[Federal Register Volume 85, Number 192 (Friday, October 2, 2020)]
[Rules and Regulations]
[Pages 62199-62207]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2020-21505]
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NUCLEAR REGULATORY COMMISSION
10 CFR Part 50
[NRC-2017-0151]
RIN 3150-AK07
Reactor Vessel Material Surveillance Program
AGENCY: Nuclear Regulatory Commission.
ACTION: Direct final rule.
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SUMMARY: The U.S. Nuclear Regulatory Commission (NRC) is amending the
reactor vessel material surveillance program requirements for
commercial light-water power reactors. This direct final rule revises
the requirements associated with the testing of specimens contained
within surveillance capsules and reporting the surveillance test
results. This direct final rule also clarifies the requirements for the
design of surveillance programs and the capsule withdrawal schedules
for surveillance capsules in reactor vessels purchased after 1982.
These changes reduce regulatory burden, with no effect on public health
and safety.
DATES: This direct final rule is effective February 1, 2021, unless
significant adverse comments are received by November 2, 2020. If this
direct final rule is withdrawn as a result of such comments, timely
notice of the withdrawal will be published in the Federal Register.
Comments received after this date will be considered if it is practical
to do so, but the NRC is able to ensure consideration only for comments
received on or before this date. Comments received on this direct final
rule will also be considered to be comments on a companion proposed
rule published in the Proposed Rules section of this issue of the
Federal Register.
ADDRESSES: Please refer to Docket ID NRC-2017-0151 when contacting the
NRC about the availability of information for this action. You may
obtain publicly-available information related to this action by any of
the following methods:
Federal Rulemaking website: Go to https://www.regulations.gov and search for Docket ID NRC-2017-0151. Address
questions about NRC dockets to Carol Gallagher; telephone: 301-415-
3463; email: [email protected]. For technical questions, contact
the individuals listed in the FOR FURTHER INFORMATION CONTACT section
of this document.
NRC's Agencywide Documents Access and Management System
(ADAMS): You may obtain publicly-available documents online in the
ADAMS Public Documents collection at https://www.nrc.gov/reading-rm/adams.html. To begin the search, select ``ADAMS Public Documents'' and
then select ``Begin Web-based ADAMS Search.'' For problems with ADAMS,
please contact the NRC's Public Document Room (PDR) reference staff at
1-800-397-4209, at 301-415-4737, or by email to [email protected].
For the convenience of the reader, instructions about obtaining
materials referenced in this document are provided in the
``Availability of Documents'' section.
Attention: The PDR, where you may examine and order copies
of public documents is currently closed. You may submit your request to
the PDR via email at [email protected] or call 1-800-397-4209
between 8:00 a.m. and 4:00 p.m. (EST), Monday through Friday, except
Federal holidays.
FOR FURTHER INFORMATION CONTACT: Stewart Schneider, Office of Nuclear
Material Safety and Safeguards, 301-415-4123, email:
[email protected], or On Yee, Office of Nuclear Reactor
Regulation, telephone: 301-415-1905, email: [email protected]. Both are
staff of the U.S. Nuclear Regulatory Commission, Washington, DC 20555-
0001.
SUPPLEMENTARY INFORMATION:
[[Page 62200]]
Table of Contents
I. Obtaining Information and Submitting Comments
II. Procedural Background
III. Background
IV. Discussion
V. Section-by-Section Analysis
VI. Regulatory Flexibility Certification
VII. Regulatory Analysis
VIII. Backfitting and Issue Finality
IX. Cumulative Effects of Regulation
X. Plain Writing
XI. Environmental Impact--Categorical Exclusion
XII. Paperwork Reduction Act Statement
XIII. Congressional Review Act
XIV. Compatibility of Agreement State Regulations
XV. Voluntary Consensus Standards
XVI. Availability of Documents
I. Obtaining Information and Submitting Comments
A. Obtaining Information
Please refer to Docket ID NRC-2017-0151 when contacting the NRC
about the availability of information for this action. You may obtain
publicly-available information related to this action by any of the
following methods:
Federal Rulemaking website: Go to https://www.regulations.gov and search for Docket ID NRC-2017-0151.
NRC's Agencywide Documents Access and Management System
(ADAMS): You may obtain publicly-available documents online in the
ADAMS Public Documents collection at https://www.nrc.gov/reading-rm/adams.html. To begin the search, select ``Begin Web-based ADAMS
Search.'' For problems with ADAMS, please contact the NRC's 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.
Attention: The PDR, where you may examine and order copies
of public documents, is currently closed. You may submit your request
to the PDR via email at [email protected] or call 1-800-397-4209
between 8:00 a.m. and 4:00 p.m. (EST), Monday through Friday, except
Federal holidays.
B. Submitting Comments
Please include Docket ID NRC-2017-0151 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. Procedural Background
Because the NRC anticipates that this action will be non-
controversial, the NRC is using the ``direct final rule process'' for
this rule. The direct final rule will become effective on February 1,
2021. However, if the NRC receives significant adverse comments on this
direct final rule by November 2, 2020, then the NRC will publish a
document that withdraws this action and will subsequently address the
comments received in a final rule as a response to the companion
proposed rule published in the Proposed Rule section of this issue of
the Federal Register. Absent significant modifications to the proposed
revisions requiring republication, the NRC will not initiate a second
comment period on this action.
A significant adverse comment is a comment where the commenter
explains why the rule would be inappropriate, including challenges to
the rule's underlying premise or approach, or would be ineffective or
unacceptable without a change. A comment is adverse and significant if:
(1) The comment opposes the rule and provides a reason sufficient
to require a substantive response in a notice-and-comment process. For
example, a substantive response is required when:
(a) The comment causes the NRC to reevaluate (or reconsider) its
position or conduct additional analysis;
(b) The comment raises an issue serious enough to warrant a
substantive response to clarify or complete the record; or
(c) The comment raises a relevant issue that was not previously
addressed or considered by the NRC.
