must be set forth as terms or conditions of the
COL.
(g) If the COL application references the use of one or more
manufactured reactors licensed under Sec. 53.1270, then the following
requirements apply:
(1) The FSAR need not contain information or analyses submitted to
the Commission in connection with the ML, provided, however, that the
FSAR must either include or incorporate by reference the ML FSAR and
must contain, in addition to the information and analyses otherwise
required, information sufficient to demonstrate that the site
characteristics fall within the site parameters specified in the ML. In
addition, the plant-specific information of the PRA, other SREs, or a
combination thereof must use the information of the PRA, other SREs, or
a combination thereof for the manufactured reactor and must be updated
to account for site-specific design information and any design changes
or departures.
(2) The FSAR must demonstrate that the interface requirements
established for the design have been met.
(3) The FSAR must demonstrate that all terms and conditions that
have been included in the ML will be satisfied by the date of issuance
of the COL. Any terms or conditions of the ML that could not be met by
the time of issuance of the COL, must be set forth as terms or
conditions of the COL.
(h) Each applicant for a COL under this part must protect
Safeguards Information against unauthorized disclosure in accordance
with the requirements in Sec. Sec. 73.21 and 73.22 of this chapter, as
applicable.
\1\ A physical security plan that contains all the information
required in both Sec. 73.55 or Sec. 73.100 of this chapter and
appendix C to 10 CFR part 73 demonstrates compliance with the
requirement for a contingency plan.
Sec. 53.1419 Contents of applications for combined licenses; other
application content.
(a) In addition to the FSAR, the application must also include the
following:
(1) Environmental report. (i) An environmental report either in
accordance with Sec. 51.50(c) of this chapter if an LWA under Sec.
53.1130 is not requested in conjunction with the COL application, or in
accordance with Sec. Sec. 51.49 and 51.50(c) of this chapter if an LWA
is requested in conjunction with the COL application; or
(ii) If the applicant wishes to request that an LWA under Sec.
53.1130 be issued before issuance of the COL, the information otherwise
required by Sec. 53.1130, in accordance with either Sec. 2.101(a)(1)
through (a)(4), or Sec. 2.101(a)(9) of this chapter;
(2) Availability controls (if not included in the FSAR). A
description of the controls on plant operations, including availability
controls, to provide reasonable confidence of safe operation and that
the configurations and special treatments for SR SSCs and NSRSS SSCs
provide the capabilities and reliabilities required to satisfy the
safety criteria of Sec. 53.220 if not addressed by Technical
Specifications under Sec. 53.1416(a)(25); and
(3) Inspections, tests, analyses, and acceptance criteria. The
proposed inspections, tests, and analyses, including those applicable
to emergency planning, that the licensee must perform, and the
acceptance criteria that are necessary and sufficient to provide
reasonable assurance that, if the inspections, tests, and analyses are
performed and the acceptance criteria met, the facility has been
constructed and will be operated in conformity with the COL, the
provisions of the Act, and the Commission's rules and regulations.
(i) If the application references an early site permit with ITAAC,
the early site permit ITAAC must apply to those aspects of the COL
which are approved in the early site permit.
(ii) If the application references a standard design certification,
the ITAAC contained in the certified design must apply to those
portions of the facility design which are approved in the standard
design certification.
(iii) If the application references an ML, the ITAAC contained in
the ML must apply to those portions of the facility design which are
approved in the ML.
(iv) If the application references an early site permit with ITAAC,
a standard design certification, an ML, or a combination thereof, the
application may include a notification that a required inspection,
test, or analysis in the ITAAC has been successfully completed and that
the corresponding acceptance criterion has been met. The Federal
Register notice required by Sec. 53.1422 of this chapter must indicate
that the application includes this notification.
(b) [Reserved]
Sec. 53.1422 Review of applications.
(a) Standards for review of applications. Applications filed under
this part will be reviewed according to the standards set out in 10 CFR
parts 20, 51, 53, 73, and 140.
(b) Administrative review of applications; hearings. A proceeding
on a COL is subject to all applicable procedural requirements contained
in 10 CFR part 2, including the requirements for docketing (Sec. 2.101
of this chapter) and issuance of a notice of hearing (Sec. 2.104 of
this chapter). If an applicant requests a Commission finding on certain
ITAAC with the issuance of the COL, then those ITAAC will be identified
in the notice of hearing. All hearings on COLs are governed by the
procedures contained in 10 CFR part 2.
Sec. 53.1425 Finality of referenced NRC approvals.
If the application for a COL under this part references an early
site permit, standard design certification rule, standard design
approval, or ML, issued under this part, the scope and nature of
matters resolved for the application and any COL issued are governed by
the relevant provisions addressing finality, including Sec. Sec.
53.1188, 53.1221, 53.1263, and 53.1288.
Sec. 53.1431 Referral to the Advisory Committee on Reactor
Safeguards.
The Commission must refer a copy of the application to the ACRS.
The ACRS must report on those portions of the application that concern
safety and must apply the standards referenced in Sec. 53.1422, in
accordance with the finality provisions in Sec. 53.1425.
Sec. 53.1434 Authorization to conduct limited work authorization
activities.
(a) If the application for a COL under this part does not reference
an early site permit which authorizes the holder to perform the
activities under Sec. 53.1130(b), the applicant may not perform those
activities without obtaining the separate authorization required by
Sec. 53.1130(a). Authorization may be granted only after the presiding
officer in the proceeding on the application has made the findings and
determination required by Sec. 53.1130(b)(1)(ii) and (b)(1)(iv), and
the Director, Office of Nuclear Reactor Regulation makes the
determination required by Sec. 53.1130(b)(1)(iii).
(b) If, after an applicant has performed the activities permitted
by a LWA issued under Sec. 53.1130, the application for the COL is
withdrawn or denied, then the applicant must implement the approved
site redress plan.
[[Page 15852]]
Sec. 53.1437 Exemptions, departures, and variances.
(a) An applicant for a COL, or any amendment to a COL, may include
in the application a request for an exemption from one or more of the
Commission's regulations.
(1) If the request is for an exemption from any part of a
referenced standard design certification rule, the Commission may grant
the request if it determines that the exemption complies with any
exemption provisions of the referenced standard design certification
rule, or with Sec. 53.1263 if there are no applicable exemption
provisions in the referenced standard design certification rule.
(2) For all other requests for exemptions, the Commission may grant
a request if it determines that the exemption complies with Sec.
53.080.
(b) An applicant for a COL who has filed an application referencing
an early site permit issued under Sec. 53.1158 may include in the
application a request for a variance from one or more site
characteristics, design parameters, or terms and conditions of the
permit, or from the Site Safety Analysis Report. In determining whether
to grant the variance, the Commission must apply the same technically
relevant criteria as were applicable to the application for the
original or renewed site permit. Once a COL referencing an early site
permit is issued, variances from the early site permit will not be
granted for that CP or COL.
(c) An applicant for a COL who has filed an application referencing
use of a manufactured reactor may include in the application a request
for a departure from one or more design characteristics, site
parameters, terms and conditions, or approved design of the
manufactured reactor under the ML issued under Sec. 53.1287. The
Commission may grant such a request only if it determines that the
departure will comply with the requirements of Sec. 53.080, and that
the special circumstances outweigh any decrease in safety that may
result from the reduction in standardization caused by the departure.
(d) Issuance of a variance under paragraph (b) of this section or a
departure under paragraph (c) of this section is subject to litigation
during the COL proceeding in the same manner as other issues material
to that proceeding.
Sec. 53.1440 Issuance of combined licenses.
(a)(1) After conducting a hearing under Sec. 53.1422(b) and
receiving the report submitted by the ACRS, the Commission may issue a
COL if the Commission finds that--
(i) The applicable standards and requirements of the Act and the
Commission's regulations have been met;
(ii) Any required notifications to other agencies or bodies have
been duly made;
(iii) There is reasonable assurance that the facility will be
constructed and will operate in conformity with the license, the
provisions of the Act, and the Commission's regulations;
(iv) The applicant is technically and financially qualified to
engage in the activities authorized; however, no finding of financial
qualification is necessary for an electric utility applicant for a COL;
(v) Issuance of the license will not be inimical to the common
defense and security or to the health and safety of the public; and
(vi) The findings required by 10 CFR part 51 have been made.
(2) The Commission may also find, at the time it issues the COL,
that certain acceptance criteria in one or more of the ITAAC in a
referenced early site permit, standard design certification, or ML have
been met. This finding will finally resolve that those acceptance
criteria have been met, those acceptance criteria will be deemed to be
excluded from the COL, and findings under Sec. 53.1452(g) with respect
to those acceptance criteria are unnecessary.
(b) The Commission must identify within the COL the inspections,
tests, and analyses, including those applicable to emergency planning,
that the licensee must perform, and the acceptance criteria that, if
met, are necessary and sufficient to provide reasonable assurance that
the facility has been constructed and will be operated in conformity
with the license, the provisions of the Act, and the Commission's rules
and regulations.
(c) A COL must contain the terms and conditions, including
technical specifications, as the Commission deems necessary and
appropriate.
Sec. 53.1443 Finality of combined licenses.
(a) After issuance of a COL, the Commission may not modify, add, or
delete any term or condition of the COL, the design of the facility,
the ITAAC contained in the license that are not derived from a
referenced standard design certification or ML, except under the
provisions of Sec. 53.1452 or Sec. 53.1590.
(b) If the COL does not reference a standard design certification,
then a licensee may make changes in the facility as described in the
FSAR (as updated) and make changes in the procedures as described in
the FSAR (as updated) under the applicable change processes in Sec.
53.1550.
(c) If the COL references a certified design, then--
(1) Changes to or departures from information within the scope of
the referenced standard design certification rule are subject to the
applicable change processes in that rule; and
(2) Changes that are not within the scope of the referenced
standard design certification rule are subject to the applicable change
processes in subpart I of this part, unless they also involve changes
to or noncompliance with information within the scope of the referenced
standard design certification rule. In these cases, the applicable
provisions of this section and the standard design certification rule
apply.
(d) [Reserved]
(e) The Commission may issue and make immediately effective any
amendment to a COL upon a determination by the Commission that the
amendment involves no significant hazards consideration,
notwithstanding the pendency before the Commission of a request for a
hearing from any person. The amendment may be issued and made
immediately effective in advance of the holding and completion of any
required hearing. The amendment will be processed under the procedures
specified in Sec. 53.1515.
(f) Any modification to, addition to, or deletion from the terms
and conditions of a COL, including any modification to, addition to, or
deletion from the inspections, tests, and analyses, or related
acceptance criteria contained in the license is a proposed amendment to
the license. There must be an opportunity for a hearing on the
amendment.
Sec. 53.1449 Inspection during construction.
(a) Licensee schedule for inspections, tests, or analyses. The
licensee must submit to the NRC, no later than 1 year after issuance of
the COL or at the start of construction as defined at Sec. 53.020,
whichever is later, its schedule for completing the inspections, tests,
or analyses in the ITAAC. The licensee must submit updates to the ITAAC
schedules every 6 months thereafter and, within 1 year of its scheduled
date for initial loading of fuel (or, for a fueled manufactured
reactor, within 1 year of its scheduled date for initiating the removal
of the features to prevent criticality required under Sec.
53.620(d)(1)), the licensee must submit updates to the ITAAC schedule
every 30 days until the final notification is provided to the NRC under
paragraph (c)(1) of this section.
[[Page 15853]]
(b) Licensee and applicant conduct of activities subject to ITAAC.
With respect to activities subject to an ITAAC, an applicant for a COL
may proceed at its own risk with design and procurement activities, and
a licensee may proceed at its own risk with design, procurement,
construction, and preoperational activities, even though the NRC may
not have found that any one of the prescribed acceptance criteria are
met.
(c) Licensee notifications--(1) ITAAC closure notification. The
licensee must notify the NRC that prescribed inspections, tests, and
analyses have been performed and that the prescribed acceptance
criteria are met. The notification must contain sufficient information
to demonstrate that the prescribed inspections, test, and analyses have
been performed and that the prescribed acceptance criteria are met.
(2) ITAAC post-closure notifications. Following the licensee's
ITAAC closure notifications under paragraph (c)(1) of this section
until the Commission makes the finding under Sec. 53.1452(g), the
licensee must notify the NRC, in a timely manner, of new information
that materially alters the basis for determining that either
inspections, tests, or analyses were performed as required, or that
acceptance criteria are met. The notification must contain sufficient
information to demonstrate that, notwithstanding the new information,
the prescribed inspections, tests, and analyses have been performed as
required, and the prescribed acceptance criteria are met.
(3) Uncompleted ITAAC notification. If the licensee has not
provided, by the date 225 days before the scheduled date for initial
loading of fuel (or, for a fueled manufactured reactor, by the date 225
days before the scheduled date for initiating the removal of the
features to prevent criticality required under Sec. 53.620(d)(1)), the
notification required by paragraph (c)(1) of this section for all
ITAAC, then the licensee must notify the NRC that the prescribed
inspections, tests, or analyses for all uncompleted ITAAC will be
performed and that the prescribed acceptance criteria will be met prior
to operation. The notification must be provided no later than the date
225 days before the scheduled date for initial loading of fuel (or, for
a fueled manufactured reactor, no later than the date 225 days before
the scheduled date for initiating the removal of the features to
prevent criticality required under Sec. 53.620(d)(1)), and must
provide sufficient information to demonstrate that the prescribed
inspections, tests, or analyses will be performed and the prescribed
acceptance criteria for the uncompleted ITAAC will be met, including,
but not limited to, a description of the specific procedures and
analytical methods to be used for performing the prescribed
inspections, tests, and analyses and determining that the prescribed
acceptance criteria are met.
(4) All ITAAC complete notification. The licensee must notify the
NRC that all ITAAC are complete.
(d) Licensee determination of noncompliance with ITAAC. (1) In the
event that an activity is subject to an ITAAC derived from a referenced
standard design certification and the licensee has not demonstrated
that the prescribed acceptance criteria are met, the licensee may take
corrective actions to successfully complete that ITAAC or request an
exemption from the standard design certification ITAAC, as applicable.
A request for an exemption must also be accompanied by a request for a
license amendment under subpart I.
(2) In the event that an activity is subject to an ITAAC not
derived from a referenced standard design certification and the
licensee has not demonstrated that the prescribed acceptance criteria
are met, the licensee may take corrective actions to successfully
complete that ITAAC or request a license amendment under subpart I.
(e) NRC inspection, publication of notices, and availability of
licensee notifications. The NRC must ensure that the prescribed
inspections, tests, and analyses in the ITAAC are performed.
(1) At appropriate intervals until the last date for submission of
requests for hearing under Sec. 53.1452, the NRC must publish notices
in the Federal Register of the NRC staff's determination of the
successful completion of inspections, tests, and analyses.
(2) The NRC must make publicly available the licensee notifications
under paragraph (c) of this section. The NRC must, no later than the
date of publication of the notice of intended operation required by
Sec. 53.1452(a), make publicly available those licensee notifications
under paragraph (c) of this section that have been submitted to the NRC
at least 7 days before that notice.
Sec. 53.1452 Operation under a combined license.
(a) The licensee must notify the NRC of its scheduled date for
initial loading of fuel no later than 270 days before the scheduled
date and must notify the NRC of updates to its schedule every 30 days
thereafter.\1\ Not less than 180 days before the date scheduled for
initial loading of fuel into a plant by a licensee that has been issued
a COL under this part, the Commission must publish notice of intended
operation in the Federal Register.\2\ The notice must provide that any
person whose interest may be affected by operation of the plant may,
within 60 days, request that the Commission hold a hearing on whether
the facility as constructed complies, or on completion will comply,
with the acceptance criteria in the COL, except that a hearing must not
be granted for those ITAAC that the Commission found were met under
Sec. 53.1440(a)(2).
(b) A request for hearing under paragraph (a) of this section must
show, prima facie--
(1) That one or more of the acceptance criteria of the ITAAC in the
COL have not been, or will not be, met; and
(2) The specific operational consequences of nonconformance that
would be contrary to providing reasonable assurance of adequate
protection of the public health and safety.
(c) The Commission, acting as the presiding officer, must determine
whether to grant or deny the request for hearing under the applicable
requirements of Sec. 2.309 of this chapter. If the Commission grants
the request, the Commission, acting as the presiding officer, must
determine whether during a period of interim operation there will be
reasonable assurance of adequate protection to the public health and
safety. The Commission's determination must consider the petitioner's
prima facie showing and any answers thereto. If the Commission
determines there is such reasonable assurance, it must allow operation
during an interim period under the COL.
(d) The Commission, in its discretion, must determine appropriate
hearing procedures, whether informal or formal adjudicatory, for any
hearing under paragraph (a) of this section, and must state its reasons
therefor.
(e) The Commission must, to the maximum possible extent, render a
decision on issues raised by the hearing request within 180 days of the
publication of the notice provided by paragraph (a) of this section or
by the anticipated date for initial loading of fuel into the reactor,
whichever is later.
(f) A petition to modify the terms and conditions of the COL will
be processed as a request for action under Sec. 2.206 of this chapter.
The petitioner must file the petition with the Secretary of the
Commission. Before the licensed activity allegedly affected by the
petition (fuel loading, low power testing, etc.) commences, the
Commission must determine whether any immediate action is required. If
the
[[Page 15854]]
petition is granted, then an appropriate order will be issued. Fuel
loading and operation under the COL will not be affected by the
granting of the petition unless the order is made immediately
effective.
(g) The licensee must not operate the facility until the Commission
makes a finding that the acceptance criteria in the COL are met, except
for those acceptance criteria that the Commission found were met under
Sec. 53.1440(a)(2). If the COL is for a modular design, each reactor
unit may require a separate finding as construction proceeds.
