[Federal Register Volume 86, Number 4 (Thursday, January 7, 2021)]
[Proposed Rules]
[Pages 1022-1030]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2020-29151]
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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. 86, No. 4 / Thursday, January 7, 2021 /
Proposed Rules
[[Page 1022]]
NUCLEAR REGULATORY COMMISSION
10 CFR Part 50
[Docket Nos. PRM-50-93 and PRM-50-95; NRC-2009-0554]
Calculated Maximum Fuel Element Cladding Temperature
AGENCY: Nuclear Regulatory Commission.
ACTION: Petitions for rulemaking; denial.
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SUMMARY: The U.S. Nuclear Regulatory Commission (NRC) is denying two
related petitions for rulemaking (PRMs), PRM-50-93 and PRM-50-95,
submitted by Mark Edward Leyse. The petitioner requested that the NRC
amend its regulations for the domestic licensing of production and
utilization facilities. The petitioner asserted that data from multirod
(assembly) severe fuel damage experiments indicate that specific
aspects of the NRC's regulations on emergency core cooling systems
acceptance criteria and evaluation models are not conservative and that
additional regulations are necessary. The NRC is denying these
petitions because existing NRC regulations provide reasonable assurance
of adequate protection of public health and safety. The petitioner did
not present sufficient new information or arguments to support the
requested changes.
DATES: The dockets for the petitions for rulemaking, PRM-50-93 and PRM-
50-95, are closed on January 7, 2021.
ADDRESSES: Please refer to Docket ID NRC-2009-0554 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-2009-0554. Address
questions about NRC dockets to Dawn Forder; telephone: 301-415-3407;
email: [email protected]. For technical questions, contact the
individual 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, 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 Section IV, ``Availability
of Documents.''
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: Daniel Doyle, Office of Nuclear
Material Safety and Safeguards, telephone: 301-415-3748, email:
[email protected], U.S. Nuclear Regulatory Commission, Washington,
DC 20555-0001.
SUPPLEMENTARY INFORMATION:
I. Background and Summary of the Petitions
II. Public Comments on the Petitions
III. NRC Technical Evaluation and Reasons for Denial
IV. Availability of Documents
V. Conclusion
I. Background and Summary of the Petitions
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 November 17, 2009, Mark
Edward Leyse submitted a PRM under Sec. 2.802. The NRC assigned docket
number PRM-50-93 to this petition and published a notice of receipt and
request for public comment in the Federal Register on January 25, 2010
(75 FR 3876).
The petitioner asserted that data from multirod (assembly) severe
fuel damage experiments indicate that specific aspects of the NRC's
regulations and associated regulatory guidance on Emergency Core
Cooling Systems (ECCS) acceptance criteria and evaluation models are
not conservative and that additional regulations are necessary.
Therefore, the petitioner requested that the NRC: (1) Amend its
regulations to require that the calculated maximum fuel element
cladding temperature not exceed a limit based on data from cited
experiments; (2) amend its regulations and associated regulatory
guidance to require that the rates of energy release, hydrogen
generation, and Zircaloy cladding oxidation from the metal-water
reaction of zirconium with steam considered in the evaluation models
used to calculate ECCS cooling performance be based on data from cited
experiments; and (3) issue a new regulation that requires minimum
allowable core reflood rates in the event of a loss-of-coolant accident
(LOCA).
On June 7, 2010, Mark Edward Leyse, on behalf of the New England
Coalition, submitted a petition for enforcement action under Sec.
2.206, ``Requests for action under this subpart.'' The petitioner
requested that the NRC order the Vermont Yankee Nuclear Power Station
to lower its licensing basis peak cladding temperature to provide an
adequate margin of safety in the event of a LOCA. The NRC staff
concluded that this petition did not meet the criteria for review under
Sec. 2.206 because it identified generic issues that could require
revisions to existing NRC regulations. Therefore, the NRC decided to
review it as a PRM under Sec. 2.802 and assigned it docket number PRM-
50-95. Because PRM-50-93 and PRM-50-95 address similar issues, the NRC
staff consolidated its review into a single activity. On October 27,
2010, the NRC published a notice of consolidation of PRM-50-93 and PRM-
50-95 in the Federal Register (75 FR 66007) and requested public
comment.
The NRC identified three main issues in the two petitions. The
remaining paragraphs of Section I summarize the following information
for each main issue: (1) Relevant background information; (2) arguments
in the petitions; and (3) specific requests the petitioner made to
address each issue.
[[Page 1023]]
Issue 1: Calculated Maximum Fuel Element Cladding Temperature Limit
Background for Issue 1
Under Sec. 50.46, ``Acceptance criteria for emergency core cooling
systems for light-water nuclear power reactors,'' of 10 CFR, light-
water nuclear power reactors fueled with uranium oxide pellets within
cylindrical Zircaloy cladding must be provided with an ECCS that must
be designed so that its calculated cooling performance following
postulated loss of coolant accidents (LOCAs) \1\ conforms to the
criteria specified in Sec. 50.46(b).\2\ Under Sec. 50.46(b)(1), the
calculated maximum fuel element cladding temperature shall not exceed
2,200 [deg]F. In addition, Sec. 50.46(b)(2) through (5), respectively,
contain requirements for calculations involving: Maximum cladding
oxidation, maximum hydrogen generation, changes in core geometry, and
long-term cooling.
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\1\ Under Sec. 50.46(c), LOCAs are hypothetical accidents that
would result from the loss of reactor coolant, at a rate that
exceeds the capability of the reactor coolant makeup system, from
breaks in pipes in the reactor coolant pressure boundary.
\2\ Criterion 35 of appendix A to 10 CFR part 50, ``General
Design Criteria for Nuclear Power Plants,'' further requires that a
system to provide abundant emergency core cooling shall be provided
and that the system safety function shall be to transfer heat from
the reactor core following any loss of reactor coolant at a rate
such that: (1) Fuel and cladding damage that could interfere with
continued effective core cooling is prevented and (2) the cladding
metal-water reaction is limited to negligible amounts.
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Petitioner's Arguments and Requests Related to Issue 1
The petitioner asserted that data from multirod (assembly) severe
fuel damage experiments indicate that the calculated maximum fuel
element cladding temperature limit of 2,200 [deg]F specified in Sec.
