[Federal Register Volume 91, Number 135 (Thursday, July 16, 2026)]
[Proposed Rules]
[Pages 44560-44716]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2026-14341]
[[Page 44559]]
Vol. 91
Thursday,
No. 135
July 16, 2026
Part V
Nuclear Regulatory Commission
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10 CFR Parts 2, 50, 51 et al.
Modernizing Reactor Licensing, Safety Oversight, and Siting Practices;
Proposed Rule
Federal Register / Vol. 91 , No. 135 / Thursday, July 16, 2026 /
Proposed Rules
[[Page 44560]]
NUCLEAR REGULATORY COMMISSION
10 CFR Parts 2, 50, 51, 52, 53, 54, 71, and 100
[NRC-2025-0975]
RIN 3150-AL44
Modernizing Reactor Licensing, Safety Oversight, and Siting
Practices
AGENCY: Nuclear Regulatory Commission.
ACTION: Proposed rule and guidance; request for comment.
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SUMMARY: Consistent with Executive Order 14300, ``Ordering the Reform
of the Nuclear Regulatory Commission,'' the U.S. Nuclear Regulatory
Commission (NRC) is conducting a review and wholesale revision of its
regulations. This proposed rule aims to modernize reactor licensing,
safety oversight, and siting practices addressing sections 5(f), 5(h),
and 5(i) of Executive Order 14300, and additional items that contribute
to adding additional generation to the electrical grid. Additionally,
as part of the NRC's overarching review of all of its regulations, the
agency identified a number of further changes to the NRC's regulations
that will improve the efficiency and efficacy of its licensing process
that are also included in this rulemaking.
DATES: Comments must be submitted electronically using https://www.regulations.gov by 11:59 p.m. eastern time on August 31, 2026.
Comments received after this date will be considered if it is practical
to do so, but the Commission is able to ensure consideration of only
comments received before this date.
ADDRESSES: Submit your comments, identified by Docket ID NRC-2025-0975,
at https://www.regulations.gov. If your material cannot be submitted
using https://www.regulations.gov, call or email the individuals listed
in the FOR FURTHER INFORMATION CONTACT section of this document for
alternate instructions.
Do not include any personally identifiable information (such as
name, address, or other contact information) or confidential business
information that you do not want publicly disclosed. All comments are
public records; they are publicly displayed exactly as received, and
will not be deleted, modified, or redacted. Comments may be submitted
anonymously.
Follow the search instructions on https://www.regulations.gov to
view public comments.
You can read a plain language description of this proposed rule at
https://www.regulations.gov/docket/NRC-2025-0975. For additional
direction on obtaining information and submitting comments, see
``Obtaining Information and Submitting Comments'' in the SUPPLEMENTARY
INFORMATION section of this document.
FOR FURTHER INFORMATION CONTACT: Daniel Doyle, U.S. Nuclear Regulatory
Commission, Washington, DC 20555-0001, telephone: 301-415-3748, email:
[email protected].
SUPPLEMENTARY INFORMATION:
Executive Summary
A. Need for the Regulatory Action
On May 23, 2025, President Donald J. Trump signed Executive Order
(E.O.) 14300, ``Ordering the Reform of the Nuclear Regulatory
Commission.'' Section 5, ``Reforming and Modernizing the NRC's
Regulations,'' requires the NRC to undertake a review and wholesale
revision of its regulations in title 10 of the Code of Federal
Regulations (10 CFR) and guidance documents as guided by the policies
set forth in section 2 of the E.O. This rulemaking addresses section
5(f) of E.O. 14300, which directs the NRC to ``[e]stablish stringent
thresholds for circumstances in which the NRC may demand changes to
reactor design once construction is underway''; section 5(h) of E.O.
14300, which directs the NRC to ``[a]dopt revised and, where feasible,
determinate and data-backed thresholds to ensure that reactor safety
assessments focus on credible, realistic risks''; and section 5(i) of
E.O. 14300, which directs the NRC to ``[r]econsider the regulations
governing the time period for which a renewed license remains
effective, and extend that period as appropriate based on available
technological and safety data.'' Additionally, as part of the NRC's
overarching review of all of its regulations, the agency identified a
number of additional changes to the NRC's regulations that will improve
the efficiency and efficacy of its licensing process that are also
included in this rulemaking.
These changes are the culmination of decades of combined
experience, feedback from nuclear experts, lessons learned by the NRC
and industry, international experience, and prior efforts to modernize
the regulatory framework. They have undergone thoughtful preparation
and internal vetting by the NRC technical experts. While some changes
in this rule had not been advanced as regulatory priorities, the
direction in the E.O. catalyzed agency efforts to accelerate
modernization of 10 CFR part 50, ``Domestic Licensing of Production and
Utilization Facilities,'' and 10 CFR part 52, ``Licenses,
Certifications, and Approvals for Nuclear Power Plants,'' and make
conforming changes to 10 CFR part 53, ``Risk-Informed, Technology-
Inclusive Regulatory Framework for Commercial Nuclear Plants.''
Therefore, this proposed rule aims to modernize reactor licensing,
safety oversight, and siting practices addressing sections 5(f), 5(h),
and 5(i) of E.O. 14300 and additional items that would contribute to
adding generation to the electrical grid. In developing the proposed
changes, the NRC has considered the benefits of increased availability
of, and innovation in, nuclear power to our economic and national
security consistent with section 501(a) of the Accelerating Deployment
of Versatile, Advanced Nuclear for Clean Energy Act of 2024 (ADVANCE
Act) and section 3 of E.O. 14300.
B. Major Provisions
Major provisions of this proposed rule include changes in the
following areas:
Expedited Construction of Certain Structures, Systems, and Components
The NRC is proposing to amend its regulations applicable to the
definition of construction in 10 CFR 50.10, ``License required; limited
work authorization,'' 10 CFR 51.4, ``Definitions,'' and 10 CFR 53.020,
``Definitions,'' in order to focus the scope of activities that are
considered construction on structures, systems, and components (SSCs)
for which construction can affect attributes of the SSC material to the
SSC's capability to perform a safety-significant function and thereby
reduce the cost impact of the current definition of construction.
Additionally, the NRC proposes to include new paragraphs in 10 CFR
50.10(h) and 53.1130(e), which would issue a general license for
beginning construction upon docketing an application for a license that
would authorize construction of a nuclear plant, subject to conditions
that ensure safety, security, and appropriate environmental review.
Determinate and Data-Backed Thresholds for Reactor Safety Assessments
The NRC is proposing to amend its regulations by revising 10 CFR
50.2, ``Definitions,'' to add the terms ``design basis event'' (DBE)
and ``beyond design basis event'' (BDBE). The proposed changes also
include a conforming revision to the definition of DBE in 10 CFR 50.49,
``Environmental
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qualification of electric equipment important to safety for nuclear
power plants.'' In parallel with the proposed changes, the NRC has
developed draft regulatory guidance (DG)-1454, ``Implementation of
Determinate and Data-Backed Thresholds for Reactor Safety
Assessments,'' which (1) establishes determinate and data-backed
thresholds for categorizing events into defined bins, (2) outlines
graded assessment approaches for DBEs and BDBEs, and (3) clarifies the
process for selecting ``design bases'' controlling parameters used as
reference bounds in the design of SSCs.
Removal of IEEE-323-1974 Reference in Footnote 3 of 10 CFR 50.49
The NRC is proposing to revise the regulations in 10 CFR 50.49 to
delete footnote 3, which clarifies that safety-related electric
equipment is referred to as Class 1E equipment in Institute of
Electrical and Electronics Engineers (IEEE) Standard 323-1974. The
reference in this footnote is now unnecessary because the NRC has
established the connection between ``safety-related'' electric
equipment and ``Class 1E'' equipment elsewhere.
Expanded Alternative Requests Under 10 CFR 50.55a(z)
The NRC is proposing to amend its regulations to allow licensees to
request a broader scope of alternatives to the requirements in 10 CFR
50.55a, ``Codes and standards.'' Currently, alternatives under 10 CFR
50.55a(z), ``Alternatives to codes and standards requirements,'' are
limited to the requirements in 10 CFR 50.55a(b), ``Use and conditions
on the use of standards,'' through (h), ``Protection and safety
systems.'' Expanding the scope of alternatives permitted under 10 CFR
50.55a to all requirements in 10 CFR 50.55a would allow for added
flexibility without requiring exemptions, while relying on the well-
understood existing criteria of acceptable level of quality and safety
(10 CFR 50.55a(z)(1)) and hardship without a compensating increase in
quality or safety (10 CFR 50.55a(z)(2)) for consistent and predictable
regulatory outcomes. This proposed action would allow nuclear power
plant licensees and applicants for construction permits (CP), operating
licenses (OL), combined licenses (COL), standard design certifications,
standard design approvals, and manufacturing licenses (ML) to request
authorization of voluntary alternatives to a broader scope of
requirements in 10 CFR 50.55a.
Risk-Informing 10 CFR 50.59 and Allowing Flexibility for Changes to
Methods
There are two proposed changes to 10 CFR 50.59, ``Changes, tests,
and experiments.'' The first proposed change to the regulation would
allow the use of quantitative risk results to demonstrate a change to
the facility does not result in a ``more than minimal increase'' as the
phrase is used in 10 CFR 50.59(c)(2)(i) and (ii). The second proposed
change would allow licensees to make changes to methods that would
previously have required NRC review under 10 CFR 50.59(c)(2)(viii) or
53.1550(a)(2)(iv), provided the licensee adopts an acceptable
verification, validation, and uncertainty quantification (VVUQ) program
in accordance with a proposed new 10 CFR 50.221, ``Credibility
requirements for modeling and simulation.'' This shift would enable
licensee-led evaluations and allow for the evaluation of advanced
modeling methods through structured processes rather than fixed
requirements.
Minimum Decommissioning Funding Assurance for Non-Large Light-Water
Reactors
The NRC is proposing rule changes in 10 CFR 50.75, ``Reporting and
recordkeeping for decommissioning planning,'' to allow certain new
reactor applicants and licensees to submit a design-specific
decommissioning cost estimate that is less than the approved table of
minimum amounts (i.e., ``formula'') values in 10 CFR 50.75(c). The
current regulations restrict minimum funding assurance for
decommissioning to the table of minimum amounts values or greater. The
NRC is proposing similar changes to 10 CFR part 53, which currently
only allows for a site-specific decommissioning cost estimate as the
certification amount. These changes would more broadly accommodate new
reactor technologies.
Incorporation of Streamlined Quality Assurance Criteria for Nuclear
Power Plants and Fuel Reprocessing Plants
This proposed rule would incorporate an appendix T, ``Streamlined
Quality Assurance Criteria for Nuclear Power Plants and Fuel
Reprocessing Plants,'' to 10 CFR part 50. The proposed appendix T to 10
CFR part 50 would provide a voluntary alternative to appendix B,
``Quality Assurance Criteria for Nuclear Power Plants and Fuel
Reprocessing Plants,'' to 10 CFR part 50 that applicants, who meet
certain conditions, could use in their respective applications. This
action would support implementation of E.O. 14300, section 5, through
(1) the establishment of performance-based quality assurance (QA)
criteria that provide explicit direction on the use of a graded
approach for applying QA requirements to SSCs relative to their safety
and risk contributions to the overall nuclear facility; (2) the
incorporation of QA terminology and methodologies used across various
industries; (3) enhanced regulatory certainty during the application
process; and (4) the removal, when appropriate, of NRC oversight of
suppliers and vendors of products and services related to nuclear power
plant and fuel reprocessing plant SSCs subject to this proposed rule.
Updates to Construction Permit Requirements and Related Licenses
The proposed rule would revise the content of applications relevant
to technical information required for CPs in 10 CFR 50.34, ``Contents
of applications; technical information,'' to remove some overly-
prescriptive wording; clarify that the level of detail provided in a
preliminary safety analysis report should be sufficient to permit the
NRC to make the findings in 10 CFR 50.35(a), 10 CFR 50.40, ``Common
standards,'' and 10 CFR 50.50, ``Issuance of licenses and construction
permits''; and adjust some of the light-water reactor (LWR)-centric
language to be more technology-inclusive. Conforming changes to
parallel sections in 10 CFR part 52 are also proposed.
Alternative Risk-Informed and Performance-Based Acceptance Criteria for
10 CFR Parts 50 and 52
The NRC is proposing to amend its regulations by adding new,
standalone provisions in 10 CFR 50.220 and 10 CFR 52.220, entitled
``Use of risk-informed and performance-based alternatives to acceptance
criteria.'' These provisions would allow licensees and applicants to
voluntarily submit and use technology-inclusive, risk-informed or
performance-based acceptance criteria as alternatives to existing
prescriptive requirements.
In addition, the NRC is proposing revisions to appendix A to 10 CFR
part 50 to update and clarify its regulations for General Design
Criteria (GDC). The proposed changes would explicitly allow deviations
from the GDCs without requiring exemptions and would revise GDC 28,
``Reactivity limits,'' to remove the prescriptive requirement to
evaluate control rod ejection and drop accidents. Instead, applicants
would be permitted to propose an alternative design basis accident for
reactivity control systems.
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Establishing Thresholds for Changes to Reactor Designs During
Construction and Operation Under 10 CFR Parts 52 and 53
In response to E.O. 14300, section 5(f), the NRC is proposing to
amend its regulations related to reactor design changes made during
construction and operation under the 10 CFR parts 52 and 53 licensing
approaches. These changes would impact licensees that reference a
certified design or manufacturing license under 10 CFR part 52 or 53.
The objectives of these proposed changes are to establish appropriate
thresholds for NRC-initiated changes as well as provide additional
flexibility and reduce unnecessary regulatory burden for licensee-
initiated changes.
Revision of the Emergency Preparedness Regulations for Nuclear Power
Reactors
The NRC is proposing to amend its regulations to create adaptable
licensing pathways for emergency preparedness (EP). Consistent with
E.O. 14300, the NRC's objectives for this proposed rule are to
streamline the licensing process, provide regulatory certainty for the
deployment of new reactor technologies, and remove prescriptive
language of lesser safety significance for licensed facilities. Central
to these proposed changes is a strengthened, more risk-informed
approach to EP for providing reasonable assurance that adequate
protective measures can and will be taken in the event of a
radiological emergency.
Optional Submittal of Operational Programs
This rulemaking would allow a developer the option to voluntarily
submit operational programs for NRC review and approval with an ML
application. The intent is to allow construction permit/operating
license (CP/OL) and COL applicants the flexibility to reference the
standardized programs approved in the ML, use their own approved
programs, or use a combination of both. Early review of these programs
would support streamlined CP/OL or COL reviews.
Early Site Permit for Nuclear Power Plants
The NRC is proposing to amend its regulations by revising the
provisions applicable to early site permit (ESP) licensing and approval
processes for nuclear power plants. These amendments would eliminate
the requirement for an ESP expiration date, clarify the applicability
of various requirements to ESPs, and propose necessary conforming
amendments throughout the NRC's regulations to enhance the NRC's
necessary regulatory effectiveness and efficiency in implementing its
licensing and approval processes.
Manufacturing License Term Extension
The NRC is proposing to amend the regulations in 10 CFR 52.173,
``Duration of manufacturing license,'' and 52.181, ``Duration of
renewal,'' to change the duration of an ML to 40 years and the duration
of the renewed ML to 40 years. By amending the regulations with these
proposed changes, the ML would be consistent with the durations for
certified designs, thereby increasing efficiency in building new
reactors.
Nuclear Power Plant License Renewal
The NRC is proposing to amend its regulations for renewing nuclear
power plant OLs. The revisions would extend the duration of renewed
licenses, allow applicants to voluntarily propose alternative risk-
informed and performance-based criteria, and remove several
prescriptive requirements related to the application process and post-
approval recordkeeping. These changes would enhance regulatory
flexibility and efficiency to facilitate operational extensions for the
current nuclear fleet.
Enhancing Flexibility of Reactor Site Criteria
The existing regulatory framework requires all stationary power
reactor applications submitted after January 10, 1997, to follow the
siting criteria in subpart B to 10 CFR part 100, ``Evaluation Factors
for Stationary Power Reactor Site Applications on or After January 10,
1997,'' without consideration of reactor type, size, output,
radiological consequence, or other factors that can widely vary given
the breadth of power reactor designs considered for future construction
and deployment in the United States. Subpart A, ``Evaluation Factors
for Stationary Power Reactor Site Applications Before January 10, 1997
and for Testing Reactors,'' to 10 CFR part 100, ``Reactor Site
Criteria,'' provides less prescriptive regulatory requirements for
reactor siting but only applies to power reactor applications submitted
prior to January 10, 1997, or an application for a testing reactor, as
defined in 10 CFR 50.2. To increase the flexibility of regulatory
requirements for the full spectrum of prospective reactor technologies,
including non-stationary reactors, this proposed change would (1)
revise subpart A to 10 CFR part 100 to include Tier 1 power reactors,
as defined in proposed 10 CFR 100.3, ``Definitions,'' and DG-4036,
``Graded Approach to Site Characterization for New Reactor
Applications,'' in addition to testing reactors; (2) remove appendix A
to 10 CFR part 100 that applies to subpart A to 10 CFR part 100; and
(3) revise subpart B to 10 CFR part 100 to include Tier 2 power
reactors, which are those reactors that do not meet the entry criteria
for subpart A to 10 CFR part 100. This proposed change will be
accompanied by draft guidance on application content, including the
entry criteria for subpart A to 10 CFR part 100, a clarification on
site parameters to be included in a site parameter envelope, and an
explanation of a graded approach to site characterization for all
external hazards to be considered under both subparts A and B to 10 CFR
part 100. In addition, the NRC proposes to revise 10 CFR part 100 to
maintain the agency's long-standing preference for siting reactors in
areas of low population density, while providing flexibility to allow
siting reactors in areas of greater population density when justified
by an assessment comparing the societal risks and societal benefits of
siting reactors in those areas. Implementing guidance for these
assessments will be developed.
Increased Enrichment of Conventional and Accident Tolerant Fuel Designs
for Light-Water Reactors
The NRC is proposing to amend its regulations related to the use of
conventional and accident tolerant LWR fuel designs. The NRC's goal is
to establish effective and efficient licensing of the use of fuels
enriched to greater than 5.0 weight percent uranium-235 while
continuing to provide reasonable assurance of adequate protection of
public health and safety. The new requirements also would address fuel
fragmentation, relocation, and dispersal in relation to the key
accident tolerant fuel components of increased enrichment and burnup
limits.
C. Cost and Benefits
The NRC prepared a draft regulatory analysis to determine the
expected quantitative and qualitative costs of the proposed rule and
associated guidance. The draft regulatory analysis concluded that the
proposed rule and associated guidance would result in undiscounted
total net savings of $1.86 billion to the NRC and industry ($802.10
million using a 7 percent discount rate and $1.26 billion using a 3
percent discount rate).
The draft regulatory analysis also considers qualitative factors to
be considered in the NRC's rulemaking decision. Qualitative factors
include
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regulatory efficiency. The proposed rule would enable the NRC to better
maintain and administer the new reactor licensing process and ensure
that the requirements for the licensing of new reactors are clear and
appropriate.
For more information, the draft regulatory analysis is available as
indicated in the ``Availability of Documents'' section of this
document.
Table of Contents
I. Obtaining Information and Submitting Comments
A. Obtaining Information
B. Submitting Comments
II. Executive Order 14300: Ordering the Reform of the Nuclear
Regulatory Commission
III. Background--Expedited Construction of Certain Structures,
Systems, and Components
A. Definition of Construction
B. General Licenses
IV. Discussion--Expedited Construction of Certain Structures,
Systems, and Components
A. Definition of Construction
B. Safety Review
C. Environmental Review
D. General Licenses and Generic Finality
V. Background--Determinate and Data-Backed Thresholds for Reactor
Safety Assessments
VI. Discussion--Determinate and Data-Backed Thresholds for Reactor
Safety Assessments
VII. Background--Removal of IEEE-323-1974 Reference in Footnote 3 of
10 CFR 50.49
VIII. Discussion--Removal of IEEE-323-1974 Reference in Footnote 3
of 10 CFR 50.49
IX. Background--Expanded Alternative Requests Under 10 CFR 50.55a(z)
X. Discussion--Expanded Alternative Requests Under 10 CFR 50.55a(z)
XI. Background--Risk-Informing 10 CFR 50.59 and Allowing Flexibility
for Changes to Methods
A. Use of Quantitative Risk Results
B. Improved Flexibility for Changes to Methods of Evaluation
XII. Discussion--Risk-Informing 10 CFR 50.59 and Allowing
Flexibility for Changes to Methods
A. Use of Quantitative Risk Results
B. Improved Flexibility for Changes to Methods of Evaluation
XIII. Background--Minimum Decommissioning Funding Assurance for Non-
Large Light-Water Reactors
XIV. Discussion--Minimum Decommissioning Funding Assurance for Non-
Large Light-Water Reactors
XV. Background--Incorporation of Streamlined Quality Assurance
Criteria for Nuclear Power Plants and Fuel Reprocessing Plants
A. Historic Quality Assurance Requirements Perspectives and
Emergent Issues
B. NRC Responses to These Issues
XVI. Discussion--Incorporation of Streamlined Quality Assurance
Criteria for Nuclear Power Plants and Fuel Reprocessing Plants
A. Introduction and Scope
B. Definitions
C. General Requirements
D. Quality Assurance Requirements
E. Quality Assurance for Software Used in Design and Analysis,
and Digital Items Important to Safety
F. Proposed Conforming Changes to 10 CFR 50.4, 50.34, 50.54,
50.55, 52.79, 53.020, 53.040, 53.460, 53.500, 53.865, 53.1309,
53.1369, 53.1416, and 53.1565
XVII. Background--Updates to Construction Permit Requirements and
Related Licenses
XVIII. Discussion--Updates to Construction Permit Requirements and
Related Licenses
XIX. Background--Alternative Risk-Informed and Performance-Based
Acceptance Criteria for 10 CFR Parts 50 and 52
A. Need for Regulatory Flexibility
B. Enabling Risk-Informed and Performance-Based Alternatives
XX. Discussion--Alternative Risk-Informed and Performance-Based
Acceptance Criteria for 10 CFR Parts 50 and 52
XXI. Background--Establishing Thresholds for Changes to Reactor
Designs During Construction and Operation Under 10CFR Parts 52 and
53
A. Development of Tiers of Information and Processes for Changes
and Departures in Design Certification Rules
B. Licensing Experience and Improvement Initiatives Regarding
Information Designation and Change Processes for Design
Certifications
C. Severe Accidents
XXII. Discussion--Establishing Thresholds for Changes to Reactor
Designs During Construction and Operation Under 10 CFR Parts 52 and
53
A. Standardization
B. Definitions
C. Processes for Changes and Departures
XXIII. Background--Revision of the Emergency Preparedness
Regulations for Nuclear Power Reactors
A. Existing Emergency Preparedness Frameworks for Nuclear Power
Reactors
B. Protective Actions and Emergency Planning Zones
XXIV. Discussion--Revision of the Emergency Preparedness Regulations
for Nuclear Power Reactors
A. Emergency Plan Licensing Flexibility
B. Emergency Planning Zone Certainty
C. Eliminating Redundant Requirements
D. Risk-Informing the Emergency Plan Change Process
XXV. Background--Optional Submittal of Operational Programs
XXVI. Discussion--Optional Submittal of Operational Programs
XXVII. Background--Early Site Permit for Nuclear Power Plants
XXVIII. Discussion--Early Site Permit for Nuclear Power Plants
XXIX. Background--Manufacturing License Term Extension
XXX. Discussion--Manufacturing License Term Extension
XXXI. Background--Nuclear Power Plant License Renewal
XXXII. Discussion--Nuclear Power Plant License Renewal
A. Extension of the Renewal Time Period
B. Alternative Risk-Informed and Performance-Based Criteria
C. Eliminate the Limitation on Early Application Submittal
D. Eliminate the Required Application Content on Exemptions
E. Eliminate the Required Application Content on Technical
Specifications
F. Reduce Ongoing, Post-Renewal Updates to the Final Safety
Analysis Report
XXXIII. Background--Enhancing Flexibility of Reactor Site Criteria
XXXIV. Discussion--Enhancing Flexibility of Reactor Site Criteria
XXXV. Background--Increased Enrichment of Conventional and Accident
Tolerant Fuel Designs for Light-Water Reactors
A. Accident Tolerant Fuels
B. Rulemaking Development
C. Background and History of Affected Regulations
D. Regulatory Basis
XXXVI. Discussion--Increased Enrichment of Conventional and Accident
Tolerant Fuel Designs for Light-Water Reactors
A. Criticality Accident Requirements in 10 CFR 50.68
B. Uranium Fuel Cycle Environmental Data--Table S-3 in 10 CFR
51.51
C. Environmental Effects of Transportation of Fuel and Waste--
Table S-4 in 10 CFR 51.52
D. Fissile Material Packaging Requirements in 10 CFR 71.55
E. Control Room Requirements in 10 CFR 50.67 and GDC 19
F. Fuel Dispersal
XXXVII. Specific Questions
XXXVIII. Regulatory Flexibility Certification
XXXIX. Regulatory Analysis
XL. Backfitting and Issue Finality
XLI. Cumulative Effects of Regulation
XLII. Plain Writing
XLIII. National Environmental Policy Act
XLIV. Paperwork Reduction Act
XLV. 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 14267: Reducing Anti-Competitive Regulatory
Barriers
XLVI. Voluntary Consensus Standards
XLVII. Availability of Guidance
XLVIII. Availability of Documents
I. Obtaining Information and Submitting Comments
A. Obtaining Information
Please refer to Docket ID NRC-2025-0975 when contacting the NRC
about the availability of information for this action. You may obtain
publicly available information related to this action by any of the
following methods:
Federal Rulemaking Website: Go to https://www.regulations.gov and search for Docket ID NRC-2025-0975.
[[Page 44564]]
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 Public 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 of this document.
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.
Technical Library: The Technical Library, which is located
at Two White Flint North, 11545 Rockville Pike, Rockville, Maryland
20852, is open by appointment only. Interested parties may make
appointments to examine documents by contacting the NRC Technical
Library by email at [email protected] between 8 a.m. and 4 p.m.
Eastern Time, Monday through Friday, except Federal holidays.
Public Meeting: The NRC will conduct public meetings to
describe the proposed amendments and answer questions from the public
on the proposed rule. The NRC will publish a notice of the location,
time, and agenda of the meetings on the NRC's public meeting website
within 10 calendar days of the meetings. Stakeholders should monitor
the NRC's public meeting website for information about the public
meetings at: https://www.nrc.gov/public-involve/public-meetings/index.cfm.
B. Submitting Comments
Comments must be submitted using https://www.regulations.gov by
11:59 p.m. Eastern Time on August 31, 2026. Please include Docket ID
NRC-2025-0975 in your comment submission.
The NRC cautions you not to include identifying or contact
information that you do not want to be publicly disclosed in your
comment submission. The NRC will post all comment submissions at
https://www.regulations.gov as well as enter the comment submissions
into ADAMS. The NRC does not routinely edit comment submissions to
remove identifying or contact information.
If you are requesting or aggregating comments from other persons
for submission to the NRC, then you should inform those persons not to
include identifying or contact information that they do not want to be
publicly disclosed in their comment submission. Your request should
state that the NRC does not routinely edit comment submissions to
remove such information before making the comment submissions available
to the public or entering the comment into ADAMS.
II. Executive Order 14300: Ordering the Reform of the Nuclear
Regulatory Commission
On May 23, 2025, President Donald J. Trump signed Executive Order
(E.O.) 14300, ``Ordering the Reform of the Nuclear Regulatory
Commission.'' Section 5, ``Reforming and Modernizing the NRC's
Regulations,'' requires the Nuclear Regulatory Commission (NRC) to
undertake a review and wholesale revision of its regulations and
guidance documents as guided by the policies set forth in section 2 of
the E.O. This rulemaking addresses section 5(f), which directs the NRC
to ``[e]stablish stringent thresholds for circumstances in which the
NRC may demand changes to reactor design once construction is
underway''; section 5(h), which directs the NRC to ``[a]dopt revised
and, where feasible, determinate and data-backed thresholds to ensure
that reactor safety assessments focus on credible, realistic risks'';
and section 5(i), which directs the NRC to ``[r]econsider the
regulations governing the time period for which a renewed license
remains effective, and extend that period as appropriate based on
available technological and safety data.'' Additionally, as part of the
NRC's overarching review of all of its regulations, the agency
identified a number of additional changes to the NRC's regulations that
will improve the efficiency and efficacy of its licensing process that
are also included in this rulemaking. In developing the proposed
changes, the NRC has considered the benefits of increased availability
of, and innovation in, nuclear power to our economic and national
security consistent with section 501(a) of the Accelerating Deployment
of Versatile, Advanced Nuclear for Clean Energy Act of 2024 (ADVANCE
Act) and section 3 of E.O. 14300.
III. Background--Expedited Construction of Certain Structures, Systems,
and Components
A. Definition of Construction
Section 185 of the Atomic Energy Act of 1954, as amended (AEA),
requires that the NRC grant construction permits (CPs) to applicants
for licenses to construct or modify production or utilization
facilities, if the applications for such permits are acceptable to the
NRC. However, the term ``construction'' is not defined anywhere in the
AEA. Instead, construction is defined within section 50.10, ``License
required; limited work authorization'' of part 50, ``Domestic Licensing
of Production and Utilization Facilities,'' in title 10 of the Code of
Federal Regulations (10 CFR). The Commission last updated this
definition in 2007 as part of the limited work authorization (LWA)
final rule, ``Limited Work Authorizations for Nuclear Power Plants''
(72 FR 57416; October 9, 2007).
In developing this definition of construction in the 2007 LWA final
rule, the Commission concluded that the definition of construction
should parallel the agency's jurisdiction because activities outside
the definition of construction would not require prior authorization.
Therefore, the Commission ``determined that construction should include
all of the activities that have a reasonable nexus to radiological
health and safety, or common defense and security'' (72 FR 57429;
October 9, 2007). For the 2007 LWA final rule, the scope of structures,
systems, and components (SSCs) falling within the definition of
construction was derived from the scope of SSCs that are included in
the program for monitoring the effectiveness of maintenance at nuclear
power plants, as defined in 10 CFR 50.65(b), because ``the definition
is well understood and there is good agreement on its implementation''
(72 FR 57429-30; October 9, 2007). The SSCs were supplemented with
those necessary to comply with emergency preparedness and security
regulations because they also have a reasonable nexus to radiological
safety or are required for the common defense and security.
New reactor designs and deployment strategies, as well as lessons
learned from previous examples, warrant a fresh look at how application
of the 10 CFR 50.10 definition of construction may unnecessarily
restrict construction of nuclear power plants using modern construction
techniques on optimized schedules. Additionally, some prospective
applicants to construct a nuclear power reactor have indicated
[[Page 44565]]
the desire to construct portions of the facility, such as permanent
materials left in an excavation, or even balance of plant systems, in
advance of a CP or combined license (COL), and without an LWA, based on
their business needs and the lack of the SSCs' safety significance. The
Commission recognizes that the activities undertaken to build or
install some facility SSCs will not affect the safety-significant
functions of those SSCs, even though the 2007 rule could have been
understood to require a license to build or install them, and the NRC
need not authorize such activities before they occur. Moreover, the NRC
has considered and granted exemptions from the current definition of
construction to allow an applicant to build SSCs over which the NRC
exercises authority but for which construction is not material to the
safety function of the SSC.
B. General Licenses
In regard to general licenses for utilization facilities, section
109 of the AEA, ``Component and Other Parts of Facilities,'' authorizes
the Commission to ``issue general licenses for domestic activities
required to be licensed under section 101 [of the AEA]'' with respect
to those utilization facilities determined by the Commission under
section 11cc.(2) of the AEA. Section 11cc.(2) of the AEA defines such
facilities as ``any important component part especially designed for [a
utilization facility as defined under sec. 11cc.(1) of the AEA] as
determined by the Commission.'' Under section 109 of the AEA, the
Commission may issue a general license authorizing construction of such
``important component parts'' if it ``determines in writing that such
general licensing will not constitute an unreasonable risk to the
common defense and security.''
In order to accommodate the business models for new reactor designs
some prospective vendors are proposing and to enable rapid deployment
strategies for advanced reactor technologies, the NRC plans to revise
its regulations to allow the use of general licenses for construction
of important component parts of a utilization facility.
IV. Discussion--Expedited Construction of Certain Structures, Systems,
and Components
A. Definition of Construction
The NRC is proposing to revise its regulations to update the
definition of ``construction'' in 10 CFR 50.10, 10 CFR 51.4,
``Definitions,'' and 10 CFR 53.020, ``Definitions,'' to facilitate the
safe construction of nuclear power plants using modern construction
techniques on optimized schedules.
Specifically, this proposed rule would revise the definition of
construction to include SSCs for which construction can affect the
SSC's capability to perform a safety-related or safety-significant
function and will, therefore, require NRC approval before commencing
construction. An additional purpose of this change is to afford license
applicants, when justified, additional flexibility to build or install
SSCs whose safety-related or safety-significant functions are not
significantly affected by those activities at a site prior to the
issuance of a license. Those SSCs that are constructed without NRC
authorization may still be subject to additional operational
requirements as part of any subsequent operating license (OL) that
would be issued.
B. Safety Review
Since the issuance of the 2007 LWA final rule, the NRC has observed
that prospective advanced reactor applicants have designed their
facilities with separation between nuclear and balance of plant SSCs in
mind, such that many of the criteria in the current definition for a
construction activity are not met for certain SSCs. Some stakeholders
have maintained that such SSCs do not have a reasonable nexus to safety
and therefore the unmet criteria are not necessary to provide
reasonable assurance of adequate protection to the health and safety of
the public; therefore, those unmet criteria are not needed or do not
serve the underlying purpose of the rule. Lacking a revision to the
definition, the remaining unmet criteria prevent prospective applicants
from constructing such components without first obtaining an LWA, CP, a
COL, or an exemption.
The proposed construction definition would be limited to those SSCs
for which construction activities may have a significant impact on
radiological health and safety. For other SSCs, even those that may
have a nexus to radiological health and safety during operation,
operational requirements should suffice, and the NRC need not license
the construction of those SSCs.
The flexibility afforded by this proposed change would rely on an
applicant-performed analysis and categorization of the SSCs of the
facility to those that do and those that do not meet the definition of
construction.
The set of SSCs that would meet the definition of construction
should include only those SSCs that perform safety-related functions or
that perform safety-significant functions and the successful completion
of those functions may be impacted by construction. For those SSCs,
prior NRC approval for construction would be required because
inadequate design or construction of those SSCs could have a
substantial contribution to radiological risk during operation.
Activities undertaken to build onsite emergency facilities
necessary to comply with either 10 CFR 50.160, ``Emergency preparedness
for small modular reactors, non-light-water reactors, and non-power
production or utilization facilities,'' or 10 CFR 50.47, ``Emergency
plans,'' and appendix E to 10 CFR part 50, ``Emergency Planning and
Preparedness for Production and Utilization Facilities,'' or 10 CFR
53.855, ``Emergency preparedness,'' as applicable, would not be
considered SSCs that meet the definition of construction. Historically,
emergency response facilities (ERF) were included due to their
reasonable nexus to radiological health and safety, but they would not
fall under the criteria in proposed 10 CFR 50.10(a)(1)(i) through (iii)
or 10 CFR 53.020. Instead, applicants would comply with the
requirements of 10 CFR 50.160, or 10 CFR 50.47 and appendix E to 10 CFR
part 50, or 10 CFR 53.855, as applicable. The applicant would need to
be aware of all functional requirements of 10 CFR 50.160, or 10 CFR
50.47 and appendix E to 10 CFR part 50, or 10 CFR 53.855 for ERFs.
These functional requirements would need to be validated in a
preoperational exercise, which would satisfy the historical reasons why
ERFs were previously included in the construction definition.
For license applicants under 10 CFR parts 50 and 52, ``Licenses,
Certifications, and Approvals for Nuclear Power Plants,'' the proposed
revisions to the definition of construction would also support use of
the SSC categorization methodology for designs licensed under the
technology-inclusive, risk-informed, and performance-based methodology
described in regulatory guide (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,'' dated
June 2020. Under this framework, the applicant would use its
probabilistic risk assessment (PRA) of the design to analyze the
function of the SSCs. As explained in RG 1.233, the applicant's
analysis would result in the classification of SSCs into one of four
categories: ``safety-related,'' ``non-safety-
[[Page 44566]]
related with special treatment,'' ``non-safety-related with no special
treatment,'' and ``all other SSCs'' (with no special treatment
required). Within the RG 1.233 methodology, ``safety-significant'' SSCs
include all those SSCs classified as ``safety-related'' or ``non-
safety-related with special treatment.'' Those SSCs whose construction
could impact their safety-significant function, and thus would fall
under the definition of construction in the amended 10 CFR
50.10(a)(1)(i) and (ii), would likely include only those SSCs that were
analyzed and classified as safety-related or non-safety-related with
special treatment. It is possible that a subset of some non-safety-
related with special treatment SSCs could have no safety functions
which would be impacted by the construction of the SSC. For these
cases, an applicant could provide further information that justifies
the exclusion of the SSC from the definition of construction. All SSCs
classified as either ``non-safety-related with no special treatment''
or ``all other SSCs,'' would likely not be included in the definition
of construction under those criteria because, as demonstrated in the
PRA, their impact on safety as evaluated under the methodology would be
negligible. Consequently, the proposed amendments to 10 CFR 50.10 would
enable applicants referencing RG 1.233 greater flexibility to undertake
preconstruction activities.
Applicants under 10 CFR part 50 or 52, when determining whether
SSCs fall under the criteria in proposed 10 CFR 50.10(a)(1)(i) through
(iii), would confirm that other considerations do not require that the
resulting list of SSCs should otherwise be subject to NRC quality
assurance (QA) requirements for design in appendix B to 10 CFR part 50,
``Quality Assurance Criteria for Nuclear Power Plants and Fuel
Reprocessing Plants.'' Criterion III of appendix B to 10 CFR part 50
sets forth requirements for design control, and appendix B to 10 CFR
part 50 includes other requirements (e.g., for records and audits) that
apply to the design of SSCs subject to appendix B to 10 CFR part 50.
In addition, applicants under 10 CFR parts 50 and 52 would also
confirm that other considerations would not require that any other SSC
should be subject to general design criterion (GDC) 1, ``Quality
standards and records,'' in appendix A to 10 CFR part 50. Criterion 1
of appendix A to 10 CFR part 50 also imposes corresponding requirements
for SSCs important to safety but not safety-related to the extent such
requirements are commensurate with an SSC's importance to safety.
Safety-related SSCs are subject to all requirements in appendix B to 10
CFR part 50, including QA requirements applicable to facility
operation.
Application of the QA requirements for design in proposed appendix
T, ``Streamlined Quality Assurance Criteria for Nuclear Power Plants
and Fuel Reprocessing Plants,'' to 10 CFR part 50 (see section XVI of
this document), could also affect the definition of construction for
applicants under 10 CFR part 50, 52, or 53, ``Risk-Informed,
Technology-Inclusive Regulatory Framework for Commercial Nuclear
Plants.'' Specifically, the definition of construction proposed in this
rule would include SSCs subject to the requirements of proposed
appendix T to 10 CFR part 50, and, in certain instances, the structures
built for those SSCs.
Applicants under 10 CFR parts 50 and 52 must comply with criterion
1 of appendix A to 10 CFR part 50. This criterion requires that ``SSCs
important to safety be designed, fabricated, erected, and tested'' to
QA standards commensurate with the importance of the safety functions
to be performed. However, an SSC could be subject to performance
requirements for operation but would not warrant the application of QA
measures under GDC 1 such as QA for design. Specifically, the SSC could
be commercial grade but also subject to specified operational
performance requirements. Such an SSC could be constructed without a
license. The rationale for the approach rests on the fact that if no
NRC QA requirement for design, fabrication, erection, and testing
applies to an SSC, then there is nothing uniquely related to nuclear
safety for the NRC to approve with respect to construction of the SSC.
Further, design limits on the operation of the SSCs that do perform
safety functions would prevent or mitigate the safety effects of the
failure of SSCs not subject to NRC QA requirements applicable to
construction. Accordingly, construction of such SSCs would not have a
reasonable nexus to nuclear safety, even if operation of the SSC did
have a reasonable nexus to nuclear safety.
The failure of an SSC that has an effect on the safety of operation
could warrant operational requirements with respect to SSCs that
perform safety functions in response to such failures and possibly
operational requirements with respect to the SSCs themselves.
An applicant's designation of an SSC as something that does not
meet the definition of construction would not restrict the NRC from
imposing such operational requirements to address radiological health
and safety.
The NRC encourages pre-application engagement when licensees plan
to undertake significant preconstruction activities to assist the NRC
staff to further understand which SSCs have a reasonable nexus to
radiological health and safety at the construction phase. An applicant
may communicate its plan to comply with the regulations, including an
SSC classification methodology to the NRC as part of preapplication
interactions. The NRC staff will provide feedback as appropriate.
Alternatively, one or more prospective license applicants could propose
a generic SSC classification methodology, which the NRC staff could
review and endorse as acceptable guidance outside a particular
licensing action. The approved methodology could then be referenced by
multiple applicants.
Similar to the existing regulations, if an applicant under the
proposed rule determines that an SSC falls within the scope of the
definition of construction in the proposed 10 CFR 50.10 or 53.020, an
exemption request or an LWA would need to be submitted to the NRC to
allow for this construction activity to occur prior to issuance of a CP
or COL. Similar to the existing processes, the LWA would be granted if
the underlying requirements of 10 CFR 50.10(d) and (e) or 10 CFR
53.1130, ``Limited work authorizations, general licenses,'' as
applicable, are met.
C. Environmental Review
As explained in section IV.A., ``Definition of Construction,'' of
this document, a CP, COL, or LWA applicant does not need to obtain an
NRC license to build SSCs that do not meet the definition of
construction in 10 CFR 51.4. As long as there is no other Federal
action authorizing these activities, under 10 CFR 51.20 through 51.22,
an environmental review under the National Environmental Policy Act
(NEPA) is not required. Activities undertaken to build SSCs excluded
from the definition of construction are not part of an NRC licensing
action because such construction activities do not have a reasonable
nexus to nuclear safety, even if operation of a particular SSC does
have a reasonable nexus to nuclear safety. An applicant's
classification of an SSC as not safety-significant, that is built prior
to issuance of a license, would not restrict the NRC from imposing
operational requirements on those SSCs through a later action.
For SSCs that do meet the definition of construction where the NRC
would authorize construction, the NRC must perform an environmental
review in accordance with 10 CFR 50.10 or 53.610, ``Construction,'' as
applicable,
[[Page 44567]]
and the application requirements of 10 CFR part 51, ``Environmental
Protection Regulations for Domestic Licensing and Related Regulatory
Functions.''
D. General Licenses and Generic Finality
The NRC proposes to add new 10 CFR 50.10(h) and 53.1130(e), which
would implement the authority in section 109 of the AEA to create a
general license allowing construction of an important component part of
a specified class of commercial nuclear plants. The specified class of
plants would be those plants of a design the NRC previously approved in
a licensing action in which the NRC also granted ``generic finality,''
as discussed later, and for which operation has been authorized. The
important component part defined as a utilization facility under
Section 11cc.(2) of the AEA and subject to the general license would be
the portion of the plant constructed on site except for the reactor
vessel, the reactor coolant system, and associated reactivity control
and heat removal systems. The general license would authorize
construction of the important component part upon docketing of an
application for a license that would authorize construction of the
nuclear plant, subject to conditions. The conditions would provide
reasonable assurance of adequate protection of the health and safety of
the public and common defense and security, and would also ensure an
appropriate level of environmental review. The NRC also proposes a
conforming change to 10 CFR 50.10(c) and 53.610(b) to provide that
construction may occur under the general licenses issued in 10 CFR
50.10(h) and 53.1130(e), respectively.
The construction activities authorized by the general license would
be limited to those SSCs for which a previously approved design was
provided generic finality, but would not include the reactor vessel,
the reactor coolant system, and associated reactivity control and heat
removal systems. The NRC expects that the construction of SSCs approved
through a general license would be for those SSCs that are not
inherently sensitive to the site-specific characteristics of a proposed
deployment site. Although construction of many SSCs in a particular
design could conceptually be approved through a general license, the
NRC cannot approve the construction of an entire utilization facility
through this provision.
The proposed new regulations would enable future applicants to
reference previously reviewed and approved information only when
significant safety and environmental issues related to design,
construction, and operation are generically resolved in a manner that
applies to the intended use of the information. For example, to qualify
for a general license, an applicant would have to reference a nuclear
reactor design that was afforded generic finality by the NRC and
successfully constructed under NRC oversight and placed into operation.
Also, the applicant's proposed site would have to fall within the
corresponding site parameter envelope that was provided in the request
for generic finality. Therefore, an application that satisfies the
proposed new regulation would provide reasonable assurance of adequate
protection of public health and safety and common defense and security
equivalent to satisfaction of existing regulations, and there would
have been a prior hearing opportunity on the reactor design being
referenced. Further, the proposed regulation would require the general
licensee to allow for NRC inspections that the Commission deems
necessary related to activities performed under the general license.
The general license regulation would also include conditions to
address environmental considerations. The OL or COL (as applicable) of
the plant for which generic finality was approved would either have met
the criteria for categorical exclusion or had a finding of no
significant impact after preparation of an environmental assessment.
Provided that the environmental characteristics of the proposed plant
fall within the environmental parameters for the plant for which
generic finality was approved, the environmental effects of a
subsequent plant would not exceed those of the approved plant and would
be acceptable. In addition, the applicant proposing to use the general
license would have to propose a plan for redress of any adverse
environmental impact from conduct of activities under the general
license should such redress be necessary. This proposed requirement
would be similar to the requirements in 10 CFR 50.10(d)(3)(iii), which
requires a redress plan as part of an application for an LWA, and 10
CFR 50.12(b)(2), which requires the Commission to consider redress of
adverse environmental impacts in determining whether to grant an
exemption permitting the conduct of construction activities prior to
the issuance of a CP.
The proposed general license regulation would also require that the
general licensee has notified the NRC that all applicable permits,
licenses, approvals, and other entitlements in connection with the
proposed action that the general licensee was responsible for obtaining
have been obtained. In addition, the proposed general license would
require that applicable Federal environmental consultations have been
completed. This would ensure that construction activities would not
begin unless the NRC has the information it would need to fulfill its
obligations for environmental review under the AEA, NEPA, and other
relevant laws.
In addition, the proposed general license regulation would clarify
that any activities undertaken by the general licensee or on its behalf
under the general license would be entirely at the risk of the general
licensee and would have no bearing on the issuance of a license with
respect to the requirements of the AEA, and rules, regulations, or
orders issued under the AEA. However, the general licensee would be
able to mitigate this additional regulatory risk through careful site
selection to ensure that site characteristics are within the bounds of
the postulated site parameters and by performing construction
activities following appropriate QA and fitness-for-duty programs.
Based on the proposed general license requirements in 10 CFR
50.10(h) and 53.1130(e), the Commission has determined that such
general licensing would be for only parts of utilization facilities,
not constitute an unreasonable risk to the common defense and security,
and, therefore, be consistent with the authority provided to the
Commission by section 109a. of the AEA.
In addition, in order to facilitate the use of the general
licensing concept, the NRC proposes to add conforming changes to the
following regulations.
The NRC proposes to add new 10 CFR 50.34(b)(14) which would require
an OL application for those 10 CFR part 50 applicants that request the
NRC to make a finding on generic finality, to include applicable site
parameters postulated for the design, including the design-basis
external hazard levels for the relevant external hazards, and an
analysis and evaluation of the design in terms of those site
parameters. Similarly, the NRC proposes to add new paragraph (bb) to 10
CFR 53.1369, ``Contents of applications for operating licenses;
technical information,'' for 10 CFR part 53, ``Risk-Informed,
Technology-Inclusive Regulatory Framework for Commercial Nuclear
Plants,'' OLs for the same purpose. For COL applications under 10 CFR
part 52 or 53, this application content requirement would be included
in a new 10 CFR 52.79(a)(48) and 53.1416(i), respectively. The site
parameters may be the same as or more severe than the site
[[Page 44568]]
characteristics established for the site of the reactor proposed in the
OL or COL application that proposed generic finality.
The NRC also proposes to add a new 10 CFR 50.58(b)(7), which would
require the Commission to include the request for generic finality as a
proposed action in the notice of proposed action required by 10 CFR
2.105 for OL applications. Similarly, the NRC also proposes to add a
new paragraph (b)(2) to 10 CFR 53.1375, ``Review of applications,''
which would require the Commission to include the request for generic
finality as a proposed action in the notice of proposed action for a 10
CFR part 53 OL, required by 10 CFR 2.105. For COL applications, the NRC
proposes to add a new 10 CFR 52.85(b) and 53.1422(b)(2), which would
require the Commission to include the request for generic finality as a
proposed action in the notice of hearing required by 10 CFR 2.104 for
COL applications under 10 CFR parts 52 and 53, respectively. These
changes would provide a hearing opportunity to the public on the
request for generic finality. In addition, the Commission's ruling on a
request for hearing or petition for leave to intervene under 10 CFR
2.309(d)(2) would consider that a petitioner may have an interest in
the application if matters resolved in the licensing proceeding were to
be afforded generic finality. This would enable petitioners whose
property, financial, or other interests would not be directly affected
by the issuance of the OL or COL for a particular reactor to have an
opportunity to intervene on generic aspects of the design that would be
afforded finality and would therefore not be subject to hearing if
referenced in a later application that would affect the petitioner's
property, financial, or other interest.
Consistent with the previous discussion, the NRC proposes to add
new 10 CFR 50.57(d), which would permit the Commission to afford
generic finality to generic aspects of the design of a utilization
facility licensed under 10 CFR part 50, including postulated site
parameters submitted pursuant to 10 CFR 50.34(b)(14), if it finds that
the proposed generic design can be constructed and operated at sites
having characteristics that fall within the site parameters postulated
for the design. For the same reason and with the same conditions, the
NRC proposes to add new paragraph (e) to 10 CFR 53.1387, ``Issuance of
operating licenses,'' which would permit the Commission to afford
generic finality to generic aspects of the design of a commercial
nuclear plant licensed under 10 CFR part 53, including postulated site
parameters submitted pursuant to 10 CFR 53.1369(bb).
Similarly, the NRC proposes to add new 10 CFR 52.97(d) which would
permit the Commission to afford generic finality to generic aspects of
the design of a utilization facility licensed under 10 CFR part 52,
including postulated site parameters submitted pursuant to 10 CFR
52.79(a)(48), if it finds that the proposed generic design can be
constructed and operated at sites having characteristics that fall
within the site parameters postulated for the design. For the same
reason and with the same conditions, the NRC proposes to add new
paragraph (d) to 10 CFR 53.1440, ``Issuance of combined licenses,''
which would permit the Commission to afford generic finality to generic
aspects of the design of a commercial nuclear plant licensed under 10
CFR part 53, including postulated site parameters submitted pursuant to
10 CFR 53.1416(i).
The regulations in 10 CFR 50.59, ``Changes, tests and
experiments,'' that establish requirements for making changes to
portions of the facility as described in the final safety analysis
report (FSAR) for an OL are applicable to generic aspects of the design
of a utilization facility that have been afforded generic finality
because that design information would be included in the FSAR for the
OL. Similarly, the regulations in 10 CFR 50.59 are applicable to
generic aspects of the design of a utilization facility that are
described in a COL FSAR and have been afforded generic finality.
Similarly, the regulations in 10 CFR part 53 that establish
requirements for making changes to portions of the facility as
described in the FSAR are applicable to generic aspects of the design
of a utilization facility that have been afforded generic finality.
The NRC proposes to add new 10 CFR 50.58(b)(8) and 52.98(h) to
include requirements to address finality for portions of 10 CFR part 50
OLs and 10 CFR part 52 COLs with respect to NRC reviews and hearings.
Proposed 10 CFR 50.58(b)(8) would require the Commission to treat as
resolved any issues referenced in following proceedings or in
enforcement hearings (other than ones under 10 CFR 2.202(e)(1)) that
were afforded finality pursuant to 10 CFR 50.57(d). The proposed 10 CFR
50.58(b)(8) would ensure that issues resolved in an approved request
for generic finality (including, if applicable, the adequacy of a
reactor design) are not re-adjudicated in the license proceedings where
such information is referenced in the license applications. The
proposed 10 CFR 52.98(h) would include substantially the same
provisions for COLs with generic finality.
To address generic finality in 10 CFR part 53, the NRC proposes to
add similar provisions to new paragraph (b) to 10 CFR 53.1390,
``Finality of operating licenses,'' and new paragraph (g) to 10 CFR
53.1443, ``Finality of combined licenses.''
Proposed 10 CFR 53.1390(b) would require the Commission, in the
proceedings for issuance of a CP, an OL, or a COL or in any enforcement
hearing (other than one initiated under 10 CFR 53.1390(a)), to treat as
resolved those matters resolved in the proceedings on the application
or renewal of the referenced OL, including, if applicable, the adequacy
of a reactor design where the referenced OL was afforded finality
pursuant to 10 CFR 53.1387(e).
Proposed 10 CFR 53.1443(g) would require the Commission, in the
proceedings for issuance of a CP, an OL, or a COL or in any enforcement
hearing (other than one initiated under 10 CFR 53.1443(a)), to treat as
resolved those matters resolved in the proceedings on the application
or renewal of the referenced COL, including, if applicable, the
adequacy of a reactor design where the referenced COL was afforded
finality pursuant to 10 CFR 53.1440(d).
As written, the proposed generic finality provisions would allow
the NRC to take appropriate action under the applicable backfitting or
issue finality provision if the NRC determines that the generic
finality approval or associated technical information presents safety
concerns that warrant NRC action. As stated above, the generic finality
that would be afforded under the proposed rule provisions would not
apply to certain enforcement hearings. For example, proposed 10 CFR
50.58(d)(8) provides that finality would apply, in part, to ``any
enforcement hearing other than one initiated by the Commission under
Sec. 2.202(e)(1) of this chapter'' (emphasis added). Enforcement
hearings under 10 CFR 2.202(e)(1) are those which involve a backfit to
modify a 10 CFR part 50 license, and 10 CFR 50.109 must be followed for
such orders and the associated proceedings. Thus, generic finality
would not apply in such cases so that the NRC could take appropriate
action if the backfitting requirements in 10 CFR 50.109 are satisfied.
Similarly, the proposed generic finality provisions in 10 CFR 52.98(h),
53.1390(b), and 53.1443(g) would provide that generic finality applies
except in enforcement hearings initiated under the issue finality
[[Page 44569]]
provisions in 10 CFR 52.98(a), 53.1390(a), and 53.1443(a),
respectively.
V. Background--Determinate and Data-Backed Thresholds for Reactor
Safety Assessments
Section 5(h) of E.O. 14300 directs the NRC to ``[a]dopt revised
and, where feasible, determinate and data-backed thresholds to ensure
that reactor safety assessments are focused on credible, realistic
risks.''
The NRC evaluated ``reactor safety assessments,'' focusing on
assessments conducted to (1) demonstrate the capability of safety-
related SSCs during design basis events (DBEs) (e.g., as described in
Chapter 15, ``Transient and Accident Analysis,'' of NUREG-0800,
``Standard Review Plan for the Review of Safety Analysis Reports for
Nuclear Power Plants: LWR Edition''), and (2) verify the ability of
SSCs to withstand certain design basis conditions, including natural
phenomena and environmental conditions (e.g., high winds, seismic
events, and conditions during normal operation and accident scenarios).
To effectively address the direction in E.O. 14300, the NRC concluded
that the most appropriate approach would be to clarify the terminology
in 10 CFR 50.2, ``Definitions,'' to ensure that safety assessments are
focused on credible, realistic risks.
The selection of DBEs and associated design basis parameters is a
critical prerequisite for determining the safety of a nuclear facility.
DBEs serve to identify the subset of SSCs subject to more stringent QA
requirements and to establish the performance capabilities those SSCs
must demonstrate under normal operation, anticipated operational
events, and accident conditions. For example, 10 CFR 50.46,
``Acceptance criteria for emergency core cooling systems for light-
water nuclear power reactors,'' requires analysis of postulated loss-
of-coolant accidents to verify the adequacy of emergency core cooling
system (ECCS) designs. In addition to the general requirement to
analyze SSC performance during DBEs, other regulations specify
additional, event-specific accident analyses, often referred to as
beyond design basis events (BDBEs). For example, 10 CFR 50.63, ``Loss
of all alternating current power,'' provides requirements related to
plants' abilities to withstand for a specified duration and recover
from a station blackout (SBO).
The technical information required in applications for CPs and OLs,
including the content of preliminary and FSARs, is outlined in 10 CFR
50.34, ``Contents of applications; technical information.'' Among other
requirements, an applicant is required to evaluate siting
considerations and the design and performance of SSCs that are intended
to prevent accidents and mitigate their consequences. The regulatory
processes described in 10 CFR part 52 include similar technical
information requirements for the content of applications (e.g., as
specified in 10 CFR 52.17, ``Contents of applications; technical
information,'' 52.47, ``Contents of applications; technical
information,'' 52.79, ``Contents of applications; technical information
in final safety analysis report,'' 52.137, ``Contents of applications;
technical information,'' and 52.157, ``Contents of applications;
technical information in final safety analysis report'').
The technical information associated with the performance of safety
assessments is documented in the preliminary or FSAR for CPs or OLs and
COLs, respectively. The FSAR describes the evaluation methods used to
establish design bases and perform safety analyses, the design and
performance requirements for SSCs, and the methods used to demonstrate
that those SSCs can perform their intended safety functions.
Accordingly, the FSAR serves as an essential component of the licensing
basis for a nuclear facility. It is also used to determine the
appropriate regulatory process for licensing basis changes, such as
those governed by 10 CFR 50.59, ``Changes, tests and experiments,'' or
10 CFR 50.90, ``Application for amendment of license, construction
permit, or early site permit.''
Within the power reactor licensing framework, the term ``safety-
related'' is used to identify SSCs that require special treatment,
including QA controls, environmental qualification, and compliance with
applicable industry codes and standards. The current definition of
``safety-related SSCs,'' provided in 10 CFR 50.2, uses the term
``design basis events'' to define the scope of safety assessments
needed to identify safety-related SSCs. However, 10 CFR 50.2 does not
include a corresponding definition of ``design basis event'' or provide
criteria for selecting events to be considered in the design basis.
While 10 CFR 50.49, ``Environmental qualification of electric equipment
important to safety for nuclear power plants,'' includes a definition
of DBEs, that definition does not explicitly apply to the definition of
safety-related SSCs in 10 CFR 50.2 and does not reference the use of
determinate, data-backed thresholds.
Traditionally, the spectrum of DBEs used to identify safety-related
SSCs, as defined in 10 CFR 50.2, has been based on information
contained in Chapter 15 of NUREG-0800. While this approach had been
effective for licensing large light-water reactors (LWRs) with designs
similar to the currently operating power reactor fleet, the DBEs
described in NUREG-0800 can have limited applicability to evolutionary
LWR designs and non-LWR designs. For example, the lack of a more
technology-inclusive definition for the term DBE in the current
definition has created challenges with respect to clarity and
reliability on the subset of SSCs that warrant special treatment.
Furthermore, it has created the potential to require safety assessments
that may not be focused on credible, realistic risks. Therefore,
clarifying what constitutes a DBE based on determinate, data-backed
thresholds would enhance the efficiency and consistency of future power
reactor licensing reviews. Consequently, the NRC proposes to provide a
definition for DBE in 10 CFR part 50.
Nonetheless, the NRC has not identified a need to propose a
corresponding revision to the term ``design bases'' in 10 CFR 50.2. The
term ``design bases'' is defined in 10 CFR 50.2 as information which
identifies the specific functions to be performed by a structure,
system, or component of a facility, and the specific values or ranges
of values chosen for controlling parameters as reference bounds for
design. The definition further clarifies that design basis values may
be (1) constraints derived from generally accepted ``state-of-the-art''
practices for achieving functional goals or (2) requirements based on
analyses of the effects of postulated accidents for which an SSC must
meet specified functional goals. Design bases are connected to safety
assessments in two ways: (1) the performance capabilities of SSCs, as
established through evaluations of DBEs and BDBEs, and (2) the design
parameters for SSCs, which are derived from the operational context in
which the function is to be performed, including considerations of
natural phenomena and environmental factors. The NRC has determined
that the existing definition of design bases in 10 CFR 50.2 provides
sufficient flexibility to support the use of determinate, data-backed
thresholds. In practice, the NRC has already applied determinate and
data-backed thresholds for the selection of design bases attributes in
several areas, including high winds, flooding, and seismic hazards.
Therefore, a rulemaking to revise the definition of design bases is not
necessary. However, the NRC is issuing draft guidance (DG)
[[Page 44570]]
that will provide additional detail within DG-1454, ``Implementation of
Determinate and Data-Backed Thresholds for Reactor Safety
Assessments,'' to support consistent application in this area.
The term BDBE has not previously been defined in 10 CFR 50.2.
However, lessons learned from ongoing studies of nuclear plant risks,
as well as operational experience, have historically led the NRC to
identify and address plant events and conditions beyond the originally
defined set of DBEs that could result in the release of radioactive
material sufficient to pose a hazard to public health and safety.
Accordingly, the NRC has imposed additional requirements to address
such events when risk insights emerged from operational experience
(e.g., SBO in 10 CFR 50.63 and anticipated transients without scram
(ATWS) in 10 CFR 50.62, ``Requirements for reduction of risk from
anticipated transients without scram (ATWS) events for light-water-
cooled nuclear power plants''). These requirements extended regulatory
attention beyond the traditional scope of DBEs. When developing this
proposed rulemaking, the NRC initially considered a framework where
these types of events were included in the DBE category. However,
experience with regulating events such as SBO and ATWS has demonstrated
that these types of events can be adequately addressed without the same
regulatory treatment as DBEs. Therefore, the formal inclusion of the
BDBE category in this proposed rule would provide a framework for
applying graded regulatory treatment to such events. It would enable
the NRC to address risks to public health and safety that do not
warrant mitigation exclusively through safety-related SSCs or
conservative safety assessments. Consequently, the NRC proposes to
provide a definition for BDBE in 10 CFR part 50.
VI. Discussion--Determinate and Data-Backed Thresholds for Reactor
Safety Assessments
The proposed changes would revise 10 CFR 50.2 to add definitions
for the terms ``design basis events'' and ``beyond design basis
events.'' The proposed changes would include a conforming revision to
the definition of ``design basis event'' in 10 CFR 50.49(b)(1)(ii).
These changes would apply to future 10 CFR part 50 and 52 applications
submitted on or after the date that would be 180 days after the
effective date of a final rule if this proposed rule were issued as a
final rule; however, existing applicants, licensees, and approval
holders under 10 CFR part 50 or 52 could voluntarily choose to adopt
them. In parallel with the proposed changes, the NRC has developed DG-
1454, which would (1) describe determinate and data-backed thresholds
for categorizing events as DBEs or BDBEs, (2) outline graded assessment
approaches for each event category, and (3) clarify the process for
selecting design bases controlling parameters used as reference bounds
in the design of SSCs. This guidance would use initiating event
frequencies and qualitative criteria for categorizing events,
maintaining consistency with the current safety assessment framework
described in 10 CFR part 50 (and referenced in 10 CFR part 52). In
addition, this guidance would describe acceptable approaches for
identifying, grouping, and quantifying initiating events to ensure they
are binned into appropriate categories. While alternate approaches,
such as defining thresholds in terms of event sequences could be used,
they would typically require the development of a full risk assessment
or other systematic risk evaluation to determine sequence frequencies.
To avoid imposing additional requirements not currently included in 10
CFR part 50, anchoring event selection to initiating event frequencies
would provide a determinate and data-backed approach without adding
regulatory burden.
Adding generally applicable definitions for DBE and BDBE would
improve regulatory clarity and enable the use of objective criteria in
selecting initiating events. Establishing threshold criteria, graded
assessment approaches, and the selection process for design basis
parameters within guidance would ensure that applicants and licensees
are provided with an approach acceptable to the NRC, while ensuring
flexibility for applicants to justify unique approaches, if desired,
without the need for an exemption.
The addition of a definition of BDBE and corresponding thresholds
would allow a reduction in unnecessary conservatism applied in the
safety assessments of lower frequency events. The NRC concluded that
reactor safety assessments associated with BDBEs are within the scope
of the existing contents of application requirements of 10 CFR 50.34
and analogous sections of 10 CFR part 52. Specifically, requirements
related to analysis and evaluation of the design and performance of
SSCs of the facility with the objective of assessing the risk to public
health and safety include consideration of BDBEs. Several existing
regulations already address specific events not originally considered
in the licensing basis or considered BDBEs (examples include but are
not limited to ATWS, loss of all alternating current power events
resulting in SBOs, and combustible gas control). For current
applicants, licensees, or approval holders who may opt to adopt the
proposed definitions for DBE and BDBE, as well as future applicants who
would be mandated to use the proposed definitions, this rulemaking
would not change the treatment of BDBEs specifically addressed by
regulation such as ATWS and SBO. However, adoption of the BDBE
definition could eliminate some events not specifically addressed by
regulation from consideration that are determined to be non-credible.
Similarly, evaluations that assume substantial release of fission
products would still be performed in accordance with 10 CFR 50.34(a)
and 10 CFR 50.67, ``Accident source term.'' For LWRs, the release would
be into containment. For other designs, it may be expressed as releases
to the environment considering expected demonstrable leakage rates from
potential flow paths and any fission product cleanup systems intended
to mitigate the consequences of accidents. These evaluations would
address the safety features that are engineered into a facility and
those barriers that must be breached as a result of an accident before
a radiological release to the environment can occur. Evaluations
required to comply with 10 CFR 50.34(a) and 10 CFR 50.67 rely on
conservative modeling assumptions. For example, as described in RG
1.183, ``Alternative Radiological Source Terms for Evaluating Design
Basis Accidents at Nuclear Power Reactors,'' source term fission
product release fractions are derived from a set of accident sequences
and many physical processes and phenomena are represented by bounding
assumptions rather than being modeled directly. In addition, these
evaluations credit only safety-related features in providing mitigation
capability. Therefore, this evaluation is generally included in the
spectrum of DBEs analyzed in Chapter 15 of NUREG-0800 (e.g., Sections
15.0.1 or 15.0.3).
The current definition of design bases in 10 CFR 50.2 provides that
controlling parameters may be derived either from accepted ``state-of-
the-art'' practices or from analysis (based on calculations or
experiments). The NRC has determined that this definition is
sufficiently broad to accommodate the use of determinate, data-backed
thresholds as implemented through guidance without the need for a
rulemaking change. Design bases are identified through two primary
means:
[[Page 44571]]
(1) the functional performance capabilities required to mitigate DBEs
and BDBEs; and (2) the design parameters derived from the operational
context in which the function is to be performed, including natural
hazards and environmental conditions. The first element is addressed
through the addition of proposed definitions of DBE and BDBE in 10 CFR
50.2 that would focus safety assessments on credible and realistic
risks. With regard to the second element, the NRC has gained
substantial experience in developing appropriate determinate and data-
backed thresholds to address protection against natural phenomena and
environmental conditions. Examples include assessment of tornado winds
(RG 1.76, ``Design-Basis Tornado and Tornado Missiles for Nuclear Power
Plants''), hurricane winds (RG 1.221, ``Design-Basis Hurricane and
Hurricane Missiles for Nuclear Power Plants''), and seismic hazards (RG
1.208, ``A Performance-Based Approach to Define the Site-Specific
Earthquake Ground Motion''). The NRC has developed DG-1454 to leverage
existing practices in this area and further clarify the selection of
design bases.
VII. Background--Removal of IEEE-323-1974 Reference in Footnote 3 of 10
CFR 50.49
Safety-related structures, systems and components are defined in 10
CFR 50.2, ``Definitions.'' The relationship between safety-related
electric equipment and Class 1E equipment was initially established
through footnote 3 of 10 CFR 50.49, ``Environmental Qualification of
Electric Equipment Important to Safety for Nuclear Power Plants.'' The
final rule promulgating 10 CFR 50.49, including footnote 3 of 10 CFR
50.49, (48 FR 2733; January 21, 1983) stated, in part: ``The scope of
the final rule covers that portion of equipment important to safety
commonly referred to as ``safety-related'' (which the Commission
interprets as essentially ``Class 1E'' equipment defined in [Institute
of Electrical and Electronics Engineers (IEEE)]-323-1974).''
The connection between ``safety-related'' and ``Class 1E'' is now
established in a more up-to-date standard--IEEE Standard 308, ``IEEE
Standard Criteria for Class 1E Power Systems for Nuclear Power
Generating Stations,'' which the NRC endorsed in RG 1.32, ``Criteria
for Power Systems for Nuclear Power Plants.''
VIII. Discussion--Removal of IEEE-323-1974 Reference in Footnote 3 of
10 CFR 50.49
The proposed action would remove footnote 3 of 10 CFR 50.49.
Footnote 3 references an old standard that is no longer utilized as the
sole means to establish the connection between ``safety-related'' and
``Class 1E.'' Instead, this connection is established in the more up-
to-date IEEE Standard 308, which the NRC endorsed in RG 1.32. Removal
of this footnote would improve regulatory clarity and would be
consistent with the NRC modernizing and improving its regulations to
reflect best practices and the maturity of the nuclear industry.
In addition, a minor editorial change is proposed to redesignate
the current footnote 4 of 10 CFR 50.49 as footnote 1 given the previous
and proposed deletions of the preceding footnotes.
IX. Background--Expanded Alternative Requests Under 10 CFR 50.55a(z)
In 10 CFR 50.55a, ``Codes and standards,'' the NRC incorporates by
reference certain parts of editions and addenda of specified codes and
standards through rulemaking. Upon incorporation by reference of these
specified codes and standards into 10 CFR 50.55a, the provisions of
these codes and standards are legally-binding NRC requirements as
delineated in 10 CFR 50.55a, subject to the conditions on certain
specific provisions that are set forth in 10 CFR 50.55a. Currently, in
paragraph (z), ``Alternatives to codes and standards requirements,'' of
10 CFR 50.55a, an applicant or licensee may request authorization of
alternatives to the requirements of paragraphs (b), ``Use and
conditions on the use of standards,'' through (h), ``Protection and
safety systems,'' of 10 CFR 50.55a, if the applicant or licensee
demonstrates either that the proposed alternative would provide an
acceptable level of quality and safety or that compliance with the
specified requirements would result in hardship or unusual difficulty
without a compensating increase in the level of quality and safety.
Since its initial promulgation in 1971 (36 FR 11423; June 12, 1971), 10
CFR 50.55a has allowed for the consideration of proposed alternatives
under these same two criteria. Over the years, 10 CFR 50.55a has been
periodically updated to reflect revised and updated codes and standards
for nuclear power plants.
On March 15, 1984 (49 FR 9711), the NRC issued a final rule that
made procedural changes by, among other things, clarifying the
procedures for alternatives, expressly noting that alternatives can be
authorized by the Director of the Office of Nuclear Reactor Regulation.
In the November 5, 2014, final rule, ``Approval of American Society of
Mechanical Engineers' [ASME] Code Cases'' (79 FR 65776), the NRC
restructured 10 CFR 50.55a to align with the Office of the Federal
Register's guidelines for incorporation by reference and to allow
proposed alternatives to NRC-approved Code Cases rather than only to
ASME Code provisions. In this restructuring, the proposed alternatives
provisions were moved from their prior location in paragraph (a)(3) of
10 CFR 50.55a to a newly designated paragraph (z) of 10 CFR 50.55a.
However, these rulemakings addressed only procedural clarifications and
a restructuring of existing regulations, not changes in the scope of
opportunities for alternatives. In the July 17, 2024, final rule,
``American Society of Mechanical Engineers Code Cases and Update
Frequency'' (89 FR 58039), the NRC added paragraph (y),
``Definitions,'' to 10 CFR 50.55a. These definitions provide
consistency and clarity throughout 10 CFR 50.55a and accommodate new
opportunities to change code of record intervals. However, these
definitions were added outside the scope of paragraph (z) of 10 CFR
50.55a because the Commission had not approved the use of 10 CFR
50.55a(z) for definitions or the newly defined intervals.
X. Discussion--Expanded Alternative Requests Under 10 CFR 50.55a(z)
Currently, the proposed alternative provisions of paragraph (z) of
10 CFR 50.55a apply to the codes and standards requirements in
paragraphs (b) through (h) of 10 CFR 50.55a. The addition of paragraph
(y) to 10 CFR 50.55a in the July 17, 2024, final rule, without an
associated expansion of the scope of paragraph (z), has resulted in
unanticipated exemptions under 10 CFR 50.12, ``Specific exemptions,''
to use alternate definitions to those included in paragraph (y) of 10
CFR 50.55a. Moreover, the criteria of paragraph (z)(1), ``Acceptable
level of quality and safety,'' or (z)(2), ``Hardship without a
compensating increase in quality and safety,'' of 10 CFR 50.55a provide
appropriate controls for requested alternatives to all requirements in
10 CFR 50.55a, so there is no need to restrict the application of 10
CFR 50.55a(z) to only some paragraphs in 10 CFR 50.55a. Therefore, the
NRC proposes to remove the restriction limiting proposed alternatives
to paragraphs (b) through (h) of 10 CFR 50.55a so that proposed
alternatives
[[Page 44572]]
would now be permitted for all regulatory requirements in 10 CFR 50.55a
using the existing criteria in paragraphs (z)(1) and (2) of 10 CFR
50.55a. The proposed removal of this scope restriction would also allow
for the use of the proposed alternatives provision in paragraph (z) of
10 CFR 50.55a for any future additions or modifications to 10 CFR
50.55a without the need for further revision to paragraph (z).
In addition to creating paragraph (y) of 10 CFR 50.55a, the
revisions in the July 17, 2024, final rule provided more flexibility to
licensees by expanding the code of record interval from 10 years to two
consecutive inservice testing and inservice inspection intervals. In
that final rule's preamble, the Commission stated, in part, that
licensees may request future alternatives based upon the code of record
interval. This revision, coupled with the staff position in SECY-23-
0061, ``Clarification of the Staff's Position on Certain American
Society of Mechanical Engineers Code Alternatives for More Than One 10-
Year Inservice Inspection Interval Under Title 10 of the Code of
Federal Regulations 50.55a,'' dated July 21, 2023, clarified that, when
appropriately justified, the duration of an alternative need not be
limited to the length of a single inservice testing or inservice
inspection interval. Rather, the NRC may approve specific alternatives
for longer durations when the technical bases supporting the requested
alternative ensure that an acceptable level of quality and safety will
be maintained. These existing flexibilities would be unchanged by this
proposed rule.
The proposed change would provide additional flexibility to a
licensee or applicant, while maintaining the same requirements for an
acceptable level of quality and safety or the presence of a hardship
without a compensating increase in quality and safety, which have been
foundational to proposed alternatives since their initial promulgation.
XI. Background--Risk-Informing 10 CFR 50.59 and Allowing Flexibility
for Changes to Methods
A. Use of Quantitative Risk Results
The AEA requires a licensee to seek an amendment for significant
changes to its facility or procedures. The Commission possesses
substantial discretion to define, by rule, the threshold that
constitutes a change significant enough to require a license amendment.
The current regulation in 10 CFR 50.59 was developed in response to
issues involving inconsistency in how licensees applied the previous
criteria to determine whether changes, tests, or experiments require
prior NRC approval. The NRC is proposing to amend 10 CFR 50.59 to allow
licensees to consider risk insights from PRAs when applying the
criteria in that provision. The statements of consideration for the 10
CFR 50.59 final rule, ``Changes, Tests, and Experiments'' (64 FR 53582;
October 4, 1999), did not allow licensees to use PRA insights at that
time, but the Commission recognized the possibility that the NRC could
one day develop the regulatory infrastructure to support the use of
PRAs in 10 CFR 50.59 analyses.
The NRC now proposes to incorporate the use of quantitative risk
results, like Core Damage Frequency (CDF) and Large Early Release
Frequency (LERF), to evaluate changes under 10 CFR 50.59(c)(2)(i) and
(ii).
The current state of practice in the nuclear fleet for quantitative
risk assessment is the use of Level 1/limited Level 2 PRAs. The
quantitative risk metrics output by these PRAs are CDF and LERF, and
they, along with their changes (i.e., [Delta]CDF and [Delta]LERF), are
the metrics used as part of risk-informed decision-making processes. As
described in RG 1.174, Revision 3, ``An Approach for Using
Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-
Specific Changes to the Licensing Basis,'' dated January 2018, these
risk metrics are based on the Commission's safety goals and the
associated quantitative health objectives. Similarly, the NRC proposes
to use the small changes in CDF and LERF jointly as means to determine
the importance of the effect of the proposed change on accident
frequency and SSC malfunction rate under 10 CFR 50.59.
B. Improved Flexibility for Changes to Methods of Evaluation
Currently, 10 CFR 50.59 allows licensees to make certain changes to
their facility or procedures without prior NRC approval, provided those
changes do not meet specific thresholds that would require a license
amendment. One of those thresholds, stated in 10 CFR 50.59(c)(2)(viii),
requires NRC review of any change in a methodology that results in a
departure from a method of evaluation described in the FSAR (as
updated) used in establishing the design bases or in the safety
analysis. As explained in the preamble of the 10 CFR 50.59 final rule
in 1999, this language was chosen to ensure NRC oversight of the safety
margins and conservatisms that form the basis of the NRC's licensing
decision. The Commission stated that the language of 10 CFR
50.59(c)(2)(viii) was selected ``to allow licensees only a small degree
of flexibility in methods where the results are tending in the non-
conservative direction'' (64 FR 53598; October 4, 1999).
In SECY-97-035, ``Proposed Regulatory Guidance Related to
Implementation of 10 CFR 50.59 (Changes, Tests, and Experiments),''
dated February 12, 1997, which transmitted proposed 10 CFR 50.59
guidance to the Commission ahead of the proposed rulemaking, the staff
recognized that, ``as the knowledge base increases and computing power
increases, new methods of analysis will more accurately predict the
actual plant response.'' However, the staff found that a comparison of
the analytical results from two different methodologies was not valid
to make a 10 CFR 50.59 determination. To make the 10 CFR 50.59
determination using a new methodology, the new methodology must be
valid (e.g., previously approved by the NRC) and the analysis in
question must be performed for the situation before the change and the
situation after the change using the same methodology.
In the decades since the original rule was written, there have been
substantial advancements in computational capabilities and modeling
practices. New data, improved understanding, and increased computing
power now allow for faster iteration and refinement of methods used in
safety analyses. Due to this substantial increase in computational
power and the rapid growth in the use of modeling and simulation, other
industries are beginning to shift away from model-by-model reviews and
are instead focusing on the processes by which organizations establish
the credibility of their models--specifically through verification,
validation, and uncertainty quantification (VVUQ) programs. The U.S.
Food and Drug Administration has taken the most prominent step in this
direction by issuing new guidance centered on credibility assessments
(FDA-2021-D-0980; November 17, 2023). The aviation industry is actively
developing a VVUQ standard through an industry-led, Federal Aviation
Administration-supported initiative.
The existing requirements of 10 CFR 50.59(c)(2)(viii) mandate a
license amendment for any departure from a method of evaluation
described in the FSAR. As defined in the regulation, a departure is a
change to any element of a methodology, unless the results are
[[Page 44573]]
``conservative or essentially the same,'' or any change from one method
to another method, unless the method has been previously approved by
the NRC for the specific application. Paragraph (a)(2)(iv) of 10 CFR
53.1550 limits changes to methods similarly to 10 CFR 50.59, allowing
changes without NRC review and approval only when results 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-approved
consensus code or standard. The rapid advancement in computational
modeling and simulation can offer more realistic and accurate safety
analyses than the legacy methods documented in many plants' original
FSARs. Thus, the current regulations can create an unnecessary
regulatory burden by requiring licensees to seek amendments to use
superior analytical tools, since these new tools would constitute a
departure from a method of evaluation described in the FSAR as
contemplated in the existing regulatory frameworks. This can
disincentivize the adoption of better technology and expend both
licensee and NRC resources on reviewing license amendment requests that
ultimately enhance, rather than degrade, safety analysis quality. The
purpose of the proposed rule change, therefore, would be to increase
regulatory efficiency and flexibility, consistent with the objectives
of the 1999 rule revision, the ADVANCE Act, and the Executive orders.
XII. Discussion--Risk-Informing 10 CFR 50.59 and Allowing Flexibility
for Changes to Methods
A. Use of Quantitative Risk Results
The NRC proposes rulemaking to establish an alternative pathway
that would allow the use of quantitative risk metrics, like CDF and
LERF, along with consideration of safety margins and defense in depth,
to evaluate a proposed change, test, or experiment against the criteria
of 10 CFR 50.59(c)(2)(i) and (ii), while leaving all other criteria in
10 CFR 50.59(c)(2) in place. New proposed 10 CFR 50.59(e) would
establish a risk-informed alternative to the existing regulation and
would not alter or impede the current practice for evaluating proposed
changes against the text of 10 CFR 50.59(c)(2)(i) and (ii), as written.
A licensee could continue to use qualitative assessments, engineering
judgement, and other existing practices and techniques to evaluate a
proposed change, test, or experiment.
Under the proposed 10 CFR 50.59(e), a licensee could demonstrate
that a change would not result in a ``more than a minimal increase''
under 10 CFR 50.59(c)(2)(i) and (ii) by using quantitative risk results
based on a PRA of appropriate scope and quality that provides
appropriate risk metrics. The change would also need to maintain
defense-in-depth and safety margins. ``Appropriate scope and quality''
in this context would mean that the licensee's model fully encompasses
the proposed change and that the model has been found to be acceptable
for use in a previous NRC-approved application. ``Appropriate risk
metrics'' in this context would mean quantitative results that
demonstrate the effects on the proposed change and can provide a
baseline for judging facility risk. For traditional PRAs, these metrics
are CDF, LERF, and the changes ([Delta]s) to CDF and LERF. Extensive
discussion of maintaining defense in depth and safety margins can be
found in RG 1.174, Revision 3. The NRC has proposed guidance for 10 CFR
50.59(e) in DG-1466, draft Revision 4 to RG 1.187, ``Guidance for
Implementation of 10CFR50.59, `Changes, Tests, And Experiments.'''
The use of PRA would not replace or supplant the deterministic
licensing basis but would supplement it with a powerful analytical
tool. Since the initial licensing of the current fleet, the NRC and the
industry have developed and matured PRA methodologies, which provide a
holistic, integrated assessment of plant safety. PRA can identify
contributors to risk and potential vulnerabilities that may not be
apparent from a purely deterministic analysis.
Incorporating CDF and LERF into the 10 CFR 50.59 process would not
be an attempt to re-license plants on a probabilistic basis. Instead,
it would provide an alternative methodology licensees could voluntarily
choose for conducting analyses under 10 CFR 50.59. It would use risk
insights to inform the judgment of the safety significance of changes
to the existing deterministic design. A change that results in a very
small, quantifiable increase in calculated risk could be reasonably
judged not to undermine the fundamental safety basis established
through deterministic principles. This approach would allow for a more
consistent, predictable, and efficient screening process, directly
fulfilling the original purpose of 10 CFR 50.59 to differentiate
between changes that require prior NRC review and those that do not.
This integration would be consistent with decades of evolving NRC
policy and practice. The agency has successfully used risk-informed
approaches in many other regulatory applications, including 10 CFR
50.65, ``Requirements for monitoring the effectiveness of maintenance
at nuclear power plants''; 10 CFR 50.48(c), ``National Fire Protection
Association Standard NFPA 805''; the Reactor Oversight Process; and RG
1.174. Using quantitative risk results, such as from a PRA, in the 10
CFR 50.59 process would be a logical evolution that would enhance the
existing framework by leveraging modern analytical tools to better
focus licensee and agency resources on issues of genuine safety
significance.
B. Improved Flexibility for Changes to Methods of Evaluation
The NRC proposes two regulatory amendments, which would work in
concert to allow licensees greater flexibility to implement changes to
analytical methods described in the FSAR.
First, the NRC proposes a targeted revision to 10 CFR
50.59(c)(2)(viii) and 53.1550(a)(2)(iv). This change would allow
licensees to implement certain changes to analytical methods described
in the FSAR (as updated) without prior NRC approval, provided those
changes are undertaken pursuant to an NRC-approved VVUQ program under
10 CFR 50.221, ``Credibility requirements for modeling and
simulation.'' Proposed guidance for compliance with proposed 10 CFR
50.221 is in DG-1468, ``Guidance for Implementation of 10 CFR 50.221,
`Credibility requirements for modeling and simulation.''' This revision
would clarify that appropriate changes made using a risk-informed and
graded VVUQ framework would be permissible under 10 CFR 50.59 without
prior NRC approval. The proposed rule would allow such changes without
prior NRC approval only if the VVUQ program were approved by the NRC
for the method of evaluation in question, and the new method of
evaluation met the credibility criteria established in the approved
VVUQ. These measures would maintain safety and ensure appropriate
controls over licensee changes to methods or evaluations while
affording flexibility through reliance on the NRC-approved VVUQ
program.
Second, the NRC proposes to adopt an optional regulation on VVUQ at
10 CFR 50.221 that would establish the requirements a VVUQ program must
meet. The new regulation would establish clear requirements and
structure for VVUQ activities used to support regulatory decisions.
Specifically, the proposed 10 CFR
[[Page 44574]]
50.221 would specify that licensees and applicants could voluntarily
establish a VVUQ program and determine the scope of the models and
simulations the VVUQ program would cover. Licensees and applicants that
establish a VVUQ program would be required to establish a process under
the VVUQ program for demonstrating the credibility of models and
simulations through VVUQ activities and assessments and could use only
models and simulations within the scope of the program for which the
licensee or applicant has successfully completed VVUQ activities and
assessments. Additionally, the licensee or applicant adopting a VVUQ
program would need to ensure that the VVUQ activities and assessments
were commensurate with the overall risk from the model or simulation.
The NRC has prepared guidance in DG-1468 on an acceptable approach for
a VVUQ program that would conform to the proposed regulation. This
approach would provide a consistent, technology-neutral framework for
demonstrating model credibility across the NRC's regulatory structure.
Via these proposed changes, the NRC would shift focus away from
evaluating each individual model or simulation directly and toward
evaluating the process by which models are determined to be credible--
specifically through structured VVUQ programs. In this context,
credibility would refer to the level of trust in a model's ability to
produce accurate and appropriate predictions for its intended use. An
NRC-approved VVUQ program would establish a new licensing basis that
focuses on how methods are selected and applied, rather than the
characteristics of the specific method and the inherent conservatisms.
Where appropriate, VVUQ-based credibility assessments could serve as a
viable alternative to full NRC review of each new or revised model.
However, a single generic VVUQ process would not be appropriate due to
the significant variability in physical phenomena, modeling
assumptions, numerical techniques, and uncertainties across different
reactor technologies and methods of evaluation. A tailored VVUQ process
approved by the NRC for the intended application would be required for
each method of evaluation to provide the necessary specificity to
ensure credible, defensible assessments of model performance for each
unique application for all possible reactor technologies and their
vastly different physical domains.
This approach could be particularly beneficial for new and advanced
reactor designs. Unlike the current fleet, which has decades of
operational data and analytical stability, advanced reactors often lack
extensive experimental databases at the time of initial licensing.
Requiring them to demonstrate method maturity at the level of existing
plants would demand significant upfront testing and analysis, delaying
deployment and increasing cost. Many of these designs are being
developed as test reactors specifically to generate such data. A rule
change would provide a clear, structured mechanism for these reactors
to update their methods over time, based on data collected during
operation, without needing to go through repeated full NRC reviews--so
long as the updates are made through an approved VVUQ process.
XIII. Background--Minimum Decommissioning Funding Assurance
Requirements for Non-Large Light-Water Reactors
The regulation in 10 CFR 50.75, ``Reporting and recordkeeping for
decommissioning planning,'' establishes requirements for indicating to
the NRC how an applicant or licensee will provide reasonable assurance
that funds will be available for the decommissioning process. During
the operational phase of a reactor facility, an applicant or licensee
must certify that funding is being provided in an amount that may be
more, but not less, than the amount described in 10 CFR 50.75(c)(1) and
(2) (also known as the table of minimum amounts, minimum funding
assurance, or ``formula'' amount). As the NRC stated in the 1988
decommissioning rule (53 FR 24018-24030; June 27, 1988), the
``formula'' amount in 10 CFR 50.75(c) does not represent the actual
cost of decommissioning for specific reactors but rather serves as a
reference level established to ensure that the bulk of the funds
necessary for a safe decommissioning is being considered and planned
for early in facility life by licensees. This provides assurance that
the facility will not become a risk to public health and safety when it
is decommissioned.
The table of minimum amounts and the associated adjustment factors
were developed and designed specifically for the large light-water
reactor technologies (boiling water reactors and pressurized water
reactors) that make up the current commercial power reactor fleet in
the U.S. However, new reactors may incorporate different technologies
and output capacities that may not require the amount of
decommissioning funding assurance described in 10 CFR 50.75(c). In
order to address different decommissioning funding needs for these new
technologies without requiring an exemption from NRC regulations, the
NRC is proposing updates to its regulations to allow certain new
reactor applicants and licensees the flexibility to certify adequate
decommissioning funding assurance during operations through the use of
either the table of minimum amounts or the submission of a design-
specific decommissioning cost estimate. Allowing for the use of a
design-specific decommissioning cost estimate that may be less than the
table of minimum amounts would provide a path for certain new reactor
applicants and licensees to demonstrate financial responsibility for
safe decommissioning based on factors specific to the reactor facility.
XIV. Discussion--Minimum Decommissioning Funding Assurance Requirements
for Non-Large Light-Water Reactors
The NRC proposes an amendment to 10 CFR 50.75 to allow certain new
reactor applicants and licensees to submit a design-specific
decommissioning cost estimate to demonstrate minimum decommissioning
funding assurance during operations that may be less than the table of
minimum amounts provided in 10 CFR 50.75(c). The values in the current
table of minimum amounts are based on funding assumptions associated
with the decommissioning of large light-water reactor facilities. This
proposed rule would allow certain new reactor applicants and licensees
to certify financial assurance for decommissioning through the use of
either the minimum formula amount or through the submission of a
design-specific decommissioning cost estimate. Thus, the proposed rule
would provide flexibility for new reactor applications that represent
smaller output and size considerations than large light-water reactor
designs.
Specifically, the NRC is proposing to add a new paragraph (b)(2) to
10 CFR 50.75 that describes the certification amount process and
minimum requirements, including reliance on design-specific
decommissioning cost estimates, for new reactor applicants and
licensees seeking to use the alternative pathway. Additionally, the NRC
is proposing to delete language in the table of minimum amounts in 10
CFR 50.75(c)(1) that requires reactors of less than 1200 megawatts
thermal (MWt) to use the certification amount for a 1200 MWt reactor.
It is conceivable that new LWR designs could have an output less than
1200 MWt. Therefore, requiring a new reactor applicant or
[[Page 44575]]
licensee to assure to an amount well above what is needed to establish
decommissioning funding assurance could be financially limiting and
unnecessary to meet the intent of decommissioning funding assurance
regulations.
A certification relying on a design-specific decommissioning cost
estimate would be required to include a description of the factors used
to develop the design-specific decommissioning cost estimate, including
generic activities performed in the major decommissioning phases of a
decommissioning project (e.g., pre-decommissioning engineering and
planning, reactor deactivation, and dismantlement). Similar to the
table of minimum amounts for large light-water reactor designs, design-
specific decommissioning cost estimates should represent the bulk of
funds necessary to safely decommission a facility, as applicable to the
specific reactor technology being utilized and the design of the
facility. Additionally, similar to the table of minimum amounts for
large light-water reactor designs, the design-specific decommissioning
cost estimate would have to be adjusted annually at a rate at least
equal to the formula in 10 CFR 50.75(c)(2). This certification process
would include NRC review and approval. However, once an initial design-
specific decommissioning cost estimate is approved by the agency as a
sufficient certification amount for financial assurance for
decommissioning, other applicants or licensees using similar technology
could reference and justify use of this amount (escalated in accordance
with NRC regulations and guidance) as the certification amount required
by proposed 10 CFR 50.75(b) for a different application. Once a
licensee nears permanent cessation of operations, a site-specific
decommissioning cost estimate that encompasses the design-specific cost
as well as costs associated with the site and operational period of the
facility, would be required for funding assurance purposes, as
described in current regulations in 10 CFR 50.82, ``Termination of
license.'' Finally, the NRC is proposing to revise 10 CFR
50.75(e)(1)(i) and (ii) to allow reactor licensees that have prepaid or
collected funds based on a design-specific estimate to take credit for
projected earnings on the prepaid or collected decommissioning funds
using up to a 2-percent annual real rate of return up to the time of
permanent termination of operations.
The NRC is proposing a similar change to 10 CFR part 53. Currently,
10 CFR part 53 only allows for a site-specific decommissioning cost
estimate as the certification amount. Therefore, the NRC is proposing
to add conforming language to 10 CFR 53.1010, ``Financial assurance for
decommissioning,'' and 53.1020, ``Cost estimates for decommissioning,''
to allow new reactor applicants and licensees to submit a design-
specific decommissioning cost estimate to demonstrate minimum
decommissioning funding assurance during operations. Finally, the NRC
is proposing to add conforming language to 10 CFR 53.1040, ``Methods
for providing financial assurance for decommissioning,'' to allow new
reactor applicants and licensees that have prepaid or collected funds
based on a design-specific estimate to take credit for projected
earnings on the prepaid or collected decommissioning funds using up to
a 2-percent annual real rate of return up to the time of permanent
termination of operations.
Additionally, conforming changes would be made to 10 CFR 50.75(e)
to include references to the requirements of proposed 10 CFR
50.75(b)(2), where appropriate. In addition, this proposed rule would
revise 10 CFR 50.75(e) to include ``applicant or'' in all appropriate
places where currently only ``licensee'' is referenced, as directed by
the Commission in staff requirements memorandum (SRM)-SECY-23-0021:
Enclosure 4, ``Table of Typographical errors and Inconsistencies,''
dated March 4, 2024. Similar changes are proposed in 10 CFR 53.1040 and
10 CFR 53.1050, ``NRC oversight'' for consistency. Collectively, these
changes would clarify that applicants and licensees would be subject to
the requirements under 10 CFR 50.75(e), 53.1040, and 53.1050, as
applicable.
This proposed rule also would make minor editorial changes in 10
CFR 50.75(e), (g), and (h) by removing errant commas, correcting
capitalization errors, correcting references by indicating paragraphs
instead of sections, and removing ``of this part,'' where necessary.
XV. Background--Incorporation of Streamlined Quality Assurance Criteria
for Nuclear Power Plants and Fuel Reprocessing Plants
A. Historic Quality Assurance Requirements Perspectives and Emergent
Issues
During the early days of nuclear power (1950s-1960s), the Atomic
Energy Commission (AEC), the NRC's predecessor agency, focused on
developing and licensing nuclear reactors. As the nuclear industry
grew, it became clear that systematic quality assurance (QA) was
essential to ensure nuclear safety, especially given the complexity and
potential hazards of nuclear technology. By the 1960s, nuclear power
plants were becoming more complex, and the consequences of failures
were more severe. Incidents and near-misses highlighted the need for
formalized QA programs to prevent design, fabrication, and construction
errors. Appendix B was added to 10 CFR part 50 by the AEC in 1970 (35
FR 10498; June 27, 1970) to (1) establish minimum QA requirements for
safety-related structures, systems, and components (SSCs) and (2)
ensure that these SSCs are designed, fabricated, constructed, and
tested to perform their intended safety functions.
Applicants for CPs, OLs, early site permits (ESPs), COLs, design
certifications, standard design approvals, and manufacturing licenses
(MLs) must include in their respective application a description of the
QA program that discusses how the applicable requirements of appendix B
to 10 CFR part 50 are satisfied.
Although appendix B to 10 CFR part 50 is foundational to nuclear
safety, commenters have expressed concerns over its implementation,
flexibility, and alignment with modern practices. These concerns
include the following topics.
Inflexibility: Appendix B to 10 CFR part 50 (1) is
prescriptive and has not been substantively updated since 1970; (2)
lacks risk-informed or performance-based flexibility, which modern
quality systems increasingly emphasize; and (3) does not facilitate
tailoring of QA programs to low-risk activities or innovative
technologies.
Outdated Language: The language in appendix B to 10 CFR
part 50 is reflective of technologies from the 1970s and does not
explicitly address digital systems, software QA, and additive
manufacturing. As a result, applicants must rely on guidance, which
could result in inconsistent implementation across applicants due to
applicants interpreting the guidance differently based on their
specific technologies.
Vendor and Supply Chain Challenges: Many suppliers,
especially non-nuclear vendors, are unfamiliar with appendix B to 10
CFR part 50, and thus applicants and licensees have challenges in
procuring products and services for the nuclear power plants.
Lack of Harmonization with International Standards:
Appendix B to 10 CFR part 50 is United States-specific and not aligned
with international standards and best practices and, thus, creates
challenges for international collaboration and global supply chains.
[[Page 44576]]
In light of these considerations, commenters have advocated for a
modernized, risk-informed QA framework that retains the safety rigor of
appendix B to 10 CFR part 50 while allowing for graded application
based on safety significance that is integrated with modern quality
systems.
B. NRC Responses to These Issues
As a result of these issues, the NRC is proposing to add appendix T
to 10 CFR part 50 as a voluntary alternative to appendix B to 10 CFR
part 50. The proposed appendix T would draw on international QA
standards to incorporate the following elements:
Performance-based QA criteria that provide explicit
direction on use of a graded approach for applying QA requirements to
SSCs relative to their safety and risk contributions to the overall
nuclear facility.
Quality assurance requirements specific to software used
in digital items and for design and analysis.
Quality assurance terminology and methodologies used
across various safety critical industries and in international
standards for quality management, thus allowing applicants to leverage
cross-industry and global supply chains.
XVI. Discussion--Incorporation of Streamlined Quality Assurance
Criteria for Nuclear Power Plants and Fuel Reprocessing Plants
This proposed rule would add a new appendix T to 10 CFR part 50 to
provide streamlined QA criteria that could be used for applications of
COLs, CPs, and OLs under certain eligibility requirements.
A. Introduction and Scope
As discussed in section XV, ``Background--Incorporation of
Streamlined Quality Assurance Criteria for Nuclear Power Plants and
Fuel Reprocessing Plants,'' of this document, NRC stakeholders have
expressed interest in utilizing a streamlined approach to QA that
better aligns with international standards. Therefore, the NRC has
developed a proposed appendix T to 10 CFR part 50 that would provide an
alternative to the current QA requirements in appendix B to 10 CFR part
50 based on International Standard, ISO 19443, ``Quality management
systems--Specific requirements for the application of ISO 9001:2015 by
organizations in the supply chain of the nuclear energy sector
supplying products and services important to nuclear safety [ITNS],''
(2018-05). The International Standard Organization collaborated closely
with the International Atomic Energy Agency in developing ISO 19443.
The proposed section I, ``Introduction and Scope,'' of appendix T
to 10 CFR part 50 would provide the eligibility requirements for using
appendix T to 10 CFR part 50 as a voluntary alternative to appendix B
to 10 CFR part 50. Specifically, applicants for CPs, OLs, and COLs
would have the option to use appendix T to 10 CFR part 50 as an
alternative to appendix B to 10 CFR part 50, provided that the three
conditions in section I are met.
Proposed condition I.A would require that the application is for an
nth-of-a-kind (NOAK) plant and would require the application to
identify the first-of-a-kind (FOAK) reference plant.
Proposed condition I.B would require that any departures from the
FOAK reference plant in the application of the NOAK plant would not
result in a change to the classification, design, and method of
manufacture, construction, and operation of SSCs identified in
licensing basis of the referenced plant.