(h) After the Commission has made the finding in paragraph (g) of
this section, the ITAAC do not, by virtue of their inclusion in the
COL, constitute regulatory requirements either for licensees or for
renewal of the license; except for the specific ITAAC for which the
Commission has granted a hearing under paragraph (a) of this section,
all ITAAC expire upon final Commission action in the proceeding.
However, subsequent changes to the facility or procedures described in
the FSAR (as updated) must comply with the requirements in Sec.
53.1443(e) or (f), as applicable.
\1\ For licensees installing fueled manufactured reactors under
a COL, the COL holder must instead notify the NRC of its scheduled
date for initiating the removal of the features to prevent
criticality required under Sec. 53.620(d)(1) no later than 270 days
before the scheduled date and must notify the NRC of updates to its
schedule every 30 days thereafter.
\2\ For licensees installing fueled manufactured reactors under
a COL, the Commission must instead publish notice of intended
operation in the Federal Register not less than 180 days before the
date scheduled for initiating the removal of the features to prevent
criticality required under Sec. 53.620(d)(1).
Sec. 53.1455 Duration of combined license.
A COL is issued for a specified period not to exceed 40 years from
the date on which the Commission makes a finding that acceptance
criteria are met under Sec. 53.1452(g) or allowing operation during an
interim period under the COL under Sec. 53.1452(c).
Sec. 53.1456 Transfer of a combined license.
A COL may be transferred under Sec. 53.1570.
Sec. 53.1458 Application for renewal.
The filing of an application for a renewed license must be in
accordance with Sec. 53.1595.
Sec. 53.1461 Continuation of combined license.
Each COL for a facility that has permanently ceased operations
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 license is terminated. During this period
of continued effectiveness, the licensee must--
(a) Take actions necessary to decommission and decontaminate the
facility and continue to maintain the facility, including, where
applicable, the storage, control and maintenance of the spent fuel, in
a safe condition; and
(b) Conduct activities in accordance with all other restrictions
applicable to the facility in accordance with the NRC's regulations and
the provisions of the COL for the facility.
Sec. 53.1470 Standardization of commercial nuclear plant designs:
licenses to construct and operate nuclear power reactors of identical
design at multiple sites.
(a) Except as otherwise specified in this section, the provisions
of this section apply to CP, OL, and COL applications for commercial
nuclear plants of identical design (the ``common design'') under this
part.
(b) Each application for a CP, OL, or COL submitted pursuant to
this section must be submitted as specified in Sec. 53.1300, Sec.
53.1360, or Sec. 53.1410, respectively, and Sec. 2.101 of this
chapter. Each application must state that the applicant wishes to
construct a facility identical to a facility proposed for one or more
sites other than the applicant's (the ``common design''), and the
applicant wishes to have the application considered under this section.
Each application must list each of the other applications to be treated
together under this section or specify that such other applications
will be submitted to the NRC within 12 months of submittal of the first
application.
(c) Each application must include the information required by the
applicable sections of this subpart, provided, however, that the
application must identify the common design, and, if applicable,
reference a standard design certification or standard design approval
under this part, or the use of a reactor manufactured under this part.
The FSAR for each application must either incorporate by reference or
include the final safety analysis of the common design, including, if
applicable, the FSAR for the referenced standard design certification,
standard design approval, or the manufactured reactor.
(d) Each application submitted pursuant to this section must
contain an environmental report under Sec. 53.1312(a)(1), Sec.
53.1372(a), or Sec. 53.1419(a)(1), as applicable, that complies with
the applicable provisions of 10 CFR part 51, provided, however, that
the application may incorporate by reference a single environmental
report on the environmental impacts of the common design that are
applicable to each site.
(e) Upon a determination that each application is acceptable for
docketing under Sec. 2.101 of this chapter, each application will be
docketed and a notice of docketing for each application will be
published in the Federal Register, under Sec. 2.104 of this chapter,
provided, however, that the notice must state that the application will
be processed under the provisions of this section and subpart D of 10
CFR part 2. At the discretion of the Commission, a single notice of
docketing for multiple applications may be published in the Federal
Register.
(f) The NRC must prepare an environmental assessment or draft and
final environmental impact statements for each of the applications
under 10 CFR part 51. Scoping under Sec. Sec. 51.28 and 51.29 of this
chapter for each of the license applications may be conducted
simultaneously and joint scoping may be conducted with respect to the
environmental issues relevant to the common design. If the applications
reference a standard design certification, then the environmental
assessment or environmental impact statement for each of the
applications must incorporate by reference the standard design
certification environmental assessment. If the applications do not
reference a standard design certification, then the NRC must prepare
environmental assessments or draft and final supplemental environmental
impact statements which address severe accident mitigation design
alternatives for the common design, which must be incorporated by
reference into the environmental assessment or environmental impact
statement prepared for each application. Scoping under Sec. Sec. 51.28
and 51.29 of this chapter for the supplemental environmental impact
statement may be conducted simultaneously and may be part of the
scoping for each of the applications.
(g) The ACRS must report on each of the applications as required by
the applicable sections of this subpart. Each report must be limited to
those safety matters for each application that are not relevant to the
common design. In addition, the ACRS must separately report on the
safety of the common design, provided, however, that the report need
not address the safety of a referenced standard design certification or
reactor manufactured under this part.
(h) The Commission must designate a presiding officer to conduct
the
[[Page 15855]]
proceeding with respect to the health and safety, common defense and
security, and environmental matters relating to the common design and
affecting at least two applications. The hearing will be governed by
the applicable provisions of subparts A, C, G, L, N, and O of 10 CFR
part 2 relating to applications for CPs, OLs, and COLs. The presiding
officer must issue a partial initial decision on the common design.
(i) If the design for the power reactor(s) proposed in a particular
application is not identical to the others, that application may not be
processed under this section and subpart D of 10 CFR part 2.
(j) As used in this section, the design of a nuclear power reactor
included in a single referenced Safety Analysis Report means the design
of those SSCs important to radiological health and safety and the
common defense and security.
Subpart I--Maintaining and Revising Licensing-Basis Information
Sec. 53.1500 Licensing-basis information.
This subpart provides the requirements for each holder of a license
for a commercial nuclear plant licensed under this part to maintain
licensing-basis information as defined in Sec. 53.020; evaluate
changes to site characteristics, plant design features, and
programmatic controls to determine needed approvals and revisions; and
submit appropriate updates to the U.S. Nuclear Regulatory Commission
(NRC).
Sec. 53.1502 Specific terms and conditions of licenses.
(a) Each license issued under this part is subject to the
provisions of the Atomic Energy Act of 1954, as amended, (the Act) and
to all rules, regulations, and orders of the Commission. The terms and
conditions of the license will be subject to amendment, revision, or
modification, by reason of amendments of the Act or by reason of rules,
regulations, and orders issued in accordance with the terms of the Act.
(b) Each license issued under this part must be subject to all
conditions imposed as a matter of law by sections 401(a)(2) and 401(d)
of the Federal Water Pollution Control Act, as amended (33 U.S.C.A.
1341(a)(2) and (d)).
(c) A holder of an operating license (OL) or combined license (COL)
under this part may take reasonable action that departs from a license
condition or a technical specification included in a license issued
under this part in a national security emergency established by a law
enacted by the Congress or by an order or directive issued by the
President pursuant to statutes or the Constitution of the United
States. The authority under this paragraph (c) must be exercised in
accordance with law, including section 57e of the Act, and is in
addition to the authority granted under Sec. 53.740(h), which remains
in effect unless otherwise directed by the Commission during a national
security emergency. The authority under this paragraph (c) may be
exercised--
(1) When this action is immediately needed to implement national
security objectives as designated by the national command authority
through the Commission; and
(2) No action consistent with license conditions and technical
specifications that can satisfy national security objectives is
immediately apparent.
(d)(1) If the NRC finds that the state of emergency preparedness
does not provide reasonable assurance that adequate protective measures
can and will be taken in the event of a radiological emergency
(including findings based on requirements of 10 CFR part 50, appendix
E, section IV.D.3) and if the deficiencies (including deficiencies
based on requirements of 10 CFR part 50, appendix E, section IV.D.3)
are not corrected within 4 months of that finding, the Commission will
determine whether the facility must be shut down or cease operations
until such deficiencies are remedied or whether other enforcement
action is appropriate. In determining whether a shutdown or other
enforcement action is appropriate, the Commission will take into
account, among other factors, whether the licensee can demonstrate to
the Commission's satisfaction that the deficiencies in the plan are not
significant for the plant in question, or that adequate interim
compensating actions have been or will be taken promptly, or that that
there are other compelling reasons for continued operation.
(2) If the planning standards for radiological emergency
preparedness apply to offsite emergency response plans, or if the
planning activities in Sec. 50.160(b)(1)(iv)(B) apply, then the NRC
will base its finding on a review of the Federal Emergency Management
Agency findings and determinations as to whether State, participating
Tribal, and local emergency plans are adequate and capable of being
implemented, and on the NRC assessment as to whether the licensee's
emergency plans are adequate and capable of being implemented. Nothing
in this paragraph (d)(2) must be construed as limiting the authority of
the Commission to take action under any other regulation or authority
of the Commission or at any time other than that specified in this
paragraph (d)(2).
Sec. 53.1505 Changes to licensing-basis information requiring prior
NRC approval.
(a) Sections 53.1510 through 53.1520 provide the process for a
licensee to request and the NRC to issue amendments to licenses,
including any conditions contained therein, technical specifications or
other attachments to a license, and any orders issued by the NRC
modifying a license. Sections 53.1525 and 53.1530 govern proposed
changes to a commercial nuclear plant referencing a certified design or
manufacturing license (ML).
(b) A licensee may propose changing licensing-basis information
established by NRC regulations by requesting an exemption in accordance
with Sec. 53.080.
Sec. 53.1510 Application for amendment of license.
Whenever a holder of a license under this part desires to amend the
license, an application for an amendment must be filed with the
Commission, as specified in Sec. 53.040, that fully describes the
changes desired and, following as far as applicable, the form
prescribed for original applications. Applications for amendments
involving changes to plant structures, systems, and components (SSCs),
programmatic controls, or the role of plant personnel must include an
assessment of the changes in relation to the safety requirements in
subpart B of this part and the analyses requirements of Sec. 53.450 as
applicable, an analysis of whether the amendment involves no
significant hazards consideration using the standards in Sec. 53.1520,
and a consideration of environmental factors.
Sec. 53.1515 Public notices; State consultation.
The Commission will use the following procedures for an application
requesting an amendment to an OL or COL issued under this part.
(a) Public notices. (1)(i) The Commission may publish in the
Federal Register under Sec. 2.105 of this chapter an individual notice
of proposed action for an amendment for which it makes a proposed
determination that no significant hazards consideration is involved,
or, at least once every 30 days, publish a periodic Federal Register
notice of proposed actions, which identifies each amendment issued and
each amendment proposed to be issued since the last such periodic
notice, or it may publish both such notices.
(ii) For each amendment proposed to be issued, the notice will
[[Page 15856]]
(A) Contain the staff's proposed determination under the standards
in Sec. 53.1520;
(B) Provide a brief description of the amendment and of the
facility involved;
(C) Solicit public comments on the proposed determination; and
(D) Provide for a 30-day comment period.
(iii) The comment period will begin on the day after the date of
the publication of the first notice, and, normally, the amendment will
not be granted until after this comment period expires.
(2) The Commission may inform the public about the final
disposition of an amendment request for which it has made a proposed
determination of no significant hazards consideration either by issuing
an individual notice of issuance under Sec. 2.106 of this chapter or
by publishing such a notice in its periodic system of Federal Register
notices. In either event, it will not make and will not publish a final
determination of no significant hazards consideration unless it
receives a request for a hearing on that amendment request.
(3) Where the Commission makes a final determination that no
significant hazards consideration is involved and that the amendment
should be issued, the amendment will be effective on issuance, even if
adverse public comments have been received and even if an interested
person meeting the provisions for intervention called for in Sec.
2.309 of this chapter has filed a request for a hearing. The Commission
need hold any required hearing only after it issues an amendment,
unless it determines that a significant hazards consideration is
involved, in which case the Commission will provide an opportunity for
a prior hearing.
(4) Where the Commission finds that an emergency situation exists,
in that failure to act in a timely way would result in derating or
shutdown of a commercial nuclear reactor, or in prevention of either
resumption of operation or of increase in power output up to the
plant's licensed power level, it may issue a license amendment
involving no significant hazards consideration without prior notice and
opportunity for a hearing or for public comment. In such a situation,
the Commission will not publish a notice of proposed determination on
no significant hazards consideration but will publish a notice of
issuance under Sec. 2.106 of this chapter providing for opportunity
for a hearing and for public comment after issuance. The Commission
expects its licensees to apply for license amendments in a timely
fashion. It will decline to dispense with notice and comment on the
determination of no significant hazards consideration if it determines
that the licensee has abused the emergency provision by failing to make
timely application for the amendment and thus itself creating the
emergency. Whenever an emergency situation exists, a licensee
requesting an amendment must explain why this emergency situation
occurred and why it could not avoid this situation, and the Commission
will assess the licensee's reasons for failing to file an application
sufficiently in advance of that event.
(5) Where the Commission finds that exigent circumstances exist, in
that a licensee and the Commission must act quickly and that time does
not permit the Commission to publish a Federal Register notice allowing
30 days for prior public comment, and it also determines that the
amendment involves no significant hazards considerations, it--
(i)(A) Will either issue a Federal Register notice providing notice
of an opportunity for hearing and allowing at least 2 weeks from the
date of the notice for prior public comment; or
(B) Will use local media to provide reasonable notice to the public
in the area surrounding a licensee's facility of the licensee's
amendment and of its proposed determination as described in paragraph
(a)(1) of this section, consulting with the licensee on the proposed
media release and on the geographical area of its coverage;
(ii) Will provide for a reasonable opportunity for the public to
comment, using its best efforts to make available to the public
whatever means of communication it can for the public to respond
quickly, and, in the case of telephone comments, have these comments
recorded or transcribed, as necessary and appropriate;
(iii) When it has issued a local media release, may inform the
licensee of the public's comments, as necessary and appropriate;
(iv) Will publish a notice of issuance under Sec. 2.106 of this
chapter;
(v) Will provide a hearing after issuance, if one has been
requested by a person who satisfies the provisions for intervention
specified in Sec. 2.309 of this chapter; and
(vi) Will require the licensee to explain the exigency and why the
licensee cannot avoid it and use its normal public notice and comment
procedures in paragraph (a)(1) of this section if it determines that
the licensee has failed to use its best efforts to make a timely
application for the amendment in order to create the exigency and to
take advantage of this procedure.
(6) Where the Commission finds that significant hazards
considerations are involved, it will issue a Federal Register notice
providing an opportunity for a prior hearing even in an emergency
situation, unless it finds an imminent danger to the health or safety
of the public, in which case it will issue an appropriate order or rule
under 10 CFR part 2.
(b) State consultation. (1) At the time a licensee requests an
amendment, it must notify the State in which its facility is located of
its request by providing that State with a copy of its application and
its reasoned analysis about no significant hazards considerations and
indicate on the application that it has done so.
(2) The Commission will advise the State of its proposed
determination about no significant hazards consideration normally by
sending it a copy of the Federal Register notice.
(3) The Commission will make the names of the Project Manager or
other NRC personnel it designated to consult with the State available
to the State official designated to consult about its proposed
determination. The Commission will consider any comments of that State
official. If it does not hear from the State in a timely manner, it
will consider that the State has no interest in its determination;
nonetheless, to ensure that the State is aware of the application,
before it issues the amendment, it will make a good faith effort to
communicate directly with that official. (Inability to consult with a
responsible State official following good faith attempts will not
prevent the Commission from making effective a license amendment
involving no significant hazards consideration.)
(4) The Commission will make a good faith attempt to consult with
the State before it issues a license amendment involving no significant
hazards consideration. If, however, it does not have time to use its
normal consultation procedures because of an emergency situation, it
will attempt to communicate directly with the appropriate State
official. (Inability to consult with a responsible State official
following good faith attempts will not prevent the Commission from
making effective a license amendment involving no significant hazards
consideration, if the Commission deems it necessary in an emergency
situation.)
(5) After the Commission issues the requested amendment, it will
send a copy of its determination to the State.
(c) Caveats about State consultation. (1) The State consultation
procedures in
[[Page 15857]]
paragraph (b) of this section do not give the State a right--
(i) To veto the Commission's proposed or final determination;
(ii) To a hearing on the determination before the amendment becomes
effective; or
(iii) To insist upon a postponement of the determination or upon
issuance of the amendment.
(2) These procedures do not alter present provisions of law that
reserve to the Commission exclusive responsibility for setting and
enforcing radiological health and safety requirements for commercial
nuclear plants.
Sec. 53.1520 Issuance of amendment.
(a) In determining whether an amendment to a license will be issued
to the applicant, the Commission will be guided by the considerations
which govern the issuance of initial licenses to the extent applicable
and appropriate. If the application is for amendment of an OL or COL
and involves the material alteration of a commercial nuclear plant, a
construction permit (CP) will be issued before the issuance of the
amendment to the license, provided however, that if the application
involves a material alteration to a manufactured reactor under this
part before its installation at a site, or a COL before the date that
the Commission makes the finding under Sec. 53.1452(g), no application
for or issuance of a CP is required. If the amendment involves a
significant hazards consideration, the Commission will give notice of
its proposed action--
(1) Under Sec. 2.105 of this chapter before acting thereon; and
(2) As soon as practicable after the application has been docketed.
(b) 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 commercial nuclear plant).
(c) The Commission may make a final determination, under the
procedures in Sec. 53.1515, that a proposed amendment to an OL or a
COL for a commercial nuclear plant under this part involves no
significant hazards consideration, if operation of the plant in
accordance with the proposed amendment would not--
(1) Involve a significant increase in the probability or
consequences of an accident previously evaluated; or
(2) Create the possibility of a new or different kind of an
accident from any accident previously evaluated; or
(3) Involve a significant reduction in a margin of safety.
