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.

Proposed Rules Federal Register

1022 

Vol. 86, No. 4 

Thursday, January 7, 2021 

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. 

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: Dawn.Forder@nrc.gov. 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 publiclyavailable 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 pdr.resource@nrc.gov. 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 pdr.resource@nrc.gov 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: Daniel.Doyle@nrc.gov, 

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 § 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 § 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 § 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 § 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. 


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1Under § 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. 

2Criterion 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. 

3Regulatory 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.’’ 

4RG 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.’’ 

5RG 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.’’ 

6RG 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.’’ 

Issue 1: Calculated Maximum Fuel 

Element Cladding Temperature Limit 

Background for Issue 1 

Under § 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 § 50.46(b).2 Under 

§ 50.46(b)(1), the calculated maximum 

fuel element cladding temperature shall 

not exceed 2,200 °F. In addition, 

§ 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. 

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 °F specified 

in § 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 °F. The petitioner provided this 

assertion as evidence that the 2,200 °F 

limit is not conservative, and requested 

that the NRC amend § 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 °F 

limit specified in § 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 § 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 § 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 § 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. 

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 § 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 

§ 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. 

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 °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 °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. 


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7Extrapolation 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. 

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 BakerJust 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 BakerJust 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 °F 

limit in § 50.46(b)(1) if, at the onset of 

reflood, the cladding temperature was 

greater than or equal to 1,200 °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 °F limit, if the 

cladding temperature was greater than 

or equal to 1,200 °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. 

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

Category 

Number of 

comment 

submissions 

Number of 

individual 

comments 

Comments from the Petitioner ................................................................................................................................. a 13 a 97 

Comments from Nuclear Industry Representatives ................................................................................................ 3 9 

Comments from Public Interest Groups or Other Interested Individuals ................................................................ 17 19 

Total .................................................................................................................................................................. 33 125 

aThe 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. 


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8Robert 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). 

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. 

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 metalwater 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 thermalhydraulic 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 °F, indicating the regulatory limit 

of 2,200 °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 


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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. 

10TRAC: Transient Reactor Analysis Code. 

RELAP: Reactor Excursion and Leak Analysis 

Program. 

escalation rates that demonstrated the 

occurrence of autocatalytic or runaway 

oxidation reactions below 2,200 °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 CathcartPawel 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 metalwater 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 °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 °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 HighTemperature (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 °F. This 

provides additional evidence indicating 

that the regulatory limit of 2,200 °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 °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 °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 °F could 

significantly reduce the safety margins 

for the following types of materials 

interactions: (1) Degradation of boilingwater 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) 

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 °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 °F (1,204 

°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 

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 °F, despite its low reflood rate. In 

addition, FLECHT run 9573 was a lowreflood-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, 


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11Committee 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.oecdnea.org/nsd/docs/1999/csni-r99-23.pdf. 

12Haskin 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. 

‘‘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 °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 °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 °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 °F to 

2,012 °F. (2–7) 

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 °F for 

experiments simulating conditions for 

design-basis accidents. Section 3.7, 

‘‘Issue Related to Whether Runaway 

Oxidation Temperatures Start at 1100 °C 

(2012 °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 metalwater 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-designbasis 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 °F in a 

postulated station blackout scenario at 

Grand Gulf Nuclear Station, which 

indicates the regulatory limit of 2,200 °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 


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2,199 °F, this indicates that runaway 

oxidation has occurred at temperatures 

below the 2,200 °F limit. (31–1, 31–2) 

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 °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 °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 °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 thermalhydraulic 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 °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 °F limit under LOCA conditions 

and (2) the 2,200 °F limit is therefore not 

conservative. The NRC staff has further 

determined that the 2,200 °F limit in 

§ 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 °F limit in § 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—wellcharacterized isothermal tests are more 

important than the complex thermal 

hydraulics suggested in the petitions. 