50.46(b)(1) is not conservative. Although not its intended purpose, the
NRC previously determined that this limit provides a conservative
safety margin from an area of Zircaloy cladding oxidation behavior
known as the autocatalytic regime. An autocatalytic condition occurs
when the heat released by the metal-water reaction of zirconium with
steam is greater than the heat that can be transferred away from the
Zircaloy cladding. This causes the Zircaloy cladding temperature to
rise, thereby increasing the diffusion of oxygen into the metal, which
in turn raises the rate at which the zirconium-steam oxidation reaction
occurs. As the metal-water reaction rate continues to increase, the
temperature of the Zircaloy cladding continues to rise, eventually
resulting in an uncontrolled reaction and temperature excursion. The
petitioner asserted that data from cited experiments indicate that such
autocatalytic metal-water oxidation reactions and uncontrolled
temperature excursions involving Zircaloy cladding have occurred at
temperatures below 2,200 [deg]F. The petitioner provided this assertion
as evidence that the 2,200 [deg]F limit is not conservative, and
requested that the NRC amend Sec. 50.46 to require that the calculated
maximum fuel element cladding temperature not exceed a limit based on
data from cited experiments, instead of the 2,200 [deg]F limit
specified in Sec. 50.46(b)(1).
Issue 2: Metal-Water Reaction Rate Equations for ECCS Evaluation Models
Background for Issue 2
To evaluate conformance with the criteria specified in Sec.
50.46(b), ECCS cooling performance must be calculated using an
acceptable evaluation model \3\ for a range of postulated LOCAs of
different sizes, locations, and other properties sufficient to provide
assurance that the most severe postulated LOCAs are evaluated. On
September 16, 1988, the NRC amended the requirements of Sec. 50.46 and
appendix K, ``ECCS Evaluation Models,'' to 10 CFR part 50 to reflect an
improved understanding of ECCS performance during reactor transients
that was obtained through extensive research performed after
promulgation of the original requirements (53 FR 35996). Under Sec.
50.46(a)(1), licensees or applicants may use one of two acceptable ECCS
evaluation model options: (1) A best-estimate or realistic evaluation
model \4\ or (2) a conservative evaluation model. Each ECCS evaluation
model option is summarized below.
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\3\ Regulatory Guide (RG) 1.157, ``Best-Estimate Calculations of
Emergency Core Cooling System Performance,'' issued May 1989, states
that ``the term `evaluation model' refers to a nuclear plant system
computer code or any other analysis tool designed to predict the
aggregate behavior of a reactor during a loss of coolant accident.
It can be either best-estimate or conservative and may contain many
correlations or models.''
\4\ RG 1.157 states that ``the terms `best-estimate' and
`realistic' have the same meaning. Both terms are used to indicate
that the techniques attempt to predict realistic reactor system
thermal-hydraulic response.''
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Option 1: Best-Estimate or Realistic ECCS Evaluation Model
Section 50.46(a)(1)(i) of 10 CFR specifies that a best-estimate
evaluation model must include sufficient supporting justification to
show that the analytical technique realistically describes the behavior
of the reactor system during a LOCA. Comparisons to applicable
experimental data must be made and uncertainties must be identified and
assessed so that the uncertainty in the calculated results can be
estimated to (1) account for the uncertainty in comparing the
calculated ECCS cooling performance to the criteria specified in Sec.
50.46(b); and (2) assure that there is a high probability of not
exceeding these criteria.
RG 1.157 describes models,\5\ correlations,\6\ data, model
evaluation procedures, and methods that are acceptable to the NRC staff
for meeting the requirements for: (1) A realistic or best-estimate
calculation of ECCS cooling performance during a LOCA; (2) estimating
the uncertainty in that calculation; and (3) including uncertainty in
the comparisons of the calculated results to the criteria of Sec.
50.46(b) to assure a high probability that the criteria would not be
exceeded. Other models, data, model evaluation procedures, and methods
can be considered if they are supported by appropriate experimental
data and technical justification.
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\5\ RG 1.157 states that ``the term `model' refers to a set of
equations derived from fundamental physical laws that is designed to
predict the details of a specific phenomenon.''
\6\ RG 1.157 states that ``the term `correlation' refers to an
equation having empirically determined constants such that it can
predict some details of a specific phenomenon for a limited range of
conditions.''
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To be considered acceptable under RG 1.157, evaluation models
should account for identified sources of heat--including the metal-
water reaction rate--in performing best-estimate calculations. In
particular, the rates of energy release, hydrogen generation, and
Zircaloy cladding oxidation from the metal-water reaction of zirconium
with steam should be calculated in a best-estimate manner using one of
two procedures, depending on the cladding temperature:
(1) If the cladding temperature is less than or equal to 1,900
[deg]F, correlations to be used to calculate metal-water reaction rates
should: (a) Be checked against a set of relevant data and (b) recognize
the effects of steam pressure, pre-oxidation of the cladding,
deformation during oxidation, and internal oxidation from both steam
and uranium oxide fuel.
(2) If the cladding temperature is greater than 1,900 [deg]F, the
Cathcart-Pawel equation and the underlying empirical data used to
derive it are considered acceptable for calculating the rates of energy
release, hydrogen generation, and cladding oxidation.
[[Page 1024]]
Option 2: Conservative ECCS Evaluation Model
Alternatively, a conservative evaluation model may be developed in
conformance with the required and acceptable features of appendix K,
``ECCS Evaluation Models,'' to 10 CFR part 50. Under appendix K,
section I.A., evaluation models must account for various sources of
heat during LOCA conditions including the metal-water reaction rate. In
particular, section I.A.5, ``Metal-Water Reaction Rate,'' of appendix K
requires use of the Baker-Just equation to calculate the rates of
energy release, hydrogen generation, and Zircaloy cladding oxidation
from the metal-water reaction of zirconium with steam, assuming that
the reaction is not steam limited.
