12254 Federal Register / Vol. 87, No. 42 / Thursday, March 3, 2022 / Proposed Rules 

NUCLEAR REGULATORY 

COMMISSION 

10 CFR Parts 20, 26, 50, 51, 52, 72, 73, 

140 

[NRC–2015–0070] 

RIN 3150–AJ59 

Regulatory Improvements for 

Production and Utilization Facilities 

Transitioning to Decommissioning 

AGENCY: Nuclear Regulatory 

Commission. 

ACTION: Proposed rule. 

SUMMARY: The U.S. Nuclear Regulatory 

Commission (NRC) is proposing to 

amend its regulations that relate to the 

decommissioning of production and 

utilization facilities. The NRC’s goals in 

amending these regulations are to 

maintain a safe, effective, and efficient 

decommissioning process; reduce the 

need for license amendment requests 

and exemptions from existing 

regulations; address other 

decommissioning issues deemed 

relevant by the NRC; and support the 

NRC’s Principles of Good Regulation, 

including openness, clarity, and 

reliability. The NRC will hold a public 

meeting to promote full understanding 

of this proposed rule and to facilitate 

public comments. 

DATES: Submit comments by May 17, 

2022. Comments received after this date 

will be considered if it is practical to do 

so, but the Commission is able to ensure 

consideration only for comments 

received before this date. 

ADDRESSES: You may submit comments 

by the following method (unless this 

document describes a different method 

for submitting comments on a specific 

subject); however, the NRC encourages 

electronic comment submission through 

the Federal rulemaking website: 

• Federal Rulemaking Website: Go to 

https://www.regulations.gov and search 

for Docket ID NRC–2015–0070. 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. 

• Email comments to: 

Rulemaking.Comments@nrc.gov. If you 

do not receive an automatic email reply 

confirming receipt, then contact us at 

301–415–1677. 

• Mail comments to: Secretary, U.S. 

Nuclear Regulatory Commission, 

Washington, DC 20555–0001, ATTN: 

Rulemakings and Adjudications Staff. 

For additional direction on obtaining 

information and submitting comments, 

see ‘‘Obtaining Information and 

Submitting Comments’’ in the 

SUPPLEMENTARY INFORMATION section of 

this document. 

FOR FURTHER INFORMATION CONTACT: 

Daniel I. Doyle, Office of Nuclear 

Material Safety and Safeguards, U.S. 

Nuclear Regulatory Commission, 

Washington, DC 20555–0001; telephone: 

301–415–3748; email: Daniel.Doyle@

nrc.gov. 

SUPPLEMENTARY INFORMATION: 

Executive Summary 

A. Need for the Regulatory Action 

The NRC is proposing to amend its 

regulations related to the 

decommissioning of production and 

utilization facilities. The Commission 

directed the NRC staff to proceed with 

an integrated rulemaking on nuclear 

power reactor decommissioning to 

address the following: A graded 

approach to emergency preparedness 

(EP), lessons learned from the licensees 

that have already gone through (or are 

currently going through) the 

decommissioning process, the 

advisability of requiring a licensee’s 

post-shutdown decommissioning 

activities report (PSDAR) to be approved 

by the NRC, the appropriateness of 

maintaining the three existing options 

for decommissioning and the 

timeframes associated with those 

options, the appropriate role of State 

and local governments and nongovernmental stakeholders in the 

decommissioning process, and any 

other issues deemed relevant by the 

NRC staff. 

Compared to an operating nuclear 

power reactor, the risk of an offsite 

radiological release is significantly 

lower, and the types of possible 

accidents are significantly fewer, at a 

nuclear power reactor that has 

permanently ceased operations and 

removed fuel from the reactor vessel. As 

a direct result, there is no need for the 

NRC to impose new requirements in the 

areas identified in this rulemaking to 

address safety or security concerns. 

Instead, the requirements in 

decommissioning should be aligned 

with the reduction in risk that occurs 

over time, while maintaining safety and 

security. The decommissioning process 

can be improved and made more 

efficient, open, and predictable by 

reducing the reliance on licensing 

actions (i.e., license amendment and 

exemption requests) that reflect this 

reduction in risk to achieve a 

sustainable regulatory framework during 

decommissioning. 

The NRC has also determined that 

changes to the regulations are 

appropriate with respect to drug and 

alcohol testing; cyber security; and 

foreign ownership, control, or 

domination of a production or 

utilization facility undergoing 

decommissioning. 

