[Federal Register Volume 90, Number 139 (Wednesday, July 23, 2025)]
[Proposed Rules]
[Pages 34609-34612]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2025-13817]
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Proposed Rules
Federal Register
________________________________________________________________________
This section of the FEDERAL REGISTER contains notices to the public of
the proposed issuance of rules and regulations. The purpose of these
notices is to give interested persons an opportunity to participate in
the rule making prior to the adoption of the final rules.
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Federal Register / Vol. 90, No. 139 / Wednesday, July 23, 2025 /
Proposed Rules
[[Page 34609]]
NUCLEAR REGULATORY COMMISSION
10 CFR Part 50
[Docket No. PRM-50-120; NRC-2019-0180]
Alternative Method for Calculating Embrittlement for Steel
Reactor Vessels
AGENCY: Nuclear Regulatory Commission.
ACTION: Petition for rulemaking; denial.
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SUMMARY: The U.S. Nuclear Regulatory Commission (NRC) is denying a
petition for rulemaking, dated August 19, 2019, submitted by Thomas A.
Bergman on behalf of NuScale Power, LLC. The petition was docketed by
the NRC on September 11, 2019, and was assigned Docket No. PRM-50-120.
The petitioner requested that the NRC revise its regulations to add an
alternative formula for calculating the mean value of the transition
temperature shift described in American Society for Testing and
Materials Standard E900-15 to the NRC's regulations and guidance
documents. The NRC is denying the petition because the petitioner did
not demonstrate the immediacy of any safety issues in the concerns
raised in the petition and did not provide any new information that
would warrant revision of the NRC's regulations.
DATES: The docket for the petition for rulemaking PRM-50-120 is closed
on July 23, 2025.
ADDRESSES: Please refer to Docket ID NRC-2019-0180 when contacting the
NRC about the availability of information for this action. You may
obtain publicly-available information related to this action by any of
the following methods:
Federal Rulemaking Website: Go to https://www.regulations.gov and search for Docket ID NRC-2019-0180. Address
questions about NRC dockets to Helen Chang; telephone: 301-415-3228;
email: [email protected]. For technical questions, contact the
individuals listed in the FOR FURTHER INFORMATION CONTACT section of
this document.
NRC's Agencywide Documents Access and Management System
(ADAMS): You may obtain publicly-available documents online in the
ADAMS Public Documents collection at https://www.nrc.gov/reading-rm/adams.html. To begin the search, select ``Begin Web-based ADAMS
Search.'' For problems with ADAMS, please contact the NRC's Public
Document Room (PDR) reference staff at 1-800-397-4209, at 301-415-4737,
or by email to [email protected]. For the convenience of the reader,
instructions about obtaining materials referenced in this document are
provided in the ``Availability of Documents'' section.
NRC's PDR: The PDR, where you may examine and order copies
of publicly available documents, is open by appointment. To make an
appointment to visit the PDR, please send an email to
[email protected] or call 1-800-397-4209 or 301-415-4737, between 8
a.m. and 4 p.m. eastern time, Monday through Friday, except Federal
holidays.
FOR FURTHER INFORMATION CONTACT: Aaron Kwok, Office of Nuclear Material
Safety and Safeguards, telephone: 301-415-1371, email:
[email protected]; or Dan Widrevitz, Office of Nuclear Reactor
Regulation, telephone: 301-415-2620, email: [email protected]. Both
are staff of the U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001.
SUPPLEMENTARY INFORMATION:
Table of Contents
I. The Petition
II. Public Comments on the Petition
III. Reasons for Denial
IV. Availability of Documents
V. Conclusion
I. The Petition
Section 2.802 of title 10 of the Code of Federal Regulations (10
CFR), ``Petition for rulemaking--requirements for filing,'' provides an
opportunity for any interested person to petition the Commission to
issue, amend, or rescind any regulation. On August 19, 2019, the NRC
received a petition for rulemaking (PRM) from Thomas A. Bergman on
behalf of NuScale Power, LLC (NuScale). The petitioner requested that
the NRC revise its regulations to add an alternative formula for
calculating the mean value of the transition temperature shift
described in American Society for Testing and Materials (ASTM) E900-15,
``Standard Guide for Predicting Radiation-Induced Transition
Temperature Shift in Reactor Vessel Materials.''