Petitioner's Arguments and Requests Related to Issue 2
The petitioner argued that data from multirod (assembly) severe
fuel damage experiments indicate that the equations used to calculate
the metal-water reaction rate in ECCS evaluation models that the NRC
has determined to be acceptable for use in evaluating ECCS cooling
performance are not conservative. In particular, the petitioner
asserted that data from cited experiments indicate that use of the
Cathcart-Pawel equation in realistic evaluation models or use of the
Baker-Just equation in conservative evaluation models would: (1)
Overestimate the temperature at which autocatalytic metal-water
oxidation reactions would occur during a LOCA; and (2) underestimate
the rate of Zircaloy cladding oxidation from the metal-water reaction
of zirconium with steam and, therefore, underestimate the heatup,
heatup rate, and maximum temperature of the Zircaloy cladding during a
LOCA. Therefore, the petitioner requested that the NRC amend RG 1.157
and appendix K to 10 CFR part 50 to require that the rates of energy
release, hydrogen generation, and Zircaloy cladding oxidation from the
metal-water reaction of zirconium with steam considered in evaluation
models used to calculate ECCS cooling performance be calculated based
on data from cited experiments, instead of using the Cathcart-Pawel or
Baker-Just equations.
Issue 3: Minimum Allowable Core Reflood Rate
Background for Issue 3
Section 50.46(b) of 10 CFR does not include criteria for calculated
ECCS cooling performance pertaining to the core reflood rate following
postulated LOCAs.
Petitioner's Arguments and Requests Related to Issue 3
The petitioner asserted that a constant core reflood rate of
approximately 1 inch per second or lower would not, with high
probability, prevent Zircaloy cladding from exceeding the 2,200 [deg]F
limit in Sec. 50.46(b)(1) if, at the onset of reflood, the cladding
temperature was greater than or equal to 1,200 [deg]F. In particular,
the petitioner asserted that: (1) Although reflood rates would vary
throughout the reactor core during a LOCA, local reflood rates could be
approximately 1 inch per second or lower; and (2) extrapolation of data
from the cited experiments indicates that a constant core reflood rate
of approximately 1 inch per second or lower would not, with high
probability, prevent Zircaloy cladding from exceeding the 2,200 [deg]F
limit, if the cladding temperature was greater than or equal to 1,200
[deg]F at the onset of reflood.\7\ Therefore, the petitioner requested
that the NRC issue a new regulation that would require minimum
allowable core reflood rates in the event of a LOCA.
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\7\ Extrapolation of the experimental data was necessary because
the referenced tests were started with relatively low initial
cladding temperatures. The petitioner hypothesized that, if these
tests had started with higher initial cladding temperatures,
autocatalytic oxidation and failure of the Zircaloy cladding would
have occurred with high probability.
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II. Public Comments on the Petitions
II.A. Overview of Public Comments
The NRC received a total of 33 comment submissions that
collectively included 125 individual comments. The NRC reviewed and
considered all 125 comments in its evaluation of the petitions. Table I
identifies the number of comment submissions and individual comments
submitted, grouped by three main categories of comments. These
categories are used only to facilitate presenting a high-level summary
and totals for the comments that different stakeholder groups
submitted; the NRC staff used the same approach for addressing all
submitted comments, regardless of category or who submitted them. The
paragraphs that follow provide a high-level overview of each category
of comments.
Table I--Number of Comment Submissions and Individual Comments by
Category
------------------------------------------------------------------------
Number of Number of
Category comment individual
submissions comments
------------------------------------------------------------------------
Comments from the Petitioner............ \a\ 13 \a\ 97
Comments from Nuclear Industry 3 9
Representatives........................
Comments from Public Interest Groups or 17 19
Other Interested Individuals...........
-------------------------------
Total............................... 33 125
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\a\ The petitioner provided nine comment submissions after the public
comment period that closed on November 26, 2010. Although not required
to do so, the NRC also considered all the comment submissions that
were submitted after the public comment period closed.
Category 1: Comments From the Petitioner
Petitioner Mark Edward Leyse provided 13 comment submissions in
support of PRM-50-93 and PRM-50-95. He provided nine of these comment
submissions after the comment period closed. The NRC considered all 13
comment submissions in its evaluation. In general, the petitioner's
comments further supported the petitions by either: (1) Repeating
information that had already been provided; (2) providing additional
details to clarify specific issues; or (3) citing other references that
the petitioner believed further substantiated the arguments in the
petitions. In some comments, the petitioner identified additional
technical issues that were relevant to the subject matter, but were not
directly related to the requested changes to the NRC's regulations. As
discussed in Section III, the NRC staff addressed these additional
technical issues in its final technical safety analysis report.
[[Page 1025]]
Category 2: Comments From Nuclear Industry Representatives
The Nuclear Energy Institute (NEI) provided two comment submissions
that oppose PRM-50-93 and PRM-50-95. Overall, NEI recommended that the
NRC deny PRM-50-93 and PRM-50-95 because the experiments identified in
the petitions--whether considered individually or in conjunction with
other experiments--do not substantiate the assertions or requests made
in the petitions. NEI further provided additional experimental evidence
that indicates the NRC's regulations and associated regulatory guidance
on ECCS acceptance criteria and evaluation models are adequate.
Exelon Corporation provided one comment submission that opposes
PRM-50-93 and PRM-50-95, stating that: (1) It did not consider the
proposed amendments to the NRC's regulations or associated regulatory
guidance to be necessary and (2) it agreed with the comments that NEI
submitted.
Category 3: Comments From Public Interest Groups or Other Interested
Individuals
Three public interest groups (Don't Waste Michigan, Beyond Nuclear,
and Union of Concerned Scientists (UCS)) each provided one comment
submission in support of PRM-50-93 and PRM-50-95. In general, these
comments provided high-level statements of support for the petitions
but did not cite relevant evidence to substantiate the petitions.
Other interested individuals provided a total of 10 comment
submissions on PRM-50-93 and PRM-50-95. In general, these individual
comments also provided high-level statements of support for the
petitions but did not cite relevant evidence to substantiate the
petitions. In addition, several comments identified unrelated concerns
about the NRC's regulations or practices that the NRC staff determined
to be outside the scope of PRM-50-93 and PRM-50-95.