Proposed condition I.C would require the application to include
procedures and work processes for implementing the requirements in
proposed appendix T to 10 CFR part 50.
The NRC would define FOAK nuclear power plants and fuel
reprocessing plants in proposed appendix T as the initial
implementation of a new reactor design or technology or new fuel
reprocessing plant design that has not been previously constructed and
operated at commercial scale, either within the U.S. or
internationally. The FOAK plant would serve as a reference plant for
NOAK nuclear power plants and fuel reprocessing plants. The NRC would
define NOAK nuclear power plants and fuel reprocessing plants in
proposed appendix T to be any subsequent implementation of a FOAK
reactor design or technology or fuel reprocessing plant design after
the FOAK plant has been designed, constructed, and operated.
Proposed conditions I.A, I.B, and I.C would limit the use of
proposed appendix T to 10 CFR part 50 to those applications that could
potentially leverage the following:
the design maturity of the FOAK completed plant design
information;
standardized components and systems to streamline
procurement and construction;
skilled labor and contractors from the FOAK projects;
FOAK operational data to help NOAK commissioning
procedures, standard operating procedures, emergency operating
procedures, training, design, and operational programs; and
established processes and procedures for implementing the
requirements in proposed appendix T to 10 CFR part 50, including an
established mechanism to ensure deviations from FOAK reliability in
NOAK would be identified and promptly corrected.
In recent cases, the NRC has noticed that design details and QA
procedures are not available during the licensing review of a CP, OL,
or COL for a FOAK plant. Therefore, during licensing and construction
for FOAK plants, the NRC conducts vendor inspections to confirm the
detailed design and as-built SSCs meet the technical and quality
requirements committed to by the applicant. However, because a NOAK
applicant could reference design details and QA procedures developed
during the FOAK licensing, these additional QA activities, such as
vendor inspections, may be unnecessary, especially for vendors
previously inspected by the NRC. Therefore, when the conditions I.A,
I.B, and I.C are met, the proposed appendix T to 10 CFR part 50 may
eliminate the need for NRC oversight of suppliers and vendors who
supply applicants with approved appendix T to 10 CFR part 50 compliant
QA programs because (1) the list of SSCs that would be governed by
proposed appendix T to 10 CFR part 50 would be standardized and
verified to be acceptable using information from the reference FOAK
plant; (2) the design of these SSCs would be complete and verified to
be acceptable using information from the reference FOAK plant; and (3)
the manufacturing and construction methods for these SSCs would be
established and verified to be acceptable during oversight of
manufacturing and construction processes for the referenced FOAK plant.
B. Definitions
The proposed section II, ``Definitions,'' of proposed appendix T to
10 CFR part 50, would include definitions for terms used in the
proposed appendix.
The NRC would define ``first-of-a-kind'' nuclear power plants and
fuel reprocessing plants as the initial implementation of a new reactor
design or technology or new fuel reprocessing plant design that has not
been previously constructed and operated at commercial scale.
The NRC would define ``nth-of-a-kind'' nuclear power plants and
fuel reprocessing plants as any subsequent implementation of a reactor
design or technology or fuel reprocessing plant
[[Page 44577]]
design after the FOAK has been designed, constructed, and operated.
The NRC would define ``quality assurance'' as all those planned and
systematic actions necessary to provide adequate confidence that a
structure, system, or component will perform satisfactorily in service.
Quality assurance includes quality control, which comprises those
actions related to the physical characteristics of a material,
structure, component, or system that provide a means to ensure the
material, structure, component, or system meets predetermined
requirements. This proposed definition is equivalent to the definition
used in appendix B to 10 CFR part 50.
The NRC would define ``quality assurance program'' as the overall
program established to assign responsibilities and authorities, define
policies and requirements, and provide for the performance and
assessment of work necessary to achieve QA.
The NRC would define ``quality management system'' (QMS) as a
structured framework that documents an organization's processes,
procedures, and responsibilities for ensuring quality. This term and
definition are used broadly by other safety-critical industries,
nuclear regulatory bodies and industry abroad, and vendors and third-
party suppliers to these industries. The QMS is different from the
terminology ``quality assurance program description'' (QAPD) used in
appendix B to 10 CFR part 50 in that the QMS has a broader scope and is
a system framework that includes quality planning, controls, assurance,
and improvement; whereas a QAPD is a descriptive document that is
narrowly focused on QA.
The NRC would define ``functional design criteria'' as metrics for
the performance of SSCs. For safety-related SSCs, these criteria define
performance metrics necessary to demonstrate compliance with the safety
criteria in 10 CFR 53.210, ``Safety criteria for design-basis
accidents.'' For non-safety-related but safety-significant SSCs, these
criteria define performance metrics necessary to demonstrate compliance
with the safety criteria in 10 CFR 53.220, ``Safety criteria for
licensing-basis events other than design-basis accidents.'' This
proposed definition would be added to proposed appendix T to 10 CFR
part 50 to align with the definition and use of this term in 10 CFR
part 53.
The NRC would define ``non-safety-related but safety-significant
SSCs'' as those SSCs that are not safety-related but are relied on to
achieve adequate defense in depth or perform risk-significant functions
and warrant special treatment. This proposed definition would be added
to proposed appendix T to 10 CFR part 50 to align with the definition
and use of this term in 10 CFR part 53.
C. General Requirements
The proposed section III, ``General Requirements,'' of proposed
appendix T to 10 CFR part 50 would provide general requirements for
establishing and maintaining a QA program for applicants that choose to
meet proposed appendix T.
Proposed section III.A, ``Integrated Quality Assurance Program,''
of proposed appendix T to 10 CFR part 50 would require that the
integrated QA program ensures that safety-related and non-safety-
related but safety-significant SSCs are designed, fabricated, erected,
and tested to quality standards commensurate with the importance of the
safety functions those SSCs perform. Proposed section III.A of in
proposed appendix T to 10 CFR part 50 would include the following seven
items that any application using in proposed appendix T to 10 CFR part
50 would be required to identify and explain in the integrated QA
program:
Responsibilities (1) are properly assigned to specific
individuals or teams in charge of executing QA activities and (2)
ensure any delegated responsibilities are properly identified and
controlled.
The design requirement of SSCs are sufficiently captured
in corresponding documents; the design bases requirements are
adequately translated into specifications, drawings, procedures, and
instructions; outputs reflect the correct design inputs; the design is
properly verified and validated; the as-built and as-operated SSC
properly meet the intended function and safety margin.
Means and methods are established to communicate relevant
technical, quality, and regulatory requirements, expectations, and
concerns between the applicant and its vendors and third-party
suppliers.
Measures are established to (1) ensure that procured SSCs
and related services meet technical and quality requirements and (2)
assess the capability of vendors or third-party suppliers that supply
the SSCs and related services.
Measures are established to (1) verify and validate that
products and services meet the technical and quality requirements of
the procured products and services, and (2) audit the vendors or third-
party suppliers that are providing the products and services.
Processes are implemented to address reoccurrence of
issues and failures.
Recordkeeping and documentation protocols for the QA
program are established.
Proposed section III.B.1 of proposed appendix T to 10 CFR part 50
would require the applicant document the QA program in the QMS and
submit the QMS to the NRC for review and approval. Proposed section
III.B.1 would require the QA program, as documented in the QMS, to
contain a graded approach for implementing the requirements of the QA
program.
Proposed section III.B.2 of proposed appendix T to 10 CFR part 50
would require the QMS to describe how the requirements in section IV of
the appendix would be met. Proposed section IV of appendix T to 10 CFR
part 50 would identify general QA criteria and software QA criteria.
Proposed section III.B.3 of appendix T to 10 CFR part 50 would require
the applicant to invoke the applicant's QA requirements in procurement
documents to all relevant contractors, vendors, suppliers, and third-
parties.
Proposed section III.B.4 of proposed appendix T to 10 CFR part 50
would require the applicant to select the appropriate industry
standards that are used to achieve quality consistent with regulatory
requirements and the NRC's policies. This proposed section would state
that the applicant would need to document the selection of ASME Nuclear
Quality Assurance (NQA)-1, ``Quality Assurance Requirements for Nuclear
Facility Applications'' or another appropriate industry standard.
Proposed section III.B.4 of appendix T to 10 CFR part 50 would also
require that gaps between the selected industry standards and the
proposed section IV, ``Quality Assurance Requirements,'' of appendix T
to 10 CFR part 50 are addressed within the QMS.
D. Quality Assurance Requirements
Proposed section IV.A, ``Quality Assurance Criteria,'' of proposed
appendix T to 10 CFR part 50, identifies QA criteria that would be
applicable to all applications that reference proposed appendix T to 10
CFR part 50. This proposed section would include 11 criteria that cover
topical areas in management, performance, and assessment. The proposed
requirements in these topical areas are consistent with International
Standards for QMSs such as ISO 9001, ``Quality Management System--
Requirements,'' which are used by many safety-critical industries, and
ASME NQA-1, ``Quality Assurance Requirements for Nuclear Facility
[[Page 44578]]
Applications,'' which is used by the nuclear industry.
(i) Management
Proposed section IV.A.1, ``Criterion 1--Management: Program,'' of
proposed appendix T to 10 CFR part 50, would include requirements for
the applicant to define the organizational structure, functional
responsibilities, levels of authority, and interfaces for performing
the work necessary to implement the QA program, and develop management
processes to plan, schedule, and assign resources to perform this work.
Proposed section IV.A.2, ``Criterion 2--Management: Personnel
Training and Qualifications,'' of proposed appendix T to 10 CFR part
50, would include requirements for the applicant to develop processes
for indoctrination and continuous training of employees for performing
the work necessary to implement the QA program.
Proposed section IV.A.3, ``Criterion 3--Management: Quality
Improvement,'' of proposed appendix T to 10 CFR part 50, would include
requirements for the applicant to establish and implement a process for
identifying and controlling issues and failures that could adversely
impact quality, safety, and regulatory compliance. The proposed section
IV.A.3 would also require applicants to include prevention of
recurrence of issues as part of corrective actions and implement
processes for continuous improvement of the QA program.
Proposed section IV.A.4, ``Criterion 4--Management: Documents and
the Associated Records,'' of proposed appendix T to 10 CFR part 50,
would include requirements for the applicant to prepare, review,
approve, issue, use, and revise documents that prescribe processes,
specify requirements, or establish the design of the SSC, and maintain
these documents as records for the QA program.
(ii) Performance
Proposed section IV.A.5, ``Criterion 5--Performance: Work
Processes,'' of proposed appendix T to 10 CFR part 50, would include
requirements for the applicant to perform work, including hazard
controls. Hazard controls are systematic measures designed to prevent,
detect, and correct issues that could compromise quality, safety, and
regulatory compliance. Examples of hazard controls applicable to QA
programs include:
embedded work processes to ensure quality such as
inspection and testing protocols, hold points and witness points, and
nonconformance reporting protocols;
monitoring and detection programs to ensure detection of
deviations and malfunctions in real time such as surveillance and
audits;
structured approaches to identify quality issues and
implement corrective and preventive actions.
Proposed section IV.A.5 of proposed appendix T to 10 CFR part 50,
would also require the applicant to establish and implement work
processes for identifying and controlling items to ensure proper use;
maintain items to prevent damage, loss, or deterioration; and calibrate
and maintain equipment used for activities affecting quality.
Proposed section IV.A.6, ``Criterion 6--Performance: Design,'' of
proposed appendix T to 10 CFR part 50, would include requirements for
the applicant to establish and implement measures for controlling the
design of SSCs, including requirements for controlling design changes,
design interfaces, and verifying and validating the adequacy of the
design. Verify in the context of design control means to perform the
set of activities to demonstrate that design conforms to specifications
and occurs during the design and development process. Examples include
reviews, inspections, and unit tests. Validate in the context of design
control means to perform the set of activities to demonstrate that the
as-developed or as-built SSC performs the intended safety-functions and
occurs after the development process. Examples include integrated
tests, analysis, and simulations.
Proposed section IV.A.7, ``Criterion 7--Performance: Procurement,''
of proposed appendix T to 10 CFR part 50, would include requirements
for the applicant to establish and implement processes for procurement
of items and services, including processes to verify that procured
items and services meet established requirements, evaluate and select
prospective suppliers, and verify that the approved suppliers continue
to provide acceptable items and services.
Proposed section IV.A.8, ``Criterion 8--Performance: Inspection and
Acceptance Testing,'' of proposed appendix T to 10 CFR part 50, would
include requirements for the applicant to establish and implement
processes for performing inspection and acceptance testing for procured
items and services.
Proposed section IV.A.9, ``Criterion 9--Performance: Maintenance of
Structures, Systems, and Components,'' of proposed appendix T to 10 CFR
part 50, would include requirements for the applicant to establish and
implement processes to control the storage of SSCs in accordance with
cleanliness and environmental standards. These requirements would
ensure that a process is used to prevent foreign material from being
introduced to the SSC during storage and to store SSCs in accordance
with the required environmental conditions (e.g., humidity,
temperature).
(iii) Assessment
Proposed section IV.A.10, ``Criterion 10--Assessment: Management
Assessment,'' of proposed appendix T to 10 CFR part 50, would include
requirements for the applicant to establish and implement processes for
management of the organization to assess the continued effectiveness of
the QA program.
Proposed section IV.A.11, ``Criterion 11--Assessment: Independent
Assessment,'' of proposed appendix T to 10 CFR part 50, would include
requirements for the applicant to establish and implement processes for
independent assessment of each aspect of the QA program. These
requirements would ensure that those performing these assessments have
sufficient authority and freedom from their management and are
technically qualified and knowledgeable to perform the assessment.
E. Quality Assurance for Software Used in Design and Analysis, and
Digital Items Important to Safety
Proposed section IV.B, ``Quality Assurance for Software Used in
Design and Analysis, and Digital Items Important to Safety,'' of
proposed appendix T to 10 CFR part 50, would identify QA criteria for
software used for design and analysis of SSCs and for digital items
important to safety.
Proposed section IV.B.1 of proposed appendix T to 10 CFR part 50,
would require that applicants establish and implement processes within
the QA program to ensure that (1) software used in digital items that
perform a safety function, (2) software used for design verification
for any SSC, and (3) software used for design analysis for any SSC, are
documented, managed, and controlled throughout the software life cycle
to ensure that the related SSCs perform their intended safety function.
Proposed section IV.B.2 would require the applicant to use
appropriate national or internal software engineering standards.
Examples of such standards include ASME, Institute for Electrical and
Electronics Engineers (IEEE), National Institutes of Standards and
Technology (NIST), and American Nuclear Society (ANS).
[[Page 44579]]
F. Proposed Conforming Changes to 10 CFR 50.4, 50.34, 50.54, 50.55,
52.79, 53.020, 53.040, 53.460, 53.500, 53.865, 53.1309, 53.1369,
53.1416, and 53.1565
Proposed conforming changes to 10 CFR 50.34(a)(7) would allow a CP
applicant who meets the eligibility requirements included in proposed
section I of appendix T to 10 CFR part 50, to include in its
application for a CP, a QMS that meets proposed appendix T to 10 CFR
part 50, as an alternative to satisfying the requirement in 10 CFR
50.34(a)(7) for submittal of a description of the QA program that meets
appendix B to 10 CFR part 50.
Proposed conforming changes to 10 CFR 50.34(b)(6)(ii) would allow
an OL applicant who meets the eligibility requirements included in
proposed section I of appendix T to 10 CFR part 50, to include in its
application for an OL, a QMS that meets proposed appendix T to 10 CFR
part 50, as an alternative to satisfying the requirement in 10 CFR
50.34(b)(6)(ii) for submittal of a description of the QA program that
meets appendix B to 10 CFR part 50.
Proposed conforming changes to 10 CFR 50.34(f)(3)(ii) would add a
reference to the proposed appendix T to 10 CFR part 50 for the
requirement on ensuring all SSCs important to safety are included in
the QA list.
Proposed conforming changes to 10 CFR 50.54(a)(1) would incorporate
requirements for:
Each nuclear power plant or fuel reprocessing plant
licensee subject to the QA criteria in proposed appendix T of 10 CFR
part 50, to implement, under 10 CFR 50.34(b)(6)(ii) or 52.79, the QMS
described or referenced in the safety analysis report, including
changes to that report.
For holders of a COL under 10 CFR part 52, to implement
the QMS described or referenced in the safety analysis report
applicable to operation 30 days prior to the scheduled date for initial
loading of the fuel.
The proposed conforming addition of 10 CFR 50.54(a)(5) would
include requirements for changes to a QMS to be submitted to the NRC
and receive NRC approval prior to implementation.
Proposed conforming changes to 10 CFR 50.55(f)(1) would incorporate
requirements for nuclear power plant or fuel reprocessing plant CP
holders subject to the QA criteria in proposed appendix T of 10 CFR
part 50, to implement, pursuant to 10 CFR 50.34(a)(7), the QMS
described or referenced in the safety analysis report, including
changes to that report.
The proposed conforming addition of 10 CFR 50.55(f)(5) would add
requirements for changes to a QMS to be submitted to the NRC and
receive NRC approval prior to implementation.
The proposed conforming addition of paragraph (b)(7)(iii) to 10 CFR
50.4, ``Written communications,'' would require a change to the safety
analysis report QMS under the proposed 10 CFR 50.54(a)(5) or 10 CFR
50.55(f)(5), or a change to a licensee's NRC-accepted QMS topical
report under 10 CFR 50.54(a)(5) or 10 CFR 50.55(f)(5), to be submitted
to the NRC's Document Control Desk, with a copy to appropriate Regional
Office, and a copy to the appropriate NRC Resident Inspector if one has
been assigned to the site of the facility.
Proposed conforming changes to 10 CFR 52.79(a)(25) and (27) would
allow a COL applicant who meets the eligibility requirement included in
proposed section I of appendix T to 10 CFR part 50, to include in its
application for COL, a QMS that meets proposed appendix T to 10 CFR
part 50, as an alternative to satisfying the requirement in 10 CFR
52.79(a)(25) and (27) for submittal of a description of the QA program
that meets appendix B to 10 CFR part 50.
Proposed conforming changes to 10 CFR 53.020 would modify the
definition of QA to align with the definition of QA in the proposed
appendix T to 10 CFR part 50.
The proposed conforming addition of paragraph (b)(7)(iii) to 10 CFR
53.040, ``Written communications,'' would require a change to the
safety analysis report QMS under the proposed 10 CFR 53.1565,
``Evaluating changes to programs included in licensing-basis
information,'' or a change to a licensee's NRC-accepted QMS topical
report under 10 CFR 53.1565, to be submitted to the NRC's Document
Control Desk, with a copy to appropriate Regional Office, and a copy to
the appropriate NRC Resident Inspector if one has been assigned to the
site of the facility.
Proposed conforming changes to paragraphs (b)(1) and (2) of 10 CFR
53.460, ``Safety categorization and special treatments,'' would allow,
for applicants that meet the eligibility requirements included in
proposed section I of appendix T to 10 CFR part 50, the special
treatments for safety-related SSCs (under proposed 10 CFR
53.460(b)(1)), and non-safety-related safety-significant SSCs and
safety-related SSCs beyond 10 CFR 53.460(b)(1) (under proposed 10 CFR
53.460(b)(2)), to meet applicable QA requirements from proposed
appendix T to 10 CFR part 50, as an alternative to these SSCs having to
meet the applicable QA requirement in appendix B to 10 CFR part 50.
Proposed conforming changes to paragraph (b) of 10 CFR 53.500,
``General siting and siting assessment,'' would allow, for applicants
that meet the eligibility requirements included in proposed section I
of appendix T to 10 CFR part 50, activities performed to identify site
characteristics or otherwise needed to determine site-specific
contributors to functional design criteria or analysis assumptions
under subpart C of 10 CFR part 53 to satisfy the QA requirements from
proposed appendix T to 10 CFR part 50, as an alternative for these
activities to meet the applicable QA requirement in appendix B to 10
CFR part 50.
Proposed conforming changes to 10 CFR 53.865, ``Quality
assurance,'' for holders of an OL or COL under 10 CFR part 53 that meet
the eligibility requirements included in proposed section I of appendix
T to 10 CFR part 50, to develop, implement, and maintain a QA program
in accordance with proposed appendix T to 10 CFR part 50, as an
alternative to appendix B to 10 CFR part 50.
Proposed conforming changes to 10 CFR 53.1309(a)(2)(i) would allow
a CP applicant under 10 CFR part 53 who meets the eligibility
requirement included in proposed section I of appendix T to 10 CFR part
50, to include in its application for a CP, a QMS that meets proposed
appendix T to 10 CFR part 50, as an alternative to satisfying the
requirement in 10 CFR 53.109(a)(2)(i) for submittal of a description of
the QA program that meets appendix B to 10 CFR part 50.
Proposed conforming changes to 10 CFR 53.1369(l) would allow an OL
applicant under 10 CFR part 53 who meets the eligibility requirement
included in proposed section I of appendix T to 10 CFR part 50, to
include in its application for an OL, a QMS that meets proposed
appendix T to 10 CFR part 50, as an alternative to satisfying the
requirement in 10 CFR 53.1369(l) for submittal of a description of the
QA program that meets appendix B to 10 CFR part 50.
Proposed conforming changes to 10 CFR 53.1416(a)(12) would allow a
COL applicant under 10 CFR part 53 who meets the eligibility
requirement included in proposed section I of appendix T to 10 CFR part
50, to include in its application for a COL, a QMS that meets proposed
appendix T to 10 CFR part 50, as an alternative to satisfying the
requirement in 10 CFR 53.1416(a)(12) for submittal of a description of
the QA program that meets appendix B to 10 CFR part 50.
[[Page 44580]]
Proposed conforming changes to 10 CFR 53.1565(d)(1)(i) would
clarify the applicability of QA criteria of appendix B of 10 CFR part
50 for each holder of an OL or COL under 10 CFR part 53, after the
Commission makes the finding under 10 CFR 53.1452(g).
Proposed conforming addition of 10 CFR 53.1565(d)(1)(iii) would
include requirements for changes to a QMS to be submitted to the NRC
and receive NRC approval prior to implementation for each holder of an
OL or COL under 10 CFR part 53, after the Commission makes the finding
under 10 CFR 53.1452(g).
Proposed conforming changes to10 CFR 53.1565(d)(2) would modify the
numbering scheme and clarify the applicability of QA criteria of
appendix B to 10 CFR part 50 for each holder of a CP or COL under 10
CFR part 53, before the Commission makes the finding under 10 CFR
53.1452(g).
The proposed conforming addition of 10 CFR 53.1565(d)(2)(ii) would
include requirements for changes to a QMS to be submitted to the NRC
and receive NRC approval prior to implementation for each holder of a
CP or COL under 10 CFR part 53, before the Commission makes the finding
under 10 CFR 53.1452(g).
XVII. Background--Updates to Construction Permit Requirements and
Related Licenses
The regulations in 10 CFR 50.34 specify the requirements for
technical information to accompany an application for a CP or an OL.
These regulations were amended in 1968 (33 FR 18610; December 17, 1968)
to add paragraph (a) of 10 CFR 50.34 to require an applicant for a CP
to submit a preliminary safety analysis report. Paragraph (a) of 10 CFR
50.34 specifies the minimum technical information in the preliminary
safety analysis report, including preliminary design information and a
description and safety assessment of the site on which the facility is
to be located. As stated in the 1968 final rule, the preliminary safety
analysis report requirement was ``intended to provide early and
adequate information which is expected to expedite the processing of CP
applications by reducing the time-consuming exchanges between the
applicant and the AEC staff required to fill information gaps.''
Subsequent changes to 10 CFR 50.34 from 2007 to the present were
additions due to new requirements, as well as clarifications and
relaxations, but the majority of 10 CFR 50.34 is unchanged since 1968.
The NRC may issue the CP if the agency makes the findings specific
to a CP that are listed in paragraph (a) of 10 CFR 50.35, ``Issuance of
construction permits,'' as well as the more general findings for
issuance of licenses and permits in 10 CFR 50.40, ``Common standards,''
and 10 CFR 50.50, ``Issuance of licenses and construction permits.''
The findings in 10 CFR 50.35(a) stem from the early practices of the
AEC, when a ``provisional'' CP would be issued when an applicant had
not submitted all the technical information necessary to complete the
application and to approve all proposed design features. Since almost
all issued ``provisional'' CPs were never converted to a ``final'' CP,
the AEC proposed codifying this practice (34 FR 6540; April 16, 1969).
The final amendment to the regulations in 10 CFR 50.35 eliminated the
term ``provisional'' CP, but the criteria in 10 CFR 50.35(a) for
issuing a CP remained the same as those previously required for a
``provisional'' CP (35 FR 5317; March 31, 1970). The current
regulations for issuing a CP in 10 CFR 50.35(a) have not been modified
since 1970.
The NRC issued 10 CFR part 52 on April 18, 1989 (54 FR 15372), to
reform the NRC's licensing process for future nuclear power plants. The
rule established new approval processes in 10 CFR part 52 for ESPs,
standard design certifications, and COLs that were additions to the
two-step licensing process that already existed in 10 CFR part 50. This
10 CFR part 52 rule also included processes for standard design
approvals and MLs. On August 28, 2007 (72 FR 49352), the NRC issued a
final rule with changes to 10 CFR part 52 to clarify the applicability
of various requirements to each of the 10 CFR part 52 approval
processes.
XVIII. Discussion--Updates to Construction Permit Requirements and
Related Licenses
This proposed rule would include updates to the language in 10 CFR
50.34(a) to more clearly link the level of detail required to be
submitted with a CP application to the findings the NRC is required to
make in 10 CFR 50.35(a), 50.40, and 50.50 to issue a CP. During its
review of recent CP applications, the NRC has noted that applicants may
provide a higher level of detail for some of the technical areas listed
in 10 CFR 50.34(a), and a lower level of detail for others, while still
providing sufficient information for the NRC to make the findings
required by 10 CFR 50.35(a), 50.40, and 50.50 to issue the CP. The
proposed revisions to footnote 1 of 10 CFR 50.34(a) would clarify that
the level of detail provided in a preliminary safety analysis report to
satisfy the minimum technical requirements in 10 CFR 50.34(a) would be
deemed sufficient if the provided information allows the NRC to make
the findings required by 10 CFR 50.35(a), 50.40, and 50.50.
This proposed rule would also remove a sentence in 10 CFR
50.34(a)(4) specifying in detail the need to perform loss of coolant
accident (LOCA) analyses required in 10 CFR 50.46 and the need for
high-point vents in proposed 10 CFR 50.46b, ``Acceptance criteria for
reactor coolant system venting systems.'' The NRC considers that these
requirements are already implicitly included in the preceding sentence
of 10 CFR 50.34(a)(4), which specifies the need to include an
evaluation of ``the adequacy of structures, systems, and components
provided for the prevention of accidents and the mitigation of the
consequences of accidents.'' These requirements are also referred to in
10 CFR 50.34(b)(4), and the NRC proposes to remove a sentence from 10
CFR 50.34(b)(4) that similarly specifies in detail the need to perform
LOCA analyses required in 10 CFR 50.46.
Conforming changes are also proposed to the similar regulatory text
and footnotes in each subpart of 10 CFR part 52 to ensure consistency
between the power reactor licensing and approval pathways. These
changes would be made in 10 CFR 52.47, which applies to standard design
certifications; 10 CFR 52.79, which applies to COLs; 10 CFR 52.137,
which applies to standard design approvals; and 10 CFR 52.157, which
applies to MLs. This change would reflect a more technology-inclusive
approach and remove prescriptive language that could be read to mean
that a LOCA analysis methodology is fully developed and validated at
the CP stage.
The proposed rule would also adjust the wording in 10 CFR
50.34(a)(1)(ii)(D) to be meaningful for designs with functional
containments which are evaluated as a release barrier or series of
barriers taken together to perform the containment safety function.
Specifically, with the proposed changes, the regulation would no longer
prescriptively state that the assumed fission product release be ``from
the core into the containment'' and would be replaced with technology-
inclusive language that refers to ``leakage rates from potential flow
paths'' rather than a ``containment leak rate.'' Light-water reactor
designs would continue to use fission product release paths from the
core into the containment and containment leak rates. The proposed rule
would also revise footnotes 3 and
[[Page 44581]]
6--which provide additional information on the fission product release
to be used in the analyses described in 10 CFR 50.34(a)(1)(ii)(D) and
in several items in 10 CFR 50.34(f)(1)--to be more technology-inclusive
and allow for evaluation of designs with mechanistic source terms and
functional containments. In addition, the proposed rule would revise
footnote 4, which discusses the use of 25 roentgen equivalent man (rem)
(0.25 sieverts (Sv)) total effective dose equivalent (TEDE) in 10 CFR
50.34(a)(1)(ii)(D)(1) as a reference value to remove outdated
information regarding recommendations included in a 1959 National
Bureau of Standards handbook.
Conforming changes would be made to the similar regulatory text and
footnotes in each subpart of 10 CFR part 52 to ensure consistency
between the power reactor licensing and approval pathways.
Specifically, these proposed changes would be made in 10 CFR 52.17,
which applies to ESPs, 10 CFR 52.47, 52.79, 52.137, and 52.157. These
changes would reflect a more technology-inclusive approach, eliminate
unnecessary exemptions that may otherwise be needed for some designs,
and ensure consistency in power reactor applications.
XIX. Background--Alternative Risk-Informed and Performance-Based
Acceptance Criteria for 10 CFR Parts 50 and 52
A. Need for Regulatory Flexibility
Many existing NRC regulations include prescriptive acceptance
criteria expressed as specific numerical limits (e.g., temperature,
pressure, dose). These criteria were developed based on the state of
knowledge and technology at the time the rules were promulgated and do
not always reflect the significant advancements in nuclear safety
analysis, PRA, and reactor design that have occurred in the decades
since. Additionally, the codification of these criteria has limited the
ability of licensees and applicants to propose alternative approaches
without seeking exemptions, which can introduce cost and regulatory
uncertainty.
Licensees and applicants have consistently identified unduly
prescriptive requirements as a deterrent to innovation. This has been
particularly challenging for U.S. companies developing new reactor
designs and seeking to compete in global markets. Nonetheless, the NRC
has determined that a top-down approach--modifying individual
prescriptive requirements throughout 10 CFR parts 50 and 52--would be
resource-intensive and could have unintended consequences, especially
for the licensing bases of currently operating reactors. Therefore,
instead of a top-down approach, the NRC is proposing to expand the use
of risk-informed and performance-based alternatives to existing
prescriptive requirements.
B. Enabling Risk-Informed and Performance-Based Alternatives
The NRC has long supported the use of risk-informed and
performance-based approaches in its regulatory decision-making, as
reflected in the Commission's policy statements and strategic goals,
particularly, SRM-SECY-98-144, ``Staff Requirements--SECY-98-144--White
Paper on Risk-Informed and Performance-Based Regulation,'' dated March
1, 1999. Since then, the agency has encouraged the use of such
approaches to improve regulatory decision-making, enhance safety, and
reduce unnecessary regulatory burden.
The proposed rule would build on this foundation by providing a
structured pathway for licensees and applicants to propose alternative
acceptance criteria that would be tailored to demonstrate the safety of
their specific technologies without the need for exemptions.
Additionally, the proposed rule would further utilize risk-informed and
performance-based methodologies to update appendix A to 10 CFR part 50
to clarify the application of general design criteria (GDCs) during the
licensing of new LWR designs.
XX. Discussion--Alternative Risk-Informed and Performance-Based
Acceptance Criteria for 10 CFR Parts 50 and 52
The NRC is proposing to add new, standalone regulations, 10 CFR
50.220 and 10 CFR 52.220, entitled ``Use of risk-informed and
performance-based alternatives to acceptance criteria,'' to allow
licensees and applicants to voluntarily submit and use technology-
inclusive, risk-informed, or performance-based acceptance criteria as
alternatives to existing prescriptive requirements. These new
provisions would support the expanded and accelerated use of acceptance
criteria reflective of innovative nuclear technologies without the need
for exemptions, while continuing to ensure reasonable assurance of
adequate protection of public health and safety.
In addition, the NRC is proposing to update appendix A to 10 CFR
part 50 to clarify that (1) deviations from GDCs could be identified
and justified within licensing submittals, with no need for a separate
exemption request; and (2) demonstrating compliance with Criterion 28,
``Reactivity limits'' (GDC 28), of appendix A to 10 CFR part 50 could
be based on a different design basis accident than the control rod
ejection or control rod drop accident.
The proposed 10 CFR 50.220 and 10 CFR 52.220 would address
regulatory inefficiencies and foster innovation by offering a flexible,
voluntary alternative to the current approach, which in many instances
relies on prescriptive requirements that applicants must seek
exemptions from when proposing to adopt innovative methodologies. The
NRC has identified several cases in which this type of framework could
have enabled more timely and efficient regulatory decisions. These
experiences highlight the value of reducing the number of exemptions
and rulemakings required, thereby improving efficiency and supporting
the deployment of new technologies. The proposed rule would also
provide opportunities for increased operational flexibility at existing
facilities that choose to propose and adopt alternative criteria.
The proposed approach would enhance regulatory flexibility,
efficiency, and reliability by providing a voluntary pathway for the
use of alternative criteria. This approach would maintain the existing
licensing basis for currently operating reactors and potential restart
units, thereby avoiding unintended impacts associated with a broad,
top-down revision of regulatory requirements. Additionally, the
proposed rule would establish a more transparent and structured process
for NRC review and acceptance of alternative criteria.
This approach would be consistent with the NRC's commitment to
enabling the safe use of nuclear technology for the benefit of society,
while maintaining reasonable assurance of adequate protection of public
health and safety. It would also be consistent with the approach of
previous NRC rulemakings that provided voluntary pathways for risk-
informed and performance-based alternatives to existing requirements,
such as the promulgation of 10 CFR 50.69, ``Risk-informed
categorization and treatment of structures, systems and components for
nuclear power reactors'' (69 FR 68008; November 22, 2004) and 10 CFR
part 53 (91 FR 15696; March 30, 2026).
The NRC recognizes that successful implementation of this proposed
approach would require broad and flexible guidance to accommodate the
range of potential alternative criteria. As a result, NRC encourages
the increased
[[Page 44582]]
use of pre-application engagement by applicants and licensees who would
wish to exercise the flexibility provided by the proposed 10 CFR 50.220
and 52.220. Draft guidance on the content of applications submitted
under this proposed rule is provided in DG-1464, ``Guidance for Content
of Applications Under 10 CFR 50.220 and 52.220 Proposing Risk-Informed
and Performance-Based Alternative Acceptance Criteria,'' which
accompanies this rulemaking.
Looking forward, the NRC would maintain a record of NRC-approved
alternative acceptance criteria, along with references to the
associated bases for approval to support streamlined use of approved
alternative acceptance criteria by potential applicants and serve as a
means of regulatory recordkeeping for the agency.
The proposed changes would include revising appendix A to 10 CFR
part 50 to clarify that exemptions are not required for deviations from
the GDCs. Instead, such deviations could be identified and justified
directly within the licensing application. While this clarification
would not change the NRC's review of the justification itself, it would
reduce regulatory burden for applicants proposing innovative designs in
which certain GDCs may be tailored to be better risk-informed.
In addition, the NRC is proposing to revise GDC 28 of appendix A to
10 CFR part 50 to allow applicants to propose and justify alternative
design basis accidents for reactivity control, rather than
prescriptively requiring evaluation of control rod ejection or control
rod drop accidents. Although applicants and licensees may be able to
accomplish this through the flexibilities in the proposed 10 CFR 50.220
and 52.220, this change would more explicitly enable the use of design-
specific accident scenarios and support removal of unnecessary
conservatism in safety analyses without the need to provide the
information that would be required in proposed 10 CFR 50.220(b) and
52.220. The NRC is issuing, for public comment along with this proposed
rule, DG-1464, ``Guidance for Content of Applications Under 10 CFR
50.220 and 52.220 Proposing Risk-Informed and Performance-Based
Alternative Acceptance Criteria,'' which would be used to support
applicants in determining appropriate design basis accidents based on
credible, realistic risks.
XXI. Background--Establishing Thresholds for Changes to Reactor Designs
During Construction and Operation Under 10 CFR Parts 52 and 53
Section 5(f) of E.O. 14300 directs the NRC to establish stringent
thresholds for circumstances in which the NRC may demand changes to a
reactor design once construction of the reactor is underway. In
response, the NRC is proposing to raise the threshold for changes
required during construction by eliminating ``increased
standardization'' as a criterion for Commission-directed modification
of design certification information on either a plant-specific or
generic basis. In addition, this proposed rule would provide additional
flexibility and reduce unnecessary regulatory burden in the regulations
governing licensee-requested changes during construction under 10 CFR
parts 52 and 53. The NRC is also proposing changes to 10 CFR parts 52
and 53 to provide flexibility and efficiencies for licensee-requested
changes during operation.
A. Development of Tiers of Information and Processes for Changes and
Departures in Design Certification Rules
In 1987, the NRC issued a policy statement on nuclear power plant
standardization (52 FR 34884; September 15, 1987). In 1989, 10 CFR part
52 was issued (54 FR 15372; April 18, 1989), which established that
design certification would be accomplished by rulemaking. The NRC
ultimately adopted a two-tiered system for design information. These
tiers are designated in a design control document (DCD), which the NRC
incorporates by reference into its regulations, and each information
tier is subject to a specified process for changes and departures from
design certification information.
Tier 1 information is the portion of the DCD that is approved and
certified. It includes definitions and general provisions; design
descriptions; inspections, tests, analyses, and acceptance criteria
(ITAAC); significant site parameters; and significant interface
requirements. Tier 1 design descriptions were intended to be applicable
for the life of the facility. Tier 2 and Tier 2* information is the
portion of the DCD that is approved but not certified. Tier 2 includes
information like that found in a final safety analysis report (FSAR)
for 10 CFR part 50 licenses. Tier 2* was created to minimize
information in Tier 1 while requiring that this information could not
be changed without prior NRC approval. If the Tier 2* designation were
not available, this information would have been designated Tier 1.
While the Tier 2* category was used in the first five design
certifications, from the U.S. Advanced Boiling Water Reactor (ABWR) to
the Economic Simplified Boiling-Water Reactor (ESBWR), the later
APR1400 and NuScale DCDs did not designate any information as Tier 2*.
Each design certification in appendix A, ``Design Certification
Rule for the U.S. Advanced Boiling Water Reactor,'' appendix D,
``Design Certification Rule for the AP1000 Design,'' appendix E,
``Design Certification Rule for the ESBWR Design,'' appendix F,
``Design Certification Rule for the APR1400 Design,'' and appendix G,
``Design Certification Rule for NuScale,'' of 10 CFR part 52 includes a
section VIII, ``Processes for Changes and Departures,'' that specifies
processes to change Tier 1, Tier 2, and, where applicable, Tier 2*
information. The requirements in section VIII of these appendices are
essentially identical except for certain certified designs that do not
have information designated as Tier 2*. Under these section VIII
requirements, plant-specific changes or departures from Tier 1 require
an exemption. Plant-specific departures from Tier 2* require NRC
approval by license amendment. Changes or departures from Tier 2
information are evaluated using a process like the one provided for 10
CFR part 50 licensees in 10 CFR 50.59, which provides criteria for
determining whether a licensee-initiated change requires prior NRC
approval. This similar departure process for Tier 2 information is,
therefore, often described as a 10 CFR 50.59-like process for COL
holders referencing a design certification. The change processes in
section VIII sought to balance standardization with flexibility and
apply during both construction and operation.
For currently certified designs, Tier 2* is defined in section II.F
of appendices A, D, and E to 10 CFR part 52. Appendices F and G to 10
CFR part 52 do not contain Tier 2* information. Tier 2* is the portion
of Tier 2 information designated with brackets, italicized text, and an
asterisk in the generic DCD. Per sections VIII.B.6.b and VIII.B.6.c of
appendices A, D, and E to 10 CFR part 52, any licensee who references
these appendices may not depart from Tier 2* matters without prior NRC
approval and any request for such a departure will be treated as a
request for a license amendment under 10 CFR 50.90. Sections VIII.B.6.b
and VIII.B.6.c of appendices A, D, and E to 10 CFR part 52 list all of
the Tier 2* matters for the respective designs. The Tier 2* matters
listed in section
[[Page 44583]]
VIII.B.6.b of appendices A, D, and E to 10 CFR part 52 retain this Tier
2* designation for the lifetime of the facility, while the Tier 2*
matters listed in section VIII.B.6.c of these appendices revert to Tier
2 after the plant first achieves full power.
Another key aspect of the standard design certification is
finality. Issuance of the design certification rule allows the design
to be incorporated by reference into a COL application. Issue finality
rules limit the types of changes that may be imposed on the
certification information. During the COL application review, design
information codified by rule is not subject to NRC review, but the NRC
staff would review applicant-requested departures from the certified
design that require NRC approval. The scope of a hearing for a COL
application that references a certified design does not include the
design certified by NRC rule but would encompass any departures from
the certified design requiring NRC approval.
The final rule promulgating 10 CFR part 53 established a risk-
informed, performance-based, and technology-inclusive regulatory
framework for commercial nuclear plants, including advanced reactor
designs. Part 53 of 10 CFR dispensed with the Tier 1 and Tier 2
terminology. Rather, 10 CFR 53.1525, ``Revising certification
information within a design certification rule,'' uses the term
``certification information'' in place of Tier 1. Information that is
``not certification information'' is equivalent to Tier 2 information
under 10 CFR part 52. The change control processes for ``certification
information'' and ``not certification information'' in 10 CFR part 53
are similar to those for Tier 1 and Tier 2 in 10 CFR part 52.
B. Licensing Experience and Improvement Initiatives Regarding
Information Designation and Change Processes for Design Certifications
The NRC has periodically considered improvements to the
effectiveness and efficiency of information designations for standard
design certifications and associated change processes. Many of these
improvements, developed through internal and external reviews and
interactions, are addressed in this proposed rule.
The NRC staff internally reviewed and considered improvements to
the content of Tier 2* and Tier 1 information. In SECY-17-0075,
``Planned Improvements in Design Certification Tiered Information
Designations,'' dated July 24, 2017, the NRC staff examined license
amendment requests (LARs) affecting Tier 2* information from AP1000 COL
licensees to assess the effectiveness of the Tier 2* designation and
the 10 CFR 50.59-like change process. While the NRC staff concluded
that there was a benefit to maintaining the use of Tier 2* information
in certified designs, it noted that for the AP1000 there were several
non-safety significant Tier 2* changes requested in LARs that probably
would not have triggered the 10 CFR 50.59-like criteria requiring prior
NRC approval of the change. In SECY-19-0034, ``Improving Design
Certification Content,'' dated April 8, 2019, the NRC staff refined the
general principles for Tier 1 content so that in future design
certifications, NRC approval would not be required for design changes
of minimal safety significance. These refinements also apply to Tier
2*.
Through the experience from Vogtle Units 3 and 4, the NRC staff has
gained insights into the Tier 2* information for the AP1000 design
referenced in appendix D to 10 CFR part 52. These insights were
summarized in the ``10 CFR part 52 Construction Lessons Learned
Report,'' issued January 16, 2024.
The AP1000 design certification contained a significant amount of
Tier 2* information, more than would be identified should the process
be repeated today. The construction experience at Vogtle Units 3 and 4
showed that, in some cases, LARs were needed to change Tier 2*
information that had minimal if any safety significance (e.g., to make
edits to a bibliography in a document that was identified as Tier 2*
information in its entirety). In these cases, submittal of an LAR to
change the information resulted in an inefficient use of resources for
both the licensee and the NRC. The use of Tier 2* designations should
be consistent with the approaches described in SECY-17-0075 and SECY-
19-0034. When this designation is used in future licensing
applications, the Tier 2* information should be carefully selected to
minimize the potential to require LARs for non-safety-significant
changes to this information.
Following the issuance of several 10 CFR part 52 COLs, the NRC
engaged with industry representatives and members of the public to
consider their perspectives while developing guidance for standardized
ITAAC and corresponding Tier 1 information. The NRC considered a draft
industry guideline developed by the Nuclear Energy Institute (NEI),
dated May 27, 2015, which discussed ``first principles'' for developing
Tier 1 information and a set of standard ITAAC that could be used in
future design certification applications. While discussions between the
NRC and the NEI continued, a final version of the guideline was never
endorsed.
The NRC issued COLs to Southern Nuclear Operating Company (SNC) for
Vogtle Units 3 and 4 on February 10, 2012. Vogtle Units 3 and 4 are
currently the only two nuclear plants to be licensed and constructed
and to enter into commercial operation using the 10 CFR part 52
licensing process. In the years following the issuance of the COLs, the
NRC and SNC had a series of interactions regarding exemptions related
to Tier 1 and Tier 2* information for Vogtle Units 3 and 4, including
the challenges, lessons learned, and experience gained during licensing
and construction.
During construction, SNC sought adjustments to the Tier 2* process
on several occasions. By letter dated August 7, 2014, SNC submitted an
amendment and exemption request to apply the existing departure
evaluation process for Tier 2 changes to Tier 2* changes. SNC withdrew
that request by letter dated December 15, 2014. Subsequently, on
February 1, 2016, the NRC granted amendment and exemption requests to
reclassify fire-protection related Tier 2* information as Tier 2. On
December 21, 2017, SNC submitted LAR-17-037, and an associated
exemption request, seeking changes to the Vogtle Units 3 and 4 COLs to
add a license condition that would apply the change process for Tier 2
information for a proposed departure from Tier 2* information provided
that specific criteria are not met. If one of the criteria were met for
the proposed departure, then the proposed departure would continue to
require prior NRC approval. In this request, SNC noted that LAR-17-037
``arises from SNC's nearly six years' experience with the departure
evaluation processes outlined in 10 CFR part 52, Appendix D'' and that
``SNC has identified an approach to alleviate some of the
administrative burden for both the NRC and the Licensee.'' In addition,
SNC stated that this proposal was intended to be in line with the Tier
2* lessons learned in SECY-17-0075. On September 20, 2018, the NRC
granted the amendments and exemptions to Vogtle Units 3 and 4 that
effectively replaced the Tier 2* change process with an alternative
process (reflected in a license condition) that applied the Tier 2
change process along with nine additional criteria to determine whether
a license amendment was required.
SNC submitted LARs, both during construction and after the start of
commercial operation, proposing to change the contents of Tier 2 and
Tier
[[Page 44584]]
2* information, and, where applicable, requested exemptions from Tier 1
information in accordance with the change processes specified in the
design certification appendix. During construction, there were nearly
200 license amendments, some with related exemptions, and various ASME
Code alternatives that required NRC review and approval. For each of
the numerous Tier 1 exemption requests approved for Vogtle Units 3 and
4, the NRC determined that the special circumstances outweighed any
decrease in safety from the reduction in standardization. Lessons
learned from these licensing activities indicate that the need for the
submittal and an evaluation of how standardization is maintained do not
result in significant insights that support the implementation of the
standardization policy; rather, experience shows that the requirement
for maintaining standardization as a criterion for allowing changes is
often burdensome to a licensee without significant benefit. The NRC has
not imposed these changes on the certification information in appendix
D of 10 CFR part 52 via rulemaking under 10 CFR 52.63, because these
deviations from standardization are not significant reductions in
safety. In addition, the NRC never found an exemption request submitted
by SNC to be unacceptable based solely on the criterion that the
special circumstances did not outweigh the decrease in safety from the
reduction in standardization. During construction, SNC also sought
efficiencies in the content of both Tier 1 and Tier 2* information.
Specifically, SNC requested an amendment to consolidate several ITAAC,
with corresponding changes to related Tier 1 information, in the plant-
specific DCD to improve the efficiency of the ITAAC completion and
closure process.
During a September 20, 2023, public meeting, SNC staff acknowledged
that the ``existing Tier 2* requirements are not overly burdensome, but
SNC would benefit from being able to implement a similar change process
across its fleet of plants.''
After the start of commercial operation, SNC submitted a license
amendment and exemption request dated July 25, 2024, for Vogtle Units 3
and 4 to remove all Tier 1 and Tier 2* information and associated
requirements from its license. SNC proposed, in part, to convert Tier 1
and Tier 2* information to Tier 2, so that the Tier 2 change process
would apply to all this information. At that time, the NRC communicated
to SNC that the request contained significant questions of policy more
appropriately addressed through the petition for rulemaking process and
that revisions to the licensing documents SNC proposed were significant
licensing changes that raised policy implications appropriate for
Commission consideration. Subsequently, SNC withdrew the license
amendment and exemption request on September 25, 2024. This proposed
rulemaking would address the issue raised by SNC's request to convert
Tier 1 and Tier 2* information to Tier 2.
On January 15, 2025, the NRC held another public meeting to seek
feedback on preliminary options to provide regulatory flexibility in 10
CFR part 52 during construction and operational phases, including a
change process for Tier 1 and Tier 2* information and the adjustment of
tier designations. Members of the public who spoke at the meeting
provided no adverse feedback and supported the NRC's approaches
presented at the meeting.
In its rulemaking efforts to align the 10 CFR parts 50 and 52
licensing processes, the NRC staff proposed various changes regarding
standardization and Tier 1 principles in SECY-22-0052, ``Proposed Rule:
Alignment of Licensing Processes and Lessons Learned from New Reactor
Licensing (RIN 3150-AI66),'' dated June 6, 2022. In SRM-SECY-22-0052,
``Staff Requirements--SECY-22-0052--Proposed Rule: Alignment of
Licensing Processes and Lessons Learned from New Reactor Licensing (RIN
3150-AI66),'' dated November 20, 2024, the Commission approved
publication of a revised proposed rule in the Federal Register that
would eliminate requirements for evaluating the impact on
standardization when approving departures from information in a design
certification or ML. This proposed rule would incorporate that change.
In SECY-22-0052, the NRC staff also proposed to add a definition in
10 CFR part 52 for Tier 1 information that would have defined Tier 1
information as the qualitative and functional level portion of the
design-related information in the generic DCD. This proposed rule also
incorporates this change and would further include ITAAC in the
definition of Tier 1. Relevant guidance documents would be updated to
clarify that Tier 1 information should not include detail that could
necessitate NRC approval for departures from certified designs that
have minimal safety significance.
This proposed rule would also revise regulations governing
licensee-requested changes during construction and operation under 10
CFR part 52 to provide additional flexibility and reduce unnecessary
regulatory burden. This would be accomplished through changes in each
of the certified designs described in appendices A, D, E, F, and G of
10 CFR part 52. Appendix B (System 80+) and appendix C (AP600) to 10
CFR part 52 have expired.
C. Severe Accidents
In SECY-12-0081, ``Risk-Informed Regulatory Framework for New
Reactors,'' dated June 6, 2012, the NRC staff described a potential
``gap'' in the Tier 2 change process regarding severe accident features
that are not related to ex-vessel severe accident prevention and
mitigation. Unless such non-ex-vessel severe accident design features
also happen to have a dual function such as also addressing design
basis accidents or aircraft impacts, risk-significant Tier 2 changes
(e.g., information in Chapter 19 of Tier 2 of the DCD and FSAR related
to prevention and mitigation of severe accidents other than those
considered ``ex-vessel'') could be screened out altogether and not
receive prior NRC approval. In SECY-12-0081, the NRC staff observed
that Tier 1 descriptions usually have sufficient detail that
necessitates prior NRC review for major changes to severe accident
design features. However, the NRC staff also noted that (1) changes may
be screened out or less appropriate criteria applied when determining
if prior NRC approval is needed and (2) whether prior NRC approval is
obtained may be highly dependent on the degree of detail in Tier 1, if
any. The current change process does not address all of the severe
accidents defined in 10 CFR 52.47(a)(23) and 52.79(a)(38). The current
regulation and its implementation through the guidance in NEI 96-07,
Appendix C, ``Guideline for Implementation of Change Processes for New
Nuclear Power Plants Licensed under 10 CFR part 52,'' Revision 0-
Corrected, issued March 2014, could result in the licensee screening
out changes in Chapter 19 (and other sections) of Tier 2 of the DCD and
FSAR that do not affect ex-vessel severe accident design features. In a
worst-case scenario, significant Tier 2 changes to non-ex-vessel severe
accident features, up to and including permanent removal from service,
could be made without prior NRC approval.
In SRM-SECY-12-0081, ``Staff Requirements--SECY-12-0081--Risk-
Informed Regulatory Framework for New Reactors,'' dated October 22,
2012, the Commission approved the NRC staff's plan to address the
potential gap in the Tier 2 change process by (a) ensuring that there
are sufficient details
[[Page 44585]]
on all key severe accident features in Tier 1 and (b) including a
change process in future design certification rulemaking for non-ex-
vessel severe accident features similar to the process for ex-vessel
severe accident features. While no current certified designs include a
change process for non-ex-vessel severe accident features, reverting
Tier 1 information to Tier 2 information once the plant first achieves
full power (a proposed change discussed in section XXII.B. of this
document) may introduce a gap regarding changes to non-ex-vessel severe
accident features. To address this potential gap, the proposed rule
would revise the 10 CFR 50.59-like process in section VIII of
appendices A, D, E, F, and G of 10 CFR part 52 to consider all severe
accidents and not just ex-vessel severe accidents.
XXII. Discussion--Establishing Thresholds for Changes to Reactor
Designs During Construction and Operation Under 10 CFR Parts 52 and 53
To respond to the direction in E.O. 14300 to ``establish stringent
thresholds'' for changes required during construction, this proposed
rule would revise the finality provisions for design certifications in
10 CFR 52.63, ``Finality of standard design certifications,'' and 10
CFR 53.1263, ``Finality of standard design certifications,'' to (1)
eliminate ``increased standardization'' as a criterion for modifying
design certification information on either a generic or plant-specific
basis, and (2) eliminate whether special circumstances outweigh any
decrease in safety that may result from a reduction in standardization
as considerations for plant-specific orders and licensee requests for
exemptions. Similarly, the proposed rule would revise finality
provisions for MLs in 10 CFR 52.171, ``Finality of manufacturing
licenses; information requests,'' and 10 CFR 53.1437, ``Exemptions,
departures, and variances,'' to eliminate whether ``special
circumstances outweigh any decrease in safety that may result from the
reduction in standardization'' as a consideration for requests for
departures from the applicant referencing or using the manufactured
reactor. Further, the proposed rule would include other rule changes to
provide additional flexibility and reduce unnecessary regulatory burden
for licensee-initiated changes.
A. Standardization
The NRC is proposing to amend its regulations to remove unnecessary
requirements to consider standardization as a criterion to justify
generic changes to certified designs, plant-specific orders regarding
information from a referenced standard design certification, or
requested departures from a standard design certification or ML.
As explained in section XXI.B. of this document, recent reactor
licensing experience with Vogtle Units 3 and 4 has shown that the
requirement for maintaining standardization as a criterion for allowing
requested changes is often burdensome to a licensee without significant
offsetting benefit. With regard to changes imposed by the NRC, 10 CFR
52.63(a)(1), 52.63(a)(4), 53.1263(a)(1), and 53.1263(a)(4) set out
criteria that must be met before the Commission modifies, rescinds, or
imposes new requirements on certification information by rulemaking or
plant-specific order. Sections 52.63(a)(1)(vii) and 53.1263(a)(1)(vii)
of 10 CFR, which provide that changes contributing to increased
standardization of certification information is an exception to
finality that the Commission may use as a basis to impose new
requirements on the certification information, would be deleted from
the NRC's regulations. Before imposing new requirements by plant-
specific order, 10 CFR 52.63(a)(4)(ii) and 53.1263(a)(4)(ii) require
that special circumstances be present and the Commission consider
whether these special circumstances outweigh any decrease in safety
that results from the effects of decreased standardization. The
criterion of whether the special circumstances outweigh any decrease in
safety that results from a reduction in standardization, which apply to
the Commission, would be deleted from the NRC's regulations. The
criterion that special circumstances are present would remain in the
regulations.
With regard to changes proposed by licensees and applicants, 10 CFR
52.63(b)(1) and 53.1263(b) permit an applicant or licensee that
references a design certification rule to request an exemption from one
or more elements of the certification information. These regulations
require the Commission to consider whether the special circumstances
that the exemption regulation requires to be present outweigh any
decrease in safety that may result from the reduction in
standardization caused by the exemption. However, as described in
section XXI.B. of this document, experience has shown that for all
cases, the required special circumstances outweigh any decrease in
safety that may result from the reduction in standardization caused by
the requested exemption. Therefore, the NRC is proposing to revise 10
CFR 52.63(b)(1) and 53.1263(b) to eliminate the requirements for the
NRC to review the impact of the requested exemption on standardization.
For similar reasons, the NRC is also proposing to revise 10 CFR
52.93(c), 52.171(b)(2), and 53.1437(c) to remove the requirement to
discuss the impact of the change on standardization as a criterion for
the justification for departures from ML information.
On September 15, 1987, the NRC issued a revised policy statement on
nuclear power plant standardization (52 FR 34884). The purpose of this
policy statement was for the Commission to encourage standardization by
providing a regulatory framework for the certification of nuclear power
plant designs that can be referenced in individual plant applications.
As stated in this policy, ``[t]he Commission believes that the use of
certified standardized designs can benefit the public health and safety
by concentrating resources on specific design approaches without
stifling ingenuity; by stimulating standardized programs of
construction practice, quality assurance, and personnel training; and
by fostering more effective maintenance and improved operation.'' The
use of certified designs would also improve the efficiency of NRC
reviews and reduce uncertainty in the regulatory process.
The proposed rule changes would not be contrary to this policy
goal. The policy goal of standardization along with the efficiencies
and safety benefits associated with it would be maintained because the
overall regulatory framework would still allow for certified designs to
be referenced by applicants and licensees in CP applications or COL
applications. The deletion of 10 CFR 52.63(a)(1)(vii) and
53.1263(a)(1)(vii) would ensure that changes to certified designs could
not be imposed on licensees for the sole purpose of establishing or
maintaining standardization among reactors referencing the same design.
However, the Commission could still impose new requirements on
certification information using one of the six other criteria that
would remain in 10 CFR 52.63(a)(1) and 53.1263(a)(1). Since the design
certification changes imposed by the NRC for reasons that meet these
other criteria would apply to all plants referencing the certified
design, standardization would still be achieved.
Regarding the proposed changes to the requirements in 10 CFR
52.63(b)(1), 52.93(c), 52.171(b)(2), 53.1263(b), and 53.1437(c)
requiring an analysis of the effects of a change on standardization,
[[Page 44586]]
experience has shown that it is challenging for an applicant or the NRC
to evaluate whether any one change proposed would decrease safety
solely as a result of a reduction in standardization. The NRC
recognizes that increased standardization remains a policy goal of 10
CFR parts 52 and 53, and requirements supporting that goal should be
maintained when there is an appropriate benefit to doing so. The
proposed changes to the NRC's regulations that remove the requirement
to justify requested changes to the design, based on the changes'
effects on standardization, would eliminate unnecessary burden on
applicants and licensees while maintaining beneficial aspects of the
NRC's policy on standardization. The proposed changes to 10 CFR
52.63(a)(4) and 53.1263(a)(4) would eliminate similar requirements for
the NRC to consider whether special circumstances outweigh any decrease
in safety that results from a reduction in standardization before
issuing a plant-specific order.
Even though the proposed rule revisions described in section
XXII.C, ``Processes for Changes and Departures,'' of this document
would provide licensees with additional flexibility to make plant-
specific changes to certified designs, the NRC does not expect the
scope and extent of plant-specific changes to result in such drastic
and significant differences that would negate the advantages and safety
benefits of standardization. The regulatory structure and change
control processes in 10 CFR parts 52 and 53 would continue to support
the policy goal of standardization.
B. Definitions
The NRC is proposing to change its regulations to add the
definitions of tier information to 10 CFR 52.1, ``Definitions,'' and to
make the definitions consistent with the principles in SECY-19-0034.
The terms ``Tier 1,'' ``Tier 2,'' and ``Tier 2*'' are defined in
section II, ``Definitions,'' of each design certification in appendices
A, D, E, F and G to 10 CFR part 52. A design certification applicant is
free to define this information. For example, an application can define
no tier information, include more than three tiers of information, or
define tiers with definitions that are different than those in current
10 CFR part 52 design certification appendices. This flexibility can
lead to inconsistencies and increased burden for design certification
and COL applicants in preparing applications and increased burden to
the NRC in reviewing applications. Recent reactor licensing experience
has shown that some applications have included more information in Tier
1 than is necessary for the purpose of Tier 1. This has resulted in the
need for licensees to request NRC review and approval of Tier 1
departures from information that, because of its minimal safety
significance, could more appropriately have been handled under the 10
CFR 50.59-like change process currently applicable to Tier 2
information.
To address these problems, the NRC is proposing to change its
regulations to add a definition of ``Tier 1'' to 10 CFR 52.1 that would
state that Tier 1 information is the qualitative and functional-level
portion of the design-related information in the generic DCD and also
includes ITAAC. Relevant guidance documents would be updated to clarify
that Tier 1 information should not include detail that could
necessitate NRC approval for departures from the certified design that
have minimal safety significance. The NRC proposes the same refinements
to the proposed 10 CFR 52.1 definition of Tier 2* information because
information should be designated as Tier 2* only if it qualifies for
inclusion in Tier 1.
The NRC also notes the following regarding the definitions of
``Tier 2'' and ``Tier 2*'' proposed to be added to 10 CFR 52.1:
The definition of ``Tier 2'' in section II of appendices
A, D, E, F and G to 10 CFR part 52 goes on to list specific information
that is included in Tier 2. The proposed definition of ``Tier 2'' in 10
CFR 52.1 would not contain this listed information.
The definition of ``Tier 2*'' in section II of appendices
A, D, E, F and G to 10 CFR part 52 states, ``This designation expires
for some Tier 2* information under paragraph VIII.B.6.'' The proposed
definition of ``Tier 2*'' in 10 CFR 52.1 would state, in part, that
after the plant first achieves full power, the Tier 2* designation
reverts to Tier 2 status for all Tier 2* matters.
These proposed 10 CFR 52.1 definitions of Tier 1, Tier 2, and Tier
2* would apply to design certifications issued after the effective date
of the final rule.
For currently certified designs, the NRC is proposing to clarify
its definitions of Tier 1 information in section II.D of appendices A,
D, E, F, and G to 10 CFR part 52. Currently, section II.D in each of
these appendices states that Tier 1 is the portion of the design-
related information in the generic DCD that is approved and certified
and includes: definitions and general provisions; design descriptions;
ITAAC; significant site parameters; and significant interface
requirements. The proposed revisions to section II.D would provide
further clarification by identifying specific sections and tables
within the generic DCD that correspond to each of these categories of
Tier 1 information. Since the proposed rule would implement different
change control processes for these various categories of Tier 1
information, there would be a need to clearly identify what Tier 1
information in the generic DCD corresponds to each category. For
definitions and general provisions (section II.D.1), significant site
parameters (section II.D.4), and significant interface requirements
(section II.D.5), the specific sections in the generic DCD would be
identified. For ITAAC (section II.D.3), the specific tables in the
generic DCD would be listed. The proposed rule would also specify that
only the inspections, tests, and analyses column and the acceptance
criteria column of these tables would be considered ITAAC information.
For design descriptions (section II.D.2), the proposed rule would
identify the specific sections in the generic DCD and would clarify
that figures and non-ITAAC tables referenced in these sections would
also be considered design description information. Non-ITAAC tables
would be any tables not specifically listed as an ITAAC table in the
definition.
C. Processes for Changes and Departures
(i) Tier 1 Design Description Information
The NRC is proposing to change its regulations regarding licensee-
requested changes during construction for COL holders referencing a
certified design. Several of these proposed rule changes would impact
the 10 CFR 50.59-like criteria in the appendices of 10 CFR part 52.
These include proposed changes to the 10 CFR 50.59-like criteria
themselves as well as their application to design information. The NRC
acknowledges that section XII, ``Discussion--Risk-Informing 10 CFR
50.59 and Allowing Flexibility for Changes to Methods,'' of this
document describes proposed changes to 10 CFR 50.59 that would allow
consideration of risk insights from PRAs and provide increased
flexibility for changes to methods of evaluation. As part of the
development of the final rule, the NRC will consider the rule revisions
to 10 CFR 50.59 and their applicability to the 10 CFR 50.59-like
process in 10 CFR part 52.
The NRC proposes to amend section VIII.A, ``Tier 1 Information,''
of
[[Page 44587]]
appendices A, D, E, F, and G to 10 CFR part 52 regarding licensee-
requested changes to Tier 1 design description information. Currently,
exemptions from all Tier 1 information are governed by the requirements
in 10 CFR 52.63(b)(1) and 52.98(f). The proposed rule would revise the
change process for Tier 1 design descriptions and provide a conforming
change in 10 CFR 52.63(b)(1) reflecting that some Tier 1 design
description changes within the scope of the new change process would
not require an exemption. Licensee-requested exemptions from
definitions and general provisions, significant site parameters, and
significant interface requirements would be unchanged and any departure
from this Tier 1 information would continue to require an exemption.
The NRC recognizes that some of the change requests for Tier 1
information based on construction experience from Vogtle Units 3 and 4
may not have been important to safety. Providing flexibility in
determining which changes to Tier 1 design description information
require NRC approval would reduce burden on licensees and minimize
possible construction delays due to licensing reviews without impacting
safety.
The NRC proposes to add section VIII.A.5 to appendices A, D, E, F,
and G to 10 CFR part 52 to allow an applicant or licensee to depart
from Tier 1 design description information without NRC approval if
certain criteria are met. The criteria for determining whether a
proposed departure from Tier 1 design description information requires
an exemption would be listed in the proposed new section VIII.A.6 of
appendices A, D, E, F, and G of 10 CFR part 52. These criteria are
based upon the 10 CFR 50.59-like criteria currently listed in sections
VIII.B.5.b and VIII.B.5.c in appendices A, D, E, F, and G of 10 CFR
part 52 for determining whether changes to Tier 2 information require
NRC approval. Since Tier 1 design description information is based on
Tier 2 information, these criteria would also be appropriate to use for
changes to Tier 1 design description information. The criteria in the
proposed new section VIII.A.6 would be identical to those in section
VIII.B.5.b. The criteria in the proposed new section VIII.A.6 would
also include criteria similar to those in current section VIII.B.5.c
but these criteria would be revised to consider all severe accident
design features and not just ex-vessel severe accident design features,
consistent with proposed changes to section VIII.B.5.c that would be
made by this proposed rule. Additional discussion for this change is in
section XXII.C.(iii), ``Severe Accidents,'' of this document. If any of
the criteria in the proposed section VIII.A.6 were met, then the
proposed departure from the Tier 1 design description would require an
exemption request for NRC approval.
Section VIII.A.2 of appendices A, D, E, F, and G to 10 CFR part 52
states that generic changes to Tier 1 information are applicable to all
applicants or licensees who reference the applicable appendix, except
those for which the change has been rendered technically irrelevant by
plant-specific actions under paragraphs VIII.A.3 or VIII.A.4. Because
the NRC proposes to add paragraphs VIII.A.5 and VIII.A.6 as means for
making plant-specific changes to Tier 1 design descriptions, the NRC
also proposes a conforming change that would amend section VIII.A.2 of
appendices A, D, E, F, and G to 10 CFR part 52 to add the new
paragraphs VIII.A.5 and VIII.A.6 to the list of plant-specific actions
that can render a generic change to Tier 1 information technically
irrelevant.
Section VI.B of appendices A, D, E, F, and G to 10 CFR part 52
lists matters the Commission considers resolved in subsequent
proceedings for issuance of a COL, amendment of a COL, or renewal of a
COL, proceedings held under 10 CFR 52.103, and enforcement proceedings
involving plants referencing the applicable appendix. Section VI.B.4 of
these appendices identifies one resolved matter as all exemptions under
and in compliance with the change processes in paragraphs VIII.A.4 and
VIII.B.4 of the applicable appendix. Section VI.B.6 of these appendices
identifies another resolved matter as all departures from Tier 2 under
and in compliance with the change processes in paragraph VIII.B.5 of
the applicable appendix that do not require prior NRC approval. Because
the NRC proposes to add paragraphs VIII.A.5 and VIII.A.6 as means for
making plant-specific changes to Tier 1 design descriptions, the NRC
also proposes a conforming change that would amend section VI.B.4 of
appendices A, D, E, F, and G to 10 CFR part 52 to add exemptions under
and in compliance with section VIII.A.6 to the list of resolved
matters. Another conforming change would amend section VI.B.6 of
appendices A, D, E, F, and G to 10 CFR part 52 to add departures from
Tier 1 under and in compliance with section VIII.A.5 that do not
require NRC approval to the list of resolved matters. In addition, the
NRC proposes to make a correction in section VI.B.6 of appendices A and
F to 10 CFR part 52 by changing the reference, ``paragraph
VIII.B.5.f,'' to ``paragraph VIII.B.5.g,'' of the applicable appendix.
Analogous changes are proposed for 10 CFR part 53. Specifically,
the NRC proposes to revise 10 CFR 53.1525(b) to allow a holder of a
license that references a design certification issued under 10 CFR part
53 to make changes to certification information that has not been
incorporated into the license without requesting an exemption if
certain criteria are met. Additional changes are proposed for 10 CFR
53.1535, ``Amendments and exemptions during construction,'' and 10 CFR
53.1550 to reflect the proposed change to 10 CFR 53.1525(b). Paragraph
(a) of 10 CFR 53.1535 would be revised to add provisions that holders
of a CP or LWA may also request an exemption, if an exemption is
required. Paragraph (b) of 10 CFR 53.1535 would be revised to require
any COL holders for which the 10 CFR 53.1452(g) finding has not yet
been made to submit an exemption request within 45 days from the date
the licensee begins the construction to implement a change requiring
NRC approval. This requirement currently applies to requested license
amendments, and the NRC is proposing to also apply it to exemptions to
reflect the proposed change to 10 CFR 53.1525(b) under which certain
departures from certification information would require an exemption
but not an amendment. The proposed changes to 10 CFR 53.1550 would
clarify that the evaluation of changes to the facility applies to
certification information as well as FSARs. Finally, the NRC proposes
to delete the current requirement in 10 CFR 53.1525(b) that a request
for an exemption be included with a LAR since departures under proposed
10 CFR 53.1525(b) that require NRC approval would be subject to
exemption requests and not require a license amendment unless the
departure would change the license itself. A conforming change is
proposed to 10 CFR 53.1530, ``Revising information within a Final
Safety Analysis Report associated with a manufacturing license,'' to
eliminate the requirement that a holder of an ML referencing a design
certification request an exemption from the design certification rule
as part of an amendment application. The proposed 10 CFR 53.1530 would
state that, in these cases, the provisions of 10 CFR 53.1525 would
apply.
(ii) Tier 1 ITAAC and Certification Information
The NRC proposes to amend section VIII.A of appendices A, D, E, F,
and G to 10 CFR part 52 regarding licensee-
[[Page 44588]]
requested changes to Tier 1 ITAAC information. Under the current change
control requirements, any change to generic ITAAC in a COL application
referencing a certified design requires the COL holder to submit both
an exemption request and a LAR to the NRC for approval. An exemption
request is required to modify generic ITAAC because, like other Tier 1
information, ITAAC are certified information. Changes to either generic
or plant-specific ITAAC also require a license amendment because
section 185b. of the AEA requires the Commission to include the ITAAC
within the COL.
The proposed rule would add a new section VIII.A.7 to appendices A,
D, E, F, and G to 10 CFR part 52. Proposed section VIII.A.7 in each
appendix would require licensees who reference that appendix to request
a license amendment under 10 CFR 50.90 for any change to Tier 1 ITAAC
information. However, unlike current requirements, licensees requesting
a departure from Tier 1 ITAAC information would no longer be required
to submit an exemption request in addition to the LAR due to the
proposed requirements in section VIII.A.7 of appendices A, D, E, F, and
G to 10 CFR part 52. Removing the requirement to submit an exemption
request would reduce the unnecessary burden on licensees by eliminating
the need to submit both an exemption request and a LAR for the same
change. The NRC would still be required to review and approve any ITAAC
changes to ensure that the requirements of 10 CFR 52.97(b)--namely,
that the COL contains the ITAAC that 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
AEA, and the Commission's regulations--continue to be met. Experience
from construction of Vogtle Units 3 and 4 demonstrated that there is no
enhanced assurance of safety by evaluating these same ITAAC changes
against the exemption criteria specified in 10 CFR 52.7 (which
reference 10 CFR 50.12), 10 CFR 52.63(b), and section VIII.A.4 of the
applicable appendix to 10 CFR part 52. The need for the exemption
request for ITAAC information was simply due to the designation of
ITAAC as Tier 1 (e.g., certified) information.
Section VIII.A.2 of appendices A, D, E, F, and G to 10 CFR part 52
states that generic changes to Tier 1 information are applicable to all
applicants or licensees who reference the applicable appendix, except
those for which the change has been rendered technically irrelevant by
plant-specific actions under paragraphs VIII.A.3 or VIII.A.4. In a
conforming change, the NRC proposes to amend section VIII.A.2 of
appendices A, D, E, F, and G to 10 CFR part 52 to add the new paragraph
VIII.A.7 to the list of plant-specific actions that can render a
generic change to Tier 1 information technically irrelevant.
Analogous changes are proposed for 10 CFR part 53. Specifically,
the NRC proposes to amend 10 CFR 53.1525(a) to allow a holder of an OL
or COL that references a design certification issued under 10 CFR part
53 to request a license amendment, in lieu of an exemption, if
proposing changes to certification information that has been
incorporated into the license. A conforming change is also proposed to
10 CFR 53.1550(a)(1) to clarify that changes could be made without a
license amendment if the change to certification information
incorporated into the license is not required.
The requirements for licensee-requested changes to Tier 1
information in the appendices of 10 CFR part 52 are not identical to
those for certified information in 10 CFR part 53 because the Tier 1
definition in the appendices of 10 CFR part 52 states that Tier 1
information includes definitions and general provisions; design
descriptions; ITAAC; significant site parameters; and significant
interface requirements, whereas 10 CFR part 53 does not specify
certification information. The 10 CFR part 52 change control process
for Tier 1 ITAAC is similar to that for 10 CFR part 53 certified
information that has been incorporated into the license. The 10 CFR
part 52 change control process for Tier 1 design descriptions is
similar to that for 10 CFR part 53 certified information that has not
been incorporated into the license. The NRC proposes to amend section
VIII.A.4 of the appendices of 10 CFR part 52 to state that licensee-
requested departures from definitions and general provisions,
significant site parameters, and significant interface requirements
would continue to require an exemption from Tier 1 information. 10 CFR
part 53 does not and would not have a similar requirement because it
does not define this type of information. Despite this difference, the
proposed requirements in 10 CFR parts 52 and 53 would establish
appropriate thresholds for licensee-requested changes. During a review
of a design certification under 10 CFR part 53, if the NRC determined
that certain certification information always warranted prior NRC
approval, this information could be designated as information required
to be incorporated into a license referencing the certified design.
(iii) Severe Accidents
The NRC proposes to amend the criteria described in section
VIII.B.5.c of appendices A, D, E, F, and G to 10 CFR part 52 used to
determine if a proposed departure from Tier 2 information affecting the
resolution of an ex-vessel severe accident design feature identified in
the plant-specific DCD requires a license amendment.
The proposed rule would address the gap described in SECY-12-0081
and section XXI.C. of this document by revising section VIII.B.5.c of
appendices A, D, E, F, and G to 10 CFR part 52 by removing the phrase
``ex-vessel'' from the description of severe accidents considered. This
change would allow the revised criteria to consider all severe accident
design features identified in the DCD, not just ex-vessel severe
accident design features.
As discussed in section XXII.C.(i) of this document, another change
proposed in this rulemaking would be to add a new section VIII.A.5 to
appendices A, D, E, F, and G to 10 CFR part 52. Proposed section
VIII.A.5 would allow an applicant or licensee to depart from Tier 1
design description information without NRC approval if certain criteria
are met. The criteria for determining whether a proposed departure from
Tier 1 design description information requires an exemption would be
listed in the proposed section VIII.A.6. These criteria were based upon
the 10 CFR 50.59-like criteria currently listed in sections VIII.B.5.b
and VIII.B.5.c of appendices A, D, E, F, and G of 10 CFR part 52 for
determining whether changes to Tier 2 information require NRC approval.
The criteria in the proposed section VIII.A.6 would also include
criteria similar to those in current section VIII.B.5.c but these
criteria would be revised to consider all severe accident design
features and not just ex-vessel severe accident design features.
(iv) Process for Changes and Departures During Commercial Operation
The NRC proposes to amend section VIII.A in appendices A, D, E, F,
and G to 10 CFR part 52 regarding the treatment of Tier 1 information
during commercial operation. A proposed section VIII.A.8 would be added
to each appendix stating that after the plant first achieves full
power, licensee-initiated plant-specific departures from Tier 1
information would be subject to the same requirements as licensee-
initiated plant-specific departures from Tier 2 information. Per 10 CFR
52.103(h), after
[[Page 44589]]
the Commission has made the 10 CFR 52.103(g) finding, ITAAC do not
constitute regulatory requirements except for any specific ITAAC for
which the Commission has granted a hearing under 10 CFR 52.103(a).
Also, all ITAAC expire upon final Commission action in the proceeding
under 10 CFR 52.103(a). Therefore, after the 10 CFR 52.103(g) finding
and any proceedings held under 10 CFR 52.103(a), the only remaining
Tier 1 requirements are the definitions and general provisions, design
descriptions, significant site parameters, and significant interface
requirements. As discussed in section XXII.C.(i) of this document,
proposed section VIII.A.5 would allow an applicant or licensee, even
during construction, to depart from Tier 1 design description
information without NRC approval if certain criteria are met. Thus, the
criteria for determining whether a proposed departure from Tier 1
design description information would require an exemption would be the
10 CFR 50.59-like criteria that apply to departures from Tier 2
information. With the NRC's additional proposal to treat licensee-
initiated departures from Tier 1 the same as licensee-initiated
departures from Tier 2 after first achieving full power, an exemption
would be needed for a plant-specific departure from Tier 1 design
description information during construction and start-up testing up to
full power if the 10 CFR 50.59-like criteria are met, but a license
amendment would be needed for such departures after full power
operation is first achieved. The criteria for determining whether the
change would need NRC approval would remain the same.
Since plant-specific departures from Tier 2 information are
governed by the 10 CFR 50.59-like criteria, the proposed section
VIII.A.8 would allow licensees, after first achieving full power, to
use the 10 CFR 50.59-like process to determine if changes to
definitions and general provisions, design descriptions, significant
site parameters, and significant interface requirements would require
prior NRC approval. Tier 1 definitions and general provisions, design
descriptions, significant site parameters, and significant interface
requirements would still remain requirements for the lifetime of the
facility, but the proposed rule would provide additional flexibility to
10 CFR part 52 licensees by allowing them to use the 10 CFR 50.59-like
process to determine if changes to this information would require prior
NRC approval.
(v) Tier 2* Information
The NRC proposes to amend sections II.F and VIII.B of appendices A
and E to 10 CFR part 52 to revert all Tier 2* information to Tier 2
status after the plant first achieves full power. Proposed amendments
to section II.F of appendices A and E to 10 CFR part 52 would clarify
that the Tier 2* designation expires for all Tier 2* matters, not just
for those listed in section VIII.B.6.c. In addition, the proposed rule
would modify section VIII.B.6.b of appendices A and E to 10 CFR part 52
to add the Tier 2* matters listed in section VIII.B.6.c and
correspondingly remove those items from section VIII.B.6.c. Finally,
section VIII.B.6.c of appendices A and E to 10 CFR part 52 would be
revised to state that after the plant first achieves full power, all
Tier 2* matters would revert to Tier 2 status and would thereafter be
subject to the departure provisions in section VIII.B.5. These proposed
changes would reduce the regulatory burden during operation for COL
holders who reference appendix A or E to 10 CFR part 52 by allowing
them to use the 10 CFR 50.59-like criteria to determine if a departure
from Tier 2* information would require a LAR. This would also provide
COL licensees under 10 CFR part 52 the same flexibility afforded to OL
licensees under 10 CFR part 50 during plant operation.
The NRC also proposes to amend sections II.F and VIII.B of appendix
D to 10 CFR part 52 regarding Tier 2* information. These changes would
allow the use of the 10 CFR 50.59-like criteria in sections VIII.B.5.b
and B.5.c to determine if licensees who reference appendix D to 10 CFR
part 52 may depart from Tier 2* information without prior NRC approval.
The proposed rule would revise sections VIII.B.5.a, VIII.B.5.b, and
VIII.B.5.c to apply to both Tier 2* information as well as Tier 2
information. Section VIII.B.6 would be deleted in its entirety, along
with cross-references to section VIII.B.6 elsewhere in appendix D to 10
CFR part 52. With these proposed changes, the Tier 2* information for
the AP1000 design would effectively be treated as Tier 2 information
during construction as well as operation. The bases for these proposed
rule changes are as follows. First, SNC's experience with the Tier 2*
departure evaluation process for Vogtle 3 and 4 during the first six
years of construction demonstrated that the Tier 2* change process for
the AP1000 design imposed an administrative burden to both the licensee
and the NRC because LARs were needed to change Tier 2* information that
had minimal safety significance. This was also documented in SECY-17-
0075 and the ``10 CFR part 52 Construction Lessons Learned Report,''
issued after the completion of construction of Vogtle units 3 and 4.
Second, the 10 CFR 59.59-like process is an acceptable approach to
determine which changes are safety significant and require prior NRC
approval. The 10 CFR 50.59 process is used successfully by the
operating fleet of plants licensed under 10 CFR part 50. In addition,
as part of this proposed rule, the NRC is proposing to apply the 10 CFR
59.59-like process to the Tier 1 design description information for
existing certified designs.
Section VI.B of appendix D to 10 CFR part 52 lists matters the
Commission considers resolved in subsequent proceedings for issuance of
a COL, amendment of a COL, or renewal of a COL, proceedings held under
10 CFR 52.103, and enforcement proceedings involving plants referencing
appendix D. Section VI.B.6 of appendix D identifies one resolved matter
as all departures from Tier 2 under and in compliance with the change
processes in paragraph VIII.B.5 of appendix D that do not require prior
NRC approval. Because the NRC proposes to amend section VIII.B of
appendix D to 10 CFR part 52 to apply the change process in paragraph
VIII.B.5 to Tier 2* information, the NRC also proposes a conforming
change that would amend section VI.B.6 of appendix D to 10 CFR part 52
to add departures from Tier 2* under and in compliance with section
VIII.B.5 that do not require NRC approval to the list of resolved
matters.
At this time, the NRC is not proposing that Tier 2* information for
the ABWR and ESBWR designs be treated as Tier 2 information during
construction. The NRC acknowledges some inconsistency in the treatment
of Tier 1 design description and Tier 2* information but wants to
provide considerations for the construction and licensing experience
from COLs referencing the AP1000. However, the NRC has posed a specific
question in section XXXVI, ``Specific Questions,'' of this document
asking if the flexibility afforded the AP1000 design should include the
ABWR and ESBWR designs and for the basis of a response to that
question.
Section VIII.B.2 of appendix A to 10 CFR part 52 states, in part,
that generic changes to Tier 2* information are applicable to all
applicants or licensees who reference that appendix, except those for
which the change has been rendered technically irrelevant by certain
plant-specific actions. The NRC proposes to amend section VIII.B.2 of
appendix A to 10 CFR part 52 to add
[[Page 44590]]
paragraph VIII.B.6 to the list of plant-specific actions that can
render a generic change to Tier 2* information technically irrelevant.
This proposed addition would make appendix A to 10 CFR part 52 similar
in this regard to other appendices in 10 CFR part 52 and avoid an
unnecessary imposition of a generic change to a plant for which the
underlying issue has already been addressed.
(vi) Section IX of Appendix D to 10 CFR Part 52
The NRC proposes to delete the requirements in section IX,
``Inspections, Tests, Analyses, and Acceptance Criteria (ITAAC),'' of
appendix D to 10 CFR part 52 and reserve this section for future use.
The ITAAC requirements currently listed in this section were
incorporated into 10 CFR 52.99 and 52.103 in a 2007 rulemaking (72 FR
49352; August 28, 2007). Therefore, the language in this section is no
longer needed, and removal of it is consistent with previous Commission
direction. As stated in the rulemaking for the ESBWR design
certification (79 FR 61944; October 15, 2014), ``The language of the
ESBWR design certification rule differs from the rule language of other
DCRs in two substantive areas. First, paragraph IX was reserved for
future use because the substantive requirements in this paragraph (for
other DCRs) has since been incorporated into 10 CFR part 52 in a 2007
rulemaking (72 FR 49352; August 28, 2007) and thus are no longer needed
in the four existing DCR appendices. The NRC intends to remove these
requirements from Section IX of the four existing DCR appendices in
future amendment(s) separate from this rulemaking.''
This change would also make appendix D to 10 CFR part 52 similar in
this regard to the other appendices in 10 CFR part 52.
(vii) Manufacturing Licenses
The NRC proposes to revise 10 CFR 52.171(b)(1) to allow the holder
of an ML to use the regulations in 10 CFR 50.59 to determine whether
changes to the facility or procedures as described in the FSAR would
require prior Commission approval of an amendment to the ML. If prior
Commission approval is required, then the change would need to be
submitted in the form of a license amendment per 10 CFR 50.90, 50.91,
and 50.92. The NRC also proposes a conforming change to 10 CFR 50.71(f)
regarding FSAR updates for ML holders. Instead of requiring FSAR
updates that reflect only design modifications approved by the
Commission, the revised 10 CFR 50.71(f) would require that FSAR updates
reflect safety analyses and evaluations that support approved
amendments to the ML or support conclusions that changes did not
require a license amendment. This revision would be consistent with the
requirements in 10 CFR 53.1530 and would provide additional flexibility
to licensees and reduce unnecessary regulatory burden in the
regulations governing licensee-requested changes.
(viii) Applicants and Licensees That Reference Manufacturing Licenses
The NRC proposes to revise its regulations to allow licensees who
reference an ML license to use the applicable change processes in 10
CFR part 50 to determine whether changes to the facility or procedures
as described in the FSAR would require prior Commission approval. This
would be accomplished with two proposed revisions to 10 CFR 52.98. The
first proposed change would be to delete 10 CFR 52.98(d), which
establishes requirements for changes or departures for COLs that
reference an ML under subpart F of 10 CFR part 52. The second proposed
change would be to revise 10 CFR 52.98(b). The current requirement in
10 CFR 52.98(b) is that only COLs that do not reference either a design
certification or an ML may make changes to the facility using the
applicable 10 CFR part 50 change processes. The NRC proposes to
eliminate the restriction for COLs that reference a reactor
manufactured under an ML by deleting the phrase ``or a reactor
manufactured under a manufacturing license issued under subpart F of
this part.'' The NRC also proposes to modify 10 CFR 52.171(b)(2) so
that it would not apply to licensees referencing an ML. These proposed
changes would allow a COL that references an ML to make changes to the
facility using the applicable 10 CFR part 50 change processes.
The NRC proposes to add a new 10 CFR 50.59(f) that would allow the
holder of an OL or COL that references a reactor manufactured under an
ML to make changes in the facility or procedures as described in the
FSAR without requesting a license amendment if the changes would be the
same as changes approved by amendment to the ML and upon a
determination that implementing the changes would be consistent with
the basis for the Commission's approval of the amendment to the ML and
would not involve any additional changes that would require an
amendment to the OL or COL. The NRC proposes to add a similar provision
as 10 CFR 53.1550(c) for holders of OLs or COLs that reference an ML.
These proposed requirements would prevent OL and COL holders and the
NRC from having to duplicate the amendment process for each
manufactured reactor.
Sections 50.59(d)(1) and 53.1550(d) of 10 CFR would also be revised
to require that the licensee maintain records of these changes.
The NRC proposes to amend 10 CFR 52.93(c), 52.171(b)(2),
53.1288(b), and 53.1437(c) to eliminate requirements that the
Commission determine that departures from MLs by an applicant
referencing the ML comply with the requirements for specific
exemptions. Applicants that reference an ML would still request
departures from the design characteristics, site parameters, terms and
conditions, or approved design of the manufactured reactor. The NRC
would review these proposed departures as part of its review of the
application. Removing the requirement that approval of these departures
also meets the requirements for exemptions would eliminate an
unnecessary regulatory burden for applicants.
All of these proposed revisions would provide additional
flexibility to licensees and applicants and reduce unnecessary
regulatory burden in the regulations governing licensee-requested
changes.
XXIII. Background--Revision of the Emergency Preparedness Regulations
for Nuclear Power Reactors
A. Existing Emergency Preparedness Frameworks for Nuclear Power
Reactors
Before December 18, 2023, appendix E, ``Emergency Planning and
Preparedness for Production and Utilization Facilities,'' to 10 CFR
part 50 identified the minimum requirements for emergency plans.
Additionally, the regulations in 10 CFR 50.47, ``Emergency plans,''
provided emergency preparedness (EP) requirements for nuclear power
reactors, including planning standards for onsite and offsite emergency
response plans. Other relevant regulations included paragraphs (q),
(s), and (t) of 10 CFR 50.54, ``Conditions of licenses.''
Efforts to develop a performance-based approach for EP have been
ongoing for over two decades. In SECY-06-0200, ``Results of the Review
of Emergency Preparedness Regulations and Guidance,'' dated September
20, 2006, the staff sought Commission approval to begin activities to
develop a new voluntary performance-based EP regulatory regimen. On
November 16, 2023, the NRC published a final rule creating 10 CFR
50.160, ``Emergency
[[Page 44591]]
preparedness for small modular reactors, non-light-water reactors, and
non-power production or utilization facilities'' (88 FR 80050)
(referred to herein as the ``2023 EP final rule''). Section 50.160 of
10 CFR provides a performance-based, technology-inclusive, risk-
informed, and consequence-oriented EP framework as an alternative to 10
CFR 50.47 and appendix E to 10 CFR part 50. The alternative EP
framework recognizes advances in reactor design and technology and
credits the potential benefits of smaller sized reactors and non-light-
water reactors (non-LWRs) associated with postulated accidents,
including slower transient response times, and relatively small and
slow release of fission products, as compared to large LWRs. The
regulations in 10 CFR 50.160 are applicable to small modular reactors
(SMRs) of rated power less than 1000 megawatts thermal (MWt), non-LWRs,
and certain non-power production or utilization facilities. The NRC did
not include large LWRs in the scope of 10 CFR 50.160 because an EP
licensing framework already existed for those reactors, and licensees
for those plants had not expressed a clear interest in changing that
framework during development of the 2023 EP final rule. However, the
work underpinning 10 CFR 50.160, originating from the motivations of
SECY-06-0200, included considerations of large LWRs. Consistent with
Commission direction in SRM-SECY-14-0038, ``Staff Requirements--SECY-
14-0038--Performance-Based Framework for Nuclear Power Plant Emergency
Preparedness Oversight,'' dated August 2, 2015, which directed the
staff to ``be vigilant in continuing to assess the NRC's emergency
preparedness program and should not rule out the possibility of moving
to a performance-based framework in the future,'' the changes in this
proposed rulemaking would expand the voluntary applicability of 10 CFR
50.160 to large LWRs. Part 53 of 10 CFR provided another opportunity to
increase the use of performance-based regulation. The regulations in 10
CFR 53.855, ``Emergency preparedness,'' require each holder of an OL or
COL under 10 CFR part 53 to have an emergency plan that complies with
either the requirements in 10 CFR 50.160 or the requirements in
appendix E to 10 CFR part 50 and the planning standards of 10 CFR
50.47(b).
B. Protective Actions and Emergency Planning Zones
EP is an operational safety program that provides reasonable
assurance that adequate protective measures can and will be taken in
the unlikely event of a radiological emergency. For radiological
emergencies, protective actions should be carefully planned to balance
protection with other important factors and ensure that actions result
in more benefit than harm. The use of precautionary protective action
strategies that rely on predetermined, prompt protective measures,
including prompt evacuation, should be reserved for only the most
severe incidents as protective actions are not without risk (see, for
example, NUREG/CR-7285, ``Nonradiological Health Consequences from
Evacuation and Relocation,'' issued September 2021). To balance these
risks, the NRC applies a graded approach to EP in which the
requirements and criteria are based on the relative radiological risks
and hazards of the facility, among other considerations. The most
detailed level of planning is associated with the implementation of
predetermined, prompt protective actions and is reserved for only the
most severe events. For less severe events, the level of planning can
be scaled commensurately.
The emergency planning zone (EPZ) is a planning tool for
implementing predetermined, prompt protective actions. It simplifies
decision-making, particularly when such decisions may be time-
constrained or require coordination across a large area involving
multiple jurisdictions. Currently, protective actions within the EPZ
are initiated promptly at the declaration of a General Emergency (i.e.,
the highest emergency classification level that indicates that events
are occurring or have occurred at the facility with the potential for
an offsite release) as a precaution. However, a precautionary approach
is not the only strategy that can be used for implementing protective
measures. Risk-informed strategies, which make use of the best
available information in a robust, transparent, and repeatable process
to arrive at a decision can also be used to respond to radiological
emergencies. Risk-informed protection strategies are particularly
useful to reduce or avoid the risk of stochastic effects from radiation
or when there is ample time to make a decision based on the actual
conditions associated with the emergency event.
Under 10 CFR 50.33(g), 50.47(c)(2), and 53.1109(g), the plume
exposure pathway EPZ for a nuclear power reactor consists of an area
about 10 miles (16 km) in radius and the ingestion pathway EPZ for such
facilities consists of an area about 50 miles (80 km) in radius. These
regulations also provide that the size of plume exposure pathway and
ingestion pathway EPZs for gas-cooled nuclear reactors and for reactors
with an authorized power level less than 250 MWt may be determined on a
``case-by-case basis.''
For small modular reactors, non-LWRs, and other non-power
production or utilization facilities, the size of the EPZ can be
determined on a case-by-case basis under 10 CFR 50.33(g)(2).
Specifically, 10 CFR 50.33(g)(2)(i) provides two criteria for
determining whether an EPZ is needed, and if so, the size of the EPZ.
The first criterion, located in 10 CFR 50.33(g)(2)(i)(A), is that the
plume exposure pathway EPZ is the area within which public dose, as
defined in 10 CFR 20.1003, ``Definitions,'' is projected to exceed 1
rem (10 millisieverts (mSv)) 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
second criterion, located in 10 CFR 50.33(g)(2)(i)(B), is that the
plume exposure pathway EPZ is the area in which predetermined, prompt
protective measures are necessary. These criteria were added in the
2023 EP final rule and were based, in part, on the methodology
described in NUREG-0396, ``Planning Basis for the Development of State
and Local Government Radiological Emergency Response Plans in Support
of Light Water Nuclear Power Plants,'' dated December 1978. Similar
provisions for a case-by-case EPZ determination are contained in 10 CFR
53.1109(g)(2).
The EPZ is one element of the emergency planning basis, which is
operational in nature. The EPZ is not a design feature of the reactor
or a part of the ``design bases'' as defined in 10 CFR 50.2 (i.e., the
EPZ is not a plant structure, system, or component). The purpose of the
EPZ analysis required by 10 CFR 50.33(g)(2) is to arrive at a robust
and resilient protection strategy for managing radiological
emergencies. The EPZs are scalable in size and commensurate with the
planning needs for the facility. However, the EPZ size does not change
the requirements for emergency planning; it only sets bounds on the
planning for a predetermined, prompt response. The capabilities within
the emergency plan can support expanding the response beyond the EPZ
during an actual emergency, should that prove necessary.
When an EPZ extends beyond the site boundary, the NRC requires
additional findings and determinations on the adequacy of offsite plans
to implement a prompt response. In such cases, the
[[Page 44592]]
NRC considers Federal Emergency Management Agency (FEMA) findings and
determinations on the adequacy of the offsite planning in its overall
reasonable assurance determination. FEMA maintains a voluntary offsite
radiological emergency preparedness (REP) program to administer EP for
the areas surrounding commercial nuclear power plants. The EP rule
changes in this proposed rule would not impact the ability of State,
local, and Tribal governments to receive support from FEMA or any other
Federal agency regardless of whether the NRC requires reasonable
assurance of the offsite plans. For current facilities without offsite
EPZs (e.g., non-power reactors, decommissioning power reactors,
independent spent fuel storage installations), the NRC does not require
findings and determinations of the adequacy of the offsite plans as the
risks from radiation exposure are manageable under comprehensive
emergency management plans and local emergency response. Emergency
plans for these types of facilities provide reasonable assurance that
adequate protective measures can and will be taken in coordination with
offsite response organizations.
XXIV. Discussion--Revision of the Emergency Preparedness Regulations
for Nuclear Power Reactors
A. Emergency Plan Licensing Flexibility
(i) Flexible EP Licensing Paths
The proposed rule would provide all applicants and licensees under
10 CFR parts 50, 52, and 53, including power reactor applicants and
licensees, with the option to use the performance-based requirements in
10 CFR 50.160. Specifically, the proposed rule would remove the
provision in the current rule limiting the applicability of 10 CFR
50.160 to only small modular reactors, non-LWRs, and non-power
production or utilization facilities. This proposed change would reduce
the need for exemptions to use the performance-based EP regulations.
For example, a reactor would not be limited in power to 1000 MWt to
meet the definition of ``small modular reactor'' as defined in 10 CFR
50.2 and utilize 10 CFR 50.160. In addition, the proposed rule would
revise paragraph I.5 of appendix E to 10 CFR part 50 to provide for a
scalable approach for power reactors that choose to comply with the
planning standards of 10 CFR 50.47(b). The proposed changes would
ensure applicants are not restricted in licensing options for EP and
would provide the appropriate regulatory flexibility to support the
various licensing pathways.
The planning standards in 10 CFR 50.47(b) and the requirements in
appendix E to 10 CFR part 50 were originally developed for the hazards
and emergency planning needs of large LWRs. Additionally, the
prescriptive planning elements of appendix E to 10 CFR part 50 can be
applied to non-power production or utilization facilities on a case-by-
case basis. Basic emergency planning functions are similar across NRC
licensed facilities, as the operational aspects of EP are well
established in regulation and in practice across all hazards. There is
no technology-specific language used in the planning standards for
required EP functions that limit application to a particular
technology. Therefore, the proposed rule would revise the section
heading of 10 CFR 50.160 from ``Emergency preparedness for small
modular reactors, non-light-water reactors, and non-power production or
utilization facilities,'' to ``Performance-based emergency preparedness
standards,'' to be applicable to all reactor types. In addition, the
proposed rule would amend 10 CFR 50.33(g), 50.34(b)(6), 50.54(q)(2),
50.54(q)(3), 52.79(a)(21), 53.1109(g), and 53.1565(d)(3) to remove the
distinction among applicants and licensees of small modular reactors,
non-LWRs, and non-power production or utilization facilities in
complying with 10 CFR 50.160. The proposed rule would not change the
definition of ``small modular reactor'' in 10 CFR 50.2 and 53.020
because the definition would no longer have specific relevance to EP
regulations.
The proposed rule would amend paragraph I.5 of appendix E to 10 CFR
part 50 to determine the degree to which compliance with the
requirements in certain sections of appendix E is necessary on a case-
by-case basis for power reactors with a site-boundary EPZ or no EPZ.
The proposed revision would reduce the need for exemptions and would
provide applicants with the flexibility to propose the planning
elements appropriate to their facility. For licensing efficiency, the
NRC and specific applicants could achieve agreement on applicable
planning elements in pre-application interactions. Alternatively,
generic guidelines could be developed and endorsed by the NRC to
support the rapid deployment of similar reactor types that wish to
apply appendix E to 10 CFR part 50 in lieu of 10 CFR 50.160.
The proposed rule would add new 10 CFR 50.47(h) to allow a licensee
to submit a license amendment to comply with the requirements of 10 CFR
50.160 in lieu of 10 CFR 50.47 and appendix E to 10 CFR part 50.
Similarly, the proposed rule would add new 10 CFR 50.160(c)(4) to allow
a licensee to submit a license amendment to comply with the
requirements of appendix E to 10 CFR part 50 and, for nuclear power
reactor licensees, the requirements of 10 CFR 50.47 in lieu of 10 CFR
50.160. The proposed rule would also add new 10 CFR 53.855(d) to
provide provisions for complying with alternative EP requirements.