(2) The comment proposes a change or an addition to the rule, and
it is apparent that the rule would be ineffective or unacceptable
without incorporation of the change or addition.
(3) The comment causes the NRC staff to make a change (other than
editorial) to the rule.
For detailed instructions on filing comments, please see the
ADDRESSES section of this document.
III. Background
A. Description of a Reactor Vessel Material Surveillance Program
The reactor vessel and its internal components support and align
the fuel assemblies that make up the reactor core and provide a flow
path to ensure adequate heat removal from the fuel assemblies. The
reactor vessel also provides containment and a floodable volume to
maintain core cooling in the event of an accident causing loss of the
primary coolant. It is a cylindrical shell with a welded hemispherical
bottom head and a removable hemispherical upper head. Some vessel
shells were fabricated from curved plates that were joined by
longitudinal and circumferential welds. Others were manufactured using
forged rings and, therefore, only have circumferential welds that join
the rings. These plate and forging materials are referred to as base
metals. Maintenance of the structural integrity of the reactor vessel
is essential in ensuring plant safety, because there is no redundant
system to maintain core cooling in the event of a vessel failure.
One characteristic of reactor vessel steels is that their material
properties change as a function of temperature and neutron irradiation.
The primary material property of interest for the purposes of reactor
vessel integrity is the fracture toughness of the reactor vessel
material. Extensive experimental work determined that Charpy impact
energy tests, which measure the amount of energy required to fail a
small material specimen, can be correlated to changes in fracture
toughness of a material. Thus, the Charpy impact specimens \1\ from the
beltline \2\ materials (i.e., base metal, weld metal, and heat-affected
zone) became the standard to assess the change in fracture toughness in
ferritic steels.
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\1\ A Charpy impact specimen is a bar of metal, or other
material, having a V-groove notch machined across the 10 mm
thickness dimension.
\2\ A definition of the beltline or beltline region is provided
in appendix G to 10 CFR part 50.
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The fracture toughness of reactor vessel materials decreases with
decreasing temperature and with increasing irradiation from the
reactor. The decrease in fracture toughness due to neutron irradiation
is referred to as ``neutron embrittlement.'' The fracture toughness of
reactor vessel materials is determined by using fracture toughness
curves in the American Society of Mechanical Engineers (ASME) Code,
[[Page 62201]]
which are indexed to the reference temperature for nil-ductility
transition (RTNDT), as specified in ASME Boiler and Pressure
Vessel Code, Section II, ``Materials.'' To account for the effects of
neutron irradiation, the increase in RTNDT is equated to the
increase in the 30 ft-lb index temperature from tests of Charpy-V notch
impact specimens irradiated in capsules as a part of the surveillance
program. The surveillance program includes Charpy impact specimens of
the base and weld metals for the reactor vessel in each surveillance
capsule. These surveillance capsules are exposed to the same operating
conditions as the reactor vessel, and because the capsules are located
closer to the reactor core than the reactor vessel inner diameter, the
surveillance specimens are generally exposed to higher neutron
irradiation levels than those experienced by the reactor vessel at any
given time.
As a result of the surveillance capsule's location within the
reactor vessel, the test specimens generally reflect changes in
fracture toughness due to neutron embrittlement in advance of what the
reactor vessel experiences and provide insight to the future condition
of the reactor vessel. Therefore, the NRC instituted reactor vessel
material surveillance programs as a requirement of appendix H,
``Reactor Vessel Material Surveillance Program Requirements'' (appendix
H), to part 50 of title 10 of the Code of Federal Regulations (10 CFR),
``Domestic Licensing of Production and Utilization Facilities,'' so
that the placement and testing of Charpy impact specimens in capsules
between the inner diameter vessel wall and the core can provide data
for assessing and projecting the change in fracture toughness of the
reactor vessel.
The purpose for requiring a reactor vessel material surveillance
program is to monitor changes in the fracture toughness properties in
the beltline region of the reactor vessel and to use this information
to analyze the reactor vessel integrity. Surveillance programs are
designed not only to examine the current status of reactor vessel
material properties but also to predict the changes in these properties
resulting from the cumulative effects of neutron irradiation.
The determination as to whether a commercial nuclear power reactor
vessel requires a material surveillance program under appendix H to 10
CFR part 50 is made at the time of plant licensing under 10 CFR part 50
or 10 CFR part 52, ``Licenses, Certifications, and Approvals for
Nuclear Power Plants.'' If this surveillance program is required, it is
designed and implemented at that time using the existing requirements.
Certain aspects of the program, such as the specific materials to be
monitored, the number of required surveillance capsules to be inserted
in the reactor vessel, and the initial capsule withdrawal schedule were
designed for the original licensed period of operation (i.e., 40
years). The editions of the ASTM International (ASTM) E 185, which are
incorporated by reference in appendix H to 10 CFR part 50, recommend
three, four, or five surveillance capsules to be included in the design
of reactor vessel material surveillance programs for the original
licensed period of operation, based on the irradiation sensitivity of
the material used to fabricate the reactor vessel.\3\ Most plants have
included several additional surveillance capsules beyond the number
recommended by ASTM E 185. These capsules are referred to as ``standby
capsules.'' The surveillance program for each reactor vessel provides
assurance that the plant's operating limits (e.g., the pressure-
temperature limits) continue to meet the provisions in Appendix G of
ASME Boiler and Pressure Vessel Code, Section XI, ``Rules for Inservice
Inspection of Nuclear Power Plant Components,'' as required by appendix
G, ``Fracture Toughness Requirements,'' to 10 CFR part 50. The program
also provides assurance that the reactor vessel material upper shelf
energy meets the requirements of appendix G to 10 CFR part 50. These
assessments are used to ensure the integrity of the reactor vessel.
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\3\ The requirements in appendix H to 10 CFR part 50 are based,
in part, on the information contained within ASTM E 185-73,
``Standard Recommended Practice for Surveillance Tests for Nuclear
Reactor Vessels;'' ASTM 185-79, ``Standard Practice for Conducting
Surveillance Tests for Light-Water Cooled Nuclear Power Reactor
Vessels;'' and ASTM E 185-82, ``Standard Practice for Conducting
Surveillance Tests for Light-Water Cooled Nuclear Power Reactor
Vessels,'' which are incorporated by reference.