Robert Leyse, a relative of petitioner Mark Edward Leyse, provided
four comment submissions in support of PRM-50-93 and PRM-50-95. Robert
Leyse had previously submitted a related petition for rulemaking (PRM-
50-76) that the NRC denied on September 6, 2005.\8\ In general, his
comments either repeated information provided in the petitions or
expressed his view that the NRC did not appropriately consider all
relevant information in its denial of PRM-50-76.
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\8\ Robert Leyse petitioned the NRC on May 1, 2002, requesting
the NRC to amend Appendix K of 10 CFR part 50 and RG 1.157 to
correct asserted technical deficiencies in the Baker-Just and
Cathcart-Pawel equations used to calculate the metal-water reaction
rate in ECCS evaluation models. The NRC denied PRM-50-76,
determining that: (1) None of the specific technical issues raised
by the petitioner showed safety-significant deficiencies in the
research, calculation methods, or data used to support ECCS cooling
performance evaluations; and (2) the NRC's regulations and
regulatory guidance on ECCS cooling performance evaluations were
based on sound science and did not need to be amended (70 FR 52893).
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II.B. NRC Response to Public Comments
Two main factors influenced the NRC's approach to developing and
documenting its response to public comments submitted on PRM-50-93 and
PRM-50-95: (1) The substantial number, length, and complexity of the
comments that were submitted; and (2) the limited availability of NRC
resources due to competing, higher-priority work. In this approach,
individual comments that addressed similar subject categories were
grouped into one of 16 high-level comment bins. The following
paragraphs provide for each bin of comments: (1) A high-level summary
of the main subject category addressed in the grouped comments,
including a listing in parentheses of the unique identifiers for
individual comments that were assigned to the bin; and (2) the NRC's
response to the grouped comments, including--if appropriate--a high-
level summary of the basis for the response and reference to the
relevant section(s) of the NRC's final technical safety analysis report
that provide(s) additional details to support the NRC's position. A
separate document consolidates all 33 comment submissions and 125
individual comments, and provides the following information: (1) A
table that lists the unique identifier and ADAMS accession number
assigned to each comment submission document and (2) markings that
clearly assign unique identifiers to portions of each comment
submission that were identified as distinct individual comments.
Information about how to access this consolidated document is provided
in Section IV.
1. General Support for Petitions Without Providing Rationale
Comment: The NRC should initiate rulemaking to address the issues
raised in the petitions. (5-1, 6-1, 7-1, 8-1, 9-1, 10-1, 11-1, 12-1,
15-1, 19-1, 23-1)
NRC response: Because these comments generally supported the
petitions without providing a rationale to substantiate this support,
the NRC's overall response to the petitions applies to this bin of
comments. The final technical safety analysis report provides
additional details to support the NRC staff's position.
2. General Opposition to Petitions Without Providing Rationale
Comment: The requested amendments to NRC's regulations are not
necessary. (18-1)
NRC response: Because this comment generally opposed the petitions
without providing a rationale to substantiate this opposition, the
NRC's overall response to the petitions applies to this bin of
comments. The final technical safety analysis report provides
additional details to support the NRC staff's position.
3. Comments Related to PRM-50-76
Comment: As stated in PRM-50-76, the Cathcart-Pawel and Baker-Just
equations are not conservative because they were not developed to
consider how complex thermal-hydraulic phenomena would affect the
metal-water reaction rate in the event of a LOCA. (2-1, 17-2)
NRC response: The NRC disagrees with these comments. Consistent
with the technical safety analysis that was performed for PRM-50-76,
the NRC staff determined that--for the development of metal-water
reaction rate equations--well-characterized isothermal tests are more
important than considering the effects of complex thermal-hydraulic
phenomena. The suggested use of complex thermal-hydraulic conditions
would be counterproductive in tests that experimentally derive reaction
rate correlations because temperature control is required to develop a
consistent set of data for correlation derivation. Isothermal tests
provide this needed temperature control. Section 1.1, ``Similar
Petition Previously Considered by NRC (ML041210109),'' of the final
technical safety analysis report provides additional details to support
the NRC staff's position.
4. Peak Cladding Temperature Limit Is Not Conservative
Comment: Data from cited experiments indicate that autocatalytic
metal-water oxidation reactions and uncontrolled temperature excursions
involving Zircaloy cladding have occurred at temperatures below 2,200
[deg]F, indicating the regulatory limit of 2,200 [deg]F is not
conservative. (2-6, 2-10, 3-1, 4-1, 14-5, 14-7, 14-11, 16-2, 16-4, 20-
1, 20-5, 20-6, 20-10, 20-14, 20-15, 21-4, 21-14, 23-2, 24-1, 25-1, 26-
11, 32-1, 32-7)
NRC response: The NRC disagrees with these comments. The NRC staff
reviewed experimental data and information from the cited experiments
and found no evidence of temperature
[[Page 1026]]
escalation rates that demonstrated the occurrence of autocatalytic or
runaway oxidation reactions below 2,200 [deg]F under LOCA conditions.
Section 2.1, ``Peak Cladding Temperature Limit is Nonconservative,'' of
the final technical safety analysis report provides additional details
to support the NRC staff's position.
5. Baker-Just and Cathcart-Pawel Equations Are Not Conservative
Comment: Data from cited experiments indicate that the Baker-Just
and Cathcart-Pawel equations used to calculate the metal-water reaction
rate in ECCS evaluation models that the NRC has determined to be
acceptable for use in evaluating ECCS cooling performance are not
conservative. (1-1, 2-5, 14-1, 14-8, 14-9, 14-10, 14-12, 14-13, 14-14,
16-1, 20-4, 20-7, 20-8, 20-9, 20-11, 20-12, 20-16, 20-17, 21-3, 21-10,
21-13, 24-2, 26-1, 27-1, 27-3, 28-2, 29-3, 29-5, 29-6, 30-1, 30-2, 32-
2, 32-9)
NRC response: The NRC agrees in part and disagrees in part with
these comments. The NRC agrees that the Cathcart-Pawel equation is
generally not conservative. However, consistent with its intended use,
the NRC staff has determined that use of the Cathcart-Pawel equation
generally results in sufficiently accurate calculations of the metal-
water reaction rate that are appropriate for realistic ECCS evaluation
models. The NRC disagrees that the Baker-Just equation is not
conservative. Consistent with its intended use, the NRC staff has
determined that use of the Baker-Just equation results in sufficiently
conservative calculations of the metal-water reaction rate that are
appropriate for conservative ECCS evaluation models. Section 2.2,
``Baker-Just and Cathcart-Pawel Equations are Nonconservative,'' of the
final technical safety analysis report provides additional details to
support the NRC staff's position.