Sec. 53.1525 Revising certification information within a design
certification rule.
(a) A holder of an OL or COL who references a design certification
rule issued under this part must request an exemption if proposing to
change one or more elements of the certification information. The
Commission may grant such a request only if it determines that the
exemption will comply with the requirements of Sec. 53.080 and that
the special circumstances outweigh any decrease in safety that may
result from the reduction in standardization caused by the departure.
(b) The request for an exemption must be included with any
associated license amendment request, which must be requested and
processed in accordance with Sec. Sec. 53.1510, 53.1515, and 53.1520.
(c) Licensees must evaluate changes to the design as described in
the Final Safety Analysis Report (FSAR) not involving changes to the
certification information using the criteria in Sec. 53.1550.
Sec. 53.1530 Revising information within a Final Safety Analysis
Report associated with a manufacturing license.
(a) The holder of an ML may make changes to the facility or
procedures as described in the Final Safety Analysis Report (FSAR)
associated with the ML without obtaining a license amendment pursuant
to Sec. 53.1510 if the change meets the criteria in Sec.
53.1550(a)(1) and (2) using the specifications in Sec. 53.1550(b). If
needed, applications for amending an ML must be submitted and processed
in accordance with Sec. Sec. 53.1510, 53.1515, and 53.1520. In those
cases where an ML references a design certification rule, the amendment
application from the holder of the ML must also request an exemption
from the design certification rule under Sec. 53.1525 if proposing to
change one or more elements of the certification information.
(b)(1) The holder of an ML must maintain records of changes to the
facility or procedures made without a license amendment under paragraph
(a) of this section. These records must include a written evaluation
which provides the bases for the determination that the change does not
require a license amendment under the criteria in paragraph Sec.
53.1550(a)(2).
(2) The records of changes in the facility must be maintained until
the expiration of an ML issued under this part, or the expiration of a
renewed license issued under Sec. 53.1295--whichever is later. Records
of changes in procedures must be maintained for a period of 5 years.
Sec. 53.1535 Amendments during construction.
(a) The holder of a CP or limited work authorization (LWA) under
this part may request an amendment to the CP or LWA in order to gain
Commission approval of the safety of selected design features or
specifications, including proposed departures from a design
certification rule or ML. Amendments to CPs or LWAs under this part
must be requested and processed under Sec. Sec. 53.1510 and 53.1520.
(b) The holder of a COL under this part for which the NRC has not
yet made a finding in accordance with Sec. 53.1452(g) must request
amendments required by Sec. 53.1525 or Sec. 53.1550 no later than 45
days from the date the licensee begins the construction of the SSCs to
implement the change or departure requiring NRC approval. The licensee
proceeds with such changes at its own risk recognizing that there is a
possibility that the amendment will not be granted.
Sec. 53.1540 Updating licensing-basis information and determining the
need for NRC approval.
(a) Sections 53.1545 through 53.1565 provide the process for a
holder of an OL or COL to modify licensing-basis information and to
evaluate potential changes to its facilities, procedures, programs, and
organizations to determine if NRC approval is required.
(b) Definitions for the purposes of Sec. Sec. 53.1545 through
53.1565--
Change means a modification or addition to, or removal from, the
commercial nuclear plant or procedures that affects a design feature or
related functional design criteria, method of performing or controlling
the functions of design features, or an evaluation that demonstrates
that intended functions will be accomplished.
Departure from a method of evaluation described in the Final Safety
Analysis Report (FSAR) (as updated) used in establishing the functional
design criteria for safety-related structures, systems, or components
or in the safety analyses means--
Changing any of the elements of the method described in the FSAR
(as updated) unless the results of the analysis are conservative or
essentially the same; or
Changing from a method described in the FSAR to another method
unless that method has been approved by NRC for the intended
application.
Facility as described in the FSAR (as updated) means--
The SSCs that are described in the FSAR (as updated),
[[Page 15858]]
The design and performance requirements for such SSCs described in
the FSAR (as updated), and
The evaluations or methods of evaluation included in the FSAR (as
updated) for such SSCs which demonstrate that their intended
function(s) will be accomplished.
Final Safety Analysis Report (as updated) means the FSAR submitted
under Sec. 53.1279, Sec. 53.1369 or Sec. 53.1416, as amended and
supplemented, and as updated under Sec. 53.1545, as applicable.
Procedures as described in the Final Safety Analysis Report (as
updated) means those procedures that contain information described in
the FSAR (as updated) such as how SSCs are operated and controlled
(including assumed operator actions and response times).
Sec. 53.1545 Updating Final Safety Analysis Reports.
(a) Each holder of an OL or COL under this part for which the
Commission has made the finding under Sec. 53.1452(g) must update the
FSAR originally submitted as part of the application for the license
every 24 months or more frequently to assure that the information
included in the report contains the latest information developed. The
submittal must include the effects on the content of the FSAR of--
(1) Changes made to the facility or procedures as described in the
FSAR;
(2) Safety analyses and evaluations performed by the licensee
either in support of approved license amendments or in support of
conclusions that changes did not require a license amendment under
Sec. 53.1550;
(3) Updates to the probabilistic risk assessment (PRA), other
systematic risk evaluations, or a combination thereof required under
Sec. 53.450(a);
(4) The cumulative effects of the changes to the facility or
procedures on the margins to the safety criteria in Sec. Sec. 53.210,
53.220, and 53.450(e) since the last FSAR update; and
(5) Analyses of new safety issues performed by or on behalf of the
licensee at Commission request.
(b)(1) The licensee must submit revisions containing updated
information to the Commission, under Sec. 53.040, identifying the
location of revised or new information.
(2) The submittal must include--
(i) 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
(ii) An identification of changes made under the provisions of
Sec. 53.1550 but not previously submitted to the Commission.
(c) Each applicant for or holder of a COL under this part for which
the Commission has not made the finding under Sec. 53.1452(g) must
submit an update to the FSAR annually by providing the information
required in paragraphs (a)(1) through (a)(5) of this section and
meeting the requirements of paragraph (b) of this section. Combined
license applicants who have requested the NRC to suspend its review of
the COL application and COL holders who have informed the NRC that they
do not plan to pursue construction need not submit an annual update of
the FSAR. If a COL applicant requests that the NRC resume its review,
or a COL holder notifies the NRC that the COL holder plans to commence
or resume construction, then the COL applicant or holder must submit to
NRC an update to its FSAR within 90 days of the request or
notification, as applicable, and annually thereafter.
(d) The FSAR (as updated) must be retained by the licensee until
the Commission terminates its license.
(e) Each holder of an ML under this part must submit an update of
the FSAR every 24 months or more frequently as necessary to facilitate
dependent COL or CP applications. The submittal must include the
effects of changes on the content of the FSAR as described in
paragraphs (a)(1), (a)(3) through (a)(5), and (b) of this section and
safety analyses and evaluations performed by the licensee either in
support of approved license amendments or in support of conclusions
that changes did not require a license amendment under Sec. 53.1530.
Sec. 53.1550 Evaluating changes to facility as described in Final
Safety Analysis Reports.
(a) The holder of an OL or COL may make changes in the facility as
described in the FSAR (as updated) and make changes in the procedures
as described in the FSAR (as updated) without obtaining a license
amendment pursuant to Sec. 53.1510 only if--
(1) A change to the technical specifications incorporated in the
license is not required; and
(2) The change meets all of the following criteria:
(i) Does not result in an increase to the frequency or consequences
of an event sequence such that an event sequence not previously
identified as risk significant becomes risk significant by the analyses
performed in accordance with Sec. 53.450(e).
(ii) Does not result in an increase to the frequency or
consequences of an event sequence such that an event sequence exceeds
the licensing-basis event evaluation criteria required to be
established in accordance with Sec. 53.450(e).
(iii) Does not involve either of the following:
(A) A change to the NRC-approved comprehensive risk metric(s) or
associated risk performance objective under Sec. 53.220(b); or
(B) An increase to the frequency or consequences of one or more
event sequences such that any calculated comprehensive risk metric
exceeds the associated risk performance objective established in
accordance with Sec. 53.220.
(iv) Does not involve a departure from a method of evaluation
described in the FSAR (as updated) used in assessing design-basis
accidents in accordance with Sec. 53.450(f) unless the results of the
analysis under Sec. 53.450(f) are conservative or essentially the
same, the revised method of evaluation has been previously approved by
the NRC for the intended application, or the revised method of
evaluation can be used under an NRC-endorsed consensus code or
standard.
(v) Does not result in a change to the safety classification of an
SSC from safety-related to either non-safety-related but safety-
significant or non-safety-related.
(vi) Does not result in more than a minimal decrease in defense in
depth.
(vii) [Reserved]
(viii) Does not result in the identification of a new design-basis
accident in accordance with Sec. 53.450(f).
(ix) Does not result in more than a minimal increase in the
consequences of any design-basis accident.
(3) In implementing this paragraph (a), the FSAR (as updated) is
considered to include FSAR changes since submittal of the last update
of the FSAR under Sec. 53.1545.
(4) The provisions in this section do not apply to changes to the
facility or procedures when the applicable regulations establish more
specific criteria for accomplishing such changes.
(b)(1) A licensee who references a design certification rule may
make departures from the standard design, without prior Commission
approval, unless the proposed departure involves a change to the design
as described in the rule certifying the design, in which case the
requirements of Sec. 53.1525 are applicable.
(2) The licensee must maintain records of all departures from the
certified design of the facility and these records must be maintained
and
[[Page 15859]]
available for audit until the termination of the license. The licensee
must identify the location and nature of departures from licensing-
basis information within supporting documents for a certified design
within the updates to the Safety Analysis Report required by Sec.
53.1545.
(3) Licensees for which the NRC has docketed the certifications
required under Sec. 53.1070 need not retain records of departures from
the design of the facility associated with SSCs that have been
permanently removed from service using an NRC-approved change process.
(c)(1) The licensee must maintain records of changes in the
facility and procedures made under paragraph (a) of this section. These
records must include a written evaluation which provides the bases for
the determination that the change does not require a license amendment
under paragraph (a)(2) of this section.
(2) The licensee must submit, as specified in Sec. 53.040, a
report containing a brief description of any departures and changes,
including a summary of the evaluation of each. A report must be
submitted at intervals not to exceed 24 months. For COLs, the report
must be submitted at intervals not to exceed 6 months during the period
from the date of application for a COL to the date the Commission makes
its findings under Sec. 53.1452(g).
(3) The records of changes in the facility must be maintained until
the termination of an OL or COL issued under this part, or the
termination of a renewed license issued under Sec. 53.1595--whichever
is later. Records of changes in procedures must be maintained for a
period of 5 years.
Sec. 53.1560 Updating program documents included in licensing-basis
information.
(a) Each holder under this part of an OL or COL for which the
Commission has made the finding under Sec. 53.1452(g) must biennially
or more frequently update the program documents submitted as part of an
application to obtain or maintain the license to assure that the
information included in the documents contains the latest information
developed. The submittals must include the effects on the content of
the program documents of--
(1) Changes made in the facility, procedures, licensee's
organization, or site environs;
(2) 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. 53.1550;
(3) Analyses of new safety issues performed by or on behalf of the
licensee at Commission request; and
(4) Changes to the programs as a result of operating experience,
corrective actions, or other reasons deemed appropriate to ensure the
programs serve their underlying purpose to support the requirements in
subpart B of this part or other NRC regulations.
(b)(1) The licensee must submit revisions containing updated
information to the Commission, as specified in Sec. 53.040,
identifying the location of revised or new information.
(2) The submittal must include--
(i) A certification by a duly authorized officer of the licensee
that either the information accurately presents changes made since the
previous submittals, necessary to reflect information and analyses
submitted to the Commission or prepared pursuant to Commission
requirement, or that no such changes were made; and
(ii) An identification of changes made under the provisions of
Sec. 53.1550 but not previously submitted to the Commission.
(c) The updated program documents must be retained by the licensee
until the Commission terminates their license.
Sec. 53.1565 Evaluating changes to programs included in licensing-
basis information.
(a) A licensee may make changes to the facility, procedures, or
organizations or address changes to site environs as described in the
program documents included in licensing-basis information without
obtaining prior NRC approval only if--
(1) A change to the technical specifications incorporated in the
license is not required;
(2) An exemption from an NRC regulation is not required; and
(3) The change conforms to program-specific requirements included
in regulations in this part, technical specifications, or the NRC-
approved program document included and reviewed as part of a license
application under subpart H or an amendment under this subpart.
(b) In implementing this section, the program documents (as
updated) include changes since submittal of the last updates of the
program documents pursuant to Sec. 53.1560.
(c) The provisions in this section do not apply to changes to the
program documents when the applicable regulations establish more
specific criteria for accomplishing such changes.
(d) To make changes to the facility, procedures, or organizations
or to address changes to site environs as described in the program
documents included in licensing-basis information for individual
programs, the following requirements must be satisfied:
(1) Quality assurance program--operation. (i) Each holder under
this part of an OL or COL, after the Commission makes the finding under
Sec. 53.1452(g), may make a change to a previously accepted quality
assurance program (QAP) description included or referenced in the
Safety Analysis Report without prior NRC approval, provided the change
does not reduce the commitments in the program description as accepted
by the NRC. Changes to the QAP description that do not reduce the
commitments must be submitted to the NRC in accordance with the
requirements of Sec. 53.1545. In addition to QAP changes involving
administrative improvements and clarifications, spelling corrections,
punctuation, or editorial items, the following changes are not
considered to be reductions in commitment:
(A) The use of a quality assurance (QA) standard approved by the
NRC which is more recent than the QA standard in the licensee's QAP at
the time of the change;
(B) The use of a QA alternative or exception approved by an NRC
safety evaluation, provided that the bases of the NRC approval are
applicable to the licensee's facility;
(C) The use of generic organizational position titles that clearly
denote the position function, supplemented as necessary by descriptive
text, rather than specific titles;
(D) The use of generic organizational charts to indicate functional
relationships, authorities, and responsibilities, or, alternately, the
use of descriptive text;
(E) The elimination of QAP information that duplicates language in
QA regulatory guides and QA standards to which the licensee is
committed; and
(F) Organizational revisions that ensure that persons and
organizations performing QA functions continue to have the requisite
authority and organizational freedom, including sufficient independence
from cost and schedule when opposed to safety considerations.
(ii) Changes to the QAP description that do reduce the commitments
must be submitted to the NRC and receive NRC approval prior to
implementation, as follows:
(A) Changes made to the QAP description as presented in the Safety
Analysis Report or in a topical report must be submitted as specified
in Sec. 53.040.
[[Page 15860]]
(B) The submittal of a change to the Safety Analysis Report QAP
description must include all pages affected by that change and must be
accompanied by a forwarding letter identifying the change, the reason
for the change, and the basis for concluding that the revised program
incorporating the change continues to satisfy the criteria of appendix
B to part 50 of this chapter and the Safety Analysis Report QAP
description commitments previously accepted by the NRC (the letter need
not provide the basis for changes that correct spelling, punctuation,
or editorial items).
(C) A copy of the forwarding letter identifying the change must be
maintained as a facility record for 3 years.
(D) Changes to the QAP description included or referenced in the
Safety Analysis Report shall be regarded as accepted by the Commission
upon receipt of a letter to this effect from the appropriate reviewing
office of the Commission or 60 days after submittal to the Commission,
whichever occurs first.
(2) Quality assurance program--siting, construction, and
manufacturing. Each holder of an LWA, early site permit, CP, ML, or
COL, before the Commission makes the finding under Sec. 53.1452(g) of
this chapter, under this part may make a change to a previously
accepted QAP description included or referenced in the Safety Analysis
Report without prior NRC approval, provided the change does not reduce
the commitments in the program description previously accepted by the
NRC. Changes to the QAP description that do not reduce the commitments
must be submitted to NRC within 90 days. Changes to the QAP description
that reduce the commitments must be submitted to NRC and receive NRC
approval before implementation, as follows:
(i) Changes to the Safety Analysis Report must be submitted for
review as specified in Sec. 53.040. Changes made to NRC-accepted QA
topical report descriptions must be submitted as specified in Sec.
53.040.
(ii) The submittal of a change to the Safety Analysis Report QAP
description must include all pages affected by that change and must be
accompanied by a forwarding letter identifying the change, the reason
for the change, and the basis for concluding that the revised program
incorporating the change continues to satisfy the criteria of appendix
B of part 50 of this chapter and the Safety Analysis Report QAP
description commitments previously accepted by the NRC (the letter need
not provide the basis for changes that correct spelling, punctuation,
or editorial items).
(iii) A copy of the forwarding letter identifying the changes must
be maintained as a facility record for 3 years.
(iv) Changes to the QAP description included or referenced in the
Safety Analysis Report shall be regarded as accepted by the Commission
upon receipt of a letter to this effect from the appropriate reviewing
office of the Commission or 60 days after submittal to the Commission,
whichever occurs first.
(3) Emergency preparedness program. (i) Definitions for the purpose
of paragraph (d)(3) of this section:
(A) Change means an action that results in modification or addition
to, or removal from, the licensee's emergency plan. All such changes
are subject to the provisions of this section except where the
applicable regulations establish specific criteria for accomplishing a
particular change.
(B) Emergency plan means the document(s), prepared and maintained
by the licensee, that identify and describe the licensee's methods for
maintaining emergency preparedness and responding to emergencies. An
emergency plan includes the plan as originally approved by the NRC and
all subsequent changes made by the licensee with, and without, prior
NRC review and approval under paragraph (d)(3) of this section.
(C) Emergency planning function means a capability or resource
necessary to prepare for and respond to a radiological emergency.
