Calculated Maximum Fuel Element Cladding Temperature, 1022-1030 [2020-29151]

Agencies

• NUCLEAR REGULATORY COMMISSION

[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
------------------------------------------------------------------------
\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: https://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