In several areas, the current 

regulations do not distinguish between 

provisions that apply to a nuclear power 

reactor that has permanently ceased 

operations and provisions that apply to 

an operating nuclear power reactor. To 

address this, the NRC is proposing to 

amend its regulations in several areas to 

provide a regulatory framework for the 

transition from operating to 

decommissioning. This proposed rule is 

a four-step graded approach that is 

commensurate with the reduction in 

radiological risk at four levels of 

decommissioning: (1) Permanent 

cessation of operations and permanent 

removal of all fuel from the reactor 

vessel, (2) sufficient decay of fuel in the 

spent fuel pool (SFP) such that it would 

not reach ignition temperature within 

10 hours under adiabatic heatup 

conditions (i.e., a complete loss of SFP 

water inventory with no heat loss), (3) 

transfer of all fuel to dry storage, and (4) 

removal of all fuel from the site. The 

graded approach is a fundamental 

concept for this proposed rule. 

Because the current regulatory 

framework for decommissioning is 

adequate to protect public health and 

safety and the common defense and 

security, many of the new requirements 

in this proposed rule are alternatives to 

current requirements. 

B. Major Provisions 

Major provisions of this proposed rule 

include changes in the following areas: 

• Emergency preparedness. This 

proposed rule offers an alternative, 

graded approach to the current 

requirements for onsite and offsite 

radiological emergency preparedness at 

a nuclear power reactor. This approach 

would provide four levels of emergency 

planning standards that coincide with 

significant milestones in 

decommissioning that reflect the 

gradual reduction of the radiological 

risk during decommissioning. 

• Physical security. This proposed 

rule would make certain changes that 

would apply once a nuclear power 

reactor enters decommissioning. These 

proposed changes would (1) permit a 

certified fuel handler (CFH) to approve 

the temporary suspension of security 

measures during certain emergency 

conditions or during severe weather, (2) 

remove the requirement that a licensee’s 

physical protection program be 


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designed to prevent significant core 

damage, (3) remove the requirement that 

a licensee must designate the reactor 

control room as a ‘‘vital area,’’ and (4) 

replace the requirement for maintaining 

continuous communications between 

the alarm stations and the control room 

with a requirement for maintaining 

communications between alarm stations 

and the CFH or senior on shift licensee 

representative, or both. This last change 

would clarify the management role of 

the CFH in a manner that is consistent 

with § 50.54(y) of title 10 of the Code of 

Federal Regulations (10 CFR). The NRC 

is also proposing to revise § 50.54(p) to 

add definitions for ‘‘change’’ and 

‘‘decrease in safeguards effectiveness,’’ 

as those terms apply to the process for 

making changes to the security plans of 

licensees under 10 CFR part 50, 

‘‘Domestic Licensing of Production and 

Utilization Facilities,’’ and 10 CFR part 

52, ‘‘Licenses, Certifications, and 

Approvals for Nuclear Power Plants,’’ 

with operating, decommissioning, or 

decommissioned reactor units. In 

addition, this proposed rule would 

provide an option for a licensee to 

protect a general license independent 

spent fuel storage installation (ISFSI) 

under the physical security 

requirements in § 73.51, ‘‘Requirements 

for the physical protection of stored 

spent nuclear fuel and high-level 

radioactive waste,’’ for a specific license 

ISFSI instead of the physical security 

requirements in § 73.55, ‘‘Requirements 

for physical protection of licensed 

activities in nuclear power reactors 

against radiological sabotage,’’ for a 

nuclear power reactor once all spent 

fuel has been moved to dry storage. 

• Cyber security. This proposed rule 

would provide that the cyber security 

requirements in § 73.54, ‘‘Protection of 

digital computer and communication 

systems and networks,’’ continue to 

apply to a nuclear power reactor after 

the licensee’s permanent cessation of 

operations, until all the fuel has been 

removed from the reactor vessel and 

there has been sufficient decay of the 

fuel in the SFP such that it would not 

reach ignition temperature within 10 

hours under adiabatic heatup 

conditions, at which point no digital 

computer and communications systems 

would be required to meet the criteria 

of § 73.54. This proposed rule would 

also provide for the removal of the cyber 

security license condition for 10 CFR 

part 50 nuclear power reactor licensees 

after the spent fuel decay period. 