(D) Reduction in effectiveness means a change in an emergency plan
that results in reducing the licensee's capability to perform an
emergency planning function in the event of a radiological emergency.
(ii)(A) Except as provided in paragraph (d)(3)(ii)(B) of this
section, a holder of an OL under this part, or a COL under this part
after the Commission makes the finding under Sec. 53.1452(g), must
follow and maintain the effectiveness of an emergency plan that meets
the requirements in appendix E to part 50 of this chapter and the
planning standards of Sec. 50.47(b).
(B) A holder of an OL under this part for a commercial nuclear
plant consisting of small modular reactors (SMRs) or non-light-water
reactors, or a holder of a COL under this part after the Commission
makes the finding under Sec. 53.1452(g) for a commercial nuclear plant
consisting of either SMRs or non-light-water reactors, must follow and
maintain the effectiveness of either an emergency plan that meets the
requirements in Sec. 50.160 or an emergency plan that meets the
requirements in appendix E to part 50 of this chapter and the planning
standards of Sec. 50.47(b).
(iii)(A) Except as provided in paragraph (d)(3)(iii)(B) of this
section, the licensee may make changes to its emergency plan without
NRC approval only if the licensee performs and retains an analysis
demonstrating that the changes do not reduce the effectiveness of the
plan and the plan, as changed, continues to meet the requirements in
appendix E to part 50 of this chapter and the planning standards of
Sec. 50.47(b).
(B) A license under this part for a commercial nuclear plant
consisting of either SMRs or non-light-water reactors may make changes
to its emergency plan without NRC approval only if the licensee
performs and retains an analysis demonstrating that the changes do not
reduce the effectiveness of the plan and the plan, as changed,
continues to meet either the requirements in Sec. 50.160 or the
requirements in appendix E to part 50 and the planning standards of
Sec. 50.47(b).
(iv) The changes to a licensee's emergency plan that reduce the
effectiveness of the plan as defined in paragraph (d)(3)(i)(D) of this
section may not be implemented without prior approval by the NRC. A
licensee desiring to make such a change must submit an application for
an amendment to its license. In addition to the filing requirements of
Sec. Sec. 53.1510 and 53.1515, the request must include all emergency
plan pages affected by that change and must be accompanied by a
forwarding letter identifying the change, the reason for the change,
and the basis for concluding that the licensee's emergency plan, as
revised, will continue to meet either the requirements in Sec. 50.160
to this chapter or the requirements in appendix E to part 50 of this
chapter and the planning standards of Sec. 50.47(b) of this chapter.
(v) The licensee must retain a record of each change to the
emergency plan made without prior NRC approval for a period of three
years from the date of the change and shall submit, as specified in
Sec. 53.040, a report of each such change, including a summary of its
analysis, within 30 days after the change is put in effect.
(vi) The licensee must retain the emergency plan and each change
for which prior NRC approval was obtained pursuant to paragraph
(d)(3)(iv) of this section as a record until the Commission terminates
the license for the nuclear power reactor.
[[Page 15861]]
(vii)(A) The licensee must provide for the development, revision,
implementation, and maintenance of its emergency preparedness program.
The licensee must ensure that all program elements are reviewed by
persons who have no direct responsibility for the implementation of the
emergency preparedness program either--
(1) At intervals not to exceed 12 months; or
(2) As necessary, based on an assessment by the licensee against
performance indicators, and as soon as reasonably practicable after a
change occurs in personnel, procedures, equipment, or facilities that
potentially could adversely affect emergency preparedness, but no
longer than 12 months after the change. In any case, all elements of
the emergency preparedness program must be reviewed at least once every
24 months.
(B) The review must include an evaluation for adequacy of
interfaces with State, participating Tribal, and local governments and
of licensee drills, exercises, capabilities, and procedures. The
results of the review, along with recommendations for improvements,
must be documented, reported to the licensee's corporate and plant
management, and retained for a period of 5 years. The part of the
review involving the evaluation for adequacy of interface with State,
participating Tribal, and local governments must be available to the
appropriate State, participating Tribal, and local governments.
(4) Security programs. (i) The licensee must prepare and maintain
safeguards contingency plan procedures in accordance with appendix C of
part 73 of this chapter for affecting the actions and decisions
contained in the Responsibility Matrix of the safeguards contingency
plan. The licensee may not make a change that would decrease the
safeguard effectiveness of a physical security plan, or guard training
and qualification plan, or cybersecurity plan submitted under subpart H
or part 73 of this chapter, or of the first four categories of
information (Background, Generic Planning Base, Licensee Planning Base,
Responsibility Matrix) contained in a licensee safeguards contingency
plan submitted under subpart H or part 73 of this chapter, as
applicable, without prior approval of the Commission. A licensee
desiring to make such a change must submit an application for amendment
to the licensee's license under Sec. Sec. 53.1510, 53.1515, and
53.1520.
(ii) The licensee may make changes to the plans referenced in
paragraph (d)(4)(i) of this section without prior Commission approval
if the changes do not decrease the safeguards effectiveness of the
plan. The licensee must maintain records of changes to the plans made
without prior Commission approval for a period of 3 years from the date
of the change, and must submit, as specified in Sec. 53.040, a report
containing a description of each change within 2 months after the
change is made. Prior to the safeguards contingency plan being put into
effect, the licensee must have--
(A) All safeguards capabilities specified in the safeguards
contingency plan available and functional;
(B) Detailed procedures developed according to appendix C to part
73 of this chapter available at the licensee's site; and
(C) All appropriate personnel trained to respond to safeguards
incidents as outlined in the plan and specified in the detailed
procedures.
(iii) The licensee must provide for the development, revision,
implementation, and maintenance of its safeguards contingency plan. The
licensee must ensure that all program elements are reviewed by
individuals independent of both security program management and
personnel who have direct responsibility for implementation of the
security program either--
(A) At intervals not to exceed 12 months; or
(B) As necessary, based on an assessment by the licensee against
performance indicators, and as soon as reasonably practicable after a
change occurs in personnel, procedures, equipment, or facilities that
potentially could adversely affect security, but no longer than 12
months after the change. In any case, all elements of the safeguards
contingency plan must be reviewed at least once every 24 months.
(iv) The review must include a review and audit of safeguards
contingency procedures and practices, an audit of the security system
testing and maintenance program, and a test of the safeguards systems
along with commitments established for response by local law
enforcement authorities. The results of the review and audit, along
with recommendations for improvements, must be documented, reported to
the licensee's corporate and plant management, and kept available at
the plant for inspection for a period of 3 years.
Sec. 53.1570 Transfer of licenses.
(a) No commercial nuclear plant license issued under this part, or
any right thereunder, shall be transferred, assigned, or in any manner
disposed of, either voluntarily or involuntarily, directly or
indirectly, through transfer of control of the license to any person,
unless the Commission gives its consent in writing.
(b)(1) An application for transfer of a license must include--
(i) As much of the information described in Sec. Sec. 53.1109,
53.1306, 53.1366, and 53.1413 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 license. 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.
(ii) A statement of the purposes for which the transfer of the
license is requested, the nature of the transaction necessitating or
making desirable the transfer of the license, and an agreement to limit
access to Restricted Data or Classified National Security Information
pursuant to Sec. 53.1115. The Commission may require any person who
submits an application for 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 Act and these regulations) to possession of the
facility or site involved.
(2) [Reserved]
(c) After appropriate notice to interested persons, including the
existing licensee, and observance of such procedures as may be required
by the Act or regulations or orders of the Commission, the Commission
will approve an application for the transfer of a 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 license is otherwise consistent with
applicable provisions of law, regulations, and orders issued by the
Commission pursuant thereto.
Sec. 53.1575 Termination of licenses.
(a) When the holder of an OL or COL under this part 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. 53.1070.
(b) Once fuel has been permanently removed from the reactor system,
the licensee must submit a written
[[Page 15862]]
certification to the NRC that meets the requirements of Sec. 53.1070.
(c)(1) Upon docketing of the certifications for permanent cessation
of operations and permanent removal of fuel from the reactor system, or
when a final legally effective order to permanently cease operations
has come into effect, the license no longer authorizes operation of the
reactor or emplacement or retention of fuel into the reactor system.
(2) Activities associated with decommissioning will be carried out
in accordance with the requirements and procedures in subpart G of this
part.
(3) The Commission shall terminate the license if it determines
that--
(i) The remaining dismantlement has been performed in accordance
with the approved license termination plan required in subpart G of
this part; and
(ii) The final radiation survey and associated documentation,
including an assessment of dose contributions associated with parts
released for use before approval of the license termination plan,
demonstrate that the facility and site have met the criteria for
decommissioning in subpart E of 10 CFR part 20.
(d) A holder of a CP or COL under this part may request the
termination of the license as well as licenses issued by the NRC under
parts 30, 40, or 70 of this chapter prior to plant operations. Such
requests may support an immediate NRC approval of the site for
unrestricted use.
Sec. 53.1580 Information requests.
Each licensee under this part must at any time before termination
of the license, upon request of the Commission, submit, as specified in
Sec. 53.040 written statements, signed under oath or affirmation, to
enable the Commission to determine whether or not the license should be
modified, suspended, or revoked. Except for information sought to
verify licensee compliance with the current licensing basis for that
facility, the NRC must prepare the reason or reasons for each
information request prior to 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 his
or her designee prior to issuance of the request.
Sec. 53.1585 Revocation, suspension, modification of licenses and
approvals for cause.
A license or standard design approval issued under this part may be
revoked, suspended, or modified, in whole or in part, for any material
false statement in the application or in the supplemental or other
statement of fact required of the applicant; or because of conditions
revealed by the application or statement of fact of any report, record,
inspection, or other means which would warrant the Commission to refuse
to grant a license or approval on an original application; or for
failure to manufacture a reactor, or construct or operate a facility in
accordance with the terms of the license, provided, however, that
failure to make timely completion of the proposed construction or
alteration of a facility under a CP under this part shall be governed
by the provisions of Sec. 53.1342(b); or for violation of, or failure
to observe, any of the terms and provisions of the Act, regulations,
license, approval, or order of the Commission.
Sec. 53.1590 Backfitting.
(a)(1) Backfitting means the modification of or addition to
systems, structures, components, or design of a facility; or the design
approval 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 the date of the commercial nuclear
plant license issued under this part.
(2) Except as provided in paragraph (a)(4) of this section, the
Commission shall require a systematic and documented analysis pursuant
to paragraph (b) of this section for backfits which it seeks to impose.
(3) Except as provided in paragraph (a)(4) of this section, the
Commission shall require the backfitting of a facility only when it
determines, based on the analysis described in paragraph (b) 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.
(4) The provisions of paragraphs (a)(2) and (a)(3) of this section
are inapplicable and, therefore, backfit analysis is not required and
the standards in paragraph (a)(3) of this section do not apply where
the Commission or staff, as appropriate, finds and declares, with
appropriate documented evaluation for its finding, either--
(i) 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
(ii) 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
(iii) 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.
(5) The Commission must 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.
(6) The documented evaluation required by paragraph (a)(4) 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.
(7) 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.
(b) In reaching the determination required by paragraph (a)(3) of
this section, the Commission will consider how the backfit should be
scheduled in light of 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:
[[Page 15863]]
(1) The statement of the specific objectives that the proposed
backfit is designed to achieve;
(2) The general description of the activity that would be required
by the licensee or applicant in order to complete the backfit;
(3) The potential change in the risk to the public from the
accidental off-site release of radioactive material;
(4) The potential impact on radiological exposure of facility
employees;
(5) The installation and continuing costs associated with the
backfit, including the cost of facility downtime or the cost of
construction delay;
(6) The potential safety impact of changes in plant or operational
complexity, including the relationship to proposed and existing
regulatory requirements;
(7) The estimated resource burden on the NRC associated with the
proposed backfit and the availability of such resources;
(8) The potential impact of differences in facility type, design or
age on the relevancy and practicality of the proposed backfit;
(9) Whether the proposed backfit is interim or final and, if
interim, the justification for imposing the proposed backfit on an
interim basis.
(c) No licensing action will be withheld during the pendency of
backfit analyses required by the Commission's rules.
(d) The Executive Director for Operations shall be responsible for
implementation of this section, and all analyses required by this
section shall be approved by the Executive Director for Operations or
his or her designee.
Sec. 53.1595 Renewal.
Licenses may be renewed by the Commission upon expiration of the
period of the license.
Subpart J--Reporting and Other Administrative Requirements
Sec. 53.1600 General information.
Each applicant and licensee under this part must ensure that U.S.
Nuclear Regulatory Commission (NRC) inspectors have unfettered access
to sites and facilities licensed or proposed to be licensed in Sec.
53.1610, must maintain records and make reports to the NRC in
accordance with requirements in Sec. Sec. 53.1620 through 53.1650,
must satisfy financial qualification and reporting requirements in
Sec. Sec. 53.1660 through 53.1700, and must obtain and maintain
required financial protections in case of an accident in Sec. Sec.
53.1720 and 53.1730.
Sec. 53.1610 Unfettered access for inspections.
(a) Each applicant for or holder of a manufacturing license (ML),
operating license (OL), combined license (COL), construction permit
(CP), or early site permit must permit inspection, by duly authorized
representatives of the Commission, of its records, premises,
activities, and of licensed materials in possession or use, related to
the license or CP or early site permit as may be necessary to
effectuate the purposes of the Atomic Energy Act of 1956, as amended,
(the Act) and the Energy Reorganization Act of 1974, as amended.
(b)(1) Each holder of an ML, OL, COL, or CP must, upon request by
the Director, Office of Nuclear Reactor Regulation, provide rent-free
office space for the exclusive use of the Commission inspection
personnel. Heat, air conditioning, light, electrical outlets, and
janitorial services must be furnished by each licensee and each holder
of a CP. The office must be convenient to and have full access to the
facility and must provide the inspectors both visual and acoustic
privacy.
(2) For a site or facility with an assigned resident inspector, the
space provided must be adequate to accommodate a full-time inspector, a
part-time secretary, and transient NRC personnel and must be generally
commensurate with other office facilities at the site. For sites or
facilities assigned multiple resident inspectors, additional space may
be requested. The office space that is provided must be subject to the
approval of the Director, Office of Nuclear Reactor Regulation. All
furniture, supplies, and communication equipment will be furnished by
the Commission.
(3) For a site or facility without an assigned resident inspector,
temporary space to accommodate periodic or special inspections must be
provided. The office space must be generally commensurate with other
office accommodations at the site.
(4) The licensee or permit holder must afford any NRC resident
inspector assigned to that site, or other 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.
(5) 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)(4) of this section, is not
announced or otherwise communicated by its employees or contractors to
other persons at the facility unless specifically requested by the NRC
inspector.
Sec. 53.1620 Maintenance of records, making of reports.
(a) Each holder of an ML, OL, COL, CP, or early site 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 Act and the Energy Reorganization Act of 1974, as
amended. Reports must be submitted in accordance with Sec. 53.040.
(b) [Reserved]
(c) Records that are required by the regulations in 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 or, in the case of an early site
permit, until the permit expires.
(d)(1) Records which 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.
(2) 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 in this part for such records shall apply
unless the Commission, under Sec. 53.080 of this part, has granted a
specific exemption from the record retention requirements in the
regulations in this part.
(e) Each licensee must notify the Commission as specified in Sec.
53.040 of
[[Page 15864]]
this part, of successfully completing power ascension testing or
startup testing as applicable within 30 calendar days of completing the
testing.
Sec. 53.1630 Immediate notification requirements for operating
commercial nuclear plants.
(a) General requirements.\1\ (1) Each holder of an OL under this
part or a COL under this part after the Commission makes the finding
under Sec. 53.1452(g), must notify the NRC Headquarters Operations
Center via the Emergency Notification System (ENS) of--
(i) The declaration of any of the Emergency Classes specified in
the licensee's approved Emergency Plan; or
(ii) Those non-emergency events specified in paragraph (b) of this
section that occurred within 3 years of the date of discovery.
(2) If the ENS is inoperative, the licensee must make the required
notifications via commercial telephone service, other dedicated
telephone system, or any other method which will ensure that a report
is made as soon as practical to the NRC Headquarters Operations Center
at the numbers specified in appendix A to part 73 of this chapter.
(3) The licensee must notify the NRC immediately after notification
of the appropriate State or local agencies and not later than 1 hour
after the time the licensee declares one of the Emergency Classes.
(4) The licensee must activate the data links with the NRC as
specified in their emergency plans after declaring an Emergency Class
for events of actual or potential substantial degradation of plant
safety or security, probable risk to site personnel life, or site
equipment damage caused by hostile action. The data links may also be
activated by the licensee during emergency drills or exercises if the
licensee's computer system has the capability to transmit the exercise
data.
(5) When making a report under paragraph (a)(1) of this section,
the licensee must identify--
(i) The Emergency Class declared; or
(ii) Paragraph (b)(1), ``One-hour reports,'' paragraph (b)(2),
``Four-hour reports,'' or paragraph (b)(3), ``Eight-hour reports,'' as
the paragraph of this section requiring notification of the non-
emergency event.
(6) In lieu of submitting a report required under paragraph (b)(2)
or (b)(3) of this section through the Emergency Notification System,
the licensee may submit the report using other methods, provided the
licensee submits the report to the NRC Headquarters Operations Center
within the required timeframe and confirms receipt of the report by the
NRC.
(b) Non-emergency events--(1) One-hour reports. If not reported as
a declaration of an Emergency Class under paragraph (a) of this
section, the licensee must notify the NRC as soon as practical and in
all cases within one hour of the occurrence of any deviation from the
plant's Technical Specifications authorized under Sec. 53.740(h) of
this part.
(2) Four-hour reports. If not reported under paragraphs (a) or
(b)(1) of this section, the licensee must notify the NRC as soon as
practical, and in all cases within 4 hours of the occurrence of any of
the following:
(i) The initiation of any commercial 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) [Reserved]
(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 commercial nuclear plant.