(ii) Preoperational Exercises
The NRC is proposing to remove specific time requirements and add
flexibility in the performance of preoperational exercises. Prior to
initial loading of fuel, or power operations, successful completion of
a preoperational exercise is required to ensure that the licensee staff
is ready to implement the emergency plan and that the emergency plan,
as written, is acceptable. The proposed rule would amend paragraphs
IV.F.2.a.(i) through (iii) of appendix E to 10 CFR part 50 and 10 CFR
50.160(c)(1) and (c)(2) by removing the requirement to demonstrate
compliance within 2 years before issuance of an OL or the scheduled
date of initial loading of fuel. Regulatory flexibility in the
performance of this exercise may be desirable for applicants and
holders of a COL that are co-located on, or adjacent to, an existing
site and that may be able to subsume this demonstration requirement
within the emergency plan exercise program that already exists, up to
and including offsite exercise requirements. The specificity of the
requirement to conduct the initial exercise within 2 years of certain
milestones creates the potential to repeat the demonstration if
schedules change. The proposed rule would ensure that the initial
exercise need only be performed one time. The proposed rule would also
provide flexibility to applicants and holders of a COL in the
performance of preoperational exercises if they have preexisting
licensed power reactors at the same site with similar onsite and
offsite emergency plan elements in place. In such cases, the applicant
or holder of a COL and the licensee may credit the same exercise as
both the preoperational exercise required under proposed paragraphs
IV.F.2.a.(i) and IV.F.2.a.(iii) and the onsite exercise required under
paragraph IV.F.2.b of appendix E to 10 CFR part 50.
(iii) Alert and Notification System
The proposed rule would provide clarity for when backup alert and
notification system (ANS) methods are
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required. Traditional ANS strategies typically consist of a single
primary method based upon fixed sirens and a single backup method using
route-alerting strategies, but alternative methods are available. For
example, the Integrated Public Alert & Warning System (IPAWS) is FEMA's
national system for local alerting that includes parallel methods of
providing authenticated emergency and life-saving information to the
public typically through mobile phones using Wireless Emergency Alerts,
to radio and television via the Emergency Alert System, and on the
National Oceanic and Atmospheric Administration's Weather Radio among
others. IPAWS typically provides multiple, simultaneous, primary
methods of alerting the public implemented in parallel, and as such,
does not require a designated backup. Because of technologies like
IPAWS, the proposed rule would amend paragraph IV.D.3 of appendix E to
10 CFR part 50 to provide flexibility in meeting the requirements based
upon the ANS chosen by the applicable State as approved in the ANS
Design Report.
(iv) Use of Modern Terminology
Certain terminology used in the current regulations is no longer
common or does not align with advances in technology. The NRC proposes
to revise 10 CFR 50.160(b)(1)(iv)(A)(2) to provide clarity and to
simplify the language for the requirement to implement the emergency
plan in response to a security event. The NRC also proposes to revise
paragraph IV.D.2 of appendix E to 10 CFR part 50 to replace ``local
broadcast services'' with ``media sources.''
B. Emergency Planning Zone Certainty
The proposed rule would provide greater certainty and a streamlined
approach to EPZ determinations for all applicants and licensees. The
NRC is proposing to amend the requirements in 10 CFR 50.33(g),
50.47(c)(2), and 53.1109(g) to simplify EPZ determinations for new
reactors and to ensure the plume exposure pathway EPZ is no larger than
needed to implement predetermined, prompt protective measures.
Specifically, the proposed rule would (1) establish the bounds of an
EPZ to generally be about 2 to 10 miles (3.2 km to 16 km) in radius;
(2) provide certainty for a site-boundary EPZ for facilities with an
authorized power level less than 300 MWt; and (3) allow for case-by-
case determinations for all facilities. The proposed rule would make
conforming changes to 10 CFR 50.160(b)(3) in referring to the
requirements in proposed 10 CFR 50.33(g)(1) and 53.1109(g)(1) to
determine and describe the boundary and physical characteristics of the
EPZ in the emergency plan.
The recommended 10-mile (16-km) plume exposure pathway EPZ is a
generic planning distance for pressurized water reactor (PWR) and
boiling water reactor (BWR) technologies based on the analyses in
NUREG-0396. In the decades since, many studies have provided additional
risk insights and analyses that demonstrate the conservatism in the
NUREG-0396 analyses. These analyses include insights from the NRC's
State-of-the-Art Reactor Consequence Analyses (SOARCA) regarding the
magnitude and timing of severe accidents; insights from the NRC's Level
3 Probabilistic Risk Assessment (PRA) Project regarding the margin to
quantitative health objectives, multi-unit events, and integrated plant
risks; and specific studies to inform NRC's EP program including: (1)
NUREG/CR-7160, ``Emergency Preparedness Significance Quantification
Process: Proof of Concept,'' dated June 2013; (2) Task 1.5-1.6 Report
to User Need Request NSIR-2017-002, ``Analyses Informing Emergency
Planning Zone Size Determinations: Identification of Parameter
Sensitivities,'' dated December 2019; (3) SAND2022-3706, ``Scoping
Analysis of MACCS Modeling Improvements for the Study of Protective
Action Recommendations, dated March 2022; and (4) SAND2025-08913,
``Dose Exceedance Distance Sensitivity Based on Parametric
Uncertainty,'' dated July 2025. Combined, these analyses support the
use of an EPZ that is generally no less than 2 miles (3.2 km) and no
more than 10 miles (16 km) for implementing predetermined, prompt
protective measures. Certain analyses were also based, in part, on
conservative source terms for designs with an authorized power level
less than 300 MWt, which suggest that there is a very low risk of
exceeding acute doses offsite requiring a predetermined, prompt
response. Additionally, numerous research reports from national
laboratories, peer-reviewed published articles, and studies on small
modular reactors, microreactors, and advanced reactor designs
consistently demonstrate these designs would not require extensive
plume exposure pathway EPZs. Combined, these analyses support the
proposed changes to 10 CFR 50.33(g), 50.47(c)(2), and 53.1109(g) to
reduce the conservatism in the current 10-mile (16-km) EPZ size and
provide certainty for a smaller EPZ or a site-boundary EPZ without
requiring extensive analyses on the part of the applicant or licensee.
The proposed rule would also simplify the case-by-case EPZ
determination under 10 CFR 50.33(g)(1) and (g)(2) and 53.1109(g)(1) and
(g)(2). The EPZ determination should be a simple evaluation of the
consequences of a spectrum of accidents to inform protective action
strategies. Tools like PRA are useful to help inform EPZ determinations
and the risk insights can be used to produce a robust and resilient
emergency plan to deal with the residual risk of the facility and to
account for uncertainty and unknowns in the design and operation of the
facility. While a risk-informed design can provide valuable information
to develop a risk-informed EP program, including sizing the EPZ, the EP
program is not part of the design and the EPZ determination is not an
analysis to reiterate the safety case of the design, demonstrate
compliance with the Licensing Modernization Project (LMP) process, or
enforce design changes on an applicant or licensee to ensure that no
accident will ever exceed 1 rem (10 mSv) at the EPZ boundary.
Accordingly, the EPZ criteria for the case-by-case analysis are not
design criteria or dose limits. The 1 rem (10 mSv) TEDE value in 10 CFR
50.33(g)(2)(i)(A) is a threshold dose quantity below which it may be
demonstrated that no EPZ is needed to manage the radiological risks and
hazards of the facility. The EPZ criteria does not require an applicant
to demonstrate that doses cannot exceed 1 rem (10 mSv) TEDE over 96
hours beyond the EPZ boundary for all accidents. When doses can exceed
1 rem (10 mSv) TEDE, it is generally necessary to risk-inform the EPZ
size by considering additional factors as specified in 10 CFR
50.33(g)(2)(i)(A) and by considering the need for predetermined, prompt
protective measures as specified in 10 CFR 50.33(g)(2)(i)(B). The EPZ
criteria in 10 CFR 53.1109(g)(2) are applied in the same way. For
analysis purposes, the dose criteria is applied to a reference
individual, as is done with other dose criteria (e.g., 10 CFR
50.34(a)(1)(ii)(D)(1) and (2)) and with the Environmental Protection
Agency (EPA) protective action guides (PAGs) (see EPA, ``PAG Manual:
Protective Action Guides and Planning Guidance for Radiological
Incidents,'' issued in 2017). Therefore, the proposed rule would amend
10 CFR 50.33(g)(2)(i)(A) and 53.1109(g)(2)(i)(A) to remove the
reference to public dose as defined in 10 CFR 20.1003 and to
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apply the dose to a reference individual. The EPZ dose criteria is not
a limit, so the term ``public dose'' would be removed to avoid
confusion.
When the radiological consequences can exceed 1 rem (10 mSv), dose
criteria should not be strictly imposed at an EPZ boundary. There is
significant inherent variability in any consequence analysis. For dose
quantities less than 25 rem (250 mSv), this inherent variability can be
on the order of kilometers. This creates a significant challenge for
defining an EPZ boundary based on low dose, especially for EPZs less
than 2 miles (3.2 km) in radius as the inherent variability in the
consequence analysis could be as large or larger than the defined EPZ.
In developing the 2023 EP final rule, the NRC extensively reviewed
available studies and performed additional sensitivity analyses to
understand the inherent variability of consequence analyses that inform
the EPZ determination. The NRC is issuing, for public comment along
with this proposed rule, DG-1430, ``Performance-Based Emergency
Preparedness,'' which would be Revision 1 to the existing RG 1.242,
with an updated methodology for a case-by-case EPZ determination. The
revised guidance would provide applicants with clarification on the
spectrum of accidents, including the consideration of security-related
events and severe seismic events; additional guidance on risk-informing
dose-distance evaluations to account for uncertainty and inherent
variability in the calculation; and additional guidance for assessing
the need for predetermined, prompt protective measures. These guidance
updates, based on NRC analyses of parameter uncertainty and NRC staff
experience with the review of EPZ methodologies, would provide simple
methods to make better risk-informed EPZ determinations.
The proposed rule would reinforce the purpose of the EPZ as a
planning tool and encourage the use of risk-informed protection
strategies for situations that do not rely on EPZs. The proposed rule
would amend 10 CFR 50.47(b)(10) by removing the word ``EPZ'' wherever
it appears to ensure protection strategies are developed for the
potential pathways of exposure and are not limited to a predefined
zone. The proposed rule would also amend 10 CFR 50.33(g)(1),
50.47(c)(2), and 53.1109(g)(1) to state that emergency plans would need
to describe such actions as are appropriate to avoid or reduce dose
within and beyond the EPZ or site boundary. The EPZ concept is not
limited to protective actions for the public, but applies to anyone
within the EPZ, including emergency workers and onsite personnel. The
proposed rule would ensure that the planning necessary to implement
predetermined, prompt protective measures would be focused within the
area most at-risk surrounding a nuclear power reactor, including areas
within the site boundary. Consistent with the historical planning basis
for the EPZ, this is an area where there is potential for acute doses
or early health effects from the accidental release of radioactive
material in addition to the risk of stochastic effects from radiation
exposure. An EPZ may also be defined for purposes of managing the risk
of stochastic effects, but the proposed rule would encourage the use of
risk-informed protection strategies to make informed protective action
decisions for the management of such risks, rather than rely on a
predetermined, prompt response that could do more harm than benefit.
The NRC proposes to clarify the requirements for use of evacuation time
estimates (ETE) to inform protective action recommendations and
protective action strategies to distinguish between the level of
planning required for predetermined, prompt protective actions, as
opposed to taking action as conditions warrant. Specifically, the
proposed rule would amend the requirements in paragraphs IV.3 and IV.4
in appendix E to 10 CFR part 50 by adding the phrase, ``predetermined,
prompt'' to clarify which protective action recommendations and
strategies within the EPZ are informed by the ETE.
Consistent with the rationale for revision of the EPZ criteria, the
proposed rule would amend 10 CFR 50.33(g)(1) and 53.1109(g)(1) by
eliminating the requirement to submit response plans of State, local,
and participating Tribal governmental entities. The proposed change to
10 CFR 50.33(g)(1) and 53.1109(g)(1) would require applicants to
coordinate with offsite organizations with responsibilities for coping
with emergencies, including State, local, and Tribal governmental
agencies, as applicable. This coordination would ensure that response
organizations are aware of potential radiological consequences of the
facility and have been consulted on appropriate protective measures,
including the extent of any EPZ. The proposed rule would require
applicants to include information that describes the extent of
interaction with these agencies. The NRC would not require FEMA
findings and determinations on the extent of interaction between the
applicant and offsite response organizations because this proposed
change is not part of the FEMA review of offsite plans under FEMA's
regulations in 44 CFR part 350, ``Review and Approval of State and
Local Radiological Emergency Plans and Preparedness.''
The proposed rule changes to the plume exposure pathway EPZ would
be consistent with Federal guidance and international standards.
Section 2.2.4 of the EPA PAG Manual discusses the relation between PAGs
and EPZs and states, ``The pre-designated areas for immediate
protective action may be reserved for use only in the most severe
incidents and in cases when the facility operator cannot provide a
quick estimate of projected dose based on actual releases. For lesser
incidents, or if the facility operator is able to provide prompt off-
site dose projections, the area for immediate protective action may be
specified at the time of the incident instead of using a pre-designated
area.'' Similarly, the International Atomic Energy Agency (IAEA) Safety
Standards in General Safety Requirements (GSR) No. 7, ``Preparedness
and Response for a Nuclear or Radiological Emergency,'' specify use of
a precautionary action zone (PAZ) for taking urgent protective actions
before any significant release occurs on the basis of conditions at the
facility in order to avoid or to minimize severe deterministic effects.
The recommended size for the PAZ within IAEA standards is 1.9 to 3.1
miles (3 to 5 km) for reactors greater than 1000 MWt and 0.3 to 1.9
miles (0.5 to 3 km) for reactors 100 to 1000 MWt. The proposed change
to 10 CFR 50.33(g), 50.47(c)(2), and 53.1109(g) would establish EPZs
that remain aligned with EPA guidelines and international standards to
meet the purpose of the EPZ as a planning tool for implementation of
predetermined, prompt protective actions. Consistent with the PAG
Manual, the zones for response could also be specified at the time of
the emergency, rather than predesignated.
The proposed rule would add new 10 CFR 50.47(g), 50.160(c)(3), and
53.855(c) to allow a licensee to submit a license amendment to change
its plume exposure pathway EPZ. The proposed rule would require the
change to be agreed on by the applicable State, local, and Tribal
governmental authorities before submission to the NRC. The FEMA REP
Program Manual provides a process for offsite response organizations to
follow for changes to the EPZ boundary in accordance with FEMA
regulations in 44 CFR 350.14.
Consistent with the 2023 EP final rule, the proposed rule would
remove
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the requirement to define an ingestion pathway EPZ. The proposed rule
would remove reference to the ingestion pathway EPZ in 10 CFR
50.33(g)(1) and 53.1109(g)(1), 50.47(b)(10), 50.47(c)(2), and paragraph
IV.F.2.a.(i) and footnote 1 of appendix E to 10 CFR part 50. In lieu of
a defined zone, the proposed rule would amend 10 CFR 50.33(g)(1),
50.47(c)(2), and 53.1109(g)(1) to require emergency plans to describe
such actions as are appropriate to protect the ingestion pathway. The
capabilities described in the emergency plan would need to address
major exposure pathways associated with the ingestion of contaminated
food and water. The duration of any exposure to contaminated food or
water could range from weeks to months and represents a long-term
response need. The current 50-mile ingestion pathway EPZ is based on
the planning assumptions in NUREG-0396 and does not reflect technology
advancements and modern response capabilities for interdicting to
prevent ingestion of contaminated food and water, such as the use of
Geographic Information System (GIS) tools and unmanned aerial vehicles
for monitoring.
Currently, paragraph IV.E.8.b of appendix E to 10 CFR part 50
requires the licensee's emergency operation facility (EOF) to be
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. The 10 miles is
based on the current 10-mile EPZ requirement and does not account for
scalable EPZs under current and proposed 10 CFR 50.33(g) and
53.1109(g). The proposed rule would amend paragraph IV.E.8.b of
appendix E to 10 CFR part 50 to specify the location of the emergency
operations facility in relation to the EPZ boundary. The proposed rule
would not change the requirement for a licensee to request Commission
approval to locate an EOF more than 25 miles from a nuclear power
reactor site.
C. Eliminating Redundant Requirements
(i) Preliminary Emergency Plans
The NRC proposes to eliminate the requirements for applicants to
submit preliminary plans for coping with emergencies because these
submittals do not significantly enhance licensing efficiency. Many
details of the emergency plan depend on conditions of the as-built
facility and site-specific parameters, including agreements with
offsite emergency response organizations. These details are often not
available early in the licensing process. The NRC reviewed the content
of CPs and ESPs and found that the level of information contained in
the associated preliminary emergency plans or major features of
emergency plans was very limited, often nothing more than a general
structure and placeholder for more detailed information. The proposed
rule would eliminate 10 CFR 50.34(a)(10) and 53.1309(a)(4) that require
applicants to submit the preliminary plans for coping with emergencies
as part of the preliminary safety analysis report for a CP application.
The proposed change would reduce the burden on applicants to provide
preliminary plans of limited benefit in licensing. The proposed rule
would not change 10 CFR 52.17(b)(2)(i) because applicants may choose
not to submit major features of the emergency plan in the ESP site
safety analysis report if such details are not available. Instead, the
NRC encourages applicants to engage the NRC in pre-application
activities in preparation for submission of emergency plans as part of
the FSAR or in preparation for submission of major features of the
emergency plan as part of the ESP site safety analysis report. As a
conforming change, the proposed rule would amend paragraphs I.1 and I.2
of appendix E to 10 CFR part 50 to remove the requirements to submit
emergency plan information in the preliminary safety analysis report
and would eliminate section II, ``The Preliminary Safety Analysis
Report,'' of appendix E to 10 CFR part 50. Many of the elements in
section II of appendix E to 10 CFR part 50 are redundant to other
criteria or would be considered as part of the development of emergency
plans in coordination with offsite response organizations under
proposed 10 CFR 50.33(g) and 53.1109(g).
(ii) Independent Program Element Reviews
The proposed rule would amend 10 CFR 50.54(t) and
53.1565(d)(3)(vii) to eliminate the requirement for licensees to ensure
that all program elements are reviewed by people who have no direct
responsibility for implementation of the EP program. Since the
implementation of this regulation, the NRC's oversight program has
evolved such that this regulation is redundant to the NRC Reactor
Oversight Program and adds no additional oversight benefit to the EP
program. However, the proposed rule would retain the requirement in 10
CFR 50.54(t)(2) and 53.1565(d)(3)(vii)(B) for an annual review of the
interface between licensee and offsite response organizations to ensure
the adequacy of the interface with State and local governments.
(iii) Evacuation Time Estimate Updates
The NRC is proposing to eliminate requirements related to updates
of the ETE that have limited utility in ensuring effective
implementation of protective action recommendations. The proposed rule
would eliminate the requirement to estimate EPZ permanent resident
population changes during the years between decennial censuses and to
update the ETE based on the criteria of paragraphs IV.5 and IV.6 of
appendix E to 10 CFR part 50. This proposed change is based, in part,
on an analysis of the limited number of licensees over the past
decennial periods that met the criteria to perform the ETE update and
an analysis of the impact of updated ETEs on protective action
strategies, following the guidance in Supplement 3, ``Guidance for
Protective Action Strategies,'' to NUREG-0654/FEMA-REP-1. Decennial ETE
updates include sensitivity analyses that provide the expected change
to the ETE as the permanent resident population changes. In addition,
section 5.4.1, ``Extreme Conditions,'' of NUREG/CR-7002, ``Criteria for
Development of Evacuation Time Estimate Studies,'' Revision 1, contains
guidance for updating the ETE if conditions within the EPZ change
significantly. The NRC has determined that the sensitivity analyses and
the ETE guidance would ensure that ETEs remain adequate for use in
protective action recommendations and in developing offsite protective
action strategies in the years between decennial censuses. For similar
reasons, a review of changes in the EPZ population under paragraph IV.7
of appendix E to 10 CFR part 50 would not be required for parts 52 and
53 licensees. As such, the proposed rule would eliminate paragraphs
IV.5, IV.6, and IV.7 of appendix E to 10 CFR part 50.
(iv) Periodic Communication Tests With NRC
The proposed rule would eliminate the requirement to test each NRC
Regional Office Operations Center under paragraph IV.E.9.d of appendix
E to 10 CFR part 50. A monthly test between each licensee and the
appropriate Regional Office Operations Center would not be required
because the regional Operations Centers are not typically staffed by
the NRC unless it is warranted by the escalation of an emergency event.
Monthly testing of communication between the licensee's primary
response location and the NRC
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Headquarters Operations Center would be sufficient.
D. Risk-Informing the Emergency Plan Change Process
The NRC proposes to revise and risk-inform the emergency plan
change process in 10 CFR 50.54(q) and 53.1565(d)(3) regarding which
proposed changes would need to receive prior approval from the NRC.
Licensees would be required to evaluate a reduction in effectiveness
for only changes to the emergency plan involving risk-significant
planning standards. Under the proposed rule, the licensee would analyze
changes related to the other planning standards and applicable
requirements in appendix E to 10 CFR part 50, or 10 CFR 50.160 only to
ensure they continue to meet the regulatory requirements.
The proposed rule would add new 10 CFR 50.54(q)(1)(v) and
53.1565(d)(3)(i)(E) to define risk-significant planning standards as
those providing the most essential functions of EP to ensure adequate
protective measures are taken to protect the public in the event of a
radiological emergency, including classification, notification,
assessment, and protective actions. The proposed definition would also
include the standards for providing adequate staffing and facilities as
risk-significant for the purposes of evaluating changes to the
emergency plan as these standards have a direct impact on the ability
to effectively implement the risk-significant planning standards. The
proposed definition would apply to the standards of 10 CFR 50.160 or
50.47 and applicable requirements in appendix E to 10 CFR part 50.
Consistent with the proposed changes to allow licensees the option
to comply with either the requirements in appendix E to 10 CFR part 50
and, for nuclear power reactor licensees, the planning standards of 10
CFR 50.47(b), or the requirements in 10 CFR 50.160, the proposed rule
would combine 10 CFR 50.54(q)(2)(i) and (ii) into a single paragraph
(q)(2) and would revise (q)(3)(i) and (ii) for the evaluation of
changes to the emergency plan. The proposed rule would similarly
combine 10 CFR 53.1595(d)(3)(ii)(A) and (B) and revise 10 CFR
53.1595(d)(3)(iii).
XXV. Background--Optional Submittal of Operational Programs
For several years, stakeholders have expressed growing interest in
the rapid, widespread deployment of reactors of a standard design
(e.g., July 31, 2024, letter from NEI to the NRC, ``Regulation of Rapid
High-Volume Deployable Reactors in Remote Applications (RHDRA) and
Other Advanced Reactors,'' followed by a July 14, 2025, supplement to
that letter). At the same time, the NRC staff developed strategies to
provide for the predictable and efficient licensing and regulation of
microreactors that would use standard designs. The staff sought
Commission approval for some of these proposals in SECY-24-0008,
``Micro-Reactor Licensing and Deployment Considerations: Fuel Loading
and Operational Testing at a Factory,'' dated January 24, 2024, and
SECY-25-0052, ``Nth-of-a-Kind Microreactor Licensing and Deployment
Consideration,'' dated June 18, 2025. The latter paper recommended a
change in Commission policy to allow the NRC to review, approve, and
afford finality to, as appropriate, standard operational programs
submitted to the NRC in connection with a design certification or ML
application. These agency actions would support the rapid licensing and
high-volume deployment of new microreactors and other low consequence
reactors and the direction in E.O. 14300 to adopt shorter timeframes
tailored to particular licensing pathways.
The Commission approved the staff's recommendations in SRM-SECY-
0008, ``Staff Requirements--SECY-24-0008--Micro-Reactor Licensing and
Deployment Considerations: Fuel Loading and Operational Testing at a
Factory,'' dated June 17, 2025, and SRM-SECY-25-0052, ``Staff
Requirements--SECY-25-0052--Nth-of-a-Kind Microreactor Licensing and
Deployment Consideration,'' dated November 13, 2025. The NRC included
many of these proposals, including the proposal to allow an ML
applicant to include standard operational program information in its
application, in the ``Licensing Requirements for Microreactors and
Other Reactors With Comparable Risk Profiles'' proposed rule (91 FR
23628; May 1, 2026). The NRC is proposing conforming changes to 10 CFR
parts 52 and 53 in this rulemaking to allow microreactor and other
developers the option to submit essentially complete programmatic
controls, operational programs, or operational requirements with an ML
application under 10 CFR part 52 or 53.
XXVI. Discussion--Optional Submittal of Operational Programs
The NRC proposes to modify 10 CFR 52.158, ``Contents of
application; additional technical information,'' to allow ML applicants
the option to submit essentially complete operational program
information with their applications and 10 CFR 52.171, ``Finality of
manufacturing license; information requests,'' to provide finality to
such program information that is reviewed and approved by the NRC as
part of the ML review. As stated in proposed 10 CFR 52.158(c), this
optional program information would be submitted ``to satisfy
requirements for license applications that may reference a
manufacturing license,'' and this program information submitted with
the ML application would be assessed against the pertinent requirements
for these other license applications.
The appropriate change control mechanism for changes to this
information by an ML holder would be established through the issuance
of the ML. Under current regulations, the change control requirements
in 10 CFR 52.171(b)(1) for changes sought by the holder of an ML apply
to design information, not operational program information. Elsewhere
in this proposed rule, the NRC proposes to modify 10 CFR 52.171(b)(1),
but as modified, the change control provision would apply to
information in the FSAR submitted under 10 CFR 52.157 and not to non-
FSAR information submitted under 10 CFR 52.158. Moreover, the change
control process in proposed 10 CFR 52.171(b)(1) relies on the change
control criteria in 10 CFR 50.59, whereas some operational programs are
subject to other change control provisions (e.g., 10 CFR 50.54(a), (p),
and (q)). Therefore, the NRC would establish the appropriate change
control process for the optionally submitted operational program
information as part of issuing the ML.
An applicant or licensee who references or uses a nuclear power
reactor manufactured under an ML who wishes to depart from or omit (in
whole or part) the optional operational program information approved in
the ML would not be subject to the process in 10 CFR 52.171(b)(2)
because this optional operational program information would not
constitute design characteristics, site parameters, terms and
conditions, or approved design information subject to 10 CFR
52.171(b)(2). Departures from, or omissions of, the optional
operational program information approved in the ML would be reviewed in
accordance with the applicable operational program requirements for the
license being applied for (e.g., OL or COL).
This proposed change would enhance regulatory certainty and
expedite review timelines for CP/OL and COL applicants wishing to
reference operational programs approved with an ML but would still
allow flexibility for these applicants to submit their own
[[Page 44597]]
programs. There would be no additional obligations imposed on license
applicants as this would be optional information for a developer to
provide and optional for a CP/OL or COL applicant to reference.
Additionally, the hearing opportunity provided as part of the ML review
process would allow for public engagement on the program information
included in the ML application. For licensing efficiency, agreement on
operational programs being considered to be essentially complete could
be achieved between the NRC and applicants through pre-application
interactions.
The NRC is proposing a similar change to 10 CFR part 53 to allow
for the optional submittal with the ML application of operational
program information not material to the design. The NRC is proposing to
modify 10 CFR 53.1282, ``Contents of applications for manufacturing
licenses; other application content,'' to allow for the submittal of
such program information and 10 CFR 53.1288, ``Finality of
manufacturing licenses,'' to provide finality to the information that
would be reviewed and approved by the NRC.
XXVII. Background--Early Site Permit for Nuclear Power Plants
Section 5(i) of E.O. 14300 directs the NRC to ``[r]econsider the
regulations governing the time period for which a renewed license
remains effective, and extend that period as appropriate based on
available technological and safety data.'' In response, the NRC is
proposing to amend its regulations to remove the ESP fixed term
requirement.
The NRC issued 10 CFR part 52 on April 18, 1989 (54 FR 15372), to
provide procedures for the early resolution of safety and environmental
issues in commercial power reactor licensing proceedings and to provide
for the standardization of the design of nuclear power plants. To
further those ends, the Commission issued 10 CFR part 52 to add
alternative licensing processes for early site permits (ESP), standard
design certifications, and COLs. These alternatives in 10 CFR part 52
were in addition to the two-step licensing process that already existed
in 10 CFR part 50. The processes in 10 CFR part 52 allow for resolving
safety and environmental issues early in licensing proceedings, which
would result in regulatory stability, and were intended to enhance the
safety and reliability of nuclear power plants through standardization.
Part 53 of 10 CFR also provides for ESPs.
In particular, ESPs provide a process for applicants to resolve the
majority of site-specific safety and environmental issues prior to
applying for a CP or COL. Subsequent to promulgating 10 CFR part 52,
the NRC has issued six ESPs,--the Exelon Generation Company, LLC ESP
Site, ESP-001, March 15, 2007; System Energy Resources, Inc., Grand
Gulf ESP Site, ESP-002, April 5, 2007; Dominion Nuclear North Anna,
LLC, North Anna ESP Site, ESP-003, November 27, 2007; Southern Nuclear
Operating Company Vogtle Electric Generating Plant ESP Site, ESP-004,
August 26, 2009; PSEG Power, LLC and PSEG Nuclear, LLC PSEG Site Early
Site Permit, ESP-005, May 5, 2016; and Tennessee Valley Authority,
Clinch River Nuclear Site Early Site Permit, ESP-006, December 19,
2019.
XXVIII. Discussion--Early Site Permit for Nuclear Power Plants
Under the current regulations in 10 CFR 52.26, ``Duration of
permit,'' the NRC may issue an ESP with a term no longer than 20 years.
Upon the elapse of 20 years, the ESP expires, but an ESP holder has the
option to renew it for an additional term under current 10 CFR 52.29,
``Application for amendment to update an early site permit.'' The term
for an ESP is similarly limited in 10 CFR 53.1164, ``Duration of
permit.''
This proposed rule would remove the requirement in 10 CFR 52.26 and
53.1164 for an ESP to include a fixed term. Because each ESP would be
issued without a fixed term, renewal would no longer be needed.
Nonetheless, the revised regulations in 10 CFR 52.29 would provide an
option by which an ESP holder may choose to update its ESP, which could
maintain the preclusive effect of the ESP if the ESP is referenced in a
COL or CP application. The proposed rule would similarly revise 10 CFR
53.1173, ``Application for renewal.'' The requirements specific to
renewal at 10 CFR 52.33, ``Duration of renewal,'' and 10 CFR 53.1179,
``Duration of renewal,'' would be removed and reserved.
As noted in the 1989 10 CFR part 52 final rule, the ESP provides
for early resolution of site-related issues, making possible the
``banking'' of the site, to enable more efficient licensing of a future
nuclear power plant. The current rule mandates renewal of the ESP at a
set point in time, not to exceed 20 years after issuance, and therefore
requires the holder of the ESP to undertake updating the ESP to retain
the benefits of issue finality. However, there are a number of factors
that may influence the time that elapses between NRC's issuance of an
ESP and a determination by the ESP holder to reference the ESP in a CP
or COL application. Indeed, two of the issued ESPs are nearing 20 years
since issuance without having been referenced in a CP or COL
application. Therefore, this maximum 20-year duration of an ESP and the
associated renewal requirement may artificially constrain the ESP
holder's development planning because the renewal-requirement lacks a
direct linkage to either a change in the ESP identified by the holder
or the holder's plans to reference the ESP in a CP or COL application.
By removing the fixed term duration and providing a process for the
permit holder to update the ESP through an amendment, the proposed rule
would provide permit holders with flexibility to retain the site as
``banked,'' rather than referencing the ESP in a CP or COL application
in a shorter time frame, and determine when or if the ESP will be
updated. Removing the fixed term, however, would necessitate a change
to the finality provisions in 10 CFR 52.39, ``Finality of early site
permit determinations,'' to provide clarity regarding the limited
duration of finality for site-specific characteristics of the initial
ESP application and to account for stale environmental evaluations.
Accordingly, the proposed rule would establish a 20-year timeframe for
information in the ESP to retain finality, which would account for the
limited time for which site data remains valid and the potential for
environmental evaluations to become stale. The proposed rule would
similarly revise 10 CFR 53.1188, ``Finality of early site permit
determinations.''
The proposed 20-year timeframe would provide a boundary for both
preparing a CP or COL application that references an ESP and NRC
review, to avoid the level of uncertainty that would exist with
providing analyses without a fixed term. However, ESP holders would
have the option to update the information in the ESP to refresh the
site data and environmental evaluations and thereby extend the 20-year
timeframe at their discretion. Thus, when developing an ESP
application, applicants for future ESPs should consider the 20-year
expected duration for potentially time sensitive information, such as
seismic, meteorological, hydrologic, and geologic characteristics; the
presence of nearby facilities such as industrial, military, or
transportation; the population profile; and environmental data and
evaluations.
Under this proposed rule, the ESP holder may choose to update the
ESP by submitting an amendment to update the potentially time sensitive
information, which would provide an additional 20 years of finality for
the updated safety
[[Page 44598]]
and environmental information if the NRC grants the amendment.
Alternatively, an ESP holder could forgo finality and address those
issues directly in a CP or COL application that references the ESP. The
request for amendment of the ESP could be made by the ESP holder at any
time. Unlike the current regulatory framework, where information must
be updated prior to expiration of the ESP at year 20 through renewal to
maintain issue resolution, the proposed changes would enable the ESP
holder to retain the ESP for as long as desired and identify needed
updates only when and if the holder determines the update would be
advantageous to support a future CP or COL application. The proposed
rule would allow the ESP holder to submit an amendment to update the
time-sensitive information and request an extension for an already
extended permit, providing flexibility for the ESP holder to update the
ESP at such time and as many times as needed. To support the removal of
a fixed term, a provision would be added to enable the ESP holder to
request termination of an ESP at any time, with consideration for the
status of any required redress for authorized activities. The changes
would enhance flexibility for the ESP holder and reduce the effort
associated with ESP renewal at year 20. NRC resources, likewise, would
not be expended for review of ESP renewal applications. To be sure,
some portions of an ESP, such as design parameters specified in the
permit, are not time-sensitive and would continue in effect unless the
ESP holder seeks to amend them or an applicant for a CP or COL
referencing the ESP proposes to change them.
This proposed rule would remove and reserve paragraph (c) of 10 CFR
2.109, ``Effect of timely renewal application,'' which provides for
timely renewal of the ESP.
This proposed rule would revise 10 CFR 52.15, ``Filing of
applications,'' to remove the reference to ``renewal'' and replace it
with ``amendment,'' reflecting that the permit would be issued with no
fixed term and would not expire.
This proposed rule would revise 10 CFR 52.18, ``Standards for
review of applications,'' to clarify its applicability to the initial
issuance of the ESP. This proposed rule would revise 10 CFR 52.25,
``Extent of activities permitted,'' to remove the reference to the
duration of the permit, reflecting that an ESP would no longer have a
fixed term. The requirement for an ESP to remain in effect for site
redress by the ESP holder if any activities authorized by 10 CFR
52.24(c) are performed would be unchanged. The proposed rule would
similarly revise 10 CFR 53.1161, ``Extent of activities permitted.''
This proposed rule would revise 10 CFR 52.26(a) to remove the
current duration, identifying that the permit would be issued with no
fixed duration. The proposed rule would similarly revise 10 CFR
53.1164. Removal of the expiration date from the ESP would enable the
permit holder to retain the site as banked until such time as either
the permit is subsumed into another application or a decision is made
to terminate the ESP. Because some site-specific characteristics are
potentially time dependent, 10 CFR 52.39 and 53.1188 would limit
finality for safety issues to 20 years from issuance of the ESP or an
amendment to update the information for those items. Finality for
environmental issues would also be limited to 20 years from issuance of
the ESP or an amendment. Because the ESP would be issued with no fixed
term, this proposed rule would revise 10 CFR 52.35, ``Use of site for
other purposes,'' to provide a provision for the permit holder to
request termination and to ensure the ESP holder identifies site
redress activities to the Commission as part of the termination
request. The ESP would not be terminated until site redress activities
have been performed in accordance with 10 CFR 52.25. Termination of the
ESP would not preclude the same or another applicant from submitting
another application for an ESP or another license for the same site at
a future time. The proposed rule would similarly revise 10 CFR 53.1182,
``Use of site for other purposes.''
This proposed rule would rename 10 CFR 52.29 and 53.1173 to,
``Application for amendment to update an early site permit,'' and
revise these sections to provide the requirements for submitting an
amendment. Following NRC review and completion of the hearing process,
issuance of the amended ESP would extend finality for future use of the
ESP for reference in a CP or COL application. The specific information
limited to a 20-year duration for finality would be specified in 10 CFR
52.39 and 53.1188.
This proposed rule would rename 10 CFR 52.31, ``Criteria for
renewal,'' as ``Issuance of amendment to update an early site permit,''
and revise it to include specific provisions for the Commission's
issuance of an amended ESP. An application for amendment would be
reviewed against similar criteria to that previously identified for
renewal, which reflect the AEA, the Commission's regulations, and
orders applicable and in effect at the time the ESP was originally
issued. New requirements to be imposed during the amendment would be
limited to those necessary for adequate protection to public health and
safety or common defense and security. Additionally, for the purpose of
site characteristic data and environmental data and evaluations which
may be time dependent, the proposed rule would limit finality for those
issues to 20 years while providing flexibility for the ESP holder to
update that information through an amendment at any time. Therefore,
the NRC's review of those potentially time-dependent characteristics
would encompass the 20-year timeframe that aligns with the finality
that would be afforded to that information. The proposed rule would
similarly revise 10 CFR 53.1176, ``Criteria for renewal,'' and rename
it as, ``Issuance of amendment to update an early site permit.''
This proposed rule would revise 10 CFR 52.39 to remove reference to
a renewed ESP and replace it with an amended ESP. The site
characteristic data on which an ESP is based, however, is valid for
only a limited duration. Accordingly, the finality provisions in 10 CFR
52.39 would be revised to reflect that an ESP does not resolve siting
issues in 10 CFR part 100, ``Reactor Site Criteria,'' specifically
including site characteristics in the ESP, in a proceeding on a
referencing application submitted more than 20 years after the date of
issuance of the ESP or the date of an update to the ESP in which a
matter has been resolved, whichever is later. This revised provision
would not affect the finality of the source term included in the ESP.
The duration limitation on finality of safety issues would be updated
by revising 10 CFR 52.39(c)(1)(v) and finality of environmental issues
would be updated by revising 10 CFR 52.39(c)(1)(vi). The proposed rule
would similarly revise 10 CFR 53.1188.
The new provisions in 10 CFR 52.29 would provide for amendments to
update ESPs in paragraph (e) of 10 CFR 52.39. The proposed rule would
similarly revise 10 CFR 53.1188(e) to add 10 CFR 53.1173 as a
requirement.
This proposed rule would amend 10 CFR 52.79 by redesignating
paragraphs (b)(3) through (5) as paragraphs (b)(4) through (6) and
revising paragraph (3) to add a requirement for a referencing COL
application to confirm whether the potentially time-dependent site
characteristics were updated in the ESP within the 20 years prior to
filing the COL application and thus resolved under 10 CFR 52.39. If
not, the COL applicant would need to evaluate and update that
information in the FSAR. The information to be evaluated would be that
information in the site safety analysis report required by paragraphs
[[Page 44599]]
10 CFR 52.17(a)(1)(vi) through (ix), which includes seismic,
meteorological, hydrologic, and geologic characteristics; the presence
of nearby facilities such as industrial, military, or transportation;
and the population profile. The safety assessment based on these
characteristics would also require evaluation and update, except that
the referencing COL application would not need to update the postulated
source term stated in the ESP if it fell within the source term derived
from the design. If the postulated source term stated in the ESP did
not fall within the source term derived from the design, then the
referencing application would need to propose a variance from the ESP
in accordance with 10 CFR 52.39 and 10 CFR 52.93. The proposed rule
would similarly revise 10 CFR 53.1416.
This provision parallels the requirement in 10 CFR 52.39 and
53.1188 that specifies the 20-year limitation for finality on this type
of potentially time-dependent information and recognizes that the
holder of an ESP may choose not to update the ESP to retain finality
once the ESP is 20 years past issuance. Together, the proposed revised
provisions of 10 CFR 52.39 and 52.79, as well as 10 CFR 53.1188 and
53.1416, would provide flexibility for an ESP holder to determine when
and what information to update while retaining the ESP for reference
until such time as the holder chooses to terminate, and for a
referencing COL applicant to have clarity regarding which aspects of
the ESP would need to be evaluated and updated in the FSAR submitted as
part of the application.
XXIX. Background--Manufacturing License Term Extension
Section 5(i) of E.O. 14300 directs the NRC to ``[r]econsider the
regulations governing the time period for which a renewed license
remains effective, and extend that period as appropriate based on
available technological and safety data.'' In response, the NRC is
proposing to amend its regulations to extend the term for initial
issuance and the renewal of an ML to enhance the NRC's regulatory
effectiveness and efficiency in implementing its licensing and approval
processes. This change would align the term of MLs with the current
term of design certifications.
On August 28, 2007, the NRC revised the provisions applicable to
the licensing and approval processes for nuclear power plants with the
final rule, ``Licenses, Certifications, and Approvals for Nuclear Power
Plants'' (72 FR 49352). In that final rule, a new subpart F,
``Manufacturing Licenses,'' of 10 CFR part 52 replaced former appendix
M of 10 CFR parts 50 and 52 that previously governed MLs. Under subpart
F, an ML would be issued upon the submission of an acceptable final
reactor design, equivalent to that required for a design certification
under 10 CFR part 52, and certain information on the manufacturing
process. The term for an ML was set to be for not less than 5, nor more
than 15 years from the date of issuance, and was established to be
consistent with the maximum term for a standard design certification in
2007. The Commission's stated intent was ``to encourage the use of a
manufacturing license for the manufacture of more than one nuclear
power reactor.'' However, no ML has been issued under subpart F of 10
CFR part 52.
In the July 2, 2025, direct final rule, ``Revising the Duration of
Design Certifications'' (90 FR 28869), the NRC replaced the 15-year
duration for design certifications with a 40-year duration period, both
for certifications currently in effect and generically for future
certifications, including renewals. As explained in that direct final
rule, extending the term of design certifications reduces unnecessary
regulatory burden on applicants and saves NRC resources without any
reduction in safety or security.
XXX. Discussion--Manufacturing License Term Extension
The proposed amendments to the regulations in 10 CFR 52.173,
``Duration of manufacturing license,'' and 52.181, ``Duration of
renewal,'' would change the duration of an ML to a maximum of 40 years
and the duration of the renewed ML to a maximum of 40 years. The
minimum terms of initial and renewed MLs would remain at 5 years.
Amending the regulations to extend the term of an ML would make the
duration of the ML consistent with the current durations of design
certifications. The current landscape for nuclear power reactors
designs, which encompasses potential microreactors and small modular
reactors in addition to traditional large LWRs, may significantly
benefit from the regulatory stability afforded by a longer ML term.
This proposed change would allow ML holders to significantly decrease
the burden associated with license renewal as the frequency of renewals
would decrease. No other impact on licensees or other stakeholders is
expected.
The NRC proposes to align the maximum term for an ML with the
maximum term for a design certification because of the similarities
between the scope of the two approvals. Most of the required
information for an ML pertains to the same final design information
required for a design certification. While part of an ML application
pertains to organizational and QA information on manufacturing
activities, this information is similar in kind to the organizational
and QA information required for reactor COLs, which are also subject to
a 40-year license term. Finally, consistent with the explanation
provided by the NRC when it extended the term of a design certification
to 40 years, the proposed extension to the term of MLs would not lead
to any reduction in safety or security. If a safety or security issue
is discovered during the term of an ML, the issue finality provisions
in 10 CFR 52.171 allow the NRC to impose requirements to address the
issue, as well as allow the holder of an ML or the applicant or
licensee who references or uses a nuclear power reactor manufactured
under an ML to seek a change to address the issue.
XXXI. Background--Nuclear Power Plant License Renewal
Section 5(i) of E.O. 14300 directs the NRC to ``[r]econsider the
regulations governing the time period for which a renewed license
remains effective, and extend that period as appropriate based on
available technological and safety data.'' Also, as set forth in
section 2 of the E.O., it is the policy of the United States to
facilitate appropriate operational extensions of the current nuclear
fleet. In response, the NRC is proposing to make several amendments to
its nuclear power reactor license renewal regulations.
The NRC first issued the license renewal rule (10 CFR part 54) on
December 13, 1991 (56 FR 64943) and later revised it on May 8, 1995 (60
FR 22461). The rule establishes the procedures, criteria, and standards
governing the renewal of nuclear power plant OLs. It also references
the environmental protection requirements in 10 CFR part 51 for the
evaluation of the environmental effects of the extended plant
operation.
Since publishing the 1995 rule, the NRC has issued renewed licenses
for nearly all currently operating plants, extending their operation
from 40 to 60 years. The NRC has also issued second, subsequent renewed
licenses for several plants, allowing them to operate for up to 80
years. Based on the experience gained through these renewals, the NRC
has identified several opportunities to revise the regulations to make
the process more flexible, efficient, and better able to support
operational extensions.
[[Page 44600]]
XXXII. Discussion--Nuclear Power Plant License Renewal
This rulemaking proposes revising requirements in the following
areas:
A. Extension of the Renewal Time Period
When the NRC issued the 1991 license renewal rule, it noted that
the 40-year limit on licenses in section 103 of the AEA was not based
on safety or security reasons; rather, it ``was adopted for antitrust
and financial reasons'' (56 FR 64960; December 13, 1991). Section 103
of the AEA also provides that OLs may be renewed. However, the NRC
noted that, because existing plants were licensed for 40-year periods,
some aging-related issues expected to arise beyond 40 years of
operation may not have been fully considered. (Id. at 64946, 64954).
Therefore, while the NRC determined that there was no reason to assess
all aspects of a plant's licensing basis during license renewal, the
NRC should ensure that aging will be effectively managed during the
period of extended operation. Moreover, this aging management does not
ensure that a plant will operate until the end of its period of
extended operation; rather it ensures that aging-related degradation
will be identified and addressed before it becomes a safety issue.
The current regulation in 10 CFR 54.31(b) limits a renewed license
to no more than 20 years beyond the existing license expiration. The
NRC explained that the intent of the 20-year limit was to provide ``a
useful opportunity to validate and reassess, if necessary, the current
understanding of age-related degradation effects.'' (Id. at 64964). At
that time, the NRC also noted that it might revisit the limit once
additional experience is acquired and the NRC gains confidence in
licensee programs that manage age-related degradation.
The NRC has examined its experience with aging management and in
light of that experience is proposing to revise 10 CFR 54.31(b) to
extend the maximum renewal term to 40 years, consistent with the
maximum license period in section 103 of the AEA. Most U.S. reactors
are already operating beyond 40 years under renewed licenses and are
following their approved aging management programs. The NRC's oversight
has confirmed that licensees are effectively detecting and resolving
aging issues in a timely manner. The NRC has not identified any
fundamental gaps in the understanding of aging that licensees have not
been able to adequately address through the current renewal process or
other existing regulatory processes.
The NRC has long recognized that issuing a renewed license for a
full 40-year term is legally consistent with the terms of the AEA. When
the NRC issued the 1991 rule, it considered two approaches to renewing
licenses: (1) a ``tack-on'' license, which would take effect at the end
of the current OL term, and (2) a ``supersession'' license, which would
be immediately effective upon renewal and would add a number of years,
at the time up to 20, to the existing license. The NRC chose to use the
supersession approach. At that time, the NRC also noted that future
efforts to extend the 20-year limit closer to the 40-year maximum would
require a reappraisal of the use of supersession licensing. The
supersession license approach that the Commission currently uses never
results in a license greater than 40 years because the renewal term is
20 years and licensees currently may not apply for renewal until there
are 20 years or less remaining on the license. In contrast, a renewed
license that includes any number of years remaining on the existing
license plus a 40-year extension would necessarily extend the term of
the renewed license past the limit in section 103 of the AEA.
Therefore, the NRC concludes that granting a 40-year extension would
require the use of a tack-on license. Moreover, as the NRC noted when
it initially promulgated its license renewal rules, a tack-on license
may be more compatible with the terms of the AEA that state that
licenses ``may be renewed upon the expiration'' of the license term.
(Id. at 64964).
Maintaining a tack-on license may require additional administrative
steps, particularly if it is issued well before the beginning of the
period of extended operation. Therefore, these amendments to 10 CFR
part 54 would also include a requirement that the licensee take steps
to ensure the renewed license remains up to date, consistent with the
current licensing basis as defined in 10 CFR 54.3, prior to the period
of extended operation. Also, currently, certain activities are
implemented prior to the period of extended operation in accordance
with the conditions of the renewed license that is immediately in
effect upon issuance, such as implementing new aging management
programs and enhancing existing programs. The NRC's oversight program
verifies that these activities are completed, including taking
advantage of the plant's refueling outage prior to the period of
extended operation to observe licensee activities that take place at
reduced power levels. The NRC is considering alternative means of
conditioning these activities, given that a tack-on license would not
be in effect until the start of the period of extended operation, and
the NRC is seeking the public's input on this topic.
With the change to a longer renewal term, applicants may need to
address potential gaps in the current safety and environmental review
guidance until the guidance is updated. For example, the current
revision of NUREG-2191, ``Generic Aging Lessons Learned for Subsequent
License Renewal (GALL-SLR) Report,'' was based on evaluations of up to
one term of subsequent license renewal under the current regulation, or
80 total years of plant operation. Likewise, NRC regulations at subpart
A of appendix B to 10 CFR part 51 codify the conclusions of NUREG-1437,
``Generic Environmental Impact Statement for License Renewal of Nuclear
Plants,'' (LR GEIS) for one initial term of renewal and one term of
subsequent renewal. However, the NRC staff noted in SECY-22-0109,
``Proposed Rule: Renewing Nuclear Power Plant Operating Licenses--
Environmental Review,'' dated December 6, 2022, that, when it provided
the draft LR GEIS to the Commission, the underlying analysis of
environmental issues in that document ``could apply to any license
renewal term,'' even one beyond the first subsequent license renewal
term.
However, with the proposed change to allow license extensions of 40
years, plants currently operating in their initial renewal period (40
to 60 years) could apply for a 40-year subsequent renewal that would
bring the plant to a total of 100 years of operation. Similarly,
licensees that will be operating in their approved subsequent renewal
period (60 to 80 years under existing regulations) could apply for a
third 40-year renewal that would bring the plant to a total of 120
years of operation. As a result, applicants seeking to operate beyond
80 years would need to consider whether to supplement the guidance to
address any aging and environmental effects specific to the longer
operating life. No such gap analysis would be needed for applicants for
an initial renewal, extending operation from 40 to 80 years, as the
current subsequent renewal guidance readily applies to 80 total years
of plant operation. For environmental effects, because the analysis in
the LR GEIS is term neutral, licensees could consider using the LR GEIS
as a starting point for that environmental analysis (although for
environmental impacts past year 80, that analysis is not codified in 10
CFR part 51). The NRC has prepared draft
[[Page 44601]]
interim staff guidance, LR-ISG-2026-01, ``Updated Review Criteria for
License Renewal,'' on this topic as part of this rulemaking. The NRC is
also seeking the public's input on this topic.
B. Alternative Risk-Informed and Performance-Based Criteria
The current license renewal rule is often described as a
deterministic rule because it prescribes the specific structures and
components whose function must be demonstrated to be maintained. Under
10 CFR 54.21(a)(1), in-scope structures and components are included in
the aging management review if they perform their function without
moving parts or without a change in configuration or properties (i.e.,
they are ``passive'') and if they are long-lived. In addition, 10 CFR
54.21(a)(3) requires applicants to demonstrate that the effects of
aging will be adequately managed so that intended function(s) will be
maintained for ``each structure and component identified in'' 10 CFR
54.21(a)(1). These regulations do not explicitly allow applicants to
use risk insights to weigh the importance of the need to evaluate aging
for license renewal.
When the NRC issued the 1995 license renewal rule, it considered
allowing risk-informed methods, specifically PRA. The NRC decided
against it due to concerns about the quality of risk data and models at
that time. However, in SRM-SECY-98-144, ``Staff Requirements--SECY-98-
144--White Paper on Risk Informed and Performance-Based Regulation,''
dated March 1, 1999, the Commission defined its expectations for risk-
informed and performance-based regulation. Since then, the agency has
encouraged the use of such approaches to improve regulatory decision-
making, enhance safety, and reduce unnecessary regulatory burden. There
has since been widescale adoption of risk initiatives such as 10 CFR
50.69 and risk-informed technical specification completion times and
surveillance frequencies that require PRAs to have a high level of
acceptability. Since 1995, PRA standards have matured, data collection
has improved, and NRC reviews of these activities have strengthened.
As a result, the NRC is proposing to add 10 CFR 54.21(a)(4) to
allow applicants to voluntarily propose risk-informed and performance-
based alternatives to the current prescriptive requirements of the
aging management review. Applicants could use these alternative
criteria to decide which structures and components need an aging
management review and how to show that aging effects will be managed.
Similarly, the NRC is also proposing to revise 10 CFR 54.29,
``Standards for issuance of a renewed license,'' to clarify that the
NRC's safety finding could rely on risk insights in assessing the
adequacy of an applicant's actions, or lack thereof, to maintain the
functionality of structures and components.
The purpose of the license renewal rule is to ensure the
preservation of intended functions during long-term operations for
which the understanding of aging is more limited. The NRC focused the
aging management requirements on passive structures and components
because they generally lack performance and condition indicators that
can be easily monitored. Although operating experience shows that
passive components are typically more reliable than active ones, and
PRA often assigns them a low contribution to plant risk, this does not
guarantee their future performance over extended operating periods. New
degradation mechanisms may emerge, existing mechanisms may progress
faster than expected, or safety margins may be reduced. In addition,
common cause aging issues that impact overall system function may
emerge.
For these reasons, alternative risk-informed or performance-based
criteria could not inappropriately weaken the capability of aging
management programs to preserve intended functions over long-term
operations. These criteria could not be used to eliminate aging
management review requirements solely because past operating experience
suggests passive structures and components are reliable. The criteria
would need to be applied carefully to preserve the intent of 10 CFR
part 54. The NRC has prepared draft interim staff guidance, LR-ISG-
2026-01, ``Updated Review Criteria for License Renewal,'' on this topic
as part of this rulemaking.
C. Eliminate the Limitation on Early Application Submittal
The current regulation in 10 CFR 54.17(c) states that a licensee
cannot apply for renewal more than 20 years before its current license
expires. When the NRC issued the 1995 license renewal rule, it stated
that the purpose of this limitation was to ensure that plants had
sufficient operating experience to reveal any aging concerns before
applying for renewal. At the same time, this 20-year limit was also
intended to give utilities enough time to plan for alternatives, like
replacing a plant, if the license was not renewed.
The NRC is proposing to remove the 20-year limit. Instead, the
proposed regulations in 10 CFR 54.17(c) would permit licensees to apply
for a renewed license anytime within their current operating period.
For example, licensees could apply for initial renewal anytime within
their initial 40-year license, and they could apply for subsequent
renewal anytime within the period of extended operation. Removing the
20-year limit would give licensees more flexibility, such as
coordinating their renewal applications with other licensing actions at
the same plant or across their fleet to make the process more
efficient. This change would not change the timely renewal provisions
in 10 CFR 2.109.
The NRC has already granted exemptions from this regulation to
several plants, based on the availability of operating experience from
similar facilities (i.e., similar materials and service environments)
to inform the renewal decision. At this time, most U.S. plants have
been operating past 40 years, and many for more than 50 years. As a
result, there is a large amount of industry-wide operating experience
to inform the NRC's renewal reviews, even if a specific plant has been
operating for a short time. Also, the regulatory process will continue
to provide the NRC with the means to verify that plants are effectively
addressing emergent aging issues after a license is renewed. As
described in NUREG-2191, Appendix B, ``Operating Experience for Aging
Management Programs,'' holders of renewed licenses should have
processes to capture and review operating experience to assess the need
to enhance their aging management programs, as appropriate.
The NRC understands that newer reactor designs might use unique
materials in new service environments, so they may not immediately have
a complete base of operating experience. If licensees with newer
designs request a renewed license early in plant life, the NRC's review
guidance (e.g., GALL Report) might not fully apply in some areas. In
those cases, applicants would be expected to explain how they will
address any unique issues and uncertainties in their demonstration of
adequate aging management.
D. Eliminate the Required Application Content on Exemptions
The current regulation in 10 CFR 54.21(c)(2) requires that the
application for a renewed license include a list of plant-specific
exemptions granted under 10 CFR 50.12 that are based on time-limited
aging analyses (TLAAs), along with a justification for continuing those
exemptions during the extended operating period. When the NRC issued
the 1995 rule, it explained that this
[[Page 44602]]
requirement was important because it was necessary to make an
independent assessment that all exemptions based on TLAAs had been
evaluated as part of the license renewal process.
However, the NRC's experience in reviewing applications has found
that this requirement leads to duplication of submitted information and
adds unnecessary burden for applicants in their preparations to support
the NRC's application review. The regulation in 10 CFR 54.21(c)(1)
already requires applicants to identify and evaluate all TLAAs, whether
or not they involve an exemption. Repeating the same information to
meet the separate 10 CFR 54.21(c)(2) requirement does not aid the NRC's
review. As a result, the NRC is proposing to remove 10 CFR 54.21(c)(2)
to streamline the process and reduce the burden on applicants.
E. Eliminate the Required Application Content on Technical
Specifications
The current regulation in 10 CFR 54.22, ``Contents of application--
technical specifications,'' requires that applicants include and
justify any technical specification changes needed to manage the
effects of aging. However, based on its experience in reviewing renewal
applications, the NRC has found that this requirement is not necessary.
As the renewal process is meant to maintain a plant's current licensing
basis (not change it), technical specification changes are very rarely
needed. Also, even without this explicit application content
requirement, the review process is capable of ensuring that any needed
technical specification changes are identified. As a result, the NRC is
proposing to remove 10 CFR 54.22. In addition, this regulation would no
longer be cited in 10 CFR 54.9, ``Information collection requirements:
OMB approval,'' and 10 CFR 54.43, ``Criminal penalties.''
F. Reduce Ongoing, Post-Renewal Updates to the Final Safety Analysis
Report
The license renewal rule requires applicants to include a
supplement to the plant's final safety analysis report (FSAR). This
supplement must provide a summary description of the programs and
activities credited for managing the effects of aging and the
evaluation of TLAAs. In addition, after the NRC issues a renewed
license, the regulation in 10 CFR 54.37(b) requires a licensee to keep
the FSAR updated. These required updates are to ``include systems,
structures, and components newly identified'' and ``describe how the
effects of aging will be managed.'' The NRC Regulatory Issue Summary
(RIS) 2007-16, Revision 1, ``Implementation of the Requirements of 10
CFR 54.37(b) for Holders of Renewed Licenses,'' dated April 28, 2010,
explained the definition of ``newly identified'' systems, structures,
and components (SSCs) as those that are either related to a change to
the plant's licensing basis or were already installed during the
original license renewal review but were not included in that review.
When the NRC issued the 1995 license renewal rule, some commenters
pointed out that 10 CFR 54.37(b) requires more detail in the ongoing
FSAR updates than the summary description provided in the original
renewal application. At that time, the NRC justified the discrepancy on
the grounds that, after a renewed license is issued, the ongoing FSAR
updates serve dual purposes. They function as both the renewal
application (in that they describe how the licensee evaluated the
effects of aging for newly identified SSCs) and the FSAR supplement
(establishing appropriate regulatory controls on aging management
programs).
However, based on its oversight experience, the NRC believes the
current level of detail in the required FSAR update is not necessary,
provided that licensees keep details of their evaluation of newly
identified SSCs available for NRC inspection. As a result, the NRC is
proposing to revise 10 CFR 54.37(b) to only require a summary
description of aging management activities, consistent with what is
required in the renewal application. Under this change, a newly
identified SSC would trigger an FSAR update only if it affects the
summary description of aging management programs or TLAAs. However,
regardless of whether an FSAR update is needed, the licensee would be
required to keep the full evaluation of newly identified SSCs available
in a form that could be audited and retrieved.
Finally, the other currently proposed revisions to 10 CFR part 54
(described in sections XXXII.A. and C. of this document) would change
how licensees would need to address some physical plant additions or
modifications made prior to the start of the period of extended
operation. To date, newly installed SSCs have not been considered to be
subject to 10 CFR 54.37(b), based on the idea that they would not be in
service for more than 40 years during the extended license period. In
its responses to comments on draft RIS-2007-16, Revision 1,
``Implementation of the Requirements of 10 CFR 54.37(b) for Holders of
Renewed Licenses'' dated May 27, 2009, the NRC explained that plant
operating programs and NRC regulatory activities are adequate to
address aging during the first 40 years of service, and aging
management is only necessary when that service life is exceeded.
However, with the other proposed changes to 10 CFR part 54--extending
the renewal term from 20 to 40 years with a tack-on license and
removing the 20-year limit on early applications--SSCs newly installed
after the NRC completes its license renewal review, but prior to start
of the period of extended operation, could significantly exceed 40
years of service during the period of extended operation. Because of
that, any of these SSCs that are installed before the start of the
period of extended operation would be subject to the provisions of 10
CFR 54.37(b).
XXXIII. Background--Enhancing Flexibility of Reactor Site Criteria
The NRC acknowledges the importance of site characterization in the
determination of site suitability to demonstrate reasonable assurance
of adequate protection of public health and safety, particularly as it
relates to external hazards with broad impacts and the potential to
cause failure of all preventative and mitigative controls with one
initiating event. Site investigations performed to determine the site-
specific external hazards and to adequately characterize the geology,
seismology, meteorology, and hydrology of a proposed power reactor site
can be time consuming and expensive. The NRC further recognizes that
the emergence of new reactor technologies, including non-stationary
reactors, and alternate fuels may justify greater flexibility within
the siting requirements, and site investigations performed to meet
those requirements.
In the current regulatory framework, 10 CFR part 100, ``Reactor
Site Criteria,'' consists of two subparts: subpart A, ``Evaluation
Factors for Stationary Power Reactor Site Applications Before January
10, 1997 and for Testing Reactors,'' and subpart B, ``Evaluation
Factors for Stationary Power Reactor Site Applications on or After
January 10, 1997.'' The requirements in subpart B to 10 CFR part 100
are more prescriptive than the requirements of subpart A to 10 CFR part
100 and impose an increased burden to obtain the required information.
For example, subpart A of 10 CFR part 100 requires an applicant to
consider the seismology, meteorology, geology, and hydrology of a site,
but the requirements in subpart B of 10 CFR part 100 demand greater
specificity, such as maximum probability wind speed, data on site
[[Page 44603]]
foundation materials, earthquake recurrence rates, and groundwater
velocity. Within the current framework, all power reactor applicants,
even those for new reactor designs with smaller physical footprints,
potentially lower radiological risk, and enhanced safety features,
would perform the same site characterization and hazard analysis under
the requirements in subpart B to 10 CFR part 100.
Subpart B to 10 CFR part 100 and the supporting guidance on
acceptable site characterization methods impose a higher burden because
they were developed primarily with large LWR sites in mind and were
informed by the scale and potential radiological consequences of these
facilities. Conversely, subpart A to 10 CFR part 100, including
appendix A, contains less prescriptive requirements applicable to older
LWR sites not reflective of updated method of external hazards analysis
or testing reactors with a lower potential radiological consequence.
Applying this regulatory framework and associated site characterization
guidance as it currently exists to new reactor designs may not reflect
a risk-informed, performance-based approach to site characterization
that fully considers the site, design, and hazard-specific conditions
at a prospective site.
Therefore, the NRC is proposing changes to these criteria to
introduce a more risk-informed, performance-based approach to siting.
The proposed changes to the regulations would clearly define a graded
approach to site characterization for nuclear power reactor applicants,
including non-stationary reactors, by introducing a tiered approach in
the regulatory framework that would allow a lower consequence power
reactor applicant to demonstrate site suitability through an
appropriately scaled site investigation rather than the potentially
more burdensome requirements in the current subpart B to 10 CFR part
100. Further, the draft guidance documents issued for public comment
along with this proposed rule would provide a risk-informed approach
that could be leveraged by reactor applicants under any licensing
pathway to optimize their site characterization to achieve an
acceptable level of risk for external hazards. This proposed approach
is consistent with other recent rulemaking actions, such as the 10 CFR
part 53 rulemaking, and would provide similar requirements for
applicants across licensing frameworks.
The NRC has a long-standing preference to site reactors in areas of
low population density but recognizes that safety, environmental,
economic, or other factors may justify siting nuclear plants in areas
with greater population densities. Therefore, the NRC proposes to
revise its regulations in subparts A and B of 10 CFR part 100 to allow
reactors to be sited in areas with greater population densities when
justified by an assessment that compares the societal risks and
societal benefits of siting reactors in those areas. This proposed
change is consistent with the recent 10 CFR part 53 rulemaking and
would provide similar requirements for applicants across licensing
frameworks.
XXXIV. Discussion--Enhancing Flexibility of Reactor Site Criteria
The proposed changes to 10 CFR part 100 would allow power reactor
applicants under 10 CFR part 50 or 52 to have greater flexibility to
determine the appropriate level of site characterization. Specifically,
these proposed changes would utilize existing flexibility in the
regulatory framework for licensing testing and power reactors,
including non-stationary reactors, by allowing certain lower
consequence power reactor applicants to use the same siting criteria
used for testing reactors under subpart A to 10 CFR part 100.
Additionally, these proposed changes would remove outdated information
on a deterministic approach to seismic hazards analysis in appendix A
to 10 CFR part 100 that no longer applies to new reactor applications.
The proposed changes to 10 CFR 100.10, ``Factors to be considered when
evaluating sites,'' would streamline the regulation and be accompanied
by guidance on how to meet the regulatory requirements for subpart A to
10 CFR part 100. The proposed title change to subpart B to 10 CFR part
100 would retain the more prescriptive regulatory requirements that
would remain applicable to higher consequence power reactor applicants.
These more prescriptive regulatory requirements would continue to
ensure appropriate siting criteria for this class of applications. The
existing titles of subparts A and B to 10 CFR part 100 are currently
reflected in 10 CFR 50.34(a) and (b). Accordingly, conforming changes
to those sections would be needed to align the proposed changes in 10
CFR part 100 with the requirements for content of applications
described in 10 CFR 50.34(a) and (b). These conforming changes include
updating 10 CFR 50.34(a) and (b) to remove references to applications
on or after January 10, 1997, and stationary power reactors.
Specific proposed changes to 10 CFR part 100 implementing this
approach are described below. The title of subpart A to 10 CFR part 100
would be changed to ``Subpart A--Evaluation Factors for Tier 1 Power
and Testing Reactors.'' Appendix A to 10 CFR part 100 would be deleted
in its entirety as it provides a description of an approach to
determine the safe shutdown earthquake that is outdated and is not
consistent with the current practice for performing a probabilistic
seismic hazards analysis at nuclear power plant sites. The information
necessary to perform an acceptable probabilistic seismic hazards
analysis would be included in guidance. The proposed revisions to 10
CFR 100.3, ``Definitions,'' would be conforming changes to define the
new terms introduced in proposed revisions to 10 CFR part 100: ``Tier 1
reactor'' and ``Tier 2 reactor.'' These terms would be defined based on
an unmitigated consequence. The proposed revision to 10 CFR 100.8,
``Information collection requirements: OMB approval,'' would be a
conforming change to remove the mention of appendix A, which would be
deleted. The proposed revisions to 10 CFR 100.10 would remove the
references to appendix A to 10 CFR part 100 and delete explanatory
text. That explanatory text would instead be contained in the draft
guidance documents issued for public comment along with this proposed
rule. With the deletion of appendix A to 10 CFR part 100, a new
reference to appendix S to 10 CFR part 50 would be added to 10 CFR
100.10(b) to provide earthquake engineering criteria for applicants
pursuing siting under subpart A to 10 CFR part 100. Additional changes
to 10 CFR 100.11, ``Determination of exclusion area, low population
zone, and population center distance,'' related to the exclusion area,
low population zone, and population center distance are being proposed
to align with the dose reference values in subpart B to 10 CFR part
100, which references 10 CFR 50.34(a)(1). Other proposed changes to the
text describing the consequence analysis in 10 CFR 100.11 would align
with proposed revisions to the text in 10 CFR 50.34(a)(1)(ii)(D) to be
more technology-inclusive and meaningful for designs with functional
containments. The proposed rule would also revise footnote 1 to 10 CFR
100.11, which provides additional information on the fission product
release, to be more technology inclusive and allow for the evaluation
of designs with mechanistic source terms and functional containments.
In addition, the proposed rule would remove outdated information from
footnote 2 to 10 CFR
[[Page 44604]]
100.11, which discusses the use of 25 rem (0.25 Sv) TEDE as a reference
value.
These proposed changes to subpart A to 10 CFR part 100 would enable
power reactor applicants that meet the Tier 1 entry criterion for
subpart A to 10 CFR part 100 to develop the appropriate level of site
characterization information to support the permit or license
application. The proposed entry criterion for a Tier 1 reactor would be
to demonstrate an unmitigated consequence of less than 25 rem (0.25 Sv)
TEDE at the site exclusion area boundary. The draft guidance documents
issued for public comment along with these proposed rule changes would
provide guidance on how to determine the unmitigated consequence,
select the site parameters to include when defining a site parameter
envelope and apply graded approaches for seismic and other external
hazard characterization for any power reactor applicant. The draft
guidance documents also would describe appropriate uses of existing
site characterization information and alternative site investigation
techniques, which would increase regulatory clarity and review
efficiency.
Proposed entry criteria for subpart A to 10 CFR part 100 and
guidance on graded approaches for external hazards with the potential
to impact all safety controls with a single initiating event (e.g.,
seismic) would be based on an unmitigated consequence analysis. The
unmitigated consequence analysis would be necessary to account for the
common-cause failure aspect of external hazards and to determine the
appropriate target performance criteria for acceptable risk.
Information on appropriate methods for performing unmitigated
consequence analyses will be provided in the draft guidance documents
being issued for public comment along with this proposed rule.
Additionally, an appendix within the draft guidance would provide
considerations for optimizing seismic design bases using the
unmitigated consequence analysis. The NRC recognizes that the seismic
design basis for a nuclear facility often has a significant impact on
cost and schedule. This appendix would enable applicants to optimize
the seismic design basis, reducing the burden of seismic design while
still ensuring safety.
With the proposed introduction of the concept of a Tier 1 reactor
in subpart A to 10 CFR part 100, the title to subpart B to 10 CFR part
100 would be changed to ``Subpart B--Evaluation Factors for Tier 2
Power Reactor Site Applications.'' The NRC proposes to make additional
changes to 10 CFR 100.20, ``Factors to be considered when evaluating
sites,'' and 100.23, ``Geologic and seismic siting criteria,'' to
reflect the new ``Tier 2'' terminology. Tier 2 reactors would be power
reactors that have not been demonstrated to meet the Tier 1 entry
criterion. This would include power reactors with an unmitigated
consequence of greater than 25 rem (0.25 Sv) TEDE at the site exclusion
area boundary or where the unmitigated consequence is undetermined. No
other changes are proposed to subpart B as part of this rulemaking, and
the framework would be maintained to accommodate higher consequence
reactor applications.
The NRC proposes to maintain its long-standing preference for
siting reactors in areas of low population density, while providing
flexibility to allow siting reactors in areas of greater population
density when warranted. The NRC recognizes that safety, environmental,
economic, or other factors may justify siting nuclear plants in areas
with higher population densities or within a densely populated center
containing more than about 25,000 residents. Therefore, the NRC is
proposing to revise 10 CFR 100.10, 100.11(a)(3), 100.21(b), and
100.21(h) to allow applicants to justify siting reactors at such sites
by performing 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) and comparing those risks to the societal benefits of a
specific proposed site (e.g., ability to use existing infrastructure
for a retired fossil fuel power plant). Implementing guidance for these
assessments will be developed.
The proposed changes to 10 CFR part 100 and the explanations
provided in the draft guidance documents issued for public comment
along with this rulemaking could result in timelier and more efficient
site investigations, more appropriate seismic design bases, and more
streamlined reviews for new power reactor applications.
XXXV. Background--Increased Enrichment of Conventional and Accident
Tolerant Fuel Designs for Light-Water Reactors
A. Accident Tolerant Fuels
Accident tolerant fuels (ATFs) are advanced nuclear fuel
technologies that have the potential to enhance safety at U.S. nuclear
power plants by offering better performance during normal operation,
transient conditions, and accident scenarios. Section 107, ``Commission
Report on Accident Tolerant Fuel,'' of the Nuclear Energy Innovation
and Modernization Act defines ATF as a new technology that makes an
existing commercial nuclear reactor more resistant to a nuclear
incident (as defined in section 11 of the AEA (42 U.S.C. 2014)) and
lowers the cost of electricity over the licensed lifetime of an
existing commercial nuclear reactor.
While this proposed rule would not make any fuel-design-specific
conclusions, it would facilitate the adoption of increased enrichment
and higher burnup to enable entities, licensed to use conventional or
ATF designs, to use LWR fuel containing uranium enriched to greater
than 5.0 weight percent uranium-235 (U-235), in most cases, without the
use of exemptions.
B. Rulemaking Development
(i) Early Considerations of Increased Enrichment
From interactions with stakeholders, the NRC is aware that
licensees and applicants plan to request to deploy ATF concepts and
operate fuel to extract more energy from each fuel rod (i.e., operate
fuel at higher burnups). Specifically, the NRC expects requests to
raise fuel burnup limits to higher than the 62 gigawatt-days per metric
ton of uranium (GWd/MTU) rod-average burnup limit set by the NRC in
most safety analysis methodologies. To achieve higher burnup limits,
licensees and applicants would need to request increases in fuel
enrichment from the current standard of 5.0 weight percent U-235 up to
approximately 10.0 weight percent U-235. Additionally, in February
2019, nuclear power industry representatives identified potential
advantages of increased enrichment fuel for LWRs in the NEI white
paper, ``The Economic Benefits and Challenges with Utilizing Increased
Enrichment and Fuel Burnup for Light-Water Reactors.''
In September 2021, the NRC issued version 1.2 of the ``Project Plan
to Prepare the U.S. Nuclear Regulatory Commission for Efficient and
Effective Licensing of Accident Tolerant Fuels,'' which describes the
pursuit of higher burnup and increased enrichment as key components of
nuclear power industry ATF efforts. Currently, the industry plans to
deploy batch loads of fuels enriched to levels greater than the current
standard of 5.0 weight percent U-235 by the mid-to-late 2020s.
Paragraph (b)(7) of 10 CFR 50.68, ``Criticality accident
requirements,'' requires that U-235 enrichment levels in power reactor
fuel be no more than 5.0
[[Page 44605]]
percent by weight, unless the NRC approves exemptions from this limit.
The development of the current regulatory framework did not foresee use
of enrichments greater than 5.0 weight percent U-235. The NRC
established the current weight percent limits as reasonable bounding
assumptions for its safety analysis methodologies. As part of this
rulemaking, the NRC evaluated the framework and considered whether the
current weight percent limits can be adjusted while maintaining
reasonable assurance of adequate protection of public health and
safety. In addition, the NRC considered whether rulemaking would
support a more efficient review of licensing actions.
In response to industry interest in LWR fuels enriched to between
5.0 to 10.0 weight percent U-235, the NRC staff submitted to the
Commission a rulemaking plan in SECY-21-0109, ``Rulemaking Plan on Use
of Increased Enrichment of Conventional and Accident Tolerant Fuel
Designs for Light-Water Reactors,'' dated December 20, 2021. The staff
requested Commission approval to initiate rulemaking to amend NRC
requirements to facilitate the use of LWR fuel containing uranium
enriched to greater than 5.0 weight percent U-235. In SECY-21-0109, the
staff recommended rulemaking to reduce the number of exemption requests
and facilitate increased regulatory efficiency and consistency. The
staff explained that rulemaking on this topic would allow the staff to
thoroughly review the potential regulatory implications of fuels
enriched to greater than 5.0 weight percent U-235 and identify and
assess the potential costs and benefits of changing regulatory
requirements that impact the use of these fuels. Rulemaking also would
provide options for a generic resolution of these issues and invite
stakeholder participation in decisions affecting this regulatory area,
rather than deciding issues on a case-by-case basis as in the current
regulatory framework.
In SRM-SECY-21-0109, ``Staff Requirements--SECY-21-0109--Rulemaking
Plan on Use of Increased Enrichment of Conventional and Accident
Tolerant Fuel Designs for Light-Water Reactors,'' dated March 16, 2022,
the Commission approved the staff's plan to initiate rulemaking to
amend requirements for the use of LWR fuel containing uranium enriched
to greater than 5.0 weight percent U-235. The Commission stated that
the provisions of the rule should apply only to high-assay, low-
enriched uranium (HALEU) fuel, both for nonproliferation and safeguards
reasons, and that the staff's analysis should focus on the range of
enrichment most likely to be contemplated in future applications.
In addition, the Commission directed that (1) fuel fragmentation,
relocation, and dispersal (FFRD) issues relevant to fuels of higher
enrichment and burnup levels should be appropriately addressed and
analyzed in the regulatory basis for this rulemaking; and (2) staff
should take a risk-informed approach when developing this rule and the
associated regulatory basis and guidance.
(ii) Other Considerations
On April 11, 2024, the Commission returned the 10 CFR 50.46c draft
final rule to the staff in SRM-SECY-16-0033, ``Staff Requirements--
SECY-16-0033--Draft Final Rule--Performance-Based Emergency Core
Cooling System Requirements and Related Fuel Cladding Acceptance
Criteria.'' As a result, this rulemaking proposes to leverage the
previously proposed performance-based approach to emergency core
cooling system (ECCS) requirements, including the expanded
applicability to advanced fuels, and incorporate the embrittlement
research findings from the 10 CFR 50.46c draft final rule into the new
voluntary 10 CFR 50.46a proposed rule. This is discussed in sections
XXXV.C.(v)(c), ``10 CFR 50.46c Rulemaking and Cladding Embrittlement
Research Findings,'' and XXXVI.F.(v), ``Alternative ECCS Analysis
Requirements and Acceptance Criteria,'' of this document.
C. Background and History of Affected Regulations
This section provides the regulatory history and the background of
each of the affected regulatory areas and its relationship to fuel
enrichment.
(i) 10 CFR 70.24 and 10 CFR 50.68 Criticality Accident Requirements
One regulation initially evaluated by the NRC for its relationship
to fuel enrichment is 10 CFR 70.24, ``Criticality accident
requirements.'' This regulation was established as part of the
rulemaking for 10 CFR part 70, ``Domestic Licensing of Special Nuclear
Material,'' in 1974 (39 FR 39020; November 5, 1974). Section 70.24
ensures that licensees handling special nuclear material (SNM) have the
appropriate monitoring systems in place for detecting a criticality
accident and establish an emergency plan.
During the 1980s and 1990s, nuclear power reactor licensees sought
exemptions from several requirements in 10 CFR 70.24, including those
related to active criticality monitoring alarm capabilities,
criticality emergency drills, and plans for evacuations,
decontamination, medical treatment for radiation exposure, and reentry
protocols, which must also be demonstrated through the regular
criticality safety drills. As discussed in SECY-97-155, ``Staff's
Action Regarding Exemptions from 10 CFR 70.24 for Commercial Nuclear
Power Plants,'' dated July 21, 1997, in response to these numerous
exemption requests, the staff evaluated the likelihood of an
inadvertent criticality accident during fuel handling operations at
nuclear power plants and concluded that such events would be unlikely
for power reactor facilities because existing administrative and design
controls were based on no more than 5.0 weight percent U-235 fuel
enrichment. As a result, in 1998 the NRC issued 10 CFR 50.68 to offer
these licensees an alternative to 10 CFR 70.24, provided that they meet
certain criteria (63 FR 63127; November 12, 1998).
Under 10 CFR 50.68, licensees may decide not to comply with the
requirements of 10 CFR 70.24 and, instead, must comply with the
requirements in 10 CFR 50.68(b), if they can demonstrate appropriate
subcriticality margins that are expressed in terms of k-effective
(keff), the estimated ratio of neutron production to neutron absorption
and leakage, for new and fresh fuel storage facilities. Section
50.68(b) specifies the following keff limits:
10 CFR 50.68(b)(2): keff <= 0.95 at 95 percent probability
and 95 percent confidence level for fresh fuel wet storage racks in
unborated water.
10 CFR 50.68(b)(3): keff <= 0.98 at 95 percent probability
and 95 percent confidence level for fresh fuel storage in low-density
hydrogenous fluid (fog). This keff limit is only applicable to the
current fleet of operating LWRs if they use a dry new fuel storage
vault that is susceptible to moderation by natural weather or fire
suppression fogging conditions. If new fuel is stored submerged in the
fuel pool (i.e., 10 CFR 50.68(b)(2) wet storage conditions), then 10
CFR 50.68(b)(3) does not apply.
10 CFR 50.68(b)(4): keff <= 0.95 at 95 percent probability
and 95 percent confidence level for spent fuel wet storage racks in
unborated water.
10 CFR 50.68(b)(4): keff <= 0.95 at 95 percent probability
and 95 percent confidence level for spent fuel wet storage racks in
borated water; furthermore, keff must remain < 1.0 at 95 percent
probability and 95 percent confidence level without soluble boron
credit taken.
[[Page 44606]]
In addition to the keff requirements, 10 CFR 50.68(b)(6) requires
radiation monitoring only during fuel handling and movement. Finally,
10 CFR 50.68(b)(7) limits current applications to 5.0 weight percent U-
235 for fresh fuels. The NRC added 10 CFR 50.68(c) in 2006 (71 FR
66648; November 16, 2006) to delineate independent spent fuel storage
installation dry storage cask loading applicability.
Most operating reactors currently adhere to 10 CFR 50.68(b)
requirements in lieu of the criticality monitoring requirements in 10
CFR 70.24. When fuel transition changes are made, if enrichments are
being increased, the licensee must perform a safety analysis of spent
fuel pool criticality as part of the fuel transition because increasing
enrichments may cause a decrease in margin to the applicable keff
safety limits of 10 CFR 50.68(b)(4). There is also a plant-specific
criticality safety limit for the new and spent fuel storage areas
listed in Standard Technical Specification Design Feature 4.3, ``Fuel
Storage,'' in NUREG-1430, ``Standard Technical Specifications--Babcock
and Wilcox Plants''; NUREG-1431, ``Standard Technical Specifications--
Westinghouse Plants''; NUREG-1432, ``Standard Technical
Specifications--Combustion Engineering Plants''; NUREG-1433, ``Standard
Technical Specifications--General Electric Plants (BWR/4)''; NUREG-
1434, ``Standard Technical Specifications--General Electric Plants BWR/
6''; and NUREG-2194, ``Standard Technical Specifications--Westinghouse
Advanced Passive 1000 (AP1000) Plants.'' The guidance in these NUREGs
assists the NRC's review of any proposed fuel changes that could affect
10 CFR 50.68 compliance.
Nonetheless, given that 10 CFR 50.68(b)(7) limits current
applications to 5.0 weight percent U-235, if reactor licensees
transition to using fuel enriched above 5.0 weight percent U-235, then
10 CFR 50.68 would not be available as an alternative to 10 CFR 70.24.
Absent rulemaking, the NRC expects that reactor licensees seeking to
transition to fuel enriched above 5.0 weight percent U-235 would likely
request exemptions from 10 CFR 50.68.
Unlike 10 CFR 50.68, 10 CFR 70.24 does not have an enrichment
limit, so a licensee or applicant could comply with 10 CFR 70.24 and
implement enrichments beyond 5.0 weight percent U-235. Licensees or
applicants complying with 10 CFR 70.24 that seek exemption from that
section's active criticality monitoring and emergency planning
requirements could not justify the exemptions with the subcriticality
margin approach methodology used in 10 CFR 50.68. Furthermore, under 10
CFR 70.24(d)(2), any exemption from 10 CFR 70.24 held by a licensee
becomes ineffective once that licensee elects to comply with 10 CFR
50.68. Industry stakeholders have also acknowledged the greater
regulatory flexibility of the 10 CFR 50.68 subcriticality margin
approach, and licensees have expressed interest in applying the 10 CFR
50.68 methodology to higher enrichments in lieu of meeting the 10 CFR
70.24 criticality safety approach. Proposed amendments to address this
issue are discussed in section XXXVI.A., ``Criticality Accident
Requirements in 10 CFR 50.68,'' of this document.
(ii) Environmental Requirements in 10 CFR 51.51 and 51.52
In 10 CFR 51.51, table S-3, ``Table of Uranium Fuel Cycle
Environmental Data,'' references the original environmental assessments
in WASH-1248, ``Environmental Survey of the Uranium Fuel Cycle,''
issued April 1974, and NUREG-0116, ``Environmental Survey of the
Reprocessing and Waste Management Portions of the LWR Fuel Cycle: A
Task Force Report,'' Supplement 1 to WASH-1248, issued October 1976.
In 10 CFR 51.52, summary table S-4, ``Environmental Impact of
Transportation of Fuel and Waste to and From One Light-Water-Cooled
Nuclear Power Reactor,'' references the environmental assessments in
WASH-1238, ``Environmental Survey of Transportation of Radioactive
Materials to and from Nuclear Power Plants,'' issued December 1972, and
Supplement 1 to NUREG-75/038, ``Environmental Survey of Transportation
of Radioactive Materials to and from Nuclear Power Plants,'' issued
April 1975.
The aforementioned environmental surveys were based on enrichments
up to 4.0 weight percent U-235. Subsequent analysis by the NRC,
incorporated into section 4.12.1.1 of NUREG-1437, ``Generic
Environmental Impact Statement for License Renewal of Nuclear Plants:
Final Report,'' Revision 1, issued June 2013, confirmed that table S-3
of 10 CFR 51.51 and table S-4 of 10 CFR 51.52 are also bounding for
enrichments up to 5.0 weight percent U-235. The NRC does not have an
approved assessment of environmental impacts related to the uranium
fuel cycle in current 10 CFR 51.51, or to transportation of enriched
fresh, nonirradiated fuel to a reactor in current 10 CFR 51.52, for
enrichments greater than 5.0 weight percent U-235.
Under 10 CFR 51.51 and related to 10 CFR 51.50, ``Environmental
report--construction permit, early site permit, or combined license
stage,'' the environmental data of table S-3 apply for enrichments up
to 5.0 weight percent U-235 and apply to the environmental report for
the CP stage, ESP stage, or COL stage of an LWR. Also, as required in
10 CFR 51.50, an applicant for a CP, ESP, or COL must submit an
environmental report that contains information specified in 10 CFR
51.45, ``Environmental report,'' 10 CFR 51.51, and 10 CFR 51.52.
Under 10 CFR 51.52, an environmental report prepared for the CP
stage, ESP stage, or COL stage of an LWR must contain a statement
concerning the environmental impacts of transportation of fuel and
radioactive waste to and from the LWR. If the conditions in 10 CFR
51.52(a) as extended to 5.0 weight percent U-235 are met, then table S-
4 gives these environmental impacts and can be so stated in the
applicant's environmental report. If an LWR applicant for a CP, ESP, or
COL cannot meet the conditions for using table S-4 (for example, the
plant has enrichments greater than 5.0 weight percent U-235), then the
environmental report must contain a full description and detailed
analysis of the environmental effects of transportation of fuel and
radioactive waste to and from the reactor, including values for the
environmental impact under normal conditions of transport and for the
environmental risk from accidents in transport.
For licensing actions other than CP, ESP, and COL applications,
which may not require submission of an environmental report (e.g.,
license amendment requests), the uranium fuel cycle and the
transportation of the fuel and waste could be considered necessary to
support the operation of the plant. In such situations, the NRC would
assess the environmental effects of the licensing action with respect
to the uranium fuel cycle or transportation of fuel and waste or both.
Sections XXXVI.B., ``Uranium Fuel Cycle Environmental Data--Table
S-3 in 10 CFR 51.51,'' and XXXVI.C., ``Environmental Effects of
Transportation of Fuel and Waste--Table S-4 in 10 CFR 51.52,'' of this
document include discussions of proposed amendments to support the use
of fuel with enrichments greater than 5.0 weight percent U-235 for the
uranium fuel cycle in 10 CFR 51.51 and transportation and waste
requirements in 10 CFR 51.52. In a separate rulemaking, the NRC is
considering changing the types of environmental impacts considered in
tables S-3 and S-4.
[[Page 44607]]
(iii) Fissile Packaging Requirements
In an effort to fully utilize ATF capabilities, licensees and
applicants are expected to seek approval for the transport of fissile
material with enrichments that range between 5.0 and 10.0 weight
percent U-235. The fuel cycle for commercial LWR fuel includes
enrichment of uranium hexafluoride (UF6) and shipment of the
enriched UF6 to a fuel fabricator for deconversion to
uranium dioxide (UO2). The enriched UF6 is
transported in NRC-approved packages that incorporate 30-inch
cylinders.
The regulation in paragraph (b) of 10 CFR 71.55, ``General
requirements for fissile material packages,'' requires that a
transportation package be designed and constructed, and its contents
limited, so that it would be subcritical if water were to leak into the
containment system. This criticality analysis with moderation ensures
criticality safety in transport in the unanticipated event that water
leaks into the containment vessel and provides moderating materials for
the fissile contents.
In the 2004 amendments to 10 CFR part 71 (69 FR 3698; January 26,
2004), the NRC implemented an exception to 10 CFR 71.55(b) in 10 CFR
71.55(g), which codified a longstanding NRC and worldwide practice for
evaluating the leakage of water into UF6 packages. This
exception for UF6 transportation packages can be used if all
of the following conditions are met:
The UF6 cylinder remains leak tight following
the tests specified for hypothetical accident conditions.
The valve body of the cylinder does not impact any other
part of the package, other than where it is attached to the cylinder.
There is adequate quality control in the manufacture,
maintenance, and repair of packagings.
Each package is tested to demonstrate closure before each
shipment.
The uranium is enriched to not more than 5.0 weight
percent U-235.
This exception is a performance-based assessment of the structural
and containment integrity of the UF6 cylinder, which is
independent of the enrichment level of the contents.
Similarly, 10 CFR 71.55(c) also provides for an exception to the
requirements in 10 CFR 71.55(b) if the applicant specifies that the
package incorporates special design features that ensure that no single
packaging error would permit leakage and that appropriate measures are
taken before each shipment to ensure that the containment system does
not leak. This exception does not limit the enrichment of the package
contents.
In the 2004 amendments to 10 CFR part 71, the NRC explained the
basis for the specific exception in 10 CFR 71.55(g) and its enrichment
limit of 5.0 weight percent U-235 as follows: (1) it would maintain
consistency with worldwide practice, (2) operation experience and
history demonstrate safe shipment of fuel enriched to less than or
equal to 5.0 weight percent U-235, and (3) it is necessary to transport
an essential commodity (UF6 feed material).
Currently, the regulations in 10 CFR part 71 are sufficiently
performance-based and, except for 10 CFR 71.55(g), do not directly
reference or limit the enrichment level of the radioactive contents.
The regulatory issue with 10 CFR 71.55(g) is that it specifies an
enrichment limit (5.0 weight percent U-235) for UF6 that
does not bound the range of enrichment that applicants may choose to
ship in their UF6 transportation packages in the future.
Section XXXVI.D., ``Fissile Material Packaging Requirements in 10
CFR 71.55,'' of this document includes a discussion of proposed
amendments to the fissile packaging requirements in 10 CFR 71.55 to
support the use of fuel with enrichments greater than 5.0 weight
percent U-235.
(iv) Appendix A to 10 CFR Part 50 (General Design Criterion 19) and 10
CFR 50.67(b)(2)(iii)
The general design criteria (GDC) in appendix A to 10 CFR part 50,
``General Design Criteria for Nuclear Power Plants,'' Criterion 19,
``Control room'' (GDC 19), provide minimum design, fabrication,
construction, testing, and performance requirements for structures,
systems, and components (SSCs) that provide reasonable assurance that
the facility can be operated without undue risk to public health and
safety. Additionally, 10 CFR 50.67, ``Accident source term,'' allows
applicable licensees to voluntarily revise the accident source term
used in design basis radiological consequences analyses if certain
requirements in 10 CFR 50.67(b)(2) are met.
Both GDC 19 and 10 CFR 50.67(b)(2)(iii) provide a specific dose-
based criterion of 5 rem (0.05 Sv) TEDE for demonstrating the
acceptability of the control room design. They represent a distinct
layer of defense-in-depth that assumes a major accident that results in
substantial meltdown of the reactor core with subsequent release of
appreciable quantities of fission products. In application, GDC 19 and
10 CFR 50.67(b)(2)(iii) are performance based and require that a
licensee or applicant provide a control room habitability design using
traditional deterministic radiological consequence analyses methods to
judge the acceptability of the design.
An acceptable level of control room habitability for design basis
events (DBEs) is necessary to provide reasonable assurance that the
control room would continue to be staffed and operated effectively to
mitigate the effects of the postulated accident and protect public
health and safety. GDC 19 and 10 CFR 50.67(b)(2)(iii) are design
criteria and should not be construed as operational limits. While the
design criteria are computed in terms of dose, they are figures of
merit used to characterize the minimum requirements for design,
fabrication, construction, testing, and performance for SSCs. The
design criteria do not represent actual occupational exposures received
during normal and emergency conditions, which are primarily controlled
by 10 CFR part 20, ``Standards for Protection Against Radiation.''
The preamble for the 1971 final rule (36 FR 3255; February 20,
1971) that first published the GDC addressed the criteria only in the
aggregate; the individual criteria were not discussed. However, there
is a record of a change made to the proposed GDC 11 (32 FR 10213; July
11, 1967), which became the final GDC 19. The proposed GDC 11 referred
to the occupational exposure limits of 10 CFR part 20 rather than
specifying a numeric dose criterion. Industry comments on that proposal
generally recommended deletion of the reference to 10 CFR part 20 (see
SECY-R-143, ``Amendment to 10 CFR 50--General Design Criteria for
Nuclear Power Plants,'' dated January 28, 1971). The Commission
resolved these comments by deleting the reference to 10 CFR part 20
occupational exposure limits and providing the current ``5 rem whole-
body, or its equivalent to any part of the body, for the duration of
the accident'' in its place.
Section 50.67 of 10 CFR was established shortly after a revision to
10 CFR part 20 was issued in 1991, providing the voluntary regulatory
mechanism for licensees to replace the original design criteria of
whole body and thyroid with the new TEDE criteria. The preamble for the
10 CFR 50.67 final rule included the Commission's rationale for
establishing 5 rem (0.05 Sv) TEDE as the GDC 19 numeric design
criterion for licensees using an alternative source term. That
rationale was composed of the following:
[[Page 44608]]
The criteria in GDC 19 were based on a primary
occupational exposure limit.
The use of 5 rem (0.05 Sv) TEDE as the control room
criterion did not imply that this value would be an acceptable exposure
during emergency conditions, or that other radiation protection
standards of 10 CFR part 20, including individual organ dose limits,
might not apply. This criterion was provided only to assess the
acceptability of design provisions for protecting control room
operators under postulated design-basis accident (DBA) conditions. The
DBA conditions assumed in these analyses, although credible, generally
did not represent actual accident sequences but were specified as
conservative surrogates to create bounding conditions for assessing the
acceptability of engineered safety features.
The regulations at 10 CFR 20.1206, ``Planned special
exposures,'' permitted a planned special dose of five times the annual
dose limits. Also, the pertinent U.S. Environmental Protection Agency
(EPA) guidance at that time, ``Manual of Protective Action Guides and
Protective Actions for Nuclear Incidents,'' EPA-400/R-92-001, issued
May 1992, set a limit of five times the annual dose limits for workers
performing emergency services such as lifesaving or protection of large
populations. The Commission did not suggest that control room dose
during an accident can be treated as a planned special exposure or that
the EPA emergency worker dose limits are an alternative to GDC 19.
However, the Commission stated that these provisions offer a useful
perspective that supports the conclusion that the organ doses implied
by the 5 rem (0.05 Sv) criterion can be considered to be acceptable due
to the relatively low probability of the events that could result in
doses of this magnitude.
Development of the current control room design criterion did not
foresee how licensees are currently operating their facilities and
managing their fuel or considering fuel enrichments up to but less than
20.0 weight percent U-235. The history of fuel utilization for the
current large LWR fleet has seen a gradual progression toward higher
fuel burnups and increased enrichments. The original control room
design criteria were developed during the late 1960s when burnup rates
and enrichments were relatively low. During that time, there was enough
margin in the facilities' design bases to accommodate the control room
design criteria, even for power uprates of up to 120 percent of the
originally licensed steady-state thermal power level. Today, vendors,
licensees, and other members of the nuclear power industry have
indicated to the NRC that they are looking for further power uprates
using fuel enrichments up to 10.0 weight percent U-235 with fuel burnup
limits higher than the 62 GWd/MTU rod-average burnup.
Depending on how the reactor core is designed with increased U-235
enrichment fuel elements and operation at higher burnup levels to reach
longer cycle time, the results of a licensee's DBA radiological
consequence analysis results would increase. The impact of this
increase would decrease the retained margin maintained by the licensee
to provide operational flexibility. An unjustifiably low design
criteria can unnecessarily burden licensees for seeking increased
enrichments with extensive analyses to preserve margin for operational
flexibility purposes. These additional analyses may not result in
safety benefits and can increase actual operational exposure to workers
due to increased maintenance activities.
Under the current definition of ``safety-related structures,
systems and components'' in 10 CFR 50.2, ``Definitions,'' any SSCs
credited with providing mitigation functions during a DBE (or accident)
must be designated as ``safety-related.'' The traditional radiological
consequence analyses performed to demonstrate compliance with the
control room design criterion assess the performance of safety-related
SSCs because they are relied upon to remain functional during and
following DBAs to ensure the capability to prevent or mitigate the
consequences of accidents that could result in potential exposure. The
analyses are not intended to be actual event sequences but, rather, are
intended to be surrogates to enable deterministic evaluation of the
response of the plant-engineered safety features. These accident
analyses are intentionally conservative in order to address
uncertainties in accident progression, fission product transport, and
atmospheric dispersion. With few exceptions, these analyses do not
credit non-safety-related SSCs or operator actions that would otherwise
lower the radiological consequence results. This analysis approach can
ensure conservative results, but the results can also have large
uncertainties. Due to the modeling approach and inherent uncertainty,
overly conservative results can lead licensees to perform extensive re-
analyses to preserve margin for operational flexibility purposes that
do not necessarily enhance safety.
The radiological consequence analyses also confirm several aspects
of the facility's design- and licensing-basis when safety-related SSCs
are credited as input parameters. These input parameters are often
specific values and limits found in the facility's updated FSAR and
technical specifications, pursuant to 10 CFR 50.36, ``Technical
specifications.'' Deviations from the Technical Specification
identified during maintenance or testing (e.g., higher leakage rates or
lower filter efficiencies) indicate a non-conformance issue with the
facility's licensing basis. In such cases, additional maintenance must
be performed to correct the discrepancy to bring the facility back into
compliance with established licensing requirements. The degree of
maintenance necessary to ensure the facility is in compliance can
significantly influence the amount of radiation exposure incurred by
workers. The numerical value of the control room design criteria,
through the radiological consequence analyses that utilize specific
values and limits found in the facility's technical specifications,
factors into the licensee's decisions when performing maintenance
activities and thus directly impacts the amount of workers' exposure to
ionizing radiation. A very low design criteria value can result in an
excessive amount of maintenance, leading to potentially avoidable
occupational exposure and unnecessary operational disturbances.
Conversely, a very high value may allow for unacceptable degradation,
potentially compromising overall safety and performance over time.