6. Need for a Minimum Allowable Reflood Rate
Comment: Extrapolation of data from cited experiments indicates
that a new regulation that requires minimum allowable core reflood
rates in the event of a LOCA is necessary to prevent Zircaloy cladding
from exceeding the regulatory limit of 2,200 [deg]F under certain
conditions. (2-2, 2-3, 2-4, 16-3, 20-2, 20-3, 20-13, 20-18, 21-2, 24-3,
26-2, 26-7, 26-9, 32-6)
NRC response: The NRC disagrees with these comments. The NRC staff
has determined--using simulations of a Zircaloy cladding bundle with
the geometry and design that was used for the cited experiments--that
steam cooling would be sufficient to maintain Zircaloy cladding
temperatures below the 2,200 [deg]F limit. Section 2.3, ``Need for a
Minimum Allowable Reflood Rate,'' of the final technical safety
analysis report provides additional details to support the NRC staff's
position.
7. Issues Related to National Research Universal Full-Length High-
Temperature (FLHT) In-Reactor Tests
Comment: In the FLHT-1 test, the test conductors were unable to
prevent a temperature excursion and runaway oxidation by increasing the
coolant flow rate when peak cladding temperatures reached approximately
2,200 [deg]F. This provides additional evidence indicating that the
regulatory limit of 2,200 [deg]F is not conservative. (21-5, 26-4, 26-
8, 28-3, 29-1, 29-4)
NRC response: The NRC disagrees with these comments. The NRC staff
determined that excessive heatup rates were not experienced during the
FLHT-1 experiment until temperatures exceeded 2,420 [deg]F. Section
3.1, ``Issues Related to National Research Universal (NRU) full-length
high-temperature (FLHT) In-reactor Tests,'' of the final technical
safety analysis report provides additional details to support the NRC
staff's position.
8. Eutectic Behavior at Temperatures Below 2,200 [deg]F
Comment: In a design-basis LOCA, eutectic reactions \9\ between
various fuel assembly components (the Zircaloy cladding, control rods,
and spacer grids) at temperatures below 2,200 [deg]F could
significantly reduce the safety margins for the following types of
materials interactions: (1) Degradation of boiling-water reactor (BWR)
control blades due to the eutectic reaction of boron carbide (B4C),
stainless steel, and Zircaloy; (2) degradation of pressurized-water
reactor (PWR) cladding due to the eutectic reaction between Inconel
grids and Zircaloy cladding; and (3) degradation of PWR control rods
that contain silver, indium, and cadmium. (21-1, 21-6, 21-7, 21-8, 21-
9, 24-4, 26-10)
---------------------------------------------------------------------------
\9\ In this context, a eutectic reaction is a reaction in which
two materials in contact with one another at relatively high
temperatures can liquefy at a temperature that is lower than the
melting temperatures of the two individual materials.
---------------------------------------------------------------------------
NRC response: The NRC disagrees with these comments. These
assertions are not supported by available experimental evidence. In its
review of available information, the NRC staff was unable to find any
evidence that loss of a coolable geometry had occurred at temperatures
below 2,200 [deg]F. Test results and analyses have shown that
insignificant eutectic reactions occur for times and maximum
temperatures assumed in a design-basis LOCA. Section 3.2, ``Eutectic
Behavior at Temperatures below 2,200 [deg]F (1,204 [deg]C),'' of the
final technical safety analysis report provides additional details to
support the NRC staff's position.
9. TRAC/RELAP \10\ Advanced Computational Engine (TRACE) Code
Simulation of (Full Length Emergency Cooling Heat Transfer) FLECHT Run
9573
---------------------------------------------------------------------------
\10\ TRAC: Transient Reactor Analysis Code. RELAP: Reactor
Excursion and Leak Analysis Program.
---------------------------------------------------------------------------
Comment: NRC's TRACE simulations of FLECHT Run 9573 are invalid
because they did not simulate the section of the test bundle that
incurred runaway oxidation. Therefore, since NRC's conclusions
regarding the reflood rate are based on its TRACE simulations of FLECHT
Run 9573, these conclusions are also invalid. (31-4, 32-3, 32-5, 33-1)
NRC response: The NRC disagrees with these comments. The NRC staff
determined that the experimental data from FLECHT run 9573 do not show
evidence of runaway oxidation below 2,200 [deg]F, despite its low
reflood rate. In addition, FLECHT run 9573 was a low-reflood-rate
experiment in which thermocouple measurements were taken at five
elevations. All five elevations were included in the NRC's TRACE
simulation of FLECHT run 9573. Section 3.3, ``TRACE simulation of
FLECHT run 9573,'' of the final technical safety analysis report
provides additional details to support the NRC staff's position.
10. Stainless Steel and Zircaloy Heat Transfer Coefficients
Comment: The heat transfer coefficients used in appendix K ECCS
evaluation models are based on data from thermal-hydraulic experiments
conducted with stainless steel rod bundles and therefore should not be
used to infer what would happen in a reactor core with Zircaloy bundles
in the event of a LOCA. (2-9, 22-1, 26-3, 26-5, 26-6, 32-4)
NRC response: The NRC disagrees with these comments. The NRC staff
determined that models for convective heat transfer are dependent upon
the properties of the fluid--not the material properties of the heat
transfer surface. Therefore, the heater rod material used in the
experiments is irrelevant to developing correlations based on the
experimental data. Section 3.5,
[[Page 1027]]
``Stainless Steel and Zircaloy Heat Transfer Coefficients,'' of the
final technical safety analysis report provides additional details to
support the NRC staff's position.
11. Issues Related to the PHEBUS B9R Test
Comment: Oxidation models are unable to predict autocatalytic
oxidation reactions that occurred below 2,200 [deg]F in the PHEBUS B9R-
2 test. (32-8, 32-10)
NRC response: The NRC disagrees with these comments. The NRC staff
determined that data from the cited PHEBUS B9R test does not
demonstrate that an autocatalytic oxidation reaction occurred at
temperatures below 2,200 [deg]F. Section 3.6, ``Issues Related to the
PHEBUS B9R Test,'' of the final technical safety analysis report
provides additional details to support the NRC staff's position.