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In addition to the Charpy impact specimens for determining the
embrittlement in the reactor vessel, the surveillance capsules
typically contain neutron dosimeters, thermal monitors, and tension
specimens.\4\ Surveillance capsules may also contain correlation
monitor material, which is a material with composition, properties, and
response to radiation that have been well characterized. The overall
accuracy of neutron fluence measurements is dependent upon knowledge of
the neutron spectrum. Therefore, a variety of neutron detector
materials (dosimetry wires) are included in each surveillance capsule
and used in the determination of neutron fluence for the vessel. The
thermal monitors that are placed in the capsules (e.g., low-melting-
point elements or eutectic alloys) are used to identify the irradiated
specimen's maximum exposure temperature.
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\4\ Tension specimens have a standardized sample cross-section,
with two shoulders and a gage (section) in between.
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B. Current Requirements Under Appendix H to 10 CFR Part 50
Appendix H to 10 CFR part 50 requires light-water nuclear power
reactor licensees to have a reactor vessel material surveillance
program to monitor changes in the fracture toughness properties of the
reactor vessel materials adjacent to the reactor core in the beltline
region. Unless it can be shown that the end of design life neutron
fluence is below certain criteria, the NRC requires licensees to
implement a materials surveillance program that tests irradiated
material specimens that are located in surveillance capsules in the
reactor vessels. The program evaluates changes in material fracture
toughness and thereby assesses the integrity of the reactor vessel. For
each capsule withdrawal, the test procedures and reporting requirements
must meet the requirements of ASTM E 185-82, ``Standard Practice for
Conducting Surveillance Tests for Light-Water Cooled Reactor Vessels,''
to the extent practicable for the configuration of the specimens in the
capsule.
The design of the surveillance program and the withdrawal schedule
must meet the requirements of the edition of ASTM E 185 that is current
on the issue date of the ASME Code to which the reactor vessel was
purchased. Later editions of ASTM E 185, up to and including those
editions through 1982, may be used. Appendix H to 10 CFR part 50
specifically incorporates by reference ASTM E 185-73, ``Standard
Recommended Practice for Surveillance Tests for Nuclear Reactor
Vessels;'' ASTM E 185-79, ``Standard Practice for Conducting
Surveillance Tests for Light-Water Cooled Nuclear Power Reactor
Vessels,'' and ASTM E 185-82. In sum, the surveillance program must
comply with ASTM E 185, as modified by appendix H to 10 CFR part 50.
The number, design, and location of these surveillance capsules within
the reactor vessel are established during the design of the program,
before initial plant operation.
Appendix H to 10 CFR part 50 also specifies that each capsule
withdrawal and subsequent test results must be the subject of a summary
technical report to be submitted to the NRC within one year of the date
of capsule withdrawal,
[[Page 62202]]
unless an extension is granted by the Director, Office of Nuclear
Reactor Regulation. The NRC uses the results from the surveillance
program to assess licensee submittals related to pressure-temperature
limits under appendix G to 10 CFR part 50 and to assess pressurized
water reactor licensee's compliance with either Sec. 50.61, ``Fracture
toughness requirements for protection against pressurized thermal shock
events,'' or Sec. 50.61a, ``Alternate fracture toughness requirements
for protection against pressurized thermal shock events.''
C. The Need for Rulemaking
When appendix H to 10 CFR part 50 was established as a requirement
(38 FR 19012; July 17, 1973), limited information and data were
available on the subject of reactor vessel embrittlement. Thus,
appendix H to 10 CFR part 50 required the inclusion of a comprehensive
collection of specimen types representing the reactor vessel beltline
materials in each surveillance capsule. Since 1973, a significant
number of surveillance capsules have been withdrawn and tested.
Analyses of these results support reconsidering the specimen types
required for testing, and the required time for reporting the results
from surveillance capsule testing. One outcome of this effort was that
some specimen types were found to contribute to the characterization of
reactor vessel embrittlement, while others did not. Therefore, the NRC
determined that these latter types were unnecessary to meet the
objectives of appendix H to 10 CFR part 50 and should no longer be
required. Revising appendix H to 10 CFR part 50 to address this
situation reduces the regulatory burden on licensees of data
collection, with no effect on public health and safety.
In 1983, appendix H to 10 CFR part 50 was revised to require
licensees to submit test results to the NRC within one year of the date
of capsule withdrawal, unless an extension is granted by the Director,
Office of Nuclear Reactor Regulation (48 FR 24008; May 27, 1983). As
stated in the 1983 rulemaking, the reason for the requirement was the
need for timely reporting of test results and notification of any
problems. At that time, there was a limited amount of data from
irradiated materials from which to estimate embrittlement trends of
reactor vessels at nuclear power plants, making it important to receive
timely reporting of test results.
Licensees that participate in an integrated surveillance program
have found it challenging to meet this one-year requirement. This is
related to the fact that an integrated surveillance program requires
coordination among the multiple licensees participating in the
program.\5\ A significant number of test specimens have been analyzed
since 1983, the results of which support a reduced need for prompt
reporting of the test results. Based on this, the NRC has determined
that the reporting requirement in appendix H to 10 CFR part 50 should
be revised. Extending the reporting period allows for more time for
licensee coordination and should help eliminate the need for licensees
to prepare and submit extension requests and for the NRC to review such
requests. This revision has no effect on public health and safety.
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\5\ Appendix H to 10 CFR part 50 permits the use of an
integrated surveillance program (ISP) as an alternative to a plant-
specific surveillance program. In an ISP, the representative
materials chosen for surveillance of a reactor vessel are irradiated
in one or more other reactor vessels that have similar design and
operating features. The data obtained from these test specimens may
then be used in the analysis of other plants participating in the
program.