• Drug and alcohol testing. This 

proposed rule would correct 

inconsistencies in the NRC’s regulations 

for fitness-for-duty (FFD) programs and 

clarify provisions regarding a nuclear 

power reactor licensee’s insider 

mitigation program (IMP). 

• Certified fuel handler definition 

and elimination of the shift technical 

advisor. This proposed rule would 

retain the existing definition for 

‘‘certified fuel handler’’ and add an 

alternative that would eliminate the 

need for nuclear power reactor licensees 

to seek the Commission’s approval of a 

fuel handler training program. The 

proposed provision would require the 

training program to address the safe 

conduct of decommissioning activities, 

safe handling and storage of spent fuel, 

and appropriate response to plant 

emergencies. The proposed alternative 

specifies that a CFH must be qualified 

in accordance with a fuel handler 

training program that meets the same 

requirements as training programs for 

non-licensed operators required by 

§ 50.120, ‘‘Training and qualification of 

nuclear power plant personnel.’’ This 

proposed rule would also clarify that a 

Shift Technical Advisor (STA) is not 

required for decommissioning nuclear 

power reactors. 

• Decommissioning funding 

assurance. This proposed rule 

recommends several changes regarding 

decommissioning funding for nuclear 

power reactors. It would modify the 

reporting frequency in § 50.75 to be 

consistent with the decommissioning 

funding assurance reporting frequency 

for ISFSIs in § 72.30(c). For ISFSI 

funding reports, this proposed rule 

would allow licensees to combine the 

reports that are required by 

§ 50.82(a)(8)(v), § 50.82(a)(8)(vii), and 

§ 72.30 and remove the requirement for 

NRC approval of ISFSI reports filed 

under § 72.30(c). It also would clarify 

that although the regulations establish a 

continuing obligation to provide 

reasonable assurance of 

decommissioning funding, when a 

licensee identifies a shortfall in the 

report required by § 50.75(f)(1), the 

licensee must obtain additional 

financial assurance to cover the shortfall 

and discuss that information in the next 

report. In addition, this proposed rule 

would make administrative changes to 

ensure consistency with § 50.4, ‘‘Written 

communications,’’ regarding the 

submission of notifications and to 

eliminate § 50.75(f)(2) because 

§ 50.75(f)(1) fully encompasses 

paragraph (f)(2). Besides proposing 

conforming changes to 10 CFR part 52, 

the NRC is asking whether the NRC 

should maintain identical requirements 

in § 52.110 and § 50.82. 

• Offsite and onsite financial 

protection requirements and indemnity 

agreements. This proposed rule would 

allow certain nuclear power reactor 

licensees in decommissioning to reduce 

the insurance amounts that they are 

required to maintain without obtaining 

exemptions from the NRC’s regulations. 

• Environmental considerations. This 

proposed rule would clarify that 

licensees must evaluate the 

environmental impacts of 

decommissioning and whether they are 

bounded by previous environmental 

reviews in the PSDAR. The proposed 

rule would also clarify environmental 

reporting requirements. 

• Record retention requirements. This 

proposed rule would remove certain 

record retention requirements for 

structures, systems, and components 

(SSCs) that no longer remain in service 

during decommissioning and would 

remove requirements to keep multiple 

copies of certain spent fuel storage 

records. The NRC is also asking a 

specific question concerning the 

recordkeeping requirements for facilities 

licensed under 10 CFR part 52. 

• Low-level waste transportation. 

This proposed rule would allow a 45- 

day window for notification of receipt of 

shipments of low-level radioactive 

waste (LLW). This increase from the 

current 20-day notification window is 

based on operating experience that 

shows that 45 days is an appropriate 

amount of time for notification of LLW 

shipments. 

• Spent fuel management planning. 

This proposed rule would clarify 

requirements that the decommissioning 

documents contain information on 

spent fuel management planning in 

accordance with the regulatory 

requirements in § 72.218, ‘‘Termination 

of licenses.’’ 

• Backfit rule. This proposed rule 

would clarify how the NRC applies 

§ 50.109, ‘‘Backfitting,’’ to nuclear 

power reactor licensees in 

decommissioning and would make 

conforming changes to § 72.62. 

• Foreign ownership, control, or 

domination. This proposed rule would 

specify the criteria for when a facility is 

no longer a production or utilization 

facility and that the foreign ownership, 

control, or domination (FOCD) 

prohibition found in § 50.38, 

‘‘Ineligibility of certain applicants,’’ no 

longer applies to a person seeking a 

license for such a facility. 