12. Whether Runaway Oxidation Begins at 2,012 [deg]F
Comment: Information in a report about degraded core quench
experiments \11\ indicates that temperatures at which temperature
excursions associated with runaway oxidation occur range from 1,922
[deg]F to 2,012 [deg]F. (2-7)
---------------------------------------------------------------------------
\11\ Committee on the Safety of Nuclear Installations, Nuclear
Energy Agency, Organisation for Economic Co-operation and
Development. Degraded Core Quench: Summary of Progress 1996-1999.
NEA/CSNI/R(99)23. Paris, France: Organisation for Economic Co-
operation and Development; 2000. Available at: http://www.oecd-nea.org/nsd/docs/1999/csni-r99-23.pdf.
---------------------------------------------------------------------------
NRC response: The NRC disagrees with this comment. The NRC staff
examined the cited report and found no data to support a determination
that runaway oxidation occurs at cladding temperatures less than 2,200
[deg]F for experiments simulating conditions for design-basis
accidents. Section 3.7, ``Issue Related to Whether Runaway Oxidation
Temperatures Start at 1100 [deg]C (2012 [deg]F),'' of the final
technical safety analysis report provides additional details to support
the NRC staff's position.
13. Experimental Methods Used To Derive the Baker-Just Metal-Water
Oxidation Reaction Correlation
Comment: The Baker-Just equation is not conservative because it is
partly derived using experimental data from inductive heating
experiments that included radiative heat losses. These radiative heat
losses would affect the oxidation behavior such that the experiment is
not representative of reactor behavior in the event of a LOCA and would
cause the Baker-Just equation to be not conservative. (13-1, 14-2, 14-
3, 14-4, 14-6, 17-1, 27-2)
NRC response: The NRC disagrees with these comments. The NRC staff
determined that the subject experimental data are consistent with data
obtained using other methods and concluded that radiative heat losses
are not relevant in correlating the data to develop the metal-water
reaction rate equation. The NRC staff further concluded that use of the
Baker-Just equation results in sufficiently conservative calculations
of the metal-water reaction rate that are appropriate for conservative
ECCS evaluation models. Section 3.9, ``Experimental Methods Used to
Derive the Baker-Just Metal-Water Oxidation Reaction Correlation,'' of
the final technical safety analysis report provides additional details
to support the NRC staff's position.
14. Issues Related To Cladding Oxidation and Hydrogen Production
Comment: The Cathcart-Pawel and Baker-Just equations are unable to
determine the increased hydrogen production that occurred in the CORA
and LOFT LP-FP-2 experiments. (29-2, 31-3)
NRC response: The NRC neither agrees nor disagrees with these
comments. The cited experiments were performed to better understand
reactor behavior under severe accident conditions. Increased hydrogen
production under such beyond-design-basis conditions is not relevant in
determining the suitability of the Cathcart-Pawel or Baker-Just
equations when used in evaluations of ECCS cooling performance for
design-basis LOCAs. Section 3.10, ``Issues Related to Cladding
Oxidation and Hydrogen Production,'' of the final technical safety
analysis report provides additional details to support the NRC staff's
position.
15. Issues Related to the Fuel Rod Failure (FRF) Tests Conducted in the
Transient REActor Test (TREAT) Facility Reactor
Comment: Data from the FRF-1 experiment for the TREAT facility
indicate that ECCS evaluation models underpredicted the amount of
hydrogen produced in that experiment. This means that ECCS evaluation
models would underpredict the amount of hydrogen produced in the event
of a LOCA and therefore are not conservative. In addition, neither
Westinghouse nor the NRC applied the Baker-Just equation to
metallurgical data from the locations of FLECHT run 9573 that incurred
autocatalytic oxidation in their application of the Baker-Just equation
under LOCA conditions to evaluate its suitability. For this reason, it
was incorrect for Westinghouse and the NRC to conclude that there is
sufficient conservatism in applying the Baker-Just equation to LOCA
conditions. (2-8, 21-11, 21-12, 28-1)
NRC response: The NRC disagrees with these comments. The NRC
considered the information about the FRF-1 experiment in the TREAT
facility in the 1971 Indian Point Unit 2 licensing hearing and
determined that the ECCS evaluation models were adequate. In addition,
while it is true that the Baker-Just equation has not been applied to
metallurgical data from the locations of FLECHT run 9573 that incurred
autocatalytic oxidation, these data were not collected at the time of
the experiment, and therefore do not exist. However, the NRC staff has
determined that the inability to apply the Baker-Just equation to such
data is an inadequate basis for asserting that it was incorrect for
Westinghouse and the NRC to conclude that there is sufficient
conservatism in applying the Baker-Just equation to LOCA conditions.
Several independent studies have shown that use of the Baker-Just
equation results in sufficiently conservative calculations of the
metal-water reaction rate under design-basis LOCA conditions. Section
3.11, ``Issues Related to the FRF Tests Conducted in the TREAT
Reactor,'' of the final technical safety analysis report provides
additional details to support the NRC staff's position.
16. Issues Raised at the Public Commission Meeting in January 2013
Comment: An NRC document \12\ states that runaway zirconium
oxidation would commence at 1,832 [deg]F in a postulated station
blackout scenario at Grand Gulf Nuclear Station, which indicates the
regulatory limit of 2,200 [deg]F is not conservative. In addition, a
report about best-estimate predictions for the LOFT LP-FP-2 experiments
\13\ states that runaway oxidation would commence if fuel-cladding
temperatures were to start increasing at a rate of 3.0 kelvins/second
(K/s). Since an analysis in support of the NRC staff's interim
evaluation of the petitions showed heatup rates of 10.3 K/s and 11.9 K/
s at
[[Page 1028]]
2,199 [deg]F, this indicates that runaway oxidation has occurred at
temperatures below the 2,200 [deg]F limit. (31-1, 31-2)
---------------------------------------------------------------------------
\12\ Haskin FE, Camp AL. Perspectives on Reactor Safety. NUREG/
CR-6042 (SAND93-0971). Washington, DC: U.S. Nuclear Regulatory
Commission; 1994. Available at: https://www.nrc.gov/docs/ML0727/ML072740014.pdf.