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D. Regulatory Basis To Support Rulemaking
In January 2019, the Commission issued Staff Requirements
Memorandum (SRM)-COMSECY-18-0016, ``Request Commission Approval to Use
the Direct Final Rule Process to Revise the Testing and Reporting
Requirements in 10 CFR part 50, Appendix H, Reactor Vessel Material
Surveillance Program Requirements (RIN 3150-AK07),'' approving
publication of the supporting regulatory basis and use of the direct
final rule process. On April 3, 2019, the NRC issued the regulatory
basis which provides an in-depth discussion on the technical merits of
this rulemaking (84 FR 12876).\6\ The regulatory basis includes
additional information on the regulatory framework, types of reactor
vessel material surveillance programs, regulatory topics that initiated
this rulemaking effort, and options to address these topics. The
regulatory basis shows that there is sufficient justification to
proceed with rulemaking to amend appendix H to 10 CFR part 50 to reduce
certain test specimens and extend the period to submit surveillance
capsule reports to the NRC. In addition, in SRM-COMSECY-18-0016, the
Commission directed the staff to clarify the requirements for the
design of surveillance programs and the withdrawal schedules for
reactor vessels purchased after 1982. These revisions will not
establish any additional requirements for the current fleet of
operating reactors.
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\6\ A subsequent notification was published on April 12, 2019
(84 FR 14845), to correct the ADAMS accession number for the
regulatory basis.
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IV. Discussion
The purpose of this action is to reduce the regulatory burden on
reactor licensees and the NRC that is associated with test specimens
contained within surveillance capsules and the reporting of
surveillance test results, with no effect on public health and safety.
This action also clarifies the requirements for the design of
surveillance programs and the withdrawal schedules for reactor vessels
purchased after 1982. The NRC has determined that the following
revisions to appendix H to 10 CFR part 50 achieve the goal of reducing
regulatory burden. These revisions do not establish any additional
requirements for the current fleet of operating reactors.
1. Heat-Affected Zone Specimens
The editions of ASTM E 185 incorporated by reference in appendix H
to 10 CFR part 50 specify that the surveillance test specimens shall
include base metal, weld metal, and heat-affected zone materials. Heat-
affected zone specimens were first required in reactor vessel material
surveillance programs in 1966 (ASTM E 185-66, ``Recommended Practice
for Surveillance Tests on Structural Materials in Nuclear Reactors'').
Cracks in heat-affected zone material had been observed to cause the
failure of components in non-nuclear applications, and from early
research, these failures were in heat-affected zone materials with high
hardness measurements, which is associated with low fracture toughness.
The heat-affected zone has been shown to exhibit superior fracture
toughness compared to the base metal. In addition, test results from
surveillance specimens have shown significant scatter of the heat-
affected zone Charpy test data because of the inhomogeneous nature of
the heat-affected zone material. This was the basis for eliminating the
requirement for heat-affected zone specimens after the 1994 edition of
ASTM E 185; thus, it is no longer prudent to require the inclusion or
testing of heat-affected zone materials.
For these reasons, the NRC is revising appendix H to 10 CFR part 50
to make optional the requirement to include or test heat-affected zone
specimens as part of the reactor vessel material surveillance program.
For existing capsules that are currently in the reactor vessel,
licenses can continue their practice to test the heat-affected zone
specimens. For new and reconstituted
[[Page 62203]]
capsules \7\ that may be inserted into the reactor vessel in the
future, licensees are no longer required to have heat-affected zone
specimens in the capsules but could choose to continue this practice.
This revision has no effect on public health and safety.
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\7\ A reconstituted capsule contains specimens from previously
tested capsules.
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2. Tension Specimens
The editions of ASTM E 185 currently incorporated by reference in
appendix H to 10 CFR part 50 specify the following with respect to
tensile testing:
(1) For unirradiated material, tension specimens shall be tested
for both the base and weld material at specified temperatures.
(2) For irradiated material, tension specimens shall be included
for both the base and weld material and tested at specified
temperatures.
(3) Tensile testing shall be conducted in accordance with ASTM
Method E 8, ``Methods of Tension Testing of Metallic Materials,'' and
ASTM E 21, ``Recommended Practice for Elevated Temperature Tension
Tests of Metallic Materials.''
The variation of tensile properties (e.g., yield strength, tensile
strength, and elongation) with test temperatures is established by
testing tension specimens over a range of temperatures. Performing
tensile tests before and after irradiation permits quantification of
the hardening effect due to irradiation using the change in yield
strength. Tensile data provide an indication of the radiation-induced
strength property changes in the reactor vessel material and serve as a
consistency check relative to Charpy data.
Past experience and test results have demonstrated that the
differences in the test temperatures specified in ASTM E 185 can be
small, which could yield small differences in tensile properties and
redundant tensile information. Eliminating one test temperature and
testing at room temperature and service temperature at all irradiation
levels, allows for the comparison of the change in strength properties
due to irradiation and temperature.
For these reasons, the NRC is revising appendix H to 10 CFR part 50
to require the inclusion or testing of only one tension specimen at
room temperature and one tension specimen at service temperature, for
all materials and irradiation levels as part of the reactor vessel
material surveillance program. This reduces the number of tension
specimens required in new and reconstituted surveillance capsules and
for testing in existing surveillance capsules. For existing capsules
that are currently in the reactor vessel, licensees can continue their
practice of testing the tension specimens in accordance with ASTM E
185. For new and reconstituted capsules that may be inserted into the
reactor vessel in the future, licensees could choose to continue this
practice. This revision has no effect on public health and safety.
3. Correlation Monitor Material
Correlation monitor material is a well characterized reactor vessel
material that has been included in many surveillance capsules.
Correlation monitor material is selected so that it has a comparable
composition and processing history to the reactor vessel material. The
purpose of a correlation monitor material in a surveillance capsule is
to provide reference data for comparison to the established trends for
the correlation monitor material.