(v) [Reserved]
(3) Eight-hour reports. If not reported under paragraphs (a),
(b)(1), or (b)(2) of this section, the licensee must notify the NRC as
soon as practical 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 commercial nuclear plant, including its
principal safety barriers, being seriously degraded; or
(B) The commercial nuclear plant being in a condition not analyzed
under Sec. 53.450 that significantly degrades plant safety.
(ii)-(iv) [Reserved]
(v) Any event that results in a major loss of emergency assessment
capability, offsite response capability, or offsite communications
capability (e.g., significant portion of control room indication, ENS,
or offsite notification system).
(c) Follow-up notification: With respect to the notifications made
under paragraphs (a) and (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 the
declaration of any of the Emergency Classes, if such a declaration has
not been previously made; or
(ii) Any change from one Emergency Class to another; or
(iii) A termination of the Emergency Class.
(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
Headquarters Operation Center upon request by the NRC.
\1\ Other requirements for immediate notification of the NRC by
licensed operating commercial nuclear plants are contained elsewhere
in this chapter, in particular Sec. Sec. 20.1906, 20.2202, 72.216,
73.77, and 73.1200 of this chapter.
Sec. 53.1640 Licensee event report system.
(a) Reportable events. (1) Each commercial nuclear plant licensee
holding an OL under this part or a COL under this part after the
Commission makes the finding under Sec. 53.1452(g), must submit a
Licensee Event Report (LER) for any event of the type described in this
paragraph (a) within 60 days after discovery of the event. In the case
of an invalid actuation reported under Sec. 53.1640(a)(2), 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 LER. 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)(A) The completion of any commercial nuclear plant shutdown
required by the plant's Technical Specifications.
(B) Any operation or condition which was prohibited by the plant's
Technical Specifications except when--
(1) The Technical Specification is administrative in nature;
(2) 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
(3) The Technical Specification was revised prior to discovery of
the event
[[Page 15865]]
such that the operation or condition was no longer prohibited at the
time of the event.
(C) Any deviation from the plant's Technical Specifications
authorized under Sec. 53.740(h).
(ii) Any event or condition that resulted in--
(A) The condition of the commercial nuclear plant, including its
principal safety barriers, being seriously degraded; or
(B) The commercial nuclear plant being in a condition not analyzed
under Sec. 53.450 that significantly degrades plant safety.
(iii) Any natural phenomena or other external condition that posed
an actual threat to the safety of the commercial nuclear plant or
significantly hampered site personnel in the performance of duties
necessary for the safe operation of the commercial nuclear plant.
(iv) Any event or condition that resulted in inadvertent operation
of any structures, systems, and component classified as safety-related
(SR) for an identified safety function under Sec. 53.460 or the
unplanned sole reliance on an SR system for those systems that are in
constant operation, 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.
(v) Any event or condition that could have prevented the
fulfillment of the safety functions identified under Sec. 53.230.
(vi) 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.
(vii)(A) Any event or condition that as a result of a single cause
could have prevented the fulfillment of any of the safety functions
identified under Sec. 53.230.
(B) Events covered in paragraph (a)(2)(vii)(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)(vii)(A) of this section if the event
results from--
(1) A shared dependency among trains or channels that is a natural
or expected consequence of the approved plant design; or
(2) Normal and expected wear or degradation.
(viii)(A) Any 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 10 CFR part
20, table 2, column 1.
(B) Any liquid effluent release that, when averaged over a time
period of 1-hour, exceeds 20 times the applicable concentrations
specified in appendix B to 10 CFR part 20, 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.
(ix) Any event that posed an actual threat to the safety of the
commercial 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 LER must contain--
(1) A brief abstract describing the major occurrences during the
event, including all component or system failures that contributed to
the event and significant corrective action taken or planned to prevent
recurrence.
(2)(i) A clear, specific narrative description of what occurred so
that knowledgeable readers conversant with the design of commercial
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 systems, structures, or components 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) [Reserved]
(G) For failures of components with multiple functions, include a
list of systems or secondary functions that were also affected.
(H) For failure that rendered a component or system classified as
SR or non-safety-related but safety-significant inoperable, an estimate
of the elapsed time from the discovery of the failure until the
component or system was returned to service.
(I) The method of discovery of each component or system failure or
procedural error.
(J) For each human performance related root cause, the licensee
must discuss the cause(s) and circumstances.
(K) Automatically and manually initiated safety system responses.
(L) 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 probability 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. 53.040, the requested information as a supplement to
the initial LER.
(d) Submission of reports. Licensee Event Reports must be prepared
on NRC Form 366 and submitted to the NRC, as specified in Sec. 53.040.
(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
[[Page 15866]]
reproduction and micrographic processing.
Sec. 53.1645 Reports of radiation exposure to members of the public.
(a) Each holder of an OL, and each holder of a COL after the
Commission has made the finding under Sec. 53.1452(g), must submit
radiological reports as required by 10 CFR part 20, as well as an
Annual Radioactive Effluent Release Report and an Annual Radiological
Environmental Operating Report. The Annual Radioactive Effluent Release
Report must specify the quantity of each of the principal radionuclides
released to unrestricted areas in liquid and in gaseous effluents and
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 calendar year. The Annual
Radiological Environmental Operating Report must provide data on
measurable levels of radiation and radioactive materials in the
environment, must include an evaluation of the relationship between
quantities of radioactive material released in effluents and resultant
radiation doses to individuals from principal pathways of exposure, and
must include the results of environmental monitoring during the
previous calendar year. These reports must also include any other
information as may be required by the Commission to estimate maximum
potential annual radiation doses to the public. The reports must be
submitted as specified in Sec. 53.040 by May 15 of each successive
year. If the total effective dose equivalent to members of the public
in unrestricted areas during the reporting period is greater than the
design objectives established under Sec. 53.425, the report must
specify the causes for exceeding the design objective and describe any
corrective actions. On the basis of these reports and any additional
information the Commission may obtain from the licensee or others, the
Commission may require the licensee to take action as the Commission
deems appropriate.
(b) If during any calendar quarter the radiation exposure to a
member of the public in the unrestricted areas, calculated on the same
basis as the respective design objective exposure, exceeds one-half of
the annual design objective exposure, the licensee must submit a report
as specified in Sec. 53.040. The report shall specify the causes for
exceeding one-half the annual design objective exposure in a quarter
and describe corrective actions that the licensee will take to maintain
radiation exposure to levels within the design objectives for the
remainder of the year. The report shall be submitted within 30 days
from the end of the quarter when one-half of the annual design
objective exposure was exceeded.
Sec. 53.1650 Facility information and verification.
(a) In response to a written request by the Commission, each
applicant for a CP or license and each recipient of a CP or a license
must submit facility information, as described in Sec. 75.10 of this
chapter, on International Atomic Energy Agency (IAEA) Design
Information Questionnaire forms and site information on DOC/NRC Form
AP-A and associated forms;
(b) As required by the Additional Protocol, must submit location
information described in Sec. 75.11 of this chapter on DOC/NRC Form
AP-1 and associated forms; and
(c) Must permit verification thereof by the IAEA and take other
action as necessary to implement the US/IAEA Safeguards Agreement, as
described in part 75 of this chapter.
Sec. 53.1660 Financial requirements.
Sections 53.1670 through 53.1700 set out the requirements and
procedures related to financial qualifications and related reporting
requirements.
Sec. 53.1670 Financial qualifications.
Except for an electric utility applicant for a license to operate a
commercial nuclear plant, an applicant for a CP, OL, or COL under this
part must appear to be financially qualified for the activities for
which the permit or license is sought.
Sec. 53.1680 [Reserved]
Sec. 53.1690 Licensee's change of status; financial qualifications.
(a) An electric utility licensee holding an OL or COL (including a
renewed license) for a commercial nuclear plant, no later than seventy-
five (75) days prior to ceasing to be an electric utility in any manner
not involving a license transfer under Sec. 53.1399 or Sec. 53.1456
must provide the NRC with the financial qualifications information that
would be required for obtaining an initial OL under this part. The
financial qualifications information must address the first full 5
years of operation after the date the licensee ceases to be an electric
utility.
(b)(1) Any holder of a license issued under this part must notify
the appropriate NRC Regional Administrator, in writing, immediately
following the filing of a voluntary or involuntary petition for
bankruptcy under any chapter of title 11 (Bankruptcy) of the United
States Code by or against--
(i) The licensee;
(ii) An entity (as 11 U.S.C. 101(14) defines that term) controlling
the licensee or listing the license or licensee as property of the
estate; or
(iii) An affiliate (as 11 U.S.C. 101(2) defines that term) of the
licensee.
(2) This notification must indicate--
(i) The bankruptcy court in which the petition for bankruptcy was
filed; and
(ii) The date of the filing of the petition.
Sec. 53.1700 Creditor regulations.
(a) Pursuant to section 184 of the Act, the Commission consents,
without individual application, to the creation of any mortgage,
pledge, or other lien upon any facility not owned by the United States
which is the subject of a license or upon any leasehold or other
interest in such facility; provided--
(1) That the rights of any creditor so secured may be exercised
only in compliance with and subject to the same requirements and
restrictions as would apply to the licensee pursuant to the provisions
of the license, the Act, and regulations issued by the Commission under
the Act; and
(2) That no creditor so secured may take possession of the facility
pursuant to the provisions of this section prior to either the issuance
of a license from the Commission authorizing such possession or the
transfer of the license.
(b) Any creditor so secured may apply for transfer of the license
covering such facility by filing an application for transfer of the
license under Sec. 53.1570. The Commission will act upon such
application under subpart I of this part.
(c) Nothing contained in this regulation shall be deemed to affect
the means of acquiring, or the priority of, any tax lien or other lien
provided by law.
(d) As used in this section--
(1) License includes any license under this part, which may be
issued by the Commission with regard to a facility.
(2) Creditor includes, without implied limitation, the trustee
under any mortgage, pledge or lien on a facility made to secure any
creditor, any trustee or receiver of the facility appointed by a court
of competent jurisdiction in any action brought for the benefit of any
creditor secured by such mortgage, pledge or lien, any purchaser of
such facility at the sale thereof upon foreclosure of such mortgage,
pledge, or lien or upon exercise of any power of sale contained
therein, or any assignee of any such purchaser.
(3) Facility includes, but is not limited to, a site which is the
subject of an early
[[Page 15867]]
site permit under this part, and a reactor manufactured under an ML
under this part.
Sec. 53.1710 Financial protection.
Sections 53.1720 and 53.1730 set out the requirements and
procedures related to licensees obtaining and maintaining insurance to
cover stabilization and decontamination activities in the event of an
accident and financial protection in accordance with part 140,
``Financial Protection Requirements and Indemnity Agreements,'' of this
chapter.
Sec. 53.1720 Insurance required to stabilize and decontaminate plant
following an accident.
Each commercial nuclear plant licensee under this part must take
reasonable steps to obtain insurance available at reasonable costs and
on reasonable terms from private sources or to demonstrate that it
possesses an equivalent amount of protection covering the licensee's
obligation, in the event of an accident at the licensee's commercial
nuclear reactor, to stabilize and decontaminate the plant and the plant
site at which such an accident may occur, provided that--
(a) The insurance required by this section must have a minimum
coverage limit for each commercial nuclear plant site of $1.06 billion,
an amount based on plant-specific estimates of costs to stabilize and
decontaminate a plant, or whatever amount of insurance is generally
available from private sources, whichever is less. The required
insurance must clearly state that, as and to the extent provided in
paragraph (d)(1) of this section, any proceeds must be payable first
for stabilization of the plant and next for decontamination of the
plant and the plant site. If a licensee's coverage falls below the
required minimum, the licensee must within 60 days take all reasonable
steps to restore its coverage to the required minimum. The required
insurance may, at the option of the licensee, be included within
policies that also provide coverage for other risks, including, but not
limited to, the risk of direct physical damage.
(b)(1) With respect to policies issued or annually renewed, the
proceeds of such required insurance must be dedicated, as and to the
extent provided in this paragraph (b), to reimbursement or payment on
behalf of the insured of reasonable expenses incurred or estimated to
be incurred by the licensee in taking action to fulfill the licensee's
obligation, in the event of an accident at the licensee's plant, to
ensure that the plant is in, or is returned to, and maintained in, a
safe and stable condition and that radioactive contamination is removed
or controlled such that personnel exposures are consistent with the
occupational exposure limits in 10 CFR part 20. These actions must be
consistent with any other obligation the licensee may have under this
chapter and must be subject to paragraph (d) of this section. As used
in this section, an ``accident'' means an event that involves the
release of radioactive material from its intended place of confinement
within the commercial nuclear plant such that there is a present danger
of release off site in amounts that would pose a threat to the public
health and safety.
(2) The stabilization and decontamination requirements set forth in
paragraph (d) of this section must apply uniformly to all insurance
policies required under this section.
(c) The licensee shall report to the NRC on April 1 of each year
the current levels of this insurance or financial security it maintains
and the sources of this insurance or financial security.
(d)(1) In the event of an accident at the licensee's plant,
whenever the estimated costs of stabilizing the licensed plant and of
decontaminating the plant and the plant site exceed one tenth of the
minimum insurance under paragraph (a) of this section, the proceeds of
the insurance required by this section must be dedicated to and used,
first, to ensure that the licensed plant is in, or is returned to, and
can be maintained in, a safe and stable condition so as to prevent any
significant risk to the public health and safety and, second, to
decontaminate the plant and the plant site in accordance with the
licensee's cleanup plan as approved by order of the Director, Office of
Nuclear Reactor Regulation. This priority on insurance proceeds must
remain in effect for 60 days or, upon order of the Director, for such
longer periods, in increments not to exceed 60 days except as provided
for activities under the cleanup plan required in paragraphs (d)(3) and
(d)(4) of this section, as the Director may find necessary to protect
the public health and safety. Actions needed to bring the plant to and
maintain the plant in a safe and stable condition may include one or
more of the following, as appropriate:
(i) Shutdown of the reactor(s) and other processes at the plant;
(ii) Establishment and maintenance of long-term cooling with stable
decay heat removal;
(iii) Maintenance of sub-criticality;
(iv) Control of radioactive releases; and
(v) Securing of structures, systems, or components to minimize
radiation exposure to onsite personnel or to the offsite public or to
facilitate later decontamination or both.
(2) The licensee must inform the Director, Office of Nuclear
Reactor Regulation in writing when the plant is and can be maintained
in a safe and stable condition so as to prevent any significant risk to
the public health and safety. Within 30 days after the licensee informs
the Director that the plant is in this condition, or at such earlier
time as the licensee may elect or the Director may for good cause
direct, the licensee must prepare and submit a cleanup plan for the
Director's approval. The cleanup plan must identify and contain an
estimate of the cost of each cleanup operation that will be required to
decontaminate the reactor sufficiently to permit the licensee either to
resume operation of the reactor or to apply to the Commission under
subpart G of this part for authority to decommission the reactor and to
surrender the license voluntarily. Cleanup operations may include one
or more of the following, as appropriate:
(i) Processing any contaminated materials generated by the accident
and by decontamination operations to remove radioactive materials;
(ii) Decontamination of surfaces inside the plant buildings to
levels consistent with the Commission's occupational exposure limits in
10 CFR part 20, and decontamination or disposal of equipment;
(iii) Decontamination or removal and disposal of internal parts,
damaged fuel from the reactor coolant or fuel systems, or related
process or waste systems; and
(iv) Cleanup of the reactor coolant or fuel systems or related
process or waste systems.
(3) Following review of the licensee's cleanup plan, the Director
will order the licensee to complete all operations that the Director
finds are necessary to decontaminate the reactor sufficiently to permit
the licensee either to resume operation of the reactor or to apply to
the Commission under subpart G of this part for authority to
decommission the reactor and to surrender the license voluntarily. The
Director must approve or disapprove, in whole or in part for stated
reasons, the licensee's estimate of cleanup costs for such operations.
Such order may not be effective for more than one year, at which time
it may be renewed. Each subsequent renewal order, if imposed, may be
effective for not more than 6 months.
(4) Of the balance of the proceeds of the required insurance not
already expended to place the plant in a safe and stable condition
under paragraph (b)(1) of this section, an amount
[[Page 15868]]
sufficient to cover the expenses of completion of those decontamination
operations that are the subject of the Director's order must be
dedicated to such use, provided that, upon certification to the
Director of the amounts expended previously and from time to time for
stabilization and decontamination and upon further certification to the
Director as to the sufficiency of the dedicated amount remaining,
policies of insurance may provide for payment to the licensee or other
loss payees of amounts not so dedicated, and the licensee may proceed
to use in parallel (and not in preference thereto) any insurance
proceeds not so dedicated for other purposes.
Sec. 53.1730 Financial protection requirements.
Commercial nuclear plant licensees must satisfy the applicable
provisions of part 140, ``Financial Protection Requirements and
Indemnity Agreements,'' of this chapter.
Subparts K and L [Reserved]
Subpart M--Enforcement
Sec. 53.9000 Violations.
(a) The Commission may obtain an injunction or other court order to
prevent a violation of the provisions of--
(1) The Atomic Energy Act of 1954, as amended (the Act);
(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 Act:
(1) For violations of--
(i) Sections 53, 57, 62, 63, 81, 82, 101, 103, 104, 107, or 109 of
the Act;
(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 Act.
Sec. 53.9010 Criminal penalties.
(a) Section 223 of the Act provides for criminal sanctions for
willful violation of, attempted violation of, or conspiracy to violate,
any regulation issued under sections 161b, 161i, or 161o of the Act.
For purposes of section 223, all the regulations in part 53 are issued
under one or more of sections 161b, 161i, or 161o, except for the
sections listed in paragraph (b) of this section.