Adequate protection of public health and safety and occupational
radiological safety can still be achieved at a higher and safe control
room design criteria performance level while balancing both dose-
savings to workers and providing some regulatory relief to maintain
operational flexibilities. As discussed in section XXXVI.E., ``Control
Room Requirements in 10 CFR 50.67 and GDC 19,'' of this document, the
NRC proposes to address these issues so licensees would not need to
perform potentially extensive re-analyses or excessive maintenance
activities, or possibly request exemptions, to demonstrate compliance
without a commensurate increase in safety.
The NRC's Radiation Protection and Emergency Response Framework
The NRC's comprehensive radiation protection and emergency response
framework, which covers both normal operations and accident conditions,
is another important aspect of the control room design criteria
rulemaking efforts. This framework helps to protect occupational
workers from ionizing radiation as well as prepare the licensee
[[Page 44609]]
to respond to abnormal and emergency conditions to protect the public
health and safety.
At the time that GDC 19 was established in 1971, 10 CFR part 20
limited occupational radiation exposure to 3 rem (0.03 Sv) whole body
dose per calendar quarter, provided the total lifetime dose was
verified not to exceed 5 rem (0.05 Sv) times the individual's age in
years minus 18. Thus, a worker could receive a radiation exposure of up
to 12 rem (0.12 Sv) in a given year.
The current annual limit on occupational radiation dose exposure in
10 CFR 20.1201, ``Occupational dose limits for adults,'' is 5 rem (0.05
Sv) TEDE. Under 10 CFR 20.1201, an adult worker could receive
occupational radiation exposure of up to 10 rem (0.10 Sv) TEDE over a
12-month period straddling two calendar years. The current 10 CFR
20.1206 also permits an adult worker to receive doses in addition to,
and accounted for separately from, the doses received under the limits
specified in 10 CFR 20.1201 of five times the annual dose limits during
the individual's lifetime, not to accumulate faster than 5 rem (0.05
Sv) TEDE in any one year. As such, an adult worker could receive
radiation exposure of up to 10 rem (0.10 Sv) TEDE within a single
calendar year period. In setting these standards in the 1991 amendment
of 10 CFR part 20 (56 FR 23360; May 21, 1991), the Commission concluded
that an infrequent exposure of workers up to twice the occupational
dose limit was adequately protective of radiation workers.
The NRC's emergency planning regulations in appendix E to 10 CFR
part 50, ``Emergency Planning and Preparedness for Production and
Utilization Facilities,'' and planning standards for nuclear power
reactors in 10 CFR 50.47, ``Emergency plans,'' require each nuclear
power reactor licensee to have an emergency plan that gives the NRC
reasonable assurance that adequate protective measures can and will be
taken in the event of a radiological emergency. The regulation at 10
CFR 50.47(b)(11) requires licensees to establish the means for
controlling radiological exposures in an emergency and states that the
means for controlling radiological exposures must include exposure
guidelines consistent with EPA Emergency Worker and Lifesaving Activity
Protection Action Guides (PAG). The EPA exposure guidelines found in
the current version of its ``PAG Manual: Protective Action Guides and
Planning Guidance for Radiological Incidents,'' recommend that doses
received under emergency conditions should be maintained as low as
reasonably achievable and, to the extent practicable, limited to 5 rem
(0.05 Sv). The guideline for actions to protect valuable property is 10
rem (0.10 Sv) where a lower dose is not practicable, the guideline for
actions to save a life or to protect large populations is 25 rem (0.25
Sv) where a lower dose is not practicable, and exposures greater than
25 rem (0.25 Sv) may be appropriate for lifesaving or protecting large
populations if the workers are volunteers who are fully aware of the
risks involved.
The events that could result in control room radiation exposures
comparable to the 10 CFR part 20 normal occupational exposure limit of
5 rem (0.05 Sv) TEDE would result in the activation of the facility's
emergency response plan and the emergency response organization. These
emergency actions include establishing higher exposure limits for
control room operators if necessary to provide public health and
safety, as permitted by paragraph (x) of 10 CFR 50.54, ``Conditions of
licenses,'' and paragraph (b) of 10 CFR 20.1001. The emergency
coordinator can also authorize issuing potassium-iodide tablets for
thyroid protection or use of emergency respiratory protection
equipment.
The Commission's framework for emergency planning and response
encompasses a combination of regulatory requirements and industry
commitments tailored to address a spectrum of potential events, from
design-basis scenarios to extremely low-probability severe accident
events. This comprehensive approach ensures preparedness to effectively
protect public health and safety by enabling robust planning and
response capabilities.
Scientific Recommendations for Radiation Protection for Worker and
Regulations Under Accident and Emergency Conditions
The NRC reviewed several source materials to understand the current
recommendations from national and international organizations
responsible for making recommendations for radiation protection
standards. The purpose of this review was to determine whether
reexamining the scientific and technical basis for the numerical value
of the control room design criteria would be warranted. Section
XXXVI.E., ``Control Room Requirements in 10 CFR 50.67 and GDC 19,'' of
this document describes how the NRC used this review to inform the
development of this proposed rule.
ICRP Publication 109, ``Application of the Commission's
Recommendations for the Protection of People in Emergency Exposure
Situations,'' issued in 2009, specifies a reference range of 2 to 10
rem (0.02 to 0.10 Sv) acute, or per year, for emergency exposure
situations. The reference level represents the level of residual dose
or risk above which it is generally judged to be inappropriate to plan
to allow exposures to occur. The ICRP considers that a dose rising
towards 10 rem (0.10 Sv) will almost always justify protective measures
and that protection against all exposures, above or below the reference
level, should be optimized.
The IAEA 2024 guidance, ``Portable Digital Assistant for First
Responders to a Radiological Emergency: Emergency worker turn-back dose
guidance,'' specifies a range of 5 to 100 rem (0.05 to 1 Sv), depending
on the severity of the actions needed.
The 2018 NCRP Report No. 180, ``Management of Exposure to Ionizing
Radiation: Radiation Protection Guidance for the United States,''
specifies the following: (1) during lifesaving activities or actions to
prevent a catastrophic situation, which includes other urgent rescue
activities, 50 radiation-absorbed dose (rad) (0.5 gray (Gy)) cumulative
whole-body absorbed dose (50 rad) should be implemented at the command
level, and (2) for other emergency activities, including extended
activities following initial lifesaving, rescue, and damage control
response, an effective dose to emergency workers should not exceed 10
rem (0.10 Sv).
Modern Health Physics and Radiation Epidemiology Knowledge
The NRC's comprehensive radiation protection and emergency response
framework for protecting individuals during normal and emergency
conditions is informed by scientific recommendations by national and
international organizations. These recommendations are based on
fundamental modern health physics and radiation epidemiology knowledge.
The NRC's consideration of these organizations' recommendations has
contributed to developing this proposal to amend the control room
design criteria value from 5 rem (0.05 Sv) TEDE to 10 rem (0.10 Sv)
TEDE with the additional provisions to justify a higher numerical value
up to 25 rem (0.25 Sv) TEDE.
The range of proposed control room design criteria values is
significantly below the threshold for observable deterministic health
effects such as acute radiation syndrome and
[[Page 44610]]
hematopoietic syndrome, which occurs at a dose around 70 to 100 rad
(0.7 to 1 Gy). This range is also far below the mean lethal dose of
ionizing radiation without medical treatment, which is estimated to be
approximately 300 to 500 rad. This demonstrates that the proposed
design criteria range is well within safe limits relative to acute
radiation effects that could impair workers from performing their
safety function in response to an event. Additionally, the criteria
range results in a small radiation risk for cancer mortality, which
would be further mitigated by radiation protection and emergency
planning actions during an actual event. This ensures a high level of
protection is still provided, thereby minimizing long-term health
impacts.
In RG 8.29, Revision 1, ``Instruction Concerning Risks from
Occupational Radiation Exposure,'' the NRC adopted a risk value, for an
occupational dose of 1 rem (0.01 Sv) TEDE, of 4 in 10,000 of developing
a fatal cancer, or approximately 1 chance in 2,500 of fatal cancer per
rem of TEDE received. The uncertainty associated with this risk
estimate does not rule out the possibility of higher risk, or the
possibility that the risk may even be zero at low occupational doses
and dose rates. The radiation risk incurred by a worker depends on the
amount of dose received. Under current health physics models, a worker
who receives 5 rem (0.05 Sv) in a year incurs 10 times as much risk as
another worker who receives only 0.5 rem (0.005 Sv).
Thus, in a group of 10,000 people, each exposed to 1 rem (0.01 Sv)
of ionizing radiation, and using the risk factor of 4 effects per
10,000 rem (100 Sv) of dose, 4 of the 10,000 people might die from
delayed cancer because of that 1 rem (0.01 Sv) dose in addition to the
2,000 normal cancer fatalities expected to occur in that group from all
other causes. From an individual perspective, a 1 rem (0.01 Sv) dose
may increase an individual worker's chances of dying from cancer from
20 percent to 20.04 percent. If one's lifetime occupational dose is 10
rem (0.1 Sv), the estimate would increase to 20.4 percent. A lifetime
dose of 100 rem (1.0 Sv) may increase chances of dying from cancer from
20 to 24 percent. This small increase in cancer risk could be inferred
over the lifetime, however it is unlikely that an increased incidence
of cancer due to irradiation would be discernible. This is because the
normal variability in baseline rates of cancer incidence is much larger
than the inferred radiation-associated cancer rates. As a point of
reference, according to NUREG-0713, Volume 43, ``Occupational Radiation
Exposure at Commercial Nuclear Power Reactors and other Facilities,''
published in 2021, the average measurable dose for radiation workers
reported to the NRC was 0.16 rem (0.0016 Sv) for 2021.
(v) Fuel Dispersal
(a) 10 CFR 50.46 and Fuel Dispersal
In 1974, the Atomic Energy Commission (AEC) established ECCS
acceptance criteria during postulated loss-of-coolant accidents (LOCAs)
in 10 CFR 50.46, ``Acceptance criteria for emergency core cooling
systems for light-water nuclear power reactors'' (39 FR 1001; January
4, 1974). The core cooling acceptance criteria in 10 CFR 50.46 were
based on the available research, operating experience, and fuel
operating conditions applicable to that era. Potential impacts of FFRD
phenomena were not understood at that time and were not referenced in
10 CFR 50.46 or the accompanying analysis methods described in appendix
K to 10 CFR part 50, ``ECCS Evaluation Models.'' The lack of reference
to such phenomena in these regulations may be attributed to the fact
that, in the early 1970s, fuel discharge burnups were well below the
threshold local burnup (i.e., 55 GWd/MTU) at which FFRD phenomena are
now recognized to be a risk. Now that increased enrichment and higher
fuel discharge burnups are being contemplated, the NRC is examining the
original intent of the rulemaking as well as the current state of
knowledge and operational experience to date to develop a performance-
based regulatory framework that addresses fuel dispersal in a manner
that maintains reasonable assurance of adequate protection of public
health and safety.
The acceptance criteria in the original 10 CFR 50.46(b) included
limits on peak cladding temperature (PCT) and maximum local oxidation
(MLO) in 10 CFR 50.46(b)(1) and (b)(2), respectively, as well as the
requirement in 10 CFR 50.46(b)(4) that the core should remain amenable
to cooling. As stated in the AEC's 1973 opinion announcing its decision
on the 10 CFR 50.46 final rule, the limits on PCT and MLO were intended
``to ensure the zircaloy cladding would remain sufficiently intact to
retain the UO2 fuel pellets in their separate fuel rods and
therefore remain in an easily coolable array.'' In other words, these
criteria were intended to prevent the fuel from leaving the confines of
the cladding.
Regarding the coolability criterion, the AEC envisioned two
scenarios that were deemed unacceptable: (1) the ballooning of the
cladding to the extent that the coolant passages are blocked and (2)
allowing the fuel pellets to fall together into a heap that would be
difficult to cool. The 1973 AEC opinion stated that the coolability
criterion should be superfluous because of the PCT and MLO criteria,
but that the AEC maintained it as a basic objective in view of its
fundamental and historical importance. In other words, the objective of
the core coolability criterion should not be viewed as different from
that of the PCT and MLO criteria. While brittle failure is precluded
under the PCT and MLO criteria of 10 CFR 50.46, the AEC understood and
accepted that ductile failure (i.e., ballooning and burst) would occur,
but it would be limited such that it would not block the coolant flow,
as stated in the first unacceptable scenario considered in the
formulation of the coolability criterion. Fuel dispersal is typically
associated with the ductile failure of cladding encapsulating finely
fragmented fuel at local burnups in excess of 55 GWd/MTU. Based on the
historical record, the AEC expected the fuel would remain confined by
the cladding following ductile failure, as fine fuel fragmentation and
dispersal was not a known phenomenon at the time and further would not
have been operative at the fuel discharge burnups attained in the early
1970s.
After the AEC's approval of 10 CFR 50.46 in 1973, fuel discharge
burnups at operating reactors continued to increase. Fuel fragmentation
and relocation were first discovered in the early 1980s, when
experiments conducted at several test facilities showed that irradiated
fuel could fragment into small pieces during a LOCA and may relocate
axially, settling into the ballooned regions. In 1984, the NRC decided
to consider the implications of the phenomena in the generic issue (GI)
program, specifically as GI-92, ``Fuel Crumbling During LOCA,'' as
described in NUREG-0933, ``Resolution of Generic Safety Issues,''
issued September 2021. The NRC found that the known conservatisms in
appendix K to 10 CFR part 50 would more than offset the heat generation
in the balloon region because of the fragmentation and relocation of
fuel in calculations performed under appendix K to 10 CFR part 50.
In the early 1990s, the conclusion of GI-92 was that fuel
fragmentation and relocation should be placed no higher than the low
priority category of the GI program. This meant that there was
insufficient justification for starting a major re-review of existing
ECCS performance analyses conducted in adherence to appendix K to 10
CFR part
[[Page 44611]]
50. However, there were ongoing efforts to develop and license more
realistic (and thus less conservative) ECCS performance models.
Therefore, the NRC expected that the ECCS evaluation methodologies
would appropriately address fuel fragmentation and relocation in their
calculations. As a result, the NRC decided that a separate GI was not
necessary, and the issue was later dropped from the GI program in 1995.
Until 2006, there was no expectation from any prior research that
fragmentation and relocation into the balloon region could result in
the loss of fuel particles through the rupture opening. In 2006,
several research tests challenged this assumption. Integral LOCA tests
conducted at the U.S. Department of Energy's (DOE's) Argonne National
Laboratory on rods up to local burnups of 64 GWd/MTU observed a small
amount of fuel loss (about the quantity of one fuel pellet). Since the
amount of material was small, this discovery was not thought to be of
safety significance. In April 2006, a LOCA test was run in the Halden
Reactor Project in Norway on a fuel rod segment with a very high local
burnup of 91.5 GWd/MTU. Results from this test showed gross loss of
fuel material from above the rupture opening. In this very-high-burnup
fuel specimen, more than 40 percent of the fuel material was in a
nearly powdered form, as described in NUREG-2121, ``Fuel Fragmentation,
Relocation, and Dispersal During the Loss-of-Coolant Accident,'' issued
March 2012.
In 2008, the NRC's Office of Nuclear Regulatory Research (RES)
issued Research Information Letter (RIL)-0801, ``Technical Basis for
Revision of Embrittlement Criteria in 10 CFR 50.46,'' dated May 30,
2008, which discussed research findings in the area of high-burnup fuel
performance during postulated LOCAs. The RIL-0801 noted that additional
research on fuel dispersal was being conducted but concluded that ``the
current NRC burnup limit of 62 GWd/MTU (average for the peak rod) is
probably low enough to prevent significant fuel loss during a LOCA.''
The RIL-0801 recommended rulemaking be pursued to revise the criteria
in 10 CFR 50.46(b) to account for high burnup phenomena that may cause
the current criteria to be non-conservative.
In 2012, the NRC published NUREG-2121 to capture the state of
knowledge and history of FFRD as of that time. NUREG-2121 concluded
that additional experimental research was needed to quantify the extent
and downstream effects of fuel dispersal. Additionally in SECY-15-0148,
``Evaluation of Fuel Fragmentation, Relocation and Dispersal Under
Loss-of-Coolant Accident (LOCA) Conditions Relative to the Draft Final
Rule on Emergency Core Cooling System Performance During a LOCA
(50.46c),'' dated November 30, 2015, the staff concluded that there was
``no imminent safety concern'' for operating reactors with respect to
FFRD and that the 10 CFR 50.46c rulemaking should not be delayed to
address FFRD. The SECY paper also stated that additional research was
ongoing, and that future regulatory action could be initiated, if
needed, to address FFRD after more research was conducted.
In December 2021, RES published RIL 2021-13, ``Interpretation of
Research on Fuel Fragmentation, Relocation, and Dispersal at High
Burnup,'' to inform the NRC's Office of Nuclear Reactor Regulation
about RES's interpretation of the FFRD research to date. In RIL 2021-
13, the staff defines conservative empirical boundaries for FFRD-
related phenomena, such as the amount of fuel that is expected to be
dispersed during a LOCA. Additionally, RIL 2021-13 identifies data gaps
associated with FFRD-related phenomena. While the models in RIL 2021-13
can be used to estimate the potential mass of fuel that could be
dispersed to the coolant, that RIL does not attempt to address the
consequences of fuel dispersal into the coolant. To fully characterize
such consequences, the NRC needs to better understand the behavior of
dispersed fuel particles in the coolant and their impact on core
coolability and safety under LOCA conditions that may involve
significant core geometry changes due to rod ballooning, significant
and varying single- and two-phase core flows, and dynamic LOCA loads.
The NRC is involved in several collaborative domestic and international
research programs, such as the Studsvik Cladding Integrity Project,
which are, in part, addressing some of the data gaps presented in RIL
2021-13. Additionally, the NRC sponsored a phenomena identification and
ranking table (PIRT) exercise concerning the consequences of fuel
dispersal. The findings of the expert PIRT panel are documented in
NUREG/CR-7307, ``Phenomena Identification and Ranking Tables on High
Burnup Fuel Fragmentation, Relocation, Dispersal, and Its Consequences
for Design-Basis Accidents in Pressurized- and Boiling-Water
Reactors,'' and have informed DG-1434, ``Addressing the Consequences of
Fuel Dispersal in Light-Water Reactor Loss-of-Coolant Accidents,''
which is being issued for comment with this proposed rule. The PIRT
will also help inform future research efforts and NRC review of
applications that may evaluate FFRD.
In 2022, in SRM-SECY-21-0109, as part of its approval of the
staff's plan to begin this increased enrichment rulemaking, the
Commission directed that FFRD should be appropriately addressed and
analyzed in the rulemaking's regulatory basis. The NRC considered
several alternatives in the regulatory basis, along with other options
received in the public comments on the regulatory basis, in developing
the following path forward to addressing FFRD.
(b) 10 CFR 50.46a Rulemaking
In 2010, the staff sent to the Commission for approval via SECY-10-
0161, ``Final Rule: Risk-Informed Changes to Loss-of-Coolant Accident
Technical Requirements (10 CFR 50.46a) (RIN 3150-AH29),'' a draft final
rule that would have created alternative ECCS requirements in 10 CFR
50.46a. The rule would have divided the current spectrum of LOCA break
sizes into two regions. The division between the two regions would have
been delineated by the transition break size (TBS). The first region
included small breaks, up to and including the TBS. The second region
included breaks larger than the TBS, up to and including the double-
ended guillotine break (DEGB) of the largest reactor coolant system
(RCS) pipe. The likelihood of these larger breaks is much lower than
the smaller breaks in the first region, which was used to support a
different regulatory treatment for the larger breaks.
In this rulemaking, the NRC proposes to build on the 10 CFR 50.46a
rulemaking as a means of analytically resolving FFRD issues. The
analytical margins gained from the treatment of large-break LOCAs as
beyond-design-basis are expected to eliminate or greatly reduce the
calculated quantity of fuel dispersal, as described in section
XXXVI.F.(ii), ``Original Determination of the Transition Break Size,''
of this document. While the NRC has attempted to update this proposed
rule with information available since SECY-10-0161 was issued, the NRC
expects stakeholders will provide significant additional information in
their comments on this proposed rule that could support the NRC further
risk-informing these aspects of the rule.
(c) 10 CFR 50.46c Rulemaking and Cladding Embrittlement Research
Findings
In 2016, the staff sent to the Commission for approval via SECY-16-
[[Page 44612]]
0033 a draft final rule to create 10 CFR 50.46c. The draft final rule
would have established performance-based regulatory requirements for
determining the acceptability of an ECCS for a nuclear power reactor.
Similar to existing regulations in 10 CFR 50.46, the draft ECCS
performance requirements in 10 CFR 50.46c were largely based upon
acceptance criteria for fuel rod cladding performance. The final rule
would have expanded the applicability of the 10 CFR 50.46 acceptance
criteria from only uranium oxide pellets within cylindrical zircaloy or
ZIRLOTM 1 cladding to any LWR fuel, regardless of fuel
design or cladding material. The draft final rule also incorporated
improved performance-based requirements to address research findings on
fuel cladding integrity and degradation mechanisms that were described
in RIL-0801.
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\1\ ZIRLO is a registered trademark of Westinghouse Electric
Company LLC.
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The Commission returned the 10 CFR 50.46c draft final rule to the
staff in SRM-SECY-16-0033, ``Staff Requirements--SECY-16-0033--Draft
Final Rule--Performance-Based Emergency Core Cooling System
Requirements and Related Fuel Cladding Acceptance Criteria,'' on April
11, 2024. In this SRM, the Commission stated that the staff should
reconsider the topics presented in the 10 CFR 50.46c draft final rule
and provided the following directions:
1. The staff should apply an appropriate risk-informed regulatory
approach to address the research findings on cladding embrittlement
effects under LOCA conditions described in SECY-16-0033.
2. The staff should evaluate Item 1 with other associated technical
issues being addressed, such as FFRD and risk-informed treatment of
LOCAs, including the 10 CFR 50.46a draft final rule.
3. The staff should evaluate whether specific ECCS criteria such as
cladding temperature should be codified or instead addressed in
regulatory guidance.
This rulemaking proposes to leverage the previously proposed
performance-based approach to ECCS requirements, including the expanded
applicability to advanced fuels, and incorporate the embrittlement
research findings from the 10 CFR 50.46c draft final rule into the new
voluntary 10 CFR 50.46a proposed rule and associated guidance. The
following discussion describes the embrittlement research findings that
the 10 CFR 50.46c draft final rule planned to address and that this
rulemaking proposes to address.
All licensees who adopt the new voluntary 10 CFR 50.46a proposed
rule would address the embrittlement research findings through their
updated analyses and associated acceptance criteria. The NRC would
continue to conduct annual safety assessments for licensees that do not
choose to adopt the new requirements to confirm reasonable assurance of
adequate protection. After the completion of this rulemaking, the NRC
would continue to consider whether any new regulatory requirements are
needed for these licensees that do not adopt proposed 10 CFR 50.46a
because the research findings show that the current criteria may not
always ensure that the fuel cladding remains ductile after the reactor
is reflooded, referred to as post-quench ductility (PQD). The NRC
expects that its assessment of further actions would consider (1) the
resource needs for and results of the ECCS annual safety assessments,
(2) the industry's plans for adoption of 10 CFR 50.46a, and (3) whether
there is a safety basis for additional requirements beyond the
promulgation of this voluntary alternative proposed rule. If the
majority of the industry adopts 10 CFR 50.46a, then little to no future
regulatory action may be needed to resolve these matters.
1. Overview of Cladding Embrittlement Research Findings
Since 1997, the NRC has undertaken a fuel cladding research program
to investigate the behavior of high-exposure fuel cladding under
accident conditions. This research program included an extensive LOCA
research and testing program at Argonne National Laboratory, as well as
jointly funded programs at the Kurchatov Institute (supported by the
French Institute for Radiological Protection and Nuclear Safety and the
NRC) and the Halden Reactor Project (a jointly funded program under the
auspices of the Organization for Economic Cooperative Development--
Nuclear Energy Agency, sponsored by national organizations in 18
countries). The effects of both alloy composition and fuel burnup on
cladding embrittlement (e.g., loss of ductility) under accident
conditions were studied in these research programs. The research
programs identified new cladding embrittlement mechanisms that were not
previously known and expanded the NRC's knowledge of previously
identified mechanisms.
2. Major Research Findings Cladding Embrittlement
These research findings have been summarized in RIL-0801, and the
detailed experimental results from the program at Argonne National
Laboratory are contained in NUREG/CR-6967, ``Cladding Embrittlement
during Postulated Loss-of-Coolant Accidents,'' dated July 31, 2008.
Since the publication of NUREG/CR-6967 and RIL-0801, additional testing
was conducted related to the embrittlement phenomenon, which has been
documented in supplemental reports. Where the additional testing
relates to conclusions and recommendations in RIL-0801, RIL-0801 has
been supplemented to reference the additional reports and incorporate
findings (``Update to Research Information on Cladding Embrittlement
Criteria in 10 CFR 50.46,'' dated December 29, 2011).
i. Hydrogen-Enhanced Beta-Layer Embrittlement
In current 10 CFR 50.46, the preservation of cladding ductility,
via compliance with regulatory criteria on PCT (10 CFR 50.46(b)(1)) and
local cladding oxidation (10 CFR 50.46(b)(2)), provides a level of
assurance that fuel cladding will not experience gross failure and that
the fuel rods will remain within their coolable lattice arrays. The
1997-2016 LOCA research program, as summarized in NUREG/CR-7219,
``Cladding Behavior During Postulated Loss-of-Coolant Accidents,''
identified new cladding embrittlement mechanisms that demonstrated that
the current combination of PCT (2200 degrees Fahrenheit ([deg]F) (1204
degrees Celsius ([deg]C)) and local cladding oxidation (17 percent
equivalent cladding reacted (ECR)) criteria may not always ensure PQD.
As explained in section 1.5 of NUREG/CR-7219, oxygen diffusion into the
base metal under LOCA conditions promotes a reduction in the thickness
(referred to as beta-layer thinning) and ductility (referred to as
beta-layer embrittlement) of the metallurgical structure within the
cladding that provides its macroscopic mechanical behavior. The
presence of hydrogen within the cladding accelerates this embrittlement
process. Hydrogen is produced from the corrosion of zirconium in water,
some of which is absorbed by the cladding, which is frequently referred
to as ``hydrogen pickup.''
The NRC's cladding embrittlement program did not investigate
cladding degradation mechanisms or develop the technical basis for
performance-based
[[Page 44613]]
requirements beyond the existing 2200 [deg]F (1204 [deg]C) PCT
criterion. Examples of degradation mechanisms beyond cladding
embrittlement (via oxygen diffusion) include excessive exothermic
metal-water reaction, alloy-specific eutectics, and loss of fuel rod
geometry due to plastic deformation. As a result, the existing 2200
[deg]F (1204 [deg]C) limit remains an upper limit on PCT for zirconium-
based alloys. However, as reflected in this embrittlement criterion in
DG-1263, ``Establishing Analytical Limits for Zirconium-Based Alloy
Cladding,'' and based on the results of the fuel cladding research
program, a lower PCT may be required to preserve ductility. Although
the 2200 [deg]F (1204 [deg]C) limit in 10 CFR 50.46 remains unchanged
in this rulemaking, the acceptance criteria in proposed 10 CFR 50.46a
would establish a requirement to address cladding degradation
phenomena, which would include cladding embrittlement. The 2200 [deg]F
(1204 [deg]C) temperature is proposed as an acceptable limit in DG-1263
rather than codified in 10 CFR 50.46a.
ii. Oxygen Ingress From Cladding Inside Diameter
As explained in section 1.5.6 of NUREG/CR-7219, oxygen sources may
be present on the inner surface of irradiated cladding due to gas-phase
uranium trioxide transport prior to gap closure, fuel-cladding-bond
formation (uranium dioxide in solid solution with zirconium dioxide),
and the fuel bonded to this layer. Under LOCA conditions, this
available oxygen may diffuse into the base metal of the cladding, which
could cause the cladding to become more brittle.
iii. Breakaway Oxidation
As explained in section 1.5.5 of NUREG/CR-7219, zirconium dioxide
can exist in several crystallographic forms, or allotropes. During
normal operation, the zirconium dioxide layer that develops has a
monoclinic crystallographic structure, which is neither fully dense nor
fully protective. During LOCA conditions, the monoclinic oxide will
transform to a tetragonal structure, and the oxide that newly forms
under LOCA conditions is also tetragonal. The tetragonal oxide is
dense, adherent, and protective against hydrogen pickup. However, there
are conditions, both mechanical (e.g., local regions of tensile stress)
and chemical (e.g., impurities at the metal surface), that promote a
transformation of the zirconium dioxide from the tetragonal back to the
monoclinic phase. The tetragonal-to-monoclinic transformation is an
instability that initiates at local regions of the metal-oxide
interface and grows rapidly throughout the oxide layer. Because this
transformation results in an increase in oxidation rate, it is referred
to as breakaway oxidation. Along with this increase in oxidation rate
resulting from cracks in the monoclinic oxide, significant hydrogen
pickup also occurs. Hydrogen that enters in this manner during a LOCA
transient promotes rapid embrittlement of the cladding.
While all zirconium alloys will eventually experience breakaway
oxidation when exposed to long enough durations of high-temperature
steam oxidation, the fuel cladding research program demonstrated that
alloying composition and manufacturing process (e.g., surface
roughness) influence the timing of this phenomenon.
iv. Applicability of Ductility-Based Analytical Limits to Burst Region
During a postulated LOCA, a portion of the fuel rod population may
be predicted to experience fuel rod ballooning and cladding rupture as
a result of rapid depressurization of the RCS in combination with
elevated cladding temperature. The number of burst rods depends on
several variables including initial conditions (e.g., fuel rod design,
rod internal pressure, rod power) and accident conditions (e.g., LOCA
break size, cladding temperature). A burst section of the fuel rod may
experience degradation mechanisms beyond oxygen diffusion embrittlement
encountered in the remaining portions of the fuel rod, including
significant amounts of hydrogen uptake from steam entering the fuel rod
through the rupture.
To investigate the mechanical behavior of ruptured fuel rods, the
NRC conducted testing, designed to result in the ballooning and burst
of as-fabricated and hydrogen-charged cladding specimens and high-
burnup fuel rod segments exposed to high-temperature steam oxidation
followed by rapid cooling by liquid water, or quench. The research
results and conclusions are documented in the NUREG-2119, ``Mechanical
Behavior of Ballooned and Ruptured Cladding.'' This testing confirms
that continued exposure to a high-temperature steam environment weakens
the already flawed region of the fuel rod surrounding the cladding
rupture. Hence, limitations on PCT and integral time-at-temperature are
necessary to preserve an acceptable amount of mechanical strength and
fracture toughness to maintain a coolable fuel geometry. Integral time-
at-temperature is related to the time spent at the elevated
temperatures seen during the LOCA and typically expressed in terms of
the ECR. In addition, the research demonstrated that the degradation in
strength and fracture toughness with prolonged exposure to steam
oxidation was increased with pre-existing cladding hydrogen content.
These research findings have been summarized in RIL-0801, and the
detailed experimental results from the program at Argonne National
Laboratory are contained in NUREG/CR-6967, ``Cladding Embrittlement
during Postulated Loss-of-Coolant Accidents,'' dated July 31, 2008.
Since the publication of NUREG/CR-6967 and RIL-0801, additional testing
was conducted related to the embrittlement phenomenon, which has been
documented in supplemental reports. Where the additional testing
relates to conclusions and recommendations in RIL-0801, RIL-0801 has
been supplemented to reference the additional reports and incorporate
findings (``Update to Research Information on Cladding Embrittlement
Criteria in 10 CFR 50.46,'' dated December 29, 2011).
These research findings presented the NRC with two options for
revising the fuel performance requirements: (1) establish a separate
performance requirement within the burst region (i.e., analytical
limits that preserve sufficient fracture toughness to ensure burst
region survival), or (2) apply the hydrogen-based embrittlement
analytical limits to the entire fuel rod.
In the absence of a credible analysis of loads, cladding stresses,
and cladding strains for a core degraded by LOCA conditions, there are
no absolute metrics to determine how much ductility or strength would
be needed to provide assurance that fuel rod cladding would maintain
its geometry during and following post-LOCA quench. It is also not
clear what impact breakage of some fuel rods into two pieces, owing to
potential loads following a hypothetical LOCA, would have on core
coolability. Fragmentation of fuel rod cladding would be more
detrimental to core coolability than severance of rods into two pieces.
Even minimal ductility ensures that cladding will have high strength
and toughness and, therefore, high resistance to fracturing. Brittle
cladding, on the other hand, might fail at low strength and shatter.
Therefore, the intent to maintain ductility is beneficial even with
limited knowledge of LOCA loads. The research documented in NUREG-2119
showed that if wall thinning and double-sided oxidation are accounted
for, then hydrogen-based embrittlement limits, such as the limit
provided in Figure 2
[[Page 44614]]
of DG-1263, are sufficient to ensure reasonable behavior of the
ballooned and ruptured region.
Therefore, the NRC elected to propose a revision to the fuel
performance requirements by applying a single performance-based
criterion to the entire fuel rod. This decision recognizes that
portions of the cladding within the burst region may not maintain
ductility. This position is reflected in DG-1263 and supported by the
technical basis documented in NUREG-2119.
D. Regulatory Basis
The NRC published the regulatory basis to support a rulemaking for
the ``Increased Enrichment of Conventional and Accident Tolerant Fuel
Designs for Light-Water Reactors'' in the Federal Register on September
8, 2023 (88 FR 61986). In the regulatory basis, the NRC presented draft
recommendations that focused on those requirements needed for LWR high-
assay, low-enriched uranium fuel, specifically with approved
conventional or ATF designs. The NRC requested public comment on these
recommendations and asked specific questions associated with the
identified regulatory topics. The NRC concluded that there was
sufficient regulatory basis to proceed with rulemaking to address the
regulatory issues associated with the use of fuel enriched to greater
than 5.0 weight percent U-235. The NRC held a public meeting on October
25, 2023, to discuss the regulatory basis and issued a summary of the
meeting on November 21, 2023.
The public comment period for the regulatory basis closed on
January 22, 2024. The NRC received 15 public comment submissions on the
regulatory basis, which are available for review at www.regulations.gov
under Docket ID NRC-2020-0034. Table 1 of this document provides ADAMS
references for these public comment submissions. Section XXXVI.F.(xiv),
``Discussion of Public Comments on the Fuel Dispersal Aspects of the
Regulatory Basis,'' of this document includes the NRC's summaries of,
and responses to, the comments that the NRC used to inform the
development of this proposed rule and the draft regulatory analysis.
BILLING CODE 7590-01-P
Table 1--ADAMS References for Public Comment Submissions on the
Regulatory Basis.
[GRAPHIC] [TIFF OMITTED] TP16JY26.507
BILLING CODE 7590-01-C
XXXVI. Discussion--Increased Enrichment of Conventional and Accident
Tolerant Fuel Designs for Light-Water Reactors
This proposed rule would amend the current regulations related to
the use of conventional and accident tolerant LWR fuel designs. From
interactions with stakeholders, the NRC is aware that licensees and
applicants plan to request higher fuel burnup limits (i.e., above 62
GWd/MTU rod average) along with the deployment of ATF concepts. To
achieve higher burnup limits, licensees and applicants would need to
request increases in fuel enrichment above the current standard of 5.0
weight percent U-235.
One of the NRC's goals in this rulemaking is to establish effective
and efficient licensing of applications using fuels enriched to greater
than 5.0 weight percent U-235, including reducing the need for requests
for exemptions from existing regulations, while continuing to provide
reasonable assurance of adequate protection of public health and
safety.
The NRC proposes revising requirements in six technical areas.
[[Page 44615]]
A. Criticality Accident Requirements in 10 CFR 50.68
The NRC proposes to amend 10 CFR 50.68(b)(7) to allow licensees or
applicants that comply or propose to comply with that section's
criticality safety requirements to enrich their fuel beyond the current
limit of 5.0 weight percent U-235. Each licensee or applicant would
have the option between the existing 5.0 weight percent U-235
enrichment limit or the value specified in their OL. The enrichment
limit is most often specified as a technical specification design
feature as defined in 10 CFR 50.36(c)(4) and is currently part of
generic standard technical specifications for various reactor
technologies, as described in section XXXV.C.(i), ``10 CFR 70.24 and 10
CFR 50.68 Criticality Accident Requirements,'' of this document.
The proposed change would not affect safety. Licensees and
applicants electing higher enrichments would be required to ensure the
same minimum margin to subcriticality that is required for licensees
implementing 10 CFR 50.68 with fuel enriched up to 5.0 weight percent
U-235. Maintaining these margins of safety would be accomplished by
applying the same keff safety limits for higher enriched fuels as for
fuel enriched up to 5.0 weight percent U-235. Licensees or applicants
that would seek to implement the requirements of 10 CFR 50.68 while
using fuel enriched above 5.0 weight percent U-235 would need to submit
for NRC review and approval a fuel transition license amendment
request. These requests would need to include calculations that show
that the new and spent fuel storage applications demonstrate compliance
with the keff safety limits specified in 10 CFR 50.68.
The proposed revision also would meet one of the purposes of this
rulemaking: allow a licensee the option to implement fuel enriched to
greater than 5.0 weight percent U-235 without requiring the licensee to
seek specific exemptions from 10 CFR 50.68 requirements.
This change would not result in any significant radiological
consequences that could impact plant workers or members of the public.
While increased enrichment would add more radioactive material to the
spent fuel pool, this increase would not change the anticipated dose
rates for occupational dose to workers in and around the fuel storage
areas or for members of the public beyond the site boundary because the
material in the pool is completely shielded by water.
The keff safety limits specified in 10 CFR 50.68(b)(2), (3), and
(4) would be maintained at their current levels with the same required
probability and confidence levels. The feasibility study contracted by
the NRC, ORNL/TM-2024/3350, ``Scoping Studies on the Impacts of
Increased Enrichment on Nuclear Criticality Safety,'' May 2024,
indicates that existing fuel technologies, like integral fuel burnable
adsorber coatings and gadolinium burnable poisons, should be able to
maintain compliance with 10 CFR 50.68 requirements for the entire range
of low enriched uranium (i.e., up to 19.75 weight percent U-235).
However, if the increased enrichment were to adversely impact the
subcriticality requirements, then additional modifications to the fuel
storage facilities would be required to restore compliance.
The NRC received several comments on the regulatory basis that
suggested that the NRC revise RG 1.240, ``Fresh and Spent Fuel Pool
Criticality Analyses,'' in coordination with any changes to 10 CFR
50.68. In March 2021, the NRC issued RG 1.240, which endorses NEI 12-
16, Revision 4, ``Guidance for Performing Criticality Analyses of Fuel
Storage at Light-Water Reactor Power Plants,'' dated September 2019.
The NRC reviewed this guidance during the development of the regulatory
basis and determined that neither RG 1.240 nor NEI 12-16, Revision 4 is
specifically dependent on enrichment levels, whether at the current
levels or those considered in this proposed rule. During this review,
the NRC also noted that RG 1.240, section C.1.o specifies that the
document's recommendations are based on existing fuel applications
currently in widespread industry use, and that new and novel
configurations and concepts implemented in the future may require
additional justification for continued use of the assumptions and
recommendations. However, the NRC did not identify any cases where the
regulatory guidance would not be applicable to the enrichment levels
being considered in this proposed rule. Further, licensees may provide
any necessary justification as part of license amendment requests.
Therefore, the NRC concluded that the guidance does not need to be
immediately updated as a part of this rulemaking effort. The NRC
observes that future revisions of RG 1.240 desired by industry
stakeholders should be pursued separately under the standard regulatory
guide maintenance and revision process.
B. Uranium Fuel Cycle Environmental Data--Table S-3 in 10 CFR 51.51
The NRC proposes to provide a regulatory justification in 10 CFR
51.51(b) for the use of table S-3 for fuel enrichment up to 20.0 weight
percent U-235. This rulemaking action is predicated on the information
provided by current LWR licensees to use enriched nuclear fuel of up to
10.0 weight percent U-235 and by new reactor developers to use high-
assay low enriched uranium with enrichment levels greater than 10.0
weight percent U-235 and less than 20.0 weight percent U-235. The
regulatory justification would be included in a proposed amendment of
10 CFR 51.51(b), table S-3, note 1. Specifically, note 1 would be
amended to add a discussion of NUREG-2249, ``Generic Environmental
Impact Statement for Licensing of New Nuclear Reactors--Final Report,''
for the environmental effects of up to 20.0 weight percent U-235 on the
uranium fuel cycle as still bounded by table S-3, and text explaining
the rationale for the addition of this document. NUREG-2249 discusses
the High-Assay Low-Enriched Uranium (HALEU) Availability Program under
the DOE with the direction to secure a domestic supply of HALEU fuel
following the Energy Act of 2020. As outlined in NUREG-2249 section
3.14.1.3, DOE has identified and contracted with partners for
enrichment services for the production of HALEU as UF6.
Additionally, DOE has identified and contracted with partners for
deconversion of HALEU stored as UF6 to other chemical forms
(i.e., metal or oxide) for fuel fabrication purposes. The DOE HALEU
Availability program is ongoing and has produced the first quantities
of HALEU with further expansion of production capacities expected over
the next several years. NUREG-2249 concludes that for the enrichment of
uranium, table S-3 would bound the environmental impacts from a
centrifuge enrichment facility to produce HALEU and the impact would be
SMALL.
Note 1 would also be amended to add a reference to NUREG-2266,
``Environmental Evaluation of Accident Tolerant Fuels with Increased
Enrichment and Higher Burnup Levels,'' for the environmental effects of
up to 80,000 MWd/MTU maximum assembly averaged burnup as still bounded
by table S-3, and text explaining the rationale for the addition of
this document. This is because the analysis in WASH-1248 was based on
12-month refueling cycles and lower enrichment and burnup levels than
are used for the
[[Page 44616]]
current fleet of LWRs. The higher burnup levels achieved since issuance
of WASH-1248 result in greater utilization of the uranium fuel (i.e.,
greater efficiency in extracting energy from the fuel). This also has
resulted in extended time between refueling operations and the removal
of fewer fuel assemblies on a per reactor-year basis for many of the
operating nuclear power plants. Deployment and use of nuclear fuels
with increased enrichment and higher burnup levels would result in
further increases in fuel efficiency in extracting energy, resulting in
further reductions in the number of spent nuclear fuel assemblies
removed during refueling operations and further extending the time
between refueling operations. Thus, the use of nuclear fuels with
increased enrichment and higher burnup levels would reduce the annual
nuclear fuel needs to support refueling to below the figures that were
used in the analysis supporting Table S-3.
C. Environmental Effects of Transportation of Fuel and Waste--Table S-4
in 10 CFR 51.52
The NRC proposes to permit the use of table S-4 of 10 CFR 51.52 for
fresh nuclear fuel shipments with increased enrichment up to 8.0 weight
percent U-235 by applying the supporting transportation analysis from
NUREG-2266. The maximum enrichment level in the supporting core
analysis relied upon in NUREG-2266 went as high as 8.0 weight percent
U-235. This proposed amendment would include a change to the enrichment
and burnup conditions under 10 CFR 51.52(a), adding a new note 2 to
table S-4 along with moving the current note 2 to note 3, and adding a
new note 4 to table S-4 to replace the current note 3. The
transportation analyses in NUREG-2266 would be referenced in table S-4,
note 2, with text explaining the rationale for the addition of this
document for addressing this enrichment level as well as a level of
burnup of up to 80,000 MWd/MTU for UO2 fuel. Any licensing
action with enrichments above 8.0 weight percent U-235 would have to be
addressed on a case-by-case basis, in accordance with 10 CFR 51.52(b),
by providing a full description and detailed analysis of the
environmental effects of the transportation of fuel and waste. The
methodology in NUREG-2266, section 3, ``Transportation,'' could be
applied for such a detailed analysis. Note 2 would also include the
environmental effects of burnup levels of up to 133,000 MWd/MTU for
TRISO fuel as still bounded by table S-4.
This rulemaking also proposes revisions and updates to 10 CFR
51.52(a), (b), and (c), which provide for the evaluation of the
environmental impacts of transportation of fuel and waste to and from
the reactor for LWRs. Reactors other than light-water-cooled nuclear
power reactors (i.e., non-LWRs) will utilize the same uranium fuel
cycle as LWRs with the transportation of material between uranium fuel
cycle stages plus transportation to and from the nuclear power plant.
The transportation of fuel to and waste from non-LWRs, like
transportation of fuel and waste for LWRs, is necessary to support the
operation of the plant. Therefore, the environmental impacts of
transportation, regardless of whether for an LWR or non-LWR, must be
assessed by the NRC. Similarly, the NRC's regulations in 10 CFR
51.50(b)(3) and (c) already provide for the evaluation of the
environmental impacts of the uranium fuel cycle for non-LWRs. To ensure
there is regulatory clarity for the transportation of non-LWR fuel and
waste in a similar manner as there is for the uranium fuel cycle for
non-LWRs, and for the transportation of fuel and waste for LWRs and
non-LWRs, the words ``other than light-water-cooled nuclear power
reactors'' from 10 CFR 51.50(b)(3) and (c), would be added to the
introductory paragraph of 10 CFR 51.52 and 10 CFR 51.52(b).
The NRC also proposes to update the information in paragraph (c),
table S-4, to the current conditions for the transportation of
radioactive material and the impacts important to such transportation.
The transportation weight values would be adjusted for trucks to the
current U.S. Department of Transportation regulations of 80,000 lb per
truck (23 CFR part 658, ``Truck Size and Weight, Route Designations--
Length, Width and Weight Limitations'') and for rail cars to 240 tons
per cask per rail car based on DOE information on the developed Atlas
spent fuel rail car (DOE Article, ``New Railcar Designed to Transport
Spent Nuclear Fuel Cleared for Operation,'' June 4, 2024). The NRC also
would remove the cumulative dose values from table S-4 because this
information is redundant to the exposed individual doses also being
provided in table S-4 and is a very small fraction of the average
natural background annual radiation exposure of 310 millirem (0.0031
Sv) per person. For example, the 1,469,000 persons along the route for
the updated table S-4 would receive on an annual basis approximately
1469 person-rem from natural radiation sources. Thus, the 3 or 4
person-rem in the current table S-4 would only add the insignificant
amount of approximately 0.3 percent of additional cumulative radiation
exposure from the transportation of fuel and waste for the 1,469,000
persons along the route for all annual shipments.
The $475 property damage per reactor year would be removed from
table S-4 because this value dates back to 1972 in appendix C of WASH-
1238, the risk of a transportation accident is very small as shown in
NUREG-2266, and there has not been a radiological transportation
accident in the United States that would have resulted in property
damage greater than the amount of damages from a commercial hazardous
material transportation accident. If necessary, to determine the
property damage risk from transportation accidents, the same
methodology in appendix C of WASH-1238 can be applied on a case-by-case
basis by assessing the annual shipment miles multiplied by the
probability of an accident per mile and multiplied by the average worth
of property damage per accident with property damage in that year.
D. Fissile Material Packaging Requirements in 10 CFR 71.55
The NRC proposes to increase the allowable enrichment range in its
fissile packaging requirements through a graded approach and include an
additional design requirement for enrichment levels from 5.0 to 10.0
weight percent U-235 to enable the use of the exception in 10 CFR
71.55(g) and remove the requirement to consider water in-leakage. The
existing 5.0 weight percent U-235 enrichment limitation is based on
standard industry and worldwide practice rather than a calculated
effect on criticality of moderator in-leakage for UF6
enriched to this level. Because the basis for approval of the exception
relies on the performance requirement for the cylinder during the
specified tests in 10 CFR 71.73, ``Hypothetical accident conditions,''
the NRC has concluded that the addition of a design feature (e.g.,
valve protection device) of the individual cylinders containing
UF6 with enrichments greater than 5.0 weight percent U-235
would be consistent with the current performance-based requirements
contained in 10 CFR 71.55(g) as well as risk insights gained from
operational experience. This design enhancement requirement would
provide additional defense-in-depth against water in-leakage that
considers the relative consequences of an inadvertent criticality as a
function of increasing enrichment level.
[[Page 44617]]
The NRC considered including in the proposed rule enrichments
between 10.0 weight percent U-235 up to but less than 20.0 weight
percent U-235. Although the NRC could consider additional prescriptive
and non-technology inclusive defense-in-depth design requirements to
mitigate the increasing risk of a criticality event as a function of
this range of enrichment levels, the NRC could not formulate a
technical or regulatory justification for providing a redundant,
prescriptive, and non-technology inclusive exception from 10 CFR
71.55(b), which was already functionally present in 10 CFR 71.55(c) for
all fissile material transportation packages. The provisions in 10 CFR
71.55(c) could be used to achieve the same desired regulatory outcome,
including enrichments for UF6 packages up to but less than
20.0 weight percent U-235, while maintaining the principles of good
regulation of efficiency, clarity, and reliability.
This proposed amendment to 10 CFR 71.55(g) differs from the
regulatory basis recommendation to not pursue a change in the
regulations. Part of the rationale for that recommendation was based on
the staff's assumption that the rule change would be cost-neutral based
on only rulemaking and licensing costs. However, that analysis did not
consider indirect shipping costs to the industry, an exclusion that the
nuclear industry, in comments on the regulatory basis, described as a
departure from the principles of good regulation. The NRC received
public comments on the regulatory basis that demonstrated that there
was sufficient net positive effect for fuel enrichment facilities and
fuel fabricators to consider a rulemaking approach that would be cost-
beneficial and maintain adequate protection of the public health and
safety.
During the development of the regulatory basis for this proposed
rule, the nuclear industry and the NRC identified that a change to 10
CFR 71.55(g) could create a misalignment with U.S. Department of
Transportation and IAEA standards. The American National Standards
Institute (ANSI) was in the process of revising its N14.1 standard,
``Nuclear Materials-Uranium Hexafluoride-Packagings for Transport''
(ANSI N14.1), when the regulatory basis was developed. The standard
ANSI N14.1 was subsequently finalized, and its incorporation into U.S.
Department of Transportation regulations will likely remove any
potential misalignment that was identified in the regulatory basis.
Specifically, ANSI N14.1 maintains a reference to a 5.0 weight percent
U-235 limit but now allows for additional flexibility to the regulatory
authority to make the final safety determination independent of an
enrichment limit. In addition, future routine harmonization activities
for transportation regulations with the IAEA will also likely remove
any misalignment.
E. Control Room Requirements in 10 CFR 50.67 and GDC 19
This proposed rule would increase the numerical value of the
control room design criteria from 5 rem to 10 rem (0.05 to 0.10 Sv);
the value may range up to 25 rem (0.25 Sv) TEDE with consideration of
the plant-specific risk profile or risk information. Increasing this
value would support increased fuel enrichments and the expected
associated increases in power levels and fuel burnup by preserving
operational flexibility by providing additional safety margin and
avoiding occupational exposures. In addition, the NRC would make minor
editorial changes to 10 CFR 50.67 to remove the term ``total effective
dose equivalent'' after the abbreviation ``TEDE'' is provided for that
term.
Comments on the regulatory basis document indicated potential
misunderstanding and concern about potential outcomes resulting from
the proposal to amend the control room design criteria. Although the
control room design criteria are distinct from operational dose limits,
the NRC recognizes the two concepts share some similarities.
Specifically, both the operational occupational exposure limit in 10
CFR part 20 and the control room design criteria are numerically
equivalent and use the same unit of ``rem TEDE.'' As part of this
rulemaking effort, the NRC attempts to clearly explain that the
proposed changes would only be to the control room design criterion and
would not change normal operational and emergency exposure limits of 10
CFR part 20.
The NRC recognizes the challenges that licensees face to retain
margin within their licensing bases for the purposes of operational
flexibility and the small amount of margin to the control room design
criteria itself. The key driver behind the proposal to amend the
control room design criteria is to facilitate increased regulatory
efficiency and consistency while continuing to provide adequate
protection of public health and safety. An unjustifiably low design
criteria could unnecessarily burden licensees for seeking increased
enrichments by requiring extensive analyses to preserve margin for
operational flexibility purposes. As discussed in section XXXV.C.(iv),
``Appendix A to 10 CFR part 50 (General Design Criterion 19) and 10 CFR
50.67(b)(2)(iii),'' of this document, these analyses do not necessarily
result in safety benefits and can increase actual operational exposure
to workers due to increased maintenance activities.
The proposed numerical value of the control room design criteria in
10 CFR 50.67 and GDC 19 would increase from 5 to 10 rem (0.05 to 0.10
Sv) with a consideration of the plant-specific risk profile or risk
information. Nuclear power reactor licensees would benefit from a
higher, and safe, performance level between 5 and 10 rem (0.05 and 0.10
Sv) TEDE when implementing advanced nuclear fuel technologies and
operational flexibility. If additional operational flexibilities are
needed beyond 10 rem (0.10 Sv) TEDE, facility-specific risk profile or
risk information can be leveraged to justify a higher numerical value
up to 25 rem (0.25 Sv) TEDE.
Under the current regulations in 10 CFR 20.1201 and 20.1206, an
adult worker can receive radiation exposure of up to 10 rem (0.10 Sv)
TEDE within a single calendar year or over a 12-month period straddling
two calendar years under normal operations. The control room design
criterion of 5 rem (0.05 Sv) TEDE, which is intended to assess the
acceptability of a given control room design for a potential reactor
accident of exceedingly low probability, is at least a factor of two
lower than what is found to be acceptable under normal operations.
Thus, the proposed rule increase in the control room design criterion
to 10 rem (0.10 Sv) TEDE would be consistent with the Commission's
current regulations for normal operations.
The proposed rule would enable a higher control room design
criteria, ranging from 10 to 25 rem (0.10 to 0.25 Sv) TEDE, for
licensees whose facility-specific risk profiles warrant them. This
range is consistent with recommendations from national and
international organizations responsible for radiation protection
standards. These recommendations are based on fundamental modern health
physics and radiation epidemiology knowledge. These organizations
generally recommend emergency exposure doses up to 25 rem (0.25 Sv)
TEDE or 50 rad (0.5 gray) whole body. Thus, the proposed control room
design criterion of 10 rem (0.10 Sv) TEDE intended to assess the
acceptability of a given control room design for a potential reactor
accident of exceedingly low probability is generally bounded by
recommended values to protect against radiation exposure during an
accident. As described in section XXXV.C., ``Background and History of
Affected
[[Page 44618]]
Regulations,'' of this document, updated scientific recommendations for
radiation protection for workers under accident and emergency
conditions help form the technical basis for the proposal to increase
the control room design value from 5 rem (0.05 Sv) TEDE originally
based on the occupational exposure limit in 10 CFR part 20.
The upper range of the proposed numerical values would be
consistent with the Commission's use of the 25 rem (0.25 Sv) TEDE limit
primarily in regulations for power reactor siting to protect the public
during emergencies, as specified in 10 CFR 100.11, ``Determination of
exclusion area, low population zone, and population center distance'';
10 CFR 50.34, ``Contents of applications; technical information''; 10
CFR 50.67; and 10 CFR part 52 for the exclusion area boundary and low
population zone. As discussed in the preamble for the final rule
updating the NRC's siting criteria (61 FR 65157; December 11, 1996),
the Commission's use of 25 rem (0.25 Sv) TEDE does not imply that the
Commission considers it to be an acceptable limit for an emergency dose
to the public under accident conditions, but only that it represents a
reference value to be used for evaluating plant features and site
characteristics intended to mitigate the radiological consequences of
accidents in order to provide assurance of low risk to the public under
postulated accidents. The Commission, based upon extensive experience
in applying this criterion and in recognition of the conservatism of
the assumptions in its application (i.e., a large fission product
release within containment associated with major core damage; maximum
allowable containment leak rate; a postulated single failure of any of
the fission product cleanup systems, such as the containment sprays;
adverse site meteorological dispersion characteristics; an individual
presumed to be located at the boundary of the exclusion area at the
centerline of the plume for two hours without protective actions),
determined that the 25 rem (0.25 Sv) TEDE criterion clearly resulted in
an adequate level of protection. As an illustration of the conservatism
of this assessment, the Commission noted that the maximum whole-body
dose received by an actual individual during the accident at Three Mile
Island Nuclear Station in March 1979, which involved major core damage,
was estimated to be about 0.1 rem (0.001 Sv).
A review of modern health physics and radiation epidemiology
knowledge provides further technical background for proposing to amend
the control room design criteria to a higher, and safe, performance
level. An important distinction from this review (see section XXXV.C.,
``Background and History of Affected Regulations,'' of this document)
highlights that the control room design criterion radiation unit of
``rem TEDE'' does not technically correspond with the expected measured
deterministic health effects from a reactor accident that would prevent
operators from performing their safety function of protecting the
public health and safety. These deterministic health effects are best
expressed in the radiation unit of ``rad.'' The 10 CFR part 20 annual
occupational exposure limit of 5 rem (0.05 Sv) TEDE, which is
applicable under accident and emergency conditions, is set sufficiently
low that no deterministic threshold dose would be reached.
To clarify the purpose of the control room design criteria
contained in GDC 19 and repeated in 10 CFR 50.67, and to distinguish
the control room design criteria from the radiation protection and EP
frameworks, the NRC is proposing several editorial changes to both
provisions. The phrase ``Adequate radiation protection'' in 10 CFR
50.67(b)(2)(iii) and GDC 19 would be replaced with ``The necessary
design, fabrication, construction, testing, and performance criteria
for structures, systems, and components important to safety.'' As
explained in the introduction to appendix A to 10 CFR part 50,
``General Design Criteria for Nuclear Power Plants,'' SSCs important to
safety are those SSCs that provide reasonable assurance that the
facility can be operated without undue risk to the health and safety of
the public. In the case of control room design, the original role of
GDC 19 was to ensure that adequate SSCs were provided to permit
occupancy of the control room during an accident. The adequacy of the
control room SSCs was to be determined by the ability of workers to
occupy the control room for the duration of an accident without
exceeding the radiological design criteria specified in GDC 19.
However, over time the phrase ``adequate radiation protection'' in GDC
19 has been conflated with the NRC's statutory standard of adequate
protection. Adequate protection is achieved through a licensee's
compliance with the NRC's comprehensive regulatory framework, of which
the GDC is one part, and not just a particular design criterion. To
increase consistency between these rules, the phrase ``Adequate
radiation protection'' in 10 CFR 50.67(b)(2)(iii) and GDC 19 would be
replaced with ``The necessary design, fabrication, construction,
testing, and performance'' to be consistent with appendix A to 10 CFR
part 50. Appendix A to 10 CFR part 50 requires that the principal
design criteria establish the necessary design, fabrication,
construction, testing, and performance requirements for SSCs important
to safety.
To further clarify the purpose of 10 CFR 50.67 and GDC 19 as they
relate to the radiation protection and emergency response frameworks,
the phrase ``personnel receiving'' in 10 CFR 50.67 and GDC 19 would be
replaced with ``calculated'' and the phrase ``access to and'' would be
deleted because the traditional DBA radiological consequence analyses
performed to demonstrate compliance with the criteria do not assess
actual ``personnel receiving'' radiation exposures or plant personal
traveling from the site boundary to the control room. The use of a
dose-based control room design criterion does not imply that it would
be an acceptable exposure during emergency conditions, or that the
radiation protection standards and emergency response standards of 10
CFR part 20 and part 50 might not apply. Rather, these analyses assess
the acceptability of design provisions for protecting control room
operators under postulated DBA conditions. The DBA conditions assumed
in these analyses, although credible, generally do not represent actual
accident sequences. These DBA conditions are specified as conservative
surrogates to create bounding conditions for assessing the
acceptability of engineered safety features.
However, rare events (e.g., events involving multiple failures) can
exceed the design basis of the facility originally envisioned by the
designers. During such events, the Commission's regulations for
radiation protection and emergency response programs require licensees
to take measures to minimize actual radiation exposures. The on-shift
emergency coordinator has the authority and responsibility to
immediately and unilaterally initiate any emergency actions. These
emergency actions include establishing higher exposure limits if
necessary to provide public health and safety. Furthermore,
arrangements are made not only with respect to the detection and
assessment of dose or intake of ionizing radiation, but also with
respect to the mitigating interventions that may have to be applied to
further protect workers. The traditional DBA radiological consequence
analyses do not necessarily credit these mitigative interventions as
they do not directly assess the performance of the control room
[[Page 44619]]
habitability envelop design itself. As a result, actual doses received
during an event are expected to be significantly lower than the
computed results in realistic accident scenarios.
The graded, risk-informed, and performance-based framework
developed for DG-1425 (proposed revision 2 of RG 1.183), ``Alternative
Radiological Source Terms for Evaluating Design-Basis Accidents at
Nuclear Power Reactors,'' would enable a performance-based evaluation
using traditional deterministic radiological consequence analysis
methods within defined risk-informed boundaries as described in
``Method for Graded Risk-Informed Performance-Based Control Room Design
Criteria Framework,'' dated September 2024. These boundaries would be
defined by acceptable radiation exposure guidelines for radiation
workers during accident and emergency conditions and acceptable
contemporary nuclear facility risk profiles using modern PRA methods.
Such a framework would provide flexibility when determining how to meet
an established acceptance criterion in a way that encourages and
rewards safety of the facility consistent with the Commission's policy
in SRM-SECY-98-144, ``Staff Requirements--SECY-98-144--White Paper on
Risk-Informed and Performance-Based Regulation,'' dated March 1, 1999.
In practice, the method would produce a framework that uses, in part,
the facility's safe design and operations to justify a higher control
room design criterion with a lower plant-specific risk metric.
The DG-1425 framework leverages a licensee's existing PRA model.
Acceptability of the PRA model used to demonstrate that the specified
criterion is commensurate with the risk of the plant is determined for
the following aspects: scope, level of detail, conformance with PRA
technical elements (i.e., technical robustness), and plant
representation and PRA configuration control. The PRA model would be
consistent with the philosophy in RG1.174 and the technical adequacy
expectations for the model in RG 1.200. For instance, the use of
overall core damage frequency (CDF) results from an NRC-approved
license amendment request that incorporates the risk-informed
completion time program into the facility's technical specifications
(i.e., Technical Specifications Task Force Traveler 505, ``Provide
Risk-Informed Extended Completion Times--RITSTF Initiative 4b'') would
be acceptable. The baseline PRA model would estimate the overall CDF
for all significant sources of risk both internal and external to the
plant (e.g., internal, flood, fires, seismic, high winds, and others).
The CDF risk metric would be the most appropriate for the purposes
of a graded, risk-informed, and performance-based control room design
criteria framework. This is because CDF accounts for a broad range of
accident scenarios and can generally encompass the risk relevant to the
sequences considered for control room habitability when deriving the
maximum hypothetical accident source term. There is also consistency
between the CDF risk metric, which does not consider radiation
protection protective actions, and the traditional DBA radiological
consequence analysis performed to demonstrate compliance. Additionally,
control room habitability design primarily concerns the ability of
personnel to maintain reactor safety during and after accidents, which
aligns closely with the overarching goals of preventing core damage and
mitigating radiological releases, which are captured by the CDF risk
metric.
In the regulatory basis for this proposed rule, the NRC sought
comments on the alternatives proposed in that document's appendix A,
``Control Room Requirements.'' Additionally, the NRC asked two
questions. The first question sought input as to whether the numerical
selection of the control room design criteria would be better aligned
with regulations designed to limit occupational exposures during
emergency conditions or regulations designed to limit annual
occupational radiation exposures during normal operations. The second
question sought input as to whether a graded, risk-informed method to
demonstrate compliance with a range of acceptable control room design
criterion values instead of a single selected value, such as the
current 5 rem (0.05 Sv) TEDE, provides the necessary flexibilities for
current and future nuclear technologies.
Overall, public comments were supportive of the staff's
recommendation to amend the control room design criteria. The comments
in support of alternative 2 generally suggested a value of 25 rem (0.25
Sv) TEDE be applied in amended regulations. Commenters stated that
using a value of 25 rem (0.25 Sv) TEDE would be more consistent with
the various U.S. and international organizations' recommendations for
emergency dose limitations up to 25 rem (0.25 Sv) TEDE. Nearly all
comments included suggestions to develop a graded, risk-informed
approach to the control room design criteria. Several comments relied
on PRA technology and methods and contemporary understandings of
facility risk to justify a higher numerical value for low-probability,
high-consequence events.
Based, in part, on the comments received, the NRC is proposing to
increase the numerical value of the control room design criteria from 5
to 10 rem (0.05 to 0.10 Sv) TEDE but range up to 25 rem (0.25 Sv) TEDE
with a consideration of the plant-specific risk profile or risk
information. In response to stakeholder interest, the NRC developed a
graded, risk-informed, and performance-based control room design
criteria framework for the supporting regulatory guidance.
F. Fuel Dispersal
(i) Overview
Based on the Commission's direction in SRM-SECY-21-0109, the public
comments on the regulatory basis (see section XXXVI.F.(xiv),
``Discussion of Public Comments on the Fuel Dispersal Aspects of the
Regulatory Basis,'' of this document) for details on the public
comments received on the fuel dispersal portions of the regulatory
basis), the anticipated impacts on this rulemaking's schedule from each
of the alternatives described in the regulatory basis, the technical
maturity of those alternatives, and the anticipated impact on safety of
each of the alternatives, the NRC proposes to address fuel dispersal in
this rulemaking. Specifically, the proposed rule would designate LOCAs
above the TBS as beyond-design-basis, allowing for best-estimate
analysis for such LOCAs, while instituting performance-based cladding
embrittlement criteria, and clarifying and updating the NRC's
definition of coolability for the LOCA event to encompass both fuel in
the reactor core and any fuel dispersed into the RCS or containment.
The proposed flexibility in ECCS analyses for LOCAs above the TBS
might enable entities using the proposed rule to demonstrate that no
fuel dispersal occurs for LOCAs above the proposed TBS. These entities
would be able to use the best-estimate (i.e., based on conditions
consistent with expected, nominal operating conditions without biases
or uncertainties) modeling under proposed 10 CFR 50.46a(e)(3) for this
category of LOCAs until more data or analyses are developed to address
fuel dispersal in another way, such as a demonstration that the fuel
remains coolable if there is fuel dispersal and the other downstream
consequences of dispersal do not have any significant
[[Page 44620]]
deleterious impacts. The proposed rule would redesignate existing 10
CFR 50.46a (which contains acceptance criteria for RCS venting systems)
as 10 CFR 50.46b and establish an alternative set of risk-informed
requirements in a new proposed 10 CFR 50.46a with which entities could
choose to comply in lieu of meeting the current emergency core cooling
system requirements in 10 CFR 50.46. Using these alternative ECCS
requirements would provide some entities with opportunities to change
various aspects of their facility design and operation, although
potential impacts of the changes pertaining to plant physical security
or cybersecurity would be evaluated during license amendment reviews.
As used in proposed 10 CFR 50.46a and this discussion of proposed
10 CFR 50.46a, ``entities'' would include applicants for and holders of
CPs, OLs, COLs, standard design approvals, and MLs, and applicants for
standard design certification rules (including such applicants after
NRC issuance of a final standard design certification rule).
The proposed rule would divide the current spectrum of LOCA break
sizes into two regions. The division between the two regions would be
delineated by the TBS. The first region would include small size
breaks, up to and including the TBS. The second region would include
breaks larger than the TBS, up to and including the DEGB of the largest
RCS pipe. While both sets of breaks are unlikely to occur, the larger
breaks are considered to have a much lower likelihood of occurring than
the smaller breaks in the first region. Under the proposed rule, the
ECCS design requirements for breaks smaller than the TBS would remain
the same as the requirements for all breaks under the current 10 CFR
50.46 ECCS rule. By contrast, under the proposed rule, the ECCS design
requirements for the pipe breaks larger than the TBS could be analyzed
using less conservative assumptions based on their lower likelihood of
occurrence. Although LOCAs for break sizes larger than the TBS would be
classified as ``beyond-design-basis accidents'' for entities that
implement the proposed 10 CFR 50.46a, these break sizes in license
applications would still be subject to regulatory evaluation. The
proposed rule would require that entities maintain the ability to
mitigate all LOCAs, up to and including the DEGB of the largest RCS
pipe. Mitigation analyses for LOCAs larger than the TBS would not need
to assume the loss of offsite power or the occurrence of a coincident
single failure event. Entities also would be allowed to credit the use
of non-safety-grade systems.
Entities who perform LOCA analyses using the proposed risk-informed
alternative requirements could find that their plant design or
operation is no longer limited by certain parameters associated with
previous DEGB analyses. Reducing the DEGB limitations would allow some
entities to propose a wide scope of design or operational changes until
another parameter in required accident analyses becomes limiting.
Potential design changes could include fuel burnup increases and other
management improvements; power uprates; and changes to the required
number of accumulators, diesel start times, sequencing of equipment,
valve stroke times, and containment spray system setpoints. Some of
these design and operational changes could increase plant safety
because an entity could modify its systems to better mitigate the more
likely, but still very rare, smaller LOCAs. Other changes, such as
increasing power, could increase the overall risk of inadvertent
release of radioactive material, which may be acceptable if the overall
plant risk increase is demonstrated to be acceptably small.
The risk-informed proposed 10 CFR 50.46a would include risk
acceptance criteria for evaluating future design changes to ensure that
any risk increases would be acceptably small. These acceptance criteria
would be consistent with the guidelines for risk-informed license
amendments in RG 1.174, ``An Approach for Using Probabilistic Risk
Assessment in Risk-Informed Decisions on Plant-Specific Changes to the
Licensing Basis,'' and ensure both the acceptability of the changes
from a risk perspective and the retention of sufficient defense-in-
depth, safety margins, and performance monitoring. The requirements for
the risk-informed evaluation process are discussed in detail in section
XXXVI.F.(vi), ``Risk-Informed Changes to the Facility, Technical
Specifications, or Procedures,'' of this document.
In addition to changes to current 10 CFR 50.46a, which would be
redesignated as 10 CFR 50.46b, and establishing a new proposed 10 CFR
50.46a, the NRC would make conforming changes to existing 10 CFR 50.46,
and 50.69; GDC 17, 35, 38, 41, 44, and 50 in appendix A to 10 CFR part
50; appendix K to 10 CFR part 50; 10 CFR 52.54; and appendix G to 10
CFR part 52.
(ii) Original Determination of the Transition Break Size
To help determine the TBS in support of the prior 10 CFR 50.46a
rulemaking (see section XXXV.C.(v)(b), ``10 CFR 50.46a Rulemaking,'' of
this document), the NRC developed pipe break frequencies as a function
of break size using an expert elicitation process for degradation-
related pipe breaks in typical BWR and PWR RCSs (NUREG-1829,
``Estimating Loss-of-Coolant Accident (LOCA) Frequencies through the
Elicitation Process,'' March 2008). The elicitation process is used for
quantifying phenomenological knowledge when data or modeling approaches
are insufficient. The NUREG-1829 elicitation focused solely on
determining event frequencies that initiate from failures of the
unisolable reactor coolant pressure boundary (RCPB), or primary system
side, related to material degradation. This effort did not consider the
AP1000 and other similar passive-safety reactor designs.
A baseline TBS was established from the expert elicitation results
for each reactor type (i.e., PWR and BWR) that corresponded to a break
frequency of once per 100,000 reactor years (1x10-5, or
10-5 per reactor year). The NRC then considered uncertainty
in the elicitation process, other potential mechanisms that could cause
passive component failure that were not explicitly considered in the
expert elicitation process, and the higher susceptibility to rupture/
failure of specific locations in the RCS by adjusting the TBS upward to
account for these factors. Other mechanisms that contribute to the
overall LOCA frequency include LOCAs resulting from failures of non-
passive components and LOCAs resulting from low probability direct and
indirect events (e.g., earthquakes of magnitude larger than the safe
shutdown earthquake and dropped heavy loads). These LOCAs have a strong
dependency on plant-specific factors.
LOCAs caused by failure of non-passive components, such as stuck-
open valves and blown out seals or gaskets, have a greater frequency of
occurrence than LOCAs resulting from the failure of passive components.
LOCAs resulting from the failure of non-passive components would be
small-break LOCAs, when considering the size of the opening that could
result should components fail open or blow out (e.g., safety valves,
pump seals). LOCAs resulting from stuck-open valves are limited by the
size of the auxiliary pipe. In some PWRs, there are large loop
isolation valves in the reactor pressure vessel outlet and inlet
piping. However,
[[Page 44621]]
a complete failure of the valve stem packing is not expected to result
in a high rate of coolant loss due to the size of the resulting
penetration in the system, because the valves are sealed in such a way
that limits leaks when they are open (i.e., they are back-seated in the
open configuration). Based on these considerations, non-passive LOCAs
are relatively small in size and are bounded by the selected TBS.
LOCAs could also be caused by dropping heavy loads that could cause
a breach of the RCS piping or damage safety-related equipment. The
majority of heavy loads are lifted during refueling when the reactor is
shut down and the primary system is depressurized, further reducing the
risk of a LOCA and a loss of core cooling. During power operation,
personnel entry into the containment is typically infrequent and of
short duration and the largest cranes are generally not accessible.
There are also operational limitations designed to limit risk due to
heavy load drops. Consequently, loads moved at power are substantially
fewer than during refueling. In addition, the RCS is inherently
protected by surrounding concrete walls, floors, missile shields, and
biological shielding. For these reasons, the NRC did not consider the
contribution of heavy load drops to overall LOCA frequency to be
significant or affect the TBS.
Seismically induced LOCA break frequencies can vary greatly from
plant to plant because of factors such as site seismicity, seismic
design considerations, and plant-specific layout and spatial
configurations. Seismic break frequencies are also affected by the
amount of pipe degradation occurring prior to postulated seismic
events. Seismic PRA insights were accumulated from the NRC Seismic
Safety Margins Research Program and the Individual Plant Examination of
External Events submittals available in the late 1990s. Based on these
studies, piping and other passive RCPB components generally exhibit
high seismic capacities and, therefore, are not significant risk
contributors. However, these studies did not explicitly consider the
effect of degraded component performance on the risk contributions.
Therefore, the NRC conducted a study in the early 2000s to evaluate the
seismic performance of undegraded and degraded passive system
components (NUREG-1903, ``Seismic Considerations for the Transition
Break Size,'' February 2008). This effort examined operating
experience, seismic PRA insights, and models to evaluate the failure
likelihood of undegraded and degraded piping. The operating experience
review considered passive component failures that have occurred as a
result of strong motion earthquakes in nuclear and fossil power plants
as well as other industrial facilities. No catastrophic failures of
large pipes resulting from earthquakes between 0.2g and 0.5g (where g
is the gravitational acceleration or approximately 9.81 meters/
second\2\) peak ground acceleration have occurred in power plants.
However, piping degradation could increase the LOCA frequency
associated with seismically induced piping failures. The NUREG-1903
report evaluated seismic loadings on degraded piping and concluded that
a large, pre-existing crack on the order of 30 percent through-wall and
145 degrees around the piping circumference would have to be present
during a large, rare earthquake (i.e., corresponding to the mean annual
frequency of exceedance equivalent to 10-5 or
10-6 per year) in order for pipe failure to occur. The NRC
concluded that the likelihood of flaws large enough to fail during such
a seismic event was sufficiently low that the TBS need not be modified
to address seismically induced direct piping failures.
Indirect RCPB failures are primary system ruptures that are a
consequence of failures in primary and non-primary system components or
structural support failures (such as reactor coolant pump supports and
steam generator supports). Structural support failures could then cause
displacements in components, causing stress on the piping and potential
failure. The NRC performed studies on two plants to estimate the
conditional pipe failure probability due to structural support failure
given a large, rare earthquake (i.e., 10-5 to
10-6 per year). These studies used seismic hazard curves
from NUREG-1488, ``Revised Livermore Seismic Hazard Estimates for
Sixty-Nine Nuclear Power Plant Sites East of the Rocky Mountains,''
April 1994. The results of these studies, as described in NUREG-1903,
showed that indirectly induced piping failure attributable to major
component support failure has a mean failure probability on the order
of 10-6 per year, which was less than the TBS criterion.
However, the NRC noted in NUREG-1903 that indirect failure analyses are
highly plant-specific. Therefore, it is possible that example plants
assessed in the NRC analyses were not necessarily limiting for all
plants.
The NRC considered the importance of indirect failures on the
selection of the TBS. For the cases considered in NUREG-1903, the
likelihood of indirectly induced piping failures resulting from major
component support failures was less than 10-5 per reactor
year, the frequency criterion used to select the TBS. Also, the median
seismic capacities for both the primary piping system and primary
system components are typically higher than other safety-related
components within the nuclear power plant. Because of these relative
capacities, the NRC expected that a seismic event of sufficient
magnitude to cause consequential failure within the primary system
would also induce failure of components in multiple trains of
mitigation systems, or even induce multiple RCS pipe breaks.
Consequently, the risk contribution from seismically induced indirect
failures was expected to depend more heavily on the relative
fragilities of plant components and systems than the size of the TBS.
Therefore, the NRC determined that adjustment to the TBS for
seismically induced indirect LOCAs was not warranted.
The final consideration in selecting the TBS was actual piping
system design (e.g., piping sizes) and operating experience. For
example, due to system configuration and operating environment, certain
piping was considered to be more susceptible to degradation and failure
than other piping in the same size range.
For PWRs, the NRC determined that 6- to 10-inch inside diameter
(i.e., inside dimension) was an appropriate range of pipe break sizes
associated with the 95th percentile LOCA frequency estimates of
1x10-5/yr from NUREG-1829. This range is only slightly
smaller than the PWR surge lines, which are attached to the RCS main
loop piping (i.e., hot leg, cold leg, and crossover leg) and are
typically 12- to 14-inch diameter Schedule 160 piping with inside
diameters of 10.1 to 11.2 inches. The RCS main loop piping is in the
range of 30 inches in diameter and has substantially thicker walls than
the surge lines. The expert elicitation panel concluded that this main
loop piping is much less likely to break than other RCS piping. The
shutdown cooling lines and safety injection lines may also be 12- to
14-inch diameter Schedule 160 piping and are likewise connected to the
RCS. In some cases (e.g., Babcock and Wilcox plants), the core flood
lines may be bigger than the surge and residual heat removal lines that
are attached to the main loop piping. The difference in diameter and
thickness of the reactor coolant piping and the piping connected to it
forms a reasonable line of demarcation to define the TBS. Therefore, in
SECY-10-0161, to capture the surge, shutdown cooling, core flood,
[[Page 44622]]
and safety injection lines in the range of piping considered to be
equal to or less than the TBS, for PWRs, the NRC staff specified the
TBS as the largest cross-sectional flow area of the RCPB piping
excluding the main loop piping.
For BWRs, the NRC determined that 13- to 20-inch inside diameter
was an appropriate range of pipe break sizes associated with the 95th
percentile LOCA frequency estimates of 1x10-5/yr from NUREG-
1829. The information gathered from the elicitation for BWRs also
showed that the estimated frequency of pipe breaks dropped markedly for
break sizes beyond the range of approximately 18 to 20 inches. After
evaluating BWR designs, the NRC determined that typical residual heat
removal piping connected to the recirculation loop piping and feedwater
piping is about 18 to 24 inches in diameter. These pipe sizes are
consistent with break sizes beyond which the pipe break frequency was
expected to decrease markedly below 10-5 per year. The NRC
staff recognized that the sizes of attached pipes vary somewhat among
plants. Thus, for BWRs, in SECY-10-0161, the staff specified the TBS as
the larger cross-sectional flow area of either the feedwater or the
residual heat removal piping inside primary containment.
Because the effects of TBS breaks on core cooling vary with the
break location, the NRC evaluated whether the frequency of TBS breaks
varies with location and whether TBS breaks could, therefore, vary in
size with location. In PWRs, the pressurizer surge line is only
connected to one hot leg and the pipes attached to the cold legs are
generally smaller than the surge line. The cold legs (including the
intermediate legs) also operate at slightly cooler temperatures such
that thermally activated degradation mechanisms would be expected to
progress more slowly in the cold leg than in the hot leg. The frequency
of occurrence of a break of a given size is composed of both the
frequency of a completely severed pipe of that size (i.e., a complete
circumferential break) plus the frequency of a partial break of that
size in an equal or larger size pipe (i.e., a partial circumferential
or longitudinal break). Therefore, the NRC considered an option where
the TBS for the hot and cold legs would be distinctly different by
considering the frequency contributions of these two break components:
(1) complete breaks of the pipes attached to the hot or cold legs at
the limiting locations within each attached pipe, and (2) partial
breaks of a constant size, as appropriate for either the hot or cold
leg, at the limiting locations within the hot or cold legs. However,
the elicitation was not envisioned to develop LOCA frequencies specific
to piping systems. As a result, there was insufficient detail from the
elicitation to draw conclusions about either the difference between hot
and cold leg failure frequencies or the frequency of occurrence of
smaller LOCAs within a large diameter pipe. Therefore, the NRC
concluded that the TBS associated with partial breaks in the hot and
cold legs should remain equivalent in size to the internal cross-
sectional area of the largest piping system other than the main loop.
Similarly, the elicitation results do not contain sufficient detail to
quantify break frequency differences among the BWR recirculation,
residual heat removal, and feedwater system piping. Thus, a smaller
partial break TBS criterion also could not be established for BWR
recirculation piping. Notably, mitigating the effects of such partial
breaks up to and including a TBS break remains within the design basis
for all RCPB piping with an inner diameter equivalent to or larger than
the TBS.
During this time, the NRC also evaluated whether TBS breaks should
be analyzed as single-ended or double-ended breaks. A postulated
double-ended break assumes that the pipe rupture causes a complete
separation and displacement of both ends of the pipe at the break such
that coolant loss occurs from both sides of the displaced piping. A
single-ended break results in an orifice through which the coolant
would flow. To address this issue, the NRC reviewed the expert
elicitation process and the guidance given to the experts in developing
their frequency estimates. The NRC concluded that the expert
elicitation LOCA frequency estimates correspond to a break area having
an equivalent circular diameter at each break size. This correspondence
is representative of a single-ended break. Additionally, the experts
based their estimates on knowledge of postulated failure mechanisms in
pressure boundary components and not on the flow rates emanating from
the breaks. The flow rates are governed by the break location and
system configuration, which determines whether reactor coolant will be
discharged from both ends of the break.
The current design-basis analysis for LWRs requires analysis of a
DEGB of the largest pipe in the RCS. Under the proposed rule, all
breaks up to and including the TBS would be analyzed under existing
requirements. A possible reason for specifying the TBS for PWRs as
double-ended could be that a complete break of the pressurizer surge
line would result in reactor coolant exiting both ends of the break.
Although this occurs initially during a LOCA, core cooling requirements
are dominated by the flow rate of coolant exiting from the hot leg side
of the break, with much less contribution from the flow rate of coolant
exiting from the pressurizer side. Therefore, specifying the TBS break
as an area equivalent to a double-ended break of the surge line would
be overly conservative. For BWRs, the effect of a double-ended break
area is also considered to be overly conservative. The selected TBS for
BWRs would be based on the larger of the residual heat removal or main
feedwater lines. A single-ended break in these lines would bound
double-ended breaks of the smaller lines in the reactor recirculation
and feedwater system. Therefore, the NRC is proposing that the TBS be
based on a single-ended break, which reasonably characterizes the
expert elicitation results and represents the flow rates associated
with postulated pipe breaks within the RCS. The NRC's proposed TBS
definition is in proposed 10 CFR 50.46a(a)(9). As an option and to
allow maximum flexibility, the proposed 10 CFR 50.46a(a)(9) definition
would allow an entity to develop and justify an alternate TBS.
(iii) Determining the Ongoing Validity of the Transition Break Size
Because the work in the development of the TBS was conducted almost
20 years ago as part of the development of the earlier rulemaking to
create alternative ECCS requirements in 10 CFR 50.46a, the NRC assessed
if the TBS developed in the early 2000s is still valid today. This
research identified possible scenarios not considered, or
underestimated, in NUREG-1829 or NUREG-1903 that could result in
primary pressure boundary breaches that are larger than the TBS in
either PWR or BWR plants. These breaches could be directly due to
operational transients (e.g., anticipated transient without scram,
water hammer, pressurized thermal shock) or indirectly due to other
failures within the plant (e.g., crane drop, secondary side failures).
Age-related degradation of the RCPB components may be a contributing,
or required, causal factor. Additionally, the breach could stem from
failure of a single RCPB component or multiple common-cause RCPB
component failures (e.g., anticipated transient without scram event
leading to the rupture of multiple degraded safety injection system
lines on separate PWR loops). Improper maintenance and human factors
may
[[Page 44623]]
also be a causal factor (e.g., not properly torquing pressurizer manway
bolts following inspection). The likelihood of such breaches was also
considered up to the end of the subsequent license renewal period
(i.e., 80 years) or the maximum extent of the licensing period for
plants that could adopt this proposed rule.
In its effort to determine the ongoing validity of the TBS, the NRC
only considered reactors authorized to operate under 10 CFR part 50 on
December 31, 2015 (i.e., the NRC did not consider the AP1000 design or
any other new LWR reactor design) because the NUREG-1829 study did not
consider plant designs that were authorized to operate after December
31, 2015 or authorized to operate under 10 CFR part 52. LWRs licensed
after that date may have different piping materials, configurations,
and operational and service conditions, among other factors, that may
impact the piping break frequencies and thus the TBS. A detailed
description of the technical justification for the continued
applicability of the NUREG-1829 results is found in the ``White Paper
on Continued Applicability of NUREG-1829,'' dated November 13, 2024,
while the technical justification for the continued applicability of
NUREG-1903 is found in the ``White Paper on Continued Applicability of
NUREG-1903,'' dated November 18, 2024.
To confirm the TBS's current validity, the NRC conducted a series
of probabilistic fracture mechanics analyses using the xLPR code
Version 2.3 to confirm the base cases analyzed in NUREG-1829. For these
analyses, four of the original base case pipe systems were chosen for
this validation: 12-inch recirculation line, 28-inch recirculation
line, 30-inch hot leg, and a 10-inch surge line. The base case
conditions, assumptions, and inputs from NUREG-1829 were generally
adopted in these analyses. However, inputs were supplemented from
Technical Letter Report TLR-RES/DE/REB-2021-09, ``Probabilistic Leak-
Before-Break Evaluation of Westinghouse Four-Loop Pressurized-Water
Reactor Primary Coolant Loop Piping using the Extremely Low Probability
of Rupture Code,'' dated August 13, 2021, and Technical Letter Report
TLR-RES/DE/REB-2021-14, ``Probabilistic Leak-Before-Break Evaluations
of Pressurized-Water Reactor Piping Systems using the Extremely Low
Probability of Rupture Code,'' dated September 28, 2021, for the PWR
cases; and from NUREG-0313, ``Technical Report on Material Selection
and Processing Guidelines for BWR Coolant Pressure Boundary Piping,''
dated January 1988, NUREG/CR-6674, ``Fatigue Analysis of Components for
60-Year Plant Life,'' dated June 2000, and NUREG/CR-4792, ``Probability
of Failure in BWR Reactor Coolant Piping, Volume 1: Summary Report,''
dated December 1988, for the BWR cases as needed. Analyses were run to
80 years, and several sensitivity cases were conducted to investigate
the impacts of mitigation and inspection on the probability of failure.
The annual frequency of small-break, medium-break, and large-break
LOCAs were calculated at 25, 40, 60, and 80 calendar years as well as
the cumulative probability of a crack, leakage, and rupture at 80
years. A summary of the analyses and results can be found in section
5.3 of the ``White Paper on Continued Applicability of NUREG-1829.''
The analyses conducted show that annual frequencies calculated for the
base case problems in the current effort were either bounded by, or
representative of, those determined in NUREG-1829.
In addition to the probabilistic fracture mechanics analyses, the
NRC conducted both an internal and external elicitation similar to the
full elicitation conducted in the original development of NUREG-1829.
The purpose of this elicitation was two-fold. The first objective was
to determine scenarios that could result in primary pressure boundary
breaches that are larger than the TBS in either PWR or BWR plants. The
second objective was to determine the representativeness of NUREG-1829
to the current day. The internal elicitation included subject matter
experts within the NRC with expertise in structural integrity analysis,
materials performance, aged-related degradation, risk assessment, and
thermal-hydraulic analysis. The external elicitation queried two of the
original NUREG-1829 elicitation effort participants. The external
elicitation confirmed that the current-day frequency of LOCAs, and
specifically the frequency of LOCAs having a break size greater than
the TBS, is conservatively represented by the NUREG-1829 estimates.
Neither the internal nor external elicitations identified any generic
issues or scenarios that either were not considered in the TBS
development or have significantly changed since the TBS development
that could undermine its technical basis.
However, both the internal and external elicitations did identify
topics that should be addressed within the proposed rulemaking and
associated guidance. Some of these topics included PRA requirements;
impacts of plant changes; stress corrosion cracking in main loop and
recirculation piping; indirect piping failures; direct and indirect
seismic failure evaluations; maintaining mitigative capabilities;
NUREG-1829 uncertainties; and attributes that could increase plant-
specific LOCA frequencies. Many of these topics were already being
addressed within this rulemaking effort, and the elicitations served to
refine the proposed treatment of these topics, as well as identify some
novel issues that were not initially considered. A more detailed
summary of these elicitations is in the ``White Paper on Continued
Applicability of NUREG-1829.''
The NRC reviewed operational experience since the original NUREG-
1829 effort to use as a basis for determining both piping and non-
piping failure frequencies. There have been only a few small (i.e.,
smaller than 2-inch diameter piping) passive-system primary pressure
boundary ruptures, and the operational experience indicates that
degradation mechanisms such as cracking, wall thinning, or through-wall
leakage are a precursor to a rupture. The precursor event frequency can
be directly calculated from operational experience, while modeling is
required to estimate the likelihood that these precursor events could
lead to LOCAs of various sizes and, hence, estimate LOCA frequencies.
The NUREG-1829 LOCA frequency estimates were based, in part, on
operating experience accumulated up to approximately 2004. The NRC's
more recent effort considered operating experience trends from 1970 to
2004, and then 2005 to the present day. This binning was used to
compare precursor event frequencies and ultimately LOCA frequencies
within the two time periods to assess trending since the completion of
NUREG-1829.
Quantitative LOCA frequency estimates were calculated for each time
period for PWR and BWR systems and for break sizes greater than and
less than the TBS in this proposed rule. Initially, estimates were
developed for each degradation mechanism that is applicable for a
particular piping or non-piping primary pressure boundary system or
component. The attribute-specific frequencies were then multiplied by
the number of attributes (e.g., number of welds) for each component and
then the contributions from each applicable degradation mechanism were
combined to develop component-specific piping and non-piping
frequencies. The component-specific frequencies were then further
combined to determine global piping and non-piping LOCA frequency
estimates. Finally, the piping and non-
[[Page 44624]]
piping contributions are summed so that, ultimately, LOCA frequency
estimates are determined as a function of rupture size. Uncertainties
were initially addressed when determining degradation mechanism-
specific estimates and appropriately combined so that the final
estimates are expressed as distributions with associated mean values
and 5th and 95th percentile LOCA frequency estimates. The findings
predict that both the BWR and PWR large LOCA frequency estimates (i.e.,
the largest NUREG-1829 LOCA size categories starting just below the TBS
up to the highest category, which includes a DEGB of the largest pipe
in the plant), based on operating experience from 2005 to the present
day, are less than estimates based on operating experience from 1970-
2004. Therefore, this work supports the expectation that the NUREG-1829
LOCA frequency estimates used to establish the TBS conservatively
represent the current-day estimates, and that the TBS established for
proposed 10 CFR 50.46a is appropriate. Additional details of this
activity can be found in section 5.2 of the ``White Paper on Continued
Applicability of NUREG-1829.''
In the ``White Paper on Continued Applicability of NUREG-1829,''
the NRC performed a qualitative analysis of the elicitation results,
the operational experience, ongoing American Society of Mechanical
Engineers (ASME) code activities, and recent research findings to
determine if any of these findings impact the proposed TBS or may cause
a break larger than the TBS. The NRC analyzed topics such as thermal
embrittlement of cast austenitic stainless steel and stainless-steel
welds; stress corrosion cracking in secondary PWR stainless lines; the
effects of carbon macrosegregation, small surface breaking flaws, and
quasi-laminar defects on reactor pressure vessel integrity; radiation
embrittlement; and changes to passive-system inspection frequencies.
The NRC evaluated each item's impact on both the direct and indirect
failures and determined that these mechanisms would not impact the
proposed TBS.
The NUREG-1903 report and the original analyses by the NRC
considered the effects of direct (flawed and unflawed) and indirect
piping failures on the selection of the TBS. For the direct unflawed
piping failure, the NRC originally used the screening approach where
the probability of exceedance of stresses corresponding to a 1 percent
probability of failure was obtained for the 26 PWRs for the most highly
stressed hot leg, cold leg, or crossover (suction) legs. The NRC
concluded that failure probabilities of unflawed piping are
significantly low compared to the frequency of 10-5 per year
used as a basis to establish the TBS. For this original assessment, the
NRC used the mean Lawrence Livermore National Laboratory seismic hazard
curves corresponding to each selected site. Since then, all currently
operating U.S. nuclear power reactor licensees have re-evaluated and
submitted their Seismic Hazard and Screening Reports (SHSRs) in
response to the March 12, 2012, letter issued by the NRC under 10 CFR
50.54(f), ``Request for Information Pursuant to Title 10 of the Code of
Federal Regulations 50.54(f) Regarding Recommendations 2.1, 2.3, and
9.3, of the Near-Term Task Force Review of Insights from the Fukushima
Dai-Ichi Accident,'' following the 2011 accident at the Fukushima Dai-
Ichi nuclear power plant. As such, the original assessment results have
been updated by using the more up-to-date site hazard information.
In addition, to better determine whether seismic loading conditions
significantly increases the probability of a break above the TBS in the
current NRC assessment, an unconditional mean piping failure
probability value has been obtained by convolving a site-specific mean
hazard curve with a representative mean large LOCA piping fragility
function obtained from the Electric Power Research Institute (EPRI)
Report (3002000709), ``Seismic Probabilistic Risk Assessment
Implementation Guide.'' The results of this analysis show that failure
probabilities of unflawed piping for a limited number of plants
considered are well below the TBS frequency criterion. In addition, the
study results show that the probabilities of exceedance corresponding
to 1 percent probability of failure are all below the TBS threshold,
even using the most conservative design stress intensity value and the
most conservative failure criterion. Therefore, there is a clear
indication that unflawed piping generally has a very low probability of
failure attributable to seismic loads, which is consistent with the
original NRC assessment conclusion in NUREG-1903 and the excellent
performance experience of piping systems observed during past strong
damaging earthquakes (``Summary and Evaluation of Historical Strong-
Motion Earthquake Seismic Response and Damage to Above-Ground
Industrial Piping,'' April 1985).
The NUREG-1903 report also showed that, for the direct flawed
piping failure, the probabilities of pipe breaks larger than the TBS
are likely to be less than 10-5 per year as the critical
flaws associated with the stresses corresponding to the 10-5
and 10-6 probability of exceedance seismic events are
generally large. However, considering the limited applicability of the
results and the effects of the recent seismic hazard updates on the
TBS, this conclusion needs to be verified by an entity on a case-by-
case basis under proposed 10 CFR 50.46a(c)(1)(i).
For the two indirect piping failure cases considered in the
original NUREG-1903 analysis, the likelihood of indirectly induced
piping failures resulting from major component support failures is less
than 10-5 per year, which was the frequency criterion used
to select the TBS. Based on this frequency criterion, the NUREG-1903
report concluded that indirectly induced piping failure is unlikely to
govern the combined failure of piping. However, the NRC noted that this
conclusion is not necessarily bounding and may not be applicable to all
sites because the assessment used generic seismic hazard curves and
representative major support fragilities in lieu of plant-specific
hazard curves and fragilities. This is further complicated by the
recent seismic hazard updates documented in the aforementioned SHSRs.
The assessment results documented in NUREG/KM-0017, ``Seismic Hazard
Evaluations for U.S. Nuclear Power Plants: Near-Term Task Force
Recommendation 2.1 Results,'' dated December 16, 2021, show that some
sites have experienced noticeable changes in both seismic exceedance
frequencies and Ground Motion Response Spectrum shapes relative to
those of the prior assessments. Taken together, these changes affect
seismic demand estimates used as an input to fragility analysis of key
component supports as well as the resulting unconditional failure
probability of indirectly induced piping failure. More details on this
effort can be found in the ``White Paper on Continued Applicability of
NUREG-1903.'' Because the risk associated with indirect piping failures
is plant-specific, the NRC would require in proposed 10 CFR
50.46a(c)(1)(i) that each entity perform an assessment of indirect
piping failures using the most up-to-date seismic hazard information.
This assessment would be part of the comprehensive risk assessment
required to implement 10 CFR 50.46a and graded approaches would be
possible depending on the significance of the associated risk. More
detailed guidance on this topic would be provided in DG-
[[Page 44625]]
1426, ``An Approach for a Risk-Informed Evaluation Process Supporting
Alternative Acceptance Criteria for Emergency Core Cooling Systems for
Light-Water Reactors,'' and DG-1428, ``Plant-Specific Applicability of
the Transition Break Size.''
Periodic inservice inspections are a key performance monitoring
strategy for verifying that analyses that predict component failure
remain accurate through the time the component is analyzed, and they
provide a method to identify novel degradation that may impact the
analysis and the structural integrity of the component. Reactor coolant
boundary piping in PWRs and BWRs is both ASME Class 1 piping and risk
significant and is typically inspected by a prescribed inservice
inspection program, a risk-informed inspection program, or an augmented
inspection program under 10 CFR 50.55a, ``Codes and standards.'' For
instance, in PWRs, the dissimilar metal welds that join the hot leg to
the reactor pressure vessel nozzle are inspected through ASME Code Case
N-770, ``Alternative Examination Requirements and Acceptance Standards
for Class 1 PWR Piping and Vessel Nozzle Butt Welds Fabricated With UNS
N06082 or UNS W86182 Weld Filler Material With or Without Application
of Listed Mitigation Activities,'' as incorporated by reference in 10
CFR 50.55a, which requires different inspection frequencies depending
on the type of mitigation employed. Also, the stainless-steel welds in
the same piping system are typically covered under Category R.1.20 in
ASME Code Case N-716-2, ``Alternative Classification and Examination
Requirements,'' as incorporated by reference in 10 CFR 50.55a through
RG 1.147, Revision 20, ``Inservice Inspection Code Case Acceptability,
ASME section XI, Division 1.'' However, the wording of the code case
does not require a minimum number of these welds to be inspected in the
overall program (see section 4(b)(2) of Code Case N-716-2). In
addition, there is a concerted effort within the ASME code community to
use risk arguments to reduce inspections in piping (thereby increasing
time between inspections) and other components, such as steam generator
shell welds, which may extend to Class 1 piping for cases with no
active degradation. This industry, EPRI, and ASME effort is addressed
in the 2023 white paper entitled, ``Draft White Paper: Statistical
Approach to Optimizing a Performance Monitoring Program.''
The analyses conducted within NUREG-1829 rely on the continuing
inservice inspection of the piping considered in predicting LOCA
frequencies. The inspections assumed were historical prescribed ASME
inspection procedures and frequencies (e.g., once in a 10-year
interval). The impact of increasing the time between inspections on the
predicted LOCA frequencies estimated in the elicitation is unknown,
therefore performance monitoring is needed to confirm the continued
adequacy of the analyses used to calculate the LOCA frequencies and
identify novel degradation that may challenge the component integrity.