\13\ Guntay S, Carboneau M, Anoda Y. Best Estimate Prediction
for OECD LOFT Project Fission Product Experiment LP-FP-2. OECD LOFT-
T-3803. Idaho Falls, ID: EG&G IDAHO, INC.; 1985. Available at ADAMS
accession no. ML071940361.
---------------------------------------------------------------------------
NRC response: The NRC disagrees with the comments. First, the
postulated station blackout scenario discussed in the document is a
severe accident that involves conditions that are beyond the design
basis, and it is inappropriate to evaluate the regulatory limit of
2,200 [deg]F for design-basis LOCAs using information obtained from
models of severe accidents, which model conditions that are more severe
than those of design-basis accidents and therefore do not provide
information about how fuel cladding would respond to high temperatures
under design-basis LOCA conditions. Second, the NRC staff has
determined that the runaway oxidation described in the cited LOFT LP-
FP-2 report was initiated because of the high temperature (2,870
[deg]F), not because of the heatup rate of 3.0 K/s. Therefore, the NRC
staff concluded that there is no basis for the assertion that runaway
oxidation has occurred at temperatures below the 2,200 [deg]F limit
because heatup rates of more than 3.0 K/s have been observed at lower
temperatures. Section 3.12, ``Issues Raised at the Public Commission
Meeting in January 2013,'' of the final technical safety analysis
report provides additional details to support the NRC staff's position.
III. NRC Technical Evaluation and Reasons for Denial
The NRC staff used a special review process to evaluate these
petitions. It did this for three main reasons: (1) Additional time and
resources were needed to reevaluate more than 40 years of severe
accident and thermal-hydraulic experimental data from more than 200
technical references to address all arguments in the petitions; (2) to
promptly respond to any significant safety issues, if any were to be
identified; and (3) to keep the public informed and to publicly address
any stakeholder concerns about the adequacy of the NRC's regulations
following the accident that occurred in 2011 at the Fukushima Dai-ichi
Nuclear Power Station in Japan.
As part of this special review process, the NRC made a series of
draft interim reports available to the public. These reports informed
the public of NRC's progress in evaluating the petitions and included
the NRC staff's initial evaluation of specific issues and relevant data
that were prioritized to determine the order in which they would be
evaluated. Information about how to access these draft interim reports
is provided in Section IV.
The NRC staff completed its technical evaluation of the petitions
and prepared a final technical safety analysis report that documents
the official technical basis for the staff's evaluation. This final
technical safety analysis report includes the NRC staff's evaluation of
(1) each of the three main issues raised in the petitions and (2)
additional technical issues that are not directly related to the
requested changes to the NRC's regulations that were raised in either
the petitions or in subsequent communications (e.g., submitted public
comments, email messages, letters, and oral statements in a public
meeting with the Commission).
Overall, the NRC is denying the petitions because the petitioner
did not present sufficient new information or arguments to support the
requested changes. In addition, the NRC disagrees with the arguments in
the petitions and concludes that the requested amendments to its
regulations and associated regulatory guidance on ECCS acceptance
criteria or evaluation models are not necessary. The remaining
paragraphs of Section III summarize the staff's evaluation of each of
the three main issues identified in the petitions and identify the
relevant section of the staff's final technical safety analysis report
that provides additional details to support the NRC's position.
Information about how to access the final technical safety analysis
report is provided in Section IV.
Issue 1: Calculated Maximum Fuel Element Cladding Temperature Limit
The NRC staff reviewed experimental data and information from the
multirod (assembly) severe fuel damage experiments cited in the
petitions and found no evidence of temperature escalation rates that
demonstrated the occurrence of autocatalytic or runaway oxidation
reactions at Zircaloy cladding temperatures less than 2,200 [deg]F.
Although some rapid temperature increases were observed in the data
from the cited experiments, the NRC staff disagrees with the assertion
that these data indicate that (1) autocatalytic metal-water oxidation
reactions and uncontrolled temperature excursions involving Zircaloy
cladding have occurred at temperatures less than the 2,200 [deg]F limit
under LOCA conditions and (2) the 2,200 [deg]F limit is therefore not
conservative. The NRC staff has further determined that the 2,200
[deg]F limit in Sec. 50.46(b)(1) provides an adequate margin of safety
to preclude autocatalytic metal-water oxidation reactions.
Therefore, the NRC concludes that the petitioner did not provide
sufficient information to support amending 10 CFR 50.46 to require that
the calculated maximum fuel element cladding temperature not exceed a
limit based on data from cited experiments, instead of the 2,200 [deg]F
limit in Sec. 50.46(b)(1). Section 2.1, ``Peak Cladding Temperature
Limit is Nonconservative,'' of the final technical safety analysis
report provides additional details to support the staff's position.
Issue 2: Metal-Water Reaction Rate Equations for ECCS Evaluation Models
The NRC staff has determined that: (1) Use of the Cathcart-Pawel
equation generally results in sufficiently accurate calculations of the
metal-water reaction rate that are appropriate for realistic ECCS
evaluation models and (2) use of the Baker-Just equation results in
sufficiently conservative calculations of the metal-water reaction rate
that are appropriate for conservative ECCS evaluation models. The final
technical safety analysis report also cites several independent studies
that provide further support for these findings.
The petitioner relied on two main arguments to support the
assertion that the Cathcart-Pawel and Baker-Just equations are not
conservative. The first argument was that data from cited multirod
(assembly) severe fuel damage experiments indicate both equations are
not conservative for use in analyses that calculate the temperature at
which an autocatalytic or runaway oxidation reaction involving the
Zircaloy cladding would occur in the event of a LOCA. The NRC staff
disagrees with this argument for two reasons: (1) Autocatalytic or
runaway oxidation does not begin at a specific temperature and (2) the
petitioner made invalid comparisons between the results of specific
experiments and generic calculations that were not intended to be
applied to a specific test facility.