On November 19, 2019 (84 FR 63819), the NRC published a notice of
docketing and request for comment for PRM-50-120. The petitioner
requested that the NRC amend its regulations in Sec. 50.61(c)(1)(iv),
with the first paragraph to read as follows: ``[Delta]RTNDT
is the mean value of the transition temperature shift, or change in
RTNDT, due to irradiation, and must be calculated using
Equation 3. As an alternative, [Delta]RTNDT may be
determined in accordance with ASTM E900-15 instead of Equation 3, and
Tables 1 and 2 of this section.'' Further, the petitioner requested
that the formula for calculating the mean value of the transition
temperature shift described in ASTM E900-15 be added for use as an
alternative to Equation 2 in Regulatory Guide (RG) 1.99, Revision 2,
``Radiation Embrittlement of Reactor Vessel Materials.'' The petitioner
requested that the following text be added to paragraph 2 in Section
1.3 of RG 1.99, to read as follows: ``For new plants electing to use
ASTM E900-15 as allowed by Regulatory Position 3 for determining
[Delta]RTNDT, the correction factor is not required,
provided that the irradiation temperature is within the ASTM E900-15
applicability range.''
The NRC identified the following five main issues raised in the
petition:
Issue 1: The methodology for calculating the mean value of the
transition temperature shift ([Delta]RTNDT) in Sec. 50.61,
``Fracture toughness requirements for protection against pressurized
thermal shock events,'' and RG 1.99 is overly conservative and is based
on outdated information.
Issue 2: The 1 [deg]F/1 [deg]F adjustment methodology requires
excessive compensation for irradiation temperatures less than 525
[deg]F and has significant drawbacks.
Issue 3: The staff required NuScale to comply with Sec. 50.61 and
RG 1.99 and use the 1 [deg]F/1 [deg]F adjustment methodology.
Issue 4: ASTM E900-15 more accurately models the effects of
irradiation temperature and does not suffer the drawbacks of the 1
[deg]F/1 [deg]F adjustment methodology.
Issue 5: The current methodology for determining embrittlement in
Sec. 50.61,
[[Page 34610]]
with 1 [deg]F/1 [deg]F adjustment, is unnecessarily burdensome for
reactors like NuScale, in that it would: (a) result in unnecessarily
restrictive heat-up and cool-down rates during startups and shutdowns,
and (b) cause surveillance capsules to be withdrawn and tested
prematurely.
II. Public Comments on the Petition
The notice of docketing for PRM-50-120 requested interested persons
to submit comments. The comment period closed on December 19, 2019. The
NRC received 6 comment submissions consisting of 38 comments. The
comments were received from private citizens, individuals affiliated
with advocacy groups, and an individual affiliated with an industry
group. The comments received on PRM-50-120 and the NRC's responses to
them are available in ADAMS under Accession No. ML20304A003.
III. Reasons for Denial
The NRC is denying the petition because the petitioner did not
demonstrate the immediacy of any safety issues in the concerns raised
in the petition and did not provide any new information that would
warrant revision of the NRC's regulations.
The NRC concludes that Issue 1 does not warrant rulemaking because
the petitioner did not provide any new information that would warrant
the expenditure of limited NRC resources for rulemaking. Specifically,
the NRC found that while a significantly larger body of data for
neutron embrittlement is now available, the core assertion that RG
1.99, Revision 2, with the use of the 1 [deg]F/1 [deg]F adjustment
methodology, provides an overly conservative prediction is not correct
in cases the NRC has evaluated such as the NuScale design certification
application (DCA). The petition presents no additional information or
data to demonstrate that the current regulation is overly conservative.
The NRC concludes that Issue 2 does not warrant rulemaking because
the petitioner did not provide any new information beyond what is
approved in the NRC's final safety evaluation for the NuScale DCA. The
steels proposed to be used in the NuScale DCA, as well as those
proposed in other light-water designs known to the NRC, are represented
in the operating fleet. The petition did not present any pertinent new
information regarding embrittlement performance characteristics of
these materials. The NRC determined that the NuScale design presented
no unusual characteristics justifying a unique temperature-
embrittlement relationship for that design. In addition, RG 1.99,
Revision 2, does not prescribe a temperature adjustment; rather, it
states that any correction factor for operating conditions below 525
[deg]F should be ``justified by reference to actual data.''
Embrittlement was previously evaluated by the staff for the
specific case of a NuScale design, whose operating conditions include a
relatively low operating temperature (the embrittlement impacts of
which the 1 [deg]F/1 [deg]F adjustment compensates), for 40 years of
operation. The NRC verified, during its review of the NuScale DCA, that
a combination of the methodology in 10 CFR 50.61 and RG 1.99, Revision
2, together with the 1 [deg]F/1 [deg]F adjustment provides an
appropriate estimate of RTNDT based on a comparison to the
publicly available information. While the NRC found that the ASTM E900-
15 methodology may support improved accuracy at intermediate fluences,
these were not proposed in the NuScale DCA, nor in the petition, and
are bounded by the information presented in the NuScale DCA.
The NRC concludes that Issue 3 does not warrant rulemaking because
the staff did not require NuScale to comply with Sec. 50.61 and RG
1.99, Revision 2, and use the 1 [deg]F/1 [deg]F adjustment methodology.