(b) The regulations in 10 CFR part 53 that are not issued under
sections 161b, 161i, or 161o for the purposes of section 223 are as
follows: Sec. Sec. 53.000, 53.015, 53.020, 53.040, 53.080, 53.090,
53.100, 53.110, 53.120, 53.600, 53.725, 53.726, 53.735, 53.760, 53.775,
53.790, 53.795, 53.820, 53.910, 53.1000, 53.1050, 53.1100, 53.1103,
53.1106, 53.1109, 53.1112, 53.1115, 53.1118, 53.1120, 53.1121, 53.1124,
53.1140, 53.1144, 53.1146, 53.1149, 53.1155, 53.1158, 53.1164, 53.1170,
53.1173, 53.1176, 53.1179, 53.1188, 53.1200, 53.1206, 53.1209, 53.1210,
53.1212, 53.1215, 53.1218, 53.1221, 53.1230, 53.1236, 53.1239, 53.1241,
53.1242, 53.1245, 53.1248, 53.1251, 53.1254, 52.1257, 52.1260, 53.1263,
53.1270, 53.1276, 53.1279, 53.1282, 53.1285, 53.1286, 53.1287, 53.1288,
53.1291, 53.1293, 53.1295, 53.1300, 53.1306, 53.1309, 53.1312, 53.1315,
53.1318, 53.1324, 53.1330, 53.1333, 53.1336, 53.1348, 53.1360, 53.1366,
53.1369, 53.1372, 53.1375, 53.1381, 53.1384, 53.1387, 53.1390, 53.1396,
53.1401, 53.1405, 53.1410, 53.1416, 53.1419, 53.1422, 53.1425, 53.1431,
53.1437, 53.1440, 53.1443, 53.1452, 53.1455, 53.1456, 53.1458, 53.1461,
53.1470, 53.1500, 53.1510, 53.1515, 53.1520, 53.1525, 53.1530, 53.1535,
53.1540, 53.1560, 53.1585, 53.1590, 53.1595, 53.1600, 53.1660, 53.1670,
53.1700, 53.1710, 53.1730, 53.9000, 53.9010.
PART 70--DOMESTIC LICENSING OF SPECIAL NUCLEAR MATERIAL
0
134. 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.
Sections 70.1(c) and 70.20a(b) also issued under secs. 135, 141,
Pub. L. 97-425, 96 Stat. 2232, 2241 (42 U.S.C. 10155, 10161).
Section 70.21(g) also issued under Atomic Energy Act sec. 122
(42 U.S.C. 2152).
Section 70.31 also issued under Atomic Energy Act sec. 57(d) (42
U.S.C. 2077(d)).
Sections 70.36 and 70.44 also issued under Atomic Energy Act
sec. 184 (42 U.S.C. 2234).
Section 70.81 also issued under Atomic Energy Act secs. 186, 187
(42 U.S.C. 2236, 2237).
Section 70.82 also issued under Atomic Energy Act sec. 108 (42
U.S.C. 2138).
Sec. 70.20a [Amended]
0
135. In Sec. 70.20a, in paragraph (b), remove ``parts 30 through 36,
39, 40, 50, 72, 110'' and add in its place ``parts 30 through 36, 39,
40, 50, 53, 72, 110''.
Sec. 70.22 [Amended]
0
136. In Sec. 70.22, wherever it may appear, remove the phrase ``part
50'' and add in its place the phrase ``part 50 or part 53''.
0
137. In Sec. 70.24, revise paragraphs (d) to read as follows:
Sec. 70.24 Criticality accident requirements.
* * * * *
(d)(1) The requirements in paragraphs (a) through (c) of this
section do not apply to a holder of a construction permit or operating
license for a nuclear power reactor issued under part 50 or part 53 of
this chapter or a combined license issued under part 52 or part 53 of
this chapter, if the holder complies with the requirements of paragraph
(b) of 10 CFR 50.68 or paragraph (m)(2) of 10 CFR 53.440, as
applicable.
(2) An exemption from Sec. 70.24 held by a licensee who thereafter
elects to comply with requirements of paragraph (b) of 10 CFR 50.68 or
paragraph (m)(2) of 10 CFR 53.440 does not exempt that licensee from
complying with any of the requirements in Sec. 50.68 or Sec.
53.440(m) of this chapter but will be ineffective so long as the
licensee elects to comply with Sec. 50.68(b) or Sec. 53.440(m)(2) of
this chapter, as applicable.
Sec. 70.32 [Amended]
0
138. In Sec. 70.32, in paragraph (c)(1) introductory text, remove the
phrase ``part 50 of this chapter'' and add in its place the phrase
``part 50 or part 53 of this chapter''; and in paragraph (d) remove the
phrase ``or Sec. 70.34 of this chapter, as appropriate.'' and add in
its place the phrase ``, Sec. 70.34, or Sec. 53.1510 of this chapter,
as appropriate.''.
0
139. 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. 53.1630 of this chapter. They do apply to those
10 CFR parts 50 or 53 licensees possessing material licensed under 10
CFR part 70 that are not subject to the notification requirements in
Sec. 50.72 or Sec. 53.1630 of this chapter.
[[Page 15869]]
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
140. 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
141. In Sec. 72.3, revise the definition for ``Independent spent fuel
storage installation or ISFSI'' to read as follows:
Sec. 72.3 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 which is located on the site of another facility
licensed under this part or a facility licensed under part 50 or part
53 of this chapter and which shares common utilities and services with
that facility or is physically connected with that other facility may
still be considered independent.
* * * * *
0
142. 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 53 of this chapter or ISFSI licensees who are an
electric utility, as defined in part 50 or part 53 of this chapter,
with a specific license issued under this part, the methods of Sec.
50.75(b), (e), and (h) or Sec. 53.1010, Sec. 53.1040, Sec.
53.1045(b), and Sec. 53.1060 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 (e)(1) through (4) of this section.
* * * * *
0
143. In Sec. 72.32, revise paragraph (c)(2) to read as follows:
Sec. 72.32 Emergency plan.
* * * * *
(c) * * *
(2)(i) Located within the exclusion area as defined in 10 CFR part
100, of a nuclear power reactor licensed for operation by the
Commission, the emergency plan that meets either the requirements in
Sec. 50.160 of this chapter or the requirements in appendix E to part
50 of this chapter and Sec. 50.47(b) of this chapter shall be deemed
to satisfy the requirements of this section.
(ii) Located within the exclusion area, as defined in 10 CFR part
53, of a commercial nuclear plant licensed for operation by the
Commission, the emergency plan that meets either the requirements in
Sec. 50.160 of this chapter or the requirements in appendix E to part
50 of this chapter and Sec. 50.47(b) of this chapter shall be deemed
to satisfy the requirements of this section.
* * * * *
Sec. 72.40 [Amended]
0
144. In Sec. 72.40, in paragraph (c), remove the phrase ``under part
50 of this chapter,'' and add in its place the phrase ``under part 50
or part 53 of this chapter,''.
0
145. 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 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 53 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 53 of this chapter).
* * * * *
Sec. 72.184 [Amended]
0
146. In Sec. 72.184, in paragraph (a), remove the phrase ``under part
50 of this chapter'' and add in its place the phrase ``under part 50 or
part 53 of this chapter''.
0
147. 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, part 52, or part 53.
0
148. 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. 53.1550
of this chapter. Results of this determination must be documented in
the evaluations made in paragraph (b)(5) of this section.
* * * * *
0
149. 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. 53.1060
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 a reactor operating or
combined license under 10 CFR part 53 and submitted under Sec. 53.1070
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
150. 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.
[[Page 15870]]
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
151. In Sec. 73.1, revise paragraph (b)(1)(i) to read as follows:
Sec. 73.1 Purpose and scope.
* * * * *
(b) * * *
(1) * * *
(i) The physical protection of production and utilization
facilities licensed under part 50, part 52, or part 53 of this chapter,
* * * * *
0
152. In Sec. 73.2, revise introductory text and paragraph (a) to read
as follows:
Sec. 73.2 Definitions.
As used in this part:
(a) Terms defined in parts 50, 52, 53, 70, and 95 of this chapter
have the same meaning when used in this part.
* * * * *
0
153. In Sec. 73.8, revise paragraph (b) to read as follows:
Sec. 73.8 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,
73.100, 73.110, 73.120, 73.1200, 73.1205, 73.1210, 73.1215, and
appendices B and C to this part.
* * * * *
0
154. 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, part 52, or part 53 of
this chapter, shall comply with the following:
* * * * *
0
155. In Sec. 73.55, revise paragraphs (a)(4) and (6), (b)(8) and
(9)(ii)(C), (c)(6), (i)(4)(iii), (l)(1) and (7)(ii), (p)(1)(i) and
(ii), and (r)(2) and (r)(4)(iii) to read as follows:
Sec. 73.55 Requirements for physical protection of licensed
activities in nuclear power reactors against radiological sabotage.
(a) * * *
(4) Applicants for an operating license under the provisions of
part 50 or part 53 of this chapter or holders of a combined license
under the provisions of part 52 or part 53 of this chapter shall
implement the requirements of this section before fuel is allowed
onsite (protected area).
* * * * *
(6) Applicants for an operating license under the provisions of
part 50 or part 53 of this chapter, or holders of a combined license
under the provisions of part 52 or part 53 of this chapter that do not
reference a standard design certification or reference a standard
design certification issued after May 26, 2009, shall meet the
requirement of Sec. 73.55(i)(4)(iii).
(b) * * *
(8) The licensee shall establish, maintain, and implement a cyber
security program in accordance with Sec. 73.54 or Sec. 73.110, as
applicable.
(9) * * *
(ii) * * *
(C) The cyber security program described in Sec. 73.54 or Sec.
73.110, as applicable; and
* * * * *
(c) * * *
(6) Cyber Security Plan. The licensee shall establish, maintain,
and implement a Cyber Security Plan that describes how the criteria set
forth in Sec. 73.54 or Sec. 73.110, as applicable, will be
implemented.
* * * * *
(i) * * *
(4) * * *
(iii) Applicants for an operating license under the provisions of
part 50 of this chapter, or holders of a combined license under the
provisions of part 52 of this chapter, or licensees under part 53 of
this chapter that elect to demonstrate compliance with Sec. 73.55,
consistent with Sec. 53.860(a)(2) of this chapter, shall construct,
locate, protect, and equip both the central and secondary alarm
stations to the standards for the central alarm station contained in
this section. Both alarm stations shall be equal and redundant, such
that all functions needed to satisfy the requirements of this section
can be performed in both alarm stations.
* * * * *
(l) * * *
(1) Commercial nuclear power reactors licensed under 10 CFR part
50, part 52, or part 53 and authorized to use special nuclear material
in the form of MOX fuel assemblies containing up to 20 weight percent
PuO2 shall, in addition to demonstrating compliance with the
requirements of this section, protect un-irradiated MOX fuel assemblies
against theft or diversion as described in this paragraph (l).
* * * * *
(7) * * *
(ii) Additional measures for the physical protection of un-
irradiated MOX fuel assemblies containing greater than 20 weight
percent PuO2 shall be determined by the Commission on a
case-by-case basis and documented through license amendment in
accordance with Sec. 50.90 or Sec. 53.1510 of this chapter.
* * * * *
(p) * * *
(1) * * *
(i) Under Sec. 50.54 paragraphs (x) and (y) or Sec. 53.740(h) of
this chapter, the licensee may suspend any security measures under this
section 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. This
suspension of security measures must be approved as a minimum by a
licensed senior operator or a generally licensed reactor operator, as
applicable, before taking this action.
(ii) During severe weather when the suspension of affected security
measures is immediately needed to protect the personal health and
safety of security force personnel and no other immediately apparent
action consistent with the license conditions and technical
specifications can provide adequate or equivalent protection. This
suspension of security measures must be approved, as a minimum, by a
licensed senior operator or a generally licensed reactor operator, as
applicable, with input from the security supervisor or manager, before
taking this action.
* * * * *
(r) * * *
(2) The licensee shall submit proposed alternative measure(s) to
the Commission for review and approval under Sec. 50.4 and Sec.
50.90, or Sec. 53.040 and Sec. 53.1510 of this chapter, before
implementation.
* * * * *
(4) * * *
(iii) Based on comparison of the costs of the alternative measures
to the costs
[[Page 15871]]
of demonstrating compliance with the Commission's requirements using
the essential elements of Sec. 50.109 or Sec. 53.1590 of this
chapter, the costs of fully demonstrating compliance with the
Commission's requirements are not justified by the protection that
would be provided.
0
156. 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 of this chapter, each holder of a combined license under the
provisions of part 52 of this chapter, and applicants for an operating
license or holders of a combined license under part 53 of this chapter
who do not demonstrate compliance with 10 CFR 73.100(a)(1)(i) shall
implement the requirements of this section before fuel is allowed on
site (protected area).
* * * * *
0
157. 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 10 CFR part 50 or
part 53 or before the Commission makes its finding under Sec.
52.103(g) or Sec. 53.1452(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
158. In Sec. 73.58, revise paragraph (a) to read as follows:
Sec. 73.58 [Amended]
(a) Each operating nuclear power reactor licensee with a license
issued under part 50, part 52, or part 53 of this chapter shall comply
with the requirements of this section.
* * * * *
0
159. In Sec. 73.67, revise introductory text of paragraphs (d)
introductory text and (f) introductory text to read as follows:
Sec. 73.67 Licensee fixed site and in-transit requirements for the
physical protection of special nuclear material of moderate and low
strategic significance.
* * * * *
(d) Fixed site requirements for special nuclear material of
moderate strategic significance. Each licensee who possesses, stores,
or uses quantities and types of special nuclear material of moderate
strategic significance at a fixed site or contiguous sites, except as
allowed by paragraph (b)(2) of this section and except those who are
licensed to operate a nuclear power reactor pursuant to part 50 or part
53, provided that the special nuclear material is located within a
protected area and protected under Sec. 73.55 or Sec. 73.100, shall:
* * * * *
(f) Fixed site requirements for special nuclear material of low
strategic significance. Each licensee who possesses, stores, or uses
special nuclear material of low strategic significance at a fixed site
or contiguous sites, except those who are licensed to operate a nuclear
power reactor pursuant to part 50 or part 53, provided that the special
nuclear material is located within a protected area and protected under
Sec. 73.55 or Sec. 73.100, shall:
* * * * *
0
160. In Sec. 73.77, revise paragraphs (a), (b)(1), (c)(6) and (7) to
read as follows:
Sec. 73.77 Cybersecurity event notifications.
(a) Each licensee subject to the provisions of Sec. 73.54 or Sec.
73.110 shall notify the NRC Headquarters Operations Center via the
Emergency Notification System (ENS), under paragraph (c) of this
section:
(1) Within one hour after discovery of a cyberattack that adversely
impacted:
(i) Safety-related or important-to-safety functions, security
functions, or emergency preparedness functions (including offsite
communications); or that compromised support systems and equipment
resulting in adverse impacts to safety, security, or emergency
preparedness functions within the scope of Sec. 73.54; or,
(ii) 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. 53.210
of this chapter, or security functions performed by digital assets
necessary for implementing the physical security requirements in Sec.
53.860(a) of this chapter.
(2) Within 4 hours:
(i) After discovery of a cyberattack that could have caused an
adverse impact to:
(A) Safety-related or important-to-safety functions, security
functions, or emergency preparedness functions (including offsite
communications); or that could have compromised support systems and
equipment, which if compromised, could have adversely impacted safety,
security, or emergency preparedness functions within the scope of Sec.
73.54; or,
(B) 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. 53.210
of this chapter, or security functions performed by digital assets
necessary for implementing the physical security requirements in Sec.
53.860(a) of this chapter.
(ii) After discovery of a suspected or actual cyberattack initiated
by personnel with physical or electronic access to digital computer and
communication systems and networks within the scope of Sec. 73.54 or
Sec. 73.110.
(iii) After notification to a local, State, or other Federal agency
(e.g., law enforcement, Federal Bureau of Investigation (FBI), etc.) of
an event related to the licensee's implementation of their
cybersecurity program for digital computer and communication systems
and networks within the scope of Sec. 73.54 or Sec. 73.110 that does
not otherwise require a notification under paragraph (a) of this
section.
(3) Within 8 hours after receipt or collection of information
regarding observed behavior, activities, or statements that may
indicate intelligence gathering or pre-operational planning related to
a cyberattack against digital computer and communication systems and
networks within the scope of Sec. 73.54 or Sec. 73.110.
(b) * * *
(1) The licensee shall use the site corrective action program to
record vulnerabilities, weaknesses, failures and deficiencies in their
Sec. 73.54 or Sec. 73.110 cybersecurity program within 24 hours of
their discovery.
* * * * *
(c) * * *
(6) Declaration of emergencies. Notifications made to the NRC for
the declaration of an emergency class shall be performed in accordance
with Sec. 50.72 or Sec. 53.1630 of this chapter, as applicable.
(7) Elimination of duplication. Separate notifications and reports
are not required for events that are also reportable under Sec. Sec.
50.72 and 50.73 or Sec. Sec. 53.1630 and 53.1640 of this chapter.
However, these notifications should also indicate the applicable Sec.
73.77 reporting criteria.
* * * * *
0
161. Add subpart J, consisting of Sec. Sec. 73.100 through 73.120, to
read as follows:
[[Page 15872]]
Subpart J--Security Requirements at Commercial Nuclear Plants
Sec.
73.100 Technology-inclusive requirements for physical protection of
licensed activities at commercial nuclear plants against
radiological sabotage.
73.110 Technology-inclusive requirements for protection of digital
computer and communication systems and networks.
73.120 Access authorization program for commercial nuclear plants.
Sec. 73.100 Technology-inclusive requirements for physical protection
of licensed activities at commercial nuclear plants against
radiological sabotage.
