[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direct final rule is withdrawn as a result of such comments, timely 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