The editions of ASTM E 185 currently incorporated by reference in
appendix H to 10 CFR part 50 specify that it is optional to include
correlation monitor material in surveillance capsules. These editions
of ASTM E 185 do not explicitly indicate whether correlation monitor
material shall be tested if it was optionally included in a
surveillance capsule. Therefore, it is ambiguous whether correlation
monitor material testing is required even though it is optional to
include this material in surveillance capsules. In practice, the
testing of correlation monitor material has demonstrated variability in
the measured material properties of the correlation monitor material,
which has limited the practical use of the data.
For these reasons, the NRC is revising appendix H to 10 CFR part 50
to clarify that testing of correlation monitor material is optional
when included in existing, new, and reconstituted surveillance
capsules. This revision has no effect on public health and safety.
4. Thermal Monitors
ASTM E 185-82 specifies that the surveillance capsules shall
include one set of temperature monitors (also known as ``thermal
monitors'') that are located within the capsule where the specimen
temperature is predicted to be the maximum, and additional sets of
temperature monitors may be placed at other locations to characterize
the temperature profile. The standard specifies reporting of the
temperature monitor results and an estimate of the maximum capsule
exposure temperature.
Irradiation temperature is one of the parameters that is closely
correlated with the effects of neutron embrittlement of reactor vessel
steels, with lower embrittlement measured at higher irradiation
temperatures within a range close to the standard operating temperature
of 288 degrees Celsius (550 degrees Fahrenheit). Therefore, knowledge
of the irradiation temperature history of surveillance capsules is
important to ensure that the surveillance data are properly interpreted
and do not portray a non-conservative estimate of the reactor vessel
neutron embrittlement.
Temperature monitors are targeted to melt at specific temperatures,
normally somewhat higher than the planned operating temperature, to
identify the highest temperature seen by the surveillance capsule. The
monitors provide an indication of whether the melt temperature was
reached but they do not provide a time-based exposure history of the
monitor.
Several factors can complicate the interpretation of the
information from temperature monitors. The first complication arises
when the surveillance capsule experiences a short duration thermal
transient that increases the coolant inlet temperature. This could
result in a positive indication from the temperature monitors, which is
insignificant to the overall exposure conditions of the surveillance
capsule. A second complication is caused by possible interpretation
issues, where apparent melting of the temperature monitors is caused by
long-term exposure of the monitor to temperatures near, but below, its
melting point.
For these reasons, the NRC is revising appendix H to 10 CFR part 50
to make optional the requirement to include or evaluate temperature
monitors as part of the reactor vessel material surveillance program.
For existing capsules that are currently in the reactor vessel,
licensees can continue their practice of evaluating the temperature
monitors. For new and reconstituted capsules that may be inserted into
the reactor vessel in the future, licensees are no longer required to
include temperature monitors in the capsules but could choose to
continue this practice. As an alternative to these temperature
monitors, an estimate of the average capsule temperature during full
power operation for each reactor fuel cycle will provide the
irradiation temperature history of the surveillance capsule. This
revision has no effect on public health and safety.
5. Surveillance Test Results Reporting
Appendix H to 10 CFR part 50 currently requires that within one
year of the date of the surveillance capsule withdrawal, a summary
technical report be submitted to the NRC that contains
[[Page 62204]]
the data required by ASTM E 185, and the results of all fracture
toughness tests conducted on the beltline materials in the irradiated
and unirradiated conditions, unless an extension is granted by the
Director, Office of Nuclear Reactor Regulation.
This one-year requirement in appendix H to 10 CFR part 50 became
effective on July 26, 1983 (48 FR 24008), with the primary purpose of
timely reporting of test results and notification of any problems
determined from surveillance tests. This was important because there
was a limited amount of available data from irradiated materials from
which to estimate embrittlement trends. An extensive amount of
embrittlement data has been collected and analyzed since this time, the
results of which support the reduced need for prompt reporting of the
test results.
Licensees participating in an integrated surveillance program have
found it challenging to meet the one-year requirement to submit a
report following each capsule withdrawal. In an integrated surveillance
program, the representative materials chosen for a reactor are
irradiated in one or more other reactors that have similar design and
operating features. The data obtained from these test specimens may
then be used in the analysis of other plants participating in the
program. Implementation of the integrated surveillance program requires
significant coordination among the multiple licensees participating in
the program. Historically, these licensees have requested a 6-month
extension to this reporting requirement and, to date, the Director of
the NRC Office of Nuclear Reactor Regulation, has granted them.
Furthermore, as surveillance capsules remain in the reactor vessel to
support operation through 60 years and 80 years, longer periods of
radioactive decay may be needed before the capsules can be shipped to
testing facilities. Licensees may find it burdensome to meet the one-
year reporting requirement under these circumstances.
For these reasons, the NRC is revising appendix H to 10 CFR part 50
to increase the time given to licensees to submit a summary technical
report of each capsule withdrawal and the test results from 1 year to
18 months. This revision has no effect on public health and safety.
6. Design of the Surveillance Program
Appendix H to 10 CFR part 50 is also being revised to clarify the
edition of ASTM E 185 that is required for a reactor vessel purchased
after 1982. Currently, there is the potential to misinterpret the
regulation as requiring the use of an edition of ASTM E 185 that is not
incorporated by reference in appendix H to 10 CFR part 50. Therefore,
the NRC is revising appendix H to 10 CFR part 50 to clarify that for
reactor vessels purchased after 1982, the design of the surveillance
program and the withdrawal schedule must meet the requirements of ASTM
E 185-82 (i.e., the latest edition of ASTM E 185 that is incorporated
by reference in appendix H to 10 CFR part 50).
License Renewal and Subsequent License Renewal
Surveillance programs that include the withdrawal schedule required
by appendix H to 10 CFR part 50 were originally established and
designed for the initial 40-year operating license of a nuclear power
plant. The objective of this program during extended plant operations
\8\ remains the same as it was during the initial 40-year operating
license, which is to continue monitoring changes in fracture toughness
of the reactor vessel materials to ensure the integrity of the reactor
vessel. This direct final rule does not revise appendix H to 10 CFR
part 50 with respect to surveillance capsule withdrawal schedules
during extended plant operation.