In Section IV of the petition the petitioner states, ``The NuScale
application of RG 1.99, Rev 2 ETC, plus the 1[deg]F/1[deg]F adjustment
methodology deman[d]ed by the staff, requires an excessive compensation
for irradiation temperature less than 525[deg]F.'' In its design
certification application, NuScale proposed but declined to support its
initial proposal to use alternate methods for calculating
RTNDT. NuScale did not provide any new information beyond
what is described in the NuScale DCA in the petition. Furthermore, the
use of 1 [deg]F/1 [deg]F adjustment methodology is not required;
rather, it is a methodology that the NRC has previously accepted for
specific applications. Consequently, NuScale could have proposed an
alternate adjustment methodology for the temperature correlation.
The NRC concludes that Issue 4 does not warrant rulemaking because
ASTM E900-15 cannot be directly substituted for the methodologies
described in Sec. 50.61 and RG 1.99, Revision 2, as proposed by the
petitioner. This is because the ASTM E900-15 embrittlement trend curve
(ETC) is an embrittlement correlation; however, it lacks other
pertinent features of RG 1.99, Revision 2, such as a methodology for
utilizing plant-specific surveillance data to check prediction results.
In addition, the paucity of data at NuScale's planned operating
temperature within the dataset used to generate ASTM E900-15 would
require further considerations prior to use. Furthermore, although
NuScale asserts in its petition that ASTM E900-15 could also be used by
advanced reactors and other small modular reactors, the ASTM E900-15
ETC is based mainly on data from light-water reactors, and its
applicability is limited to the temperature range of the data used to
develop the embrittlement trend curve. NuScale is the only light-water
reactor design that has ever been reviewed by the NRC that would
operate with such a low operating temperature, and the other advanced
reactor designs the NRC is aware of would operate at substantially
higher temperatures than are addressed by the current data, and
therefore the NRC finds that ASTM E900-15 would not be useable for such
high temperature reactors without additional adjustments. Therefore,
the NRC finds that the petitioner's claim that ASTM E900-15 would
provide wide-ranging benefits for future advanced reactor designs is
not supported.
Additionally, the NRC determined that this issue does not warrant
rulemaking because the NRC has evaluated the acceptability of using
ASTM E900-15 for calculating reactor pressure vessel (RPV)
embrittlement trends. The NRC provided details of this effort at a May
19, 2020, public meeting to discuss RG 1.99, Revision 2, and appendix H
to 10 CFR part 50. During the Materials Information Exchange public
meeting on July 14, 2020, the NRC gave a status update indicating that
it had decided not to pursue an alternative to RG 1.99, Revision 2, at
this time. As part of the status update, the NRC noted that it planned
to document the results of its evaluation effort in two technical
letter reports, and that it also would complete a holistic evaluation
of RPV integrity, considering both the RG evaluation and RPV
surveillance programs, using the principles of risk-informed
decisionmaking from RG 1.174, Revision 3, ``An Approach for Using
Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-
Specific Changes to the Licensing Basis.'' The NRC also stated it would
continue to engage with stakeholders on this topic. The NRC indicated
that the documentation generated under this effort could be used by
future licensees or applicants seeking an alternative to RG 1.99,
Revision 2, based on the ASTM E900-15 ETC.
[[Page 34611]]
On October 26, 2020, the NRC issued the first technical letter
report TLR RES/DE/CIB 2020 09, ``RG 1.99 Revision 2 Update FAVOR
Scoping Study.'' In this report, the staff estimated the probability of
potential reactor vessel cracking under a variety of plant operating
transients relative to the degree of embrittlement underprediction
(i.e., how much may risk increase if embrittlement was underpredicted).
Estimates of embrittlement under RG 1.99, Revision 2 and the ASTM E900-
15 were then generated for operating plant materials. This allowed for
a comparison of the ``risk'' of using the older RG 1.99, Revision 2,
correlation versus the ASTM E900-15 correlation. The technical letter
report concluded that the risk associated with not updating the ETC of
RG 1.99, Revision 2, is relatively low. Given the low risk, the NRC
determined that there would be little benefit to updating RG 1.99,
Revision 2. The NRC evaluated this conclusion based on the information
included in the petition as well as the preliminary findings of the
evaluation process described above.
On January 19, 2021, the NRC staff issued the second technical
letter report, TLR-RES/DE/CIB-2020-11, ``Basis for a Potential
Alternative to Revision 2 of Regulatory Guide 1.99.'' The report
concluded that ASTM E900-15 is the best available alternative ETC to
the RG 1.99, Revision 2 ETC, providing more accurate predictions when
evaluated against the existing surveillance data. However, ASTM E900-15
cannot directly substitute for the methodologies described in 10 CFR
50.61 and RG 1.99, Revision 2, because the ASTM E900-15 ETC is an
embrittlement correlation and lacks other pertinent features such as a
methodology for using plant specific surveillance data to check
prediction results. More specifically, the scarcity of data at
NuScale's operating temperature within the BASELINE dataset used to
generate ASTM E900-15 would require further considerations for use.