(a) Introduction. (1) Each licensee that is licensed to operate a
commercial nuclear plant under part 53 of this chapter and elects to
implement the requirements of this section must identify achievable
target sets in accordance with paragraph (b)(5) of this section and
develop, implement, and maintain a physical protection program under
the following requirements:
(i) Each licensee that demonstrates no achievable target sets exist
in accordance with paragraph (b)(5) of this section, and does not
credit any active measures (e.g., operator action, mitigative action,
detection, assessment, armed response) in making that demonstration, is
exempt from the remaining requirements of this section.
(ii) Each licensee that demonstrates no achievable target sets
exist in accordance with paragraph (b)(5) of this section, and credits
active measures in making that demonstration, 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 the features to
prevent criticality required under Sec. 53.620(d)(1) of this chapter);
for such licensees, the requirements of paragraphs (b)(2) through (4)
of this section will be deemed satisfied if the physical protection
program is designed to ensure that the credited active measures will be
implemented in response to threats up to and including the design-basis
threat of radiological sabotage.
(iii) Each licensee that demonstrates achievable target sets exist,
in accordance with paragraph (b)(5) of this section, 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 the features to
prevent criticality required under Sec. 53.620(d)(1) of this chapter).
(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) 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. Specifically, the licensee must--
(i) Ensure that the physical protection program capabilities to
protect against the design-basis threat of radiological sabotage are
maintained at all times; and
(ii) Provide defense in depth in achieving performance requirements
through the integration of engineered systems, administrative controls,
and management measures.
(3) The physical protection program must be designed to prevent the
release of radionuclides from any source from exceeding the dose
reference values defined in Sec. 53.210 of this chapter.
(4) The physical protection program must be designed and
implemented to achieve and maintain the reliability and availability of
structures, systems, and components (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. The physical protection program must
be designed to provide layers of security response, with each layer
assuring that a single failure does not result in the loss of
capability to neutralize the design-basis threat adversary. Structures,
systems, and components relied on for delay functions must be designed
to allow for timely security responses to adversary attacks with
adequate defense in depth.
(A) The security response may rely on the use of onsite responders,
law enforcement or other offsite armed responders, or a combination
thereof, to fulfill the interdiction and neutralization functions
required by paragraph (b)(4)(iv) of this section. A licensee relying
entirely or partially on law enforcement or other offsite armed
responders must--
(1) Maintain the capability to detect, assess, interdict, and
neutralize threats as required by paragraphs (b)(4)(i), (b)(4)(ii), and
(b)(4)(iv) of this section;
(2) Provide adequate delay to enable law enforcement or other
offsite armed responders to fulfill the interdiction and neutralization
functions for threats up to and including the design-basis threat of
radiological sabotage;
(3) 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;
(4) Fully describe in the safeguards contingency plan the role that
law enforcement or other offsite armed responders will play in the
licensee's protective strategy. The description must provide sufficient
detail to enable the NRC to determine that the licensee's physical
protection program provides reasonable assurance of adequate protection
against threats up to and including the design-basis threat of
radiological sabotage; and
(5) Identify criteria and measures to compensate for the
degradation or absence of law enforcement or other offsite armed
responders and propose suitable compensatory measures that meet the
requirements of paragraph (h)(3) of this section to address this
degradation.
(B) For licensees relying entirely or partially on law enforcement
responders to fulfill the interdiction and neutralization functions
required by
[[Page 15873]]
paragraph (b)(4)(iv) of this section, the training and qualification
requirements related to armed response personnel in paragraphs (c) and
(e) of this section do not apply to law enforcement responders. The
licensee shall continue to satisfy the performance evaluation
requirements in paragraph (g) of this section for all armed response
personnel, including law enforcement.
(v) Protecting against land and waterborne vehicle bomb assaults.
The licensee must be capable of protecting the plant against the
design-basis threat vehicle bomb assault. The methods that are relied
on to protect against a design-basis threat land vehicle and waterborne
vehicle bomb assault must be designed to protect the reactor building
and structures containing safety- or security-related systems, and
components from explosive effects.
(vi) Access control portals. The licensee must be capable of
detecting and denying unauthorized access to persons and pass-through
of contraband materials (e.g., weapons, incendiary devices, explosives)
to protected areas.
(5) The licensee must identify and document complete and accurate
target sets in accordance with the following:
(i) Preventative operator actions may be credited as target set
elements when: sufficient time to implement exists; environmental
conditions allow operator actions to be completed successfully;
adversary interference is precluded; all equipment required for
operator actions is available, dedicated, staged, and maintained;
approved procedures exist specific to the task being performed; and
training is maintained for proficiency of the credited operator action.
(ii) The identification of target sets must not assume the success
of the security organization; except that licensees may consider delay
provided by the security organization when assessing the availability
of operator actions.
(iii) The licensee must consider cyberattacks in the identification
of target sets.
(iv) The licensee must further identify achievable target sets
through a site-specific analysis. Achievable target sets are those that
are within the capabilities of the design-basis threat adversary to
compromise, destroy, or render non-functional; cannot be mitigated
after adversary interference is precluded and prior to a release of
radionuclides exceeding dose reference values defined in 10 CFR 53.210;
and, if defeated, result irreversibly in exceedance of the dose
reference values in 10 CFR 53.210.
(v) The licensee must document and maintain the process used to
identify achievable target sets, to include the site-specific analyses
and methodologies used to determine and group the target set equipment
or elements, including elements not contained in a protected or vital
area.
(vi) 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.
(vii) The licensee must maintain records in accordance with
paragraph (j) of this section and, in addition, must maintain site-
specific analyses until submittal of the licensee's certifications
required by Sec. 53.1070 of this chapter.
(6) The licensee must identify and analyze site-specific
conditions, including achievable target sets, that may affect the
physical protection program needed to implement the requirements of
this section. The licensee must account for these conditions in
demonstrating compliance with the requirements of this section.
(7) 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.
(8) The licensee must establish, implement, and maintain an access
authorization program under Sec. 73.56, and must describe the program
in the physical security plan.
(9) The licensee must establish, implement, and maintain a
cybersecurity program under Sec. 73.54 or Sec. 73.110 and must
describe the program in the cybersecurity plan.
(10) The licensee must establish, implement, and maintain an
insider mitigation program and must describe the program in the
physical security plan.
(i) The insider mitigation program must monitor the initial and
continuing trustworthiness and reliability of individuals granted or
retaining unescorted access or unescorted access authorization to a
protected or vital area, and implement defense-in-depth methodologies
to minimize the potential for an insider (active, passive, or both) to
adversely affect, either directly or indirectly, the licensee's
capability to protect against radiological sabotage.
(ii) The insider mitigation program must integrate elements of--
(A) The access authorization program under Sec. 73.56 or Sec.
73.120;
(B) The fitness-for-duty program under 10 CFR part 26;
(C) The cybersecurity program under Sec. 73.54 or Sec. 73.110;
and
(D) The physical protection program under this section.
(11) 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.
(12) 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.
(13)(i) The licensee must ensure that the firearms background check
requirements of Sec. 73.17 of this part are met for all members of the
security organization whose official duties require access to covered
weapons or who inventory enhanced weapons.
(ii) The provisions of this paragraph (b)(13) 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 part.
(c) Security organization. The licensee must establish and maintain
a security organization that is staffed, trained, qualified, and
equipped to implement the physical protection program under the
requirements of this section.
(1) The licensee must establish a management system for maintaining
and implementing security policies and procedures to implement the
requirements of this section and the security plans.
(2) Implementing procedures must document the conduct of security
operations, security design and configuration controls, maintenance,
training and qualification, and contingency responses.
(3) The licensee must--
(i) Establish a process for the approval of designs, policies,
processes, and procedures and changes by the individual with overall
responsibility for the physical protection program; and
(ii) Ensure that revisions and changes to the physical protection
program and implementing policies, processes, and procedures satisfy
the requirements of this section.
(4) The licensee must retain, in accordance with Sec. 73.70, 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
[[Page 15874]]
safeguards information in Sec. Sec. 73.21 and 73.22.
(5) The licensee may not permit any individual to implement any
part of the physical protection program unless the individual has been
trained, equipped, and qualified to perform their assigned duties and
responsibilities in accordance with the training and qualification
plan.
(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) Security reviews. The licensee must establish and implement
security reviews to assess the effectiveness of the implementation of
the physical protection program. Security reviews must be performed by
individuals independent of those personnel responsible for program
management and any individual who has direct responsibility for
implementing the onsite physical protection program.
(1) The licensee must review each element of the physical
protection 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. The objective of these reviews must be maintaining
effective implementation of the engineered and administrative controls
required to achieve the physical protection program functions and the
management system required to implement programs and requirements in
this section.
(2) The licensee must establish and perform self-assessments to
ensure the effective implementation of the physical protection program
functions of detection, assessment, communication, delay, and
interdiction and neutralization to protect against the design-basis
threat of radiological sabotage. The licensee must perform design
verification and assessments of the capabilities of active and passive
engineering systems relied on to protect against the design-basis
threat.
(3) Reviews of the security program must include, but are not
limited to, an audit of the effectiveness of the physical protection
program, security plans, implementing procedures, cybersecurity
programs, safety/security interface activities, the testing,
maintenance, and calibration program, and response commitments by
local, State, and Federal law enforcement authorities.
(4) The results and recommendations of the onsite physical
protection 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.
(g) 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.
(1) The licensee must establish the frequencies for performance
evaluations commensurate with the security significance of the physical
protection program.
(2) The licensee must document processes and procedures for
implementing the performance evaluations. The licensee must maintain
records, including results, findings, and corrective actions identified
during the performance evaluations.
(h) Maintenance, testing, and calibration and corrective actions.
(1) The licensee must ensure that security SSCs, including supporting
systems, are inspected, tested, and calibrated for operability and
performance at intervals necessary and sufficient to meet the
requirements of this section.
(2) The licensee must implement corrective actions to ensure
resolution of identified vulnerabilities and deficiencies to meet the
requirements of this section.
(3) The licensee must establish and implement timely compensatory
measures for degraded or inoperable security SSCs to meet the
requirements of this section. Compensatory measures must provide a
level of protection that is equivalent to the protection that was
provided prior to the degradation or inoperability of the security
structures, systems, or components.
(4) The licensee must document processes and procedures and
maintain records for implementing the corrective actions, compensatory
measures, and maintenance, inspection, testing, and calibration of
security SSCs.
(i) Suspension of security measures. (1) The licensee may suspend
implementation of affected requirements of this section in accordance
with Sec. 53.740(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.
(j) 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.
Sec. 73.110 Technology-inclusive requirements for protection of
digital computer and communication systems and networks.
(a) Each licensee that is licensed to operate a commercial nuclear
plant under 10 CFR part 53 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
[[Page 15875]]
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. 53.210 of this chapter.
(2) Adversely impacting the security functions performed by digital
assets necessary for implementing the physical security requirements in
Sec. 53.860(a) of this chapter.
(b) To protect digital computer and communication systems and
networks associated with the functions described in paragraphs (a)(1)
and (2) of this section, 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 their 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;
(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.
Sec. 73.120 Access authorization program for commercial nuclear
plants.
(a) Introduction and scope. Each applicant for an operating license
or a holder of a combined license under 10 CFR part 53 must establish,
maintain, and implement an access authorization program before initial
fuel load into the reactor (or, for a fueled manufactured reactor,
before initiating the removal of the features to prevent criticality
required under Sec. 53.620(d)(1) of this chapter). The requirements in
this section apply to applicants and licensees who demonstrate
compliance with 10 CFR 73.100(a)(1)(i).
(b) Applicability. (1) The following individuals must be subject to
an access authorization program under this section:
(i) Any individual to whom a licensee intends to grant unescorted
access to a commercial nuclear plant protected area, vital area, or
controlled access area where licensed material is used or stored;
(ii) Any individual whose duties and responsibilities permit the
individual to take actions by electronic means, either on site or
remotely, that could adversely impact the licensee's or applicant's
operational safety, security, or emergency preparedness;
(iii) Any individual who has responsibilities for implementing a
licensee's or applicant's protective strategy, including armed security
force
[[Page 15876]]
officers, alarm station operators, and tactical response team leaders
but not including Federal, State, or local law enforcement personnel;
and
(iv) The licensee or applicant access authorization program
reviewing official or contractor or vendor access authorization program
reviewers.
(2) The licensee or applicant may subject other individuals,
including employees of a contractor or a vendor who are designated in
access authorization program procedures, to an access authorization
program that demonstrates compliance with the requirements of this
section.
(c) General performance objectives and requirements. Each
licensee's or applicant's access authorization program under this
section must demonstrate that the individuals who are specified in
paragraph (b) of this section are trustworthy and reliable, such that
they do not constitute an unreasonable risk to public health and safety
or the common defense and security. The licensee's access authorization
program must maintain the capabilities for demonstrating compliance
with the following performance requirements:
(1) Background investigation. (i)(A) Licensees and applicants must
ensure that any individual seeking initial unescorted access or to
maintain unescorted access is subject to a background investigation.
(B) Background investigations must include the program elements
contained under Sec. 37.25 of this chapter and must also include a
credit history evaluation.
(ii) Background investigations must include fingerprinting and an
FBI identification and criminal history records check in accordance
with Sec. 37.27 of this chapter.
(iii) Licensees must have the informed and signed consent of the
subject individual to initiate a background investigation. This consent
must include authorization to share personal information with other
individuals or organizations as necessary to complete the background
investigation. A signed consent must be obtained prior to any
reinvestigation. The subject individual may withdraw his or her consent
at any time. Licensees must inform the individual that--
(A) If an individual withdraws his or her consent, the licensee may
not initiate any elements of the background investigation that were not
in progress at the time the individual withdrew his or her consent; and
(B) The withdrawal of consent for the background investigation is
sufficient cause for denial or termination of unescorted access
authorization.
(2) Behavioral observation. Licensees, applicants, contractors, and
vendors must ensure the access authorization program includes
provisions that the individuals specified in paragraph (b) of this
section are subject to behavioral observation.
(i) Each person subject to behavioral observation must communicate
to the licensee or applicant observed behaviors or activities of
individuals that may constitute an unreasonable risk to the health and
safety of the public and common defense and security.
(ii) Behavioral observation must include visual observation, in
person or remotely by video, to detect and promptly report to plant
supervision any concerns arising from behavioral observation,
including, but not limited to, concerns related to any questionable
behavior patterns or activities of others.
(3) Self-reporting of legal actions. Licensees or applicants must
inform personnel who are granted and who maintain unescorted access of
their responsibilities to self-report to plant supervision legal
actions taken by a law enforcement authority or court of law against
the individual that could result in incarceration or a court order or
that requires a court appearance, including but not limited to an
arrest, an indictment, the filing of charges, or a conviction, but
excluding minor civil actions or misdemeanors such as parking
violations or speeding tickets, for any individual who has applied for
unescorted access or who maintains unescorted access.
(4) Unescorted access. Licensees or applicants must grant
unescorted access only after the licensee has verified an individual is
trustworthy and reliable. A list of persons currently approved for
unescorted access to a protected area, vital area, or controlled access
area must be maintained at all times. Unescorted access determinations
must be reviewed annually by the reviewing official. Licensees and
applicants must complete an FBI criminal history record check update
for each individual maintaining unescorted access, within 10 years of
the last review.
(5) Termination of unescorted access. Licensees and applicants must
promptly terminate unescorted access when this access is no longer
required or a reviewing official determines an individual is no longer
trustworthy and reliable in accordance with this section.
(6) Determination basis for access. (i) The licensee's or
applicant's reviewing official must determine whether to permit, deny,
unfavorably terminate, maintain, or administratively withdraw an
individual's unescorted access based on an evaluation of all of the
information collected to demonstrate compliance with the requirements
of this section.
(ii) Licensees and applicants must provide individuals subject to
this section, prior to any final adverse determination, the right to
complete, correct, and explain information obtained as a result of the
licensee's background investigation pursuant to Sec. 37.23(g) of this
chapter.
(iii) The licensee's or applicant's reviewing officials are the
only individuals authorized to make unescorted access determination
decisions. Each licensee or applicant must name one or more individuals
to be reviewing officials pursuant to the requirements of Sec.
37.23(b)(2) of this chapter.
(7) Review procedures. Review procedures must be established in
accordance with Sec. 37.23(f) of this chapter, to include provisions
for the notification in writing of individuals who are denied
unescorted access or who are unfavorably terminated.
(8) Protection of information. Licensees, applicants, contractors,
or vendors must establish and maintain a system of files and procedures
in accordance with Sec. 37.31 of this chapter, to ensure personal
information is not disclosed to unauthorized persons.
(9) Access authorization reviews and corrective action. Licensees
and applicants must develop, implement, and maintain procedures for
conduct of access authorization reviews and corrective actions in
accordance with Sec. 37.33 of this chapter to ensure the continuing
effectiveness of the access authorization program and to ensure that
the access authorization program and program elements are in compliance
with the requirements of this section. Each licensee and applicant must
be responsible for the continuing effectiveness of the access
authorization program, including access authorization program elements
that are provided by the contractors or vendors, and the access
authorization programs of any of the contractors or vendors that are
accepted by the licensee or applicant.
(10) Records. Licensees, applicants, and contractors or vendors
must document the processes and procedures for maintaining records used
or created to establish an individual's trustworthiness and reliability
or to document access determinations. Licensees, applicants, and
contractor or vendors must--
(i) Retain documentation regarding the trustworthiness and
reliability of individual employees for 3 years from the date the
individual no longer requires unescorted access;
[[Page 15877]]
(ii) Retain a copy of the current access authorization program
procedures as a record for 3 years after the procedure is no longer
needed. If any portion of the procedure is superseded, retain the
superseded material for 3 years after the record is superseded; and
(iii) Retain the list of persons approved for unescorted access for
3 years after the list is superseded or replaced. Records maintained in
any database(s) must be available for NRC review.