---------------------------------------------------------------------------
\8\ The period beyond the original license of a nuclear power
plant (i.e., during license renewal to operate for 60 years and
potentially during subsequent license renewal to operate for 80
years).
---------------------------------------------------------------------------
New Reactors
New light-water nuclear power reactor designs are substantially
similar to operating reactors with regard to the relevant
considerations for establishing adequate surveillance programs under
appendix H to 10 CFR part 50. These similarities include proposed
materials, fabrication methods, and operating environments. The
proposed withdrawal schedules from ASTM E 185 are constructed to
provide early evidence of material behavior which is of particular
interest for a new or novel design with little or no operating
experience. Consequently, the NRC is not revising appendix H to 10 CFR
part 50 to address new light-water nuclear power reactor designs
separately from existing reactors.
V. Section-by-Section Analysis
The following paragraphs describe the specific changes being made
by this direct final rule.
Appendix H to Part 50--Reactor Vessel Material Surveillance Program
Requirements
Section III. Surveillance Program Criteria
This direct final rule revises paragraph III.B.1 to clarify the
design of surveillance programs and the capsule withdrawal schedules
for reactor vessels purchased after 1982 and to include information
regarding the use of optional provisions. This direct final rule also
adds new paragraph III.B.4 that makes optional certain aspects of ASTM
E 185.
Section IV. Report of Test Results
This direct final rule revises the timeframe for the submission of
a summary technical report from 1 year to 18 months.
VI. Regulatory Flexibility Certification
Under the Regulatory Flexibility Act (5 U.S.C. 605(b)), the NRC
certifies that this direct final rule does not have a significant
economic impact on a substantial number of small entities. This direct
final 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
(Sec. 2.810).
VII. Regulatory Analysis
The NRC has prepared a regulatory analysis for this direct final
rule. The analysis examines the costs and benefits of the alternatives
considered by the NRC. Based on the analysis, the NRC concludes that
this action is cost beneficial and reduces the regulatory costs for
reactor licensees and the NRC for an issue that is not significant to
safety. This issue is not significant to safety because this direct
final rule reduces the testing of some specimens and eliminates the
testing of other specimens that were found not to provide meaningful
information to assess the integrity of the reactor vessel. Also,
extending by 6 months the period for submitting the report of test
results to the NRC is not significant to safety. This is because the
increase in neutron fluence over 6 months is very small, and therefore
the projected increase in embrittlement for the 6-month period would
also be very small. This small impact, in conjunction with the margin
of safety that is inherent in the pressure-temperature limit curves,
minimizes any impact due to the 6-month increase.
[[Page 62205]]
VIII. Backfitting and Issue Finality
The NRC's backfitting provisions for holders of construction
permits, and applicants and holders of operating licenses and combined
licenses, appear in Sec. 50.109, ``Backfitting'' (the Backfit Rule).
Issue finality provisions, which are analogous to the backfitting
provisions in Sec. 50.109, appear in Sec. 52.63, ``Finality of
Standard Design Certifications;'' Sec. 52.83, ``Finality of Referenced
NRC Approvals; Partial Initial Decision on Site Suitability;'' Sec.
52.98, ``Finality of Combined Licenses; Information Requests;'' Sec.
52.145, ``Finality of Standard Design Approvals, Information Request;''
and Sec. 52.171, ``Finality of Manufacturing Licenses; Information
Requests.''
This direct final rule: (1) Provides licensees with a nonmandatory
relaxation from the current 1 year following a capsule withdrawal to 18
months to submit surveillance capsule test results, and (2) reduces
testing requirements by amending the NRC's regulations in appendix H to
10 CFR part 50. Because these changes are nonmandatory, licensees have
the option to comply with the revised requirements for testing certain
surveillance capsule specimens or for extending the allowable period
for submitting surveillance test results to the NRC (i.e., licensees
can continue to submit surveillance capsule test results within one
year of the date of capsule withdrawal). Therefore, this direct final
rule does not constitute backfitting or raise issue finality concerns.
IX. Cumulative Effects of Regulation
Cumulative effects of regulation (CER) consists of the challenges
licensees may face in addressing the implementation of new regulatory
positions, programs, and requirements (e.g., rulemaking, guidance,
generic letters, backfits, inspections). The CER may manifest in
several ways, including the total burden imposed on licensees by the
NRC from simultaneous or consecutive regulatory actions that can
adversely affect the licensee's capability to implement those
requirements, while continuing to operate or construct its facility in
a safe and secure manner.
The goals of the NRC's CER effort were met throughout the
development of this action. The NRC has engaged external stakeholders
at public meetings held during the development of the regulatory basis
and this direct final rule. A public meeting was held on June 1, 2017,
to provide an opportunity for the exchange of information on the scope
and related costs and benefits associated with this action. Feedback
obtained at this meeting was used in developing the regulatory basis
and regulatory analysis. A second public meeting was held on April 30,
2019, to provide information on the status and scope of this direct
final rule, and to discuss implementation and CER. Summaries of both
public meetings are available in ADAMS, as provided in the
``Availability of Documents'' section of this document.
X. 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
31883).
XI. Environmental Impact--Categorical Exclusion
The Commission has determined under the National Environmental
Policy Act of 1969, as amended, and the Commission's regulations in 10
CFR part 51, subpart A, that the direct final rule will not have a
significant effect on the quality of the human environment and,
therefore, an environmental impact statement is not required. The
principal effect of this direct final rule is to amend the reactor
vessel materials surveillance program requirements for commercial
light-water power reactors. Specifically, it amends the requirements
associated with the testing of specimens contained within surveillance
capsules and reporting the surveillance test results.
The amendments to appendix H to 10 CFR part 50 that revise the
surveillance requirements for testing specimens add optional provisions
that would need to be adopted by individual licensees. In order to
adopt these optional provisions, licensees would need to either submit
a license amendment or determine whether the optional provisions can be
implemented under 10 CFR 50.59, ``Changes, tests and experiments.''