To provide appropriate performance monitoring, proposed 10 CFR
50.46a(b)(3) would require that a sampling inspection program be
conducted on the welds in piping systems whose diameter is greater than
the TBS. Under the proposed rule, credit may be given for those welds
inspected as part of an established inspection program (e.g., these
welds could be included in the sample inspected in the risk-informed
piping inspection programs in lieu of other welds in the same risk-
informed category). The dissimilar metal welds in PWRs, which are
susceptible to primary water stress corrosion cracking, are inspected
periodically per Code Case N-770, while the similar circumferential
welds are part of a risk-informed program. Due to the number of similar
metal, circumferential, Class 1 welds in a PWR reactor coolant loop, or
those circumferential welds in a BWR that are classified as Category A
welds (i.e., welds of resistant material as defined in Generic Letter
88-01), it is possible that welds from the systems whose diameter is
greater than the TBS might not be included in the sample inspected as
part of the risk-informed inspection program. In addition, future ASME
code changes might decrease the number of these circumferential welds
inspected. Therefore, the proposed rule would require licensees to
inspect an NRC-approved risk-informed sample of these similar metal
circumferential welds in a PWR or the Category A circumferential welds
in a BWR in accordance with 10 CFR 50.55a with the highest failure
potential before implementation of proposed 10 CFR 50.46a and every
subsequent in-service inspection interval. This proposed requirement,
coupled with the ongoing inspection programs, would provide for the
appropriate amount of performance monitoring data over the course of
the plant's licensed life. Additional technical background supporting
this proposed requirement can be found in the ``White Paper on
Continued Applicability of NUREG-1829.''
The NRC also considered the possibility that currently licensed
AP1000 facilities; other currently certified designs listed in
appendices A through G of 10 CFR part 52 (i.e., the AP600, ESBWR,
APR1400, System 80+, U.S. Advanced Boiling Water Reactor, and NuScale
designs); and other future LWR plants could apply the proposed TBS. The
original elicitation effort did not consider these newer reactor
designs when developing the LOCA frequencies. The current validation
effort also did not consider the effects of design differences between
these newer plants and other currently operating PWRs on the TBS. These
newer plants may have different piping materials, configurations, and
operational and service conditions, among other factors, that may
impact the piping break frequencies and, thus, the TBS. Therefore, the
NRC decided that the proposed TBS would not be applicable to newer
plant designs, and these reactors should collectively be treated as
``new reactors'' as described in section XXXVI.F.(xii), ``Applicability
to New Reactor Designs,'' of this document.
(iv) Evaluation of the Plant-Specific Applicability of the Transition
Break Size
Because both the NUREG-1829 and NUREG-1903 studies developed
representative and not bounding estimates, and the recent validation
efforts in the ``White Paper on Continued Applicability of NUREG-1829''
and ``White Paper on Continued Applicability of NUREG-1903'' only
confirmed the continued applicability of these representative
estimates, unique plant attributes may result in plant-specific LOCA
frequencies that are greater than reported in either NUREG-1829 or
NUREG-1903. Consequently, proposed 10 CFR 50.46a(c)(1)(i) would require
entities applying to implement proposed 10 CFR 50.46a for plants
authorized to operate under 10 CFR part 50 on December 31, 2015, to
conduct an evaluation to demonstrate the applicability of the TBS as
defined in 10 CFR 50.46a(a)(9) to their individual plants. In addition,
proposed 10 CFR 50.46a(a)(9) and 10 CFR 50.46a(c)(1)(i) also would
allow for an alternate TBS to be proposed and justified. Similarly,
proposed 10 CFR 50.46a(c)(2) would require that entities applying to
implement proposed 10 CFR 50.46a for all other LWRs submit an analysis
demonstrating that the proposed reactor design is similar to the
designs of reactors authorized to operate under 10 CFR part 50 on
December 31, 2015. This
[[Page 44626]]
analysis should demonstrate that the NUREG-1829 and NUREG-1903 results,
which supported the development of the proposed TBS definition as
described in section XXXVI.F.(iii), ``Determining the Ongoing Validity
of the Transition Break Size,'' of this document, are generally
applicable to these plant designs. Applicants for all other plants
would also need to propose and justify an appropriate TBS.
Additionally, proposed 10 CFR 50.46a(c)(1)(i) for reactors
authorized to operate under 10 CFR part 50 on December 31, 2015, and
proposed 10 CFR 50.46a(c)(2) for all other LWRs would require
demonstration that the TBS as defined in proposed 10 CFR 50.46a(a)(9)
remain applicable after initial plant changes such that the TBS remains
valid. Proposed 10 CFR 50.46a(d)(4) would require demonstration that
subsequent plant changes enacted under this proposed rule also do not
invalidate the TBS. Most anticipated plant changes should not impact
the TBS, so little, if any, evaluation would be necessary to
demonstrate the acceptability of such changes. However, some changes,
such as power uprates, have the potential to affect the TBS by
increasing operating temperatures, coolant flow rate, and neutronic
flux.
Guidance for conducting the plant-specific evaluations to
demonstrate the initial applicability of the TBS and the subsequent
applicability after implementing changes under this proposed rule is
provided in DG-1428.
(v) Alternative ECCS Analysis Requirements and Acceptance Criteria
For breaks at or below the TBS, proposed 10 CFR 50.46a(e)(2)(ii)
would specify that acceptance criteria be satisfied to a high level of
probability (i.e., 95 percent probability level, as explained in RG
1.157, ``Best-Estimate Calculations of Emergency Core Cooling System
Performance''), which is currently required for all breaks under 10 CFR
50.46. Commensurate with the lower probability of breaks larger than
the proposed TBS, 10 CFR 50.46a(e)(3) of the proposed rule would
specify alternative ECCS analysis requirements in addition to the
current 10 CFR 50.46 for breaks larger than the proposed TBS.
Therefore, proposed 10 CFR 50.46a(e)(3)(ii) would require entities to
analyze ECCS cooling performance for breaks up to and including a
double-ended rupture of the largest pipe in the RCS using a relaxed set
of criteria compared to current LOCA requirements. These analyses would
need to be performed by methods acceptable to the NRC and would need to
demonstrate that ECCS cooling performance conforms to the acceptance
criteria set forth in the proposed rule.
Additionally, the proposed rule would modify the ECCS acceptance
criteria from the current 10 CFR 50.46 to be more performance-based and
would address the research findings in RIL-0801 for all breaks. The
proposed rule would establish two ECCS performance criteria in proposed
10 CFR 50.46a(e)(1) and fuel system requirements in proposed 10 CFR
50.46a(f).
(a) ECCS Performance Criteria
The SSCs of the ECCS are designed to provide residual heat removal
during and following a postulated LOCA. Failure of the ECCS to perform
its intended function would result in a loss of coolable geometry
followed by core reconfiguration. While the principal ECCS performance
requirements are simple in nature (i.e., remove residual heat and
maintain a coolable geometry), the system must be designed to achieve
specified performance objectives, taking into consideration all
degradation mechanisms and any unique performance features of the
particular fuel system that the ECCS is intended to cool. Sufficient
empirical data must be available for the particular fuel system to
enable the entity to identify all degradation mechanisms (e.g.,
embrittlement, loss of structural integrity) and any unique performance
features (e.g., eutectic or exothermic reactions, combustible gas
generation). Consideration of applicable degradation mechanisms would
in turn allow specification of limits for key parameters (e.g.,
cladding or fuel temperatures) that would establish the duration for
which the ECCS must remove residual heat and ensure that a coolable
geometry is maintained. In addition, fuel-specific analytical
requirements may be necessary to accurately or conservatively model
unique phenomena that impact the ECCS performance demonstration (e.g.,
fuel rod balloon and burst, cladding inside-diameter oxygen ingress).
Section 50.46a(e)(1) of the proposed rule would establish the
following principal ECCS performance requirements:
Sufficient coolant so that the fuel remains in a coolable
geometry during and following the LOCA heatup and quench.
Sufficient long-term cooling so that decay heat will be
removed for the extended period of time required by the long-lived
radioactivity remaining in the fuel.
Compliance with these performance requirements would provide
reasonable assurance that the overall objective of maintaining a
coolable fuel geometry during and after a LOCA. In addition, the
proposed rule would dictate specific analytical requirements for
demonstrating compliance with the ECCS performance requirements. For
instance, to demonstrate compliance with these system performance
requirements, ECCS performance would be evaluated using fuel-specific
performance objectives and associated analytical limits that take into
consideration all known degradation mechanisms and unique performance
features of the particular fuel system, along with an acceptable
evaluation model.
In previous comments, Framatome suggested changes to the draft
final 10 CFR 50.46c rule to make it more performance based. Those
changes included two ECCS performance criteria that, according to
Framatome, would support new fuel types. The NRC agreed that the
performance criteria suggested by Framatome would have benefits such as
making those performance requirements more performance based. The two
performance criteria in proposed 10 CFR 50.46a(e)(1) align closely with
Framatome's recommended performance criteria. The difference between
the proposed rule and Framatome's suggestion is also a change from the
draft final 10 CFR 50.46c rule (see section XXXV.C.(v)(c), ``10 CFR
50.46c Rulemaking and Cladding Embrittlement Research Findings,'' of
this document). In that draft rule, core temperature was the first ECCS
performance requirement. In this proposed rule, the NRC replaced core
temperature with fuel coolability to potentially allow for a
demonstration of coolability and safety of fuel outside of a fuel rod,
such as dispersed fine fuel fragments resulting from FFRD.
(b) Fuel Coolability
As explained in section XXXV.C.(v)(a), ``10 CFR 50.46 and Fuel
Dispersal,'' of this document, since 1973, the AEC and NRC position has
been that the objective of the coolable geometry criterion in 10 CFR
50.46(b)(4) is to maintain fuel pellets within the cladding and fuel
rods within the fuel bundle lattice. While fuel dispersal is not
explicitly addressed in the NRC's current regulations, the objective of
the coolability criterion indicates that significant fuel dispersal
during a LOCA is not permitted, as fuel leaving the fuel rod leads to
loss of the fuel rod structure and loss of fuel bundle configuration.
However, the state-of-knowledge
[[Page 44627]]
surrounding LOCAs and the ability to simulate complex phenomena during
a LOCA have drastically improved since 1973 and is expected to continue
to do so. Therefore, it is appropriate to update the understanding of
coolability. Based on the conclusions in NUREG/CR-7307, which describes
the NRC-sponsored fuel dispersal consequences PIRT exercise, it is
reasonable to conclude that some amount of fuel dispersal could be
coolable, and it is expected that this can be demonstrated with the
tools and experimental capabilities available today. The PIRT expert
panel believed that relatively simple bounding calculations can be
performed to address the primary downstream consequences of dispersal.
While core configuration, fuel bundle array geometry, and fuel rod
structure have traditionally been considered to be challenged by any
significant amount of fuel dispersal, the NRC is open to the
possibility that core configuration, fuel bundle array geometry, and
fuel rod structure could be shown to remain generally intact even if
some amount of fuel is dispersed. While simple bounding calculations
may be possible at the present time, the models and testing necessary
to perform a more detailed analysis and a more detailed demonstration
in support of reactor safety analyses have not been developed and
specific limits concerning how much dispersed fuel would be acceptable
have not been defined. Therefore, similar to any other LOCA phenomena,
prior to approving LOCA analysis methods involving fuel dispersal under
both the existing regulations and the proposed rule, the NRC would
review the relevant modeling approaches, including their underlying
validation and experimental basis, to ensure their credibility. Should
fuel dispersal be calculated to occur, an entity would need to provide
a technical basis addressing the coolability of both fuel retained in
the core and fuel dispersed into the reactor vessel or containment.
The proposed 10 CFR 50.46a provides an explicit risk-informed
framework for treatment of breaks above the TBS that could relax
required analytic assumptions. The proposed risk-informed framework is
not the only permissible way to use risk insights in addressing the
ECCS requirements. The NRC will continue to assess regulatory
compliance consistent with the existing regulatory framework on risk-
informed regulation, including regulatory guidance such as RG 1.174. An
acceptable level of defense-in-depth and safety margins should continue
to be maintained when applying risk insights. As a result of addressing
coolability in a more performance-based way and treating dispersal as
not being incompatible with coolability, other licensing pathways may
be made possible to address coolability, such as other alternatives
postulated in the regulatory basis for this rulemaking or proposed by
industry in topical reports.
This proposed rule would represent a change in the Commission's
position on coolability to allow for demonstrations of fuel dispersal.
This proposed change could result in questions on when the core
geometry would no longer be considered coolable. The proposed
requirements to demonstrate fuel coolability (in proposed 10 CFR
50.46a(e)(1)(i) and (f)(3)) and address cladding degradation phenomena
(in proposed 10 CFR 50.46a(f)(1)) in the ECCS performance analyses
would provide reasonable assurance that unacceptable scenarios would
not occur. Unacceptable scenarios would include, but would not be
limited to the following:
Widespread brittle cladding failure. Brittle failure of
the cladding during a LOCA could lead to the shattering of the cladding
and the release of much of the fuel in the rod. While fuel may be
dispersed out of a rupture opening of a fuel rod when considering FFRD,
less fuel is expected to be dispersed than if PQD is not preserved.
Additionally, the general fuel bundle configuration of the core would
be lost in the case of widespread brittle cladding failure, which would
have major deleterious effects on core coolability. For these reasons,
the NRC would maintain PCT and PQD criteria for zirconium-alloy and
UO2 fuel systems in 10 CFR 50.46 but include them in
guidance (i.e., DG-1263) to support proposed 10 CFR 50.46a. The PCT and
PQD criteria in DG-1263 would address proposed 10 CFR 50.46a(f)(1),
which states that cladding degradation phenomena must be addressed.
Fuel/cladding melt. In addition to ensuring PQD, emergency
coolant provided by the ECCS is intended to prevent the fuel from
melting during a LOCA. If the ECCS is unable to provide sufficient
cooling to prevent the fuel from melting, then this situation could
lead to an outcome where the fuel becomes uncoolable. While this
rulemaking proposes to recategorize LOCAs above the TBS as beyond-
design-basis, the NRC continues to find that avoidance of fuel melting
during a LOCA is appropriate to ensure a coolable geometry. Similarly,
the fuel dispersal PIRT identified fuel accumulation on the spacer
grids as a concern. If fuel particles settle on spacer grids and there
is two-sided heating of the cladding, then it could potentially cause
the cladding to melt and therefore damage the configuration of the
core. This postulated scenario must be avoided. For zirconium-alloy and
UO2 fuel systems that retain their fuel, the PCT criteria in
DG-1263, which would address the proposed 10 CFR 50.46a(f)(1)
requirement to have NRC-approved limits that address cladding
degradation phenomena, would prevent fuel melt during a LOCA.
(c) Acceptable Methodologies and Analysis Assumptions
Proposed 10 CFR 50.46a(e) would retain the requirement in existing
10 CFR 50.46 that ECCS cooling performance must be calculated in
accordance with an acceptable evaluation model. Acceptable evaluation
models are currently of two types: those that realistically describe
the behavior of the RCS during a LOCA, and those that conform with the
required and acceptable features specified in appendix K to 10 CFR part
50. Appendix K to 10 CFR part 50 evaluation models incorporate
conservatism as a means to justify that the acceptance criteria are
satisfied by an ECCS design. In contrast, the realistic or best-
estimate models attempt to accurately simulate the expected phenomena
while accounting for uncertainty such that there is a high level of
probability that the ECCS acceptance criteria will not be exceeded. As
a result, comparisons to applicable experimental data must be made and
uncertainty in the evaluation model and inputs must be identified and
assessed. All these existing requirements are included in proposed 10
CFR 50.46a(e)(2) for breaks at or below the TBS.
As currently required under 10 CFR 50.46, the ECCS analysis
performed with realistic modeling must demonstrate with a high level of
probability that the acceptance criteria will not be exceeded. The NRC
position in RG 1.157 is that 95 percent probability constitutes an
acceptably high probability. Section 50.46a(e)(2) of the proposed rule
retains this high level of probability as the statistical acceptance
criterion for breaks below the TBS. For breaks at or above the TBS,
proposed 10 CFR 50.46a(e)(3) would depart from the requirement of high
probability to state that assurance to at least a best-estimate level
is acceptable. For these breaks, best-estimate would refer to nominal
and unbiased analyses. Thus, for realistic evaluation models for breaks
above the TBS, entities would not be required to account for the
[[Page 44628]]
uncertainty. The best-estimate analyses could also take credit for
offsite power, as stated in proposed 10 CFR 50.46a(e)(3).
Proposed 10 CFR 50.46a(e)(2) and (e)(3) would each require a
separate determination of the most limiting break scenarios within each
of the two break-size regions: (1) breaks at or below the TBS and (2)
breaks larger than the TBS up to and including a double-ended rupture
of the largest pipe in the RCS. Different methodologies, analytical
assumptions, and acceptance criteria could be used for each break size
region. Consistent with current 10 CFR 50.46 requirements, entities
would be required to analyze breaks at or below the TBS by assuming the
worst single failure concurrent with a loss-of-offsite power and only
crediting the mitigative capability of safety-related SSCs. For breaks
larger than the TBS, entities could credit operation of both safety and
non-safety-related SSCs (subject to system availability as supported by
plant-specific data or analysis) provided that onsite power could be
readily provided to that equipment through manual actions after a loss-
of-offsite power (e.g., within approximately 30 minutes). This
requirement for non-safety-related equipment would be a defense-in-
depth consideration for severe accident management. The SSCs that are
credited for such accidents should have at least a pedigree similar to
that of the equipment credited in other beyond-design-basis accidents.
All non-safety equipment that is credited for analyses of breaks larger
than the TBS would need to be identified as such and evaluated whether
they should be listed in the plant technical specifications in order to
satisfy criterion 4 of 10 CFR 50.36(c)(2)(ii). Criterion 4 of 10 CFR
50.36(c)(2)(ii) states that a technical specification limiting
condition for operation must be established for a ``structure, system,
or component which operating experience or probabilistic risk
assessment has shown to be significant to public health and safety.''
Despite the recategorization of LOCAs greater than the TBS as beyond-
design-basis, the NRC still believes that it is significant to public
health and safety that such events are able to be mitigated. Criterion
4 of 10 CFR 50.36(c)(2)(ii) is the basis for the establishment of
technical specifications for equipment that are used in the mitigation
of another beyond-design-basis accident, anticipated transient without
scram. For example, in the BWR/6 standard technical specifications in
Volume 2 of Revision 5.0 of NUREG-1434, ``Standard Technical
Specifications--General Electric BWR/6 Plants,'' the typically non-
safety-related anticipated transient without scram recirculation pump
trip system is stated as satisfying criterion 4 of 10 CFR
50.36(c)(2)(ii).
(d) Fuel-Specific Performance and Analytical Requirements
Section 50.46a(f) of the proposed rule would include performance
requirements for fuel designs. The fuel designs would be required to
have NRC-approved limits that do the following:
Address cladding degradation phenomena.
Maintain fuel coolability.
Avoid explosive concentration of combustible gas.
Demonstrate that, after any calculated successful initial
operation of the ECCS, the ECCS must provide sufficient coolant to
remove decay heat and prevent further cladding failure for the extended
period of time required by the long-lived radioactivity remaining in
the fuel.
Means of meeting the requirement to address cladding degradation
criteria are provided in DG-1263 for fuel designs consisting of uranium
oxide or mixed uranium-plutonium oxide fuel pellets within cylindrical
zirconium-alloy cladding. Under this requirement, entities should
address the research findings discussed in section XXXV.C.(v)(c), ``10
CFR 50.46c Rulemaking and Cladding Embrittlement Research Findings,''
of this document when establishing criteria on PCT, PQD, and breakaway
oxidation, as detailed in DG-1263. The NRC is proposing to not include
PCT and PDQ requirements in the proposed 10 CFR 50.46a, unlike the
current 10 CFR 50.46(b)(1) and (b)(2). Instead, the NRC proposes to
include PCT and PQD criteria in guidance to enable regulatory
flexibility. For example, additional criteria for fuel performance
above 2200 [deg]F (1204 [deg]C) may be needed to address other high-
temperature failure mechanisms that may not occur below 2200 [deg]F
(1204 [deg]C). Including the PCT, the maximum local oxidation, and
maximum hydrogen generation criteria in DG-1263 rather than codified in
proposed 10 CFR 50.46a would provide this flexibility.
Exemptions from proposed 10 CFR 50.46a would not be needed for new
fuels to which the current criteria may not be applicable, such as non-
zirconium-alloy cladded fuel, since the criteria would be located in
guidance. The NRC did not include guidance for fuel designs that do not
consist of uranium oxide or mixed uranium-plutonium oxide fuel pellets
within cylindrical zirconium-alloy cladding as a part of this
rulemaking due to a lack of operational experience on which to base
specific criteria and to accelerate the rulemaking schedule. The NRC
has determined that these new fuels should be able to be licensed based
on technical justifications without the need for exemptions from
regulation. For example, since the detailed criteria would be in
guidance, justification could be provided for alternative criteria for
PCT that exceed the 2200 [deg]F (1204 [deg]C) limit in 10 CFR
50.46(b)(1) and has historically been used for fuel designs consisting
of uranium oxide or mixed uranium-plutonium oxide fuel pellets within
cylindrical zirconium alloy cladding.
As discussed in section XXXVI.F.(ii), ``Original Determination of
the Transition Break Size,'' of this document, some amount of fuel
dispersal could be shown to be coolable and therefore acceptable under
proposed 10 CFR 50.46a. During FFRD, fuel is dispersed following
ductile failure of the cladding (i.e., ballooning and burst) and also
can be dispersed from brittle failure of the cladding. Historically,
brittle failure of even a single rod has been prevented by ensuring
compliance with 10 CFR 50.46(b), as described in section XXXV.C.(v)(a),
``10 CFR 50.46 and Fuel Dispersal,'' of this document. Under this
proposed rule, the NRC may review applications for safety
demonstrations of FFRD during breaks above the TBS, so it is logical to
consider safety demonstrations of fuel dispersed from brittle failure
for LOCAs above the TBS. Such brittle failures should be limited so
that there is not a widespread loss of the general fuel rod bundle
configuration in the core, as widespread destruction of the fuel rod
bundle configuration would have significant deleterious effects on core
coolability.
There has been little research on the amount of fuel dispersal
expected to occur due to brittle failure, but if coolability and safety
can be ensured following brittle failure of the cladding, then the NRC
may find it to be acceptable for breaks above the TBS. Entities would
need to justify the safety case for permitting brittle failure of fuel
rods and the resulting fuel dispersal.
Additionally, the criteria in proposed 10 CFR 50.46a(f) may be used
as an alternative to the criteria in proposed 10 CFR 50.46(b), as
applicable, without having to adopt the rest of proposed 10 CFR 50.46a,
as stated in proposed 10 CFR 50.46(a)(3)(i).
The existing regulation in 10 CFR 50.46(b)(5) requires that, for
long-term cooling, the calculated core temperature
[[Page 44629]]
be maintained at an acceptably low value following any calculated
successful initial operation of the ECCS. It also requires that decay
heat be removed for the extended period of time required by the long-
lived radioactivity remaining in the fuel. Section 50.46a(f)(4) of the
proposed rule would retain these requirements from 10 CFR 50.46(b)(4)
and specify that cladding failure should be prevented in the extended
period of time required by the long-lived radioactivity remaining in
the fuel.
(e) Restriction of Reactor Operation
Proposed 10 CFR 50.46a(i) would allow the Director of the NRC's
Office of Nuclear Reactor Regulation to impose restrictions on reactor
operation if the NRC determines that the evaluations of ECCS cooling
performance are not consistent with the requirements for evaluation
models and analysis methods specified in proposed 10 CFR 50.46a(e)(1)
through (e)(3) and 10 CFR 50.46a(f). Non-compliance could be due to
factors such as lack of a sufficient database upon which to assess
model uncertainty, use of a model outside the range of an appropriate
data base, use of models inconsistent with the requirements of appendix
K to 10 CFR part 50, or discovery of phenomena unknown at the time of
approval of the methodology. Lack of compliance with methodological
requirements would not necessarily mean that the ECCS capability is
unacceptable, but only that the analysis results using the methodology
in question cannot be relied upon to demonstrate compliance with the
appropriate acceptance criteria. Thus, depending upon the specific
circumstances, it might be necessary for the NRC to impose restrictions
on operation until these issues are resolved. This requirement would be
consistent with the current ECCS regulations in existing 10 CFR
50.46(a)(2).
(vi) Risk-Informed Changes to the Facility, Technical Specifications,
or Procedures
The proposed 10 CFR 50.46a would designate LOCAs above the TBS as
beyond-design-basis, allowing for best-estimate analysis for such
LOCAs. The proposed flexibility in ECCS analyses for LOCAs above the
TBS would enable a wide range of changes that could be implemented at
facilities seeking to apply the proposed 10 CFR 50.46a.
Entities that would request approval to use 10 CFR 50.46a would use
a risk-informed evaluation to demonstrate that facility changes would
satisfy the risk-informed acceptance criteria in proposed 10 CFR
50.46a(h). Changes that would need to be evaluated would be specified
in proposed 10 CFR 50.46a(d)(3) and would include all ``enabled''
changes (i.e., changes that would be permissible if the NRC were to
approve the entity's request to use proposed 10 CFR 50.46a) that
satisfy the alternative ECCS analysis requirements in proposed 10 CFR
50.46a but do not satisfy the ECCS analysis requirements in current 10
CFR 50.46.
Proposed 10 CFR 50.46a(h)(1), (2) and (3) would require entities to
demonstrate in their risk-informed evaluations that increases in plant
risk (if any) would meet appropriate risk acceptance criteria, defense-
in-depth would be maintained, adequate safety margins would be
maintained, and adequate performance-measurement programs would be
implemented. All changes to a plant, its technical specifications, or
its procedures that would be based upon the analyses of ECCS
performance permitted under proposed 10 CFR 50.46a(e)(3)-except for
those changes made under proposed 10 CFR 50.46a(h)(1)-would need to be
reviewed and approved by the NRC. A wide range of changes could be
implemented under proposed 10 CFR 50.46a that, if improperly
implemented by entities, could result in significant adverse impacts on
public health and safety or common defense and security. NRC review and
approval would provide verification that an entity has properly
evaluated each proposed change against the acceptance criteria in
proposed 10 CFR 50.46a. Existing 10 CFR 50.36(b) requires each license
authorizing operation of a production or utilization facility of a type
described in 10 CFR 50.21, ``Class 104 licenses; for medical therapy
and research and development facilities,'' or 50.22, ``Class 103
licenses; for commercial and industrial facilities,'' to include
technical specifications, so changes to the technical specifications
require the license to be amended. However, 10 CFR 50.36 does not
provide risk thresholds, nor does it require a risk-informed evaluation
to demonstrate that the proposed change does not result in significant
adverse impacts on public health and safety or common defense and
security. Therefore, for technical specifications changes proposed
under 10 CFR 50.46a, 10 CFR 50.46a(h)(2) would provide the risk
acceptance criteria necessary for entities to evaluate each proposed
change, beyond that required in 10 CFR 50.36. Accordingly, the NRC's
proposed rule would require NRC review and approval of all changes
initiated under proposed 10 CFR 50.46a(h)(2).
An entity other than a design certification applicant or holder of
an ML who sought to make certain changes that would be enabled by the
proposed rule without prior NRC review and approval would need to
submit for NRC review its risk-informed process that would be used in
evaluating the acceptability of these changes as described in proposed
10 CFR 50.46a(c)(1)(v). The entity's process would also need to include
a means to evaluate the continued applicability of the TBS with the
acceptance criteria used in the evaluation described in proposed 10 CFR
50.46a(c)(1)(i) for plants authorized to operate under 10 CFR part 50
on December 31, 2015 or as described in proposed 10 CFR 50.46a(c)(2)
for entities other than those authorized to operate under 10 CFR part
50 on December 31, 2015. An entity who would make only a single or a
few changes enabled by the rule would not need to submit a risk-
informed evaluation process. Instead, that entity would submit only its
risk-informed evaluation of each change it has requested. Proposed 10
CFR 50.46a(h)(1) would contain acceptance criteria for self-made
changes enabled by the rule. Under the proposed framework, if an
entity's initial application to implement proposed 10 CFR 50.46a does
not include a risk-informed evaluation process, then that entity could,
at any later time, submit another license amendment requesting approval
of a risk-informed evaluation process. The DG-1426 would provide
guidance for the risk-informed evaluation specified in proposed 10 CFR
50.46a(c)(1)(iv) and the risk-informed evaluation process specified in
proposed 10 CFR 50.46a(c)(1)(v).
(a) Requirements for the Risk-Informed Evaluation
The acceptability of all entity-initiated changes made under the
rule would be judged in a risk-informed manner. The risk-informed
assessment process would include methods for evaluating compliance with
the risk criteria, defense-in-depth criteria, safety margin criteria,
and performance measurement criteria in proposed 10 CFR 50.46a(h).
These attributes have been identified by the NRC in RG 1.174 as a set
of risk evaluation tools to ensure that changes to the facility do not
endanger public health and safety.
Compliance with the risk criteria would play a key role in the
regulatory structure of the proposed rule. Entities would be required
to use a risk assessment to determine the change in
[[Page 44630]]
risk associated with facility changes. Inasmuch as the Commission's
final policy statement on the ``Use of Probabilistic Risk Assessment
Methods in Nuclear Regulatory Activities'' (60 FR 42622; August 16,
1995) sets forth the Commission's intention to encourage the use of PRA
and to expand the scope of PRA applications in all nuclear regulatory
matters to the extent supported by the state-of-the-art in PRA methods
and data, 10 CFR 50.46a(h)(4) of the proposed rule would require that a
technically acceptable PRA be used to demonstrate compliance with the
requirements of proposed 10 CFR 50.46a if the change being assessed
could substantively increase risk. Proposed 10 CFR 50.46a(h)(4)(i)
through (iv) set forth the four general attributes of an acceptable PRA
for the purposes of this proposed -rule. However, the NRC recognizes
that nonquantitative PRA assessment methodologies and approaches could
also be used to complement or supplement the quantitative aspects of a
PRA, especially when performance of a quantitative PRA methodology of
the level needed to support a particular plant modification decision
would not be justifiable because the safety significance of the
decision would not warrant the level of technical sophistication
inherent in a PRA. Accordingly, proposed 10 CFR 50.46a(h)(5) would
establish the minimum requirements for risk assessment methodologies
other than PRA. This proposed requirement would provide flexibility for
entities to use the nonPRA risk methodology (or combination of
different methodologies) when these methodologies produce results that
are sufficient to -determine that the risk acceptance criteria in the
proposed rule have been met.
(b) Aggregation of Plant Changes When Evaluating Changes in Risk
Entities often make changes to their facilities, technical
specifications, and procedures. Some changes that entities would be
able to make after being approved to use this proposed rule would not
have been permitted without the proposed 10 CFR 50.46a ECCS
requirements (i.e., enabled changes). Other changes would be unrelated
insofar as the basis of the changes and NRC approval, when necessary,
would rely on regulations, guidelines, or facility priorities that
would not depend on the proposed 10 CFR 50.46a ECCS requirements.
Unrelated changes would indirectly influence the change in risk of the
proposed 10 CFR 50.46a related changes because they would change the
risk profile of the facility. If unrelated changes were combined
(bundled) with enabled changes in evaluating the proposed 10 CFR 50.46a
change in risk estimates, the result would typically be different than
if the unrelated changes were considered as part of the baseline risk
associated with the current design and operation of the facility.
Regulatory guide 1.174 permits bundling changes (referred to as
combined changes in RG 1.174) and provides additional acceptance
guidelines when combining unrelated plant changes that might decrease
risk together with a group of enabled changes to evaluate the total
change in risk for comparison to the acceptance guidelines.
Allowing the bundling of unrelated changes into the proposed 10 CFR
50.46a change in risk estimates would encourage entities to use risk-
informed methods to take advantage of opportunities to reduce risk.
However, in some situations, bundling could mask the creation of
significant risk outliers. To ensure that outliers would not be
created, the proposed rule would not permit bundling of unrelated
changes with enabled changes without NRC review and approval.
Specifically, proposed 10 CFR 50.46a(h)(2)(i) and (iii) would allow
changes not enabled by proposed 10 CFR 50.46a to be bundled with
changes enabled by proposed 10 CFR 50.46a in the calculation of the
change in risk when an entity would submit an application for a change
under 10 CFR 50.90, ``Application for amendment of license,
construction permit, or early site permit.''
(c) NRC approval of an Entity Process for Making Changes to an Entity's
Facility or Procedures Without NRC Review and Approval
As a general matter, the proposed rule would require an entity to
obtain NRC review and approval through an application for any changes
to its facility, technical specifications, or procedures that may be
implemented under proposed 10 CFR 50.46a. However, the proposed rule
would allow an entity, other than a design certification applicant or
holder of an ML, to make a subset of plant and procedure changes
without NRC approval if the changes would involve minimal changes in
risk and no significant impact upon defense-in-depth capabilities,
safety margins, or performance monitoring. Prior NRC review and
approval of these changes on an individual basis would be unnecessary
if the NRC had previously concluded that the entity other than a design
certification applicant or holder of an ML had an adequate technical
process for appropriately identifying this subset of changes. Plant
changes that would involve minimal changes in risk and have no
significant impact upon defense-in-depth, safety margins, and
performance monitoring (and do not involve a change to the technical
specifications), would not result in significant issues involving
public health and safety or common defense and security.
Expending entity resources to prepare, and NRC resources to review
and approve, an application for approval of plant changes involving
minimal changes in risk would not be efficient uses of resources.
Rather, if the NRC would review and approve in advance the entity's
processes (including the acceptability of the entity's PRA and other
risk assessment methods) and criteria for identifying changes that
would have minimal impact on risk and would not significantly affect
defense-in-depth, safety margin, performance monitoring, or plant
physical security, then there would be no need for the NRC to review
and approve each of the individual changes. Accordingly, the NRC is
proposing an approach in 10 CFR 50.46a(h)(1) that would allow an entity
other than a design certification applicant or holder of an ML to
obtain ``pre-approval'' of a process for identifying minimal plant and
procedure changes made possible under proposed 10 CFR 50.46a.
Proposed 10 CFR 50.46a(h) would enable an entity other than a
design certification applicant or holder of an ML to make changes based
upon the provisions of proposed 10 CFR 50.46a, without prior NRC
approval, if the requirements in proposed 10 CFR 50.46a(h)(1) and
(h)(3) were met. The proposed rule also would require that the change
be permitted under the 10 CFR 50.59, ``Changes, tests and
experiments.'' Compliance with the 10 CFR 50.59 requirements would be
necessary to ensure that facility changes made without NRC approval
would not result in plant conditions that could impact public health
and safety. Compliance with the proposed 10 CFR 50.46a(h) requirements
for risk assessments would be required to ensure that facility changes
would result in acceptable changes in risk, that adequate defense-in-
depth would be maintained, that safety margins would be maintained, and
that adequate performance-measurement programs would be implemented.
Design certification applicants would not be subject to the change
process in proposed 10 CFR 50.46a(h)(1), either before or after NRC
certification of the design. An applicant for a design certification
that has not been approved
[[Page 44631]]
by the NRC would not need this provision because it could change the
design specified in its application before NRC issuance of a final
standard design certification rule for its design. The NRC also has
determined that design certification applicants whose designs have been
certified should not be allowed to change the certified designs without
NRC review and approval via rulemaking. Allowing the design
certification applicant to make changes to the certified design without
NRC approval through proposed 10 CFR 50.46a(h)(1) would be inconsistent
with the purpose of certifying the design in a rulemaking and would
effectively reduce the NRC's regulatory control over the design
certification. The NRC would review any deviations from the certified
design when reviewing the COL application. The NRC also would exclude
ML holders from this option to avoid a reduction of NRC regulatory
control over the approved manufacturing design, consistent with the
requirement in existing 10 CFR 52.171(b)(1) that holder of an ML may
not make changes to the design of the nuclear power reactor authorized
to be manufactured without NRC approval.
(d) Risk Acceptance Criteria for Plant Changes
To make changes to the facility, technical specifications, or
procedures not permitted by proposed 10 CFR 50.46a(h)(1), proposed 10
CFR 50.46a(h)(2) would require the submission of information from the
risk-informed evaluation demonstrating, among other things, that the
total increases in risk from the proposed change would be very small
and that the overall plant risk would remain small. These
characterizations of ``very small'' and ``small'' are intended to be
consistent with their use in RG 1.174, which introduces surrogate
guidelines that provide assurance that overall plant risks remain
bounded by the Commission's Policy Statement on ``Safety Goals for the
Operations of Nuclear Power Plants'' (51 FR 28044; August 4, 1986).
1. Risk Estimate
To satisfy the Commission's requirement in proposed 10 CFR
50.46a(h)(2)(ii) that the total increases in risk are very small, an
entity would need to evaluate the change in risk for each facility
change and show that the change meets the acceptance guidelines. If a
series of changes were made over time, proposed 10 CFR
50.46a(h)(2)(iii) would require that the cumulative effect of these
changes be evaluated and shown to meet the acceptance criteria.
Proposed 10 CFR 50.46a(h)(2)(iii) also would permit an entity to
combine changes in risk from facility changes not enabled by proposed
10 CFR 50.46a with changes in risk from facility changes that would be
enabled by proposed 10 CFR 50.46a for the purposes of meeting the
acceptance guidelines. Taken together, this bundled group of enabled
changes and unrelated changes would be referred to as the ``changes
made under'' proposed 10 CFR 50.46a.
For each change requiring a risk-informed evaluation, the total
change in risk from all facility changes made under the proposed 10 CFR
50.46a would need to be evaluated and compared to the ``very small''
acceptance criterion when the change is first made and then with each
subsequent enabled change that would result in a greater than minimal
increase in risk. Requiring that the total change in risk from all
facility changes made under the proposed 10 CFR 50.46a be compared to
the proposed 10 CFR 50.46a acceptance criteria instead of allowing the
changes in risk to be partitioned and individually compared to the
acceptance criteria would ensure that the total risk increase for all
changes, as they are implemented over time, would not constitute more
than a very small increase in risk. If the total increase in the
applicable risk metrics were not compared to the acceptance criteria,
then several changes in which each individual change's risk increase
was kept below the proposed rule's risk acceptance criteria could,
considered cumulatively, result in a significant increase in risk. A
significant increase may not satisfy the Commission's proposed
criterion that the overall plant risk remains small. Also, comparing
the risk increase from each change to the acceptance criteria
independently of all previous changes would render the use of the
``very small'' criterion inadequate to monitor and control increases in
risk from a series of plant changes implemented over time.
Comparing the total risk increase to the risk increase criterion,
and allowing bundling of unrelated changes in the change in risk
estimate, would support the NRC's position as stated in RG 1.174 that,
consistent with the key principles of risk-informed integrated
decision-making, entities should have a risk management approach in
which risk insights are not just used to systematically increase risk,
but also to help reduce risk where appropriate and where it is shown to
be cost effective.
2. Acceptance Criteria
The risk acceptance guidelines proposed in 10 CFR 50.46a(h)(2)(ii)
use CDF and large early release frequency (LERF) risk metrics. As
discussed in SRM-SECY-98-015, ``Staff Requirements--SECY-98-015--Final
General Regulatory Guide and Standard Review Plan for Risk-Informed
Regulation of Power Reactors,'' dated May 20, 1998, the Commission
approved the use of RG 1.174, which incorporated Commission direction
in the March 19, 1998, SRM issued for SECY-97-287, ``Final Regulatory
Guidance on Risk-Informed Regulation: Policy Issues,'' dated December
12, 1997. Based on proposals made by the staff, the Commission approved
the use of very small increases in CDF and LERF independent of the
baseline calculated CDF/LERF, provided that licensees track and NRC
staff monitor the cumulative effect of changes. These risk metrics are
based on subsidiary objectives derived from the NRC's safety goals and
their quantitative health objectives. In particular, the CDF risk
metric is used as a surrogate for the individual latent cancer fatality
risk, and the LERF risk metric is used as a surrogate for the
individual early fatality risk. The NRC has used CDF and LERF in making
regulatory decisions for more than 30 years. The NRC endorsed the use
of CDF and LERF as appropriate measures for evaluating risk and
ensuring safety in nuclear power plants when it adopted RG 1.174 in
1998. Since the adoption of RG 1.174, the NRC has had 27 years of
experience in applying risk-informed regulation to support a variety of
applications, including amending facility procedures and programs
(e.g., inservice testing and inservice inspection programs), amending
facility OLs, making changes to the FSAR, and implementing risk-
informed technical specifications. Based on this experience, the NRC
has determined that CDF and LERF are acceptable measures for evaluating
changes in risk as the result of changes to a facility, technical
specifications, and procedures, except for certain changes that affect
containment performance but do not affect CDF or LERF. Changes that
affect containment performance are considered as part of the defense-
in-depth evaluation.
In SRM-SECY-07-0082, ``Staff Requirements--SECY-07-0082--Rulemaking
to Make Risk-Informed Changes to Loss-of-Coolant Accident Technical
Requirements; 10 CFR 50.46a, `Alternative Acceptance Criteria for
Emergency Core Cooling Systems for Light-Water Nuclear Power Reactors,'
''
[[Page 44632]]
dated August 10, 2007, the Commission concluded that, to more closely
follow the approach presented in RG 1.174, the staff, in the 10 CFR
50.46a draft final rule, should restrict changes to a plant to very
small risk increases. As discussed in RG 1.174, a very small risk
increase is independent of a plant's overall risk as measured by the
current CDF and LERF. Increases in CDF of 10-6 per reactor
year or less and increases in LERF of 10-7 per reactor year
or less are very small risk increases for existing reactor facilities.
Limiting the acceptance criteria for plant changes enabled under
proposed 10 CFR 50.46a to very small risk increases ensures that
significant plant changes will not be enabled strictly due to the
initiating event frequency for a LOCA greater than the TBS, which is on
the order of 10-5 and the same as the criteria for small
risk increases. In SRM-SECY-12-0081, ``Staff Requirements--SECY-12-
0081--Risk-Informed Regulatory Framework for New Reactors'' dated
October 22, 2012, the Commission approved the staff's recommendation to
transition new reactors from large release frequency to LERF at or
before initial fuel load and discontinue regulatory use of large
release frequency. Applicants for new reactor OLs under 10 CFR part 50
or COLs under 10 CFR part 52 may need to transition to LERF to
demonstrate in the risk-informed evaluation that the proposed changes
meet the requirements in proposed 10 CFR 50.46a(h).
Since adopting RG 1.174 in 1998, the NRC has applied the
quantitative change in risk guidelines to individual plant changes and
to sequences of plant changes implemented over time. The NRC has found
these guidelines and the CDF and LERF values (when used together with
the defense-in-depth, safety monitoring, and performance measurement
criteria) to be capable of differentiating between changes, and
sequences of changes, that are not expected to endanger public health
and safety and those that might.
When the change does not significantly increase LOCA frequencies or
invalidate the evaluation demonstrating the applicability of the TBS to
the applicant's facility and the change is permitted under 10 CFR
50.59, 10 CFR 52.98(b), 10 CFR 52.98(c), and 10 CFR 52.98(d), as
applicable, proposed 10 CFR 50.46a(h)(1) would permit entities other
than a design certification applicant or holder of an ML to make
changes without prior NRC approval if the changes involve minimal
increases in risk that also have no significant impact upon defense-in-
depth capabilities, safety margins, and performance monitoring. A
minimal risk increase is one that, when considered qualitatively by
itself or in combination with all other minimal increases, would never
become significant. A minimal increase in risk is an increase less than
10 percent of the risk increases that would be very small for any
licensee. Therefore, a minimal increase is an increase of less than
10-7 per reactor year for CDF and an increase in LERF of
less than 10-8 per reactor year. These values are two orders
of magnitude below the maximum allowed risk increase guidelines in RG
1.174 and one order of magnitude less than the very small criterion.
Although multiple changes, when evaluated separately, could each be a
minimal increase in risk, when combined and evaluated together, they
could exceed the very small criterion. Most of these changes would have
a much smaller (and, in some cases, an unmeasurable) increase in risk.
If an entity other than a design certification applicant or holder of
an ML were to implement an unexpectedly large number of minimal risk
changes, then the periodic reporting requirements in proposed 10 CFR
50.46a(j)(3) would provide adequate notice to ensure that the NRC is
aware of potentially significant changes (or any collective impact), so
that the NRC could undertake additional oversight actions as deemed
necessary and appropriate.
(e) Defense-in-Depth
Section 50.46a(h)(3)(i) of the proposed rule would require that the
risk-informed evaluation demonstrate that defense-in-depth is
maintained. Defense-in-depth is an element of the NRC's safety
philosophy that employs successive measures to prevent accidents or
mitigate damage if a malfunction, accident, or naturally caused event
occurs at a nuclear facility. As conceived and implemented by the NRC,
defense-in-depth provides, among other things, redundancy in addition
to a multiple barrier approach against fission product releases.
Defense-in-depth continues to be an effective way to account for
uncertainties in equipment and human performance. The NRC has
determined that retention of adequate defense-in-depth must be ensured
in all risk-informed regulatory activities.
(f) Safety Margins
Proposed 10 CFR 50.46a(h)(3)(ii) would require that adequate safety
margins be retained to account for uncertainties. These uncertainties
include phenomenology, modeling, plant construction, and plant
operation. Without this proposed provision, entities could make plant
changes that would inappropriately reduce safety margins, resulting in
an unacceptable increase in risk or challenge to plant SSCs. This
proposed requirement would ensure that an adequate safety margin exists
to account for these uncertainties, such that there would be no
unacceptable results or consequences (e.g., structural failure) if an
acceptance criterion or limit is exceeded.
(g) Performance Measuring Programs
Section 50.46a(h)(3)(iii) of the proposed rule would require
entities to implement adequate performance-measurement programs and
feedback strategies to ensure that the risk-informed evaluation
continues to reflect actual plant design and operation. The risk-
informed evaluation would include the risk assessment, maintenance of
adequate defense-in-depth, and maintenance of adequate safety margins.
This proposed requirement would require that the monitoring programs be
designed to detect degradation of SSCs before plant safety is
compromised.
(vii) Leak Detection Requirements
In its SRM on SECY-07-0082, the Commission directed the NRC staff
to increase the defense-in-depth against large pipe breaks provided by
the 10 CFR 50.46a draft final rule. The SRM also directed the NRC staff
to evaluate various approaches for enhancing that draft final rule with
requirements for improved leak detection methods. The NRC determined
that adequate leak detection capability meeting this direction could
enhance defense-in-depth for LOCAs larger than the TBS by reducing the
likelihood of pipe breaks in the large break region. Thus, proposed 10
CFR 50.46a(d)(2) would require that licensees have leak detection
systems available at the facility and implement actions as necessary to
identify, monitor, and quantify leakage to ensure that adverse safety
consequences do not result from leaking primary pressure boundary
components that are larger than the TBS.
Because proposed 10 CFR 50.46a would not change the design basis of
piping and components that are smaller than the TBS, the requirements
of proposed 10 CFR 50.46a(d)(2) would apply only to piping and
components that are larger than TBS. The NRC recognizes that leakage
detection methods that satisfy these proposed requirements may not be
capable of determining whether the source of leakage is from piping or
a component that is larger or smaller than the TBS.
[[Page 44633]]
Discrimination between leaks in pipes larger or smaller than the TBS
would be unnecessary as long as adequate leak detection would be
provided for all beyond-TBS piping.
(viii) Programmatic Requirements
The proposed rule would include several specific programmatic
requirements in proposed 10 CFR 50.46a(d) that would apply to entities
who are approved to implement proposed 10 CFR 50.46a. These
requirements would remain applicable to such an entity as long as the
entity is subject to the proposed 10 CFR 50.46a alternative ECCS
requirements until such time as the licensee permanently ceases
operations by submitting the certifications required under 10 CFR
50.82(a). The proposed programmatic requirements would require entities
implementing proposed 10 CFR 50.46a to do the following:
(a) Maintain ECCS models and/or analysis methods that demonstrate
compliance with the ECCS acceptance criteria.
(b) Maintain adequate reactor coolant leak detection equipment
available at the facility and identify, monitor, and quantify leakage
to ensure that adverse safety consequences do not result from leaking
primary pressure boundary components that are larger than the TBS.
(c) Perform a risk-informed evaluation for each potentially risk-
significant change (or group of changes) to the facility enabled by
proposed 10 CFR 50.46a.
(d) Perform an evaluation to determine the effect of all planned
nuclear power plant changes and do not implement any facility change
that would invalidate the applicability of the TBS to the facility.
(e) Within 120 days after the outage in which the inspection
required in proposed 10 CFR 50.46a(b)(3) has been performed, submit to
the NRC a report that details the results of the inspection and any
impact these results have on the TBS in accordance with proposed 10 CFR
50.46a(b)(3).
The following discussion describes each of the programmatic
requirements.
(a) Maintain ECCS Models and/or Analysis Methods That Demonstrate
Compliance With the ECCS Acceptance Criteria
Section 50.46a(d)(1) of the proposed rule would require that
calculated results of entity ECCS models and/or analysis methods must
demonstrate compliance with the ECCS acceptance criteria as long as the
entity is subject to the requirements in proposed 10 CFR 50.46a.
Entities also would be required to update ECCS models and/or analysis
methods by modifying them as needed to address any error corrections
and plant design changes affecting ECCS performance during this time
period.
(b) Maintain Adequate Reactor Coolant Leak Detection Equipment
Available at the Facility and Identify, Monitor, and Quantify Leakage
To Ensure That Adverse Safety Consequences Do Not Result From Leaking
Primary Pressure Boundary Components That Are Larger Than the
Transition Break Size
The requirement for adequate leak detection capability would be in
proposed 10 CFR 50.46a(d)(2) and was discussed in section
XXXVI.F.(vii), ``Leak Detection Requirements,'' of this document.
Adequate leak detection would be required for all primary coolant
pressure boundary piping and other components whose rupture could
result in a break larger than the TBS.
(c) Perform a Risk-Informed Evaluation for Each Change (or Group of
Changes) to the Facility Enabled by Proposed 10 CFR 50.46a
In addition to meeting all other applicable requirements, entities
would be required by proposed 10 CFR 50.46a(d)(3) to perform a risk-
informed evaluation for changes enabled by proposed 10 CFR 50.46a. If
an entity had a change methodology that was submitted under proposed 10
CFR 50.46a(c)(1)(iv) and approved by the NRC, that licensee would be
able to make some changes without NRC approval as long as the
acceptance criteria in proposed 10 CFR 50.46a(h)(1) were met.
Otherwise, the entity would be required to submit the results of its
risk-informed evaluation for NRC review and approval. The entity would
need to retain the results of all risk-informed evaluations made under
proposed 10 CFR 50.46a(h)(1) and periodically submit a summary of the
results to the NRC as required under proposed 10 CFR 50.46a(j)(3).
(d) Perform an Evaluation To Determine the Effect of All Planned
Facility Changes and Do Not Implement Any Facility Change That Would
Invalidate the Applicability of the TBS to the Facility
For the TBS as defined in 10 CFR 50.46a(a)(9) to properly apply to
an entity's facility, that entity would be required to perform an
evaluation to demonstrate that the TBS is applicable to that particular
facility. For those entities that decide to propose an alternate TBS, a
similar evaluation would be needed as part of the basis for the
proposed TBS. But after the initial evaluation has demonstrated the
applicability of the TBS, an entity could make significant facility
changes that would invalidate the initial evaluation. Therefore, after
a facility has been approved to use proposed 10 CFR 50.46a, the
proposed rule would require the entity to ensure that the TBS remains
applicable to the facility by reviewing all subsequent plant changes to
ensure that the facility is not modified to the extent that the results
impact the applicability of the TBS. Licensees' existing configuration
management programs, which contain a process to control plant changes
made under 10 CFR 50.59, could be modified, through screening or
evaluation, to identify future plant changes that may invalidate the
applicability of the NRC's generic studies. Most anticipated plant
changes should not impact the TBS, so little, if any, evaluation would
be necessary to demonstrate the acceptability of such changes. However,
some changes, such as power uprates, would have the potential to affect
the TBS by increasing operating temperatures, coolant flow rate, and
neutronic flux.
(e) Submit to the NRC a Report Within 120 Days After the Outage When
the Inspection Occurred Detailing the Results of the Inspections and
Any Impact These Results Have on the TBS in Accordance With Proposed 10
CFR 50.46a(b)(3)
Under proposed 10 CFR 50.46a(b)(3), for RCPB piping whose inner
diameter is greater than the TBS, licensees would be required to
inspect an NRC-approved risk-informed sample of the similar metal
piping circumferential welds in a PWR and the Category A welds (as
defined in Generic Letter 88-01) in a BWR in accordance with 10 CFR
50.55a(g) before implementing proposed 10 CFR 50.46a and in every
subsequent inservice inspection interval (as defined in 10 CFR
50.55a(y)). Any indications found during this inspection should be
dispositioned according to ASME section XI requirements and the effects
of any indications on the continued applicability of the TBS should be
evaluated.
(ix) Reporting Requirements
(a) ECCS Reporting Requirements
The ECCS reporting requirements currently provided in 10 CFR
50.46(a)(3) were added during the 1988 revision to 10 CFR 50.46 (53 FR
35996; September 16, 1988). The proposed rule (52 FR 6334; March 3,
1987) preceding that final rule prompted several public
[[Page 44634]]
comments on the reporting requirements, some of which suggested that
the reporting requirements be relaxed or even eliminated (see NRC
Summary of Public Comments, ``PR-050--52FR06334--Emergency Core Cooling
Systems: Revisions to Acceptance Criteria,'' March 3, 1987).
In response to these comments, the NRC identified several reasons
for requiring reporting of changes to, or errors in, ECCS evaluation
models or in the applications thereof. First, the Commission believed
that significant changes or errors raise potential questions about the
adequacy of an evaluation model as a whole. Second, the Commission
noted that even minor or inconsequential errors and changes constituted
a deviation from what previously had been reviewed and accepted.
Finally, the Commission also noted that applications of models to areas
not contemplated during initial review of the model could result in
errors by extending a model beyond its intended range. The Commission
stated that, overall, the reporting requirements were a clarification
and relaxation of the then-existing requirements.
More than 20 years later, in its rulemaking efforts for 10 CFR
50.46c, the NRC again received numerous public comments about reporting
requirements, some of which suggested that the reporting requirements
should be relaxed or eliminated. In response to those public comments,
the NRC is proposing in the new 10 CFR 50.46a several relaxations to,
and clarifications of, the reporting requirements compared to what
currently appears in 10 CFR 50.46(a)(3). However, the NRC proposes to
leave the deterministic reporting requirements under the 1988 revision
to the ECCS regulations largely intact, with the addition of reporting
requirements for changes and error corrections affecting the integral
time-at-temperature (i.e., ECR) calculation.
Reporting remains necessary for changes to, or errors in, ECCS
evaluation models, and applications thereof, for several reasons.
First, considering all safety analyses for DBEs, LOCA analysis requires
substantially more complex thermal-hydraulic calculations. The
reporting requirements allow minor changes to, and error corrections
for, these calculations. Second, the reporting requirements provide a
regulatory framework for communicating, and addressing the effects of,
these changes and errors. And finally, reporting requirements are less
burdensome than an alternative that would include topical report
revisions and license amendments to address changes and errors. The NRC
considered these benefits, in addition to the original, safety-related
justification for the reporting requirements set forth in the 1988
rulemaking, when deciding to retain reporting requirements in proposed
10 CFR 50.46a.
The ECCS reporting criteria in proposed 10 CFR 50.46a(j) generally
match the criteria in the 10 CFR 50.46c draft final rule in SECY-16-
0033. One of the primary differences would be that an entity could
propose an alternative definition for a ``significant change'' for
breaks above the TBS. Furthermore, proposed 10 CFR 50.46a(j)(1)(iii)
would relax the reporting requirements for applicants for a standard
design certification (including applicants after the Commission has
adopted a final design certification regulation) or applicants for or
holders of a standard design approval under 10 CFR part 52 that are
approved to use proposed 10 CFR 50.46a. Under proposed 10 CFR
50.46a(j)(1)(iii), standard design approval holders and applicants and
standard design certification applicants that are approved to use the
alternative ECCS criteria in proposed 10 CFR 50.46a would be required
to internally document the nature and estimated effect of all changes
and errors affecting ECCS evaluation models. This documentation would
be subject to NRC inspection. Also, if the cumulative effect of changes
or errors would result in an inability to ensure compliance with the
alternative ECCS criteria in proposed 10 CFR 50.46a(f), then proposed
10 CFR 50.46a(j)(1)(iv) would continue to require NRC notification of
the underlying changes or errors and associated corrective actions so
that the NRC may evaluate the potential for impacts to the affected
standard design approval or standard design certification.
Instead of the existing requirement for the applicant for or holder
of a standard design approval or applicant for a standard design
certification to report this information to the NRC, the CP, OL, or COL
applicant would be responsible for providing an acceptable analysis of
the ECCS in its application submitted to the NRC under proposed 10 CFR
50.46a(j)(1)(i)-(iii). The proposed relaxation for applicants for or
holders of standard design approvals and applicants for standard design
certifications is justified because reporting changes and errors to the
NRC before the design is referenced in an application for a CP, OL, or
COL would not produce a tangible public health and safety benefit,
provided that the changes or errors would not create the potential for
the standard design approval or standard design certification to become
noncompliant with NRC requirements. There would be no public health and
safety benefit because the change or error would not impact the
operation of an operating reactor or even the NRC's safety review of an
application.
Moreover, instead of the existing 30-day reporting requirements
that are in 10 CFR 50.46(a)(3)(ii), the reporting requirements proposed
in 10 CFR 50.46a(j)(1)(ii), (j)(1)(iv), (j)(2)(ii), and (j)(2)(iii)
would require reports to be submitted to the NRC within 60 days of
discovery to better align with the 10 CFR 50.73, ``Licensee event
report system,'' report timeframe. As stated in 10 CFR 50.46(a)(3)(ii),
discovery of a change or error that exceeds the 10 CFR 50.46 acceptance
criteria would be a reportable event as described in paragraph (e) of
10 CFR 50.55, ``Conditions of construction permits, early site permits,
combined licenses, and manufacturing licenses''; 10 CFR 50.72,
``Immediate notification requirements for operating nuclear power
reactors''; and 10 CFR 50.73. Likewise, discovery of a change or error
that exceeds the criteria in proposed 10 CFR 50.46a would also be a
reportable event under 10 CFR 50.55(e), 10 CFR 50.72, and 10 CFR 50.73.
Since exceeding the proposed 10 CFR 50.46a criteria would be reportable
under those regulations, the NRC proposes to align the report required
under proposed 10 CFR 50.46a(j)(1) and (2) with the 60-day timeframe in
10 CFR 50.73. The additional time permitted relative to 10 CFR 50.46
would be based on the low initiating event frequency for many of the
limiting LOCAs, which would be confirmed on a plant-specific basis in
order to employ the proposed 10 CFR 50.46a. The additional time may
also result in increased accuracy in the information provided to the
NRC in the report required under proposed 10 CFR 50.46a.
The 10 CFR 50.46a proposed rule would clarify existing reporting
and corrective action requirements. Proposed 10 CFR 50.46a(j)(1) would
distinguish four possible combinations of reporting criteria based upon
predicted response, level of significance (i.e., significant or not
significant, as defined by the proposed rule), and whether the error,
change, or operation would result in any exceedance of acceptance
criteria. For each scenario, the proposed rule would provide the
required actions, reports, and a time frame for providing the necessary
reports.
Presently, the reporting requirements in 10 CFR 50.46(a)(3) require
that entities report changes to, or errors in,
[[Page 44635]]
an ECCS evaluation model, or in the application of the evaluation
model, and the estimated effect of the changes or errors on predicted
PCT. Proposed 10 CFR 50.46a would expand the definition of a
significant change or error to include integral time-at-temperature
(i.e., ECR). Any changes or errors that prolong the temperature
transient may further challenge the PQD analytical limits; however,
they may not significantly change the predicted PCT. As such, this
change or error would not be captured in the existing reporting
requirements. The NRC would include the reporting requirements for
changes and errors in integral time-at-temperature in proposed 10 CFR
50.46a to improve the content and communications of reports submitted
to the NRC. The NRC also proposes this requirement to inform the
agency's response to future changes to, or errors discovered in, ECCS
evaluation models or in the applications thereof.
A significant change or error would occur when the sum of the
absolute magnitudes of the respective changes is greater than 50 [deg]F
(10 [deg]C) PCT or 1.0 percent ECR for breaks at or below the TBS, as
defined in proposed 10 CFR 50.46a(k)(1). A change of 1.0 percent ECR is
correlated to a change in calculated ECR for a 50 [deg]F (10 [deg]C)
change in cladding temperature for a typical analysis of record PCT.
The definition of significant change in proposed 10 CFR 50.46a(k)(1)
would be specific to zirconium-alloy cladding. A new definition of
significant change or error may be necessary for other cladding
materials. In addition, proposed 10 CFR 50.46a(k)(1)(ii) would require
the use of maximum local oxidation (i.e., percent ECR) to evaluate the
impact of a change or error on the predicted integral time-at-
temperature. In proposed 10 CFR 50.46a(k)(2), for breaks above the TBS,
the significant change or error is one that results in a significant
reduction in the capability to meet the requirements of proposed 10 CFR
50.46a(e)(1) and (f). For LOCAs above the TBS, this significant change
or error is higher level than the criteria for LOCAs at or below the
TBS to allow for an entity to define alternative criteria for a
significant change. If alternative criteria are not defined, then the
same reporting criteria in proposed 10 CFR 50.46(k)(1) would be
applied.
Existing reporting requirements in 10 CFR 50.46(a)(3) with respect
to any ``change to or error discovered in an acceptable evaluation
model or in the application of such a model'' have been a source of
confusion. Two areas of common misconceptions are related to (1) the
baseline PCT and integral time-at-temperature values when estimating a
significant change or error (i.e., greater than 50 [deg]F (10 [deg]C)),
and (2) the 30-day reporting requirement including ``a proposed
schedule for providing a reanalysis or taking other action as may be
needed to show compliance with Sec. 50.46 requirements.'' Similar to
the 10 CFR 50.46c draft final rule, the NRC is proposing in this
rulemaking to revise the existing reporting requirements to (1)
identify more clearly the baseline values to be used in reporting
pursuant to the requirements of proposed 10 CFR 50.46a(j), (2) require
under proposed 10 CFR 50.46a(j)(1)(ii) and (iv) that entities include,
in a report describing a significant change, a proposed scope and
schedule for providing a reanalysis, and (3) distinguish between the
requirements for proposing a reanalysis scope and schedule, and for
proposing to implement corrective action as may be needed to show
compliance with proposed 10 CFR 50.46a(e) requirements.
As is the case with 10 CFR 50.46(a)(3), flexibility would exist in
terms of the scope of reanalysis that would be required to comply with
proposed 10 CFR 50.46a(j)(1)(ii) and (iv), and with the schedule that
the reporting entity may propose. Since the promulgation of the 1988
revision to 10 CFR 50.46, the NRC has accepted multiple evaluations,
which have been performed within a scope that has been significantly
limited when compared to full-scale implementation of an ECCS
evaluation model. In these cases, the NRC has concluded that such
evaluations satisfied the requirements of 10 CFR 50.46(a)(3)(ii). The
Commission does not propose to change its approach in this regard. In
addition, a proposed schedule for reanalysis may also incorporate
appropriate flexibility. The NRC has accepted proposed schedules that
extend as far as four years into the future and that are managed using
regulatory commitments that can be changed or updated as operating
conditions require and nuclear safety considerations permit. Proposed
10 CFR 50.46a(j)(1)(ii) and (iv) would retain this flexibility.
Proposed 10 CFR 50.46a(k)(1) would provide threshold values for PCT
and integral time-at-temperature for entities to use when estimating
the effect of a significant change or error. The baseline predictions
used to assess a significant change or error should be the PCT and
integral time-at-temperature values documented in a plant's updated
final safety analysis report (UFSAR). These values should represent the
latest LOCA analyses that were submitted and reviewed by the NRC as
part of a license amendment request (e.g., power uprate, fuel
transition) or as incorporated into the facility licensing basis in
accordance with NRC-approved methods, as amended by subsequent annual
reports.
Existing 10 CFR 50.46(a)(3) requires entities to include, in a
report describing the nature of a significant error or change and its
estimated effect on the predicted PCT, a proposed schedule for
providing a reanalysis or taking other action as may be needed to show
compliance with 10 CFR 50.46. This requirement has led to a
misconception that, when a significant error is reported that does not
cause the predicted PCT to exceed its 2200 [deg]F (1204 [deg]C)
acceptance criterion, a proposed schedule for providing a reanalysis is
not required and taking other action is not needed to show compliance
with the requirements. As explained in the preamble to the 1988
revision of 10 CFR 50.46, it has long been the NRC's position that
facility operation in excess of the 10 CFR 50.46 acceptance criteria is
an immediate safety concern that requires prompt corrective action and
the ``taking other action'' language in the rule was intended to
address that concern. This position is underscored by the final
sentence of the existing 10 CFR 50.46(a)(3)(iii): ``The affected
applicant or holder shall propose immediate steps to demonstrate
compliance or bring plant design into compliance with Sec. 50.46
requirements.'' Therefore, the reporting and reanalysis requirements
would be further separated into proposed 10 CFR 50.46a(j)(1) and 10 CFR
50.46a(j)(2) to distinguish the requirements that apply when a
significant change or error is identified that results in facility
operation in excess of the proposed 10 CFR 50.46a acceptance criteria,
from the requirements that apply when a significant change or error is
identified that does not result in facility operation in excess of the
proposed 10 CFR 50.46a acceptance criteria.
When a change to, or error in, an ECCS evaluation model, or in the
application of such a model, is discovered, the entity would be
responsible for estimating the magnitude of changes in predicted
results to (1) determine if immediate steps are necessary to
demonstrate compliance or bring plant design or operation into
compliance with proposed 10 CFR 50.46a(e) requirements, and (2)
identify reporting requirements. Under proposed 10 CFR 50.46a(j), an
entity's obligation to report and take corrective action would vary
depending upon whether the licensee's
[[Page 44636]]
situation falls into one of the following three possible scenarios:
1. Change or error that does not result in any predicted response
that exceeds any acceptance criteria and is itself not significant.
In this scenario, proposed 10 CFR 50.46a(j)(1)(i) would require the
entity to do the following:
a. Submit an annual report documenting the change(s) and/or
error(s), along with the estimated magnitudes of changes in predicted
results, and the basis for the entity's determination that the change
or error is not significant.
b. Revise the UFSAR in accordance with paragraph (e) of 10 CFR
50.71, ``Maintenance of records, making of reports.''
c. Use the revised UFSAR PCT/ECR predictions as a baseline for
future evaluations.
2. Change or error that does not result in any predicted response
that exceeds any acceptance criteria but is significant.
In this scenario, proposed 10 CFR 50.46a(j)(1)(ii) would require
the entity to do the following:
a. Within 60 days of making a change, discovering an error, or
both, submit a report documenting the change, error, or both, estimated
magnitudes of changes in predicted results, the proposed scope and
schedule for providing a reanalysis, and a description of and schedule
for implementing corrective actions. The reanalysis must be performed
within a scope of detail appropriate to address the reported,
significant change to, or error in, the ECCS evaluation model, or in
its application. The proposed reanalysis schedule should reflect
consideration for both the magnitude of the change or error and the
available margin to NRC acceptance criteria, once the estimated effect
of the change or error is applied to the existing results.
b. In accordance with the schedule proposed in paragraph a.,
provide the reanalysis to the NRC. This may be accomplished by
submitting a subsequent report, pursuant to proposed 10 CFR
50.46a(j)(1), describing the reanalysis and providing the updated
results. The reanalysis may be limited in scope but must otherwise be
performed using an acceptable evaluation model.
c. Revise the UFSAR to include new evaluation model results in
accordance with 10 CFR 50.71(e).
d. Use the revised UFSAR evaluation model results as a baseline for
the future evaluations.
3. Change or error that results in any predicted response that
exceeds any of the four acceptance criteria in proposed 10 CFR
50.46a(f).
In this scenario, proposed 10 CFR 50.46a(j)(1)(iv) and 10 CFR
50.46a(j)(2)(i) would require the entity to do the following:
a. Take immediate actions to bring the plant into compliance with
the acceptance criteria.
b. Within 60 days of making a change, discovering an error, or
both, submit a report documenting the change, error, or both, estimated
magnitudes of changes in predicted results, description of corrective
actions and/or compensatory measures, and the proposed scope and
schedule for providing a reanalysis, and a description of and schedule
for implementing corrective actions. The reanalysis must be performed
within a scope of detail appropriate to address the reported change to,
or error in, the ECCS evaluation model, or in its application. The
proposed reanalysis schedule should reflect consideration for both the
magnitude of the change or error and the fact that the change or error
has caused the NRC's acceptance criteria to be exceeded. The entity may
also need to submit this information in the reports required under 10
CFR 50.55(e), 10 CFR 50.72, and 10 CFR 50.73, as applicable.
c. Provide the reanalysis to the NRC in accordance with the
schedule submitted to the NRC. This may be accomplished by submitting a
subsequent report, pursuant to proposed 10 CFR 50.46a(j)(1), describing
the reanalysis and providing the updated results. The reanalysis must
be performed using an acceptable evaluation model. Revise the UFSAR to
include new evaluation model results in accordance with 10 CFR
50.71(e).
d. Use the revised UFSAR evaluation model results as the baseline
for future evaluations.
As described in scenario 3, proposed 10 CFR 50.46a(j)(1)(iv) and
(j)(2)(i) would apply to changes to, or errors in, ECCS evaluation
models, or the applications thereof, that affect the predicted
performance relative to all of the acceptance criteria contained in
proposed 10 CFR 50.46a(f). For example, an error or change in a PWR
boron precipitation calculation that invalidates the timing for
emergency operating procedures is considered a condition in which a
plant's conformance to proposed 10 CFR 50.46a(f)(4) would be uncertain.
In this circumstance, the NRC would consider this a potentially serious
safety issue in need of immediate attention and potential correction.
In proposed 10 CFR 50.46a(j)(2)(i), if a licensee with a COL identifies
a change or error that results in the acceptance criteria in proposed
10 CFR 50.46a(f) being exceeded before the Commission has made a
finding under paragraph (g) of 10 CFR 52.103, ``Operation under a
combined license,'' immediate action to bring a facility into
compliance would not be required due to the low safety significance of
exceeding the acceptance criteria before the initial startup of the
plant. Similarly, in proposed 10 CFR 50.46a(j)(2)(ii) and (iii), for
design certification applicants (including an applicant after the
Commission has adopted a final design certification regulation) and
applicants and holders for standard design approvals, if a change or
error correction results in calculated ECCS performance that does not
conform to the criteria in proposed 10 CFR 50.46a(e), immediate action
would not be required. In this situation, the affected applicant or
holder would need to propose appropriate steps to the Commission within
60 days to demonstrate compliance with proposed 10 CFR 50.46a
requirements, along with a report of the nature of the changes or
errors that resulted in an inability to assure compliance and an
estimate of their effect on the limiting transient.
Based upon the complexity of the ECCS performance demonstration,
entities may need to report multiple estimated effects on PCT and
integral time-at-temperature to assess available margin to analytical
limits under the proposed rule. For example, if the fuel rod population
in a reactor core is subdivided, analyzed, and judged against different
analytical limits (e.g., burnup-dependent fuel rod populations), the
entity may need to report multiple estimated effects and estimated
available margin to PCT and integral time-at-temperature analytical
limits. This would enhance communication with the NRC, especially for
understanding the proposed scope and schedule for reanalysis, if
required.
(b) Risk-Informed Evaluation Process Reporting
Section 50.46a(j)(3) of the proposed rule would require periodic
reports of changes that required a risk-informed evaluation under
proposed 10 CFR 50.46a(d)(3) and were implemented without prior NRC
approval under proposed 10 CFR 50.46a(h)(1). Proposed 10 CFR
50.46a(j)(3) would not require the submission of a report if no changes
involving minimal changes in risk were made under proposed 10 CFR
50.46a(h)(1) during the reporting period.
(c) TBS Applicability Requirements
Section 50.46a(j)(3) of the proposed rule would require an entity
to provide a brief summary of the basis for the
[[Page 44637]]
entity determination under proposed 10 CFR 50.46a(h)(1)(iii) that each
change made under proposed 10 CFR 50.46a(h)(1) would not invalidate the
TBS applicability evaluations made under proposed 10 CFR
50.46a(c)(1)(i) for an operating reactor authorized to operate under 10
CFR part 50 on December 31, 2015, or under proposed 10 CFR 50.46a(c)(2)
for entities other than those authorized to operate under 10 CFR part
50 on December 31, 2015.
Proposed 10 CFR 50.46a(j)(4) would include reporting requirements
for the periodic inspections used to provide assurance that no
significant degradation would be occurring in piping systems with an
inner diameter greater than the TBS that could undermine the technical
basis of the TBS. During each inservice inspection interval, the
licensee would be required to submit a summary report within 120 days
after completing the outage when the inspections specified in proposed
10 CFR 50.46a(b)(3) are completed. If any reportable indications were
found during these inspections, then the licensee's summary report
would need to include an evaluation of the impact of these indications
on the TBS as well as any other ASME section XI requirement. This
report could be combined with the summary report required by 10 CFR
50.55a(b)(2)(xxxii), which has an identical reporting time requirement.
The level of detail in the report specified in proposed 10 CFR
50.46a(j)(4) would be consistent with the report specified in 10 CFR
50.55a(b)(2)(xxxii).
(x) Documentation Requirements
Section 50.46a(l) of the proposed rule would require that entities
maintain records sufficient to demonstrate compliance with proposed 10
CFR 50.46a requirements. When making changes under proposed 10 CFR
50.46a(h), entities would be required to document the bases for
concluding that the acceptance criteria in proposed 10 CFR 50.46a(h)(1)
and (h)(2) would be satisfied and would continue to be satisfied as
long as the entity is subject to the proposed 10 CFR 50.46a. Entities
would be required under proposed part II of appendix K to 10 CFR part
50 to document the bases of evaluation models used to perform ECCS
calculations. Entities also would be required to document plant design
changes made under proposed 10 CFR 50.46a by updating the FSAR in
accordance with the requirements in 10 CFR 50.71(e). All documentation
could be reviewed during NRC inspections and/or audits.
(xi) Submittal and Review of Applications
(a) Initial Application for Implementing Proposed 10 CFR 50.46a
Requirements
When an entity would first apply to use the proposed 10 CFR 50.46a
requirements, that entity would need to submit an application under 10
CFR 50.34, 10 CFR 50.90, or 10 CFR part 52 for NRC review and approval.
The initial application would need to contain the information specified
in proposed 10 CFR 50.46a(c)(1)(i) through (vii), as applicable. This
would include information related to the applicability of the TBS to
the facility (if the entity desires to develop an alternate TBS, the
information needed in the application would include those items
discussed in section XXXVI.F.(xii), Applicability to New Reactor
Designs,'' of this document); information identifying the ECCS analysis
methods to be used; information describing the risk-informed evaluation
for all changes enabled by the proposed rule and proposed in the
application; information describing the proposed process for making
risk-informed changes without prior NRC approval (if the applicant
would seek approval of such a process); information describing non-
safety equipment to be credited for compliance with the ECCS acceptance
criteria in proposed 10 CFR 50.46a(e); and information describing how
the leak detection program would satisfy the criteria in proposed 10
CFR 50.46a(d)(2).
An entity's initial change from its existing ECCS analysis would
not need to be reviewed by the entity under the provisions of 10 CFR
50.59 because the proposed rule would require NRC review and approval
of the initial application to implement the proposed 10 CFR 50.46a
requirements. After the proposed 10 CFR 50.46a evaluation models and
initial ECCS LOCA analyses were established by approval of the proposed
10 CFR 50.46a application, subsequent changes to ECCS analyses would be
controlled by the process in 10 CFR 50.59 (which provides criteria for
determining which changes are within the licensee's authority to make)
and the requirements in proposed 10 CFR 50.46a(j) for reporting changes
to evaluation models and analysis methods (whether from correction of
errors or changes). The initial application could request one or more
facility changes.
The initial application also would include a request for NRC
approval of a process for evaluating the acceptability of future
facility changes enabled by proposed 10 CFR 50.46a using the provisions
in proposed 10 CFR 50.46a(h)(1). If approval of a process for
evaluating future changes were requested, the application would need to
include the information described in proposed 10 CFR 50.46a(c)(1)(iv).
(b) Subsequent Applications for Changes Under Proposed 10 CFR 50.46a
After NRC approval of an entity's initial application addressing
ECCS analyses and the risk-informed evaluation processes under the
proposed rule, the entity could submit applications for proposed
changes under 10 CFR 50.90. These license amendment applications would
need to contain the following:
For licensees, the information required by 10 CFR 50.90;
Information from the risk-informed evaluation
demonstrating that the risk criteria, defense-in-depth criteria, safety
margins, and performance monitoring criteria in proposed 10 CFR
50.46a(h)(2) and (h)(3) would be met;
Information demonstrating that the ECCS acceptance
criteria in proposed 10 CFR 50.46a(e)(1) would be met; and
Information demonstrating that the proposed change would
not increase the LOCA frequency of the facility by an amount that would
invalidate the applicability of the TBS to the facility.
After reviewing the application with the proposed change, the NRC
could approve the change if it complies with the criteria in proposed
10 CFR 50.46a(h)(2) and (h)(3) and all other applicable NRC
regulations, including the current requirements for plant physical
security. In addition, the NRC would evaluate potential impacts of the
proposed change on facility security to ensure that the change would
not significantly reduce the ``built-in capability'' of the plant to
resist security threats, thus ensuring that the change would not be
inimical to the common defense and security and would provide adequate
protection to public health and safety.
Entities who would not submit a request for NRC approval of a
process for evaluating the acceptability of future changes enabled by
proposed 10 CFR 50.46a using the provisions in proposed 10 CFR
50.46a(h)(1) could make such a request at any time by submitting the
application containing the information described in proposed 10 CFR
50.46a(c)(1)(iv).
(xii) Applicability to New Reactor Designs
As discussed in sections XXXVI.F.(ii), ``Original Determination of
the Transition Break Size,'' and (iii), ``Determining the Ongoing
Validity of the Transition Break Size,'' of this
[[Page 44638]]
document, the TBS was established in NUREG-1829, which was based on the
commercial LWR designs authorized to operate under 10 CFR part 50 on
December 31, 2015. LWRs authorized to operate under 10 CFR part 52 or
after that date may have different piping materials, configurations,
and operational and service conditions, among other factors, that may
impact the piping break frequencies and thus the TBS. New LWR reactor
designs (i.e., the class of commercial LWR reactors authorized to
operate under 10 CFR part 50 after December 31, 2015, or under 10 CFR
part 52) may have similar piping materials, service conditions and
operational programs, piping designs, and mitigation and control of
age-related degradation programs as those found in currently operating
plants. There are several new LWR designs for which the NRC expects
that the frequency of LOCAs above the TBS could be as low as it is at
current LWRs. Thus, applicants can apply the proposed 10 CFR 50.46a
requirements to these new reactor designs with adequate justification.
New large LWR or small modular reactor entities under 10 CFR part 50 or
part 52 who wish to apply proposed 10 CFR 50.46a would have to submit
an analysis for NRC approval, as specified in proposed 10 CFR
50.46a(c)(2), demonstrating why it would be appropriate to apply the
alternative ECCS requirements and what the appropriate TBS would be in
order for the new design to meet the proposed 10 CFR 50.46a rule.
In its analysis, the entity would be required to demonstrate that
the proposed reactor facility is similar to reactors authorized to
operate under 10 CFR part 50 on December 31, 2015. In addressing
similarity of the proposed design to reactors authorized to operate
under 10 CFR part 50 on December 31, 2015, the entity should address
design, construction and fabrication, and operational factors that
include, but are not limited to:
The similarity of the piping materials of construction and
construction techniques for new reactors to those in the operating
fleet authorized to operate under 10 CFR part 50 on December 31, 2015;
The similarity of service conditions and operational
programs (e.g., in-service inspection and testing, leak detection, QA,
etc.) for new reactors to those for the operating fleet authorized to
operate under 10 CFR part 50 on December 31, 2015;
The similarity of piping design (e.g., pipe sizes and pipe
configuration) for new reactors to those found in the operating fleet
authorized to operate under 10 CFR part 50 on December 31, 2015;
Adherence to existing regulatory requirements, regulatory
guidance, and industry programs related to mitigation and control of
age-related degradation (e.g., aging management, fatigue monitoring,
water chemistry, stress corrosion cracking mitigation, etc.); and
Any plant-specific attributes that may increase LOCA
frequencies compared to those used to support the development of the
proposed TBS.
Proposed 10 CFR 50.46a(c)(2) would also require that the analysis
include a recommendation for an appropriate TBS and a justification
that the proposed TBS is consistent with the technical basis for the
proposed 10 CFR 50.46a (i.e., the TBS would include sufficient margin
to provide assurance that, when considering the limited availability of
operating experience data and the uncertainty in the estimation of LOCA
frequency, the estimated frequency of breaks larger than the TBS for
all initiators would not exceed 10-5 per year). For those
new reactor (large LWR and small modular reactor) designs that employ
design features that effectively increase the break size via opening of
specially designed valves to rapidly depressurize the RCS during any
size LOCA, justification of the acceptability of a TBS would also be
necessary. The justification should consider the integral impacts of
fluid conditions such as flow quality at the discharge node(s), which
can substantially affect the temperature transient experienced by the
fuel. The methodology used to determine the proposed TBS should be
described in the justification.
(xiii) Changes to 10 CFR 50.46
The NRC is proposing several changes to 10 CFR 50.46. Proposed 10
CFR 50.46(a) would include a pointer to proposed 10 CFR 50.46a as a
voluntary alternative. Proposed 10 CFR 50.46(a)(3)(i) would be revised
to permit use of the ECCS criteria in proposed 10 CFR 50.46a(f) instead
of 10 CFR 50.46(b). To align with these changes, the applicable ECCS
reporting requirements from proposed 10 CFR 50.46a would be added as
proposed 10 CFR 50.46(a)(iv) and 50.46(a)(v) for entities that are
approved to use proposed 10 CFR 50.46a(f). These changes would not
invalidate current exemptions from 10 CFR 50.46 and could reduce the
number of exemptions for non-zircaloy and non-ZIRLO claddings due to
the allowance to use the fuel technology-neutral proposed 10 CFR
50.46a(f) criteria in place of the prescriptive 10 CFR 50.46(b)
criteria.
(xiv) Discussion of Public Comments on the Fuel Dispersal Aspects of
the Regulatory Basis
In the public comments on the regulatory basis, representatives of
the nuclear power industry did not present a unanimous recommendation
among the alternatives provided by the NRC. Alternative 5 would have
had the NRC pursue rulemaking to modify 10 CFR 50.46 to allow for
insights from piping fracture mechanics to be used to disposition
large-break LOCAs and, thus, fuel dispersal during large-break LOCAs.
NEI and Westinghouse supported a modified version of alternative 5 that
would have aligned with EPRI's ``Alternate Licensing Strategy.'' The
EPRI Alternate Licensing Strategy proposed to utilize piping fracture
mechanics to show that leaks in large pipes can be detected and
operator action taken with sufficient probability before the pipe
breaks, such that the larger LOCAs that could challenge fuel integrity
would not occur. EPRI submitted its Alternate Licensing Strategy to the
NRC via topical reports in April 2024.
NEI also supported, along with the BWR Owners' Group, alternative
4, a rulemaking that would provide a generic bounding assessment of
dose and use risk insights for post-FFRD consequences. The former
alternative was not viewed by NEI and the BWR Owners' Group as a
solution for BWRs since it would utilize the leak-before-break (LBB)
concept and LBB has only ever been approved for use in PWRs.
Historically, the NRC has not allowed LBB to be applied to ECCS design,
containment design, or equipment qualification, as described in the
preamble to a 1987 final rule amending GDC 4, ``Environmental and
dynamic effects design bases'' (52 FR 41288; October 27, 1987). The PWR
Owners' Group and Framatome expressed support for an approach that
would employ integrated decision-making as done in the disposition of
in-vessel downstream effects associated with generic safety issue
(GSI)-191.
Additionally, there was support from NEI, the BWR Owners' Group,
and Westinghouse for alternative 2, a rulemaking to recategorize large-
break LOCAs as beyond-design-basis accidents, though they recommended
that a rulemaking with an updated 10 CFR 50.46c be pursued separately
from this increased enrichment rulemaking due to the perceived schedule
impact. Aspects of Framatome's response also aligned with this
alternative, such as their statement in their proposed path forward
that ``the amount of dispersal
[[Page 44639]]
and coolability evaluation in a [large-break LOCA] would be based on
fully best-estimate models (i.e., conditions consistent with expected,
nominal operating conditions without biases or uncertainties).''
Moreover, in response to question 2 posed to the public on fuel
dispersal in the regulatory basis Federal Register notice, NEI,
Westinghouse, and Framatome stated that they believed that with a true
best-estimate analysis for large-break LOCAs, it would be reasonable to
demonstrate that no high burnup rods rupture and, therefore, disperse
fuel, especially when combined with the anticipated benefits from ATF.
Industry recommendations stated schedule as a large reason for
their support of particular alternatives. The regulatory basis
attempted to qualitatively compare the rulemaking schedule impacts of
each alternative. After further consideration of these alternatives,
the NRC's accuracy of the estimated impacts has improved and the
qualitative schedule estimates presented in the regulatory basis have
been updated significantly. The NRC now considers that alternative 2
would take less time than alternatives 4 and 5 because the previous
work on the draft 10 CFR 50.46a rulemaking could be leveraged. For
example, the technical basis and rule language were already developed
by the NRC, so most of the work would be in confirming that the
technical basis is still applicable, modernizing the rule, and
developing guidance. In addition, the previous work performed on the
previous 10 CFR 50.46c rulemaking could be leveraged to develop a
performance-based rule with guidance that would facilitate future
efforts to pursue other licensing pathways. As such, the risk-informed
approach in alternative 2 and the performance-based approach in the
previous 10 CFR 50.46c rule have been adapted into this proposed rule.
The Union of Concerned Scientists and two members of the public
stated that they did not support any approach that allows for fuel
dispersal. Another member of the public recommended that the NRC wait
until more research and analysis is performed for fuel dispersal. The
NRC considered all these comments, among other factors, as discussed in
section XXXVI.F.(i), ``Overview,'' of this document, when deciding to
pursue a performance-based alternative 2.
XXXVII. Specific Questions
The NRC is seeking advice and recommendations from the public on
the proposed rule, draft guidance documents, and a draft regulatory
analysis. The NRC is particularly interested in comments and supporting
rationale from the public on the following:
Expedited Construction of Certain Structures, Systems, and Components
Question 1: Should the NRC consider granting one or more general
licenses to begin construction activities as part of an existing or
future standard design certification rule? If the NRC were to issue
such a general license, should the NRC require that the general license
holder reference another plant that has completed construction
(resulting in the general license excluding first of a kind
applicants)? If not, what conditions should the NRC include in the
general license to ensure that the construction activities authorized
by the general license can be completed safely? Please provide a basis
for your response.
Question 2: Would it be useful for the NRC to issue regulations to
provide for general licenses related to production and utilization
facilities for other activities described under section 109 of the AEA,
with the exception of import, export, construction, or operation?
Please provide a basis for your response.
Question 3: Does the proposed regulatory change achieve its
intended objective of allowing applicants to make business decisions on
the appropriate balance between expedited construction and the
regulatory risk that changes will be necessary to address safety
issues? Are there other approaches that can both provide the needed
flexibility while better allowing applicants to structure the degree of
regulatory risk they take on through interactions with the NRC staff
(e.g., hold-points during construction for NRC inspection)?
Risk-Informing 10 CFR 50.59 and Allowing Flexibility for Changes to
Methods
In accordance with Commission direction in the staff requirements
memorandum (SRM), dated July 21, 1993, on SECY-93-087, ``Policy,
Technical, and Licensing Issues Pertaining to Evolutionary and Advanced
Light-Water Reactor (ALWR) Designs,'' dated April 2, 1993, the NRC
staff considers common cause failures (CCFs) in digital instrumentation
and controls (DI&C) systems to be beyond design-basis events (BDBEs).
In addition, SECY-93-087 states that common mode failures could defeat
the redundancy achieved by the hardware architectural structure, and
could result in the loss of more than one echelon of defense-in-depth
provided by the monitoring, control, reactor protection, and engineered
safety functions performed by the DI&C systems. Therefore, a CCF could
result in ``a malfunction of an SSC important to safety with a
different result than any previously evaluated in the final safety
analysis report'' (see 10 CFR 50.59 (c)(2)(vi)), and a licensee would
have to obtain a license amendment pursuant to 10 CFR 50.90 prior to
implementing the proposed DI&C modification.
Question 4: In light of these considerations, do the proposed
revisions to 10 CFR 50.59 adequately address issues that may arise from
facilities that have transitioned to DI&C? If not, please provide
specific comments on changes that would be necessary to address these
issues.
Updates to Construction Permit Requirements and Related Licenses
Question 5: Are there any additional requirements in 10 CFR 50.34
that the NRC should consider changing to be more technology-inclusive
or provide additional clarity? Should conforming changes to 10 CFR
50.67 be made to be consistent with 10 CFR 50.34 wording updates?
Please provide a basis for your response.
Question 6: Are there any requirements in 10 CFR 50.34 that the NRC
should consider changing that are posing undue barriers to timely and
efficient submittal of CP applications, or which are unnecessary at the
CP stage? Please provide a basis for your response.
Establishing Thresholds for Changes to Reactor Designs During
Construction and Operation Under 10 CFR Part 52
Question 7: The proposed revisions to the definitions in each of
the appendices in 10 CFR part 52 identify the specific sections in the
generic DCD that contain the various categories of Tier 1 information
(i.e., definitions and general provisions, design descriptions, ITAAC,
significant site parameters, and significant interface requirements).
The proposed revisions to section VIII in each of the appendices in 10
CFR part 52 describe the change control process associated with each
category of Tier 1 information. Will there be adequate regulatory
clarity regarding the category of Tier 1 information in the generic DCD
such that the change control process can be readily identified? Please
provide a basis for your response.
Question 8: The proposed revisions to section VIII.B of appendix D
to 10 CFR part 52 would effectively treat Tier 2* information for the
AP1000 design as Tier 2 information during construction as well as
operation. However, for appendices A (ABWR) and E (ESBWR) Tier 2*
information would not be treated as Tier 2 information until after
[[Page 44640]]
the plant has achieved full power. The basis for this difference is the
construction and licensing experience gained from the COLs referencing
the AP1000 design which is lacking for the ABWR and ESBWR designs.
Should the flexibility provided to the AP1000 design in appendix D
during construction also be extended to the ABWR and ESBWR designs in
appendices A and E? Please provide a basis for your response.
Question 9: Paragraph (b) of 10 CFR 53.1535 states ``The holder of
a COL under this part for which the NRC has not yet made a finding in
accordance with 10 CFR 53.1452(g) must request amendments required by
10 CFR 53.1525 or 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.'' This differs from provisions in 10
CFR part 52 by adopting requirements that explicitly support a change
process described in RG 1.237, ``Guidance for Changes During
Construction for New Nuclear Plants Being Constructed Under a Combined
License Referencing a Certified Design Under 10 CFR part 52.'' Should
the provisions in 10 CFR part 52 be revised to also explicitly support
this change process? Please provide a basis for your response. Question
10: The proposed revisions to 10 CFR 52.98 and 53.1550 allow holders of
an OL or COL that reference a manufacturing license (ML) to make
changes to the FSAR without obtaining a license amendment if the
changes are identical to changes approved by the Commission by
amendment to the ML for the manufactured reactor and upon determining
that implementation of the changes will be consistent with the basis
for the Commission's approval of the amendment to the ML and not
involve any additional changes that would require an amendment to its
OL or COL. Should this flexibility be provided to COLs that reference a
design certification? Please provide a basis for your response.
Question 11: Should the proposed rule changes to 10 CFR 50.59
described in section XII of this document also be applied to the 10 CFR
50.59-like process in the appendices of 10 CFR part 52? Please provide
a basis for your response.
Revision of the Emergency Preparedness Regulations for Nuclear Power
Reactors
Conduct of Exercises
The NRC is considering risk-informed revisions to the conduct of
exercises in section F, ``Training,'' of appendix E to 10 CFR part 50.
Paragraph IV.F.2.b of appendix E to 10 CFR part 50 requires a biennial
exercise of the onsite plan to ensure that emergency response
organization (ERO) proficiency in key skills is maintained. The NRC is
assessing risk-informed revisions to the conduct of exercises to ensure
ERO proficiency is maintained through the appropriate frequency of
exercises and performance of the exercise cycle as required by
paragraph IV.F.2.j of appendix E to 10 CFR part 50. Licensees conduct
drills and exercises to develop and maintain key skills such as timely
and accurate classification and notification of events; assessment of
radiological releases and development of protective actions; planning,
analysis, and implementation of mitigative actions; worker protection
during emergency conditions; and coordination with offsite response
organizations to include dissemination of information to the public
through media channels. Following each exercise, licensees perform a
critique of their actions to comply with 10 CFR 50.47(b)(14) which
states, in part, ``deficiencies identified as a result of exercises or
drills are (will be) corrected.'' In 1996, the NRC revised the
emergency plan regulations for exercises from annual to biennial (61 FR
30129; June 14, 1996) to allow greater flexibility in licensee
emergency planning training activities while maintaining readiness. The
NRC is considering another revision to the frequency to either a
triennial, or quadrennial basis. For example, a quadrennial exercise
cycle could be supplemented with additional drills to ensure that
proficiency is maintained between exercise years. Additionally, the NRC
is considering additional language in paragraph IV.F.2.c of appendix E
to 10 CFR part 50 to allow for the use of tabletop exercises or other
performance enhancing methods when offsite authorities are not
participating in the exercise of the onsite emergency plan required by
paragraph IV.F.2.b. These potential changes could increase scheduling
flexibility and reduce the resource burden while maintaining reasonable
assurance that ERO proficiency will be maintained. If the exercise
frequency is changed to quadrennial for some sites, the NRC is
considering a corresponding change to the exercise cycle from 8 to 16
calendar years in paragraph IV.F.2.j.(iii) of appendix E to 10 CFR part
50 to reflect the proposed change to quadrennial exercises. However, a
change to the exercise cycle may impact the ability of the ERO to
demonstrate proficiency in the key skills necessary to respond to
various emergency events, particularly in sites with increased turnover
in ERO members. During the 1996 change from annual to biennial, the NRC
received several comments about the ability of licensees to maintain
proficiency with a longer gap between exercises and how this increased
gap could also affect the performance of State and local responders.
These comments were ultimately resolved and the NRC determined that
reasonable assurance remained with the change in exercise frequency.
With an additional three decades of experience, advances in all-hazards
emergency preparedness, and a range of potential new advanced reactor
designs in the future, the NRC is again interested in comments on the
conduct of exercises. The NRC urges stakeholders to provide comments on
the following questions:
Question 12: What are the potential benefits and challenges of a
change to the conduct of exercises to applicants and licensees and
offsite response organizations? Please provide a basis for your
response.
Question 13: What specific criteria related to maintaining onsite
ERO proficiency should be considered in evaluating whether a biennial,
triennial, or quadrennial exercise is appropriate? Please provide a
basis for your response.
Question 14: How would a potential change impact the conduct of
exercises of the offsite plans required under paragraph IV.F.2.c of
appendix E to 10 CFR part 50? Please provide a basis for your response.
Question 15: If there was a change to the exercise frequency,
should the exercise cycle be changed or otherwise be revised to ensure
there is adequate opportunity for the ERO to demonstrate adequate
performance under paragraph IV.F.2.j of appendix E to 10 CFR part 50?
Please provide a basis for your response.
Considerations for Emergency Planning Zones
The proposed change to 10 CFR 50.33(g)(1) would provide certainty
for a site-boundary EPZ for facilities less than 300 MWt. This proposed
change is based on NRC staff review of available technical analyses
that estimate radiological consequences for a variety of reactor
designs and design power levels. The NRC is considering factors that
may obviate the need for applicants of certain reactor designs to
perform
[[Page 44641]]
detailed analyses to support a site-boundary EPZ determination or reach
a determination that no EPZ is needed. Such factors may include use of
novel fuel forms or the performance of functional containment that may
generically demonstrate that predetermined, prompt protective measures
are unwarranted. The NRC is also considering factors that may
necessitate additional analysis by the applicant for facilities less
than 300 MWt.
Question 16: Are there additional design factors or analyses that
should be considered to provide certainty on EPZ determinations? Please
address within the comment the technical basis for consideration and
how they relate to EPZ determinations and, specifically, the criteria
of 10 CFR 50.33(g)(2).
Question 17: Should design features, other than power level, or
site attributes be considered for determining when additional analysis
by the applicant is necessary to justify site-boundary EPZ for
facilities less than 300 MWt? Please provide the basis for your
response.
Optional Submittal of Operational Programs
The NRC is proposing to add a new provision in 10 CFR 52.158(c)
that will allow an applicant for an ML to voluntarily submit
standardized operational program information for approval and to revise
10 CFR 52.171 to provide finality for such information that is approved
by the NRC in the ML review. The NRC is proposing these changes in
response to stakeholder interest in the context of deployment of
factory-fabricated reactors.
Question 18: Should the NRC consider making similar changes to the
requirements in 10 CFR 52.47, 52.63, 53.1239, and 53.1263 for design
certifications? Please provide a basis for your response.
Question 19: If the NRC were to add a provision for the optional
submittal of standardized operational program information in a design
certification application, please describe how change control should be
accounted for as a part of certifying the design.
Early Site Permit for Nuclear Power Plants
Question 20: With the proposed change to remove the requirement for
renewal, all changes to update the early site permit (ESP) must be made
by amendment. What other options could the staff consider to increase
the useability of an ESP? Could the provisions in the, ``Commission
Policy Statement on Deferred Plants,'' (52 FR 38077; October 14, 1987)
be applied to ESPs? Please provide a basis for your response.
Question 21: For a COL application that references an ESP, the
changes proposed in 10 CFR 52.39 and 53.1188 would limit finality on
safety and environmental issues to 20 years following ESP issuance or
approval of an update amendment. Because the proposed rule would also
retain the requirement for applicants to review for new and significant
information when referencing an ESP, the NRC requests input on whether
limiting finality to 20 years is necessary to ensure that all relevant
environmental information is considered in licensing decisions.
Manufacturing License Term Extension
Question 22: The manufacturing license term extension aligns with
the recently promulgated July 2, 2025, direct final rule, ``Revising
the Duration of Design Certifications,'' where the NRC replaced the 15-
year duration for design certifications with a 40-year duration period.
The Commission is interested in determining if a manufacturing license
or design certification term limit is necessary. Should the NRC
eliminate time limits for manufacturing licenses, design
certifications, or both? Please provide a basis for your response.
Nuclear Power Plant License Renewal
Safety Guidance
The NRC currently has separate safety guidance for initial and
subsequent renewal to address 40 to 60 years and 60 to 80 years of
operation, respectively. In addition, some licensees may soon apply for
a third renewal, for which no guidance has yet been developed. One
option would be to maintain a single set of safety guidance that
represents the state-of-the-art for aging management, covering initial
renewal, subsequent renewal, and periods beyond. This guidance would
identify an acceptable approach for managing aging and evaluating time-
limited aging analyses, regardless of the renewal period. The question
below applies to the Generic Aging Lessons Learned Report for initial
(NUREG-1801) and subsequent (NUREG-2191) license renewal and the
Standard Review Plan for initial (NUREG-1800) and subsequent (NUREG-
2192) license renewal.
Question 23: Should the NRC consolidate its safety guidance into
one collection that is time-independent (i.e., the same guidance
applying to all periods of extended operation), instead of having
separate guidance for initial renewal, subsequent renewal, and the
upcoming third renewal term? Is there a different approach that should
be considered? Please provide a basis for your response.
Implementation of Aging Management
Question 24: With the proposed change to use a 40-year ``tack-on''
license, what regulatory approach should the NRC take with respect to
aging management activities that are currently required to be
implemented prior to the period of extended operation? Should
conditions that are currently included in the renewed license be
revised or instead imposed by different means? Please provide a basis
for your response.