The suggested use of complex thermalhydraulic conditions would be 

counterproductive in tests to 

experimentally derive reaction rate 

correlations because temperature 

control is required to develop a 


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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 CathcartPawel 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 metalwater 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 °F limit specified in 

§ 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 °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

Date Document 

ADAMS accession 

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–50–76) ....................................................................... 70 FR 52893 

January 25, 2010 ............... Notice of Receipt of Petition for Rulemaking (PRM–50–93) ..................................................... 75 FR 3876 

October 27, 2010 ............... Notice of Consolidation of Petitions for Rulemaking and Re-Opening of Comment Period 

(PRM–50–93 and PRM–50–95).

75 FR 66007 

Consolidated Public Comments Document 

November 21, 2017 ............ Public Comments on Petitions for Rulemaking: Calculated Maximum Fuel Element Cladding 

Temperature.

ML17325A007 

Draft Interim Reports 

August 23, 2011 ................. Draft Interim Review of PRM–50–93/95 Issues Related to the CORA Tests ........................... ML112290888 

September 27, 2011 ........... Draft Interim Review of PRM–50–93/95 Issues Related to the LOFT LP–FP–2 Test .............. ML112650009 

October 16, 2012 ............... Draft Interim Review of PRM–50–93/95 Issues Related to Conservatism of 2200 °F, MetalWater Reaction Rate Correlations, and ‘‘The Impression Left from [FLECHT] Run 9573.’’.ML12265A277 

March 8, 2013 .................... Draft Interim Review of PRM–50–93/95 Issues Related to Minimum Allowable Core Reflood 

Rate.

ML13067A261 

Final Technical Safety Analysis Report 

August 19, 2016 ................. Technical Safety Analysis of PRM–50–93/95, Petition for Rulemaking on § 50.46 .................. ML16078A318 

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. 


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1 21 CFR 1301.13(e)(1)(iv). 

2 21 CFR 1301.13(e)(1)(1)(i)–(iii), (v)–(vi), and 

(viii)–(x). 

3 21 CFR 1301.13(e)(1)(vii). 

4 21 CFR 1309.21. 

5 21 CFR 1301.13(e)(1) and 1309.21 

6https://www.deadiversion.usdoj.gov/drugreg/ 

index.html#regapps. 

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 

DEPARTMENT OF JUSTICE 

Drug Enforcement Administration 

21 CFR Parts 1301, 1309, and 1321 

[Docket No. DEA–587] 

RIN 1117–AB58 

Amending Regulations To Require 

Online Submission of Applications for 

and Renewals of DEA Registration 

AGENCY: Drug Enforcement 

Administration, Department of Justice. 

ACTION: Notice of proposed rulemaking. 

SUMMARY: This rule proposes to amend 

the Drug Enforcement Administration 

(DEA) regulations to require all initial 

and renewal applications for DEA 

registration to be submitted online. 

DATES: Electronic comments must be 

submitted, and written comments must 

be postmarked, on or before March 8, 

2021. Commenters should be aware that 

the electronic Federal Docket 

Management System will not accept any 

comments after 11:59 p.m. Eastern Time 

on the last day of the comment period. 

All comments concerning collections 

of information under the Paperwork 

Reduction Act must be submitted to the 

Office of Management and Budget on or 

before March 8, 2021. 

ADDRESSES: To ensure proper handling 

of comments, please reference ‘‘Docket 

No. DEA–587’’ on all correspondence, 

including any attachments. 

• Electronic comments: The Drug 

Enforcement Administration (DEA) 

encourages that all comments be 

submitted electronically through the 

Federal eRulemaking Portal which 

provides the ability to type short 

comments directly into the comment 

field on the web page or attach a file for 

lengthier comments. Please go to http:// 

www.regulations.gov and follow the 

online instructions at that site for 

submitting comments. Upon completion 

of your submission, you will receive a 

Comment Tracking Number for your 

comment. Please be aware that 

submitted comments are not 

instantaneously available for public 

view on Regulations.gov. If you have 

received a Comment Tracking Number, 

your comment has been successfully 

submitted and there is no need to 

resubmit the same comment. 