When the 10 CFR 50.59 regulation was promulgated in 1999, the
Commission concluded that there would be no significant impact on the
environment for the types of changes to a nuclear power plant's
licensing basis that a licensee could make under this provision without
NRC review. If a license amendment is required to be submitted, the
environmental impacts of that future license amendment would be
evaluated by the NRC staff as part of the review of the license
amendment request. The amendments to appendix H to 10 CFR part 50 that
revise the recordkeeping and reporting requirements are categorically
excluded under 10 CFR 51.22(c)(3)(ii) and (iii). The NRC has also
determined that this action would involve no significant change in the
types or amounts of any effluents that may be released offsite; no
significant increase in individual or cumulative occupational radiation
exposure; and no significant increase in the potential for or
consequences from radiological accidents. In addition, the NRC has
determined that there are no significant impacts to biota, water
resources, historic properties, cultural resources, or socioeconomic
conditions in the region. As such, there are no extraordinary
circumstances that would preclude reliance on this categorical
exemption. Therefore, pursuant to 10 CFR 51.22(b), no environmental
impact statement or environmental assessment need be prepared in
connection with revising the reporting requirement under appendix H to
10 CFR part 50.
XII. Paperwork Reduction Act
The burden to the public for the information collection is
estimated to be reduced by 78 hours per response, including the time
for reviewing instructions, searching existing data sources, gathering
and maintaining the data needed, and completing and reviewing the
information collection. Further information about information
collection requirements associated with this direct final rule can be
found in the companion proposed rule published elsewhere in this issue
of the Federal Register.
This direct final rule is being issued prior to approval by the
Office of Management and Budget (OMB) of these information collection
requirements, which were submitted under OMB control number 3150-0011.
When OMB notifies us of its decision, we will publish a document in the
Federal Register providing notice of the effective date of the
information collections or, if approval is denied, providing notice of
what action we plan to take.
Send comments on any aspect of these information collections,
including suggestions for reducing the burden, to the Information
Services Branch (T6-A10M), U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, or by email to
[email protected]; and to OMB Office of Information and
Regulatory Affairs (3150-0011), Attn: Desk Officer for the Nuclear
Regulatory Commission, 725 17th Street NW, Washington, DC 20503; email:
[email protected].
[[Page 62206]]
Public Protection Notification
The NRC may not conduct or sponsor, and a person is not required to
respond to, a collection of information unless the document requesting
or requiring the collection displays a currently valid OMB control
number.
XIII. Congressional Review Act
This direct final rule is a rule as defined in the Congressional
Review Act (5 U.S.C. 801-808). However, the Office of Management and
Budget has not found it to be a major rule as defined in the
Congressional Review Act.
XIV. Compatibility of Agreement State Regulations
Under the ``Policy Statement on Adequacy and Compatibility of
Agreement State Programs,'' approved by the Commission on June 20,
1997, and published in the Federal Register (62 FR 46517; September 3,
1997), this rule is classified as compatibility ``NRC.'' Compatibility
is not required for Category ``NRC'' regulations. The NRC program
elements in this category are those that relate directly to areas of
regulation reserved to the NRC by the Atomic Energy Act of 1954, as
amended, or the provisions of 10 CFR chapter I, and although an
Agreement State may not adopt program elements reserved to the NRC, it
may wish to inform its licensees of certain requirements via a
mechanism that is consistent with a particular State's administrative
procedure laws, but does not confer regulatory authority on the State.
XV. Voluntary Consensus Standards
The National Technology Transfer and Advancement Act of 1995 (Pub.
L. 104-113) requires that Federal agencies use technical standards that
are developed or adopted by voluntary consensus standards bodies unless
using such a standard is inconsistent with applicable law or otherwise
impractical. In this direct final rule, the NRC is amending the reactor
vessel materials surveillance program requirements to reduce the
regulatory burden for an issue that is not significant to safety
associated with the testing of surveillance capsule specimens and
reporting the surveillance test results. It also clarifies the
requirements for the design of surveillance programs and the withdrawal
schedules for reactor vessels purchased after 1982. Specifically, this
direct final rule allows licensees to reduce the testing of some
specimens and eliminates the testing of other specimens that were found
not to provide meaningful information to assess the integrity of the
reactor vessel. It also extends by 6 months the period for licensees to
submit the report of test results to the NRC. The increase in neutron
fluence over 6 months is very small, and therefore the projected
increase in embrittlement over this period would also be very small.
This small impact, in conjunction with the margin of safety which is
inherent in the pressure-temperature limit curves, minimizes any impact
due to the 6-month increase. This action does not constitute the
establishment of new conditions on the ASTM standards that are
currently incorporated by reference in appendix H to 10 CFR part 50 nor
a standard that contains generally applicable requirements. This action
maintains the use of the ASTM standards that are currently incorporated
by reference in appendix H to 10 CFR part 50 but makes optional certain
aspects of the ASTM standards that have been determined not to be
necessary for the safe operation of nuclear power plants.
XVI. 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.
----------------------------------------------------------------------------------------------------------------
Document Adams Accession No./Web Link/Federal RegisterCitation
----------------------------------------------------------------------------------------------------------------
ASME Boiler and Pressure Vessel Code, Section II, https://www.asme.org.
``Materials''.
ASTM E 185-73, ``Standard Recommended Practice for https://www.astm.org.
Surveillance Tests for Nuclear Reactor Vessels''.
ASTM 185-79, ``Standard Practice for Conducting https://www.astm.org.
Surveillance Tests for Light-Water Cooled Nuclear
Power Reactor Vessels''.
ASTM E 185-82, ``Standard Practice for Conducting https://www.astm.org.
Surveillance Tests for Light-Water Cooled Nuclear
Power Reactor Vessels''.
ASME Boiler and Pressure Vessel Code, Section XI, https://www.asme.org.
Appendix G, ``Rules for Inservice Inspection of
Nuclear Power Plant Components''.
Federal Register notification--``Part 50 Final Rule- 38 FR 19012.