0
162. In Sec. 73.1200, revise paragraphs (a) introductory text, (c)(1)
introductory text, (e)(1) introductory text, (e)(4), (g)(1)
introductory text, (o)(5)(i) and (o)(6)(i), (r), and (s) 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 Sec. 73.100 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 Sec. 73.100 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 Sec. 73.100 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 Sec.
73.100, 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 Sec. 73.100 must notify the NRC Headquarters
Operations Center within 8 hours after time of discovery of the
following facility security program failures involving--
* * * * *
(o) * * *
(5) * * *
(i) Licensees must establish the requested continuous
communications channel once the licensee has completed other required
notifications under this section, Sec. 50.72 of this chapter, appendix
E to part 50 of this chapter, Sec. 53.1630 of this chapter, Sec.
70.50 of this chapter; or Sec. 72.75 of this chapter; as appropriate.
* * * * *
(6) * * *
(i) Licensees must establish the requested continuous
communications channel once the licensee or the movement control center
has completed other required notifications under this section, Sec.
50.72 of this chapter, appendix E to part 50 of this chapter, Sec.
53.1630 of this chapter, Sec. 70.50 of this chapter; Sec. 72.75 of
this chapter; or requested assistance from the LLEA, as appropriate.
* * * * *
(r) Declaration of emergencies. Licensees notifying the NRC of the
declaration of an emergency class must do so in accordance with
Sec. Sec. 50.72, 53.1630, 63.73, 70.50, and 72.75 of this chapter, as
applicable.
(s) Elimination of duplication. Licensees with notification
obligations under paragraphs (a) through (h), (m), and (n) of this
section and Sec. Sec. 50.72, 53.1630, 63.73, 70.50, and 72.75 of this
chapter may notify the NRC of events in a single communication. This
communication must identify each regulation under which the licensee is
reporting.
* * * * *
0
163. 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 Sec. 53.1640 of this
chapter must prepare the written follow-up report on NRC Form 366.
(ii) Licensees not subject to Sec. 50.73 or Sec. 53.1640 of this
chapter must prepare the written follow-up report in a letter format.
* * * * *
0
164. 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 Sec. 73.100, 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, 53, 60, 63, 70, and 72 of this chapter, or both.
* * * * *
0
165. In Sec. 73.1215, revise 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 Sec. 73.100, the licensees must report activities they
assess are suspicious. Examples include, but are not limited to, the
following:
* * * * *
0
166. 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, 53, 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
167. 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,
2297f); Energy Reorganization Act of 1974, secs. 201, 202 (42 U.S.C.
5841, 5842); 44 U.S.C. 3504 note.
0
168. In Sec. 74.31, revise paragraph (a) introductory text 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,
[[Page 15878]]
other than a production or utilization facility licensed pursuant to
part 50, part 53, 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
169. In Sec. 74.41, revise paragraph (a) introductory text 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
53 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
170. In Sec. 74.51, revise paragraph (a) introductory text 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 53 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
171. 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.
Section 75.4 also issued under Nuclear Waste Policy Act secs.
135 (42 U.S.C. 10155, 10161).
0
172. In Sec. 75.4, revise the second paragraph of the introductory
text and the definition for ``Facility'' to read as follows:
Sec. 75.4 Definitions.
* * * * *
Unless otherwise defined in this section, the terms defined in
Sec. Sec. 40.4, 50.2, 53.020, and 70.4 of this chapter have the same
meaning when used in this part.
* * * * *
Facility means:
(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, 53, 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
173. 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
174. In Sec. 95.5, revise the definition for ``License'' to read as
follows:
Sec. 95.5 Definitions.
* * * * *
License means a license issued under 10 CFR part 50, 52, 53, 54,
60, 63, 70, or 72.
* * * * *
Sec. 95.39 [Amended]
0
175. In Sec. 95.39(a), remove ``part 52'' and add in its place ``part
52 or part 53.''
PART 140--FINANCIAL PROTECTION REQUIREMENTS AND INDEMNITY
AGREEMENTS
0
176. 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
177. In Sec. 140.2, revise paragraphs (a)(1) and (2) to read as
follows:
Sec. 140.2 Scope.
(a) * * *
(1) To each person who is an applicant for or holder of a license
issued under 10 CFR part 50, 52, 53, or 54 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 53 of this chapter and a combined
license under part 52 or part 53 of this chapter), and to other persons
indemnified with respect to the involved facilities.
* * * * *
0
178. 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, 53 or 54 to operate a
nuclear reactor, or is the applicant for or holder of a combined
license issued under 10 CFR part 52, 53, or 54, except licenses held by
persons found by the Commission to be Federal agencies or nonprofit
educational institutions licensed to conduct educational activities.
This subpart also applies to persons licensed to possess and use
plutonium in a plutonium processing and fuel fabrication plant.
0
179. 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, 53, or 54 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
[[Page 15879]]
reactors; provided, that such primary financial protection covers all
reactors at the location.
0
180. 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, 53, or 54 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
181. 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.
Each holder of a 10 CFR part 50 or 10 CFR part 53 construction
permit, or a holder of a combined license under part 52 or part 53 of
this chapter before the date that the Commission had made the finding
under Sec. 52.103(g) or Sec. 53.1452(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
53, or a combined license under 10 CFR part 52 or 53, shall, during the
period before issuance of a license authorizing operation under 10 CFR
part 50 or 53, or the period before the Commission makes the finding
under Sec. 52.103(g) or Sec. 53.1452(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
182. In Sec. 140.20, revise paragraphs (a)(1)(i) and (ii) 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 53 of this chapter) authorizing the licensee to operate the
nuclear reactor involved; or
(ii) The date that the Commission makes the finding under Sec.
52.103(g) or Sec. 53.1452(g) of this chapter; or
* * * * *
PART 150--EXEMPTIONS AND CONTINUED REGULATORY AUTHORITY IN
AGREEMENT STATES AND IN OFFSHORE WATERS UNDER SECTION 274
0
183. 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.
Sections 150.3, 150.15, 150.15a, 150.31, 150.32 also issued
under Atomic Energy Act secs. 11e(2), 81, 83, 84 (42 U.S.C.
2014e(2), 2111, 2113, 2114).
Section 150.14 also issued under Atomic Energy Act sec. 53 (42
U.S.C. 2073).
Section 150.15 also issued under Nuclear Waste Policy Act sec.
135 (42 U.S.C. 10155, 10161).
Section 150.17a also issued under Atomic Energy Act sec. 122 (42
U.S.C. 2152).
Section 150.30 also issued under Atomic Energy Act sec. 234 (42
U.S.C. 2282).
0
184. 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 (GTCC) waste, as defined in part 72 of
this chapter, in an ISFSI or an MRS licensed under part 72 of this
chapter; the GTCC waste must originate in, or be used by, a facility
licensed under part 50, part 52, or part 53 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 53 of this chapter and is licensed under part
30 and/or part 70 of this chapter.
* * * * *
PART 170--FEES FOR FACILITIES, MATERIALS, IMPORT AND EXPORT
LICENSES, AND OTHER REGULATORY SERVICES UNDER THE ATOMIC ENERGY ACT
OF 1954, AS AMENDED
0
185. The authority citation for 10 CFR part 170 continues to read as
follows:
Authority: Atomic Energy Act of 1954, secs. 11, 161(w) (42
U.S.C. 2014, 2201(w)); Energy Reorganization Act of 1974, sec. 201
(42 U.S.C. 5841); 42 U.S.C. 2215; 31 U.S.C. 901, 902, 9701; 44
U.S.C. 3504 note.
0
186. In Sec. 170.3, revise the definitions for ``Manufacturing
License,'' ``Part 55 Reviews,'' ``Power reactor,'' and ``Special
projects'' to read as follows:
Sec. 170.3 Definitions.
* * * * *
Manufacturing license means a license under subpart F of part 52 of
this chapter or subpart H of part 53 of this chapter to manufacture a
nuclear power reactor(s) to be operated at sites not identified in the
license application.
* * * * *
Part 55 Reviews as used in this part means those services provided
by the Commission to administer requalification and replacement
examinations and tests for reactor operators licensed under 10 CFR part
55 or part 53 of the Commission's regulations and employed by part 50
or part 53 licensees. These services also include related items such as
the preparation, review, and grading of the examinations and tests.
* * * * *
Power reactor means a nuclear reactor designed to produce
electrical or heat energy licensed by the Commission under the
authority of section 103 or subsection 104b of the Act, and under the
provisions of Sec. 50.21(b), Sec. 50.22, or part 53 of this chapter.
* * * * *
Special projects means specific services provided by the Commission
for which fees are not otherwise specified in this chapter. This
includes, but is not limited to, contested hearings on licensing
actions directly related to U.S. Government national security
initiatives (as determined by the NRC), topical report reviews, early
site reviews, waste solidification activities, activities related to
the tracking and monitoring of shipment of classified matter, services
provided to certify licensee, vendor, or other private industry
personnel as instructors for 10 CFR part 55 or part 53 reactor
operators, reviews of financial assurance submittals that do not
require a license amendment, reviews of responses to Confirmatory
Action Letters, reviews of uranium recovery licensees' land-use survey
reports, and reviews of Sec. 50.71 or Sec. 53.1545 of this chapter
Final Safety Analysis Reports. Special projects does not include
activities otherwise exempt from fees under this part. It also does not
include those contested hearings for which a fee exemption is granted
in Sec. 170.11(a)(2), including those related to individual plant
security modifications.
* * * * *
0
187. In Sec. 170.12, revise paragraph (d)(1)(v) to read as follows:
Sec. 170.12 Payment of fees.
* * * * *
(d) * * *
(1) * * *
[[Page 15880]]
(v) 10 CFR 50.71 or 53.1545 Final Safety Analysis Reports;
* * * * *
Sec. 170.21 [Amended]
0
188. In Sec. 170.21, in footnote 1 remove the phrase ``(e.g., 10 CFR
50.12, 10 CFR 73.5)'' and add in its place the phrase ``(e.g., 10 CFR
50.12, 10 CFR 53.080, 10 CFR 73.5)''.
0
189. Revise Sec. 170.41 to read as follows:
Sec. 170.41 Failure by an applicant or licensee to pay prescribed
fees.
If the Commission determines that an applicant or a licensee has
failed to pay a prescribed fee required in this part, the Commission
will not process any application and may suspend or revoke any license
or approval issued to the applicant or licensee. The Commission may
issue an order with respect to licensed activities that the Commission
determines to be appropriate or necessary to carry out the provisions
of this part, parts 30, 31, 32 through 35, 40, 50, 53, 61, 70, 71, 72,
73, and 76 of this chapter, and of the Act.
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
0
190. The authority citation for 10 CFR part 171 continues to read as
follows:
Authority: Atomic Energy Act of 1954, secs. 11, 161(w), 223, 234
(42 U.S.C. 2014, 2201(w), 2273, 2282); Energy Reorganization Act of
1974, sec. 201 (42 U.S.C. 5841); 42 U.S.C. 2215; 44 U.S.C. 3504
note.
0
191. Revise Sec. 171.3 to read as follows:
Sec. 171.3 Scope.
The regulations in this part apply to any person holding an
operating license for a test reactor or research reactor issued under
part 50 of this chapter, and to any person holding an operating license
for a power reactor licensed under 10 CFR part 50 or part 53, or a
combined license issued under 10 CFR part 52 or part 53, that has
provided notification to the U.S. Nuclear Regulatory Commission (NRC)
that the licensee has successfully completed power ascension testing.
The regulations in this part also apply to any person holding a
materials license as defined in this part, a Certificate of Compliance,
a sealed source or device registration, a quality assurance program
approval, and to a Government agency as defined in this part.
Notwithstanding the other provisions in this section, the regulations
in this part do not apply to uranium recovery and fuel facility
licensees until after the Commission verifies through inspection that
the facility has been constructed in accordance with the requirements
of the license.
0
192. In Sec. 171.5, revise the definitions for ``Operating license''
and ``Power reactor'' to read as follows:
Sec. 171.5 Definitions.
* * * * *
Operating license means having a license issued under Sec. 50.57
or Sec. 53.1387 of this chapter. It does not include licenses that
only authorize possession of special nuclear material after the
Commission has received a request from the licensee to amend its
licensee to permanently withdraw its authority to operate or the
Commission has permanently revoked such authority.
* * * * *
Power reactor means a nuclear reactor designed to produce
electrical or heat energy and licensed by the Commission under the
authority of section 103 or subsection 104b of the Atomic Energy Act of
1954, as amended, and under the provisions of Sec. 50.21(b) or Sec.
50.22, or part 53 of this chapter.
* * * * *
0
193. In Sec. 171.15, revise paragraphs (a), (b)(2)(iii), (c)(1), and
(d)(1) to read as follows:
Sec. 171.15 Annual fees: Non-power production or utilization
licenses, reactor licenses, and independent spent fuel storage
licenses.
(a) Each person holding an operating license for one or more non-
power production or utilization facilities under 10 CFR part 50 that
has provided notification to the NRC of the successful completion of
startup testing; each person holding an operating license for a power
reactor licensed under 10 CFR part 50 or a combined license under 10
CFR part 52, or an operating license or combined license for a
commercial nuclear plant under 10 CFR part 53, that has provided
notification to the NRC of the successful completion of power ascension
testing; each person holding a 10 CFR part 50 or part 52, power reactor
license, or a 10 CFR part 53 commercial nuclear plant license that is
in decommissioning or possession only status, except those that have no
spent fuel onsite; and each person holding a 10 CFR part 72 license who
does not hold a 10 CFR part 50, part 52, or part 53 license and
provides notification under Sec. 72.80(g) of this chapter, shall pay
the annual fee for each license held during the Federal fiscal year in
which the fee is due. This paragraph (a) does not apply to test or
research reactors exempted under Sec. 171.11(b).
(b) * * *
(2) * * *
(iii) Generic activities required largely for NRC to regulate power
reactors (e.g., updating part 50, part 52, or part 53 of this chapter,
operating the Incident Response Center, new reactor regulatory
infrastructure). The base annual fee for operating power reactors does
not include generic activities specifically related to reactor
decommissioning.
(c)(1) The FY 2025 annual fee for each power reactor holding a 10
CFR part 50 or part 53 operating license or combined license issued
under 10 CFR part 52 or part 53 that is in a decommissioning or
possession-only status and has spent fuel onsite, and for each
independent spent fuel storage 10 CFR part 72 licensee who does not
hold a 10 CFR part 50 or part 53 operating license, or a 10 CFR part 52
or part 53 combined license, is $326,000.
* * * * *
(d)(1) Each person holding an operating license for an SMR issued
under 10 CFR part 50 or part 53, or a combined license issued under 10
CFR part 52 or part 53, that has provided notification to the NRC of
the successful completion startup testing, shall pay the annual fee for
all licenses held for an SMR site. The annual fee will be determined
using the cumulative licensed thermal power rating of all SMR units and
the bundled unit concept, during the fiscal year in which the fee is
due. For a given site, the use of the bundled unit concept is
independent of the number of SMR plants, the number of SMR licenses
issued, or the sequencing of the SMR licenses that have been issued.
* * * * *
0
194. In Sec. 171.17, revise paragraphs (a) introductory text,
(a)(1)(ii), and (a)(2) to read as follows:
Sec. 171.17 Proration.
* * * * *
(a) Reactors, 10 CFR part 72 licensees who do not hold 10 CFR part
50, 10 CFR part 52, or 10 CFR part 53 licenses, and materials licenses
with annual fees of $100,000 or greater for a single fee category. The
NRC will base the proration of annual fees for terminated and
downgraded licenses on the fee rule in effect at the time the action is
official. The NRC will base the determinations on the proration
requirements under paragraphs (a)(2) and (3) of this section.
(1) * * *
[[Page 15881]]
(ii) The annual fees for new licenses for non-power production or
utilization facilities, 10 CFR part 72 licensees who do not hold 10 CFR
part 50, part 52, or part 53 licenses, and materials licenses with
annual fees of $100,000 or greater for a single fee category for the
current FY, that are subject to fees under this part and are granted a
license to operate on or after October 1 of a FY, are prorated on the
basis of the number of days remaining in the FY. Thereafter, the full
annual fee is due and payable each subsequent FY.
(2) Terminations. The base operating power reactor annual fee for
operating reactor licensees or the annual fee for small modular reactor
licensees, who have requested amendment to withdraw operating authority
permanently during the FY will be prorated based on the number of days
during the FY the license was in effect before docketing of the
certifications for permanent cessation of operations and permanent
removal of fuel from the reactor vessel or when a final legally
effective order to permanently cease operations has come into effect.
The spent fuel storage/reactor decommissioning annual fee for reactor
licensees who permanently cease operations and have permanently removed
fuel from the site during the FY will be prorated on the basis of the
number of days remaining in the FY after docketing of both the
certifications of permanent cessation of operations and permanent
removal of fuel from the site. The spent fuel storage/reactor
decommissioning annual fee will be prorated for those 10 CFR part 72
licensees who do not hold a 10 CFR part 50, part 52, or part 53 license
who request termination of the 10 CFR part 72 license and permanently
cease activities authorized by the license during the FY based on the
number of days the license was in effect before receipt of the
termination request. The annual fee for materials licenses with annual
fees of $100,000 or greater for a single fee category for the current
FY will be prorated based on the number of days remaining in the FY
when a termination request or a request for a possession-only license
is received by the NRC, provided the licensee permanently ceased
licensed activities during the specified period. The annual fee for
non-power production or utilization facilities will be prorated based
on the number of days remaining in the FY when the authorization to
operate the facility has been permanently removed from the license
during the FY.
* * * * *
Dated: March 25, 2026.
For the Nuclear Regulatory Commission.
Tomas Herrera,
Acting Secretary of the Commission.
[FR Doc. 2026-06048 Filed 3-27-26; 8:45 am]
BILLING CODE 7590-01-P