• Paper comments: Paper comments 

that duplicate electronic submissions 

are not necessary. Should you wish to 

mail a paper comment, in lieu of an 

electronic comment, it should be sent 

via regular or express mail to: Drug 

Enforcement Administration, Attn: DEA 

Federal Register Representative/DPW, 

8701 Morrissette Drive, Springfield, 

Virginia 22152. 

FOR FURTHER INFORMATION CONTACT: 

Scott A. Brinks, Regulatory Drafting and 

Policy Support Section, Diversion 

Control Division, Drug Enforcement 

Administration; Mailing Address: 8701 

Morrissette Drive, Springfield, Virginia 

22152; Telephone: (571) 362–3261. 

SUPPLEMENTARY INFORMATION: 

Posting of Public Comments 

Please note that all comments 

received are considered part of the 

public record. They will, unless 

reasonable cause is given, be made 

available by the Drug Enforcement 

Administration (DEA) for public 

inspection online at http://

www.regulations.gov. Such information 

includes personal identifying 

information (such as your name, 

address, etc.) voluntarily submitted by 

the commenter. The Freedom of 

Information Act applies to all comments 

received. If you want to submit personal 

identifying information (such as your 

name, address, etc.) as part of your 

comment, but do not want it to be made 

publicly available, you must include the 

phrase ‘‘PERSONAL IDENTIFYING 

INFORMATION’’ in the first paragraph 

of your comment. You must also place 

all of the personal identifying 

information you do not want made 

publicly available in the first paragraph 

of your comment and identify what 

information you want redacted. 

If you want to submit confidential 

business information as part of your 

comment, but do not want it to be made 

publicly available, you must include the 

phrase ‘‘CONFIDENTIAL BUSINESS 

INFORMATION’’ in the first paragraph 

of your comment. You must also 

prominently identify the confidential 

business information to be redacted 

within the comment. 

Comments containing personal 

identifying information or confidential 

business information identified as 

directed above will be made publicly 

available in redacted form. If a comment 

has so much confidential business 

information that it cannot be effectively 

redacted, all or part of that comment 

may not be made publicly available. 

Comments posted to http://

www.regulations.gov may include any 

personal identifying information (such 

as name, address, and phone number) 

included in the text of your electronic 

submission that is not identified as 

confidential as directed above. 

An electronic copy of this proposed 

rule is available at http://

www.regulations.gov for easy reference. 

Legal Authority 

The Controlled Substances Act (CSA) 

grants the Attorney General authority to 

promulgate rules and regulations 

relating to: The registration and control 

of the manufacture, distribution, and 

dispensing of controlled substances and 

listed chemicals; reporting changes to 

professional or business addresses; and 

the efficient execution of his statutory 

functions. 21 U.S.C. 821, 822(a), 827(h), 

871(b), 957(a). The Attorney General is 

further authorized by the CSA to 

promulgate rules and regulations 

relating to the registration and control of 

importers and exporters of controlled 

substances and listed chemicals. 21 

U.S.C. 958(f). The Attorney General has 

delegated this authority to the 

Administrator of DEA. 28 CFR 0.100(b). 

DEA Form 224 applies to new 

registration applications for retail 

pharmacy, hospital/clinic, practitioner, 

teaching institution, or mid-level 

practitioner registrations.1 DEA Form 

225 applies to new registration 

applications for manufacturer, 

distributor, researcher, canine handler, 

analytical laboratory, importer, or 

exporter registrations.2 DEA Form 363 

applies to new registration applications 

for narcotic treatment program 

registrations.3 DEA Form 510 applies to 

new registration applications for 

domestic chemical registrations.4 DEA 

Forms 224a, 225a, 363a, and 510a apply 

to registration renewal applications.5

Purpose of the Proposed Rule 

The purpose of this notice of 

proposed rulemaking is to simplify the 

form submission process by requiring 

that all registration and renewal 

applications be submitted online. 

Currently, DEA regulations permit DEA 

Registration Forms (224/224a, 225/225a, 

363/363a, and 510/510a) to be 

submitted either through the secure 

online database, or by paper forms 

delivered to DEA Headquarters.6 This 

proposed rule will amend DEA 

regulations to require that all 

registration and renewal applications be 


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