Licensing of Production and Utilization Facilities;
Fracture Toughness and Surveillance Program
Requirements,'' July 17, 1973.
Federal Register notification--``10 CFR Part 50 Final 48 FR 24008.
Rule, Fracture Toughness Requirements for Light-Water
Nuclear Power Reactors,'' May 27, 1983.
Rulemaking for Appendix H to 10 CFR Part 50, ``Reactor ML19038A477.
Vessel Material Surveillance Program Requirements--
Regulatory Basis,'' April 2019.
Federal Register notification--``10 CFR Part 50, 84 FR 12876.
Reactor Vessel Material Surveillance Program:
Regulatory Basis; Availability,'' April 3, 2019.
Federal Register notification--``10 CFR Part 50, 84 FR 14845.
Reactor Vessel Material Surveillance Program:
Regulatory Basis; Availability; Correction,'' April
12, 2019.
ASTM E 185-66, ``Recommended Practice for Surveillance https://www.astm.org.
Tests on Structural Materials in Nuclear Reactors``.
ASTM Method E 8, ``Methods of Tension Testing of https://www.astm.org.
Metallic Materials,''.
ASTM E21 ``Recommended Practice for Elevated https://www.astm.org.
Temperature Tension Tests of Metallic Materials.''.
Summary of April 30, 2019, Public Meeting to Discuss ML19127A050.
the Status of the Appendix H, Reactor Vessel Material
Surveillance Program Requirements Rulemaking.
Summary of June 1, 2017, Public Meeting to Discuss the ML17173A081.
Scope and Related Costs and Benefits Associated with
the ``Reactor Vessel Materials Surveillance Program
Requirements'' Proposed Rulemaking.
Staff Requirements Memorandum (SRM)-COMSECY-18-0016, ML19009A517.
``Request Commission Approval to Use the Direct Final
Rule Process to Revise the Testing and Reporting
Requirements in 10 CFR Part 50, Appendix H, Reactor
Vessel Material Surveillance Program Requirements (RIN
3150-AK07)''.
Regulatory Analysis for the Direct Final Rule: Appendix ML20246G422.
H to 10 CFR Part 50--Reactor Vessel Material
Surveillance Program Requirements, September 2020.
----------------------------------------------------------------------------------------------------------------
[[Page 62207]]
List of Subjects in 10 CFR Part 50
Administrative practice and procedure, Antitrust, Backfitting,
Classified information, Criminal penalties, Education, Fire prevention,
Fire protection, Incorporation by reference, Intergovernmental
relations, Nuclear power plants and reactors, Penalties, Radiation
protection, Reactor siting criteria, Reporting and recordkeeping
requirements, Whistleblowing.
For the reasons set forth in the preamble, and under the authority
of the Atomic Energy Act of 1954, as amended; the Energy Reorganization
Act of 1974, as amended; and 5 U.S.C. 552 and 553, the NRC is adopting
the following amendments to 10 CFR part 50:
PART 50--DOMESTIC LICENSING OF PRODUCTION AND UTILIZATION
FACILITIES
0
1. 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; Sec. 109, Pub. L. 96-295, 94 Stat. 783.
0
2. In appendix H to part 50:
0
a. Revise paragraph III.B.1;
0
b. Add paragraph III.B.4; and
0
c. In paragraph IV.A, remove the phrase ``one year'' and add in its
place the phrase ``eighteen months''.
The revision and addition read as follows:
Appendix H to Part 50--Reactor Vessel Material Surveillance Program
Requirements
* * * * *
III. * * *
B. * * *
1. The design of the surveillance program and the withdrawal
schedule must meet the requirements of the edition of the ASTM E 185
that is current on the issue date of the ASME code to which the
reactor vessel was purchased; for reactor vessels purchased after
1982, the design of the surveillance program and the withdrawal
schedule must meet the requirements of ASTM E 185-82. For reactor
vessels purchased in or before 1982, later editions of ASTM E 185
may be used, but including only those editions through 1982. For
each capsule withdrawal, the test procedures and reporting
requirements must meet the requirements of the ASTM E 185 to the
extent practicable for the configuration of the specimens in the
capsule. If any of the optional provisions in paragraphs III.B.4(a)
through (d) of this section are implemented in lieu of ASTM E 185,
the number of specimens included or tested in the surveillance
program shall be adjusted as specified in paragraphs III.B.4(a)
through (d) of this section.
* * * * *
4. Optional provisions. As used in this section, references to
ASTM E 185 include the edition of ASTM E 185 that is current on the
issue date of the ASME Code to which the reactor vessel was
purchased through the 1982 edition.
(a) First Provision: Heat-Affected Zone Specimens--The inclusion
or testing of weld heat-affected zone Charpy impact specimens within
the surveillance program as specified in ASTM E 185 is optional.
(b) Second Provision: Tension Specimens--If this provision is
implemented, the minimum number of tension specimens to be included
and tested in the surveillance program shall be as specified in
paragraphs III.B.4(b)(i) and (ii) of this section.
(i) Unirradiated Tension Specimens--Two tension specimens from
each base and weld material required by ASTM E 185 shall be tested,
with one specimen tested at room temperature and the other specimen
tested at the service temperature; and
(ii) Irradiated Tension Specimens--Two tension specimens from
each base and weld material required by ASTM E 185 shall be included
in each surveillance capsule and tested, with one specimen tested at
room temperature and the other specimen tested at the service
temperature.
(c) Third Provision: Correlation Monitor Materials--The testing
of correlation monitor material specimens within the surveillance
program as specified in ASTM E 185 is optional.
(d) Fourth Provision: Thermal Monitor--The inclusion or
examination of thermal monitors within the surveillance program as
specified in ASTM E 185 is optional.
* * * * *
Dated at Rockville, Maryland, this 24th day of September, 2020.
For the Nuclear Regulatory Commission.
Annette Vietti-Cook,
Secretary for the Commission.
[FR Doc. 2020-21505 Filed 10-1-20; 8:45 am]
BILLING CODE 7590-01-P