NuScale is the only light-water reactor design reviewed by the NRC that
would operate with such a low temperature, and other advanced reactor
designs that the NRC is currently aware of would operate at
substantially higher temperatures than are addressed by the current
data and therefore the NRC finds that ASTM E900-15 would not be useable
for such high temperature reactors without additional work. Therefore,
the NRC finds that the petitioner's claim that ASTM E900-15 would
provide wide-ranging benefits for future advanced reactor designs is
not supported.
The NRC concludes that Issue 5 does not warrant rulemaking because
the petition did not establish the merits of its assertions regarding
unnecessary burden being imposed by the use of the RG 1.99, Revision 2,
methodology for determining the heat-up and cool-down rates during
startups and shutdowns. Consistent with the discussion concerning Issue
1, the NRC staff reviewed a forecasting of embrittlement for the
NuScale DCA and found the application of the current approach to be
acceptable and appropriate. With regards to the impact on heat-up/cool-
down curves, the staff did not have a basis to conclude that these
curves would have affected actual plant operation in a manner causing
significant unnecessary burden. Likewise, the petitioner did not
demonstrate the merits of the concern related to the withdrawal
schedules for surveillance capsules. The specific timing of removal
does not alter the associated burden of a removal and is not subject to
specific regulatory requirements.
IV. Availability of Documents
The documents identified in the following table are available to
interested persons through one or more of the following methods, as
indicated.
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ADAMS accession No./
Document Federal Register
citation/website
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NuScale, LLC Petition for Rulemaking to ML19254B848.
Revise 10 CFR Part 50--Alternative Method
for Calculating Embrittlement for Steel
Reactor Vessels, August 19, 2019.
Alternative Method for Calculating 84 FR 63819.
Embrittlement for Steel Reactor Vessels,
November 19, 2019.
NRC Response to Public Comments for PRM-50- ML20304A003.
120, October 14, 2021.
Regulatory Guide 1.174, Revision 3, ``An ML17317A256.
Approach for Using Probabilistic Risk
Assessment in Risk-Informed Decisions on
Plant-Specific Changes to the Licensing
Basis,'' January 2018.
Regulatory Guide 1.99, Revision 2, ML003740284.
``Radiation Embrittlement of Reactor Vessel
Materials,'' May 1988.
American Society for Testing and Materials, https://doi.org/10.1520/
``Standard Guide for Predicting Radiation- E0900-15E02 https://
Induced Transition Temperature Shift in www.astm.org/Standards/
Reactor Vessel Materials,'' ASTM E 900-15e2, E900.htm.
West Conshohocken, PA; ASTM International,
2015.
RG 1.99, Revision 2, and Reactor Vessel ML20168A008 (Package).
Surveillance Public Meeting, May 19, 2020.
NuScale Standard Plant Design Certification ML20224A493.
Application, Chapter 5, ``Reactor Coolant
System and Connecting Systems,'' July 2020.
American Society for Testing and Materials, https://compass.astm.org/
``Standard Practice for Conducting EDIT/
Surveillance Tests for Light-Water Cooled html_historical.cgi?E185
Nuclear Power Reactor Vessels,'' ASTM E185- +02.
82e2, E 706 (IF). West Conshohocken, PA;
ASTM International, 0 (July 1, 1982).
PHASE 6--NuScale DC Final Safety Evaluation ML20023A318 (Package).
Report (Complete with Appendices).
RG 1.99 Revision Evaluation Effort for ML20192A002.
Industry/U.S. Nuclear Regulatory Commission
Materials Programs Technical Information
Exchange Public Meeting, July 14, 2020.
RG 1.99 Revision 2 Update FAVOR Scoping ML20300A551.
Study, October 26, 2020.
TLR-RES/DE/CIB-2020-11, ``Basis for a ML20345A003.
Potential Alternative to Revision 2 of
Regulatory Guide 1.99,'' January 19, 2021.
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V. Conclusion
For the reasons cited in this document, the NRC is denying PRM-50-
120. The NRC completed an evaluation of the petition and determined
that the issues in the petition did not demonstrate the immediacy of
any safety issues and did not provide any new information that would
warrant revision of the NRC's regulations. The NRC concludes that the
arguments presented in the petition do not support the requested
revisions to its regulations. Finally, the NRC reaffirms that its
existing regulations continue to provide reasonable assurance of
adequate protection of public health and safety.
[[Page 34612]]
Dated: July 21, 2025.
For the Nuclear Regulatory Commission.
Carrie Safford,
Secretary of the Commission.
[FR Doc. 2025-13817 Filed 7-22-25; 8:45 am]
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