• Clarification of scope of license 

termination plan requirement. This 

proposed rule would clarify that the 

requirement for a license termination 

plan in §§ 50.82(a)(9) and 52.110(i) 

applies only to nuclear power reactor 

licensees that have loaded fuel into the 

reactor. 

• Removal of license conditions and 

withdrawal of orders made redundant 


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by regulation. This proposed rule would 

deem removed conditions imposed 

upon individual licensees and withdraw 

NRC orders that have been identified as 

having been made redundant by 

subsequent regulation resulting in their 

requirements being generically 

applicable. License conditions deemed 

removed would be actually removed by 

administrative license amendment 

subsequent to the effective date of the 

final rule. The NRC is interested in 

obtaining stakeholder input to identify 

potential redundant requirements not 

listed in this proposed rule. 

• Changes for consistent treatment of 

holders of combined licenses and 

operating licenses. The proposed rule 

would improve consistency in 

regulatory treatment for combined 

license (part 52) and operating license 

(part 50) holders by aligning regulatory 

applicabilities for combined license 

holders upon submittal of the 

§ 52.110(a) certifications with regulatory 

applicabilities for operating license 

holders upon submittal of the 

§ 50.82(a)(1) certifications. 

C. Costs and Benefits 

The NRC prepared a draft regulatory 

analysis to determine the expected 

quantitative costs and benefits of this 

proposed rule, as well as qualitative 

factors to be considered in the NRC’s 

rulemaking decision. The conclusion of 

the analysis is that this proposed rule 

would result in net savings to 

production and utilization facility 

licensees and the NRC. The analysis 

combines the costs and benefits from 

the decommissioning areas of EP, 

physical security, cyber security, drug 

and alcohol testing, CFH training, 

decommissioning funding assurance, 

offsite and onsite financial protection 

requirements and indemnity 

agreements, environmental 

considerations, records retention, lowlevel waste transportation, spent fuel 

management planning, application of 

the Backfit Rule, FOCD, and 

clarification of the scope of a license 

termination plan. The analysis discusses 

the economic impact to the nuclear 

industry, government, and society from 

the rulemaking and associated guidance. 

The draft regulatory analysis 

discusses the cost benefit analysis for 

the various alternatives of each area of 

decommissioning proposed by the NRC, 

and shows that the NRC’s proposed rule 

and guidance development is overall 

cost beneficial to the nuclear industry, 

government, and society as shown in 

Table 1. 

TABLE 1—SUMMARY OF COSTS AND BENEFITS (7% NPV) 

Benefits Costs Net benefit 

$18,315,000 $(401,000) $17,914,000 

The draft regulatory analysis also 

considers, in a qualitative fashion, 

regulatory efficiency, public health and 

safety, and common defense and 

security. For the regulatory efficiency 

aspect, this proposed rule would enable 

the NRC to better maintain and 

administer regulatory activities over the 

decommissioning process and ensure 

that the requirements for 

decommissioning production and 

utilization facilities are clear and 

appropriate. This proposed rule would 

also continue to provide reasonable 

assurance of adequate protection of the 

public health and safety and promote 

the common defense and security and 

protect the environment at production 

and utilization facility sites that have 

started decommissioning. 

Based on these quantitative and 

qualitative factors, the draft regulatory 

analysis concludes that the proposed 

rule should be adopted. For more 

information, please see the draft 

regulatory analysis available at the 

NRC’s Agencywide Documents Access 

and Management System (ADAMS) 

under Accession No. ML22019A132. 