The second argument was that the Cathcart-Pawel and Baker-Just
equations were not developed to consider how complex thermal-hydraulic
phenomena would affect the metal-water reaction rate in the event of a
LOCA. However, consistent with the technical safety analysis that was
performed for PRM-50-76, the NRC staff determined that--for the
development of metal-water reaction rate equations--well-characterized
isothermal tests are more important than the complex thermal hydraulics
suggested in the petitions. The suggested use of complex thermal-
hydraulic conditions would be counterproductive in tests to
experimentally derive reaction rate correlations because temperature
control is required to develop a
[[Page 1029]]
consistent set of data for correlation derivation. Isothermal tests
provide this necessary temperature control. However, previous studies
have applied the derived correlations to transients that include
complex thermal-hydraulic conditions to verify that the proposed
phenomena embodied in the correlations are limiting. These studies
showed that (1) use of the Cathcart-Pawel equation results in
conservative or best-estimate calculations of the metal-water reaction
rate and (2) use of the Baker-Just equation results in conservative
calculations of the metal-water reaction rate.
Therefore, the NRC concludes that the petitioner did not provide
sufficient information to support revising RG 1.157 and appendix K to
10 CFR part 50 to require that the rates of energy release, hydrogen
generation, and Zircaloy cladding oxidation from the metal-water
reaction of zirconium with steam considered in evaluation models used
to calculate ECCS cooling performance be calculated based on data from
cited experiments, instead of using the Cathcart-Pawel or Baker-Just
equations. Section 2.2, ``Baker-Just and Cathcart-Pawel Equations are
Nonconservative'' of the final technical safety analysis report
provides additional details to support the NRC staff's position.
Issue 3: Minimum Allowable Core Reflood Rate
NRC calculations using simulations of a Zircaloy cladding bundle
with the geometry and design that was used for the cited multirod
(assembly) severe fuel damage experiments disproved the petitioner's
assertions about the reflood rate. In particular, calculations using
simulations showed that steam cooling would be sufficient to maintain
the Zircaloy cladding temperatures below the 2,200 [deg]F limit
specified in Sec. 50.46(b)(1). Moreover, the NRC staff determined that
(1) cooling of a fuel rod bundle depends on several parameters and heat
transfer mechanisms rather than on the reflood rate alone; (2) linear
extrapolation of initial Zircaloy cladding temperatures to predict
final cladding temperature is inappropriate because of increased
radiative cooling at higher temperatures; and (3) extrapolation of
experimental data does not show ``with high probability'' that peak
cladding temperatures will exceed 2,200 [deg]F.
Therefore, the NRC staff concludes that the petitioner did not
provide sufficient information to support issuance of a new regulation
that requires minimum allowable core reflood rates in the event of a
LOCA. Section 2.3, ``Need for a Minimum Allowable Reflood Rate,'' of
the final technical safety analysis report provides additional details
to support the NRC staff's position.
IV. Availability of Documents
Table II provides information about how to access the documents
referenced in this document. The ADDRESSES section of this document
provides additional information about how to access ADAMS.
Table II--Information About How To Access Referenced Documents
----------------------------------------------------------------------------------------------------------------
ADAMS accession
Date Document No. or Federal
Register citation
----------------------------------------------------------------------------------------------------------------
Submitted Petitions
----------------------------------------------------------------------------------------------------------------
May 1, 2002......................................... Petition for Rulemaking (PRM-50-76)... ML022240009
November 17, 2009................................... Petition for Rulemaking (PRM-50-93)... ML093290250
June 7, 2010........................................ Petition for Rulemaking (PRM-50-95)... ML102770018
----------------------------------------------------------------------------------------------------------------
Federal Register Notices
----------------------------------------------------------------------------------------------------------------
September 6, 2005................................... Denial of Petition for Rulemaking (PRM- 70 FR 52893
50-76).
January 25, 2010.................................... Notice of Receipt of Petition for 75 FR 3876
Rulemaking (PRM-50-93).
October 27, 2010.................................... Notice of Consolidation of Petitions 75 FR 66007
for Rulemaking and Re-Opening of
Comment Period (PRM-50-93 and PRM-50-
95).
----------------------------------------------------------------------------------------------------------------
Consolidated Public Comments Document
----------------------------------------------------------------------------------------------------------------
November 21, 2017................................... Public Comments on Petitions for ML17325A007
Rulemaking: Calculated Maximum Fuel
Element Cladding Temperature.
----------------------------------------------------------------------------------------------------------------
Draft Interim Reports
----------------------------------------------------------------------------------------------------------------
August 23, 2011..................................... Draft Interim Review of PRM-50-93/95 ML112290888
Issues Related to the CORA Tests.
September 27, 2011.................................. Draft Interim Review of PRM-50-93/95 ML112650009
Issues Related to the LOFT LP-FP-2
Test.
October 16, 2012.................................... Draft Interim Review of PRM-50-93/95 ML12265A277
Issues Related to Conservatism of
2200 [deg]F, Metal-Water Reaction
Rate Correlations, and ``The
Impression Left from [FLECHT] Run
9573.''.
March 8, 2013....................................... Draft Interim Review of PRM-50-93/95 ML13067A261
Issues Related to Minimum Allowable
Core Reflood Rate.
----------------------------------------------------------------------------------------------------------------
Final Technical Safety Analysis Report
----------------------------------------------------------------------------------------------------------------
August 19, 2016..................................... Technical Safety Analysis of PRM-50-93/ ML16078A318
95, Petition for Rulemaking on Sec.
50.46.
----------------------------------------------------------------------------------------------------------------
V. Conclusion
For the reasons cited in this document, the NRC is denying PRM-50-
93 and PRM-50-95. The petitioner did not present sufficient new
information or arguments to support the requested changes. In addition,
the NRC disagrees with the arguments in the petitions and concludes
that the requested amendments to its regulations and associated
regulatory guidance are not necessary. The NRC's existing regulations
provide reasonable assurance of adequate protection of public health
and safety.
[[Page 1030]]
Dated: December 29, 2020.
For the Nuclear Regulatory Commission.
Annette L. Vietti-Cook,
Secretary of the Commission.
[FR Doc. 2020-29151 Filed 1-6-21; 8:45 am]
BILLING CODE 7590-01-P