Considerations for Shorter License Renewal Term
Question 25: Should the NRC retain an option for licensees seeking
shorter renewals (those that would not exceed 40 years when combined
with the number of years remaining on an existing license) to apply for
a ``supersession'' license?
Increased Enrichment of Conventional and Accident Tolerant Fuel Designs
for Light-Water Reactors
Expansion of 10 CFR 71.55(g) Enrichment Levels
The NRC is proposing to change 10 CFR 71.55(g) to expand the
exception from 10 CFR 71.55(b) for enrichment levels between 5.0 and
10.0 weight percent U-235. The NRC is not considering expanding the
exception for enrichments from 10.0 weight percent U-235 up to but less
than 20.0 weight precent U-235 for two reasons: (1) limited technical
data with respect to overall mechanisms and consequences of a canister
breach allowing water in-leakage that would provide additional risk
insights; and (2) the current transportation package certification path
using 10 CFR 71.55(c) adequately addresses the potential need to ship
UF6 enriched above 10.0 weight percent U-235, has no
enrichment limitation, and is technology neutral.
Question 26: 10 CFR 71.55(c) provides a certification pathway that
has limited prescriptive requirements and no enrichment limitations.
Should the NRC add another option to the exception in 10 CFR 71.55(g)
to account for enrichment levels from 10.0 weight percent U-235 up to
but less than 20.0 weight percent U-235? Please provide a basis for
your response.
Regulatory Certainty of Performance-Based ECCS Rod Embrittlement
Criteria
Question 27: Will there be adequate regulatory certainty if the
fuel system criteria for uranium oxide or uranium-
[[Page 44642]]
plutonium oxide pellets within cylindrical zirconium-based cladding are
located in guidance and 10 CFR 50.46a(f) of the proposed rule is
performance-based? Please provide a basis for your response.
Inspection Requirements
The NRC is proposing to add non-destructive evaluation inspection
requirements for circumferential welds in piping systems with inner
diameters that are larger than the TBS to provide assurance that an
acceptable level of performance monitoring is maintained in these
systems such that breaks larger than the TBS remain highly unlikely.
The NRC has evaluated several sampling schemes to determine the
most effective approach for providing assurance against breaks larger
than the TBS in these systems. These inspections may support plant-
specific applicability of the TBS. The option in this proposed rule is:
Inspecting a smart sample of those welds with attributes
(e.g., operating temperature, type of welding, fabrication history)
most conducive to degradation and having the highest failure potential
before implementation of the proposed rule and in every subsequent
inservice inspection interval thereafter.
Other schemes include:
Inspecting all welds at least one-time over the plant's
remaining licensing period.
Inspecting a random sample of the weld population before
implementation of the proposed rule and in every subsequent inservice
inspection interval thereafter.
Question 28: Please provide feedback on the effectiveness of the
proposed sampling scheme in providing assurance against breaks larger
than the TBS in these systems. Are there pros and cons to the other
sampling schemes, an alternative sampling scheme, or a holistic
approach relying on existing performance monitoring strategies that the
NRC should consider, and why? Should the sampling scheme be required
before or as plants enter into extended operation beyond what is
currently authorized in the operating fleet (e.g., if authorized beyond
80 years)?
The NRC is also proposing a requirement to evaluate the effect of
any identified indications found during inspections required by this
proposed rule on the TBS. The NRC recognizes that any such findings
must be dispositioned under ASME section XI, which would require a
demonstration that the indications remain acceptable for continued
service or are repaired before placing the component back into service.
However, the NRC has not included criteria in this proposed rule
regarding what constitutes an acceptable additional evaluation of any
inspection findings. The NRC is considering that an acceptable
additional evaluation would consist of two parts. The first part would
be a deterministic or probabilistic fracture mechanics analysis to
demonstrate that the failure likelihood of any indications identified
during the inspection resulting in a likelihood of ruptures greater
than the TBS is less than 10\-6\ per year. The second part would, if
the indications are evidence that active degradation is occurring,
recharacterize the risk-informed inspection examination category for
the population of all similar welds to be commensurate with the
expected degradation mechanism. All future inspections of this weld
population would then be performed according to the ASME section XI
requirements associated with that inspection category.
Question 29: The NRC is interested in obtaining feedback on the
efficacy of this proposed evaluation method and more broadly on what an
acceptable additional evaluation of any indications found during the
inspection should entail.
DG-1261--Breakaway Oxidation
Question 30: The DG-1261, Revision 1, ``Measuring Breakaway
Oxidation Behavior,'' contains breakaway oxidation testing criteria.
The NRC is interested in obtaining feedback on this regulatory
position. Please provide a basis for your response. The NRC is
interested in specific information on how quality assurance practices
and industry operating experience have evolved since the underlying
research was published, and whether these developments should inform
potential adjustments to this provision.
Regulatory Analysis
Question 31: The NRC is interested in feedback on the draft
regulatory analysis, in particular, the benefits associated with power
uprates under the proposed 10 CFR 50.46a. Please provide a basis for
your response.
DG-1426--Risk-Informed Evaluation and Risk-Informed Evaluation Process
An entity that wishes to make changes under the proposed rule would
have to perform a risk-informed evaluation and demonstrate that the
proposed change meets the acceptance criteria in proposed 10 CFR
50.46a(h). The acceptance criteria in the proposed rule would include
that the total increases in core damage frequency (CDF) and large early
release frequency (LERF) due to the proposed change are very small and
that the overall plant baseline risk remains small. ``Small'' and
``very small'' are defined in RG 1.174 and DG-1426. The acceptance
criteria are limited to very small changes instead of small changes, in
part, because the TBS initiating event frequency is the same as the
limit for small changes (i.e., change in CDF of approximately
1x10-5). If the acceptance criteria were limited to small
changes, a PRA would show that not mitigating a TBS or DEGB LOCA is
always less than the small threshold, making the risk assessment
unnecessary and overly burdensome.
An entity seeking to make changes under the proposed rule without
prior NRC approval would have to have an NRC-approved risk-informed
evaluation process and demonstrate that any increases in the estimated
risk are minimal. The proposed acceptance criteria for proposed changes
that may be made without prior NRC approval were chosen to be an order
of magnitude lower than the acceptance criteria for changes that
require prior NRC approval for consistency with other established risk-
informed programs.
The DG-1426 contains draft guidance that would be acceptable to the
NRC for performing a risk-informed evaluation and establishing a risk-
informed evaluation process that satisfies the requirements of the
proposed rule. This DG follows the structure for risk-informed changes
in RG 1.174. Entities would be able to leverage changes made for
previous risk-informed amendments, such as 10 CFR 50.69, ``Risk-
informed categorization and treatment of structures, systems and
components for nuclear power reactors,'' or Technical Specification
Task Force Traveler 505, ``Provide Risk-Informed Extended Completion
Times--RITSTF Initiative 4B,'' when performing the risk-informed
evaluation. Therefore, the NRC does not expect that the risk-informed
evaluation would be burdensome.
Under current 10 CFR 50.59, ``Changes, tests and experiments,'' a
licensee would need to obtain a license amendment if there would be
more than a minimal increase in frequency or consequences of accidents
or malfunctions. As described in the preamble for the 10 CFR 50.59
final rule (64 FR 53582; October 4, 1999), PRAs (i.e., total increases
in CDF and LERF) may not be used to determine if the minimal increase
standard is met because 10 CFR 50.59 concerns DBEs, while RG 1.174
includes risk from severe accidents beyond the design basis. Similar to
the change in proposed
[[Page 44643]]
10 CFR 50.59 as described in section XII, ``Discussion--Risk-Informing
10 CFR 50.59 and Allowing Flexibility for Changes to Methods,'' of this
document, the risk-informed evaluation process in proposed 10 CFR
50.46a would provide additional flexibility by allowing a licensee to
use the total increases in CDF and LERF to determine if the increase in
risk associated with a change enabled by this proposed rule would be
minimal.
In this proposed rule, the NRC is including the risk-informed
evaluation and risk-informed evaluation process that were included in
the previously proposed but discontinued 10 CFR 50.46a rulemaking in
SECY-10-0161. As part of the development of that previous rule, the NRC
received comments from the public and the Advisory Committee on Reactor
Safeguards on the risk-informed evaluation and risk-informed evaluation
process related to, among other things, the use of the ``very small''
threshold for changes, the definition of ``minimal increase in risk,''
PRA maintenance and upgrade, the evaluation of cumulative risk, and the
relationship of the risk-informed evaluation process to 10 CFR 50.59.
In addition, the NRC staff received direction from the Commission in
SRM-SECY-07-0082 to use the ``very small'' threshold for changes. While
some of these issues are addressed in this proposed rule, the NRC is
seeking additional information to identify if further changes are
needed.
Question 32: The NRC is interested in any additional considerations
or feedback on the risk-informed evaluation and risk-informed
evaluation process proposed in 10 CFR 50.46a and the implementing
guidance in DG-1426, including whether they should be modified or
removed from the proposed rule in their entirety. Specifically, what
are the advantages and disadvantages of including a risk-informed
evaluation as a necessary part of implementing the rule? In addition,
what are the advantages and disadvantages of restricting all changes to
very small risk increases? Would the risk-informed evaluation,
restricting changes to very small risk increases, or both, make it
burdensome to implement changes under the proposed rule? What are the
advantages and disadvantages of removing the risk-informed evaluation
and risk-informed evaluation process from the proposed rule? If the NRC
removed the risk-informed evaluation and risk-informed evaluation
process from the proposed rule, how should changes that could impact
the risk calculation be evaluated? Please provide any supporting basis
or justification for your response.
New Reactor Applicability
Question 33: The proposed rule would establish a TBS based on
nuclear power reactor licensees' operating experience with piping
materials, piping designs, service conditions, operational programs,
and mitigation and control of age-related degradation programs. The
proposed rule would allow new reactors to submit for NRC approval an
alternate TBS that includes a demonstration that the proposed reactor
design is similar to the designs of reactors authorized to operate
under 10 CFR part 50 on December 31, 2015. This provision might not be
applicable to some new reactor designs, including small modular
reactors, due to fundamental design differences. For example, some of
the new reactor designs may have novel LOCA mitigation strategies, such
as passive features or use of valves and associated flanges in the
piping system to ensure coolant inventory retention. Some applicants
may be able to demonstrate that the failure frequencies at particular
locations in the piping system are extremely low and can be maintained
extremely low during plant operation. In lieu of the TBS, the failure
frequencies of the piping system may be established and maintained
extremely low during the plant operation to justify relaxation of the
required assumptions for the analyses similar to the proposed rule
(e.g., a realistic treatment of defense-in-depth features and best-
estimate analyses). The NRC is interested in receiving feedback on the
applicability of the proposed rule to such new reactor designs. Please
include in your feedback whether it is appropriate to risk-inform the
LOCA requirements in light of novel design features, and language that
would appropriately allow entities to treat breaks at particular
locations in the piping systems for new reactor designs similarly to
break sizes larger than the TBS in the proposed rule. Please also
include in your feedback whether the option for a non-size-based TBS
would improve the applicability of the proposed rule to these designs
along with any supporting basis or justification for your response.
Alternate Transition Break Size
Question 34: The NRC incorporated in the proposed rule enabling
language to allow currently operating power reactor licensees to
recommend and justify an alternative TBS. The NRC is interested in
feedback on this option. Please include in your feedback the potential
advantages that could be realized by this alternative and if potential
benefits are more likely to be realized generically or for individual
plants. Please also provide any disadvantages that may result from a
loss of regulatory consistency if similar plants have different TBS
values and any potential impacts on fulfilling the inspection
requirements in proposed 10 CFR 50.46a(b)(3). Please also describe the
impact of any associated costs necessary to develop a risk-informed
technical basis for an alternative TBS and the associated uncertainties
due to the lack of current guidance for developing this basis.
Alternative 10 CFR 50.46a Licensing Pathways
Question 35: This proposed rule would add a voluntary alternative
to 10 CFR 50.46 designed to provide flexibility by allowing licensees
to adopt tailored approaches for ECCS performance evaluations. This
flexibility is proposed in 10 CFR 50.46(a)(1), which would provide
applicants and licensees the ability to select between two acceptance
criteria pathways. The first is presented in proposed 10 CFR 50.46(b),
which is the current 10 CFR 50.46 acceptance criteria. The second is
presented in proposed 10 CFR 50.46a(f), which contains the new
performance-based acceptance criteria. The NRC expects that the
performance-based requirements in proposed 10 CFR 50.46a(f) would
facilitate implementation of potential alternative approaches to FFRD
such as those described in the regulatory basis for this rulemaking,
some of which are already being considered by the NRC. These
alternatives could be beneficial for licensees that elect not to pursue
the remainder of proposed 10 CFR 50.46a. Please comment on any elements
of the proposed rule that could have unintended consequences for
implementation of these alternatives.
ECCS Reporting Requirements
Question 36: The industry has proposed that a graded approach could
be used to relax the ECCS reporting requirements based on context
beyond the current 10 CFR 50.46 criteria. Please provide thoughts on
how a graded approach could be applied to relax the ECCS reporting
requirements in 10 CFR 50.46 and proposed 10 CFR 50.46a. Please provide
a basis for your response.
Threshold for Significance Determination
Question 37: The proposed fixed 50-degree peak clad temperature
(PCT)
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threshold for significant errors can either overstate or understate the
safety significance of a change depending on available margin to the
acceptance criterion. Further, it does not consider designs where a
parameter other than PCT is proposed as the acceptance criterion.
Alternative approaches, such as a margin-based threshold can more
accurately reflect actual risk significance. Please provide your
thoughts, with justification and any supporting information, on
alternative approaches for determining the significance of errors in
the LOCA evaluation model (e.g., percent reduction in the margin
between the acceptance criteria and the approved maximum value of the
acceptance parameter).
ECCS Performance
Question 38: The proposed rule explicitly only provides the TBS-
based approach to address FFRD. The TBS-based approach may not provide
enough margin to demonstrate coolability at the upper end of the burnup
range allowed by the proposed rule and therefore, could result in
implementation and licensing challenges. The NRC previously included a
dose consequence-based approach to addressing FFRD in the regulatory
analysis for the Increased Enrichment Rule (Docket ID: NRC-2020-0034).
What are the advantages and disadvantages of including an explicit
option in the performance-based ECCS acceptance criteria in proposed 10
CFR 50.46a(f) that is based on dose-consequence? What implementation
challenges (e.g., acceptance threshold) would need to be addressed for
such an option? Please provide the rationale and any supporting
information for your response.
Streamlined Quality Assurance Criteria for Nuclear Power Plants and
Fuel Reprocessing Plants
Question 39: Appendix T of 10 CFR is intended to address nuclear
supply chain constraints by providing a graded approach for additional
QA flexibilities and aligning with international standards. The
proposed rule limits the applicability of appendix T to nth-of-a-kind
(NOAK) plants referencing a first-of-a-kind (FOAK) plant constructed
and operated under appendix B. How can the NRC expand the applicability
of appendix T to FOAK plants and existing licensees, including any
guardrails or restrictions that may be desirable (e.g., vendor
inspections for FOAK and new suppliers for NOAK)? Should the NRC
consider alternative approaches to appendix T such as endorsement of
ISO 19443 or changes to appendix B? Please provide the basis for your
recommendations, including any unintended consequences.
Question 40: The change process for the Quality Management System
(QMS) in proposed 10 CFR 50.54(a)(5) requires changes to the QMS be
submitted to the NRC for NRC approval prior to implementation. Should
the change process have a graded approach or mirror 10 CFR 50.54(a)(4)
in which only changes that reduce the commitments be submitted to the
NRC for prior approval? Please explain if there is a preferred method.
Alternative Risk-Informed and Performance-Based Acceptance Criteria for
10 CFR Parts 50 and 52
Question 41: The NRC is seeking views on increasing the flexibility
provided by the proposed rule. The rule currently does not include the
allowance for a licensee or applicant to adopt a previously approved
alternative without submitting an application to the NRC, provided that
the basis for the staff's approval is demonstrated and documented to be
applicable to the adoptee's plant or facility. Would adding such
flexibility increase or accelerate the use of the proposed rule? What
are some viable approaches to resolve any conflicts of such flexibility
with the requirements in 10 CFR 50.59?
Siting
Question 42: The proposed rule introduces a new graded approach to
siting that provides entry criteria for the less restrictive
requirements in subpart A to 10 CFR part 100 (i.e., demonstrating an
unmitigated consequence of less than 25 rem (0.25 Sv) TEDE at the site
exclusion area boundary) as an alternative to the prescriptive siting
requirements in subpart B. The associated draft guidance provides
detailed information on how to evaluate this entry criteria. Given the
significance of this new graded approach to siting, are there
implications in terms of implementation of this new approach that
should be further explored in guidance or addressed at the final rule
stage?
XXXVIII. Regulatory Flexibility Certification
As required by the Regulatory Flexibility Act of 1980, 5 U.S.C.
605(b), the Commission certifies that this rule, if adopted, will not
have a significant economic impact on a substantial number of small
entities. This proposed rule affects only the licensing and operation
of nuclear power plants. The companies that own these plants do not
fall within the scope of the definition of ``small entities'' set forth
in the Regulatory Flexibility Act or the size standards established by
the NRC (10 CFR 2.810).
XXXIX. Regulatory Analysis
The NRC has prepared a draft regulatory analysis on this proposed
regulation. The analysis examines the costs and benefits of the
alternatives considered by the NRC. The NRC requests public comment on
the draft regulatory analysis. The draft regulatory analysis is
available as indicated in the ``Availability of Documents'' section of
this document. Comments on the draft analysis may be submitted to the
NRC as indicated under the ADDRESSES caption of this document.
XL. Backfitting and Issue Finality
This section describes the backfitting and issue finality
implications of this proposed rule and the draft guidance documents
described in section XLVII, ``Availability of Guidance,'' of this
document, as applied to pertinent NRC approvals and certain applicants
that reference NRC approvals in their applications. The NRC's
backfitting provisions associated with nuclear power plants licensed
under 10 CFR part 50 appear in 10 CFR 50.109, ``Backfitting.'' Issue
finality provisions (analogous to the backfitting provisions in 10 CFR
50.109) for approvals under 10 CFR part 52 are located in various
provisions of 10 CFR part 52. The NRC Management Directive (MD) 8.4,
``Management of Backfitting, Forward Fitting, Issue Finality, and
Information Requests,'' describes the Commission's policies on
backfitting and issue finality.
Part 53 of 10 CFR contains backfitting and issue finality
provisions. Those provisions apply to NRC actions that would affect 10
CFR part 53 licensees and certain applicants that reference NRC
approvals under 10 CFR part 53. The NRC has not issued any licenses or
other approvals under 10 CFR part 53, so no licensees or applicants
under 10 CFR part 53 could be affected by this proposed rule.
Therefore, this proposed rule's changes to 10 CFR part 53 would not
constitute backfitting under 10 CFR part 53 or affect the issue
finality of an approval issued under 10 CFR part 53.
``Backfitting'' is defined in 10 CFR 50.109(a)(1) as, in relevant
part, a modification of or addition to the systems, structures, and
components or design of a facility; or the design approval or
manufacturing license (ML) of a facility; or the procedures or
organization required to design, construct, or operate a facility,
which
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results from a new or amended provision in the Commission's
regulations. The issue finality provision for combined licenses (COLs)
located in 10 CFR 52.98 provides, in relevant part, that the Commission
may not modify, add, or delete any term or condition of a COL except in
accordance with the provisions of 10 CFR 50.109. Essentially, if a
change does not constitute backfitting, then the change does not affect
the issue finality of a COL.
Part 52 of 10 CFR contains an issue finality provision for MLs. No
entity holds an ML under 10 CFR part 52. Therefore, the proposed
changes to 1) amend 10 CFR 52.173 to change the duration of an ML to a
maximum of 40 years, 2) amend 10 CFR 52.181 to change the duration of a
renewed ML to a maximum of 40 years, 3) amend 10 CFR 50.71(f) regarding
final safety analysis report (FSAR) updates by ML holders, and 4) amend
10 CFR 52.171(b)(1) to allow the holder of an ML to use the regulations
in 10 CFR 50.59 to determine whether changes to the facility or
procedures as described in the FSAR would require prior Commission
approval of an amendment to the ML, would not affect the issue finality
of an ML.
Similarly, the proposed change to amend 10 CFR 52.98, which would
allow COL holders who reference an ML to use the applicable change
processes in 10 CFR part 50 to determine whether changes to the
facility or procedures as described in the FSAR would require prior
Commission approval, would not affect the issue finality of a COL
referencing an ML because there are no MLs for a COL to reference.
Also, the proposed change to amend 10 CFR 50.59 to allow the holder of
an operating license (OL) under 10 CFR part 50 or a COL under 10 CFR
part 52 that references a reactor manufactured under an ML to make
changes in the facility or procedures as described in the FSAR without
requesting a license amendment if the changes would be the same as
changes approved by amendment to the ML and upon a determination that
implementing the changes would be consistent with the basis for the
Commission's approval of the amendment to the ML and would not involve
any additional changes that would require an amendment to the OL or
COL, would not constitute backfitting of a 10 CFR part 50 operating
license or affect the issue finality of a 10 CFR part 52 COL
referencing an ML because there are no MLs for an OL or a COL to
reference.
Two sets of proposed changes would affect the issue finality of
standard design certifications under 10 CFR part 52. First, the NRC
proposes to clarify the definitions of Tier 1 information in section
II.D of appendices A, D, E, F, and G to 10 CFR part 52 by identifying
specific sections and tables within the generic design certification
documents that correspond to each of these categories of Tier 1
information. Second, the NRC proposes to amend sections II.F and VIII.B
of appendices A and E to 10 CFR part 52 to revert all Tier 2*
information to Tier 2 status after the plant first achieves full power.
These proposed changes would modify the certification information of
the certified standard designs in those appendices. The issue finality
provision for design certifications is in 10 CFR 52.63(a)(1) and
prohibits the modification of certification information unless the
Commission determines in a rulemaking that certain criteria are met. In
this case, the Commission determines that these proposed changes would
reduce unnecessary regulatory burden and maintain protection to public
health and safety and the common defense and security in accordance
with 10 CFR 52.63(a)(1)(iii). Therefore, the NRC may propose to make
these changes.
Under 10 CFR 52.63(a)(2), in a rulemaking performed under 10 CFR
52.63(a)(1)(iii), 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.
The NRC prepared an analysis addressing whether the benefits of the
proposed changes would justify the costs for plants that are already
licensed or for which an application for a permit or license is under
consideration (see section XXXIX, ``Regulatory Analysis,'' of this
document).
Other proposed changes to the standard design certifications under
10 CFR part 52 would not affect the issue finality of those certified
designs or COL holders referencing the certified designs because the
proposed changes would not amend certification information of the
designs. The Commission explained in the 2007 10 CFR part 52 final rule
that 10 CFR 52.63(a) applies to changes to the certification
information (i.e., the information in the generic design control
document incorporated by reference in a design certification appendix
in 10 CFR part 52) but does not apply to changes to the certified
design rule language (72 FR 49381; August 28, 2007). Therefore,
deleting section IX of appendix D to 10 CFR part 52 because it is
redundant with 10 CFR 52.99 and 52.103 would not affect the issue
finality of the existing design certification or COLs referencing the
existing certified design because the proposed change would not amend
certification information of the design.
Of the other proposed changes, many would not constitute
backfitting under 10 CFR part 50 or affect the issue finality of a COL
under 10 CFR part 52 because they would be non-mandatory relaxations of
existing requirements. As explained in MD 8.4, non-mandatory
relaxations of regulations generally do not meet the definition of
``backfitting'' in 10 CFR 50.109(a)(1). Thus, these proposed changes
would not constitute backfitting under 10 CFR part 50 or affect the
issue finality of a COL under 10 CFR part 52. The following proposed
changes would be non-mandatory relaxations of existing requirements:
Amending paragraphs IV.F.2.a.(i) through (iii) of appendix
E to 10 CFR part 50 and paragraphs 10 CFR 50.160(c)(1) and (c)(2) to
remove the requirement to demonstrate compliance within 2 years before
issuance of an OL under 10 CFR part 50 or the scheduled date of initial
loading of fuel for a 10 CFR part 52 COL. A 10 CFR part 50 OL applicant
or 10 CFR part 52 COL holder could conduct the initial exercise within
2 years of the respective milestones and would comply with the proposed
rule.
Amending paragraph IV.D.3 of appendix E to 10 CFR part 50
to revise requirements for when backup alert and notification system
methods are required. Licensees could continue to have a single primary
method and a single backup method and would comply with the proposed
rule.
Amending 10 CFR 50.33(g), 50.47(c)(2), and 50.160(b)(3) to
ensure the plume exposure pathway emergency planning zone (EPZ) is no
larger than needed to implement predetermined, prompt protective
measures. Licensees could continue to have a plume exposure pathway EPZ
of an area about 10 miles (16 km) in radius and would comply with the
proposed rule.
Eliminating the requirement to define an ingestion pathway
EPZ by removing references to the ingestion pathway EPZ in 10 CFR
50.33(g)(1), 50.47(b)(10), 50.47(c)(2), and paragraph IV.F.2.a.(i) and
footnote 1 of appendix E to 10 CFR part 50. Licensees could continue to
have an ingestion pathway EPZ of an area about 50 miles (80 km) in
radius and would comply with the proposed rule.
Amending paragraph IV.E.8.b of appendix E to 10 CFR part
50 to require the emergency plan to specify the location of the
emergency operations facility in relation to the EPZ boundary. A
licensee could continue to locate its
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emergency operation facility between 10 and 25 miles (16 and 40 km) of
the nuclear power reactor site, or a primary facility less than 10
miles (16 km) from the nuclear power reactor site and a backup facility
between 10 and 25 miles (16 and 40 km) of the nuclear power reactor
site, based on the current 10-mile (16-km) EPZ requirement, and would
comply with the proposed rule.
Amending 10 CFR 50.54(t) to eliminate the requirement for
licensees to ensure that all program elements are reviewed by people
who have no direct responsibility for implementation of the emergency
preparedness (EP) program. Licensees could continue to have their EP
program elements reviewed by people who have no direct responsibility
for implementation of the EP program and would comply with the proposed
rule.
Amending paragraphs IV.5, IV.6, and IV.7 of appendix E to
10 CFR part 50 to eliminate certain requirements related to updates of
evacuation time estimates. Licensees could continue to update their EPZ
permanent resident population estimates and evacuation time estimate
analysis, and would comply with the proposed rule.
Amending paragraph IV.E.9.d of appendix E to 10 CFR part
50 to eliminate the requirement for monthly tests of communications
between the licensee and the appropriate NRC Regional Office Operations
Center. Licensees could try to communicate monthly with their NRC
Regional Office Operations Center and would comply with the proposed
rule.
Amending 10 CFR 71.55 to expand the exception from 10 CFR
71.55(b) for packages containing enrichment levels between 5.0 and 10.0
weight percent U-235. Packages could contain enrichment levels below
5.0 weight percent U-235 and would comply with the proposed rule.
Amending 10 CFR 50.67 and General Design Criterion (GDC)
19 of appendix A to 10 CFR part 50 to increase the numerical value of
the control room design criteria from 5 to 10 rem (0.05 to 0.10 Sv) and
enable a higher control room design criteria, ranging from 10 to 25 rem
(0.10 to 0.25 Sv) TEDE, for licensees whose facility-specific risk
profiles warrant them. Licensees could meet the current control room
habitability design requirements and would comply with the proposed
rule.
Removing the restriction limiting proposed alternatives to
paragraphs (b) through (h) of 10 CFR 50.55a so that proposed
alternatives would now be permitted for all regulatory requirements in
10 CFR 50.55a using the existing criteria in paragraphs (z)(1) and (2)
of 10 CFR 50.55a. Licensees could continue to propose alternatives to
only paragraphs (b) through (h) of 10 CFR 50.55a and would comply with
the proposed rule.
Amending appendix A to 10 CFR part 50 to clarify that (1)
deviations from GDC could be identified and justified within licensing
submittals, with no need for a separate exemption request; and (2)
demonstrating compliance with GDC 28 of appendix A to 10 CFR part 50
could be based on a different design basis accident than the control
rod ejection or control rod drop accident. Licensees could continue to
submit an exemption request or demonstrate compliance with GDC 28 using
an evaluation of control rod ejection or control rod drop accidents and
would comply with the proposed rule.
Amending 10 CFR 54.37(b) to only require that the final
safety evaluation report include a summary description of aging
management activities that addresses newly identified SSCs, as
appropriate, consistent with what is required in the renewal
application.
Some of the changes in this proposed rule would not constitute
backfitting under 10 CFR part 50 or affect the issue finality of an
approval under 10 CFR part 52 because the proposed changes would
provide a voluntary alternative set of requirements. Licensees could
continue to comply with the current applicable requirement and would
not be required to comply with the proposed rule. The following
proposed changes would not require holders of 10 CFR part 50 or 52
approvals to comply with the proposed rule changes:
Amending various regulations to allow applicants and
licensees under 10 CFR parts 50 and 52 the option to use 10 CFR 50.160.
Amending 10 CFR 52.26 to remove the requirement for an ESP
to include a fixed term and making related conforming changes to
subpart A of 10 CFR part 52.
Amending 10 CFR 50.68(b)(7) to allow enrichment above 5.0
weight percent U-235.
Amending 10 CFR 50.46a and making conforming changes to
provide alternative acceptance criteria for emergency core cooling
systems for light-water reactors.
Adding 10 CFR 50.221 to establish an optional
verification, validation, and uncertainty quantification (VVUQ)
program.
Adding appendix T to 10 CFR part 50 to establish an
alternative to the current quality assurance requirements in appendix B
to 10 CFR part 50.
Amending various paragraphs in 10 CFR 50.4, 50.34, 50.54,
50.55, and 52.79 to make conforming changes to reflect the addition of
appendix T to 10 CFR part 50.
Adding 10 CFR 50.220 and 52.220 to allow licensees and
applicants to voluntarily submit and use technology-inclusive, risk-
informed, or performance-based acceptance criteria as alternatives to
existing prescriptive requirements.
Several of the proposed changes would apply to only future
applicants and to current licensees at their discretion, and therefore
would not constitute backfitting under 10 CFR part 50 or affect the
issue finality of a 10 CFR part 52 approval. Applicants and potential
applicants (for licenses, permits, and other regulatory approvals)
generally are not within the scope of the backfitting or issue finality
regulations. Those regulations include language delineating when those
provisions begin; in general, the backfitting and issue finality
regulations begin upon the issuance of the license, permit, or other
approval. The following proposed changes would apply to only future
applicants and to current licensees at their discretion:
Amending paragraph I.5 of appendix E to 10 CFR part 50 to
determine the degree to which compliance with the requirements in
certain sections of appendix E is necessary on a case-by-case basis for
power reactors with a site-boundary EPZ or no EPZ.
Amending 10 CFR 50.33(g), 50.47(c)(2), and 50.160(b)(3) to
simplify EPZ determinations.
Amending 10 CFR 50.33(g) to eliminate the requirement to
submit response plans of State, local, and participating Tribal
governmental entities.
Amending 10 CFR 50.34(a)(10) and sections I.1, I.2, and II
of appendix E to 10 CFR part 50 to remove the requirements for
applicants to submit preliminary plans for coping with emergencies in
the preliminary safety analysis report.
Amending 10 CFR 52.158 and 52.171 to allow ML applicants
the option to submit essentially complete operational program
information with their applications and to provide finality to such
program information that is reviewed and approved by the NRC as part of
the ML review.
Amending 10 CFR 54.31 to extend the maximum renewal period
for a license to 40 years.
Adding 10 CFR 54.21(a)(4) to allow applicants to
voluntarily propose risk-informed and performance-based
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alternatives to the current prescriptive requirements of the aging
management review.
Amending 10 CFR 54.17(c) to remove the 20-year limit on
applying for license renewal.
Amending 10 CFR 54.21(c)(2) to remove the requirement to
submit a list of plant-specific exemptions granted under 10 CFR 50.12
that are based on time-limited aging analyses and a justification for
continuing those exemptions during the extended operating period.
Amending 10 CFR 54.22 to remove the requirement that
applicants include and justify any technical specification changes
needed to manage the effects of aging.
Amending 10 CFR part 100 to allow power reactor applicants
under 10 CFR part 50 or 52 to have greater flexibility to determine the
appropriate level of site characterization.
Amending or adding 10 CFR 50.10, 50.34(b), 50.57, 50.58,
51.4, 52.79(a), 52.85, 52.97(d), and 52.98 to revise the definition of
construction to better focus the definition on safety-significant
matters and reduce the cost impact of the current definition, to allow
certain applicants to request generic finality, and to allow certain
construction activities under a general license.
Amending 10 CFR 50.2 to add definitions for the terms
``design basis events'' and ``beyond design basis events'' and making a
conforming change to 10 CFR 50.49(b)(1)(ii).
Adding a new paragraph (b)(2) to 10 CFR 50.75 to allow new
reactor applicants and licensees to use an alternative decommissioning
funding assurance pathway.
Amending 10 CFR 50.75(c)(1) to delete language in the
table of minimum amounts that requires reactors of less than 1200 MWt
to use the certification amount for a 1200 MWt reactor.
Amending 10 CFR 50.75(e) to include ``applicant or'' in
all appropriate places where currently only ``licensee'' is referenced.
Amending 10 CFR 50.34(a)(4) and (b)(4) and footnote 1 of
10 CFR 50.34(a) regarding construction permit and OL application
requirements and making conforming changes to 10 CFR 52.47, 52.79,
52.137, and 52.157.
Amending paragraph (a)(1)(ii)(D) and footnotes 3 and 6 of
10 CFR 50.34 to use technology-inclusive language and making conforming
changes to 10 CFR 52.17, 52.47, 52.79, 52.137, and 52.157.
Amending footnote 4 of 10 CFR 50.34(a)(1)(ii)(D)(1) to
remove outdated information regarding recommendations included in a
1959 National Bureau of Standards handbook.
Amending 10 CFR 52.1 to add the definitions of Tier 1,
Tier 2, and Tier 2*, which would apply to design certifications issued
after the effective date of the proposed rule (if finalized).
The remaining proposed changes in this proposed rule would not
constitute backfitting under 10 CFR part 50 or affect the issue
finality of a 10 CFR part 52 COL, and they would not be non-mandatory
relaxations of existing requirements, voluntary alternative
requirements, or applicable to only future applicants. These proposed
changes would not require a licensee to modify or add to systems,
structures, components, or the design of a facility; or the design
approval or ML of a facility; or the procedures or organization
required to design, construct, or operate a facility. Therefore, the
proposed changes would not meet the 10 CFR 50.109 definition of
``backfitting'' and, thus, would not constitute backfitting and would
not affect the issue finality of a 10 CFR part 52 COL. The following
proposed rule changes would fall into this category:
Amending 10 CFR 50.160(b)(1)(iv)(A)(2) and paragraph
IV.D.2 of appendix E to 10 CFR part 50 to use modern terminology.
Amending 10 CFR 50.54(q) to revise and risk-inform the
emergency plan change process. The criteria to determine whether the
licensee's emergency plan change would require prior NRC approval and
the process to obtain that approval in 10 CFR 50.54(q) are not within
the scope of ``backfitting'' as defined in 10 CFR 50.109(a)(1) because
they are part of an NRC-designed administrative change process that is
not required to design, construct, or operate a facility. In addition,
the procedures a licensee might use to decide whether to change its
emergency plan are not required to design, construct, or operate a
facility. Other proposed changes that would amend a change process are
the following:
[cir] Adding 10 CFR 50.59(e) to establish a risk-informed
alternative to the existing 10 CFR 50.59 change process.
[cir] Amending 10 CFR 50.59(c)(2)(viii) to allow licensees to
implement certain changes to analytical methods described in the FSAR
(as updated) without prior NRC approval, provided those changes are
undertaken pursuant to an NRC-approved VVUQ program under 10 CFR
50.221.
[cir] Amending section VIII.A of appendices A, D, E, F, and G to 10
CFR part 52 to revise the change process for licensee-requested changes
to Tier 1 design description information and Tier 1 ITACC information,
and to revise the change process for changes and departures during
commercial operation.
[cir] Amending section VIII.B.5.c of appendices A, D, E, F, and G
to 10 CFR part 52 to amend the change process criteria used to
determine if a proposed departure from Tier 2 information affecting the
resolution of an ex-vessel severe accident design feature identified in
the plant-specific DCD requires a license amendment.
[cir] Amending section VIII.B of appendix D to 10 CFR part 52 to
allow the use of the 10 CFR 50.59-like criteria in sections VIII.B.5.b
and B.5.c to determine if licensees who reference appendix D to 10 CFR
part 52 may depart from Tier 2* information without prior NRC approval.
Amending table S-3 in 10 CFR 51.51(b). The information in
table S-3 provides the basis for evaluating the contribution of the
environmental effects of the uranium fuel cycle to support the NRC's
NEPA obligations at the time of a nuclear reactor licensing action.
Because table S-3 only provides information about the environmental
effects away from a nuclear reactor, the NRC's proposed revision to
table S-3 would not result in a modification or addition that meets the
definition of ``backfitting'' in 10 CFR 50.109(a)(1).
Amending table S-4 in 10 CFR 51.52. The information in
table S-4 provides the basis for evaluating the contribution of the
environmental impacts of the transportation of fuel and waste to and
from a nuclear reactor to support the NRC's NEPA obligations at the
time of a nuclear reactor licensing action. Because table S-4 only
provides information about the environmental impacts away from a
nuclear reactor, the NRC's proposed revisions to table S-4 would not
result in a modification or addition that meets the definition of
``backfitting'' in 10 CFR 50.109(a)(1).
Amending the terminology in 10 CFR 50.67 and GDC 19 of
appendix A to 10 CFR part 50 to clarify the regulations.
Amending current 10 CFR 50.34, 50.46, 50.46a, and 50.69;
GDC 17, 35, 38, 41, 44 and 50 of appendix A to 10 CFR part 50; appendix
K to 10 CFR part 50; 10 CFR 52.47, 52.54, 52.79, 52.137, and 52.157;
and appendix G to 10 CFR part 52 to reflect the proposed 10 CFR 50.46a.
Many of these proposed changes would support implementation of the
proposed 10 CFR 50.46a, which would offer voluntary alternative
criteria for emergency core cooling systems for light-water reactors.
These proposed conforming changes would not meet the definition of
backfitting for the same
[[Page 44648]]
reasons why the proposed 10 CFR 50.46a--voluntary alternative
requirements--would not meet the definition of backfitting. The other
proposed conforming changes would affect regulations for applicants
and, therefore, would not constitute backfitting under 10 CFR part 50
or affect the issue finality of a 10 CFR part 52 approval.
Removing footnote 3 from 10 CFR 50.49(b)(1) and
redesignating footnote 4 as footnote 1.
Amending 10 CFR 50.75(e), (g), and (h) to make editorial
corrections.
Amending 10 CFR 52.63(a)(1)(vii) to eliminate ``increased
standardization'' as a criterion that the Commission must meet before
modifying, rescinding, or imposing new requirements on certification
information by rulemaking.
Amending 10 CFR 52.63(a)(4)(ii) to eliminate whether
special circumstances outweigh any decrease in safety that may result
from a reduction in standardization as a criterion that the Commission
must meet before imposing new requirements on a design by plant-
specific order.
Amending 10 CFR 52.63(b)(1) to eliminate the requirement
for the Commission to review the impact of a requested exemption on
standardization.
Amending 10 CFR 52.93 and 52.171 to eliminate the
requirement for the Commission to discuss the impact of a change on
standardization as a criterion for the justification for departures
from ML information.
As described in the ``Availability of Guidance'' section of this
document, the NRC is issuing 24 draft guidance documents that, if
finalized, would provide guidance on the methods acceptable to the NRC
for complying with aspects of this proposed rule. Further, as discussed
in the guidance documents, applicants and licensees would not be
required to comply with the positions set forth in the guidance.
Therefore, issuance of the guidance documents as final guidance would
not constitute backfitting under 10 CFR parts 50 and 53 or affect the
issue finality of any approval issued under 10 CFR parts 52 and 53.
XLI. Cumulative Effects of Regulation
The NRC seeks to minimize potential negative consequences resulting
from the cumulative effects of regulation (CER). The NRC believes that
the deregulatory impacts of this rulemaking activity are unlikely to
cause implementation challenges for stakeholders. In addition, during
the pendency of this rulemaking, the NRC is deprioritizing issuance of
regulatory actions that might influence the implementation date for the
new rule requirements (e.g., orders, generic communications, license
amendment requests, and inspection findings of a generic nature).
To fully understand any potential CER implications that could
result from this rulemaking, the NRC is asking the following questions.
Response to these questions is voluntary and any input will be
considered during development of the final rule.
1. The NRC is proposing an effective date that will be 30 days
after the date of publication of a final rule. Does this provide
sufficient time to implement the proposed requirements? Please provide
a rationale for your response.
2. Are there unintended consequences related to this rulemaking and
how should they be addressed? Please provide a rationale for your
response.
3. Please comment on the NRC's cost and benefit estimates in the
regulatory analysis that supports this proposed rule.
XLII. Plain Writing
The Plain Writing Act of 2010 (Pub. L. 111-274) requires Federal
agencies to write documents in a clear, concise, and well-organized
manner. The NRC has written this document to be consistent with the
Plain Writing Act as well as the Presidential Memorandum, ``Plain
Language in Government Writing,'' published June 10, 1998 (63 FR
31885). The NRC requests comment on this document with respect to the
clarity and effectiveness of the language used.
XLIII. National Environmental Policy Act
The Commission proposes to determine under the National
Environmental Policy Act of 1969, as amended, and the Commission's
regulations in subpart A of 10 CFR part 51, that this rule, if adopted,
would not be a major Federal action significantly affecting the quality
of the human environment, and an environmental impact statement is not
required. The basis of this proposed determination regarding potential
environmental impacts is the implementation of the proposed rule for
the regulations described in this Federal Register notice would not
have a significant impact on the environment. The proposed requirements
would be administrative in application or matters of procedure or
provide an equivalent level of safety as existing requirements;
therefore, there would be similar environmental impacts from the
implementation of the regulations in this proposed rule as there are
for existing requirements.
The proposed determination of this draft environmental assessment
is that there will be no significant effect on the quality of the human
environment from this action. Public stakeholders should note, however,
that comments on any aspect of this environmental assessment may be
submitted to the NRC as indicated under the ADDRESSES caption. The
draft environmental assessment is available as indicated under the
``Availability of Documents'' section of this document. This
environmental assessment and proposed finding of no significant impact
can be tracked with identification number NEPA ID EAXX-429-00-000-
1770782365.
XLIV. Paperwork Reduction Act
This proposed rule contains new or amended collections of
information subject to the Paperwork Reduction Act of 1995 (44 U.S.C.
3501 et seq.). This proposed rule has been submitted to the Office of
Management and Budget for review and approval of the information
collections.
Type of submission: New.
The title of the information collection: Modernizing Reactor
Licensing, Safety Oversight, and Siting Practices.
OMB approval numbers: 3150-0008, 3150-0011, 3150-0093, 3150-0151,
3150-0155, 3150-0264, and 3150-0274.
The form number if applicable: Not applicable.
How often the collection is required or requested: Collections are
submitted one-time, on-occasion, and periodically. The frequency of
collections changes by adding new periodic reporting under 10
CFR50.46a, introducing event-driven submissions for quality management
system (QMS) changes and EP framework transitions, while reducing EP
program reviews from annual to biennial under 10 CFR50.54(t). Changes
to the QMS must be maintained for three years, while other records are
generally required to be maintained for the length of license.
Who will be required or asked to respond: The proposed rule would
affect a range of entities that apply for or hold licenses from the
NRC. The affected entities encompass both new applicants and current
licensees operating under 10 CFR parts 50, 52, 53, 54, 71, and 100 and
10 CFR 50.55a.
An estimate of the number of annual responses: 2,075.1 (2,063.1
reporting responses, 10.3 recordkeeping responses for 10 CFR part 50;
1.7 reporting responses for 10 CFR part 53).
The estimated number of annual respondents: 130 (120 for 10 CFR
part 50, 5 for 10 CFR part 53, 5 for 10 CFR part 100).
[[Page 44649]]
An estimate of the total number of hours needed annually to comply
with the information collection requirement or request: 33,321.6
(83,116.6 reporting + 6,150 recordkeeping for 10 CFR part 50; -835
reporting for 10 CFR part 53; -55,110 reporting for 10 CFR part 100).
Abstract: The NRC is proposing to amend its regulations to
modernize reactor licensing, safety oversight, and siting practices.
The proposed rule introduces voluntary risk-informed and performance-
based alternatives, updates EP requirements, streamlines quality
assurance criteria through a new appendix T to 10 CFR part 50, and
revises siting criteria to incorporate risk insights. These changes
affect information collections associated with license applications,
license amendment requests, periodic reporting, and recordkeeping. New
collections include ECCS evaluation model reporting and monitoring
under 10 CFR50.46a, QMS change approvals under 10 CFR50.54(a)(5),
50.55(f)(5), and 53.1565, and EP framework transition requests under 10
CFR50.160. The proposed rule also reduces burden by eliminating
duplicative requirements, such as prescriptive EP content in CP
applications and ESP renewal provisions, and by simplifying seismic/
geologic siting criteria under 10 CFR part 100. Overall, the proposed
rule results in a net increase in burden due to additional technical
analyses and documentation required for voluntary alternatives,
balanced by targeted reductions from modernization and streamlining
initiatives.
This proposed action includes amendments to several requirements in
10 CFR 50.55a and parts 50, 52, 53, 54, 71, and 100. The proposed
changes to 10 CFR parts 2 and 51 do not contain any new or amended
collections of information subject to the Paperwork Reduction Act of
1995.
The NRC is seeking public comment on the potential impact of the
information collections contained in this proposed rule and on the
following issues:
1. Is the proposed information collection necessary for the proper
performance of the functions of the NRC, including whether the
information will have practical utility? Please explain your response.
2. Is the estimate of the burden of the proposed information
collection accurate? Please explain your response.
3. Is there a way to enhance the quality, utility, and clarity of
the information to be collected? Please explain your response.
4. How can the burden of the proposed information collection on
respondents be minimized, including the use of automated collection
techniques or other forms of information technology?
A copy of the Office of Management and Budget (OMB) clearance
package and proposed rule are available in the ``Availability of
Documents'' section of this document or may be viewed free of charge by
contacting the NRC's Public Document Room reference staff at 1-800-397-
4209, at 301-415-4737, or by email to [email protected].
You may obtain information and comment on submissions related to
the OMB clearance package by searching on https://www.regulations.gov
under Docket ID NRC-2025-0975.
You may submit comments on any aspect of these proposed information
collections, including suggestions for reducing the burden and on the
above issues, by the following method:
Federal rulemaking website: Go to https://www.regulations.gov and search for Docket ID NRC-2025-0975.
Submit comments by August 17, 2026.
Public Protection Notification
The NRC may not conduct or sponsor, and a person is not required to
respond to, a collection of information unless the document requesting
or requiring the collection displays a currently valid OMB control
number.
XLV. Executive Orders
The following are Executive orders that are related to this
proposed rule:
A. Executive Order 12866: Regulatory Planning and Review (as Amended by
Executive Order 14215, Ensuring Accountability for All Agencies)
The Office of Information and Regulatory Affairs (OIRA) has
determined that this proposed rule is an economically significant
regulatory action under section 3(f) of E.O. 12866. Accordingly, NRC
submitted this proposed rule to OIRA for review. NRC is required to
conduct an economic analysis in accordance with section 6(a)(3)(B) of
E.O. 12866. More can be found in section XXXIX, ``Regulatory
Analysis,'' of this document.
B. Executive Order 14154: Unleashing American Energy
NRC has examined this proposed rule document and has determined
that it is consistent with the policies and directives outlined in E.O.
14154.
C. Executive Order 14192: Unleashing Prosperity Through Deregulation
This action is tentatively determined to be a deregulatory action
as defined by E.O. 14192. Details on the estimated costs of this
proposed rule document can be found in section XXXIX, ``Regulatory
Analysis,'' of this document.
D. Executive Order 14267: Reducing Anti-Competitive Regulatory Barriers
Executive Order 14267 requires the NRC to identify anti-competitive
regulations for recission or modification.
The NRC identified 10 CFR 50.34, and related portions of 10 CFR
part 52, because these regulations could create a barrier to market
participation by using language that may not be clear on the minimum
level of information needed to be submitted for a construction permit
and uses overly prescriptive language that is not technology inclusive.
The proposed recission/modification of the regulations would support
the objectives of E.O. 14267 by removing regulatory requirements that
could create unnecessary barriers to entry for new market entrants. See
section XVIII, ``Discussion--Updates to Construction Permit
Requirements and Related Licenses,'' of this document for more
information.
The NRC identified 10 CFR 50.54 because this regulation has an
anticompetitive effect by raising barriers to entry. It also serves an
important regulatory goal as many portions of this regulation are
necessary for reasonable assurance of adequate protection. The proposed
modifications of the regulation would support the objectives of E.O.
14267 by adding alternatives to enhance entry for new market entrants.
The proposed amendments to 10 CFR 50.54 would be conforming changes to
reflect the addition of performance-based QA criteria in the proposed
appendix T to 10 CFR part 50 as an alternative to appendix B to 10 CFR
part 50 for eligible applicants. See section XVI, Discussion--
Incorporation of Streamlined Quality Assurance Criteria for Nuclear
Power Plants and Fuel Reprocessing Plants,'' of this document for more
information.
The NRC identified 10 CFR 50.160 because this regulation creates a
barrier to market participation by limiting use of performance-based
emergency preparedness standards to small modular reactors less than
1000 MWt, non-light water reactors, and other new technologies. The
proposed modification of the regulation would support the objectives of
E.O. 14267 by making performance-based regulations available to all new
market entrants. See section XXIV, ``Discussion--Revision of the
Emergency Preparedness Regulations for Nuclear Power
[[Page 44650]]
Reactors,'' of this document for more information.
The NRC identified 10 CFR 100.3 because this regulation creates a
barrier to market participation by limiting available potential
facility sites, especially for designs that do not require low
population zones beyond site boundaries. The proposed recission/
modification of the regulation would support the objectives of E.O.
14267 by removing regulatory requirements that could create unnecessary
barriers to entry for new market entrants. See section XXXIV,
``Discussion--Enhancing Flexibility of Reactor Site Criteria,'' of this
document for more information.
XLVI. Voluntary Consensus Standards
The National Technology Transfer and Advancement Act of 1995,
Public Law 104-113, requires that Federal agencies use technical
standards that are developed or adopted by voluntary consensus
standards bodies unless the use of such a standard is inconsistent with
applicable law or otherwise impractical. In this proposed rule, the NRC
would revise the regulations associated with the usage of increased
enrichment of conventional and ATF designs for LWRs in 10 CFR parts 50,
51, 52, and 71. This action would not constitute the establishment of a
standard that contains generally applicable requirements.
XLVII. Availability of Guidance
The NRC is issuing for comment 24 draft guidance documents to
support the implementation of the proposed requirements in this
rulemaking. You may obtain information and comment submissions related
to the draft guidance by searching on http://www.regulations.gov under
Docket ID NRC-2025-0975. You may submit comments on these draft
guidance documents by the methods outlined in the ADDRESSES section of
this document.
1. The DG-1261, Revision 1, ``Measuring Breakaway Oxidation
Behavior,'' would be a new regulatory guide.
2. The DG-1262, Revision 1, ``Determining Post-Quench Ductility,''
would be a new regulatory guide.
3. The DG-1263, Revision 1, ``Establishing Analytical Limits for
Zirconium-Based Alloy Cladding,'' would be a new regulatory guide.
4. The DG-1425, ``Alternative Radiological Source Terms for
Evaluating Design-Basis Accidents at Nuclear Power Reactors,'' would be
Revision 2 to the existing RG 1.183.
5. The DG-1426, ``An Approach for a Risk-Informed Evaluation
Process Supporting Alternative Acceptance Criteria for Emergency Core
Cooling Systems for Light-Water Reactors,'' would be a new regulatory
guide.
6. The DG-1428, ``Plant-Specific Applicability of the Transition
Break Size,'' would be a new regulatory guide.
7. The DG-1430, ``Performance-Based Emergency Preparedness,'' would
be Revision 1 to the existing RG 1.242.
8. The DG-1434, ``Addressing the Consequences of Fuel Dispersal in
Light-Water Reactor Loss-of-Coolant Accidents,'' would be a new
regulatory guide.
9. The DG-1454, ``Implementation of Determinate and Data-Backed
Thresholds for Reactor Safety Assessments,'' would be a new regulatory
guide.
10. The DG-1456, ``Emergency Response Planning and Preparedness for
Nuclear Power Reactors,'' would be Revision 8 to the existing RG 1.101.
11. The DG-1457, ``Guidance on Making Changes to Emergency Plans
for Nuclear Power Reactors,'' would be Revision 2 to the existing RG
1.219.
12. The DG-1460, ``Applications for Nuclear Power Plants,'' would
be Revision 2 to the existing RG 1.206.
13. The DG-1461, ``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,'' would be
Revision 1 to the existing RG 1.237.
14. The DG-1462, ``A Performance-Based Approach to Define the Site-
Specific Earthquake Ground Motion,'' would be Revision 1 to the
existing RG 1.208.
15. The DG-1463, ``Meteorological Monitoring Programs for Nuclear
Power Plants,'' would be Revision 2 to the existing RG 1.23.
16. The DG-1464, ``Guidance for Content of Applications Under 10
CFR 50.220 and 52.220 Proposing Risk-Informed and Performance-Based
Alternative Acceptance Criteria,'' would be a new regulatory guide.
17. The DG-1465, ``Guidance for a Technology Inclusive Content of
Application Methodology to Inform the Licensing Basis and Content of
Applications for Licenses, Certifications, and Approvals for Non-Light-
Water Reactors,'' would be Revision 1 to the existing RG 1.253.
18. The DG-1466, ``Guidance for Implementation of 10 CFR 50.59,
`Changes, Tests, and Experiments,' '' would be Revision 4 to the
existing RG 1.187.
19. The DG-1467, ``Assuring the Availability of Funds for
Decommissioning Nuclear Reactors,'' would be Revision 3 to the existing
RG 1.159.
20. The DG-1468, ``Guidance for Implementation of 10 CFR 50.221,
`Credibility Requirements for Modeling and Simulation,' '' would be a
new regulatory guide.
21. The DG-4036, ``Graded Approach to Site Characterization for New
Reactor Applications,'' would be a new regulatory guide.
22. The DG-4037, ``Preparation of Environmental Reports for Nuclear
Power Stations,'' would be Revision 5 to the existing RG 4.2.
23. LR-ISG-2026-01, ``Updated Review Criteria for License
Renewal,'' would be new interim staff guidance.
24. NUREG-0800, Chapter 14, Section 14.3, ``Inspections, Tests,
Analyses, and Acceptance Criteria,'' would be a revision to the
existing standard review plan (issued March 2007).
Draft regulatory guide updates beyond this rulemaking are included
in DG-1425. These additional updates to DG-1425 include the following:
Updated steady-state release fractions for accidents other
than the maximum hypothetical accident (loss-of-coolant accident),
based on American National Standards Institute/American Nuclear Society
5.4, ``Method for Calculating the Fractional Release of Volatile
Fission Products from Oxide Fuel,'' May 2011, extending the
applicability to higher burnups and increased enrichments;
Additional guidance for modeling BWR main steam isolation
valve (MSIV) leakage and fission product removal by suppression pool
scrubbing;
Guidance for crediting holdup and retention of MSIV
leakage within the main steamlines and condenser for BWRs; and
Guidance for use of best estimate plus uncertainty
approaches to determining inputs to radiological models.
XLVIII. 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.
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BILLING CODE 7590-01-C
The NRC prepared an unofficial redline strikeout version of the
proposed changes to regulatory text that is intended to help the
readeridentifythe changes. The NRC is providing the unofficial redline
as a reader tool only, and this document is listed in the
``Availability of Documents'' section of this document. Comments on the
rule text should be made in this proposed rule.
The NRC may post materials related to this document, including
public comments, on the Federal rulemaking website at https://www.regulations.gov under Docket ID NRC-2025-0975. In addition, the
Federal rulemaking website allows members of the public to receive
alerts when changes or additions occur in a docket folder. To
subscribe: (1) navigate to the docket folder (NRC-2025-0975); (2) click
the ``Subscribe'' link; and (3) enter an email address and click on the
``Subscribe'' link.
List of Subjects
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 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 52
Administrative practice and procedure, Antitrust, Combined license,
Early site permit, Emergency planning, Fees, Inspection, Issue
finality, Limited work authorization, Manufacturing license, Nuclear
power plants and reactors, Probabilistic risk assessment, Prototype,
Reactor siting criteria, Redress of site, Penalties, Reporting and
recordkeeping requirements, Standard design, Standard design
certification.
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 54
Administrative practice and procedure, Age-related degradation,
Backfitting, Classified information,
[[Page 44668]]
Criminal penalties, Environmental protection, Nuclear power plants and
reactors, Penalties, Radiation protection, Reporting and recordkeeping
requirements.
10 CFR Part 71
Criminal penalties, Hazardous materials transportation,
Intergovernmental relations, Nuclear materials, Packaging and
containers, Penalties, Radioactive materials, Reporting and
recordkeeping requirements.
10 CFR Part 100
Nuclear power plants and reactors, Radiation protection, Reactor
siting criteria, Reporting and recordkeeping requirements.
For the reasons set out in the preamble and under the authority of
the Atomic Energy Act of 1954, as amended; the Energy Reorganization
Act of 1974, as amended; and 5 U.S.C. 552 and 553, the NRC is proposing
to amend 10 CFR parts 2, 50, 51, 52, 53, 54, 71, and 100:
PART 2--AGENCY RULES OF PRACTICE AND PROCEDURE
0
1. 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. 2.109 [Amended]
0
2. In Sec. 2.109, remove and reserve paragraph (c).
PART 50--DOMESTIC LICENSING OF PRODUCTION AND UTILIZATION
FACILITIES
0
3. The authority citation for part 50 continues 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; ADVANCE Act of 2024, sec. 301 (42 U.S.C. 2133 note).
0
4. In Sec. 50.2, add in alphabetical order the definitions ``Beyond
design basis event'' and ``Design basis event'' to read as follows:
Sec. 50.2 Definitions.
* * * * *
Beyond design basis event means an initiating event with a
frequency below the threshold for a design basis event, but greater
than or equal to a threshold for credibility that provides an
appropriate level of safety and is deemed acceptable by the NRC, or an
event (other than a design basis event) that is specifically required
by regulation to be considered in the licensing basis. This definition,
along with the corresponding definition of ``design basis event,''
applies to applications submitted on or after [DATE 180 DAYS AFTER THE
EFFECTIVE DATE OF FINAL RULE] for construction permits and operating
licenses under this part, as well as for early site permits, standard
design certifications, combined licenses, standard design approvals,
and manufacturing licenses under part 52 of this chapter. Applicants
for or holders of licenses, permits, standard design certifications, or
standard design approvals under this part or part 52 of this chapter
who are not subject to this requirement may propose to adopt this
definition and the corresponding definition of ``design basis event''
by submitting a request for NRC approval in an application, through a
license amendment request under Sec. 50.90, or through the applicable
change process under part 52 of this chapter.
* * * * *
Design basis event means an initiating event with a frequency
greater than or equal to a determinate, data-backed threshold that
provides an appropriate level of safety and is deemed acceptable by the
NRC. This definition, along with the corresponding definition for
``beyond design basis event,'' applies to applications submitted on or
after [DATE 180 DAYS AFTER THE EFFECTIVE DATE OF FINAL RULE] for
construction permits and operating licenses under this part, as well as
for early site permits, standard design certifications, combined
licenses, standard design approvals, and manufacturing licenses under
part 52 of this chapter. Applicants for or holders of licenses,
permits, standard design certifications, or standard design approvals
under this part or part 52 of this chapter who are not subject to this
requirement may propose to adopt this definition and the corresponding
definition of ``beyond design basis event'' by submitting a request for
NRC approval in an application, through a license amendment request
under Sec. 50.90, or through the applicable change process under part
52 of this chapter.
* * * * *
0
5. In Sec. 50.4, add paragraph (b)(7)(iii) to read as follows:
Sec. 50.4 Written Communications.
* * * * *
(b) * * *
(7) * * *
(iii) A change to the Safety Analysis Report quality management
system under Sec. 50.54(a)(5) or Sec. 50.55(f)(5), or a change to a
licensee's NRC-accepted quality management system topical report under
Sec. 50.54(a)(5) or Sec. 50.55(f)(5), 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 signed original must be sent.
* * * * *
0
6. In Sec. 50.10, revise paragraphs (a) and (c), and add paragraph (h)
to read as follows:
Sec. 50.10 License required; limited work authorization.
(a) Definitions. As used in this section, construction means the
activities in paragraph (a)(1) of this section.
(1) Activities constituting construction are 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 may impact
the required functions of:
(i) Safety-related structures, systems, or components (SSCs) of a
facility, as defined in Sec. 50.2;
(ii) SSCs that perform safety-significant functions; and
(iii) SSCs necessary to comply with part 73 of this chapter.
(2) With respect to production or utilization facilities, other
than testing facilities and nuclear power plants, required to be
licensed under section 104a. or section 104c. of the Act, construction
does not include the erection of buildings which will be used for
activities other than operation of a facility and which may also be
used to house a facility (e.g., the construction of a college
laboratory building with space for installation of a training reactor).
[[Page 44669]]
(3) Any activities that are determined to be outside the scope of
those defined in Sec. 50.10(a)(1) and that are undertaken by an
applicant or on its behalf are entirely at the risk of the applicant
and has no bearing on the issuance of a license with respect to the
requirements of the Act, and rules, regulations, or orders issued under
the Act.
* * * * *
(c) Requirement for construction permit, early site permit
authorizing limited work authorization activities, combined license, or
limited work authorization. Except as provided in paragraph (h) of this
section, no person may begin the construction of a production or
utilization facility on a site on which the facility is to be operated
until that person has been issued either a construction permit under
this part, a combined license under part 52 of this chapter, an early
site permit authorizing the activities under paragraph (d) of this
section, or a limited work authorization under paragraph (d) of this
section.
* * * * *
(h) Issuance of general license. A general license is hereby issued
to an applicant for a construction permit or combined license for a
utilization facility under 10 CFR part 50 or 52 for construction
activities on a site that is specified in the application, subject to
the following conditions:
(1) The applicant has submitted and the Commission has docketed a
construction permit or combined license application for a utilization
facility under 10 CFR part 50 or 52 that meets the following criteria:
(i) The application references a reactor design for which the
Commission issued an operating license or issued a combined license and
made the finding under Sec. 52.103(g) of this chapter and for which
the Commission afforded generic finality under Sec. 50.57(d) or Sec.
52.97(d);
(ii) The operating license or combined license described in
paragraph (h)(1)(i) of this section met the criteria for a categorical
exclusion or resulted in a finding of no significant impact from an
environmental assessment in accordance with part 51 of this chapter;
(iii) The application utilizing the general license includes a plan
for redress of any adverse environmental impact from conduct of
activities under the general license should such redress be necessary;
and
(iv) The application must contain information demonstrating that
the site characteristics are bounded by the site parameters postulated
for the approval of generic finality.
(2) The applicant may perform construction only upon notification
to the NRC Director of NRR using instructions in Sec. 50.4 before the
start of construction. The notice must state that all applicable
permits, licenses, approvals, and other entitlements in connection with
the proposed action have been obtained. The notice may be in the form
of a letter, but must contain the applicant's name, address, and the
name and means of contacting a person responsible for providing
additional information concerning construction under this general
license.
(3) All applicable Federal environmental consultations have been
completed.
(4) The general license authorizes construction of those generic
aspects of the design of the commercial nuclear plant for which the
Commission afforded generic finality and does not authorize
installation of the reactor vessel, the reactor coolant system, or
associated reactivity control and heat removal systems;
(5) The applicant must allow for NRC inspections that the
Commission deems necessary related to activities performed under the
general license.
(6) Any activities undertaken by the applicant or on its behalf
under the general license are entirely at the risk of the applicant and
have no bearing on the issuance of a license with respect to the
requirements of the Act, and rules, regulations, or orders issued under
the Act.
0
7. In Sec. 50.33, remove footnotes 1 and 2 and revise paragraph (g) to
read as follows:
Sec. 50.33 Contents of applications; general information.
* * * * *
(g)(1) If the application is for an operating license or combined
license for a nuclear power reactor, or if the application is for an
early site permit and contains plans for coping with emergencies under
Sec. 52.17(b)(2)(ii) of this chapter, the applicant must coordinate
radiological emergency preparedness activities with offsite
organizations with responsibilities for coping with emergencies
including State, local, and Tribal governmental agencies, as
applicable. Specifically, the applicant must ensure that these response
organizations are aware of the potential radiological consequences of
the facility and have been consulted on appropriate protective measures
including the extent of any emergency planning zone (EPZ) for
implementing predetermined, prompt protective measures. The application
must include information that describes the extent of the applicant's
interaction with these response organizations. If the application is
for an early site permit that, under Sec. 52.17(b)(2)(i) of this
chapter, proposes major features of the emergency plans describing the
EPZ, then the description of the EPZ must meet the requirements of this
paragraph (g)(1). Generally, the plume exposure pathway EPZ for nuclear
power reactors shall consist of an area about 2 to 10 miles (3.2 to 16
km) in radius. For reactors with an authorized power level less than
300 MW thermal, the plume exposure pathway EPZ may be established at
the site boundary. The need for and size of the EPZ may also be
determined on a case-by-case basis as described in Sec. 50.33(g)(2).
The exact size and configuration of the EPZ surrounding a particular
nuclear power reactor 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. Emergency plans must describe
such actions as are appropriate to avoid or reduce dose within and
beyond the EPZ or site boundary and to protect the ingestion pathway.
(2) For a case-by-case EPZ determination, the applicant or licensee
must submit an analysis used to determine whether the criteria in Sec.
50.33(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) Dose to an individual is projected to exceed 1 rem (10 mSv)
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) [Reserved]
* * * * *
0
8. In Sec. 50.34,
0
a. Remove the text ``LOCA's'' wherever it appears, and add, in its
place, the text ``LOCAs'';
0
b. Remove the text ``PWR's'' wherever it appears, and add, in its
place, the text ``PWRs'';
0
c. Remove the text ``BWR's'' wherever it appears, and add, in its
place, the text ``BWRs'';
0
d. Revise paragraphs (a)(1) introductory text, (a)(1)(ii)(D), (a)(7),
(b)(6)(ii), (b)(6)(v), (b)(10), (b)(11), and footnotes 1, 3, 4, and 6;
0
e. Remove the last sentence of paragraphs (a)(4) and (b)(4);
[[Page 44670]]
0
f. Remove and reserve paragraph (a)(10);
0
g. In paragraph (a)(12), remove the phrase ``On or after January 10,
1997, stationary power'' and add in its place ``Power''.
0
h. Add paragraph (b)(14); and
0
i. Revise paragraph (f)(3)(ii).
The revisions and additions are to read as follows:
Sec. 50.34 Contents of applications; technical information.
(a) Preliminary safety analysis report. Each application for a
construction permit shall include a preliminary safety analysis report.
The minimum information \[1]\ to be included shall consist of the
following:
(1) Power reactor applicants for a construction permit shall comply
with paragraph (a)(1)(ii) of this section. All other applicants for a
construction permit shall comply with paragraph (a)(1)(i) of this
section.
* * * * *
(ii) * * *
(D) The safety features that are to be engineered into the facility
and those barriers that must be breached as a result of an accident
before a release of radioactive material to the environment can occur.
Special attention must be directed to plant design features intended to
mitigate the radiological consequences of accidents. In performing this
assessment, an applicant shall assume a fission product release \[3]\
assuming that the facility is operated at the ultimate power level
contemplated. The applicant shall perform an evaluation and analysis of
the postulated fission product release, using the expected demonstrable
leakage rates from potential flow paths and any fission product cleanup
systems intended to mitigate the consequences of the accidents,
together with applicable site characteristics, including site
meteorology, to evaluate the offsite radiological consequences. Site
characteristics must comply with part 100 of this chapter. The
evaluation must determine that:
(1) 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 \[4]\ (0.25 Sv) total effective dose equivalent
(TEDE).
(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
(0.25 Sv) TEDE;
* * * * *
(7) A description of the quality assurance program or a quality
management system to be applied to the design, fabrication,
construction, and testing of the structures, systems, and components of
the facility. Appendix B to this part, ``Quality Assurance Criteria for
Nuclear Power Plants and Fuel Reprocessing Plants,'' sets forth the
requirements for quality assurance programs for nuclear power plants
and fuel reprocessing plants. Appendix T to this part, ``Streamlined
Quality Assurance Criteria for Nuclear Power Plants and Fuel
Reprocessing Plants,'' sets forth streamlined requirements for quality
assurance programs for nuclear power plants and fuel reprocessing
plants that an eligible construction permit applicant may voluntarily
use as an alternative to appendix B to this part. The description of
the quality assurance program for a nuclear power plant or a fuel
reprocessing plant shall include a discussion of how the applicable
requirements of appendix B will be satisfied or for eligible
construction permit applications, the quality management system for a
nuclear power plant or fuel reprocessing plant shall include
discussions of how the applicable requirements of appendix T will be
satisfied.
* * * * *
(10) [Reserved]
* * * * *
(b) * * *
(6) * * *
(ii) Managerial and administrative controls to be used to assure
safe operation. Appendix B to this part, ``Quality Assurance Criteria
for Nuclear Power Plants and Fuel Reprocessing Plants,'' sets forth the
requirements for such controls for nuclear power plants and fuel
reprocessing plants. Appendix T to this part, ``Streamlined Quality
Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing
Plants,'' sets forth streamlined requirements for such controls for
nuclear power plants and fuel reprocessing plants that an eligible
operating license applicant may voluntarily use as an alternative to
appendix B to this part. The information on the controls to be used for
a nuclear power plant or a fuel reprocessing plant shall include a
discussion of how the applicable requirements of appendix B to this
part will be satisfied or, for eligible operating license applications,
the quality management system for a nuclear power plant or fuel
reprocessing plant shall include discussions of how the applicable
requirements of appendix T to this part will be satisfied.
* * * * *
(v) Plans for coping with emergencies. The applicant must provide
the offsite response organizations that are expected to respond in an
emergency with the opportunity to provide input on the emergency plan
before submitting it to the NRC. The application must contain any input
on the emergency plan received from offsite response organizations.
* * * * *
(10) Power reactor applicants who apply for an operating license,
as partial conformance to General Design Criterion 2 of appendix A to
this part, shall comply with the earthquake engineering criteria of
appendix S to this part.
(11) Power reactor applicants who apply for an operating license
shall provide a description and safety assessment of the site and of
the facility as in Sec. 50.34(a)(1)(ii).
* * * * *
(14) An applicant may include in its application a request for
generic finality, to generic aspects of the design under this part,
such that information in the application, if approved by the NRC, is
considered resolved in other proceedings where information approved for
generic finality is referenced. An application for an operating license
that requests generic finality must include applicable site parameters
postulated for the design, including the design-basis external hazard
levels for the relevant external hazards, and an analysis and
evaluation of the design in terms of those site parameters.
* * * * *
(f) * * *
(3) * * *
(ii) Ensure that the quality assurance (QA) list required by
Criterion II of appendix B and appendix T of 10 CFR part 50 includes
all structures, systems, and components important to safety. (I.F.1)
* * * * *
\[1]\ This paragraph specifies the minimum required technical
information to be included in a preliminary safety analysis report,
while Sec. Sec. 50.35(a)(1) through (4), 50.40, and 50.50 specify
required findings for issuance of a construction permit. The level
of detail provided in a preliminary safety analysis report to
satisfy the minimum technical information requirements in paragraph
(a) of this section will be deemed sufficient if the provided
information enables the Commission to make the findings for issuance
of a construction permit in Sec. Sec. 50.35(a)(1) through (4),
50.40, and 50.50. The applicant may provide information
[[Page 44671]]
required by this paragraph in the form of a discussion, with
specific references, of similarities to and differences from,
facilities of similar design for which applications have previously
been filed with the Commission.
* * * * *
\[3]\ The fission product release assumed for this evaluation
should be based upon a major accident, hypothesized for purposes of
site analysis or postulated from considerations of possible
accidental events to bound a broad range of design basis accidents.
Such accidents have generally been assumed to result in substantial
meltdown of the core with subsequent release of appreciable
quantities of fission products.
\[4]\ The use of 25 rem (0.25 Sv) 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 such designs provide
assurance of low risk of public exposure to radiation, in the event
of an accident.
* * * * *
\[6]\ The fission product release assumed for these calculations
should be based upon a major accident, hypothesized for purposes of
site analysis or postulated from considerations of possible
accidental events to bound a broad range of design basis accidents.
Such accidents have generally been assumed to result in substantial
meltdown of the core with subsequent release of appreciable
quantities of fission products.
0
9. Revise and republish Sec. 50.46 to read as follows:
Sec. 50.46 Acceptance criteria for emergency core cooling systems for
light-water nuclear power reactors.
(a) Each boiling or pressurized light-water nuclear power reactor
must be provided with an emergency core cooling system (ECCS) that is
designed under requirements described in this paragraph.
(1) The requirements of this section or Sec. 50.46a must be
satisfied for:
(i) Holders of an operating license under this part authorized to
operate on December 31, 2015;
(ii) Holders of an operating license under this part authorized to
operate after December 31, 2015, and whose reactor design is
demonstrated under Sec. 50.46a(c)(2) to be similar to the designs of
reactors authorized to operate under this part on December 31, 2015;
(iii) Holders of a construction permit issued under this part whose
reactor design is demonstrated under Sec. 50.46a(c)(2) to be similar
to the design of reactors authorized to operate under this part on
December 31, 2015;
(iv) Holders of a combined license, standard design approval, or
manufacturing license under part 52 of this chapter whose reactor
design is demonstrated under Sec. 50.46a(c)(2) to be similar to the
designs of reactors authorized to operate under this part on December
31, 2015;
(v) Applicants for a construction permit or operating license under
this part whose reactor design is demonstrated under Sec. 50.46a(c)(2)
to be similar to the designs of reactors authorized to operate under
this part on December 31, 2015; and
(vi) Applicants for a combined license, standard design approval,
manufacturing license, or standard design certification (including such
applicants after NRC issuance of a final standard design certification
rule) under part 52 of this chapter whose reactor design is
demonstrated under Sec. 50.46a(c)(2) of this section to be similar to
the designs of reactors authorized to operate under this part on
December 31, 2015.
(2) The requirements of this section must be satisfied for:
(i) Holders of an operating license under this part that were
authorized to operate after December 31, 2015, and whose reactor design
is not demonstrated under Sec. 50.46a(c)(2) to be similar to the
designs of reactors authorized to operate under this part on December
31, 2015;
(ii) Holders of a construction permit issued under this part whose
reactor design is not demonstrated under Sec. 50.46a(c)(2) to be
similar to the design of reactors authorized to operate under this part
on December 31, 2015;
(iii) Holders of a combined license, standard design approval, or
manufacturing license under part 52 of this chapter whose reactor
design is not demonstrated under Sec. 50.46a(c)(2) to be similar to
the designs of reactors authorized to operate under this part on
December 31, 2015;
(iv) Applicants for a construction permit or operating license
under this part whose reactor design is not demonstrated under Sec.
50.46a(c)(2) to be similar to the designs of reactors authorized to
operate under this part on December 31, 2015; and
(v) Applicants for a combined license, standard design approval,
manufacturing license, or standard design certification (including such
applicants after NRC issuance of a final standard design certification
rule) under part 52 of this chapter whose reactor design is not
demonstrated under Sec. 50.46a(c)(2) to be similar to the designs of
reactors authorized to operate under this part on December 31, 2015.
(3) (i) The ECCS system must be designed so that its calculated
cooling performance following postulated loss-of-coolant accidents
(LOCAs) conforms to the criteria set forth in paragraph (b) of this
section or Sec. 50.46a(f). ECCS cooling performance must be calculated
in accordance with an acceptable evaluation model and must be
calculated for a number of postulated LOCAs of different sizes,
locations, and other properties sufficient to provide assurance that
the most severe postulated LOCAs are calculated. Except as provided in
paragraph (a)(3)(ii) of this section, the evaluation model must include
sufficient supporting justification to show that the analytical
technique realistically describes the behavior of the reactor system
during a LOCA. Comparisons to applicable experimental data must be made
and uncertainties in the analysis method and inputs must be identified
and assessed so that the uncertainty in the calculated results can be
estimated. This uncertainty must be accounted for, so that, when the
calculated ECCS cooling performance is compared to the criteria set
forth in paragraph (b) of this section or Sec. 50.46a(f), as
applicable, there is a high level of probability that the criteria
would not be exceeded. Section II, ``Required Documentation,'' of
appendix K to this part, sets forth the documentation requirements for
each evaluation model. This section does not apply to a nuclear power
reactor facility for which the certifications required under Sec.
50.82(a)(1) or Sec. 52.110(a) of this chapter have been submitted.
(ii) Alternatively, an ECCS evaluation model may be developed in
conformance with the required and acceptable features of appendix K to
this part, ``ECCS Evaluation Models.''
(4) The Director of Nuclear Reactor Regulation may impose
restrictions on reactor operation if it is found that the submitted
evaluations of ECCS cooling performance are not consistent with
paragraphs (a)(3)(i) and (ii) of this section.
(5) (i) Each applicant for or holder of an operating license or
construction permit issued under this part, applicant for a standard
design certification under part 52 of this chapter (including an
applicant after the Commission has adopted a final design certification
regulation), or an applicant for or holder of a standard design
approval, a combined license, or a manufacturing license issued under
part 52 of this chapter, must estimate the effect of any change to or
error in an acceptable evaluation model or in the application of such a
model to determine if the change or error is significant. For this
purpose, a significant change or error is one which results in a
calculated peak
[[Page 44672]]
fuel cladding temperature different by more than 50 [deg]F from the
temperature calculated for the limiting transient using the last
acceptable model, or is a cumulation of changes and errors such that
the sum of the absolute magnitudes of the respective temperature
changes is greater than 50 [deg]F.
(ii) For each change to or error discovered in an acceptable
evaluation model or in the application of such a model that affects the
temperature calculation, the applicant or holder of a construction
permit, operating license, combined license, or manufacturing license
must report the nature of the change or error and its estimated effect
on the limiting ECCS analysis to the Commission at least annually as
specified in Sec. 50.4 or Sec. 52.3 of this chapter, as applicable.
If the change or error is significant, the applicant or licensee must
provide this report within 30 days and include with the report a
proposed schedule for providing a reanalysis or taking other action as
may be needed to show compliance with Sec. 50.46 requirements. This
schedule may be developed using an integrated scheduling system
previously approved for the facility by the NRC. For those facilities
not using an NRC-approved integrated scheduling system, a schedule will
be established by the NRC within 60 days of receipt of the proposed
schedule. Any change or error correction that results in a calculated
ECCS performance that does not conform to the criteria set forth in
paragraph (b) of this section is a reportable event as described in
Sec. Sec. 50.55(e), 50.72, and 50.73. The affected applicant or
licensee must propose immediate steps to demonstrate compliance or
bring plant design or operation into compliance with Sec. 50.46
requirements.
(iii) For each change to or error discovered in an acceptable
evaluation model or in the application of such a model that affects the
temperature calculation, the applicant or holder of a standard design
approval or the applicant for a standard design certification
(including an applicant after the Commission has adopted a final design
certification rule) must report the nature of the change or error and
its estimated effect on the limiting ECCS analysis to the Commission
and to any applicant or licensee referencing the standard design
approval or standard design certification at least annually as
specified in Sec. 52.3 of this chapter. If the change or error is
significant, the applicant or holder of the standard design approval or
the applicant for the standard design certification must provide this
report within 30 days and include with the report a proposed schedule
for providing a reanalysis or taking other action as may be needed to
show compliance with Sec. 50.46 requirements. The affected applicant
or holder must propose immediate steps to demonstrate compliance or
bring plant design into compliance with Sec. 50.46 requirements.
(iv) For entities that are approved to use Sec. 50.46a(f) instead
of paragraph (b) of this section, changes or errors discovered in an
acceptable evaluation model should be reported in accordance with Sec.
50.46a(j)(1) and (2).
(b) The ECCS system of each boiling or pressurized light-water
nuclear power reactor fueled with uranium oxide pellets within
cylindrical zircaloy or ZIRLO cladding must be designed so that its
calculated cooling performance following postulated LOCAs conforms to
the following criteria:
(1) Peak cladding temperature. The calculated maximum fuel element
cladding temperature must not exceed 2200 [deg]F.
(2) Maximum cladding oxidation. The calculated total oxidation of
the cladding may nowhere exceed 0.17 times the total cladding thickness
before oxidation. As used in this subparagraph total oxidation means
the total thickness of cladding metal that would be locally converted
to oxide if all the oxygen absorbed by and reacted with the cladding
locally were converted to stoichiometric zirconium dioxide. If cladding
rupture is calculated to occur, the inside surfaces of the cladding
must be included in the oxidation, beginning at the calculated time of
rupture. Cladding thickness before oxidation means the radial distance
from inside to outside the cladding, after any calculated rupture or
swelling has occurred but before significant oxidation. Where the
calculated conditions of transient pressure and temperature lead to a
prediction of cladding swelling, with or without cladding rupture, the
unoxidized cladding thickness must be defined as the cladding cross-
sectional area, taken at a horizontal plane at the elevation of the
rupture, if it occurs, or at the elevation of the highest cladding
temperature if no rupture is calculated to occur, divided by the
average circumference at that elevation. For ruptured cladding, the
circumference does not include the rupture opening.
(3) Maximum hydrogen generation. The calculated total amount of
hydrogen generated from the chemical reaction of the cladding with
water or steam must not exceed 0.01 times the hypothetical amount that
would be generated if all of the metal in the cladding cylinders
surrounding the fuel, excluding the cladding surrounding the plenum
volume, were to react.
(4) Coolable geometry. Calculated changes in core geometry must be
such that the core remains amenable to cooling.
(5) Long-term cooling. After any calculated successful initial
operation of the ECCS, the calculated core temperature must be
maintained at an acceptably low value and decay heat must be removed
for the extended period of time required by the long-lived
radioactivity remaining in the core.
(c) As used in this section:
(1) LOCAs are hypothetical accidents that would result from the
loss of reactor coolant, at a rate in excess of the capability of the
reactor coolant makeup system, from breaks in pipes in the reactor
coolant pressure boundary up to and including a break equivalent in
size to the double-ended rupture of the largest pipe in the reactor
coolant system.
(2) An evaluation model is the calculational framework for
evaluating the behavior of the reactor system during a postulated LOCA.
It includes one or more computer programs and all other information
necessary for application of the calculational framework to a specific
LOCA, such as mathematical models used, assumptions included in the
programs, procedure for treating the program input and output
information, specification of those portions of analysis not included
in computer programs, values of parameters, and all other information
necessary to specify the calculational procedure.
(d) The requirements of this section are in addition to any other
requirements applicable to ECCS set forth in this part. The criteria
set forth in paragraph (b) of this section, with cooling performance
calculated in accordance with an acceptable evaluation model, are in
implementation of the general requirements with respect to ECCS cooling
performance design set forth in this part, including, in particular,
criterion 35 of appendix A to this part.
0
10. Revise Sec. 50.46a to read as follows:
Sec. 50.46a Alternative acceptance criteria for emergency core
cooling systems for light-water nuclear power reactors.
(a) Definitions. For the purposes of this section:
(1) Changes enabled by this section means changes to the facility,
technical specifications, and procedures that satisfy the alternative
ECCS analysis requirements under this section but do
[[Page 44673]]
not satisfy the ECCS requirements under Sec. 50.46.
(2) Cladding means the material structure surrounding and
containing the fissile material and providing a barrier to prevent
fission product transport or release to the coolant.
(3) Crud means any foreign substance deposited on the surface of
fuel cladding.
(4) Entity means an applicant for or a holder of a construction
permit, operating license, combined license, standard design approval,
or manufacturing license, or an applicant for a standard design
certification (including such applicant after NRC issuance of a final
standard design certification rule).
(5) ECCS evaluation model means the calculational framework for
evaluating the behavior of the light-water reactor system (including
fuel) during a postulated loss-of-coolant accident (LOCA). It includes
one or more computer programs and all other information necessary for
application of the calculational framework to a specific LOCA, such as
mathematical models used, assumptions included in the programs,
procedure for treating the program input and output information,
specification of those portions of analysis not included in computer
programs, values of parameters, and all other information necessary to
specify the calculational procedure.
(6) Loss-of-coolant accidents (LOCAs) means the hypothetical
accidents that would result from the loss of reactor coolant, at a rate
in excess of the capability of the reactor coolant makeup system, from
breaks in pipes in the reactor coolant pressure boundary up to and
including a break equivalent in size to the double-ended rupture of the
largest pipe in the reactor coolant system. LOCAs involving breaks at
or below the transition break size are design basis accidents. LOCAs
involving breaks larger than the transition break size are beyond-
design-basis accidents.
(7) Operating configuration means those plant characteristics, such
as power level, equipment unavailability (including unavailability
caused by corrective and preventive maintenance), and equipment
capability that affect plant response to a LOCA.
(8) Quench means the rapid cooling of the fuel cladding by liquid
coolant water.
(9) Transition break size (TBS) for reactors authorized to operate
under this part on December 31, 2015, is a break area equal to the
largest cross-sectional flow area of the reactor coolant pressure
boundary piping excluding the hot leg, cold leg, or crossover leg
piping for a pressurized water reactor; the larger cross-sectional flow
area of either the feedwater line or residual heat removal line inside
containment for a boiling water reactor; or a plant-specific
alternative break area. For reactors that are authorized to operate
under this part after December 31, 2015, and for light-water reactors
(LWRs) that are authorized to operate under part 52 of this chapter,
the TBS will be determined on a plant-specific basis.
(b) Applicability and scope.
(1) Those entities listed in subparagraphs (i) through (vi) of this
paragraph may apply under paragraph (c) of this section to use the
requirements of this section. This section does not apply to a nuclear
power reactor facility for which the certifications required under
Sec. 50.82(a)(1) or Sec. 52.110(a) of this chapter have been
submitted.
(i) Holders of an operating license under this part authorized to
operate on December 31, 2015;
(ii) Holders of an operating license under this part authorized to
operate after December 31, 2015, and whose reactor design is
demonstrated under paragraph (c)(2) of this section to be similar to
the designs of reactors authorized to operate under this part on
December 31, 2015;
(iii) Holders of a construction permit issued under this part whose
reactor design is demonstrated under paragraph (c)(2) of this section
to be similar to the design of reactors authorized to operate under
this part on December 31, 2015.
(iv) Holders of a combined license, standard design approval, or
manufacturing license under part 52 of this chapter whose reactor
design is demonstrated under paragraph (c)(2) of this section to be
similar to the designs of reactors authorized to operate under this
part on December 31, 2015.
(v) Applicants for a construction permit or operating license under
this part whose reactor design is demonstrated under paragraph (c)(2)
of this section to be similar to the designs of reactors authorized to
operate under this part on December 31, 2015.
(vi) Applicants for a combined license, standard design approval,
manufacturing license, or standard design certification (including such
applicants after NRC issuance of a final standard design certification
rule) under part 52 of this chapter whose reactor design is
demonstrated under paragraph (c)(2) of this section to be similar to
the designs of reactors authorized to operate under this part on
December 31, 2015.
(2) The requirements of this section are in addition to any other
requirements applicable to ECCS, with the exception of Sec. 50.46. The
criteria set forth in paragraph (e)(1) of this section, with cooling
performance calculated in accordance with an acceptable evaluation
model or analysis method under paragraphs (e)(2) and (3) of this
section, are in implementation of the general requirements with respect
to ECCS cooling performance design set forth in this part, including,
in particular, criterion 35 of appendix A to this part.
(3) A licensee must inspect, under Sec. 50.55a(g), for those
reactor coolant pressure boundary piping whose inner diameter is
greater than the TBS, an NRC-approved sampling of the similar metal
piping circumferential welds in a PWR and the circumferential welds in
a BWR that are classified as Category A welds before implementation of
this section and in every subsequent in-service inspection interval (as
defined in Sec. 50.55a(y)). The sampling must include those
circumferential welds with the highest failure potential. Credit may be
taken for welds inspected as part of established inspection programs
(e.g., risk-informed inservice inspection programs). The effect on the
TBS of any degradation identified during these inspections must be
evaluated.
(c) Application.
(1) An entity seeking to implement this section must submit an
application under Sec. 50.34, 50.90, or part 52 of this chapter, as
applicable, that contains the following information:
(i) A written evaluation demonstrating applicability of the TBS to
the entity's facility or a proposed alternative TBS and a justification
that the proposed TBS is consistent with the technical basis for this
section. The effects of the initial plant changes proposed in the
application must be considered as part of this evaluation.
(ii) As applicable, an inspection report that details the results
of the inspection requirements in paragraph (b)(3) of this section and
the evaluation of the impact of these results on the TBS.
(iii) Identification of the acceptable analysis method(s) for
demonstrating compliance with the ECCS criteria in paragraph (e) of
this section.
(iv) A description of the risk-informed evaluation used to
demonstrate that the proposed changes to the facility meet the
requirements in paragraph (h) of this section.
(v) For an entity other than a design certification applicant or a
holder of a manufacturing license that wishes to make changes enabled
by this section without prior NRC review and approval, a process to be
used for evaluating the
[[Page 44674]]
acceptability of these changes, including:
(A) A description of the approach, methods, and decision-making
process to be used for evaluating compliance with the acceptance
criteria in paragraphs (h)(1), (2), and (3) of this section;
(B) A description of the probabilistic risk assessment (PRA) model
and/or non-PRA risk assessment methods to be used for demonstrating
compliance with paragraphs (h)(4) and (5) of this section; and
(C) A description of the approach, methods, and decision-making
process to be used to evaluate the continued applicability of the TBS
with the acceptance criteria used in the evaluation from paragraph
(c)(1)(i) of this section for plants authorized to operate under this
part on December 31, 2015, or from paragraph (c)(2) of this section for
entities other than those authorized to operate under this part on
December 31, 2015.
(vi) A description of non-safety equipment that is credited for
demonstrating compliance with the ECCS acceptance criteria in paragraph
(e) of this section.
(vii) A written evaluation demonstrating how the leak detection
program in place at the facility satisfies the criteria in paragraph
(d)(2) of this section.
(2) Each applicant, other than one authorized to operate under this
part on December 31, 2015, seeking to implement the requirements of
this section must submit, in addition to the information required by
paragraphs (c)(1)(ii) through (vii) of this section, an analysis
demonstrating why the proposed reactor design is similar to the designs
of reactors authorized to operate under this part on December 31, 2015,
such that the provisions of this section may properly apply. The
analysis must also include a proposed TBS and a justification that the
proposed TBS is consistent with the technical basis for this section.
The effects of the initial plant changes proposed in the application
must be considered as part of this evaluation.
(3) The NRC may approve an application to use this section if:
(i) The evaluation submitted under paragraph (c)(1)(i) of this
section demonstrates the applicability of the TBS to the facility for
reactors authorized to operate under this part on December 31, 2015;
(ii) The method(s) for demonstrating compliance with the ECCS
acceptance criteria in paragraph (e)(1) of this section meet the
requirements in paragraphs (e)(2) and (3) of this section;
(iii) The risk-informed evaluation used to make changes under this
section is adequate for determining whether the acceptance criteria in
paragraph (h) of this section have been met;
(iv) If applicable, the risk-informed evaluation process proposed
for use to make changes under paragraph (h)(1) of this section is
adequate for determining whether the acceptance criteria in paragraph
(h) of this section have been met;
(v) For each reactor not authorized to operate on December 31,
2015, the evaluation submitted under paragraph (c)(2) of this section
demonstrates that the reactor design is similar to the designs of
reactors authorized to operate under this part on December 31, 2015,
and the applicant demonstrates that its proposed TBS applies to its
facility; and
(vi) The applicable standards and requirements of the Act and the
Commission's regulations have been met.
(d) Programmatic requirements. An entity whose application under
paragraph (c) of this section is approved by the NRC must comply with
the following requirements as long as the entity is subject to the
requirements in this section:
(1) The entity must maintain the ECCS evaluation models meeting the
requirements in paragraphs (e)(1), (2), and (3) of this section after
implementing any error corrections and changes;
(2) The entity must have leak detection systems available at the
facility and must implement actions during operation as necessary to
identify, monitor, and quantify leakage to ensure that adverse safety
consequences do not result from leaking primary pressure boundary
components that are larger than the TBS;
(3) Changes made under this section must, in addition to meeting
other applicable NRC requirements, be evaluated by a risk-informed
evaluation demonstrating that the acceptance criteria in paragraph (h)
of this section are met;
(4) The entity must perform an evaluation to determine the effect
of all planned facility changes and must not implement any facility
change that would significantly increase LOCA frequencies or invalidate
the evaluation demonstrating the applicability of the TBS performed
pursuant to paragraph (c)(1)(i) of this section for an operating
reactor licensee authorized to operate under this part on December 31,
2015, or the evaluation used to determine the plant-specific TBS
performed pursuant to paragraph (c)(2) of this section for entities
other than those authorized to operate under this part on December 31,
2015; and
(5) During operation, the licensees must perform the inspections
prescribed in paragraph (b)(3) of this section during every subsequent
inservice inspection interval (as defined in Sec. 50.55a(y)) on the
same samples inspected to satisfy paragraph (b)(3) of this section. The
effect on the TBS of any additional degradation identified since the
previous inspection must be evaluated.
(e) ECCS Performance.
(1) Alternative ECCS acceptance criteria. For each entity approved
by the NRC to use this section, its reactor must be provided with an
ECCS designed to satisfy the acceptance criteria in this paragraph in
the event of, and following, a postulated LOCA. The demonstration of
ECCS performance must comply with paragraph (e)(2) of this section for
breaks at or below the TBS and paragraph (e)(3) of this section for
breaks above the TBS.
(i) The ECCS provides sufficient coolant so that the fuel remains
in a coolable geometry during and following the LOCA heatup and quench.
(ii) The ECCS provides sufficient coolant so that decay heat will
be removed for the extended period of time required by the long-lived
radioactivity remaining in the fuel.
(2) ECCS evaluation performance demonstration for LOCAs involving
breaks at or below the TBS. ECCS cooling performance at or below the
TBS must be calculated in accordance with an evaluation model that
meets the requirements of either section I to appendix K to this part,
or for realistic evaluation models, the following requirements.
(i) The evaluation model must be used for a number of postulated
LOCAs of different sizes, locations (including LOCAs in piping systems
with an inner diameter that is larger than the TBS), and other
properties sufficient to provide assurance that the most severe
postulated LOCAs involving breaks at or below the TBS are analyzed.
(ii) The evaluation model must include sufficient supporting
justification to show that the analytical technique realistically
describes the behavior of the reactor system during a LOCA. Comparisons
to applicable experimental data must be made and uncertainties in the
analysis method and inputs must be identified and assessed so that the
uncertainty in the calculated results can be estimated. This
uncertainty must be accounted for, so that when the calculated ECCS
cooling performance is compared to the ECCS performance criteria set
forth in paragraph (e)(1) of this section and
[[Page 44675]]
addresses the fuel system acceptance criteria and modeling requirements
in paragraph (f) of this section, there is a high level of probability
that the criteria would be met.
(iii) The ECCS evaluation model must address changes in fuel
geometry.
(3) ECCS performance demonstration for LOCAs involving breaks
larger than the TBS. ECCS cooling performance for LOCAs involving
breaks larger than the TBS must be calculated in accordance with an
evaluation model that meets the requirements of either section I to
appendix K to this part or, for realistic evaluation models, the
following requirements. These calculations may take credit for the
availability of offsite power and do not require the assumption of a
single failure. Availability of safety-related or non-safety-related
equipment may be assumed if supported by plant-specific data or
analysis, and provided that onsite power can be readily provided
through simple manual actions to equipment that is credited in the
analysis.
(i) The evaluation model must be used for a number of postulated
LOCAs of different sizes, locations, and other properties sufficient to
provide assurance that the most severe postulated LOCAs larger than the
TBS up to the double-ended rupture of the largest pipe in the reactor
coolant system are analyzed.
(ii) The evaluation model must include sufficient supporting
justification to show that the analytical technique realistically
describes the behavior of the reactor system during a LOCA. Comparisons
to applicable experimental data must be made so that there is assurance
to at least a best-estimate level that the calculated ECCS cooling
performance meets the ECCS performance criteria set forth in paragraph
(e)(1) of this section and addresses the fuel system acceptance
criteria and modeling requirements in paragraph (f) of this section.
(iii) The ECCS evaluation model must address changes in fuel
geometry.
(4) Required documentation. The documentation requirements of this
paragraph supersede the requirements in section II, ``Required
Documentation,'' of appendix K to this part for those entities that are
approved to use this section.
(i) (A) A description of the ECCS evaluation model must be
submitted to the NRC. The description must be sufficiently complete to
permit technical review of the analytical approach, including the
equations used, their approximations in difference form, the
assumptions made, and the values of all parameters or the procedure for
their selection.
(B) A detailed source code of each computer program, in the same
form as used in the ECCS evaluation model, must be provided to the NRC
upon request.
(ii) For each computer program, solution convergence must be
demonstrated by studies of system modeling, noding and calculational
time steps, or both.
(iii) Appropriate sensitivity studies must be performed for each
ECCS evaluation model to evaluate the effect on the calculated results
of variations in noding, phenomena assumed in the calculation to
predominate, including pump operation or locking, and values of
parameters over their applicable ranges. For items to which results are
shown to be sensitive, the choices made must be justified.
(iv) To the extent practicable, predictions of the ECCS evaluation
model, or portions thereof, must be compared with applicable
experimental information. The technical adequacy of the calculational
methods used in the ECCS evaluation models must be documented. For
realistic evaluation models, the documentation must demonstrate that
the performance criteria of paragraphs (e)(1) and (f) of this section
are met. For appendix K models, this documentation must demonstrate
compliance with required features of section I of appendix K to this
part and must demonstrate that the performance criteria of paragraphs
(e)(1) and (f) of this section are met.
(v) The Director of the Office of Nuclear Reactor Regulation may
impose restrictions on reactor operation if the NRC finds that the
submitted evaluations of ECCS cooling performance are not consistent
with paragraph (e) of this section.
(f) Fuel performance criteria. Fuel system designs must have NRC-
approved limits that:
(1) Address cladding degradation phenomena;
(2) Maintain fuel coolability;
(3) Avoid explosive concentration of combustible gas; and
(4) Demonstrate that, after any calculated successful initial
operation of the ECCS, the ECCS must provide sufficient coolant to
remove decay heat and prevent further cladding failure for the extended
period of time required by the long-lived radioactivity remaining in
the fuel.
(g) Use of NRC-approved fuel in reactor.
(1) Fuel load. A licensee that is approved to use this section may
not load fuel into a reactor unless the resulting core design satisfies
the ECCS performance requirements of paragraph (e) of this section and
the fuel system acceptance criteria and modeling requirements in
paragraph (f) of this section, or otherwise complies with technical
specifications governing lead test assemblies in its license.
(2) Operation. If a licensee that is approved to use this section
determines that fuel in the reactor no longer complies with the ECCS
performance requirements of paragraph (e) of this section and the fuel
system acceptance criteria and modeling requirements in paragraph (f)
of this section, then the licensee must take immediate action to come
into compliance with paragraph (e) or (f) of this section, as
applicable.
(h) Changes to facility, technical specifications, or procedures.
An entity that wishes to make changes enabled by this section must
perform a risk-informed evaluation.
(1) An entity other than a design certification applicant or holder
of a manufacturing license may make changes enabled by this section,
other than changes to the technical specifications, without prior NRC
approval if:
(i) The change is permitted under Sec. 50.59 for holders of
operating licenses, combined licenses that do not reference a standard
design certification or standard design approval, or manufacturing
license (under Sec. 52.98(b) of this chapter), or combined licenses
that reference a standard design approval; or permitted under Sec.
52.98(c) of this chapter for holders of combined licenses that
reference a standard design certification; or permitted under Sec.
52.98(d) of this chapter for holders of combined licenses that
reference a manufacturing license;
(ii) The risk-informed evaluation process approved in accordance
with paragraph (c)(1)(v) of this section demonstrates that any
increases in the estimated risk are minimal and the criteria in
paragraph (h)(3) of this section are met; and
(iii) The change does not significantly increase LOCA frequencies
or invalidate the evaluation demonstrating the applicability of the TBS
to the applicant's facility, performed pursuant to paragraph (c)(1)(i)
of this section for an operating reactor licensee authorized to operate
under this part on December 31, 2015, or the evaluation used to
establish the plant-specific TBS, performed pursuant to paragraph
(c)(2) of this section for entities other than those authorized to
operate under this part on December 31, 2015.
(2) For implementing changes that are not permitted under paragraph
(h)(1) of
[[Page 44676]]
this section, the entity must submit an application containing the
following:
(i) For reactor licensees, the information required under Sec.
50.90;
(ii) Information from the risk-informed evaluation demonstrating
that the total increases in core damage frequency and large early
release frequency are very small, the overall risk remains small, and
the criteria in paragraph (h)(3) of this section are met;
(iii) If previous changes have been made under this section,
information from the risk-informed evaluation on the cumulative effect
on risk of the proposed change and all previous changes made under this
section. If more than one plant change is combined, including plant
changes not enabled by this section, into a group for the purposes of
evaluating acceptable risk increases, then the evaluation of each
individual change must be performed along with the evaluation of
combined changes;
(iv) Information demonstrating that the criteria in paragraph (e)
of this section are met; and
(v) Information demonstrating that the proposed change will not
significantly increase the LOCA frequencies or invalidate the
evaluation demonstrating the applicability of the TBS to the entity's
facility, performed pursuant to paragraph (c)(1)(i) of this section for
an operating reactor licensee authorized to operate under this part on
December 31, 2015, or the evaluation used to establish the plant-
specific TBS, performed pursuant to paragraph (c)(2) of this section
for entities other than those authorized to operate under this part on
December 31, 2015.
(3) All changes made under this section must meet the following
criteria:
(i) Adequate defense-in-depth is maintained;
(ii) Adequate safety margins are retained to account for
uncertainties; and
(iii) Adequate performance-measurement programs are implemented to
ensure the risk-informed evaluation continues to reflect actual plant
design and operation. These programs must be designed to detect
degradation of the system, structure, or component before plant safety
is compromised, provide feedback of information and timely corrective
actions, and monitor systems, structures, or components at a level
commensurate with their safety significance.
(4) Whenever a PRA is used in the risk-informed evaluation, the PRA
must, with respect to the area of evaluation that is the subject of the
PRA:
(i) Address initiating events from sources both internal and
external to the plant and for all modes of operation, that would affect
the regulatory decision in a substantial manner;
(ii) Reasonably represent the current configuration and operating
practices at the plant;
(iii) Have sufficient technical acceptability (including
consideration of uncertainty) and level of detail to provide confidence
that the total risk estimates and the change in total risk estimates
adequately reflect the plant and the effect of the proposed change on
risk; and
(iv) Be determined, through peer review, to meet industry standards
for PRA acceptability that have been endorsed or otherwise found
acceptable by the NRC.
(5) Whenever risk assessment methods other than PRAs are used to
develop quantitative or qualitative estimates of changes to risk in the
risk-informed evaluation, an integrated and systematic process must be
used. All aspects of the analyses must reasonably reflect the current
plant configuration and operating practices and applicable plant and
industry operating experience.
(i) Authority to impose restrictions on operation.
The Director of the Office of Nuclear Reactor Regulation may impose
restrictions on reactor operation if the NRC finds that the submitted
evaluations of ECCS cooling performance are not consistent with the
requirements of this section.
(j) Reporting. Each entity subject to the requirements of this
section must comply with the requirements of this paragraph.
(1) ECCS evaluation model: reporting.
(i) If the applicant for or holder of a construction permit,
operating license, combined license, or manufacturing license
identifies any change to, or error in, an ECCS evaluation model, or the
application of such a model, that does not result in any predicted
response that exceeds any of the acceptance criteria specified in this
section and is itself not significant as defined in paragraph (k) of
this section, then each of these entities must prepare a report
describing each such change or error, its estimated effect on predicted
response, and the basis for the entity's determination that the change
or error is not significant. This entity must submit the report to the
NRC, as specified in Sec. 50.4 or Sec. 52.3 of this chapter, at least
annually.
(ii) If the applicant for or holder of a construction permit,
operating license, combined license, or manufacturing license
identifies any change to, or error in, an ECCS evaluation model, or the
application of such a model, that does not result in any predicted
response that exceeds any of the acceptance criteria specified in this
section but is significant as defined in paragraph (k) of this section,
then each of these entities must prepare a report describing each such
change or error, its estimated effect on predicted response, proposed
corrective actions, and a proposed scope and schedule for providing a
reanalysis and for implementing the corrective actions. This entity
must submit the report to the NRC, as specified in Sec. 50.4 or Sec.
52.3 of this chapter, within 60 days of the change or discovery of the
error.
(iii) If an applicant for a standard design certification
(including an applicant after the Commission has adopted a final design
certification regulation) or an applicant for or holder of a standard
design approval under part 52 of this chapter identifies any change to,
or error in, an ECCS evaluation model, or the application of such a
model, that does not result in any predicted response that exceeds any
of the acceptance criteria specified in this section but is significant
as defined in paragraph (k) of this section, then each of these
entities must document the nature of the change or error and its
estimated effect on the limiting ECCS analysis.
(iv) If a licensee identifies any change to, or error in, an ECCS
evaluation model or the application of such a model, that results in
any of the ECCS acceptance criteria specified in this section to be
exceeded at the facility, then the licensee must submit a report
describing each such change or error, its estimated effect on predicted
response, proposed corrective actions, and a proposed scope and
schedule for providing a reanalysis and for implementing the corrective
actions. The licensee must submit the report to the NRC, as specified
in Sec. 50.4 or Sec. 52.3 of this chapter, within 60 days of the
change or discovery of the error. The report required by this paragraph
is in addition to any reporting required by Sec. 50.72.
(2) ECCS evaluation model: corrective action.
(i) If a licensee identifies any change to, or error in, an ECCS
evaluation model or the application of such a model, that results in
any of the acceptance criteria specified in this section to be exceeded
at the facility, then the licensee (in the case of a combined license
under part 52 of this chapter, after the Commission has made the
finding under Sec. 52.103(g) of this chapter) must take immediate
action to bring the facility into compliance with
[[Page 44677]]
the acceptance criteria. In addition, the corrective action as
described in the report required by paragraph (j)(1) of this section
must be implemented.
(ii) If a standard design certification applicant (including an
applicant after the Commission has adopted a final design certification
regulation) is required by paragraph (j)(1)(iv) of this section to
submit a reanalysis, or identifies a change to, or error in an ECCS
evaluation model, or in the application of such a model, that results
in any predicted response that exceeds any of the acceptance criteria
specified in this section, then the standard design certification
applicant (including an applicant after the Commission has adopted a
final design certification regulation) must propose appropriate steps
to the Commission, as specified in Sec. 52.3 of this chapter, within
60 days to demonstrate compliance with Sec. 50.46a requirements, along
with a report of the nature of the changes or errors that resulted in
an inability to assure compliance and an estimate of their effect on
the limiting transient.
(iii) If an applicant for or holder of a standard design approval
under part 52 of this chapter is required by paragraph (j)(1)(iv) of
this section to submit a reanalysis, or identifies a change to, or
error in an ECCS evaluation model, or in the application of such a
model, that results in any predicted response that exceeds any of the
acceptance criteria specified in this section, then the standard design
approval applicant or holder must propose appropriate steps to the
Commission, as specified in Sec. 52.3 of this chapter, within 60 days
to demonstrate compliance with Sec. 50.46a requirements, along with a
report of the nature of the changes or errors that resulted in an
inability to assure compliance and an estimate of their effect on the
limiting transient.
(3) Minimal changes: reporting. No later than 24 months after NRC
approval of the entity's application and every 24 months thereafter,
the entity must submit, as specified in Sec. 50.4 or Sec. 52.3 of
this chapter, a short description of each change involving minimal
changes in risk made under paragraph (h)(1) of this section in the
preceding 24 months and a brief summary of the basis for the entity's
determination pursuant to paragraph (h)(1)(iii) of this section that
the change does not invalidate the applicability evaluation made under
paragraph (c)(1)(i) of this section for an operating reactor licensee
authorized to operate under this part on December 31, 2015, or the
plant-specific TBS evaluation made under paragraph (c)(2) of this
section for entities other than those authorized to operate under this
part on December 31, 2015.
(4) Inspection: reporting. Within 120 days after completing the
outage when the inspections specified in paragraph (d)(5) of this
section were performed, the licensee must submit a summary report
detailing the results of the inspections and the evaluation of the
effect on the TBS of any additional degradation identified since the
previous evaluation. This report can be combined with the summary
report required by Sec. 50.55a(b)(2)(xxxii).
(k) Significant change or error in the ECCS evaluation model.
(1) For LOCAs at or below the TBS, a significant change or error in
the ECCS evaluation model for uranium oxide and mixed uranium-plutonium
oxide pellets within cylindrical zirconium-alloy cladding is one that
results in:
(i) A calculated peak fuel cladding temperature different by more
than 50 [deg]F from the temperature calculated for the limiting
transient using the last acceptable evaluation model, or is a
cumulation of changes and errors such that the sum of the absolute
magnitudes of the respective temperature changes is greater than 50
[deg]F; or
(ii) A calculated integral time-at-temperature different by more
than 1.0 percent equivalent cladding reacted from the oxidation
calculated for the limiting transient using the last acceptable
evaluation model, or is a cumulation of changes and errors such that
the sum of the absolute magnitudes of the respective oxidation changes
is greater than 1.0 percent equivalent cladding reacted.
(2) For LOCAs above the TBS, a significant change or error in the
ECCS evaluation model for uranium oxide and mixed uranium-plutonium
oxide pellets within cylindrical zirconium-alloy cladding is one that
results in a significant reduction in the capability to meet the
requirements of paragraphs (e)(1) and (f) of this section.
(3) For fuel that does not consist of uranium or mixed uranium-
plutonium oxide pellets within cylindrical zirconium-alloy cladding, a
significant change in the ECCS evaluation model is one that results in
a significant reduction in the capability to meet the requirements of
paragraphs (e)(1) and (f) of this section.
(l) Documentation. Following implementation of the requirements in
this section, each entity subject to this section must maintain records
sufficient to demonstrate compliance with the requirements in this
section in accordance with Sec. 50.71.
0
11. Add Sec. 50.46b under the undesignated center heading ``Standards
for Licenses, Certifications, and Regulatory Approvals'' to read as
follows:
Sec. 50.46b Acceptance criteria for reactor coolant system venting
systems.
Each nuclear power reactor must be provided with high-point vents
for the reactor coolant system, for the reactor vessel head, and for
other systems required to maintain adequate core cooling if the
accumulation of noncondensible gases would cause the loss of function
of these systems. High-point vents are not required for the tubes in U-
tube steam generators. Acceptable venting systems must meet the
following criteria:
(a) The high-point vents must be remotely operated from the control
room.
(b) The design of the vents and associated controls, instruments,
and power sources must conform to appendix A and appendix B of this
part.
(c) The vent system must be designed to ensure that:
(1) The vents will perform their safety functions; and
(2) There would not be inadvertent or irreversible actuation of a
vent.
0
12. In Sec. 50.47,
0
a. In paragraph (b)(10), remove the word ``EPZ'' wherever it appears;
0
b. Revise paragraph (c)(2); and
0
c. Add paragraphs (g) and (h) to read as follows:
Sec. 50.47 Emergency plans.
* * * * *
(c) * * *
(2) Generally, the plume exposure pathway EPZ for nuclear power
reactors shall consist of an area about 2 to 10 miles (3.2 to 16 km) in
radius. For reactors with an authorized power level less than 300 MW
thermal, the plume exposure pathway EPZ may be established at the site
boundary. The need for and size of the EPZ may also be determined on a
case-by-case basis as described in Sec. 50.33(g)(2). The exact size
and configuration of the EPZ surrounding a particular nuclear power
reactor 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. Emergency plans must describe such actions
as are appropriate to avoid or reduce dose within and beyond the EPZ or
site boundary and to protect the ingestion pathway.
* * * * *
(g) A licensee desiring to change its plume exposure pathway EPZ
must submit an application for a license
[[Page 44678]]
amendment under Sec. 50.90 and receive NRC approval before
implementing the change. Any such license amendment request must
include documentation demonstrating that the applicable State, local,
and Tribal governmental authorities have agreed to the EPZ change.
(h) A licensee desiring to comply with the requirements of Sec.
50.160 in lieu of appendix E to this part, and for nuclear power
reactor licensees, the planning standards of Sec. 50.47(b), must
submit an application for a license amendment under Sec. 50.90 and
receive NRC approval before implementing the change.
0
13. In Sec. 50.49, remove footnote 3, redesignate footnote 4 as
footnote 1, and revise and republish paragraph (b) to read as follows:
Sec. 50.49 Environmental qualification of electric equipment
important to safety for nuclear power plants.
* * * * *
(b) Electric equipment important to safety covered by this section
is:
(1) Safety-related electric equipment.
(i) This equipment is that relied upon to remain functional during
and following design basis events to ensure--
(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 that could result in potential offsite exposures comparable
to the guidelines in Sec. 50.34(a)(1), 50.67(b)(2), or 100.11 of this
chapter, as applicable.
(ii) For applications submitted on or after [DATE 180 DAYS AFTER
THE EFFECTIVE DATE OF FINAL RULE] for a license, permit, standard
design certification, or standard design approval under this part or
part 52 of this chapter, design basis events are defined in Sec. 50.2.
This definition also applies to other licenses, permits, standard
design certifications, or standard design approvals for which the NRC
has approved adopting the Sec. 50.2 definition for design basis
events. In all other cases, design basis events are defined as
conditions of normal operation, including anticipated operational
occurrences, design basis accidents, external events, and natural
phenomena for which the plant must be designed to ensure functions
(b)(1)(i)(A) through (C) of this section.
(2) Non-safety-related electric equipment whose failure under
postulated environmental conditions could prevent satisfactory
accomplishment of safety functions specified in subparagraphs
(b)(1)(i)(A) through (C) of this section by the safety-related
equipment.
(3) Certain post-accident monitoring equipment.\[1]\
* * * * *
\[1]\ Specific guidance concerning the types of variables to be
monitored is provided in Revision 2 of Regulatory Guide 1.97,
``Instrumentation for Light-Water-Cooled Nuclear Power Plants to
Assess Plant and Environs Conditions During and Following an
Accident.'' Copies of the Regulatory Guide may be purchased through
the U.S. Government Publishing Office by calling 202-512-1800 or by
writing to the U.S. Government Publishing Office, P.O. Box 37082,
Washington, DC 20013-7082.
0
14. In Sec. 50.54, add paragraphs (a)(5) and (q)(1)(v) and revise
paragraphs (a)(1), (q)(2), (q)(3), and (t) to read as follows:
Sec. 50.54 Conditions of licenses.
* * * * *
(a)(1) Each nuclear power plant or fuel reprocessing plant licensee
subject to the quality assurance criteria in appendix B or T of this
part shall implement, under Sec. 50.34(b)(6)(ii) or Sec. 52.79 of
this chapter, the quality assurance program or quality management
system, respectively, described or referenced in the safety analysis
report, including changes to that report. However, a holder of a
combined license under part 52 of this chapter shall implement the
quality assurance program or quality management system described or
referenced in the safety analysis report applicable to operation 30
days prior to the scheduled date for the initial loading of fuel.
* * * * *
(5) Changes to the quality management system must be submitted to
the NRC and receive NRC approval prior to implementation, as follows:
(i) Changes made to the quality management system as presented in
the Safety Analysis Report or in a topical report must be submitted as
specified in Sec. 50.4.
(ii) The submittal of a change to the Safety Analysis Report
quality management system 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 quality management system incorporating the change
continues to satisfy the criteria of appendix T of this part and the
Safety Analysis Report quality management system 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 change must
be maintained as a facility record for three years.
(iv) Changes to the quality management system 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.
* * * * *
(q) * * *
(1) * * *
(v) Risk significant planning standard means the most essential
functions of emergency preparedness to ensure adequate protective
measures are taken to protect the public in the event of a radiological
emergency. For the purposes of this section, the risk significant
planning standards are classification, notification, assessment,
protective actions, staffing, and facilities.
(2) A holder of a license under this part, or a combined license
under part 52 of this chapter after the Commission makes the finding
under Sec. 52.103(g) of this chapter, shall follow and maintain the
effectiveness of an emergency plan that meets the requirements in
appendix E to this part and, for nuclear power reactor licensees, the
planning standards of Sec. 50.47(b), or an emergency plan that meets
the requirements in Sec. 50.160.
(3) A licensee may make changes to its emergency plan without NRC
approval only if the licensee performs and retains an analysis
demonstrating that:
(i) For planning standards that are risk significant, the changes
do not reduce the effectiveness of the plan and the plan, as changed,
continues to meet either the risk significant requirements of Sec.
50.160 or the applicable requirements in appendix E to this part and,
for nuclear power reactor licensees, the risk significant planning
standards of Sec. 50.47(b); and
(ii) For planning standards that are not risk-significant, the
plan, as changed, continues to meet the applicable requirements.
* * * * *
(t) The licensee must provide for annual evaluation of the adequacy
of the interfaces between the licensee and the applicable State, local,
and Tribal governments, including licensee drills, exercises,
capabilities, and procedures. The results of the evaluation, along with
recommendations for improvements, must be documented, reported to the
[[Page 44679]]
licensee's corporate and plant management, retained for a period of 5
years, and must be made available to the appropriate State, local, and
Tribal governments.
* * * * *
0
15. In Sec. 50.55, revise paragraph (f)(1) and add paragraph (f)(5) to
read as follows:
Sec. 50.55 Conditions of construction permits, early site permits,
combined licenses, and manufacturing licenses.
* * * * *
(f)(1) Each nuclear power plant or fuel reprocessing plant
construction permit holder subject to the quality assurance criteria in
appendix B or T of this part shall implement, pursuant to Sec.
50.34(a)(7) of this part, the quality assurance program description or
quality management system, respectively, described or referenced in the
Safety Analysis Report, including changes to that report.
* * * * *
(5) Changes to the quality management system must be submitted to
the NRC and receive NRC approval prior to implementation, as follows:
(i) Changes made to the quality management system as presented in
the Safety Analysis Report or in a topical report must be submitted as
specified in Sec. 50.4.
(ii) The submittal of a change to the Safety Analysis Report
quality management system 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 quality management system incorporating the change
continues to satisfy the criteria of appendix T of this part and the
Safety Analysis Report quality management system 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 change must
be maintained as a facility record for three years.
(iv) Changes to the quality management system 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.
* * * * *
Sec. 50.55a [Amended]
0
16. In Sec. 50.55a, in paragraph (z) introductory text, remove the
phrase ``of paragraphs (b) through (h)''.
0
17. In Sec. 50.57, add paragraph (d) to read as follows:
Sec. 50.57 Issuance of operating license.\[1]\
* * * * *
(d) The Commission may afford generic finality to generic aspects
of the design of a utilization facility, including postulated site
parameters, and requirements submitted pursuant to Sec. 50.34(b)(14),
if it finds that the proposed generic design can be constructed and
operated at sites having characteristics that fall within the site
parameters postulated for the design in accordance with applicable
requirements and without undue risk to the health and safety of the
public.
\[1]\ The Commission may issue a provisional operating license
pursuant to the regulations in this part in effect on March 30,
1970, for any facility for which a notice of hearing on an
application for a provisional operating license or a notice of
proposed issuance of a provisional operating license has been
published on or before that date.
0
18. In Sec. 50.58, add paragraphs (b)(7) and (b)(8) to read as
follows:
Sec. 50.58 Hearings and report of the Advisory Committee on Reactor
Safeguards.
* * * * *
(b) * * *
(7) If an applicant requests generic finality under Sec.
50.34(b)(14) for an operating license under this part, the Commission
will include a request for generic finality as a proposed action in the
notice of proposed action required by Sec. 2.105 of this chapter.
(8) In a proceeding for issuance of a construction permit,
operating license, or combined license, or in any enforcement hearing
other than one initiated by the Commission under Sec. 2.202(e)(1) of
this chapter, in which an operating license issued under this subpart
is referenced, the Commission must treat as resolved those matters
resolved in the proceeding on the application for issuance or renewal
of the referenced operating license, including, if applicable, the
adequacy of a reactor design, if the referenced operating license was
afforded finality pursuant to Sec. 50.57(d).
0
19. In Sec. 50.59, revise paragraphs (c)(2)(viii) and (d)(1) and add
paragraphs (e) and (f) to read as follows:
Sec. 50.59 Changes, tests, and experiments.
* * * * *
(c) * * *
(2) * * *
(viii) Result in a departure from a method of evaluation described
in the FSAR (as updated) used in establishing the design bases or in
the safety analyses, unless the licensee has demonstrated through a
documented verification, validation, and uncertainty quantification
(VVUQ) process, conducted under a VVUQ program that meets the
requirements of Sec. 50.221 and has been approved by the NRC for the
intended application, that the departure from a method of evaluation
described in the FSAR (as updated) meets the criteria established for
credibility in the VVUQ program, including implementation of
established VVUQ activities and assessments.
* * * * *
(d) (1) The licensee shall maintain records of changes in the
facility, of changes in procedures, and of tests and experiments made
pursuant to paragraph (c) or (f) of this section. These records must
include a written evaluation which provides the bases for the
determination that the change, test, or experiment does not require a
license amendment pursuant to paragraph (c)(2) or (f) of this section.
* * * * *
(e) For the purposes of criteria Sec. 50.59(c)(2)(i) and (ii), an
increase may be demonstrated to not be a ``more than a minimal
increase'' using quantitative risk results based on a probabilistic
risk assessment of appropriate scope and quality that provides
appropriate risk metrics. The change must also maintain defense-in-
depth and safety margins.
(f) The holder of an operating license or a combined license that
authorizes operation of a manufactured reactor may make changes in the
facility as described in the final safety analysis report (as updated)
and make changes in the procedures as described in the final safety
analysis report (as updated) without obtaining a license amendment
pursuant to Sec. 50.90 if the changes are identical to changes
approved by the Commission by amendment to the manufacturing license
for the manufactured reactor and upon determining that implementation
of the changes will be consistent with the basis for the Commission's
approval of the amendment to the manufacturing license and not involve
any additional changes that would require an amendment to its operating
license or combined license.
0
20. In Sec. 50.67, revise and republish paragraph (b) to read as
follows:
Sec. 50.67 Accident source term.
* * * * *
(b) Requirements.
(1) A licensee who seeks to revise its current accident source term
in design basis radiological consequence analyses must apply for a
license amendment
[[Page 44680]]
under Sec. 50.90. The application must contain an evaluation of the
consequences of applicable design basis accidents \[1]\ previously
analyzed in the safety analysis report.
(2) The NRC may issue the amendment only if the applicant's
analysis demonstrates with reasonable assurance that:
(i) 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 \[2]\ (0.25 Sv) total effective dose equivalent
(TEDE).
(ii) 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 (0.25 Sv) TEDE.
(iii) The necessary design, fabrication, construction, testing, and
performance criteria for structures, systems, and components important
to safety are provided to permit occupancy of the control room under
accident conditions without calculated radiation exposures in excess of
10 rem (0.10 Sv) TEDE or a higher design criterion limit established in
accordance with paragraph (b)(3) of this section for the duration of
the accident.
(3) The licensee may establish a design criterion limit higher than
10 rem (0.10 Sv) TEDE but not greater than 25 rem \[3]\ (0.25 Sv) TEDE
for compliance with paragraph (b)(2)(iii) of this section provided the
licensee demonstrates that the specified limit is consistent with the
plant risk profile or commensurate with the risk of the plant.
\[1]\ The fission product release assumed for these calculations
should be based upon a major accident, hypothesized for purposes of
design analyses or postulated from considerations of possible
accidental events, that would result in potential hazards not
exceeded by those from any accident considered credible. Such
accidents have generally been assumed to result in substantial
meltdown of the core with subsequent release of appreciable
quantities of fission products.
\[2]\ The use of 25 rem (0.25 Sv) TEDE is not intended to imply
that this value constitutes an acceptable limit for emergency doses
to the public under accident conditions. Rather, this 25 rem (0.25
Sv) TEDE value has been stated in this section as a reference value,
which can be used in the evaluation of proposed design basis changes
with respect to potential reactor accidents of exceedingly low
probability of occurrence and low risk of public exposure to
radiation.
\[3]\ The use of 25 rem (0.25 Sv) TEDE as the control room
criterion is not intended to imply that this value constitutes an
acceptable limit for emergency doses under accident conditions.
Adequate radiological protection for occupationally exposed
individuals is provided by the provisions of part 20 of this chapter
and Sec. 50.47(b)(11). This criterion is provided only to assess
the acceptability of design provisions, in particular engineered
safety features that mitigate fission product release, under
postulated DBA conditions. The conditions assumed in these analyses,
although credible, are of exceedingly low probability of occurrence
and do not represent actual accident sequences but are specified as
conservative surrogates to create bounding conditions for assessing
the acceptability of engineered safety features.
0
21. In Sec. 50.68, revise paragraph (b)(7) to read as follows:
Sec. 50.68 Criticality accident requirements.
* * * * *
(b) * * *
(7) The maximum nominal U-235 enrichment of the fresh fuel
assemblies is limited to five (5.0) percent by weight or to the value
specified in the operating license which must be less than twenty
(20.0) percent by weight.
* * * * *
Sec. 50.69 [Amended]
0
22. In Sec. 50.69, revise paragraph (b)(1)(ii) by removing the
reference ``10 CFR 50.46a(b)'' and adding in its place ``Sec.
50.46b(b)''.
0
23. In Sec. 50.71, revise paragraph (f) to read as follows:
Sec. 50.71 Maintenance of records, making of reports.
* * * * *
(f) Each person licensed to manufacture a nuclear power reactor
under subpart F of part 52 of this chapter shall update the FSAR
originally submitted as part of the application to reflect the
effects\[2]\ of all changes made in the facility or procedures as
described in the FSAR; all safety analyses and evaluations performed by
the licensee either in support of approved amendments to the
manufacturing license or in support of conclusions that changes did not
require a license amendment in accordance with Sec. 50.59(c)(2); and
any new analyses of safety issues performed by or on behalf of the
licensee at the NRC's request. This submittal shall contain all the
changes necessary to reflect information and analyses submitted to the
Commission by the licensee or prepared by the licensee with respect to
the modification under Sec. 52.171 of this chapter or the analyses
requested by the Commission under Sec. 52.171 of this chapter. The
updated information shall be appropriately located within the update to
the FSAR.
* * * * *
\[2]\ See footnote 1.
0
24. In Sec. 50.75,
0
a. Revise paragraphs (b), (c), and (e)
0
b. In paragraph (g)(4)(iii), remove the phrase ``10 CFR Part 20,
Subpart E'' and add in its place the phrase ``subpart E to part 20 of
this chapter'';
0
c. In paragraph (h)(5), remove the reference ``(h)(3)'' and add in its
place ``(3)''.
The revisions are to read as follows:
Sec. 50.75 Reporting and recordkeeping for decommissioning planning.
* * * * *
(b) Each power reactor applicant for or holder of an operating
license, and each applicant for a combined license under subpart C of
part 52 of this chapter for a production or utilization facility shall
submit a decommissioning report, as required by Sec. 50.33(k).
(1)(i) For an applicant for or holder of an operating license under
this part of a type and power level specified in paragraph (c) of this
section, the report must contain a certification that financial
assurance for decommissioning will be (for a license applicant), or has
been (for a license holder), provided in an amount which may be more,
but not less, than the amount stated in the table in paragraph (c)(1)
of this section adjusted using a rate at least equal to that stated in
paragraph (c)(2) of this section. For an applicant for a combined
license under subpart C of part 52 of this chapter of a type and power
level specified in paragraph (c) of this section, the report must
contain a certification that financial assurance for decommissioning
will be provided no later than 30 days after the Commission publishes
notice in the Federal Register under Sec. 52.103(a) of this chapter in
an amount which may be more, but not less, than the amount stated in
the table in paragraph (c)(1) of this section, adjusted using a rate at
least equal to that stated in paragraph (c)(2) of this section.
(ii) The amount to be provided must be adjusted annually using a
rate at least equal to that stated in paragraph (c)(2) of this section.
(iii) The amount must be covered by one or more of the methods
described in paragraph (e) of this section as acceptable to the NRC.
(iv) The amount stated in the applicant's or licensee's
certification may be based on a site-specific cost estimate for
decommissioning the facility. As part of the certification, a copy of
the financial instrument obtained to satisfy the requirements of
paragraph (e) of this section must be submitted to NRC; provided,
however,
[[Page 44681]]
that an applicant for or holder of a combined license need not obtain
such financial instrument or submit a copy to the Commission except as
provided in paragraph (e)(3) of this section.
(2)(i) For an applicant for or holder of an operating license under
this part or subpart C of part 52 of this chapter of a type and power
level other than that specified in paragraph (c) of this section, the
report must contain a certification that financial assurance for
decommissioning will be (for a license applicant), or has been (for a
license holder), provided in an amount that may be based on a design-
specific decommissioning cost estimate or the amount stated in the
table in paragraph (c)(1) of this section adjusted using a rate at
least equal to that stated in paragraph (c)(2) of this section.
(ii) A certification relying on a design-specific decommissioning
cost estimate must demonstrate that there is reasonable assurance that
sufficient funds necessary for safely decommissioning the facility will
be available, when needed, and provide the factors used to develop the
design-specific decommissioning cost estimate, including reactor
technology, power level (in MWt), and costs related to labor, energy,
and waste burial. Additionally, design-specific decommissioning cost
estimates 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.
(iii) The amount to be provided must be adjusted annually using a
rate at least equal to that stated in paragraph (c)(2) of this section.
(iv) The amount must be covered by one or more of the methods
described in paragraph (e) of this section as acceptable to the NRC.
(v) The amount stated in the applicant's or licensee's
certification may be based on a site-specific cost estimate for
decommissioning the facility. As part of the certification, a copy of
the financial instrument obtained to satisfy the requirements of
paragraph (e) of this section must be submitted to NRC; provided,
however, that an applicant for or holder of a combined license need not
obtain such financial instrument or submit a copy to the Commission
except as provided in paragraph (e)(3) of this section.
(c) Table of minimum amounts (January 1986 dollars) required to
demonstrate reasonable assurance of funds for decommissioning by
reactor type and power level, P (in MWt); adjustment factor.\[1]\
[GRAPHIC] [TIFF OMITTED] TP16JY26.525
(2) An adjustment factor at least equal to 0.65 L + 0.13 E + 0.22 B
is to be used 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.''
* * * * *
(e)(1) Financial assurance is to be provided by the following
methods.
(i) Prepayment. Prepayment is the deposit made preceding the start
of operation or the transfer of a license under Sec. 50.80 into an
account segregated from applicant or licensee assets and outside the
administrative control of the applicant or licensee and its
subsidiaries or affiliates of cash or liquid assets such that the
amount of funds would be sufficient to pay decommissioning costs at the
time permanent termination of operations is expected. 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 shall 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. An applicant or
licensee that has prepaid funds based on a site-specific estimate under
paragraph (b)(1) or (2) of this section may take credit for projected
earnings on the prepaid decommissioning trust funds, using up to a 2-
percent annual real rate of return from the time of future funds'
collection through the projected decommissioning period, provided that
the site-specific estimate is based on a period of safe storage that is
specifically described in the estimate. This includes the periods of
safe storage, final dismantlement, and license termination. An
applicant or licensee that has prepaid funds based on a design-specific
estimate under paragraph(b)(2) of this section or on the formulas in
paragraph (c) of this section may take credit for projected earnings on
the prepaid decommissioning funds using up to a 2-percent annual real
rate of return up to the time of permanent termination of operations.
An applicant or licensee may use a credit of greater than 2 percent if
the applicant's or licensee's rate-setting authority has specifically
authorized a higher rate. However, applicants or licensees certifying
only to design-specific decommissioning cost estimates or formula
amounts (i.e., not a site-specific
[[Page 44682]]
estimate) can take a pro-rata credit during the immediate dismantlement
period (i.e., recognizing both cash expenditures and earnings the first
7 years after shutdown). Actual earnings on existing funds may be used
to calculate future fund needs.
(ii) External sinking fund. An external sinking fund is a fund
established and maintained by setting funds aside periodically in an
account segregated from applicant or licensee assets and outside the
administrative control of the applicant or licensee and its
subsidiaries or affiliates in which the total amount of funds would be
sufficient to pay decommissioning costs at the time permanent
termination of operations is expected. 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 shall 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. An applicant or licensee that has collected funds based on a
site-specific estimate under paragraph (b)(1) or (2) of this section
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 decommissioning period, provided
that the site-specific estimate is based on a period of safe storage
that is specifically described in the estimate. This includes the
periods of safe storage, final dismantlement, and license termination.
An applicant or licensee that has collected funds based on a design-
specific estimate under paragraph (b)(2) of this section or on the
formulas in paragraph (c) of this section may take credit for collected
earnings on the decommissioning funds using up to a 2-percent annual
real rate of return up to the time of permanent termination of
operations. An applicant or licensee may use a credit of greater than 2
percent if the applicant's or licensee's rate-setting authority has
specifically authorized a higher rate. However, applicants or licensees
certifying only to design-specific decommissioning cost estimates or
formula amounts (i.e., not a site-specific estimate) can take a pro-
rata credit during the dismantlement period (i.e., recognizing both
cash expenditures and earnings the first 7 years after shutdown).
Actual earnings on existing funds may be used to calculate future fund
needs. An applicant or 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 (e)(1)(ii). This method may be used as the
exclusive mechanism relied upon for providing financial assurance for
decommissioning in the following circumstances:
(A) By an applicant or 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 assumed to meet this
condition.
(B) By an applicant or 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 paragraph (b), (c), or (f) of this section, or Sec. 50.82.
(iii) A surety method, insurance, or other guarantee method.
(A) 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:
(1) 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 applicant or 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 applicant or licensee fails to provide a replacement acceptable to
the NRC within 30 days after receipt of notification of cancellation.
(2) 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.
(B) 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 part 30 of this chapter.
(C) 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 part 30 of this chapter. 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 part 30 of this chapter. For non-profit
entities, such as colleges, universities, and non-profit hospitals, a
guarantee of funds by the applicant or licensee may be used if the
guarantee and test are as contained in appendix E to part 30 of this
chapter. 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.
(iv) Statements of intent. For a Federal applicant for or holder of
a power reactor operating license, or for a Federal, State, or local
government applicant or holder of an operating license for a non-power
production or utilization facility, a statement of intent containing a
cost estimate for decommissioning, and indicating that funds for
decommissioning will be obtained when necessary.
(v) Contractual obligations. Contractual obligation(s) on the part
of an applicant's or licensee's customer(s), the total amount of which
over the duration of the contract(s) will provide the applicant's or
licensee's total share of uncollected funds estimated to be needed for
decommissioning pursuant to paragraph (b), (c), or (f) of this section,
or Sec. 50.82. To be acceptable to the NRC as a method of
decommissioning funding assurance, the terms of the contract(s) shall
include provisions that the electricity buyer(s) will pay for the
decommissioning obligations specified in the contract(s),
notwithstanding the operational status either of the licensed power
reactor 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
[[Page 44683]]
qualifications of the contracting entity or entities offered as
assurance for decommissioning funding.
(vi) Other mechanisms. Any other mechanism, or combination of
mechanisms, that provides, as determined by the NRC upon its evaluation
of the specific circumstances of each application or licensee
submittal, assurance of decommissioning funding equivalent to that
provided by the mechanisms specified in paragraphs (e)(1)(i) through
(v) of this section. Applicants or licensees who do not have sources of
funding described in paragraph (e)(1)(ii) of this section may use an
external sinking fund in combination with a guarantee mechanism, as
specified in paragraph (e)(1)(iii) of this section, provided that the
total amount of funds estimated to be necessary for decommissioning is
assured.
(2) The NRC reserves the right to take the following steps in order
to ensure an applicant's or licensee's adequate accumulation of
decommissioning funds: review, as needed, the rate of accumulation of
decommissioning funds; and, either independently or in cooperation with
the FERC and the applicant's or licensee's State PUC, take additional
actions as appropriate on a case-by-case basis, including modification
of an applicant's or licensee's schedule for the accumulation of
decommissioning funds.
(3) Each holder of a combined license under subpart C of part 52 of
this chapter shall, 2 years before and 1 year before the scheduled date
for initial loading of fuel, consistent with the schedule required by
Sec. 52.99(a) of this chapter, submit a report to the NRC containing a
certification updating the information described under paragraph (b) of
this section, including a copy of the financial instrument to be used.
No later than 30 days after the Commission publishes notice in the
Federal Register under Sec. 52.103(a) of this chapter, the licensee
shall submit a report containing a certification that financial
assurance for decommissioning is being provided in an amount specified
in the licensee's most recent updated certification, including a copy
of the financial instrument obtained to satisfy the requirements of
paragraph (e) of this section.
* * * * *
\[1]\ Amounts are based on activities related to the definition
of ``Decommission'' in Sec. 50.2 of this part and do not include
the cost of removal and disposal of spent fuel or of nonradioactive
structures and materials beyond that necessary to terminate the
license.
0
25. In Sec. 50.160, revise the section heading and paragraphs
(b)(1)(iv)(A)(2), (b)(3), and (c) to read as follows:
Sec. 50.160 Performance-based emergency preparedness standards.
* * * * *
(b) * * *
(1) * * *
(iv) * * *
(A) * * *
(2) Implement the emergency plan in response to a security event.
* * * * *
(3) Emergency planning zone. Determine and describe the boundary
and physical characteristics of the EPZ in the emergency plan as
required by Sec. 50.33(g)(1) or 53.1109(g)(1) of this chapter.
* * * * *
(c) Implementation.
(1) An applicant for an operating license issued under this part
after December 18, 2023, must establish, implement, and maintain an
emergency preparedness program that meets the requirements of paragraph
(b) of this section, as described in the emergency plan and license,
and conduct an initial exercise to demonstrate this compliance before
the issuance of an operating license for the facility described in the
license application.
(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 53.1452(g) of this chapter, 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 before the scheduled date for initial
loading of fuel, or for a fueled manufactured reactor, before the
scheduled date for initiating the removal of the features to prevent
criticality required under Sec. 53.620(d)(1) of this chapter.
(3) A licensee desiring to change its plume exposure pathway EPZ
must submit an application for a license amendment under Sec. 50.90
and receive NRC approval before implementing the change. Any such
license amendment request must include documentation demonstrating that
the applicable State and local governmental authorities have agreed to
the EPZ change.
(4) A licensee desiring to comply with the requirements in appendix
E to this part and, for nuclear power reactor licensees, the planning
standards in Sec. 50.47(b) in lieu of Sec. 50.160 must submit an
application for a license amendment under Sec. 50.90 and receive NRC
approval before implementing the change.
0
26. Add Sec. 50.220 under a new undesignated heading ``Risk-Informed
and Performance-Based Alternatives'' consisting of Sec. 50.220 and
Sec. 50.221 to read as follows:
Risk-Informed and Performance-Based Alternatives
Sec. 50.220 Use of risk-informed and performance-based alternatives
to acceptance criteria.
(a) For each regulation in this part that provides specific
acceptance criteria, licensees or applicants may propose, either
through a license amendment request or as part of an application for an
initial license, an alternative under paragraph (b) of this section. If
a proposal under paragraph (b) of this section includes alternatives to
acceptance criteria in more than one regulation in this part, it must
analyze the cumulative effect of those changes to ensure that an
appropriate level of safety is maintained.
(b) An applicant for or a holder of a construction permit or
operating license under this part; or an applicant for or holder of a
design approval, operating license, a combined license, or
manufacturing license under this chapter; voluntarily choosing to
implement this section must submit a license amendment application that
contains the following information or include the following information
in the initial application:
(1) The specific regulation and the acceptance criteria for which
the alternative criteria are being proposed.
(2) The alternative criteria proposed in place of the acceptance
criteria. If applicable, the alternative criteria must include either
the selected performance targets or a description of the methodology
that will be used to determine the performance targets.
(3) A description of the process followed to develop the
alternative acceptance criteria. The process must:
(i) Consider results and insights from the plant- or design-
specific probabilistic risk assessment (PRA) or systematic risk
assessment or combination thereof to show that the alternate proposed
criteria establish an equivalent level of safety to the existing
acceptance criteria.
(ii) Reasonably reflect the current plant configuration and
operating practices, and applicable plant, facility, and industry
operational experience.
(iii) Achieve or maintain adequate defense-in-depth.
(iv) Achieve or maintain sufficient safety margins.
[[Page 44684]]
(v) Use an integrated panel of plant- or design-knowledgeable
members whose collective expertise supports an integrated evaluation of
the alternative criteria.
(vi) Include a performance monitoring program that will be used to
periodically evaluate whether the alternative criteria remain valid and
make timely adjustments to the design and/or operation of the plant or
facility as necessary to maintain the validity.
(vii) Include an evaluation of the cumulative effect of the
proposed alternative criteria with relevant NRC-approved alternative
criteria.
0
27. Add Sec. 50.221 under the undesignated center heading ``Risk-
Informed and Performance-Based Alternatives'' to read as follows:
Sec. 50.221 Credibility requirements for modeling and simulation.
(a) Purpose and scope.
(1) Purpose. This regulation establishes requirements for licensees
or applicants to voluntarily adopt a verification, validation, and
uncertainty quantification (VVUQ) program that will ensure the
credibility of models and simulations (M&S), used to demonstrate
compliance with NRC safety criteria for nuclear power plants.
(2) Scope. A VVUQ program may apply to any model or simulation
whose results inform or directly support nuclear power plant safety
assessments, performance criteria, regulatory decisions, or otherwise
contribute to demonstrating compliance with regulatory requirements.
(b) Definitions. For the purpose of this section:
(1) Credibility means the degree to which a model's predictions for
a specific purpose can be justifiably trusted such that it can be
relied upon.
(2) Verification, validation, and uncertainty quantification (VVUQ)
means activities related to verification, validation, and uncertainty
quantification, as well as all other supporting tools and techniques
used to establish and demonstrate the credibility of a model or
simulation.
(3) M&S risk is defined as the combination of the likelihoods that
the M&S is incorrect or misleading and the consequences of relying on
the incorrect or misleading result for some decision making purpose.
Because simulations inherently include multiple potential sources of
error, assessing M&S risk is generally a cumulative process. M&S risk
can serve as a basis for establishing a tiered review framework,
organized into three levels: low, medium, and high.
(i) Low risk. Confirmation of the adequacy of VVUQ activities of
M&S determined to have low risk may be accomplished using internal
quality assurance procedures established by the licensee's or
applicant's quality assurance program.
(ii) Medium risk. Confirmation of the adequacy of VVUQ activities
of M&S determined to have medium risk may be provided by an external
organization independent of the internal modeling group.
(iii) High risk. Confirmation of the adequacy of VVUQ activities of
M&S determined to have high risk require review and approval by the
NRC, generally through the submittal of license amendments or topical
reports.
(c) Credibility requirements. The VVUQ program must establish
credibility requirements. Credibility requirements address two
complementary elements: the specific activities performed to establish
model or simulation credibility (VVUQ Activities), and the evaluation
demonstrating that the outcomes of these activities are sufficient and
appropriate for the intended use of the model or simulation (VVUQ
Assessments). Licensees and applicants may not adopt or rely on a model
or simulation covered by the VVUQ program prior to successful
completion of VVUQ activities and VVUQ assessments.
(1) VVUQ activities. The VVUQ program must specify the activities
necessary to establish and ensure the credibility of M&S used in
regulatory or safety applications for that specific VVUQ program. The
specific activities required depend on the characteristics of the model
or simulation, its intended application, and the associated M&S risk.
When considering M&S risk, individual error sources and their
associated likelihoods and consequences must be evaluated and then
combined to quantify the overall risk. In addition to verification,
validation, and uncertainty quantification, VVUQ programs may include a
range of supporting activities. These include, but are not limited to,
sensitivity analyses, scaling evaluations, comparisons against
experimental or operational data, and any other relevant methods or
techniques needed to adequately establish credibility.
(2) VVUQ assessments. The VVUQ program must require an evaluation
of the outcomes of the VVUQ activities performed to determine whether
they are sufficient to justify trust in the model or simulation for its
specific application. This assessment must consider all relevant
uncertainties and potential errors associated with the model or
simulation, including but not limited to those stemming from
computational methods, numerical approximations, input parameters,
modeling assumptions, and comparisons to experimental or operational
data. The VVUQ program must require identification of the
applicability, limitations, and key assumptions inherent to the
modeling or simulation approach, ensuring that the depth and rigor of
the assessment appropriately reflect the intended use, scope, and
associated risk of the specific model or simulation. Assessments must
not only evaluate numerical or quantitative outcomes of the VVUQ
activities, but also clearly interpret the significance of these
outcomes within the context of their intended regulatory application.
0
28. In appendix A to part 50,
0
a. Under the heading ``Introduction,'' in the last sentence of the last
paragraph, remove the word ``must'' and add in its place the word
``may'' and remove the word ``justified.'' and add in its place the
phrase ``justified within the licensing submittal without needing an
exemption.''
0
b. Under the heading ``Criteria,'' criteria 17, 19, 28, 35, 38, 41, 44,
and 50 are revised to read as follows:
Appendix A to Part 50--General Design Criteria for Nuclear Power Plants
* * * * *
Criterion 17--Electric power systems. An onsite electric power
system and an offsite electric power system shall be provided to
permit functioning of structures, systems, and components important
to safety. The safety function for each system (assuming the other
system is not functioning) shall be to provide sufficient capacity
and capability to assure that:
(1) specified acceptable fuel design limits and design
conditions of the reactor coolant pressure boundary are not exceeded
as a result of anticipated operational occurrences and
(2) the core is cooled and containment integrity and other vital
functions are maintained in the event of postulated accidents.
The onsite electric power supplies, including the batteries, and
the onsite electric distribution system, shall have sufficient
independence, redundancy, and testability to perform their safety
functions assuming a single failure, except for loss-of-coolant
accidents involving breaks in the reactor system coolant pressure
boundary larger than the transition break size under Sec. 50.46a,
where a single failure of the onsite power supplies and electrical
distribution system need not be assumed for plants under Sec.
50.46a. For those breaks only, neither a single failure nor the
unavailability of offsite power need be assumed.
Electric power from the transmission network to the onsite
electric distribution
[[Page 44685]]
system shall be supplied by two physically independent circuits (not
necessarily on separate rights of way) designed and located so as to
minimize, to the extent practical, the likelihood of their
simultaneous failure under operating and postulated accident and
environmental conditions. A switchyard common to both circuits is
acceptable. Each of these circuits shall be designed to be available
in sufficient time following a loss of all onsite alternating
current power supplies and the other offsite electric power circuit,
to assure that specified acceptable fuel design limits and design
conditions of the reactor coolant pressure boundary are not
exceeded. One of these circuits shall be designed to be available
within a few seconds following a loss-of-coolant accident to assure
that core cooling, containment integrity, and other vital safety
functions are maintained.
Provisions shall be included to minimize the probability of
losing electric power from any of the remaining supplies as a result
of, or coincident with, the loss of power generated by the nuclear
power unit, the loss of power from the transmission network, or the
loss of power from the onsite electric power supplies.
* * * * *
Criterion 19--Control room. A control room shall be provided
from which actions can be taken to operate the nuclear power unit
safely under normal conditions and to maintain it in a safe
condition under accident conditions, including loss-of-coolant
accidents. The necessary design, fabrication, construction, testing,
and performance criteria for structures, systems, and components
important to safety are provided to permit occupancy of the control
room under accident conditions without calculated radiation
exposures in excess of 10 rem (0.10 Sv) whole body, or its
equivalent to any part of the body, or a higher design criteria
limit established in accordance with paragraph (2) of this section
for the duration of the accident. Equipment at appropriate locations
outside the control room shall be provided
(1) with a design capability for prompt hot shutdown of the
reactor, including necessary instrumentation and controls to
maintain the unit in a safe condition during hot shutdown;
(2) with a potential capability for subsequent cold shutdown of
the reactor through the use of suitable procedures. Applicants for
and holders of construction permits and operating licenses under
this part who apply on or after January 10, 1997; applicants for
standard design approvals or certifications under part 52 of this
chapter who apply on or after January 10, 1997; applicants for and
holders of combined licenses or manufacturing licenses under part 52
of this chapter who do not reference a standard design approval or
certification; or holders of operating licenses using an alternative
source term under Sec. 50.67, shall meet the requirements of this
criterion. The necessary design, fabrication, construction, testing,
and performance criteria for structures, systems, and components
important to safety are provided to permit occupancy of the control
room under accident conditions without personnel receiving
calculated radiation exposures in excess of 10 rem (0.10 Sv) total
effective dose equivalent (TEDE) or a higher design criterion limit
established in accordance with paragraph (3) of this section for the
duration of the accident; and
(3) with a design criterion limit higher than 10 rem (0.10 Sv)
TEDE but not greater than 25 rem \[3]\ (0.25 Sv) TEDE for compliance
with paragraph (2) of this section provided the licensee
demonstrates that the specified limit is consistent with the plant
risk profile or commensurate with the risk of the plant.
* * * * *
Criterion 28--Reactivity limits. The reactivity control systems
shall be designed with appropriate limits on the potential amount
and rate of reactivity increase to assure that the effects of
postulated reactivity accidents can neither
(1) result in damage to the reactor coolant pressure boundary
greater than limited local yielding nor
(2) sufficiently disturb the core, its support structures or
other reactor pressure vessel internals to impair significantly the
capability to cool the core. These postulated reactivity accidents
shall include the maximum reactivity insertion and rate resulting
from a failure or malfunction of the reactivity control systems,
steam line rupture, changes in reactor coolant temperature and
pressure, and cold-water addition.
* * * * *
Criterion 35--Emergency core cooling. A system to provide
abundant emergency core cooling shall be provided. The system safety
function shall be to transfer heat from the reactor core following
any loss of reactor coolant at a rate such that
(1) fuel and clad damage that could interfere with continued
effective core cooling is prevented and
(2) clad metal-water reaction is limited to negligible amounts.
Suitable redundancy in components and features, and suitable
interconnections, leak detection, isolation, and containment
capabilities shall be provided to assure that for onsite electric
power system operation (assuming offsite power is not available) and
for offsite electric power system operation (assuming onsite power
is not available) the system safety function can be accomplished,
assuming a single failure, except for loss-of-coolant accidents
involving breaks in the reactor coolant system boundary larger than
the transition break size under Sec. 50.46a. For those breaks only,
neither a single failure nor the unavailability of offsite power
need be assumed.
* * * * *
Criterion 38--Containment heat removal. A system to remove heat
from the reactor containment shall be provided. The system safety
function shall be to reduce rapidly, consistent with the functioning
of other associated systems, the containment pressure and
temperature following any loss-of-coolant accident and maintain them
at acceptably low levels.
Suitable redundancy in components and features, and suitable
interconnections, leak detection, isolation, and containment
capabilities shall be provided to assure that for onsite electric
power system operation (assuming offsite power is not available) and
for offsite electric power system operation (assuming onsite power
is not available) the system safety function can be accomplished,
assuming a single failure, except for analysis of loss-of-coolant
accidents involving breaks in the reactor coolant pressure boundary
larger than the transition break size under Sec. 50.46a. For those
breaks only, neither a single failure nor the unavailability of
offsite power need be assumed.
* * * * *
Criterion 41--Containment atmosphere cleanup. Systems to control
fission products, hydrogen, oxygen, and other substances which may
be released into the reactor containment shall be provided as
necessary to reduce, consistent with the functioning of other
associated systems, the concentration and quality of fission
products released to the environment following postulated accidents,
and to control the concentration of hydrogen or oxygen and other
substances in the containment atmosphere following postulated
accidents to assure that containment integrity is maintained.
Each system shall have suitable redundancy in components and
features, and suitable interconnections, leak detection, isolation,
and containment capabilities to assure that for onsite electric
power system operation (assuming offsite power is not available) and
for offsite electric power system operation (assuming onsite power
is not available) its safety function can be accomplished, assuming
a single failure, except for analysis of loss-of-coolant accidents
involving breaks in the reactor coolant pressure boundary larger
than the transition break size under Sec. 50.46a. For those breaks
only, neither a single failure nor the unavailability of offsite
power need be assumed.
* * * * *
Criterion 44--Cooling water. A system to transfer heat from
structures, systems, and components important to safety, to an
ultimate heat sink shall be provided. The system safety function
shall be to transfer the combined heat load of these structures,
systems, and components under normal operating and accident
conditions.
Suitable redundancy in components and features, and suitable
interconnections, leak detection, and isolation capabilities shall
be provided to assure that for onsite electric power system
operation (assuming offsite power is not available) and for offsite
electric power system operation (assuming onsite power is not
available) the system safety function can be accomplished, assuming
a single failure, except for analysis of loss-of-coolant accidents
involving breaks in the reactor coolant pressure boundary larger
than the transition break size under Sec. 50.46a. For those breaks
only, neither a single failure nor the unavailability of offsite
power need be assumed.
* * * * *
Criterion 50--Containment design basis. The reactor containment
structure, including access openings, penetrations, and the
containment heat removal system shall be designed so that the
containment structure and its internal compartments can
[[Page 44686]]
accommodate, without exceeding the design leakage rate and with
sufficient margin, the calculated pressure and temperature
conditions resulting from any loss-of-coolant accident. This margin
shall reflect consideration of
(1) the effects of potential energy sources which have not been
included in the determination of the peak conditions, such as energy
in steam generators and as required by Sec. 50.44 energy from
metal-water and other chemical reactions that may result from
degradation but not total failure of emergency core cooling
functioning,
(2) the limited experience and experimental data available for
defining accident phenomena and containment responses, and
(3) the conservatism of the calculational model and input
parameters.
For reactors designed to comply with Sec. 50.46a, the
structural and leak tight integrity of the reactor containment
structure, including access openings, penetrations, and its internal
compartments, shall be maintained for realistically calculated
pressure and temperature conditions resulting from any loss-of-
coolant accident larger than the transition break size.
* * * * *
\[3]\ The use of 25 rem (0.25 Sv) whole body, or its equivalent
to any part of the body as the control room criterion is not
intended to imply that this value constitutes an acceptable limit
for emergency doses under accident conditions. This criterion is
provided only to assess the acceptability of design provisions, in
particular engineered safety features that mitigate fission product
release under postulated DBA conditions. The conditions assumed in
these analyses, although credible, are of exceedingly low
probability of occurrence and do not represent actual accident
sequences but are specified as conservative surrogates to create
bounding conditions for assessing the acceptability of engineered
safety features. Adequate radiological protection for occupationally
exposed individuals is provided by the provisions of part 20 of this
chapter and Sec. 50.47(b)(11).
0
29. In appendix E to part 50,
0
a. Revise section I;
0
b. Remove and reserve section II;
0
c. Revise paragraphs IV.3., IV.4., IV.D.2., IV.D.3., IV.E.8.b.
introductory text, IV.E.9.d., IV.F.2. introductory text, IV.F.2.a., and
IV.F.2.c. second sentence of the introductory text;
0
d. Remove and reserve paragraphs IV.5. through 7.;and
0
c. Revise footnote 1 to read as follows:
Appendix E to Part 50--Emergency Planning and Preparedness for
Production and Utilization Facilities
* * * * *
I. Introduction
1. Each applicant for an operating license is required by Sec.
50.34(b) or Sec. 53.1369 of this chapter to include in the final
safety analysis report plans for coping with emergencies. Each
applicant for a combined license under subpart C of part 52 of this
chapter or subpart H of part 53 of this chapter is required by Sec.
52.79 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.
2. This appendix establishes minimum requirements for emergency
plans for use in attaining an acceptable state of emergency
preparedness. These plans shall be submitted as part of the final
safety analysis report for an operating license. These plans, or
major features thereof, may be submitted as part of the site safety
analysis report for an early site permit.
3. The potential radiological hazards to the public associated
with the operation of non-power production or utilization facilities
licensed under this part and fuel facilities licensed under part 70
of this chapter involve considerations different than those
associated with nuclear power reactors. Consequently, the size of
Emergency Planning Zones \[1]\ (EPZs) for facilities other than
power reactors and the degree to which compliance with the
requirements of this section and sections III, IV, and V of this
appendix is necessary, will be determined on a case-by-case
basis.\[2]\
4. Notwithstanding the above paragraphs, in the case of an
operating license authorizing only fuel loading and/or low power
operations up to 5 percent of rated power, no NRC or FEMA review,
findings, or determinations concerning the state of offsite
emergency preparedness or the adequacy of and the capability to
implement State and local offsite emergency plans, as defined in
this appendix, are required prior to the issuance of such a license.
5. For power reactors with site-boundary EPZs or no EPZ, the
degree to which compliance with the requirements of this section and
sections III, IV, and V of this appendix is necessary, will be
determined on a case-by-case basis.
6. The Tennessee Valley Authority Watts Bar Nuclear Plant, Unit
2, holding a construction permit under the provisions of part 50 of
this chapter, shall meet the requirements of the final rule issued
November 23, 2011, as applicable to operating nuclear power reactor
licensees.
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.
II. [Reserved]
* * * * *
IV. * * *
3. Nuclear power reactor licensees shall use evacuation time
estimates (ETEs) and updates to the ETEs in the formulation of
predetermined, prompt protective action recommendations and shall
provide the ETEs and ETE updates to State and local governmental
authorities for use in developing offsite predetermined, prompt
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
predetermined, prompt protective action recommendations and
providing it to State and local governmental authorities for use in
developing offsite predetermined, prompt protective action
strategies.
5. [Reserved]
6. [Reserved]
7. [Reserved]
* * * * *
D. * * *
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 media sources 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 means have been established for alerting and
providing prompt instructions to the public within the plume
exposure pathway EPZ. The design objective of the alert and
notification system (ANS) 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 ANS. The alerting and notification capability shall
additionally include means for a backup method of ANS 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.
Alternatively, if the ANS capability, as documented in the ANS
Design Report, demonstrates that the primary method of alerting and
notifying is robust and has multiple methods implemented in
parallel, an independent backup method is
[[Page 44687]]
unnecessary. If a backup method is necessary, it 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. * * *
8. * * *
b. For a nuclear power reactor licensee's emergency operations
facility required by paragraph 8.a of this section, either a
facility located outside the EPZ or a backup facility outside the
EPZ if the primary facility is within the EPZ or onsite. 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:
* * * * *
9. * * *
d. Provisions for communications by the licensee with the NRC
Headquarters Operations Center from the nuclear power reactor
control room, the onsite technical support center, and the emergency
operations facility, as applicable. Such communications shall be
tested monthly.
* * * * *
F. * * *
2. The plan shall describe provisions for the conduct of
emergency preparedness drills and exercises.\[3]\ Licensees must
submit scenarios under Sec. 50.4 or Sec. 53.040 of this chapter no
later than 60 days before use in a required drill or exercise. These
scenarios must document the applicable regulation(s) the scenario is
intended to meet.
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 in order to
validate the effectiveness of the onsite, and if necessary, the
offsite emergency plans and offsite coordination.
(i) For an operating license issued under part 50 or part 53 of
this chapter, and if there are no preexisting licensed power
reactors at this site, this exercise must be conducted 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 must include participation by each State and local
government within the plume exposure pathway EPZ, as applicable. If
the applicant currently has an operating reactor at the site with
similar onsite and offsite emergency plan elements, this exercise
may be included within the operating reactor licensee's existing
exercise schedules.
(ii) For a combined license issued under part 52 or part 53 of
this chapter, this exercise must be conducted before the scheduled
date for initial loading of fuel. 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 with similar onsite and offsite emergency plan elements,
this exercise may be included within the operating reactor
licensee's existing exercise schedule. If the onsite emergency plan
elements are not similar, then paragraph 2.a.(ii) of this section
applies.
* * * * *
c. * * * 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 two years and shall, at least,
partially participate in other offsite plan exercises in this
period.\[5]\ * * *
* * * * *
\[1]\ EPZs for power reactors are discussed in NUREG-0396; EPA
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]\ Regulatory Guide 2.6, ``Emergency Planning for Research
and Test Reactors and Other Non-Power Production and Utilization
Facilities,'' may be used as guidance for the acceptability of non-
power production or utilization facility emergency response plans.
\[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.
* * * * *
0
30. In appendix K to part 50,
0
a. Amend section I by revising paragraph C.6. by removing the text
``BWR's'' and adding, in its place, the text ``BWRs''.
0
b. Amend section II by adding paragraph 6 to read as follows:
Appendix K to Part 50--ECCS Evaluation Models
* * * * *
II. * * *
6. If an entity is approved to implement Sec. 50.46a, then the
documentation requirements in Sec. 50.46a(e) apply and supersede
the requirements of section II of this appendix for that entity and
associated regulatory approval.
0
31. Add appendix T to read as follows:
Appendix T to Part 50--Streamlined Quality Assurance Criteria for
Nuclear Power Plants and Fuel Reprocessing Plants
I. Introduction and Scope
Nuclear power plants and fuel reprocessing plants include
structures, systems, and components (SSCs) 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 SSCs for applicants of these
facilities that meet certain eligibility criteria. Those applicants
may elect to use this appendix as an alternative to the quality
assurance requirements in appendix B to this part, ``Quality
Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing
Plants.'' To reference this appendix in an application for a
construction permit, operating license, or combined license, an
applicant must demonstrate it meets the following criteria:
A. The application is for an nth-of-a-kind construction permit,
operating license, or combined license that identifies the first-of-
a-kind reference plant in the application and any departures from
the first-of-a-kind reference plant;
B. Departures from the first-of-a-kind reference plant do not
change the classification, design, and method of manufacture,
construction, operation, including the design and operational
controls, of structures, systems, and components within the scope of
this appendix; and
C. Procedures and work processes for implementing the
requirements in this appendix, including a mechanism to identify and
promptly correct adverse changes in performance reliability of SSCs
within the scope of this appendix, have been established and
included in the application.
[[Page 44688]]
II. Definitions
First-of-a-kind (FOAK) means those nuclear power plants and fuel
reprocessing plants that are the initial implementation of a reactor
design or technology or fuel reprocessing plant design that has not
been previously constructed and operated at commercial scale.
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.
Non-safety-related but safety-significant (NSRSS) structures,
systems, and components (SSCs) means those SSCs that are not safety-
related but are relied on to achieve adequate defense in depth or
perform risk-significant functions and warrant special treatment.
Nth-of-a-kind (NOAK) means those nuclear power plants and fuel
reprocessing plants that are subsequent implementations of a reactor
design or technology or fuel reprocessing plant design that
reference the FOAK plant after the FOAK has been designed,
constructed, and operated.
Quality assurance means all those planned and systematic actions
necessary to provide adequate confidence that a structure, system,
or component will perform satisfactorily in service. Quality
assurance includes quality control, which comprises those actions
related to the physical characteristics of a material, structure,
component, or system to ensure the material, structure, component,
or system will meet predetermined requirements.
Quality assurance program means the overall program established
to assign responsibilities and authorities, define policies and
requirements, and provide for the performance and assessment of
work.
Quality management system means a structured framework that
documents an organization's processes, procedures, and
responsibilities for ensuring quality.
III. General Requirements
A. Integrated Quality Assurance Program. The applicant must
establish a quality assurance program, document the quality
assurance program in the quality management system, and submit the
quality management system to the NRC for review and approval. The
quality management system must meet the requirements found in
Section IV of this appendix. The applicant must implement the
approved quality management system. A quality assurance program must
be established that ensures that safety-related SSCs, as defined in
Sec. 50.2, and NSRSS SSCs, for applications under parts 50 and 52
of this chapter, and safety-related SSCs, as defined in Sec.
53.020, and NSRSS SSCs, for applications under part 53 of this
chapter, are designed, fabricated, erected, and tested to quality
standards commensurate with the importance of the safety functions
those structures, systems, and components perform. In describing the
quality assurance program, the quality management system must
identify, at the minimum, and explain in detail:
1. All relevant responsibilities and accountabilities of
personnel performing functions related to implementation of the
quality assurance program, including any delegation of such
functions.
2. Measures for ensuring that--
i. the design bases requirements, for those applications under
parts 50 and 52 of this chapter, and functional design criteria, for
those applications under part 53, of the SSCs within the scope of
the quality management system are adequately translated into
technical specifications, drawings, procedures, and instructions;
ii. the design is verified to meet technical, quality, and
regulatory requirements; and
iii. the as-built and as-operated SSCs are validated to meet the
intended function and safety margin.
3. Bidirectional communication pathways for ensuring that all
relevant requirements, expectations, concerns, and issues are
transmitted through contractual obligations and other means between
the applicant or licensee and vendors and third-party suppliers, as
appropriate.
4. Measures to--
i. ensure that procured SSCs and related services meet technical
and quality requirements; and
ii. to assess the capability of vendors or third-party suppliers
that supply the SSCs and related services to meet expectations for
product, service, or operational quality.
5. Measures established to--
i. verify and validate that products and services meet the
technical and quality requirements of the procured products and
services; and
ii. audit the vendors or third-party suppliers that are
providing the products and services.
6. Processes implemented to identify, correct, and prevent
reoccurrence of issues and failures that could adversely impact
quality, safety, and regulatory compliance.
7. Recordkeeping and documentation protocols for the quality
assurance program that support demonstrating that SSCs are properly
designed, fabricated, erected, and tested to quality standards
commensurate with the importance of the safety functions those
components perform.
B. Quality Assurance Program Development and Implementation. The
quality assurance program, as documented in the quality management
system, must:
1. Contain a graded approach for all quality assurance
activities based on the safety significance of the applicable
covered SSCs. A graded approach must be used to implement the
requirements of the quality assurance program.
i. The implementation of a graded approach must ensure quality
assurance efforts are focused in proportion to the risks associated
with a product, process, or project.
ii. A graded approach must provide a process for ensuring that
the level of analysis, documentation, and actions used to comply
with a requirement are commensurate with:
a. The relative importance to safety, safeguards, and security
of each structure, system, and component that is governed by the
quality assurance program.
b. The magnitude of any hazard involved.
c. The life-cycle stage of a facility.
d. The type of facility.
e. The particular characteristics of a facility.
f. The relative importance to radiological hazards.
g. Any other relevant factors.
iii. The basis of any graded approach must be documented for
each applicable quality assurance requirement of this regulation and
must be submitted as part of an application. The graded approach may
not be used to negate any other applicable requirements.
2. The quality management system must describe how the
requirements in Section IV of this appendix are met.
3. The applicable quality assurance requirements must be
included in procurement documents from the applicant to all relevant
contractors, vendors, suppliers, and third-parties.
4. Document the selection of American Society of Mechanical
Engineers (ASME) Nuclear Quality Assurance (NQA)-1, ``Quality
Assurance Requirements for Nuclear Facility Applications'' in
accordance with Sec. 50.55a(a)(1)(v)(B) or another appropriate
industry standard.
i. The necessary level of detail from the standard(s) must be
described to achieve quality consistent with regulatory
requirements.
ii. Gaps between the selected industry standard(s) and Section
IV of this appendix must be addressed within the quality management
system.
IV. Quality Assurance Requirements
A. Quality Assurance Criteria. For all applications under this
appendix, the quality assurance program must establish, identify,
and implement processes for meeting the following criteria.
1. Criterion 1--Management: Program.
i. Establish an organizational structure, functional
responsibilities, levels of authority, and interfaces for those
managing, performing, and assessing quality assurance activities
conducted pursuant to this appendix.
ii. Establish management processes, including planning,
scheduling, and providing resources, for quality assurance
activities conducted pursuant to this appendix.
2. Criterion 2--Management: Personnel Training and
Qualification.
i. Ensure that personnel receive training and qualifications to
be capable of performing their assigned work for activities covered
by this appendix.
ii. Ensure continuing training to personnel to maintain their
job proficiency for activities covered by this appendix.
3. Criterion 3--Management: Quality Improvement.
i. Ensure detection and prevention of quality problems for SSCs
and activities covered by this appendix.
ii. Identify, control, and correct materials, parts, or
components that do not meet established requirements.
iii. Identify the causes of problems and include prevention of
recurrence as a part of corrective action planning.
[[Page 44689]]
iv. Identify and select opportunities for improvement for the
quality assurance program.
4. Criterion 4--Management: Documents and the Associated
Records.
i. Prepare, review, approve, issue, use, and revise documents to
prescribe processes, specify requirements, or establish designs for
SSCs and activities covered by this appendix.
ii. Specify, prepare, review, approve, and maintain these
documents as records.
5. Criterion 5--Performance: Work Processes.
i. Perform work consistent with technical standards,
administrative controls, and other hazard controls adopted to meet
regulatory requirements using approved instructions, procedures, or
other appropriate means.
ii. Identify and control materials, parts, and components,
including partially fabricated assemblies, to ensure proper use.
iii. Maintain material and equipment to prevent damage, loss, or
deterioration.
iv. Calibrate and maintain equipment used for activities
affecting quality.
6. Criterion 6--Performance: Design.
i. Design of SSCs and design processes using sound engineering
or scientific principles and appropriate standards.
ii. Incorporate applicable requirements, design bases, as
defined in Sec. 50.2, and functional design criteria in design work
and design changes.
iii. Identify and control design interfaces.
iv. Verify and validate the adequacy of the design of
structures, systems, and components using individuals or groups
other than those who performed the work.
v. Verify the adequacy of the design of structures, systems, and
components before approval and implementation of the design.
vi. Validate the design of the structure, system and components
before relying on the structures, systems, and components to perform
their intended safety function.
7. Criterion 7--Performance: Procurement.
i. Procure items and services that meet established requirements
and verify items and services perform as specified.
ii. Evaluate and select prospective suppliers on the basis of
specified criteria that will ensure the requirements of this
appendix are met.
iii. Ensure that approved suppliers continue to provide
acceptable items and services.
8. Criterion 8--Performance: Inspection and Acceptance Testing.
i. Inspect and test specified items, services, and processes
using established acceptance and performance criteria.
ii. Calibrate and maintain equipment used for inspections and
tests.
9. Criterion 9--Performance: Maintenance of structures, systems,
and components. Control the storage of structures, systems, and
components consistent with appropriate cleanliness and environmental
standards.
10. Criterion 10--Assessment: Management Assessment. Ensure that
managers assess their management processes and identify and correct
problems that hinder the organization from achieving its objectives.
11. Criterion 11--Assessment: Independent Assessment.
i. Plan and conduct independent assessments to measure item and
service quality, the adequacy of work performance, and to promote
improvement.
ii. Establish sufficient authority and freedom from line
management for independent assessment teams.
iii. Ensure personnel who perform independent assessments are
technically qualified and knowledgeable in the areas to be assessed.
B. Quality Assurance for Software Used in Design and Analysis,
and Digital Items Important to Safety. Processes for software
quality assurance must be identified, established, and implemented
within the quality assurance program and documented in the quality
management system.
1. Software quality assurance processes must be applied to
software used--
i. in digital items for SSCs within the scope of this appendix;
ii. for design verification for SSCs within the scope of this
appendix; and
iii. for design analysis of SSCs within the scope of this
appendix.
2. Software used for the applications identified in paragraph 1
of this section must be documented, managed, and controlled
throughout its life cycle to ensure that the safety functions will
be performed under design basis conditions.
3. Appropriate national or international software engineering
standard(s) must be used (e.g., ASME, Institute for Electrical and
Electronics Engineers (IEEE), National Institutes of Standards and
Technology (NIST), American Nuclear Society (ANS), etc.). These
standards may include combinations of standards.
PART 51--ENVIRONMENTAL PROTECTION REGULATIONS FOR DOMESTIC
LICENSING AND RELATED REGULATORY FUNCTIONS
0
32. The authority citation for part 51 continues 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.
Sections 51.20, 51.30, 51.60, 51.80, and 51.97 also issued under
Nuclear Waste Policy Act secs. 135, 141, 148 (42 U.S.C. 10155,
10161, 10168).
Section 51.22 also issued under Atomic Energy Act sec. 274 (42
U.S.C. 2021) and under Nuclear Waste Policy Act sec. 121 (42 U.S.C.
10141).
Sections 51.43, 51.67, and 51.109 also issued under Nuclear
Waste Policy Act sec. 114(f) (42 U.S.C. 10134(f)).
0
33. In 51.4, revise the definition for ``Construction'' to read as
follows:
Sec. 51.4 Definitions.
* * * * *
Construction means:
(1) For production and utilization facilities, the activities in
paragraph (1)(i) of this definition.
(i) Activities constituting construction are 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:
(A) Safety-related structures, systems, or components (SSCs) of a
facility, as defined in Sec. 50.2 of this chapter;
(B) SSCs that perform safety-significant functions; and
(C) SSCs necessary to comply with part 73 of this chapter.
(ii) With respect to production or utilization facilities, other
than testing facilities and nuclear power plants, required to be
licensed under section 104a. or section 104c. of the Act, construction
does not include the erection of buildings which will be used for
activities other than operation of a facility and which may also be
used to house a facility (e.g., the construction of a college
laboratory building with space for installation of a training reactor).
(iii) Any activities that are determined to be outside the scope of
those defined in the definition of construction in Sec. 51.4 and that
are undertaken by an applicant or on its behalf are entirely at the
risk of the applicant and has no bearing on the issuance of an
environmental assessment, environmental impact statement or finding of
no significant impact with respect to NRC's regulated activities under
the requirements of the Act, and rules, regulations issued under the
Act.
(2) For materials licenses, taking any site-preparation activity at
the site of a facility subject to the regulations in parts 30, 36, 40,
and 70 of this chapter that has a reasonable nexus to radiological
health and safety or the common defense and security; provided,
however, that construction does not mean taking any other action that
has no reasonable nexus to radiological health and safety or the common
defense and security.
* * * * *
0
34. In Sec. 51.51, revise table note 1 of table S-3 to read as
follows:
Sec. 51.51 Uranium fuel cycle environmental data--Table S-3.
* * * * *
(b) * * *
[[Page 44690]]
Table S-3--Table of Uranium Fuel Cycle Environmental Data \[1]\
[Normalized to model LWR annual fuel requirement [WASH-1248] or
reference reactor year [NUREG-0116]] [See footnotes at end of this
table]
* * * * * * *
\[1]\ In some cases where no entry appears it is clear from the
background documents that the matter was addressed and that, in
effect, the table should be read as if a specific zero entry had been
made. However, there are other areas that are not addressed at all in
the table. Table S-3 does not include health effects from the
effluents described in the table or estimates of releases of Radon-222
from the uranium fuel cycle or estimates of Technetium-99 released
from waste management or reprocessing activities. These issues may be
the subject of litigation in the individual licensing proceedings.
Data supporting this table are given in the ``Environmental Survey of
the Uranium Fuel Cycle,'' WASH-1248, April 1974; the ``Environmental
Survey of the Reprocessing and Waste Management Portion of the LWR
Fuel Cycle,'' NUREG-0116 (Supp.1 to WASH-1248); the ``Public Comments
and Task Force Responses Regarding the Environmental Survey of the
Reprocessing and Waste Management Portions of the LWR Fuel Cycle,''
NUREG-0216 (Supp. 2 to WASH-1248); and in the record of the final
rulemaking pertaining to Uranium Fuel Cycle Impacts from Spent Fuel
Reprocessing and Radioactive Waste Management, Docket RM-50-3. The
contributions from reprocessing, waste management and transportation
of wastes are maximized for either of the two fuel cycles (uranium
only and no recycle). The contribution from transportation excludes
transportation of cold fuel to a reactor and of irradiated fuel and
radioactive wastes from a reactor which are considered in table S-4 of
Sec. 51.20(g). The contributions from the other steps of the fuel
cycle are given in columns A-E of table S-3A of WASH-1248.
The analysis in NUREG-2249, ``Generic Environmental Impact Statement for
Licensing of New Nuclear Reactors--Final Report,'' December 2025,
extends this table up to an enrichment of 20.0 weight percent uranium-
235 for environmental effects of uranium recovery (which replaced
uranium mining and milling), the production of uranium hexafluoride,
gaseous centrifuge isotopic enrichment (which replaced gaseous
diffusion isotopic enrichment), and fuel fabrication. The analysis in
NUREG-2266, ``Environmental Evaluation of Accident Tolerant Fuels with
Increased Enrichment and Higher Burnup Levels,'' July 2024, extends
the assembly averaged level of burnup of the irradiated fuel from the
reactor of up to 80,000 megawatt-days per metric ton as related to
uranium fuel cycle activities.
* * * * *
0
35. Revise and republish Sec. 51.52 to read as follows:
Sec. 51.52 Environmental effects of transportation of fuel and
waste--Table S-4.
Under Sec. 51.50, every environmental report prepared for the
construction permit stage or early site permit stage or combined
license stage of a light-water-cooled and other than light-water-cooled
nuclear power reactor, and submitted after February 4, 1975, must
contain a statement concerning transportation of fuel and radioactive
wastes to and from the reactor. That statement must indicate that the
reactor and this transportation meet either all of the conditions in
paragraph (a) of this section or all of the conditions of paragraph (b)
of this section.
(a)
(1) The reactor has a core thermal power level not exceeding 3,800
megawatts;
(2) The reactor fuel is in the form of sintered uranium dioxide
pellets having a uranium-235 enrichment not exceeding 8% by weight, and
the pellets are encapsulated in zircaloy rods;
(3) The average level of irradiation of the irradiated fuel from
the reactor does not exceed 80,000 megawatt-days per metric ton, and no
irradiated fuel assembly is shipped until at least 90 days after it is
discharged from the reactor;
(4) With the exception of irradiated fuel, all radioactive waste
shipped from the reactor is packaged and in a solid form;
(5) Unirradiated fuel is shipped to the reactor by truck;
irradiated fuel is shipped from the reactor by truck, rail, or barge;
and radioactive waste other than irradiated fuel is shipped from the
reactor by truck or rail; and
(6) The environmental impacts of transportation of fuel and waste
to and from the reactor, with respect to normal conditions of transport
and possible accidents in transport, are as set forth in summary table
S-4 in paragraph (c) of this section; and the values in the table
represent the contribution of the transportation to the environmental
costs of licensing the reactor.
(b) For light-water-cooled and other than light-water-cooled
reactors not meeting the conditions of paragraph (a) of this section,
the statement must contain a full description and detailed analysis of
the environmental effects of transportation of fuel and wastes to and
from the reactor, including values for the environmental impact under
normal conditions of transport and for the environmental risk from
accidents in transport. The statement must indicate that the values
determined by the analysis represent the contribution of such effects
to the environmental costs of licensing the reactor.
(c)
BILLING CODE 7590-01-P
Summary Table S-4--Environmental Impact of Transportation of Fuel and
Waste To and From One Nuclear Power Reactor [1], [2] Normal
Conditions of Transport
[GRAPHIC] [TIFF OMITTED] TP16JY26.526
[[Page 44691]]
[GRAPHIC] [TIFF OMITTED] TP16JY26.527
Accidents in Transport
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BILLING CODE 7590-01-C
PART 52--LICENSES, CERTIFICATIONS, AND APPROVALS FOR NUCLEAR POWER
PLANTS
0
36. The authority citation for part 52 continues to read as follows:
Authority: Atomic Energy Act of 1954, secs. 103, 104, 147, 149,
161, 181, 182, 183, 185, 186, 189, 223, 234 (42 U.S.C. 2133, 2134,
2167, 2169, 2201, 2231, 2232, 2233, 2235, 2236, 2239, 2273, 2282);
Energy Reorganization Act of 1974, secs. 201, 202, 206, 211 (42
U.S.C. 5841, 5842, 5846, 5851); 44 U.S.C. 3504 note.
0
37. In Sec. 52.1, in paragraph (a), add in alphabetical order the
definitions ``Tier 1'', ``Tier 2'', and ``Tier 2*'' to read as follows:
Sec. 52.1 Definitions.
(a) * * *
Tier 1 means, for design certifications issued after [EFFECTIVE
DATE OF FINAL RULE], the qualitative and functional-level portion of
the design-related information contained in the generic design control
document, including ITAAC, that is approved and certified by a standard
design certification. The design descriptions, interface requirements,
and site parameters are derived from Tier 2 information. For design
certifications issued prior to [EFFECTIVE DATE OF FINAL RULE], see the
definition of this term in the applicable appendix to this part.
Tier 2 means, for design certifications issued after [EFFECTIVE
DATE OF FINAL RULE], the portion of the design-related information
contained in the generic design control document that is approved, but
not certified, by a standard design certification. Compliance with Tier
2 is required, but generic changes to, and plant specific departures
from, Tier 2 are governed by the process set out in the applicable
appendix to this part. Compliance with Tier 2 provides a sufficient,
but not the only acceptable, method for complying with Tier 1.
Compliance methods differing from Tier 2 must satisfy the change
process in the applicable appendix to this part. Regardless of these
differences, an applicant or licensee must meet the requirement in the
applicable appendix to this part to reference Tier 2 when referencing
Tier 1. For design certifications issued prior to [EFFECTIVE DATE OF
FINAL RULE], see the definition of this term in the applicable appendix
to this part.
Tier 2* means, for design certifications issued after [EFFECTIVE
DATE OF FINAL RULE], the portion of the Tier 2 information containing
the qualitative and functional-level portion of design-related
information, designated as such in the generic design control document,
that is subject to the change process for such information that is
specified in a standard design certification rule. After a plant first
achieves full power, Tier 2* information in the plant-specific design
control document for that plant reverts to Tier 2 status and is
thereafter subject to the change and departure provisions for Tier 2
information. For design certifications issued prior to [EFFECTIVE DATE
OF FINAL RULE], see the definition of this term in the applicable
appendix to this part.
* * * * *
Sec. 52.15 [Amended]
0
38. In Sec. 52.15, in paragraph (c), remove the word ``renewal'' and
add in its place the word ``amendment''.
0
39. In Sec. 52.17, revise paragraph (a)(1)(ix) and footnotes 1 and 2
to read as follows:
Sec. 52.17 Contents of applications; technical information.
(a) * * *
(1) * * *
(ix) A description and safety assessment of the site on which a
facility is to be located. The assessment must contain an analysis and
evaluation of the major structures, systems, and components of the
facility that bear significantly on the acceptability of the site under
the radiological consequence evaluation factors identified in
paragraphs (a)(1)(ix)(A) and (a)(1)(ix)(B) of this section. In
performing this assessment, an applicant shall assume a fission product
release \[1]\ assuming that the facility is operated at the ultimate
power level contemplated. The applicant shall perform an evaluation and
analysis of the postulated fission product release, using the expected
demonstrable leakage rates from potential flow paths and any fission
product cleanup systems intended to mitigate the consequences of the
accidents, together with applicable site characteristics, including
site meteorology, to evaluate the offsite radiological consequences.
Site characteristics must comply with part 100 of this chapter. The
evaluation must determine that:
(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 \[2]\ (0.25 Sv) total effective dose equivalent
(TEDE).
(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
(0.25 Sv) TEDE;
* * * * *
\[1]\ The fission product release assumed for this evaluation
should be based upon a major accident, hypothesized for purposes of
site analysis or postulated from considerations of possible
accidental events to bound a broad range of design basis accidents.
These accidents have generally been assumed to result in substantial
meltdown of the core with subsequent release of appreciable
quantities of fission products.
\[2]\ The use of 25 rem (0.25 Sv) 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.
0
40. Revise Sec. 52.18 to read as follows:
Sec. 52.18 Standards for review of applications.
Applications for the initial issuance of an early site permit filed
under this subpart will be reviewed according to the applicable
standards set out in part 50 of this chapter and its appendices and
part 100 of this chapter. In addition, the Commission shall prepare an
environmental impact statement during review of the application, in
accordance with the provisions of part 51 of this chapter. The
Commission shall determine, after consultation with Federal Emergency
Management Agency, as applicable, whether the information required of
the applicant by Sec. 52.17(b)(1) shows that there is not a
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. 52.17(b)(2)(i) are acceptable 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, and whether any emergency plans submitted by
the applicant under Sec. 52.17(b)(2)(ii) provide reasonable assurance
that adequate protective measures can and will be taken in the event of
a radiological emergency.
Sec. 52.25 [Amended]
0
41. In Sec. 52.25, in the first sentence, remove the phrase
``performed and the site is not referenced in an application
[[Page 44693]]
for a construction permit or a combined license issued under subpart C
of this part while the permit remains valid,'' and add in its place the
phrase ``performed, and the early site permit holder has applied for
termination,''.
0
42. In Sec. 52.26, revise paragraph (a) and remove and reserve
paragraph (b) to read as follows:
Sec. 52.26 Duration of permit.
(a) An early site permit issued under this subpart will be issued
with no fixed term.
(b) [Reserved]
* * * * *
0
43. Revise Sec. 52.29 to read as follows:
Sec. 52.29 Application for amendment to update an early site permit.
(a) An early site permit holder may choose to submit an application
to amend an early site permit to update the data and information on
which the permit is based at any time after issuance of the early site
permit. The early site permit holder may provide updated information on
as many issues as the early site permit holder chooses and may request
to extend the period for which the agency will afford those issues
finality up to 20 additional years from the date of the amendment's
issuance. The early site permit holder may request such an extension
for an already extended permit. The application must meet the
requirements of Sec. Sec. 50.90 and 50.92 of this chapter.
(b) An application submitted under paragraph (a) of this section
must contain all information necessary to bring up to date the
information and data contained in the previous application for those
issues the early site permit holder has chosen to update.
(c) Each application must include an environmental report as
required by part 51 of this chapter, or a request and justification for
a categorical exclusion under part 51 of this chapter.
(d) Any person whose interest may be affected by the update of the
permit may request a hearing on the application for the update. 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. 2.309 of this chapter.
0
44. Revise Sec. 52.31 to read as follows:
Sec. 52.31 Issuance of amendment to update an early site permit.
The Commission shall grant amendment of an early site permit only
if it determines that:
(a) 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, except early site permits issued before [EFFECTIVE
DATE OF THE FINAL RULE] are no longer subject to Sec. 52.26(a); and
(b) Any new requirements the Commission may wish to impose are
necessary for adequate protection to public health and safety or common
defense and security.
Sec. 52.33 [Removed and Reserved]
0
45. Remove and reserve Sec. 52.33.
0
46. Revise Sec. 52.35 to read as follows:
Sec. 52.35 Use of site for other purposes.
(a) A site for which an early site permit has been issued under
this subpart may be used for purposes other than those described in the
permit, including the location of other types of energy facilities. The
permit holder shall 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, in accordance with the provisions of Sec. 52.39.
(b) If the permit holder no longer intends to use the site for a
nuclear power plant or for other reasons no longer wishes to hold the
permit, as described in the request, the permit holder may at any time
request the Director to terminate the early site permit. The request to
terminate the permit must comply with the filing requirements of
Sec. Sec. 52.3 and 50.30 of this chapter, and identify the applicable
requirements for site redress of Sec. 52.25. Upon request, the
Director may terminate the permit.
(c) Termination of the early site permit does not bar the permit
holder or another applicant from filing a new application for the site.
0
47. In Sec. 52.39,
0
a. In paragraph (a)(1), remove the references ``Sec. Sec. 52.26 or
52.33'' and add in its place the reference ``Sec. 52.26'';
0
b. In paragraph (a)(2) introductory text, remove the word ``renewal''
and add in its place the word ``amendment'';
0
c. Revise paragraphs (c)(1)(iv) and (v);
0
d. Add paragraph (c)(1)(vi);
0
e. In paragraph (d), remove the word ``renewed'' and add in its place
the word ``amended''; and
0
f. In paragraph (e), remove the references ``10 CFR 50.90 and 50.92''
and add in its place the references ``Sec. Sec. 50.90, 50.92, and
52.29 of this chapter''.
The revisions and additions are to read as follows:
Sec. 52.39 Finality of early site permit determinations.
* * * * *
(c) * * *
(1) * * *
(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;
(v) The information as required in the site safety analysis report
in accordance with Sec. 52.17(a)(1)(vi) through (ix) has not been
updated after 20 years from the date of early site permit issuance or a
previous update of the permit by amendment, whichever is later; and
(vi) (A) Any significant environmental issue that was not resolved
in the early site permit proceeding;
(B) For an application that references an early site permit issued
or updated by amendment, whichever is later, no more than 20 years
before the submission of the application, 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; or
(C) For an application that references an early site permit, issued
or updated by amendment, whichever is later, more than 20 years before
submission of the application, any issue involving the impacts of
construction and operation of the facility regardless of whether the
early site permit proceeding resolved the issue.
* * * * *
0
48. In Sec. 52.47,
0
a. Revise paragraph (a)(2)(iv);
0
b. In paragraph (a)(4), remove the last sentence; and
0
c. Redesignate footnotes 3 and 4 as footnotes 1 and 2 and revise
footnotes 1 and 2 to read as follows:
Sec. 52.47 Contents of applications; technical information.
* * * * *
(a) * * *
(2) * * *
(iv) The safety features that are to be engineered into the
facility and those barriers that must be breached as a result of an
accident before a release of radioactive material to the environment
can occur. Special attention must be directed to plant design features
intended to mitigate the radiological consequences of accidents. In
performing this assessment, an
[[Page 44694]]
applicant shall assume a fission product release \[1]\ assuming that
the facility is operated at the ultimate power level contemplated. The
applicant shall perform an evaluation and analysis of the postulated
fission product release, using the expected demonstrable leakage rates
from potential flow paths and any fission product cleanup systems
intended to mitigate the consequences of the accidents, together with
applicable postulated site parameters, including site meteorology, to
evaluate the offsite radiological consequences. The evaluation must
determine that:
(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 \[2]\ (0.25 Sv) total effective dose equivalent
(TEDE);
(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
(0.25 Sv) TEDE;
* * * * *
\[1]\ The fission product release assumed for this evaluation
should be based upon a major accident, hypothesized for purposes of
site analysis or postulated from considerations of possible
accidental events to bound a broad range of design basis accidents.
These accidents have generally been assumed to result in substantial
meltdown of the core with subsequent release of appreciable
quantities of fission products.
\[2]\ The use of 25 rem (0.25 Sv) 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.
0
49. In Sec. 52.54, revise paragraph (a) introductory text and
paragraph (b) to read as follows:
Sec. 52.54 Issuance of standard design certification.
(a) After conducting a rulemaking proceeding under Sec. 52.51 on
an application for a standard design certification and receiving the
report to be submitted by the Advisory Committee on Reactor Safeguards
under Sec. 52.53, 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:
* * * * *
(b) The standard design certification rule must specify the site
parameters, design characteristics, and any additional requirements and
restrictions of the standard design certification rule. A standard
design certification rule that was reviewed and approved as meeting the
requirements of Sec. 50.46a of this chapter must specify the criteria
governing departures that a referencing combined license must meet. The
criteria must ensure that the safety bases for the NRC's approval of
the certified design's compliance with Sec. 50.46a of this chapter
(including applicability of the transition break size) continue to
apply despite the departure.
* * * * *
0
50. In Sec. 52.63,
0
a. In paragraph (a)(1)(v) at the end of the sentence, add the word
``or'' after ``information;'';
0
b. In paragraph (a)(1)(vi) at the end of the sentence, remove the
phrase ``security; or'' and add in its place the word ``security.'';
0
c. Remove paragraph (a)(1)(vii); and
0
d. Revise paragraphs (a)(4)(ii) and (b)(1) to read as follows:
Sec. 52.63 Finality of standard design certifications.
(a) * * *
(4) * * *
(ii) Special circumstances as defined in Sec. 52.7 are present.
* * * * *
(b)
(1) An applicant or licensee who references a design certification
rule may request an exemption from one or more elements of the
certification information if one is required by the applicable change
process within the referenced design certification rule. The Commission
may grant such a request only if it determines that the exemption will
comply with the requirements of Sec. 52.7. The granting of an
exemption on request of an applicant is subject to litigation in the
same manner as other issues in the operating license or combined
license hearing.
* * * * *
0
51. In Sec. 52.79,
0
a. Revise and republish paragraphs (a)(1)(vi), (a)(2)(iv), (a)(5),
(a)(21), (a)(25), (a)(27), (a)(36)(i), and (b);
0
b. Add paragraph (a)(48); and
0
c. Redesignate footnotes 5 through 8 as footnotes 1 through 4 and
revise footnotes 1 through 4 to read as follows:
Sec. 52.79 Contents of applications; technical information in final
safety analysis report.
(a) * * *
(1) * * *
(vi) A description and safety assessment of the site on which the
facility is to be located. The assessment must contain an analysis and
evaluation of the major structures, systems, and components of the
facility that bear significantly on the acceptability of the site under
the radiological consequence evaluation factors identified in
paragraphs (a)(1)(vi)(A) and (a)(1)(vi)(B) of this section. In
performing this assessment, an applicant shall assume a fission product
release \[1]\ assuming that the facility is operated at the ultimate
power level contemplated. The applicant shall perform an evaluation and
analysis of the postulated fission product release, using the expected
demonstrable leakage rates from potential flow paths and any fission
product cleanup systems intended to mitigate the consequences of the
accidents, together with applicable site characteristics, including
site meteorology, to evaluate the offsite radiological consequences.
Site characteristics must comply with part 100 of this chapter. The
evaluation must determine that:
(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 \[2]\ (0.25 Sv) total effective dose equivalent
(TEDE).
(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
(0.25 Sv) TEDE; and
(2) * * *
(iv) The safety features that are to be engineered into the
facility and those barriers that must be breached as a result of an
accident before a release of radioactive material to the environment
can occur. Special attention must be directed to plant design features
intended to mitigate the radiological consequences of accidents. In
performing this assessment, an applicant shall assume a fission product
release \[3]\ assuming that the facility is operated at the ultimate
power level contemplated;
* * * * *
(5) An analysis and evaluation of the design and performance of
structures, systems, and components with the objective of assessing the
risk to public health and safety resulting from operation of the
facility and including determination of the margins of safety
[[Page 44695]]
during normal operations and transient conditions anticipated during
the life of the facility, and the adequacy of structures, systems, and
components provided for the prevention of accidents and the mitigation
of the consequences of accidents.
* * * * *
(21) Emergency plans complying 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;
* * * * *
(25) A description of the quality assurance program or a quality
management system, applied to the design, and to be applied to the
fabrication, construction, and testing, of the structures, systems, and
components of the facility. Appendix B to part 50 of this chapter sets
forth the requirements for quality assurance programs for nuclear power
plants. The description of the quality assurance program for a nuclear
power plant must include a discussion of how the applicable
requirements of appendix B to part 50 of this chapter have been and
will be satisfied, including a discussion of how the quality assurance
program will be implemented or for eligible combined license
applications. Appendix T to part 50 of this chapter, ``Streamlined
Quality Assurance Criteria for Nuclear Power Plants and Fuel
Reprocessing Plants,'' sets forth streamlined requirements for quality
assurance programs for nuclear power plants and fuel reprocessing
plants that an eligible combined license applicant may voluntarily use
as an alternative to appendix B to part 50 of this chapter. The quality
management system, required by appendix T to part 50 of this chapter,
for a nuclear power plant or fuel reprocessing plant shall include
discussions of how the applicable requirements of appendix T to part 50
of this chapter will be satisfied;
* * * * *
(27) Managerial and administrative controls to be used to assure
safe operation. Appendix B to part 50 of this chapter sets forth the
requirements for these controls for nuclear power plants. The
information on the controls to be used for a nuclear power plant shall
include a discussion of how the applicable requirements of appendix B
to part 50 of this chapter will be satisfied. Appendix T to part 50 of
this chapter, ``Streamlined Quality Assurance Criteria for Nuclear
Power Plants and Fuel Reprocessing Plants,'' sets forth streamlined
requirements for such controls for nuclear power plants and fuel
reprocessing plants that an eligible combined license applicant may
voluntarily use as an alternative to appendix B to part 50 of this
chapter. The quality management system, required by appendix T to part
50 of this chapter, for a nuclear power plant or fuel reprocessing
plant shall include discussions of how the applicable requirements of
appendix T to part 50 of this chapter will be satisfied;
* * * * *
(36) (i) A safeguards contingency plan in accordance with the
criteria set forth in appendix C to part 73 of this chapter. The
safeguards contingency plan shall include plans for dealing with
threats, thefts, and radiological sabotage, as defined in part 73 of
this chapter, relating to the special nuclear material and nuclear
facilities licensed under this chapter and in the applicant's
possession and control. Each application for this type of license shall
include the information contained in the applicant's safeguards
contingency plan.\[4]\ (Implementing procedures required for this plan
need not be submitted for approval.)
* * * * *
(48) An applicant may include in its application a request for
generic finality, to generic aspects of the design under this part,
such that information in the application, if approved by the NRC, is
considered resolved in other proceedings where information approved for
generic finality is referenced. An application for a combined license
that requests generic finality must include applicable site parameters
postulated for the design, including the design-basis external hazard
levels for the relevant external hazards, and an analysis and
evaluation of the design in terms of those site parameters.
(b) If the combined license application references an early site
permit, then the following requirements apply:
(1) The final safety analysis report need not contain information
or analyses submitted to the Commission in connection with the early
site permit, provided, however, that the final safety analysis report
must either include or incorporate by reference the early site permit
site safety analysis report and must 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 final safety analysis report does not demonstrate that
design of the facility falls within the site characteristics and design
parameters, the application shall include a request for a variance that
complies with the requirements of Sec. Sec. 52.39 and 52.93.
(3) If the early site permit site safety analysis report
information required by Sec. 52.17(a)(1)(vi) through (ix) has not been
updated after 20 years from the date of early site permit issuance or a
previous update of the permit by amendment, whichever is later, the
combined license application shall include updated information and
revised analyses, as necessary, in the final safety analysis report.
The referencing application need not update the postulated source term
stated in the early site permit if it falls within the source term
derived from the design. If the postulated source term stated in the
early site permit does not fall within the source term derived from the
design, the referencing application must propose a variance from the
early site permit that complies with the requirements of Sec. Sec.
52.39 and 52.93.
(4) The final safety analysis report must demonstrate that all
terms and conditions that have been included in the early site permit,
other than those imposed under Sec. 50.36b of this chapter, will be
satisfied by the date of issuance of the combined license. Any terms or
conditions of the early site permit that could not be met by the time
of issuance of the combined license, must be set forth as terms or
conditions of the combined license.
(5) If the early site permit approves complete and integrated
emergency plans, or major features of emergency plans, then the final
safety analysis report must include any new or additional information
that updates and corrects the information that was provided under Sec.
52.17(b), 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 reduction in effectiveness under Sec. 50.54(q) of this
chapter.
(6) If complete and integrated emergency plans are approved as part
of the early site permit, new certifications meeting the requirements
of paragraph (a)(22) of this section are not required.
* * * * *
\[1]\ The fission product release assumed for this evaluation
should be based upon a major accident, hypothesized for purposes of
site analysis or postulated from considerations of possible
accidental events to bound a broad
[[Page 44696]]
range of design basis accidents. These accidents have generally been
assumed to result in substantial meltdown of the core with
subsequent release of appreciable quantities of fission products.
\[2]\ The use of 25 rem (0.25 Sv) 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.
\[3]\ The fission product release assumed for this evaluation
should be based upon a major accident, hypothesized for purposes of
site analysis or postulated from considerations of possible
accidental events to bound a broad range of design basis accidents.
These accidents have generally been assumed to result in substantial
meltdown of the core with subsequent release of appreciable
quantities of fission products.
\[4]\ A physical security plan that contains all the information
required in both Sec. 73.55 of this chapter and appendix C to 10
CFR part 73 satisfies the requirement for a contingency.
0
52. Revise Sec. 52.85 to read as follows:
Sec. 52.85 Administrative review of applications; hearings.
(a) A proceeding on a combined license is subject to all applicable
procedural requirements contained in part 2 of this chapter, 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 combined license, then those ITAAC will be identified
in the notice of hearing. All hearings on combined licenses are
governed by the procedures contained in part 2 of this chapter.
(b) If an applicant requests generic finality under Sec.
52.79(a)(48), the Commission will include a request for generic
finality as a proposed action in the notice of hearing required by
Sec. 2.104 of this chapter.
Sec. 52.93 [Amended]
0
53. In Sec. 52.93, in paragraph (c) remove the last sentence.
0
54. In Sec. 52.97, add paragraph (d) to read as follows:
Sec. 52.97 Issuance of combined licenses.
* * * * *
(d) The Commission may afford generic finality to generic aspects
of the design of a utilization facility, including postulated site
parameters, and requirements submitted pursuant to Sec. 52.79(a)(48),
if it finds that the proposed generic design can be constructed and
operated at sites having characteristics that fall within the site
parameters postulated for the design in accordance with applicable
requirements and without undue risk to the health and safety of the
public.
0
55. In Sec. 52.98, revise paragraph (b), remove and reserve paragraph
(d), and add paragraph (h) to read as follows:
Sec. 52.98 Finality of combined licenses; information requests.
* * * * *
(b) If the combined license does not reference a design
certification, then a licensee may make changes in the facility as
described in the final safety analysis report (as updated), make
changes in the procedures as described in the final safety analysis
report (as updated), and conduct tests or experiments not described in
the final safety analysis report (as updated) under the applicable
change processes in part 50 of this chapter (e.g., Sec. 50.54, Sec.
50.59, or Sec. 50.90 of this chapter).
* * * * *
(d) [Reserved]
* * * * *
(h) In a proceeding for the issuance of a construction permit,
operating license, or combined license, or in any enforcement hearing
other than one initiated by the Commission under paragraph (a) of this
section, in which a combined license issued under this subpart is
referenced, the Commission must treat as resolved those matters
resolved in the proceeding on the application for issuance or renewal
of the referenced combined license, including, if applicable, the
adequacy of a reactor design where the referenced combined license was
afforded finality pursuant to Sec. 52.97(d).
0
56. In Sec. 52.137,
0
a. Revise and republish paragraph (a)(2)(iv);
0
b. In paragraph (a)(4), remove the word ``SSC'' and add in its place
the word ``SSCs'' and remove the last sentence; and
0
c. Redesignate footnotes 9 and 10 as footnotes 1 and 2 and revise
footnotes 1 and 2 to read as follows:
Sec. 52.137 Contents of applications; technical information.
* * * * *
(a) * * *
(2) * * *
(iv) The safety features that are to be engineered into the
facility and those barriers that must be breached as a result of an
accident before a release of radioactive material to the environment
can occur. Special attention must be directed to plant design features
intended to mitigate the radiological consequences of accidents. In
performing this assessment, an applicant shall assume a fission product
release \[1]\ assuming that the facility is operated at the ultimate
power level contemplated. The applicant shall perform an evaluation and
analysis of the postulated fission product release, using the expected
demonstrable leakage rates from potential flow paths and any fission
product cleanup systems intended to mitigate the consequences of the
accidents, together with applicable postulated site parameters,
including site meteorology, to evaluate the offsite radiological
consequences. The evaluation must determine that:
(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 \[2]\ (0.25 Sv) 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
(0.25 Sv) TEDE;
* * * * *
\[1]\ The fission product release assumed for this evaluation
should be based upon a major accident, hypothesized for purposes of
site analysis or postulated from considerations of possible
accidental events to bound a broad range of design basis accidents.
These accidents have generally been assumed to result in substantial
meltdown of the core with subsequent release of appreciable
quantities of fission products.
\[2]\ The use of 25 rem (0.25 Sv) 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.
0
57. In Sec. 52.157,
0
a. Revise and republish paragraph (d);
0
b. In paragraph (f)(1), remove the last sentence;
0
c. Redesignate footnotes 11 and 12 as footnotes 1 and 2 and revise
footnotes 1 and 2.
The addition and revisions are to read as follows:
[[Page 44697]]
Sec. 52.157 Contents of applications; technical information in final
safety analysis report.
* * * * *
(d) The safety features that are engineered into the reactor and
those barriers that must be breached as a result of an accident before
a release of radioactive material to the environment can occur. Special
attention must be directed to reactor design features intended to
mitigate the radiological consequences of accidents. In performing this
assessment, an applicant shall assume a fission product release \[1]\
assuming that the facility is operated at the ultimate power level
contemplated. The applicant shall perform an evaluation and analysis of
the postulated fission product release, using the expected demonstrable
leakage rates from potential flow paths and any fission product cleanup
systems intended to mitigate the consequences of the accidents,
together with applicable postulated site parameters, including site
meteorology, to evaluate the offsite radiological consequences. The
evaluation must determine that:
(1) 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 \[2]\ (0.25 Sv) total effective dose equivalent
(TEDE); and
(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
(0.25 Sv) TEDE;
* * * * *
\[1]\ The fission product release assumed for this evaluation
should be based upon a major accident, hypothesized for purposes of
site analysis or postulated from considerations of possible
accidental events to bound a broad range of design basis accidents.
These accidents have generally been assumed to result in substantial
meltdown of the core with subsequent release of appreciable
quantities of fission products.
\[2]\ The use of 25 rem (0.25 Sv) 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.
0
58. Revise and republish Sec. 52.158 to read as follows:
Sec. 52.158 Contents of application; additional technical
information.
(a) Inspections, tests, analyses, and acceptance criteria (ITAAC).
(1) The application must contain the proposed inspections, tests, and
analyses that the licensee who will be operating the reactor shall
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
manufacturing license, 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 which 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 notification required by Sec.
52.163 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 manufacturing license application references a standard
design certification, the environmental report need not contain a
discussion of severe accident mitigation design alternatives for the
reactor.
(c) Optional operational programs. An applicant may include in its
application descriptions of essentially complete programmatic controls,
operational programs, or operational requirements beyond those required
by Sec. 52.157 in order to satisfy requirements for license
applications that may reference a manufacturing license. If approved by
the NRC as part of the manufacturing license, such programmatic
controls, operational programs, and operational requirements would have
finality under Sec. 52.171.
0
59. In Sec. 52.171, revise paragraphs (a)(1) and (b) to read as
follows:
Sec. 52.171 Finality of manufacturing licenses; information requests.
(a)(1) Notwithstanding any provision in Sec. 50.109 of this
chapter, during the term of a manufacturing license the Commission may
not modify, rescind, or impose new requirements on the design of the
nuclear power reactor being manufactured; the requirements for the
manufacture of the nuclear power reactor; or the programmatic controls,
operational programs, or operational requirements; unless the
Commission determines that a modification is necessary to bring the
design of the reactor or its manufacture into compliance with the
Commission's requirements applicable and in effect at the time the
manufacturing license was issued, or to provide reasonable assurance of
adequate protection to public health and safety or common defense and
security.
* * * * *
(b)(1) The holder of a manufacturing license may make a change to
the facility or procedures as described in the final safety analysis
report (as updated) associated with the manufacturing license without
obtaining a license amendment pursuant to Sec. 50.90 of this chapter
if the change meets the criteria in Sec. 50.59(c)(1) of this chapter.
If needed, applications for amending an ML must be submitted and
processed in accordance with Sec. Sec. 50.90, 50.91, and 50.92 of this
chapter.
(2) An applicant who references or uses a nuclear power reactor
manufactured under a manufacturing license under this subpart may
request a departure from the design characteristics, site parameters,
terms and conditions, or approved design of the manufactured reactor.
The granting of a departure on request of an applicant is subject to
litigation in the same manner as other issues in the construction
permit or combined license hearing.
* * * * *
Sec. 52.173 [Amended]
0
60. In Sec. 52.173, remove the phrase ``15 years'' and add in its
place ``40 years''.
Sec. 52.181 [Amended]
0
61. In Sec. 52.181, remove the phrase ``15 years'' and add in its
place ``40 years''.
0
62. Revise subpart G consisting of Sec. 52.220 and add Sec. 52.220 to
read as follows:
[[Page 44698]]
Subpart G--Risk-Informed and Performance-Based Alternatives
Sec. 52.220 Use of risk-informed and performance-based alternatives
to acceptance criteria.
For each regulation in this part that provides specified acceptance
criteria, applicants may propose an alternative following the
requirements in Sec. 50.220 of this chapter.
0
63. In appendix A to part 52, revise paragraphs II.D., II.F., VI.B.4.,
VI.B.6., VIII.A., and VIII.B. to read as follows:
Appendix A to Part 52--Design Certification Rule for the U.S. Advanced
Boiling Water Reactor
* * * * *
II. * * *
D. Tier 1 means the portion of the design-related information
contained in the generic DCD that is approved and certified by this
appendix (Tier 1 information). The design descriptions, interface
requirements, and site parameters are derived from Tier 2
information. Tier 1 information includes:
1. Definitions and general provisions, which are located in the
following sections of the ABWR Design Control Document, Revision 7:
Section 1.0, ``Introduction''; Section 1.1, ``Definitions''; Section
1.2, ``General Provisions''; Appendix A, ``Legend for Figures'';
Appendix B, ``Abbreviations and Acronyms Used in the ABWR Certified
Design Material''; and Appendix C, ``Conversion to ASME Standard
Units'';
2. Design descriptions, which are located in the following
sections of the ABWR Design Control Document, Revision 7, and any
figures and non-ITAAC tables referenced in these sections: Section
2.0, ``Certified Design for ABWR Systems,'' and Section 3.0,
``Additional Certified Design Material'';
3. Inspections, tests, analyses, and acceptance criteria
(ITAAC), which are located in the inspections, tests, analyses
column and the acceptance criteria column of the following tables of
the ABWR Design Control Document, Revision 7: Tables 2.1.1d, 2.1.2,
2.1.3, 2.2.1, 2.2.2, 2.2.3, 2.2.4, 2.2.5, 2.2.6, 2.2.7, 2.2.8,
2.2.9, 2.2.10, 2.2.11, 2.3.1, 2.3.2, 2.3.3, 2.4.1, 2.4.2, 2.4.3,
2.4.4, 2.5.5, 2.5.6, 2.6.1, 2.6.2, 2.6.3, 2.7.1b, 2.7.3, 2.7.5,
2.8.4, 2.9.1, 2.10.1, 2.10.2a, 2.10.2b, 2.10.4, 2.10.7, 2.10.9,
2.10.13, 2.10.21, 2.10.22, 2.10.23, 2.11.1, 2.11.2, 2.11.3d, 2.11.4,
2.11.5, 2.11.6, 2.11.9, 2.11.10, 2.11.11, 2.11.12, 2.11.13, 2.11.20,
2.11.23, 2.12.1, 2.12.10, 2.12.11, 2.12.12, 2.12.13, 2.12.14,
2.12.15, 2.12.16, 2.12.17, 2.14.1, 2.14.4, 2.14.6, 2.14.7, 2.14.8,
2.14.9, 2.15.3, 2.15.5a, 2.15.5b, 2.15.5c, 2.15.5d, 2.15.5e,
2.15.5f, 2.15.5g, 2.15.5h, 2.15.5i, 2.15.5j, 2.15.5k, 2.15.5l,
2.15.5m, 2.15.6, 2.15.10, 2.15.11, 2.15.12, 2.15.13, 2.15.14,
2.15.15, 2.16.2, 2.17.1, 3.1, 3.2a, 3.2b, 3.3, 3.4, and 3.6.
4. Significant site parameters, which are located in the
following section of the ABWR Design Control Document, Revision 7:
Section 5.0, ``Site Parameters''; and
5. Significant interface requirements, which are located in the
following section of the ABWR Design Control Document, Revision 7:
Section 4.0, ``Interface Requirements.''
* * * * *
F. Tier 2* means the portion of the Tier 2 information,
designated as such in the generic DCD, which is subject to the
change process in paragraph VIII.B.6 of this appendix. After a plant
first achieves full power, Tier 2* information in the plant-specific
design control document for that plant reverts to Tier 2 status and
is thereafter subject to the departure provisions in paragraph
VIII.B.5 of this appendix.
* * * * *
VI. * * *
B. * * *
4. All exemptions from the DCD under and in compliance with the
change processes in paragraphs VIII.A.4, VIII.A.6, and VIII.B.4 of
this appendix, but only for that plant;
* * * * *
6. Except as provided in paragraph VIII.B.5.g of this appendix,
all departures from Tier 1 and Tier 2 under and in compliance with
the change processes in paragraphs VIII.A.5 and VIII.B.5 of this
appendix that do not require prior NRC approval, but only for that
plant; and
* * * * *
VIII. * * *
A. Tier 1 Information
1. Generic changes to Tier 1 information are governed by the
requirements in Sec. 52.63(a)(1).
2. Generic changes to Tier 1 information are applicable to all
applicants or licensees who reference this appendix, except those
for which the change has been rendered technically irrelevant by
action taken under paragraph A.3, A.4, A.5, A.6, or A.7 of this
section.
3. Departures from Tier 1 information that are required by the
Commission through plant-specific orders are governed by the
requirements in Sec. 52.63(a)(4).
4. Exemptions from Tier 1 information on definitions and general
provisions, significant site parameters, and significant interface
requirements are governed by the requirements in Sec. Sec.
52.63(b)(1) and 52.98(f). The Commission will deny a request for an
exemption from Tier 1, if it finds that the design change will
result in a significant decrease in the level of safety otherwise
provided by the design.
5. An applicant or licensee who references this appendix may
depart from Tier 1 design description information, without NRC
approval, unless the proposed departure requires an exemption under
paragraph A.6 of this section.
6. A proposed departure from Tier 1 design descriptions would
require an exemption if it would:
a. Result in more than a minimal increase in the frequency of
occurrence of an accident previously evaluated in the plant-specific
DCD;
b. Result in more than a minimal increase in the likelihood of
occurrence of a malfunction of a structure, system, or component
(SSC) important to safety and previously evaluated in the plant-
specific DCD;
c. Result in more than a minimal increase in the consequences of
an accident previously evaluated in the plant-specific DCD;
d. Result in more than a minimal increase in the consequences of
a malfunction of an SSC important to safety previously evaluated in
the plant-specific DCD;
e. Create a possibility for an accident of a different type than
any evaluated previously in the plant-specific DCD;
f. Create a possibility for a malfunction of an SSC important to
safety with a different result than any evaluated previously in the
plant-specific DCD;
g. Result in a design basis limit for a fission product barrier
as described in the plant-specific DCD being exceeded or altered;
h. Result in a departure from a method of evaluation described
in the plant-specific DCD used in establishing the design bases or
in the safety analyses;
i. Result in a substantial increase in the probability of a
severe accident such that a particular severe accident previously
reviewed and determined to be not credible could become credible; or
j. Result in a substantial increase in the consequences to the
public of a particular severe accident previously reviewed.
7. A licensee who references this appendix may not depart from
Tier 1 ITAAC without prior NRC approval. A request for a departure
will be treated as a request for a license amendment under Sec.
50.90 of this chapter and does not require an exemption from this
appendix.
8. After the plant first achieves full power, licensee-initiated
plant-specific departures from Tier 1 information are subject to the
same requirements as licensee-initiated plant-specific departures
from Tier 2 information.
B. Tier 2 and Tier 2* Information
1. Generic changes to Tier 2 or Tier 2* information are governed
by the requirements in Sec. 52.63(a)(1).
2. Generic changes to Tier 2 or Tier 2* information are
applicable to all applicants or licensees who reference this
appendix, except those for which the change has been rendered
technically irrelevant by action taken under paragraph B.3, B.4,
B.5, or B.6 of this section.
3. The Commission may not require new requirements on Tier 2 or
Tier 2* information by plant-specific order, while this appendix is
in effect under Sec. 52.55 or Sec. 52.61, unless:
a. A modification is necessary to secure compliance with the
Commission's regulations applicable and in effect at the time this
appendix was approved, as set forth in Section V of this appendix,
or to ensure adequate protection of the public health and safety or
the common defense and security; and
b. Special circumstances as defined in Sec. 50.12(a) of this
chapter are present.
4. An applicant or licensee who references this appendix may
request an exemption from Tier 2 or Tier 2* information. The
Commission may grant such a request only if it determines that the
exemption will comply with the requirements of Sec. 50.12(a) of
this chapter. The Commission will deny a request for an exemption
from Tier 2 or Tier 2*, if it finds that the design change will
result in a significant decrease in the level of
[[Page 44699]]
safety otherwise provided by the design. The granting of an
exemption to an applicant must be subject to litigation in the same
manner as other issues material to the license hearing. The granting
of an exemption to a licensee must be subject to an opportunity for
a hearing in the same manner as license amendments.
5. a. An applicant or licensee who references this appendix may
depart from Tier 2 information, without prior NRC approval, unless
the proposed departure involves a change to or departure from Tier 1
information (if prior NRC approval is required by paragraph A of
this section), Tier 2* information, or the TS, or requires a license
amendment under paragraph B.5.b or B.5.c of this section. When
evaluating the proposed departure, an applicant or licensee shall
consider all matters described in the plant-specific DCD.
b. A proposed departure from Tier 2, other than one affecting
resolution of a severe accident issue identified in the plant-
specific DCD or one affecting information required by Sec.
52.47(a)(28) to address aircraft impacts, requires a license
amendment if it would:
(1) Result in more than a minimal increase in the frequency of
occurrence of an accident previously evaluated in the plant-specific
DCD;
(2) Result in more than a minimal increase in the likelihood of
occurrence of a malfunction of a structure, system, or component
important to safety and previously evaluated in the plant-specific
DCD;
(3) Result in more than a minimal increase in the consequences
of an accident previously evaluated in the plant-specific DCD;
(4) Result in more than a minimal increase in the consequences
of a malfunction of a structure, system, or component important to
safety previously evaluated in the plant-specific DCD;
(5) Create a possibility for an accident of a different type
than any evaluated previously in the plant-specific DCD;
(6) Create a possibility for a malfunction of a structure,
system, or component important to safety with a different result
than any evaluated previously in the plant-specific DCD;
(7) Result in a design-basis limit for a fission product barrier
as described in the plant-specific DCD being exceeded or altered; or
(8) Result in a departure from a method of evaluation described
in the plant-specific DCD used in establishing the design bases or
in the safety analyses.
c. A proposed departure from Tier 2, affecting resolution of a
severe accident design feature identified in the plant-specific DCD,
requires a license amendment if:
(1) There is a substantial increase in the probability of a
severe accident such that a particular severe accident previously
reviewed and determined to be not credible could become credible; or
(2) There is a substantial increase in the consequences to the
public of a particular severe accident previously reviewed.
d. A proposed departure from Tier 2 information required by
Sec. 52.47(a)(28) to address aircraft impacts shall consider the
effect of the changed design feature or functional capability on the
original aircraft impact assessment required by Sec. 50.150(a) of
this chapter. The applicant or licensee shall describe, in the
plant-specific DCD, how the modified design features and functional
capabilities continue to meet the aircraft impact assessment
requirements in Sec. 50.150(a)(1) of this chapter.
e. If a departure requires a license amendment under paragraph
B.5.b or B.5.c of this section, it is governed by Sec. 50.90 of
this chapter.
f. A departure from Tier 2 information that is made under
paragraph B.5 of this section does not require an exemption from
this appendix.
g. A party to an adjudicatory proceeding for either the
issuance, amendment, or renewal of a license or for operation under
Sec. 52.103(a), who believes that an applicant or licensee who
references this appendix has not complied with paragraph VIII.B.5 of
this appendix when departing from Tier 2 information, may petition
to admit into the proceeding such a contention. In addition to
complying with the general requirements of Sec. 2.309 of this
chapter, the petition must demonstrate that the departure does not
comply with paragraph VIII.B.5 of this appendix. Further, the
petition must demonstrate that the change bears on an asserted
noncompliance with an ITAAC acceptance criterion in the case of a
Sec. 52.103 preoperational hearing, or that the change bears
directly on the amendment request in the case of a hearing on a
license amendment. Any other party may file a response. If, on the
basis of the petition and any response, the presiding officer
determines that a sufficient showing has been made, the presiding
officer shall certify the matter directly to the Commission for
determination of the admissibility of the contention. The Commission
may admit such a contention if it determines the petition raises a
genuine issue of material fact regarding compliance with paragraph
VIII.B.5 of this appendix.
6. a. An applicant who references this appendix may not depart
from Tier 2* information, which is designated with brackets,
italicized text, and an asterisk in the generic DCD, without NRC
approval. The departure will not be considered a resolved issue,
within the meaning of Section VI of this appendix and Sec.
52.63(a)(5).
b. A licensee who references this appendix may not depart from
the following Tier 2* matters without prior NRC approval. A request
for a departure will be treated as a request for a license amendment
under Sec. 50.90 of this chapter.
(1) Fuel burnup limit (4.2).
(2) Fuel design evaluation (4.2.3).
(3) Fuel licensing acceptance criteria (Appendix 4B).
(4) ASME Boiler & Pressure Vessel Code, Section III.
(5) ACI 349 and ANSI/AISC N-690.
(6) Motor-operated valves.
(7) Equipment seismic qualification methods.
(8) Piping design acceptance criteria.
(9) Fuel system and assembly design (4.2), except burnup limit.
(10) Nuclear design (4.3).
(11) Equilibrium cycle and control rod patterns (Appendix 4A).
(12) Control rod licensing acceptance criteria (Appendix 4C).
(13) Instrument setpoint methodology.
(14) EMS performance specifications and architecture.
(15) SSLC hardware and software qualification.
(16) Self-test system design testing features and commitments.
(17) Human factors engineering design and implementation
process.
c. After the plant first achieves full power, all Tier 2*
matters revert to Tier 2 status and are thereafter subject to the
departure provisions in paragraph B.5 of this section.
d. Departures from Tier 2* information that are made under
paragraph B.6 of this section do not require an exemption from this
appendix.
* * * * *
0
64. In appendix D to part 52, remove and reserve section IX, and revise
paragraphs II.D., II.F., VI.B.4., VI.B.6, VIII.A., and VIII.B. to read
as follows:
Appendix D to Part 52--Design Certification Rule for the AP1000 Design
* * * * *
II. * * *
D. Tier 1 means the portion of the design-related information
contained in the generic DCD that is approved and certified by this
appendix (Tier 1 information). The design descriptions, interface
requirements, and site parameters are derived from Tier 2
information. Tier 1 information includes:
1. Definitions and general provisions, which are located in the
following section of the AP1000 Design Control Document, Revision
19, and amendments thereto in Supplemental Information to Support
the AP1000 Design Certification Extension, APP-GW-GL-705 Rev. 0:
Section 1, ``Introduction'';
2. Design descriptions, which are located in the sections of the
AP1000 Design Control Document, Revision 19, identified in Table 1
to Paragraph D of this appendix, and amendments thereto in
Supplemental Information to Support the AP1000 Design Certification
Extension, APP-GW-GL-705 Rev. 0 and any figures and non-ITAAC tables
referenced in these sections:
BILLING CODE 7590-01-P
[[Page 44700]]
[GRAPHIC] [TIFF OMITTED] TP16JY26.529
BILLING CODE 7590-01-C
3. Inspections, tests, analyses, and acceptance criteria
(ITAAC), which are located in the inspections, tests, analyses
column and the acceptance criteria column of the following tables of
the AP1000 Design Control Document, Revision 19 and amendments
thereto in Supplemental Information to Support the AP1000 Design
Certification Extension, APP-GW-GL-705 Rev. 0: Tables 2.1.1-1,
2.1.2-4, 2.1.3-2, 2.2.1-3, 2.2.2-3, 2.2.3-4, 2.2.4-4, 2.2.5-5,
2.3.1-2, 2.3.2-4, 2.3.3-2, 2.3.4-2, 2.3.5-2, 2.3.6-4, 2.3.7-4,
2.3.8-2, 2.3.9-3, 2.3.10-4, 2.3.11-2, 2.3.12-1, 2.3.13-3, 2.3.14-2,
2.3.15-2, 2.3.19-2, 2.3.29-1, 3.1-1, 3.2-1, 3.3-6, 3.5-6, 3.6-1, and
3.7-3.
4. Significant site parameters, which are located in the
following section of the AP1000 Design Control Document, Revision
19, and amendments thereto in Supplemental Information to Support
the AP1000 Design Certification Extension, APP-GW-GL-705 Rev. 0:
Section 5, ``Site Parameters''; and
5. Significant interface requirements, which are located in the
following section of the AP1000 Design Control Document, Revision
19, and amendments thereto in Supplemental Information to Support
the AP1000 Design Certification Extension, APP-GW-GL-705 Rev. 0:
Section 4, ``Interface Requirements.''
* * * * *
F. Tier 2* means the portion of the Tier 2 information,
designated as such in the generic DCD, which is subject to the
change process in Section VIII.B.5 of this appendix.
* * * * *
VI. * * *
B. * * *
4. All exemptions from the DCD under and in compliance with the
change processes in paragraphs VIII.A.4, VIII.A.6, and VIII.B.4 of
this appendix, but only for that plant;
* * * * *
6. Except as provided in paragraph VIII.B.5.g of this appendix,
all departures from Tier 1, Tier 2, and Tier 2* under and in
compliance with the change processes in paragraphs VIII.A.5 and
VIII.B.5 of this appendix that do not require prior NRC approval,
but only for that plant; and
* * * * *
VIII. * * *
A. Tier 1 information.
1. Generic changes to Tier 1 information are governed by the
requirements in Sec. 52.63(a)(1).
2. Generic changes to Tier 1 information are applicable to all
applicants or licensees who reference this appendix, except those
for which the change has been rendered technically irrelevant by
action taken under
[[Page 44701]]
paragraphs A.3, A.4, A.5, A.6, or A.7 of this section.
3. Departures from Tier 1 information that are required by the
Commission through plant-specific orders are governed by the
requirements in Sec. 52.63(a)(4).
4. Exemptions from Tier 1 information on definitions and general
provisions, significant site parameters, and significant interface
requirements are governed by the requirements in Sec. Sec.
52.63(b)(1) and 52.98(f). The Commission will deny a request for an
exemption from Tier 1, if it finds that the design change will
result in a significant decrease in the level of safety otherwise
provided by the design.
5. An applicant or licensee who references this appendix may
depart from Tier 1 design description information, without NRC
approval, unless the proposed departure requires an exemption under
paragraph A.6 of this section.
6. A proposed departure from Tier 1 design descriptions would
require an exemption if it would:
a. Result in more than a minimal increase in the frequency of
occurrence of an accident previously evaluated in the plant-specific
DCD;
b. Result in more than a minimal increase in the likelihood of
occurrence of a malfunction of a structure, system, or component
(SSC) important to safety and previously evaluated in the plant-
specific DCD;
c. Result in more than a minimal increase in the consequences of
an accident previously evaluated in the plant-specific DCD;
d. Result in more than a minimal increase in the consequences of
a malfunction of an SSC important to safety previously evaluated in
the plant-specific DCD;
e. Create a possibility for an accident of a different type than
any evaluated previously in the plant-specific DCD;
f. Create a possibility for a malfunction of an SSC important to
safety with a different result than any evaluated previously in the
plant-specific DCD;
g. Result in a design basis limit for a fission product barrier
as described in the plant-specific DCD being exceeded or altered;
h. Result in a departure from a method of evaluation described
in the plant-specific DCD used in establishing the design bases or
in the safety analyses;
i. Result in a substantial increase in the probability of a
severe accident such that a particular severe accident previously
reviewed and determined to be not credible could become credible; or
j. Result in a substantial increase in the consequences to the
public of a particular severe accident previously reviewed.
7. A licensee who references this appendix may not depart from
Tier 1 ITAAC without prior NRC approval. A request for a departure
will be treated as a request for a license amendment under Sec.
50.90 of this chapter and does not require an exemption from this
appendix.
8. After the plant first achieves full power, licensee-initiated
plant-specific departures from Tier 1 information are subject to the
same requirements as licensee-initiated plant-specific departures
from Tier 2 information.
B. Tier 2 and Tier 2* information.
1. Generic changes to Tier 2 or Tier 2* information are governed
by the requirements in Sec. 52.63(a)(1).
2. Generic changes to Tier 2 or Tier 2* information are
applicable to all applicants or licensees who reference this
appendix, except those for which the change has been rendered
technically irrelevant by action taken under paragraphs B.3, B.4, or
B.5 of this section.
3. The Commission may not require new requirements on Tier 2 or
Tier 2* information by plant-specific order while this appendix is
in effect under Sec. 52.55 or Sec. 52.61, unless:
a. A modification is necessary to secure compliance with the
Commission's regulations applicable and in effect at the time this
appendix was approved, as set forth in Section V of this appendix,
or to ensure adequate protection of the public health and safety or
the common defense and security; and
b. Special circumstances as defined in Sec. 50.12(a) of this
chapter are present.
4. An applicant or licensee who references this appendix may
request an exemption from Tier 2 or Tier 2* information. The
Commission may grant such a request only if it determines that the
exemption will comply with the requirements of Sec. 50.12(a) of
this chapter. The Commission will deny a request for an exemption
from Tier 2 or Tier 2*, if it finds that the design change will
result in a significant decrease in the level of safety otherwise
provided by the design. The grant of an exemption to an applicant
must be subject to litigation in the same manner as other issues
material to the license hearing. The grant of an exemption to a
licensee must be subject to an opportunity for a hearing in the same
manner as license amendments.
5. a. An applicant or licensee who references this appendix may
depart from Tier 2 or Tier 2* information, without prior NRC
approval, unless the proposed departure involves a change to or
departure from Tier 1 information (if prior NRC approval is required
by paragraph A of this section) or the TS, or requires a license
amendment under paragraphs B.5.b or B.5.c of this section. When
evaluating the proposed departure, an applicant or licensee shall
consider all matters described in the plant-specific DCD.
b. A proposed departure from Tier 2 or Tier 2*, other than one
affecting resolution of a severe accident issue identified in the
plant-specific DCD or one affecting information required by Sec.
52.47(a)(28) to address Sec. 50.150 of this chapter, requires a
license amendment if it would:
(1) Result in more than a minimal increase in the frequency of
occurrence of an accident previously evaluated in the plant-specific
DCD;
(2) Result in more than a minimal increase in the likelihood of
occurrence of a malfunction of a structure, system, or component
(SSC) important to safety and previously evaluated in the plant-
specific DCD;
(3) Result in more than a minimal increase in the consequences
of an accident previously evaluated in the plant-specific DCD;
(4) Result in more than a minimal increase in the consequences
of a malfunction of an SSC important to safety previously evaluated
in the plant-specific DCD;
(5) Create a possibility for an accident of a different type
than any evaluated previously in the plant-specific DCD;
(6) Create a possibility for a malfunction of an SSC important
to safety with a different result than any evaluated previously in
the plant-specific DCD;
(7) Result in a design basis limit for a fission product barrier
as described in the plant-specific DCD being exceeded or altered; or
(8) Result in a departure from a method of evaluation described
in the plant-specific DCD used in establishing the design bases or
in the safety analyses.
c. A proposed departure from Tier 2 or Tier 2* affecting
resolution of a severe accident design feature identified in the
plant-specific DCD, requires a license amendment if:
(1) There is a substantial increase in the probability of a
severe accident such that a particular severe accident previously
reviewed and determined to be not credible could become credible; or
(2) There is a substantial increase in the consequences to the
public of a particular severe accident previously reviewed.
d. If an applicant or licensee proposes to depart from the
information required by Sec. 52.47(a)(28) to be included in the
FSAR for the standard design certification, then the applicant or
licensee shall consider the effect of the changed feature or
capability on the original assessment required by Sec. 50.150(a) of
this chapter. The applicant or licensee must also document how the
modified design features and functional capabilities continue to
meet the assessment requirements in Sec. 50.150(a)(1) of this
chapter in accordance with Section X of this appendix.
e. If a departure requires a license amendment under paragraph
B.5.b or B.5.c of this section, it is governed by Sec. 50.90 of
this chapter.
f. A departure from Tier 2 or Tier 2* information that is made
under paragraph B.5 of this section does not require an exemption
from this appendix.
g. A party to an adjudicatory proceeding for either the
issuance, amendment, or renewal of a license or for operation under
Sec. 52.103(a), who believes that an applicant or licensee who
references this appendix has not complied with paragraph VIII.B.5 of
this appendix when departing from Tier 2 information, may petition
to admit into the proceeding such a contention. In addition to
compliance with the general requirements of Sec. 2.309 of this
chapter, the petition must demonstrate that the departure does not
comply with paragraph VIII.B.5 of this appendix. Further, the
petition must demonstrate that the change bears on an asserted
noncompliance with an ITAAC acceptance criterion in the case of a
Sec. 52.103 preoperational hearing, or that the change bears
directly on the amendment request in the case of a hearing on a
license amendment. Any other party may file a response. If, on the
basis of the petition and
[[Page 44702]]
any response, the presiding officer determines that a sufficient
showing has been made, the presiding officer shall certify the
matter directly to the Commission for determination of the
admissibility of the contention. The Commission may admit such a
contention if it determines the petition raises a genuine issue of
material fact regarding compliance with paragraph VIII.B.5 of this
appendix.
* * * * *
0
65. In appendix E to part 52, revise paragraphs II.D., II.F, VI.B.4.,
VI.B.6., VIII.A., and VIII.B. to read as follows:
Appendix E to Part 52--Design Certification Rule for the ESBWR Design
* * * * *
II. * * *
D. Tier 1 means the portion of the design-related information
contained in the generic DCD that is approved and certified by this
appendix (Tier 1 information). The design descriptions, interface
requirements, and site parameters are derived from Tier 2
information. Tier 1 information includes:
1. Definitions and general provisions, which are located in the
following sections of the ESBWR Design Control Document, Revision
10: Section 1, ``Introduction,'' Section 1.1, ``Definitions and
General Provisions,'' Section 1.2, ``Figure Legend,'' Section 1.3,
``Table Legend,'' and Section 1.4, ``Design Acceptance Criteria'';
2. Design descriptions, which are located in the following
sections of the ESBWR Design Control Document, Revision 10, and any
figures and non-ITAAC tables referenced in these sections: Section
2, ``Design Descriptions and ITAAC,'' and Section 3, ``Non-System
Based Material'';
3. Inspections, tests, analyses, and acceptance criteria
(ITAAC), which are located in the inspections, tests, analyses
column and the acceptance criteria column of the following tables of
the ESBWR Design Control Document, Revision 10: Tables 2.1.1-3,
2.1.2-3, 2.2.1-6, 2.2.2-7, 2.2.3-4, 2.2.4-6, 2.2.5-4, 2.2.6-3,
2.2.7-4, 2.2.9-3, 2.2.12-5, 2.2.13-4, 2.2.14-4, 2.2.15-2, 2.2.16-4,
2.3.1-2, 2.3.2-2, 2.4.1-3, 2.4.2-3, 2.5.5-1, 2.5.6-1, 2.5.10-1,
2.6.1-2, 2.6.2-2, 2.10.1-2, 2.10.2-2, 2.10.3-1, 2.11.1-1, 2.11.2-1,
2.11.4-2, 2.11.5-1, 2.11.6-1, 2.11.7-1, 2.12.3-1, 2.12.5-1, 2.12.7-
1, 2.13.1-2, 2.13.3-2, 2.13.4-2, 2.13.5-2, 2.13.8-1, 2.13.9-1,
2.15.1-2, 2.15.3-2, 2.15.4-2, 2.15.5-2, 2.15.7-2, 2.15.8-1, 2.16.1-
1, 2.16.2-2, 2.16.2-4, 2.16.2-6, 2.16.2-7, 2.16.2-9, 2.16.2-10,
2.16.3-2, 2.16.3.1-1, 2.16.4-1, 2.16.5-2, 2.16.6-2, 2.16.7-2,
2.16.8-1, 2.16.9-1, 2.16.10-1, 2.16.11-1, 2.16.12-1, 2.16.13-1,
2.16.14-1, 2.19-1, 3.1-1, 3.3-2, 3.4-1, 3.6-1, 3.7-1, and 3.8-2.
4. Significant site parameters, which are located in the
following section of the ESBWR Design Control Document, Revision 10:
Section 5, ``Site Parameters''; and
5. Significant interface requirements, which are located in the
following section of the ESBWR Design Control Document, Revision 10:
Section 4, ``Interface Material.''
* * * * *
F. Tier 2* means the portion of the Tier 2 information,
designated as such in the generic DCD, which is subject to the
change process in paragraph VIII.B.6 of this appendix. After a plant
first achieves full power, Tier 2* information in the plant-specific
design control document for that plant reverts to Tier 2 status and
is thereafter subject to the departure provisions in paragraph
VIII.B.5 of this appendix.
* * * * *
VI. * * *
B. * * *
4. All exemptions from the DCD under and in compliance with the
change processes in paragraphs VIII.A.4, VIII.A.6, and VIII.B.4 of
this appendix, but only for that plant;
* * * * *
6. Except as provided in paragraph VIII.B.5.g of this appendix,
all departures from Tier 1 and Tier 2 under and in compliance with
the change processes in paragraphs VIII.A.5 and VIII.B.5 of this
appendix that do not require prior NRC approval, but only for that
plant; and
* * * * *
VIII. * * *
A. Tier 1 information
1. Generic changes to Tier 1 information are governed by the
requirements in Sec. 52.63(a)(1).
2. Generic changes to Tier 1 information are applicable to all
applicants or licensees who reference this appendix, except those
for which the change has been rendered technically irrelevant by
action taken under paragraphs A.3, A.4, A.5, A.6, or A.7 of this
section.
3. Departures from Tier 1 information that are required by the
Commission through plant-specific orders are governed by the
requirements in Sec. 52.63(a)(4).
4. Exemptions from Tier 1 information on definitions and general
provisions, significant site parameters, and significant interface
requirements are governed by the requirements in Sec. Sec.
52.63(b)(1) and 52.98(f). The Commission will deny a request for an
exemption from Tier 1, if it finds that the design change will
result in a significant decrease in the level of safety otherwise
provided by the design.
5. An applicant or licensee who references this appendix may
depart from Tier 1 design description information, without NRC
approval, unless the proposed departure requires an exemption under
paragraph A.6 of this section.
6. A proposed departure from Tier 1 design descriptions would
require an exemption if it would:
a. Result in more than a minimal increase in the frequency of
occurrence of an accident previously evaluated in the plant-specific
DCD;
b. Result in more than a minimal increase in the likelihood of
occurrence of a malfunction of a structure, system, or component
(SSC) important to safety and previously evaluated in the plant-
specific DCD;
c. Result in more than a minimal increase in the consequences of
an accident previously evaluated in the plant-specific DCD;
d. Result in more than a minimal increase in the consequences of
a malfunction of an SSC important to safety previously evaluated in
the plant-specific DCD;
e. Create a possibility for an accident of a different type than
any evaluated previously in the plant-specific DCD;
f. Create a possibility for a malfunction of an SSC important to
safety with a different result than any evaluated previously in the
plant-specific DCD;
g. Result in a design basis limit for a fission product barrier
as described in the plant-specific DCD being exceeded or altered;
h. Result in a departure from a method of evaluation described
in the plant-specific DCD used in establishing the design bases or
in the safety analyses;
i. Result in a substantial increase in the probability of a
severe accident such that a particular severe accident previously
reviewed and determined to be not credible could become credible; or
j. Result in a substantial increase in the consequences to the
public of a particular severe accident previously reviewed.
7. A licensee who references this appendix may not depart from
Tier 1 ITAAC without prior NRC approval. A request for a departure
will be treated as a request for a license amendment under Sec.
50.90 of this chapter and does not require an exemption from this
appendix.
8. After the plant first achieves full power, licensee-initiated
plant-specific departures from Tier 1 information are subject to the
same requirements as licensee-initiated plant-specific departures
from Tier 2 information.
B. Tier 2 and Tier 2* information
1. Generic changes to Tier 2 or Tier 2* information are governed
by the requirements in Sec. 52.63(a)(1).
2. Generic changes to Tier 2 or Tier 2* information are
applicable to all applicants or licensees who reference this
appendix, except those for which the change has been rendered
technically irrelevant by action taken under paragraphs B.3, B.4,
B.5, or B.6 of this section.
3. The Commission may not require new requirements on Tier 2 or
Tier 2* information by plant-specific order while this appendix is
in effect under Sec. 52.55 or Sec. 52.61, unless:
a. A modification is necessary to secure compliance with the
Commission's regulations applicable and in effect at the time this
appendix was approved, as set forth in Section V of this appendix,
or to ensure adequate protection of the public health and safety or
the common defense and security; and
b. Special circumstances as defined in Sec. 50.12(a) of this
chapter are present.
4. An applicant or licensee who references this appendix may
request an exemption from Tier 2 or Tier 2* information. The
Commission may grant such a request only if it determines that the
exemption will comply with the requirements of Sec. 50.12(a) of
this chapter. The Commission will deny a request for an exemption
from Tier 2 or Tier 2*, if it finds that the design change will
result in a significant decrease in the level of safety otherwise
provided by the design. The grant of an exemption to an applicant
must be subject to litigation in the same manner as other issues
material to the license
[[Page 44703]]
hearing. The grant of an exemption to a licensee must be subject to
an opportunity for a hearing in the same manner as license
amendments.
5. a. An applicant or licensee who references this appendix may
depart from Tier 2 information, without prior NRC approval, unless
the proposed departure involves a change to or departure from Tier 1
information (if prior NRC approval is required by paragraph A of
this section), Tier 2* information, or the TS, or requires a license
amendment under paragraph B.5.b or B.5.c of this section. When
evaluating the proposed departure, an applicant or licensee shall
consider all matters described in the plant-specific DCD.
b. A proposed departure from Tier 2, other than one affecting
resolution of a severe accident issue identified in the plant-
specific DCD or one affecting information required by Sec.
52.47(a)(28) to address aircraft impacts, requires a license
amendment if it would:
(1) Result in more than a minimal increase in the frequency of
occurrence of an accident previously evaluated in the plant-specific
DCD;
(2) Result in more than a minimal increase in the likelihood of
occurrence of a malfunction of a structure, system, or component
(SSC) important to safety and previously evaluated in the plant-
specific DCD;
(3) Result in more than a minimal increase in the consequences
of an accident previously evaluated in the plant-specific DCD;
(4) Result in more than a minimal increase in the consequences
of a malfunction of an SSC important to safety previously evaluated
in the plant-specific DCD;
(5) Create a possibility for an accident of a different type
than any evaluated previously in the plant-specific DCD;
(6) Create a possibility for a malfunction of an SSC important
to safety with a different result than any evaluated previously in
the plant-specific DCD;
(7) Result in a design-basis limit for a fission product barrier
as described in the plant-specific DCD being exceeded or altered; or
(8) Result in a departure from a method of evaluation described
in the plant-specific DCD used in establishing the design bases or
in the safety analyses.
c. A proposed departure from Tier 2 affecting resolution of a
severe accident design feature identified in the plant-specific DCD,
requires a license amendment if:
(1) There is a substantial increase in the probability of a
severe accident such that a particular severe accident previously
reviewed and determined to be not credible could become credible; or
(2) There is a substantial increase in the consequences to the
public of a particular severe accident previously reviewed.
d. A proposed departure from Tier 2 information required by
Sec. 52.47(a)(28) to address aircraft impacts shall consider the
effect of the changed design feature or functional capability on the
original aircraft impact assessment required by Sec. 50.150(a) of
this chapter. The applicant or licensee shall describe in the plant-
specific DCD how the modified design features and functional
capabilities continue to meet the aircraft impact assessment
requirements in Sec. 50.150(a)(1) of this chapter.
e. If a departure requires a license amendment under paragraph
B.5.b or B.5.c of this section, it is governed by Sec. 50.90 of
this chapter.
f. A departure from Tier 2 information that is made under
paragraph B.5 of this section does not require an exemption from
this appendix.
g. A party to an adjudicatory proceeding for either the
issuance, amendment, or renewal of a license or for operation under
Sec. 52.103(a), who believes that an applicant or licensee who
references this appendix has not complied with paragraph VIII.B.5 of
this appendix when departing from Tier 2 information, may petition
to admit into the proceeding such a contention. In addition to
compliance with the general requirements of Sec. 2.309 of this
chapter, the petition must demonstrate that the departure does not
comply with paragraph VIII.B.5 of this appendix. Further, the
petition must demonstrate that the change bears on an asserted
noncompliance with an ITAAC acceptance criterion in the case of a
Sec. 52.103 preoperational hearing, or that the change bears
directly on the amendment request in the case of a hearing on a
license amendment. Any other party may file a response. If, on the
basis of the petition and any response, the presiding officer
determines that a sufficient showing has been made, the presiding
officer shall certify the matter directly to the Commission for
determination of the admissibility of the contention. The Commission
may admit such a contention if it determines the petition raises a
genuine issue of material fact regarding compliance with paragraph
VIII.B.5 of this appendix.
6. a. An applicant who references this appendix may not depart
from Tier 2* information, which is designated with italicized text
or brackets and an asterisk in the generic DCD, without NRC
approval. The departure will not be considered a resolved issue,
within the meaning of Section VI of this appendix and Sec.
52.63(a)(5).
b. A licensee who references this appendix may not depart from
the following Tier 2* matters without prior NRC approval. A request
for a departure will be treated as a request for a license amendment
under Sec. 50.90 of this chapter.
(1) Fuel mechanical and thermal-mechanical design evaluation
reports, including fuel burnup limits.
(2) Control rod mechanical and nuclear design reports.
(3) Fuel nuclear design report.
(4) Critical power correlation.
(5) Fuel licensing acceptance criteria.
(6) Control rod licensing acceptance criteria.
(7) Mechanical and structural design of spent fuel storage
racks.
(8) Steam dryer pressure load analysis methodology.
(9) ASME Boiler and Pressure Vessel Code, Section III,
Subsections NE (Division 1) and CC (Division 2) for containment
vessel design.
(10) American Concrete Institute 349 and American National
Standards Institute/American Institute of Steel Construction--N690.
(11) Power-operated valves.
(12) Equipment seismic qualification methods.
(13) Piping design acceptance criteria.
(14) Instrument setpoint methodology.
(15) Safety-Related Distribution Control and Information System
performance
specification and architecture.
(16) Safety System Logic and Control hardware and software.
(17) Human factors engineering design and implementation.
(18) First of a kind testing for reactor stability (first plant
only).
(19) Reactor precritical heatup with reactor water cleanup/
shutdown cooling (first plant only).
(20) Isolation condenser system heatup and steady state
operation (first plant only).
(21) Power maneuvering in the feedwater temperature operating
domain (first plant only).
(22) Load maneuvering capability (first plant only).
(23) Defense-in-depth stability solution evaluation test (first
plant only).
c. After the plant first achieves full power, all Tier 2*
matters revert to Tier 2 status and are thereafter subject to the
departure provisions in paragraph B.5 of this section.
d. Departures from Tier 2* information that are made under
paragraph B.6 of this section do not require an exemption from this
appendix.
* * * * *
0
66. In appendix F to part 52, revise paragraphs II.D., VI.B.4.,
VI.B.6., VIII.A., VIII.B.5.a., and VIII.B.5.c.
Appendix F to Part 52--Design Certification Rule for the APR1400 Design
* * * * *
II. * * *
D. Tier 1 means the portion of the design-related information
contained in the generic DCD that is approved and certified by this
appendix (Tier 1 information). The design descriptions, interface
requirements, and site parameters are derived from Tier 2
information. Tier 1 information includes:
1. Definitions and general provisions, which are located in the
following section of the APR1400 Design Control Document Tier 1,
Revision 3: Section 1.0, ``Introduction'';
2. Design descriptions, which are located in the sections of the
APR1400 Design Control Document Tier 1, Revision 3, identified in
Table 1 to Paragraph D of this appendix, and any figures and non-
ITAAC tables referenced in these sections:
[[Page 44704]]
[GRAPHIC] [TIFF OMITTED] TP16JY26.530
3. Inspections, tests, analyses, and acceptance criteria
(ITAAC), which are located in the inspections, tests, analyses
column, and the acceptance criteria column of the following tables
of the APR1400 Design Control Document Tier 1, Revision 3: Tables
2.2.1-3, 2.2.2-2, 2.2.3-1, 2.2.4-1, 2.2.5-1, 2.2.6-2, 2.2.7-2,
2.2.8-1, 2.2.9-1, 2.3-3, 2.4.1-4, 2.4.2-4, 2.4.3.4, 2.4.4-4, 2.4.5-
4, 2.4.6-4, 2.4.7-1, 2.5.1-5, 2.5.2-5, 2.5.3-3, 2.5.4-5, 2.5.5-2,
2.6.1-3, 2.6.2-3, 2.6.3-3, 2.6.4-3, 2.6.5-1, 2.6.6-1, 2.6.7-1,
2.6.8-1, 2.6.9-1, 2.7.1.1-1, 2.7.1.2-4, 2.7.1.4-4, 2.7.1.5-4,
2.7.1.8-3, 2.7.2.1-4, 2.7.2.2-4, 2.7.2.3-4, 2.7.2.4-1, 2.7.2.5-4,
2.7.2.6-4, 2.7.3.1-3, 2.7.3.2-3, 2.7.3.3-3, 2.7.3.5-3, 2.7.3.6-1,
2.7.4.1-1, 2.7.4.2-1, 2.7.4.3-4, 2.7.4.4-2, 2.7.4.5-1, 2.7.5.2-3,
2.7.6.1-2, 2.7.6.2-4, 2.7.6.3-2, 2.7.6.4-3, 2.7.6.5-3, 2.8-2, 2.9-1,
2.10-1, 2.11.1-2, 2.11.2-4, 2.11.3-2, 2.11.4-3, 2.12-1, and 2.13-1.
4. Significant site parameters, which are located in the
following section of the APR1400 Design Control Document Tier 1,
Revision 3: Section 2.1, ``Site Parameters''; and
5. Significant interface requirements, which are located in the
following section of the APR1400 Design Control Document Tier 1,
Revision 3: Section 3.0, ``Interface Requirement.''
* * * * *
VI. * * *
B. * * *
4. All exemptions from the DCD under and in compliance with the
change processes in paragraphs VIII.A.4, VIII.A.6, and VIII.B.4 of
this appendix, but only for that plant;
* * * * *
6. Except as provided in paragraph VIII.B.5.g of this appendix,
all departures from Tier 1 and Tier 2 under and in compliance with
the change processes in paragraphs VIII.A.5 and VIII.B.5 of this
appendix that do not require prior NRC approval, but only for that
plant; and
* * * * *
VIII. * * *
A. Tier 1 Information
1. Generic changes to Tier 1 information are governed by the
requirements in Sec. 52.63(a)(1).
2. Generic changes to Tier 1 information are applicable to all
applicants or licensees who reference this appendix, except those
for which the change has been rendered technically irrelevant by
action taken under paragraphs A.3, A.4, A.5, A.6, or A.7 of this
section.
3. Departures from Tier 1 information that are required by the
Commission through plant-specific orders are governed by the
requirements in Sec. 52.63(a)(4).
4. Exemptions from Tier 1 information on definitions and general
provisions, significant site parameters, and significant interface
requirements are governed by the requirements in Sec. Sec.
52.63(b)(1) and 52.98(f). The Commission will deny a request for an
exemption from Tier 1, if it finds that the design change will
result in a significant decrease in the level of safety otherwise
provided by the design.
5. An applicant or licensee who references this appendix may
depart from Tier 1 design description information, without NRC
approval, unless the proposed departure requires an exemption under
paragraph A.6 of this section.
6. A proposed departure from Tier 1 design descriptions would
require an exemption if it would:
a. Result in more than a minimal increase in the frequency of
occurrence of an accident previously evaluated in the plant-specific
DCD;
b. Result in more than a minimal increase in the likelihood of
occurrence of a malfunction of a structure, system, or component
(SSC) important to safety and previously evaluated in the plant-
specific DCD;
c. Result in more than a minimal increase in the consequences of
an accident previously evaluated in the plant-specific DCD;
d. Result in more than a minimal increase in the consequences of
a malfunction of an SSC important to safety previously evaluated in
the plant-specific DCD;
e. Create a possibility for an accident of a different type than
any evaluated previously in the plant-specific DCD;
f. Create a possibility for a malfunction of an SSC important to
safety with a different result than any evaluated previously in the
plant-specific DCD;
g. Result in a design basis limit for a fission product barrier
as described in the plant-specific DCD being exceeded or altered;
h. Result in a departure from a method of evaluation described
in the plant-specific DCD used in establishing the design bases or
in the safety analyses.
i. Result in a substantial increase in the probability of a
severe accident such that a particular severe accident previously
reviewed and determined to be not credible could become credible; or
j. Result in a substantial increase in the consequences to the
public of a particular severe accident previously reviewed.
7. A licensee who references this appendix may not depart from
Tier 1 ITAAC without prior NRC approval. A request for a departure
will be treated as a request for a license amendment under Sec.
50.90 of this chapter and does not require an exemption from this
appendix.
8. After the plant first achieves full power, licensee-initiated
plant-specific departures from Tier 1 information are subject to the
same requirements as licensee-initiated plant-specific departures
from Tier 2 information.
B. * * *
5. a. An applicant or licensee who references this appendix may
depart from Tier 2 information, without prior NRC approval, unless
the proposed departure involves a change to or departure from Tier 1
information (if prior NRC approval is required by paragraph A of
this section) or the TS, or requires a license amendment under
paragraph B.5.b or B.5.c of this section. When evaluating the
proposed departure, an applicant or licensee shall consider all
matters described in the plant-specific DCD.
* * * * *
c. A proposed departure from Tier 2, affecting resolution of a
severe accident design feature identified in the plant-specific DCD,
requires a license amendment if:
(1) There is a substantial increase in the probability of a
severe accident such that a particular severe accident previously
[[Page 44705]]
reviewed and determined to be not credible could become credible; or
(2) There is a substantial increase in the consequences to the
public of a particular severe accident previously reviewed.
* * * * *
0
67. In appendix G to part 52, revise paragraphs II.D., V.B.1., V.B.9.,
VI.B.4, V.B.6., VIII.A., VIII.B.5.a., and VIII.B.5.c. to read as
follows:
Appendix G to Part 52--Design Certification Rule for NuScale
* * * * *
II. * * *
D. Tier 1 means the portion of the design-related information
contained in the generic DCD that is approved and certified by this
appendix (Tier 1 information). The design descriptions, interface
requirements, and site parameters are derived from Tier 2
information. Tier 1 information includes:
1. Definitions and general provisions, which are located in the
following sections of the NuScale Standard Plant Design
Certification Application, Certified Design Descriptions and
Inspections, Tests, Analyses, & Acceptance Criteria (ITAAC), Part
2--Tier 1, Revision 5: Section 1.0, ``Introduction,'' Section 1.1,
``Definitions,'' and Section 1.2, ``General Provisions'';
2. Design descriptions, which are located in the following
chapters of the NuScale Standard Plant Design Certification
Application, Certified Design Descriptions and Inspections, Tests,
Analyses, & Acceptance Criteria (ITAAC), Part 2--Tier 1, Revision 5
and any figures and non-ITAAC tables referenced in these sections:
Chapter 2, ``Unit Specific Structures, Systems, and Components
Design Descriptions and Inspections, Tests, Analyses, and Acceptance
Criteria,'' and Chapter 3, ``Shared Structures, Systems, and
Components and Non-Structures, Systems, and Components Design
Descriptions and Inspections, Tests, Analyses, and Acceptance
Criteria'';
3. Inspections, tests, analyses, and acceptance criteria
(ITAAC), which are located in the inspections, tests, analyses
column and the acceptance criteria column of the following tables of
the NuScale Standard Plant Design Certification Application,
Certified Design Descriptions and Inspections, Tests, Analyses, &
Acceptance Criteria (ITAAC), Part 2--Tier 1, Revision 5: Tables 2.1-
4, 2.2-3, 2.3-1, 2.5-7, 2.6-1, 2.7-2, 2.8-2, 3.0-1, 3.1-2, 3.2-2,
3.3-1, 3.4-1, 3.5-1, 3.6-2, 3.7-1, 3.8-1, 3.9-2, 3.10-2, 3.11-2,
3.12-2, 3.13-1, 3.14-2, 3.15-1, 3.16-1, 3.17-2, and 3.18-2.
4. Significant site parameters, which are located in the
following chapter of the NuScale Standard Plant Design Certification
Application, Certified Design Descriptions and Inspections, Tests,
Analyses, & Acceptance Criteria (ITAAC), Part 2--Tier 1, Revision 5:
Chapter 5, ``Site Parameters''; and
5. Significant interface requirements which are located in the
following chapter of the NuScale Standard Plant Design Certification
Application, Certified Design Descriptions and Inspections, Tests,
Analyses, & Acceptance Criteria (ITAAC), Part 2--Tier 1, Revision 5:
Chapter 4, ``Interface Requirements.''
* * * * *
V. * * *
B. * * *
1. Paragraph (f)(2)(vi) of 10 CFR 50.34 and 10 CFR 50.46b--High
point venting for the reactor coolant system and reactor pressure
vessel head.
* * * * *
9. Appendix A of 10 CFR part 50--Electric Power Systems GDCs: *
* *
* * * * *
VI. * * *
B. * * *
4. All exemptions from the DCD under and in compliance with the
change processes in paragraphs VIII.A.4, VIII.A.6, and VIII.B.4 of
this appendix, but only for that plant;
* * * * *
6. Except as provided in paragraph VIII.B.5.g of this appendix,
all departures from Tier 1 and Tier 2 under and in compliance with
the change processes in paragraphs VIII.A.5 and VIII.B.5 of this
appendix that do not require prior NRC approval, but only for that
plant; and
* * * * *
VIII. * * *
A. Tier 1 Information
1. Generic changes to Tier 1 information are governed by the
requirements in Sec. 52.63(a)(1).
2. Generic changes to Tier 1 information are applicable to all
applicants or licensees who reference this appendix, except those
for which the change has been rendered technically irrelevant by
action taken under paragraphs A.3, A.4, A.5, A.6, or A.7 of this
section.
3. Departures from Tier 1 information that are required by the
Commission through plant-specific orders are governed by the
requirements in Sec. 52.63(a)(4).
4. Exemptions from Tier 1 information on definitions and general
provisions, significant site parameters, and significant interface
requirements are governed by the requirements in Sec. Sec.
52.63(b)(1) and 52.98(f). The Commission will deny a request for an
exemption from Tier 1, if it finds that the design change will
result in a significant decrease in the level of safety otherwise
provided by the design.
5. An applicant or licensee who references this appendix may
depart from Tier 1 design description information, without NRC
approval, unless the proposed departure requires an exemption under
paragraph A.6 of this section.
6. A proposed departure from Tier 1 design descriptions would
require an exemption if it would:
a. Result in more than a minimal increase in the frequency of
occurrence of an accident previously evaluated in the plant-specific
DCD;
b. Result in more than a minimal increase in the likelihood of
occurrence of a malfunction of a structure, system, or component
(SSC) important to safety and previously evaluated in the plant-
specific DCD;
c. Result in more than a minimal increase in the consequences of
an accident previously evaluated in the plant-specific DCD;
d. Result in more than a minimal increase in the consequences of
a malfunction of an SSC important to safety previously evaluated in
the plant-specific DCD;
e. Create a possibility for an accident of a different type than
any evaluated previously in the plant-specific DCD;
f. Create a possibility for a malfunction of an SSC important to
safety with a different result than any evaluated previously in the
plant-specific DCD;
g. Result in a design basis limit for a fission product barrier
as described in the plant-specific DCD being exceeded or altered;
h. Result in a departure from a method of evaluation described
in the plant-specific DCD used in establishing the design bases or
in the safety analyses.
i. Result in a substantial increase in the probability of a
severe accident such that a particular severe accident previously
reviewed and determined to be not credible could become credible; or
j. Result in a substantial increase in the consequences to the
public of a particular severe accident previously reviewed.
7. A licensee who references this appendix may not depart from
Tier 1 ITAAC without prior NRC approval. A request for a departure
will be treated as a request for a license amendment under Sec.
50.90 of this chapter and does not require an exemption from this
appendix.
8. After the plant first achieves full power, licensee-initiated
plant-specific departures from Tier 1 information are subject to the
same requirements as licensee-initiated plant-specific departures
from Tier 2 information.
* * * * *
B. * * *
5.
a. An applicant or licensee who references this appendix may
depart from Tier 2 information, without prior NRC approval, unless
the proposed departure involves a change to or departure from Tier 1
information (if prior NRC approval is required by paragraph A of
this section) or the TS, or requires a license amendment under
paragraph B.5.b or B.5.c of this section. When evaluating the
proposed departure, an applicant or licensee shall consider all
matters described in the plant-specific DCD.
* * * * *
c. A proposed departure from Tier 2, affecting resolution of a
severe accident design feature identified in the plant-specific DCD,
requires a license amendment if:
(1) There is a substantial increase in the probability of a
severe accident such that a particular severe accident previously
reviewed and determined to be not credible could become credible; or
(2) There is a substantial increase in the consequences to the
public of a particular severe accident previously reviewed.
* * * * *
[[Page 44706]]
PART 53--RISK-INFORMED, TECHNOLOGY-INCLUSIVE REGULATORY FRAMEWORK
FOR COMMERCIAL NUCLEAR PLANTS
0
68. The authority citation for part 53 continues to read as follows:
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.
0
69. In Sec. 53.020, revise the definitions for ``Construction'' and
``Quality Assurance (QA)'' to read as follows:
Sec. 53.020 Definitions.
* * * * *
Construction means those activities which 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:
(1) 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;
(2) SSCs necessary to comply with 10 CFR part 73.
* * * * *
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 ensure the material, structure, component, or system
meets predetermined requirements.
* * * * *
0
70. In Sec. 53.040, revise paragraph (b)(7)(ii) and add (b)(7)(iii) to
read as follows:
Sec. 53.040 Written communications.
* * * * *
(b) * * *
(7) * * *
(ii) A change to an NRC-accepted QA topical report or quality
management system 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.
(iii) A change to the Safety Analysis report quality management
system under Sec. 53.1565, or a change to a licensee's NRC-accepted
quality management system 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.
* * * * *
0
71. In Sec. 53.460, revise paragraphs (b)(1) and (b)(2) to read as
follows:
Sec. 53.460 Safety categorization and special treatments.
* * * * *
(b) * * *
(1) The special treatments for SR SSCs must include meeting the
applicable quality assurance requirements from appendix B of part 50 of
this chapter or, for eligible applicants, applicable quality assurance
requirements from appendix T 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 or, for eligible applicants,
meeting selected quality assurance requirements from appendix T of part
50 of this chapter when such treatment is needed to address performance
requirements, equipment reliability, or uncertainties.
* * * * *
0
72. In Sec. 53.500, revise paragraph (b) to read as follows:
Sec. 53.500 General siting and siting assessment.
* * * * *
(b) Activities performed to identify site characteristics or
otherwise needed to determine site-specific contributors to 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 or, for eligible applicants,
meeting applicable quality assurance requirements from appendix T of
part 50 of this chapter.
0
73. In Sec. 53.610, revise paragraph (b) introductory text to read as
follows:
Sec. 53.610 Construction.
* * * * *
(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 a general license or LWA authorizing activities under Sec.
53.1130.
* * * * *
0
74. In Sec. 53.855, add paragraphs (c) and (d) to read as follows:
Sec. 53.855 Emergency preparedness.
* * * * *
(c) A licensee desiring to change its plume exposure pathway EPZ
must submit an application for a license amendment under Sec. 53.1510
and receive NRC approval before implementing the change. Any such
license amendment request must include documentation demonstrating that
the applicable State, local, and Tribal governmental authorities have
agreed to the EPZ change.
(d) A licensee desiring to change its emergency plan to comply with
either the requirements of Sec. 50.160 of this chapter or appendix E
to part 50 of this chapter and the planning standards of Sec. 50.47(b)
of this chapter, must submit an application for a license amendment
under Sec. 53.1510 and receive NRC approval before implementing the
change.
0
75. Revise Sec. 53.865 to read as follows:
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 or, for eligible OL or COL
holders, a quality management system in accordance with appendix T of
part 50 of this chapter. A written quality assurance program manual
must be developed and used to guide the conduct of the program.
0
76. In Sec. 53.1010, revise paragraph (b)(2) to read as follows:
Sec. 53.1010 Financial assurance for decommissioning.
* * * * *
(b) * * *
(2) The amount of financial assurance for decommissioning to be
provided may be based on a design-specific or site-specific cost
estimate for
[[Page 44707]]
decommissioning the facility under Sec. 53.1020.
* * * * *
0
77. Revise Sec. 53.1020 to read as follows:
Sec. 53.1020 Cost estimates for decommissioning.
(a) A certification relying on a design-specific decommissioning
cost estimate must demonstrate that there is reasonable assurance that
sufficient funds necessary for safely decommissioning the facility will
be available, when needed, and provide the factors used to develop the
design-specific decommissioning cost estimate, including reactor
technology, power level (in MWt), and costs related to labor, energy,
and waste burial. The amount to be provided must also address the
approach to annual adjustments required by Sec. 53.1030. Finally,
design-specific decommissioning cost estimates 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.
(b) Site-specific decommissioning cost estimates (DCEs) must be in
an amount that may be more, but not less, than the amount stated in
paragraph (a) of this section. Site-specific DCEs must account for the
engineering, labor, 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.
0
78. Revise Sec. 53.1040 to read as follows:
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 applicant or licensee assets and outside the
administrative control of the applicant or 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. An
applicant or licensee that has prepaid funds based on a design-specific
or site-specific decommissioning 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. An applicant or licensee may use a
credit of greater than 2 percent if the applicant's or licensee's rate-
setting authority has specifically authorized a higher rate. However,
applicants or licensees certifying only to design-specific
decommissioning cost estimates can take a pro-rata credit during the
dismantlement period (i.e., recognizing both cash expenditures and
earnings the first 7 years after shutdown). 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 applicant or licensee assets and outside the
administrative control of the applicant or 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. An applicant or licensee that has collected funds based on a
design-specific or site-specific decommissioning 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. An applicant or licensee may use a credit of greater than
2 percent if the applicant's or licensee's rate-setting authority has
specifically authorized a higher rate. However, applicants or licensees
certifying only to design-specific decommissioning cost estimates can
take a pro-rata credit during the dismantlement period (i.e.,
recognizing both cash expenditures and earnings the first 7 years after
shutdown). Actual earnings on existing funds may be used to calculate
future fund needs. An applicant or 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 an applicant or 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 an applicant or 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
[[Page 44708]]
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 applicant or 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 applicant or 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.
(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 an applicant's or licensee's customer(s), the total amount of
which over the duration of the contract(s) will provide the applicant's
or 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 application or
licensee submittal, assurance of decommissioning funding equivalent to
that provided by the mechanisms specified in paragraphs (a) through (e)
of this section. Applicants or 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.
0
79. Revise Sec. 53.1050 to read as follows:
Sec. 53.1050 NRC oversight.
The NRC reserves the right to take the following steps in order to
ensure an applicant's or 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 applicant's or licensee's State Public Utility Commission,
take additional actions as appropriate on a case-by-case basis,
including modification of an applicant's or licensee's schedule for the
accumulation of decommissioning funds.
0
80. In Sec. 53.1109, revise paragraph (g) to read as follows:
Sec. 53.1109 Contents of applications; general information.
* * * * *
(g)(1) 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 coordinate
radiological emergency preparedness activities with offsite
organizations with responsibilities for coping with emergencies
including State, local, and Tribal governmental agencies, as
applicable. Specifically, the applicant must ensure that these response
organizations are aware of the potential radiological consequences of
the facility and have been consulted on appropriate protective measures
including the extent of any emergency planning zone (EPZ) for
implementing predetermined, prompt protective measures. The application
must include information that describes the extent of the applicant's
interaction with these response organizations. 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 2 to 10 miles (3.2 to 16
km) in radius. For reactors with an authorized power level less than
300 MW thermal, the plume exposure pathway EPZ may be established at
the site boundary. The need for and size of the EPZ may also be
determined on a case-by-case basis as described in Sec. 53.1109(g)(2).
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. Emergency plans must describe
such actions as are appropriate to avoid or reduce dose within and
beyond the EPZ or site boundary and to protect the ingestion pathway.
(2) For a case-by-case EPZ determination, the applicant or licensee
must submit an 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) Dose to an individual is projected to exceed 1 rem (10
millisieverts) 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) [Reserved]
* * * * *
0
81. In 53.1130, revise the section heading and paragraphs (a)(3)(ii),
(b)(1)(i), and (c) and add paragraph (e) to read as follows:
[[Page 44709]]
Sec. 53.1130 Limited work authorizations, general licenses.
(a) * * *
(3) * * *
(ii) Information to demonstrate the applicability of a categorical
exclusion, or if a categorical exclusion is not applicable, an
environmental report in accordance with part 51 of this chapter; and
* * * * *
(b) * * *
(1) * * *
(i) The NRC staff issues the final documentation required under
NEPA and all applicable Federal environmental consultations have been
complete, in accordance with part 51 of this chapter;
* * * * *
(c) Effect of limited work authorization.
(1) 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
in a contested hearing 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).
(2) Any activities that are determined to be outside the scope of
those defined in the definition of construction in Sec. 53.020 and
that are undertaken by an applicant or on its behalf are entirely at
the risk of the applicant and have no bearing on the issuance of a
license with respect to the requirements of the Act, and rules,
regulations, or orders issued under the Act.
* * * * *
(e) Issuance of general license. A general license is hereby issued
to an applicant for a construction permit or combined license for a
utilization facility under this part for construction activities on a
site that is specified in the application, subject to the following
conditions:
(1) The applicant has submitted and the Commission has docketed a
CP or COL application for a commercial nuclear plant under this part
that meets the following criteria;
(i) The application references a reactor design for which the
Commission issued an operating license under this part or issued a
combined license under this part and made the finding under Sec.
53.1452(g) and for which the Commission afforded generic finality under
Sec. 53.1387(e) or Sec. 53.1440(d); and
(ii) The operating license or combined license described in
paragraph (e)(1)(i) of this section met the criteria for a categorical
exclusion or resulted in a finding of no significant impact from an
environmental assessment in accordance with part 51 of this chapter;
(iii) The application utilizing the general license includes a plan
for redress of any adverse environmental impact from conduct of
activities under the general license should such redress be necessary;
and
(iv) The application must contain information demonstrating that
the site characteristics are bounded by the site parameters postulated
for the approval of generic finality.
(2) The applicant may perform construction only upon notification
to the NRC Director of NRR using instructions in Sec. 53.040 before
the start of construction. The notice must state that all applicable
permits, licenses, approvals, and other entitlements in connection with
the proposed action have been obtained. The notice may be in the form
of a letter, but must contain the applicant's name, address, and the
name and means of contacting a person responsible for providing
additional information concerning construction under this general
license;
(3) All applicable Federal environmental consultations have been
completed;
(4) The general license only authorizes construction of those
generic aspects of the design of the commercial nuclear plant for which
the Commission afforded generic finality and does not authorize
installation of the reactor vessel, the reactor coolant system, or
associated reactivity control and heat removal systems;
(5) The applicant must allow for NRC inspections that the
Commission deems necessary related to activities performed under the
general license; and
(6) Any activities undertaken by the applicant or on its behalf
under the general license are entirely at the risk of the applicant and
have no bearing on the issuance of a license with respect to the
requirements of the Act, and rules, regulations, or orders issued under
the Act.
0
82. Revise Sec. 53.1161 to read as follows:
Sec. 53.1161 Extent of activities permitted.
If the activities authorized by Sec. 53.1158(c) are performed and
the early site permit holder has applied for termination, 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.
0
83. In Sec. 53.1164, revise paragraph (a) and remove and reserve
paragraph (b) to read as follows:
Sec. 53.1164 Duration of permit.
(a) An early site permit issued under this subpart will be issued
with no fixed term.
(b) [Reserved]
* * * * *
0
84. Revise Sec. 53.1173 to read as follows:
Sec. 53.1173 Application for amendment to update an early site
permit.
(a) An early site permit holder may choose to submit an application
to amend an early site permit to update the data and information on
which the permit is based at any time after issuance of the early site
permit. The early site permit holder may provide updated information on
as many issues as the early site permit holder chooses and may request
to extend the period for which the agency will afford those issues
finality up to 20 additional years from the date of the amendment's
issuance. The early site permit holder may request such an extension
for an already extended permit. The application must meet the
requirements of Sec. Sec. 53.1510 and 53.1520.
(b) An application submitted under paragraph (a) of this section
must contain all information necessary to bring up to date the
information and data contained in the previous application for those
issues the early site permit holder has chosen to update.
(c) Each application must include a complete environmental report
as required by part 51 of this chapter, or a request and justification
for a categorical exclusion under part 51 of this chapter.
(d) Any person whose interest may be affected by the update of the
permit may request a hearing on the application for the update. 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. 2.309 of this chapter.
0
85. Revise Sec. 53.1176 to read as follows:
[[Page 44710]]
Sec. 53.1176 Issuance of amendment to update an early site permit.
The Commission shall grant amendment of an early site permit only
if it determines that:
(a) 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
(b) Any new requirements the Commission may wish to impose are
necessary for adequate protection to public health and safety or common
defense and security.
Sec. 53.1179 [Removed and Reserved]
0
86. Remove and reserve Sec. 53.1179.
0
87. Revise Sec. 53.1182 to read as follows:
Sec. 53.1182 Use of site for other purposes.
(a) 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.
(b) If the permit holder no longer intends to use the site for a
nuclear power plant or for other reasons no longer wishes to hold the
permit, as described in the request, the permit holder may at any time
request the Director to terminate the early site permit. The request to
terminate the permit must comply with the filing requirements of
Sec. Sec. 53.040 and 53.1100 and identify the applicable requirements
for site redress of Sec. 53.1161. Upon request, the Director may
terminate the permit.
(c) Termination of the early site permit does not bar the permit
holder or another applicant from filing a new application for the site.
0
88. In Sec. 53.1188,
0
a. In paragraph (a)(1), remove the phrase ``or Sec. 53.1179'';
0
b. In paragraph (a)(2), wherever it appears remove the word ``renewal''
and add in its place the word ``amendment'';
0
c. Revise and republish paragraph (c)(1);
0
d. In paragraph (d), remove the word ``renewed'' and in its place the
word ``amended'';
0
e. In paragraph (e), in the last sentence add in sequential order the
reference ``53.1173''.
The revision is to read as follows:
Sec. 53.1188 Finality of early site permit determinations.
* * * * *
(c) * * *
(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;
(v) The information as required in the site safety analysis report
in accordance with Sec. 53.1146(a)(1)(vi) through (ix) has not been
updated after 20 years from the date of early site permit issuance or a
previous update of the permit by amendment, whichever is later, or
(vi)(A) Any significant environmental issue that was not resolved
in the early site permit proceeding;
(B) For an application that references an early site permit issued
or updated by amendment, whichever is later, no more than 20 years
before the submission of the application, 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; and
(C) For an application that references an early site permit, issued
or updated by amendment, whichever is later, more than 20 years before
submission of the application, any issue involving the impacts of
construction and operation of the facility regardless of whether the
early site permit proceeding resolved the issue.
* * * * *
0
89. In Sec. 53.1263,
0
a. Revise paragraphs (a)(1)(v) and (vi), and remove paragraph
(a)(1)(vii);
0
b. In paragraph (a)(4)(ii), remove the last sentence; and
0
c. Revise paragraph (b).
The revisions are to read as follows:
Sec. 53.1263 Finality of standard design certifications.
(a)(1) * * *
(v) Is necessary to correct material errors in the certification
information; or
(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.
* * * * *
(b) An applicant who references a design certification rule may
request an exemption from one or more elements of the certification
information, if one is required per Sec. 53.1525. The Commission may
grant such a request only if it determines that the exemption will
comply with the requirements of Sec. 53.080. 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.
* * * * *
0
90. In Sec. 53.1282, add paragraph (e) to read as follows:
Sec. 53.1282 Contents of applications for manufacturing licenses;
other application content.
* * * * *
(e) Optional operational programs. An applicant may include in its
application descriptions of essentially complete programmatic controls,
operational programs, or operational requirements beyond those required
by Sec. 53.1279 in order to satisfy requirements for license
applications that may reference a manufacturing license. If approved by
the NRC as part of the manufacturing license, such programmatic
controls, operational programs, and operational requirements would have
finality under Sec. 53.1288.
0
91. In Sec. 53.1288, revise paragraphs (a)(1) and (b) to read as
follows:
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; the requirements for the
manufacture of the manufactured reactor; or the programmatic controls,
operational programs, or operational requirements, 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
[[Page 44711]]
the time the ML was issued, or to provide reasonable assurance of
adequate protection to public health and safety or common defense and
security.
* * * * *
(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 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.
0
92. In Sec. 53.1309, revise paragraph (a)(2)(i) and remove and reserve
paragraph (a)(4) to read as follows:
Sec. 53.1309 Contents of applications for construction permits;
technical information.
* * * * *
(a) * * *
(2) * * *
(i) Quality assurance program. A description of the QAP or, for
eligible applicants, the quality management system 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 of part 50 of this chapter or, for eligible
applicants, how the requirements of appendix T of part 50 of this
chapter will be satisfied.
* * * * *
(4) [Reserved]
* * * * *
0
93. In Sec. 53.1369, revise paragraph (l) and add paragraph (bb) to
read as follows:
Sec. 53.1369 Contents of applications for operating licenses;
technical information.
* * * * *
(l) Quality assurance. A description of the QAP or, for eligible
applicants, the quality management system that demonstrates compliance
with the requirements under Sec. 53.865.
* * * * *
(bb) Requests for generic finality. An applicant may include in its
application a request for generic finality, to generic aspects of the
design under this part, such that information in the application, if
approved by the NRC, is considered resolved in other proceedings where
information approved for generic finality is referenced. An application
for an operating license that requests generic finality must include
applicable site parameters postulated for the design, including the
design-basis external hazard levels for the relevant external hazards,
and an analysis and evaluation of the design in terms of those site
parameters.
0
94. In Sec. 53.1375, revise paragraph (b) to read as follows:
Sec. 53.1375 Review of applications.
* * * * *
(b) Administrative review of applications; hearings.
(1) 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.
(2) If an applicant requests generic finality under Sec.
53.1369(bb) for an OL under this part, the Commission will include a
request for generic finality as a proposed action in the notice of
proposed action required by Sec. 2.105 of this chapter.
0
95. In Sec. 53.1387, add paragraph (e) to read as follows:
Sec. 53.1387 Issuance of operating licenses.
* * * * *
(e) The Commission may afford generic finality to generic aspects
of the design of a commercial nuclear plant under this part, including
postulated site parameters, and requirements submitted pursuant to
Sec. 53.1369(bb), if it finds that the proposed generic design can be
constructed and operated at sites having characteristics that fall
within the site parameters postulated for the design in accordance with
applicable requirements and without undue risk to the health and safety
of the public.
0
96. Revise Sec. 53.1390 to read as follows:
Sec. 53.1390 Finality of operating licenses.
(a) 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.
(b) In a proceeding for the issuance of a CP, OL, or COL, or in any
enforcement hearing other than one initiated by the Commission under
paragraph (a) of this section, in which an OL issued under Sec.
53.1387 is referenced, the Commission must treat as resolved those
matters resolved in the proceeding on the application for issuance or
renewal of the referenced OL including, if applicable, the adequacy of
a reactor design, where the referenced OL was afforded finality
pursuant to Sec. 53.1387(e).
0
97. In Sec. 53.1416, revise paragraphs (a)(12) and (d) and add
paragraph (i) to read as follows:
Sec. 53.1416 Contents of applications for combined licenses;
technical information.
* * * * *
(a) * * *
(12) Quality assurance. A description of the QAP or, for eligible
applicants, the quality management system under Sec. 53.865.
* * * * *
(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) If the early site permit site safety analysis report
information required by Sec. 53.1146(a)(1)(vi) through (ix) has not
been updated after 20 years from the date of early site permit issuance
or a previous update of the permit by amendment, whichever is later;
the combined license application shall include updated information and
revised analyses, as necessary, in the final safety analysis report.
(4) 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.
(5) 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
[[Page 44712]]
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.
(6) 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.
* * * * *
(i) An applicant may include in its application a request for
generic finality, to generic aspects of the design under this part,
such that information in the application, if approved by the NRC, is
considered resolved in other proceedings where information approved for
generic finality is referenced. An application for a combined license
that requests generic finality must include applicable 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.
0
98. In Sec. 53.1422, revise paragraph (b) to read as follows:
Sec. 53.1422 Review of applications.
* * * * *
(b) Administrative review of applications; hearings.
(1) 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 contested
hearings on COLs are governed by the procedures contained in 10 CFR
part 2.
(2) If an applicant requests generic finality under Sec.
53.1416(i) for a COL under this part, the Commission will include a
request for generic finality as a proposed action in the notice of
hearing required by Sec. 2.104 of this chapter.
Sec. 53.1437 [Amended]
0
99. In Sec. 53.1437, in paragraph (c), remove the last sentence.
0
100. In Sec. 53.1440, add paragraph (d) to read as follows:
Sec. 53.1440 Issuance of combined licenses.
* * * * *
(d) The Commission may afford generic finality to generic aspects
of the design of a commercial nuclear plant under this part, including
postulated site parameters, and requirements submitted pursuant to
Sec. 53.1416(i), if it finds that the proposed generic design can be
constructed and operated at sites having characteristics that fall
within the site parameters postulated for the design in accordance with
applicable requirements and without undue risk to the health and safety
of the public.
0
101. In Sec. 53.1443, add paragraph (g) to read as follows:
Sec. 53.1443 Finality of combined licenses.
* * * * *
(g) In a proceeding for the issuance of a CP, OL, or COL, or in any
enforcement hearing other than one initiated by the Commission under
paragraph (a) of this section, in which a COL issued under Sec.
53.1440 is referenced, the Commission must treat as resolved those
matters resolved in the proceeding on the application for issuance or
renewal of the referenced COL including, if applicable, the adequacy of
a reactor design, where the referenced COL was afforded finality
pursuant to Sec. 53.1440(d).
0
102. In Sec. 53.1525, revise paragraphs (a) and (b) to read as
follows:
Sec. 53.1525 Revising certification information within a design
certification rule.
(a) A holder of a license who references a design certification
rule issued under this part must request a license amendment in
accordance with Sec. Sec. 53.1510, 53.1515, and 53.1520 if proposing
to change certification information that has been incorporated into the
license.
(b) For certification information that has not been incorporated
into the license, a holder of a license who references a design
certification rule issued under this part may make changes to the
certification information without requesting an exemption, if the
changes meet the criteria in Sec. 53.1550(a)(1) and (2) using the
specifications in Sec. 53.1550(b)(2) and (3). If an exemption is
requested, the Commission may grant such a request only if it
determines that the exemption will comply with the requirements of
Sec. 53.080.
* * * * *
0
103. In Sec. 53.1530, in paragraph (a) revise the last sentence to
read as follows:
Sec. 53.1530 Revising information within a Final Safety Analysis
Report associated with a manufacturing license.
* * * In those cases where an ML references a design certification
rule, the provisions of Sec. 53.1525 apply.
* * * * *
0
104. Revise Sec. 53.1535 to read as follows:
Sec. 53.1535 Amendments and exemptions 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.
The holder of a CP or LWA under this part may also request an
exemption, if required by Sec. 53.1525, to depart from a design
certification rule.
(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
exemptions required by Sec. 53.1525 and amendments required by Sec.
53.1525 or 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 exemption or amendment will not be granted.
0
105. Revise Sec. 53.1550 to read as follows:
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 an exemption
pursuant to Sec. 53.1525 or license amendment pursuant to Sec.
53.1510 only if--
(1) A change to the technical specifications or other certification
information 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
[[Page 44713]]
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; the revised method of evaluation
can be used under an NRC-endorsed consensus code or standard; or the
licensee has demonstrated through a documented verification,
validation, and uncertainty quantification (VVUQ) process, conducted
under a VVUQ program that meets the requirements of Sec. 50.221 of
this chapter and has been approved by the NRC for the intended
application, that the departure from a method of evaluation described
in the FSAR (as updated) meets the criteria established for credibility
in the VVUQ program.
(v) Does not result in a change to the safety classification of an
SSC from non-safety-related to safety-related, from non-safety-related
but safety-significant to safety-related, or 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) For commercial nuclear plants licensed under this part for
which alternative evaluation criteria are adopted in accordance with
Sec. 53.470, does not result in a change to the frequency or
consequences of event sequences such that the alternate evaluation
criteria are exceeded.
(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 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) The holder of an OL or COL that authorizes operation of a
manufactured reactor 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 if the changes are identical to changes approved by
the Commission by amendment to the manufacturing license for the
manufactured reactor and upon determining that implementation of the
changes will be consistent with the basis for the Commission's approval
of the amendment to the manufacturing license and not involve any
additional changes that would require an amendment to the OL or COL.
(d)(1) The licensee must maintain records of changes in the
facility and procedures made under paragraphs (a) and (c) 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) or (c) 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.
0
106. In Sec. 53.1565, in paragraph (d), revise (1)(i) introductory
text, add (1)(iii), revise (2), add (3)(i)(E), and revise (3)(ii),
(iii), and (vii) to read as follows:
Sec. 53.1565 Evaluating changes to programs included in licensing-
basis information.
* * * * *
(d) * * *
(1) * * *
(i) Each holder under this part of an OL or COL, after the
Commission makes the finding under Sec. 53.1452(g), subject to the
quality assurance criteria in appendix B of part 50 of this chapter,
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:
* * * * *
(iii) Each holder of an OL or COL after the Commission makes the
finding under Sec. 53.1452(g), subject to the quality assurance
criteria in appendix T of part 50 of this chapter, must submit changes
to the quality management system to the NRC and receive NRC approval
prior to implementation as follows:
(A) Changes made to the quality management system as presented in
the Safety Analysis Report or in a topical report must be submitted as
specified in Sec. 53.040.
(B) The submittal of a change to the Safety Analysis Report quality
management system 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
quality management system incorporating the change continues to satisfy
the criteria of appendix T of part 50 of this chapter and the Safety
Analysis Report quality management system commitments previously
accepted by the NRC (the letter need not provide the basis for changes
that correct spelling, punctuation, or editorial items).
[[Page 44714]]
(C) A copy of the forwarding letter identifying the change must be
maintained as a facility record for three years.
(D) Changes to the quality management system 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.
(2) Quality assurance program--siting, construction, and
manufacturing.
(i) Each holder of an LWA, early site permit, CP, ML, or COL,
before the Commission makes the finding under Sec. 53.1452(g), subject
to the quality assurance criteria in appendix B to part 50 of this
chapter, 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:
(A) 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.
(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 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).
(C) A copy of the forwarding letter identifying the changes 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.
(ii) Each holder of a CP or COL, before the Commission makes the
finding under Sec. 53.1452(g), subject to the quality assurance
criteria in appendix T of part 50 of this chapter, must submit changes
to the quality management system to the NRC and receive NRC approval
prior to implementation as follows:
(A) Changes made to the quality management system as presented in
the Safety Analysis Report or in a topical report must be submitted as
specified in Sec. 53.040.
(B) The submittal of a change to the Safety Analysis Report quality
management system 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
quality management system incorporating the change continues to satisfy
the criteria of appendix T of part 50 of this chapter and the Safety
Analysis Report quality management system 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 three years.
(D) Changes to the quality management system 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.
(3) * * *
(i) * * *
(E) Risk significant planning standard means the most essential
functions of emergency preparedness to ensure adequate protective
measures are taken to protect the public in the event of a radiological
emergency. For the purposes of this section, the risk significant
planning standards are classification, notification, assessment,
protective actions, staffing, and facilities.
(ii) 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) of this chapter, or an emergency
plan that meets the requirements in Sec. 50.160 of this chapter.
(iii) A licensee may make changes to its emergency plan without NRC
approval only if the licensee performs and retains an analysis
demonstrating that:
(A) For planning standards that are risk significant, the changes
do not reduce the effectiveness of the plan and the plan, as changed,
continues to meet either the risk-significant requirements of Sec.
50.160 of this chapter or the applicable requirements in appendix E to
part 50 of this chapter and the risk significant planning standards of
Sec. 50.47(b) of this chapter; and
(B) For planning standards that are not risk-significant, the plan,
as changed, continues to meet the applicable requirements.
* * * * *
(vii) The licensee must provide for annual evaluation of the
adequacy of the interfaces between the licensee and the applicable
State, local, and Tribal governments, including licensee drills,
exercises, capabilities, and procedures. The results of the evaluation,
along with recommendations for improvements, must be documented,
reported to the licensee's corporate and plant management, retained for
a period of 5 years, and must be made available to the appropriate
State, local, and Tribal governments.
* * * * *
PART 54--REQUIREMENTS FOR RENEWAL OF OPERATING LICENSES FOR NUCLEAR
POWER PLANTS
0
107. The authority citation for part 54 continues to read as follows:
Authority: Atomic Energy Act of 1954, secs. 102, 103, 104, 161,
181, 182, 183, 186, 189, 223, 234 (42 U.S.C. 2132, 2133, 2134, 2136,
2137, 2201, 2231, 2232, 2233, 2236, 2239, 2273, 2282); Energy
Reorganization Act of 1974, secs. 201, 202, 206 (42 U.S.C. 5841,
5842, 5846); 44 U.S.C. 3504 note.
Section 54.17 also issued under 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; ADVANCE Act
of 2024, sec. 301 (42 U.S.C. 2133 note).
Sec. 54.9 [Amended]
0
108. In Sec. 54.9, in paragraph (b), remove the reference ``54.22''.
0
109. In Sec. 54.17, revise paragraph (c) to read as follows:
Sec. 54.17 Filing of application.
* * * * *
(c) An application for a renewed license may not be submitted to
the Commission earlier than the start of the period of operation that
directly precedes the requested renewal period.
* * * * *
0
110. In Sec. 54.21, add paragraph (a)(4) and remove and reserve
paragraph (c)(2) to read as follows:
[[Page 44715]]
Sec. 54.21 Contents of application--technical information.
* * * * *
(a) * * *
(4) An applicant may voluntarily use risk-informed and performance-
based alternatives to the requirements in paragraphs (a)(1), (a)(2),
and (a)(3) of this section. The application must describe the
alternatives, justify their basis, and include a description of the
underlying systematic analysis.
* * * * *
(c) * * *
(2) [Reserved]
* * * * *
Sec. 54.22 [Removed and Reserved]
0
111. Remove and reserve Sec. 54.22.
0
112. In Sec. 54.29, revise paragraphs (a) introductory text and (a)(1)
to read as follows:
Sec. 54.29 Standards for issuance of a renewed license.
(a) Appropriate actions have been identified and have been or will
be taken with respect to the matters identified in paragraphs (a)(1)
and (a)(2) of this section, such that there is reasonable assurance
that the activities authorized by the renewed license will continue to
be conducted in accordance with the CLB, and that any changes made to
the plant's CLB in order to comply with this paragraph are in accord
with the Act and the Commission's regulations. These matters are:
(1) managing the effects of aging during the period of extended
operation on the functionality of structures and components that have
been identified to require review under Sec. 54.21(a)(1) and
alternatives authorized by Sec. 54.21(a)(4); and
* * * * *
0
113. In Sec. 54.31, revise paragraphs (b) and (c) to read as follows:
Sec. 54.31 Issuance of renewed license.
* * * * *
(b) A renewed license will be issued for a fixed period of time,
not to exceed 40 years.
(c) A renewed license will become effective upon expiration of the
operating license or combined license previously in effect. Prior to
the renewed license becoming effective, the licensee shall take all
necessary steps to ensure the renewed license remains up to date,
consistent with the CLB.
* * * * *
0
114. In Sec. 54.37, revise paragraph (b) to read as follows:
Sec. 54.37 Additional records and recordkeeping requirements.
* * * * *
(b) After the renewed license is in effect, FSAR updates required
by 10 CFR 50.71(e) must contain any changes to summary descriptions of
how the effects of aging will be managed for any newly identified
systems, structures, and components that require aging management or an
evaluation of time-limited aging analyses in accordance with Sec.
54.21. Documentation of the related aging management review or
evaluation of time-limited aging analyses shall be retained in
accordance with Sec. 54.37(a).
Sec. 54.43 [Amended]
0
115. In Sec. 54.43, in paragraph (b), remove the reference ``54.22''.
PART 71--PACKAGING AND TRANSPORTATION OF RADIOACTIVE MATERIAL
0
116. The authority citation for part 71 continues to read as follows:
Authority: Atomic Energy Act of 1954, secs. 53, 57, 62, 63, 81,
161, 182, 183, 223, 234, 1701 (42 U.S.C. 2073, 2077, 2092, 2093,
2111, 2201, 2232, 2233, 2273, 2282, 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, sec. 180 (42 U.S.C. 10175);
44 U.S.C. 3504 note.
Section 71.97 also issued under Sec. 301, Pub. L. 96-295, 94
Stat. 789 (42 U.S.C. 5841 note).
0
117. In Sec. 71.55, revise and republish paragraph (g) to read as
follows:
Sec. 71.55 General requirements for fissile material packages.
* * * * *
(g) Packages containing uranium hexafluoride only are excepted from
the requirements of paragraph (b) of this section provided that:
(1) Following the tests specified in Sec. 71.73, there is no
physical contact between the valve body and any other component of the
packaging, other than at its original point of attachment, and the
valve remains leak tight;
(2) There is an adequate quality control in the manufacture,
maintenance, and repair of packagings; (3) Each package is tested to
demonstrate closure before each shipment;
(4) The uranium is enriched to not more than 10 weight percent
uranium-235; and
(5) A design feature is incorporated to protect the valve or other
fill device from impact for contents with uranium-235 enriched above 5
weight percent and up to 10 weight percent.
PART 100--REACTOR SITE CRITERIA
0
118. The authority citation for part 100 continues to read as follows:
Authority: Atomic Energy Act of 1954, secs. 103, 104, 161, 182
(42 U.S.C. 2133, 2134, 2201, 2232); Energy Reorganization Act of
1974, secs. 201, 202 (42 U.S.C. 5841, 5842); 44 U.S.C. 3504 note.
Sec. 100.1 [Amended]
0
119. In Sec. 100.1, in paragraphs (a) and (d), remove the word
``stationary''.
Sec. 100.2 [Amended]
0
120. In Sec. 100.2, remove the word ``stationary''.
0
121. In Sec. 100.3, add in alphabetical order the definitions ``Tier 1
reactor'' and ``Tier 2 reactor'' to read as follows:
Sec. 100.3 Definitions.
* * * * *
Tier 1 reactor means a power reactor having an unmitigated
consequence of less than 25 rem (0.25 Sv) TEDE at the site EAB.
Tier 2 reactor means a power reactor having an unmitigated
consequence of greater than 25 rem (0.25 Sv) TEDE at the site EAB or
where the unmitigated consequence is undetermined.
* * * * *
0
122. In Sec. 100.8, revise paragraph (b) to read as follows:
Sec. 100.8 Information collection requirements: OMB approval.
* * * * *
(b) The approved information collection requirements contained in
this part appear in Sec. Sec. 100.10, 100.11, 100.20, 100.21, and
100.23.
0
123. Revise subpart A section heading consisting of Sec. 100.10
through Sec. 100.11 to read as follows:
Subpart A--Evaluation Factors for Tier 1 Power and Testing Reactors
* * * * *
0
124. Revise Sec. 100.10 to read as follows:
Sec. 100.10 Factors to be considered when evaluating sites.
This approach to evaluating sites applies to Tier 1 power reactors
and testing reactors. The Commission will take the following factors
into consideration in determining the acceptability of a site for a
Tier 1 power or testing reactor:
(a) Characteristics of reactor design and proposed operation.
(b) Population density and use characteristics of the site
environs, including the exclusion area, low population zone, and
population center distance (or alternate area assessed by consideration
of societal risks and benefits).
(c) Physical characteristics of the site, including seismology,
meteorology, geology, and hydrology. Applications to
[[Page 44716]]
earthquake engineering criteria are contained in appendix S to part 50
of this chapter.
(d) Where unfavorable physical characteristics of the site exist,
the proposed site may nevertheless be found to be acceptable if the
design of the facility includes appropriate and adequate compensating
engineering safeguards.
(e) For siting criteria described in Sec. 100.11(a)(3)(i)(B), a
comparison of societal risks and societal benefits.
0
125. Revise Sec. 100.11 to read as follows:
Sec. 100.11 Determination of exclusion area, low population zone, and
population considerations.
(a) As an aid in evaluating a proposed site, an applicant should
assume a fission product release,\[1]\ the expected demonstrable leak
rates from potential flow paths, and the meteorological conditions
pertinent to the site to derive an exclusion area, a low population
zone, and population center distance. For the purpose of this analysis
the applicant should determine the following:
(1) An exclusion area of such size that an individual located at
any point on its boundary 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 \[2]\ TEDE.
(2) A low population zone of such size that an individual located
at any point on its outer boundary 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 total radiation dose
in excess of 25 rem (0.25 Sv) TEDE.
(3)(i) The reactor site must either:
(A) 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
(B) Be found acceptable to the NRC based on assessments of societal
risks in comparison to societal benefits for the specific site.
(ii) 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.
(b) [Reserved]
\[1]\ The fission product release assumed for these calculations
should be based upon a major accident, hypothesized for purposes of
site analysis or postulated from considerations of possible
accidental events to bound a broad range of design basis accidents.
\[2]\ The use of 25 rem (0.25 Sv) 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 reactor sites with respect to
potential reactor accidents of exceedingly low probability of
occurrence, and low risk of public exposure to radiation in the
event of an accident.
0
126. Revise subpart B section heading to read as follows:
Subpart B--Evaluation Factors for Tier 2 Power Reactor Site
Applications
* * * * *
0
127. In Sec. 100.20, revise the introductory paragraph to read as
follows:
Sec. 100.20 Factors to be considered when evaluating sites.
This approach to evaluating sites applies to Tier 2 power reactors.
The Commission will take the following factors into consideration in
determining the acceptability of a site for a Tier 2 power reactor:
* * * * *
0
128. In Sec. 100.21, redesignate footnote 3 as footnote 1 and revise
footnote 1 and paragraphs (b), (g), and (h) to read as follows:
Sec. 100.21 Non-seismic siting criteria.
* * * * *
(b) (1) The reactor site must either:
(i) Provide a population center distance, as defined in Sec.
100.3, of at least one and one-third times the distance from the
reactor to the outer boundary of the low population zone; or
(ii) Be found acceptable to the NRC based on assessments of
societal risks in comparison to societal benefits for the specific
site;
(2) 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;
* * * * *
(g) Physical characteristics unique to the proposed site that could
pose a significant impediment to the development of emergency plans
must be identified; and
(h) 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
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 \[1]\.
\[1]\ Examples of these factors include, but are not limited to,
such factors as the higher population density site having superior
seismic characteristics, better access to skilled labor for
construction, better rail and highway access, shorter transmission
line requirements, the ability to use existing infrastructure for a
retired fossil fuel power plant, or less environmental impact on
undeveloped areas, wetlands or endangered species, etc. Some of
these factors are included in, or impact, the other criteria
included in this section.
0
129. In Sec. 100.23, revise paragraph (a) to read as follows:
Sec. 100.23 Geologic and seismic siting criteria.
* * * * *
(a) Applicability. The requirements in paragraphs (c) and (d) of
this section apply to applicants for an early site permit or combined
license pursuant to part 52 of this chapter, or a construction permit
or operating license for a nuclear power plant pursuant to part 50 of
this chapter that do not meet the entry criteria for subpart A of this
part.
* * * * *
Appendix A to Part 100 [Removed]
0
130. Remove appendix A to part 100.
Dated: July 14, 2026.
For the Nuclear Regulatory Commission.
Jody Martin,
Secretary of the Commission.
[FR Doc. 2026-14341 Filed 7-15-26; 8:45 am]
BILLING CODE 7590-01-P