Table of Contents 

I. Obtaining Information and Submitting 

Comments 

A. Obtaining Information 

B. Submitting Comments 

II. Background 

A. 1988 Decommissioning Rule 

B. 1996 Decommissioning Rule 

C. Post-1996 Final Rule Decommissioning 

Activity 

D. Spent Fuel Pool Studies 

E. Changes in Nuclear Power Reactor 

Decommissioning at the NRC and Within 

the Nuclear Power Industry 

F. Decommissioning Lessons Learned 

Report 

G. Initiation of This Proposed Rule 

H. Advance Notice of Proposed 

Rulemaking 

I. Regulatory Basis 

III. Discussion 

A. Current Regulatory Process 

B. Objectives of This Proposed Rule 

C. Applicability 

D. Applicability to NRC Licensees During 

Operations 

E. Applicability to ISFSI-Only and 

Standalone ISFSI/Decommissioned 

Reactor Sites 

F. Graded Approach 

G. Technical Basis for Graded Approach 

H. Levels of Decommissioning 

1. Level 1 

2. Level 2 

3. Level 3 

4. Level 4 

IV. Scope of the Proposal 

A. Emergency Preparedness 

1. Introduction 

2. Graded Approach for Emergency 

Preparedness 

3. Licensee Supporting Analyses 

4. Post-Shutdown Emergency Plans 

5. Permanently Defueled Emergency Plans 

6. Independent Spent Fuel Storage 

Installation-Only Emergency Plans 

7. All Spent Fuel Removed From Site 

8. Changes to Emergency Plans 

9. Program Element Review Under 

§ 50.54(t) 

10. Reasonable Assurance and Offsite 

Radiological Emergency Preparedness 

11. Clean-up of Regulations 

12. Revisions to § 72.32 

B. Physical Security 

1. Security Plans 

2. Dry Cask Storage 

3. Significant Core Damage 

4. Vital Areas 

5. Communications 

6. Suspension of Security Measures 

C. Cyber Security 

D. Drug and Alcohol Testing 

1. Scope of 10 CFR Part 26 

2. Fitness-for-Duty Elements for Insider 

Mitigation Program 

3. Criminal Penalties 

E. Certified Fuel Handler Definition and 

Elimination of the Shift Technical 

Advisor 

1. Alternative Definition for Certified Fuel 

Handler 

2. Elimination of the Shift Technical 

Advisor 

F. Decommissioning Funding Assurance 

1. Clarification of § 50.82(a) and § 52.110(h) 

2. Changes to Reporting Requirements 

3. Shortfalls in Decommissioning Funding 

Assurance 

4. Conforming Changes to 10 CFR Part 52 

5. Change to 10 CFR Part 72 

G. Offsite and Onsite Financial Protection 

Requirements and Indemnity 

Agreements 

1. Proposed Revisions to Offsite Liability 

and Onsite Property Insurance 

Requirements 

2. Proposed Revision to Extraordinary 

Nuclear Occurrences Requirements 

3. Proposed New Rule Language in 

§ 50.54(w)(6) 

H. Environmental Considerations 


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1. Clarifying Changes to 10 CFR Parts 50 

and 52 

2. Consistency Changes to 10 CFR Part 51 

I. Record Retention Requirements 

J. Low-Level Waste Transportation 

K. Spent Fuel Management Planning 

1. Requirements for the IFMP in § 50.54(bb) 

and the PSDAR in § 50.82 and § 52.110 

2. Requirements in § 72.218 for 

Termination of the General License for 

Spent Fuel Storage 

L. Backfit Rule 

M. Foreign Ownership, Control, or 

Domination 

N. Clarification of Scope of License 

Termination Plan Requirement 

O. Removal of License Conditions and 

Withdrawal of Orders 

P. Changes for Consistent Treatment of 

Holders of Combined Licenses and 

Operating Licenses 

V. Specific Requests for Comments 

VI. Section-by-Section Analysis 

VII. Regulatory Flexibility Certification 

VIII. Regulatory Analysis 

IX. Backfitting and Issue Finality 

A. Current and Future Applicants 

B. Existing Design Certifications 

C. Existing Licensees 

D. Backfit Analysis 

1. Introduction and Background 

2. Detailed Description of the Proposed 

Change Affecting Issue Finality 

3. Benefits: Substantial Increase in Public 

Health and Safety and Common Defense 

and Security 

4. Costs 

5. Determination of Substantial Benefits 

Justifying Costs of the Proposed Change 

Affecting Issue Finality 

6. Conclusion 

7. Evaluation of Factors in § 50.109(c)(1) 

Through (9) 

E. Draft Regulatory Guidance 

X. Cumulative Effects of Regulation 

XI. Plain Writing 

XII. National Environmental Policy Act 

XIII. Paperwork Reduction Act 

XIV. Criminal Penalties 

XV. Voluntary Consensus Standards 

XVI. Availability of Guidance 

XVII. Public Meeting 

XVIII. Availability of Documents 

I. Obtaining Information and 

Submitting Comments 

A. Obtaining Information 

Please refer to Docket ID NRC–2015– 

0070 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–2015–0070. 

• NRC’s Agencywide Documents 

Access and Management System 

(ADAMS): You may obtain publicly 

available documents online in the 

ADAMS Public Documents collection at 

https://www.nrc.gov/reading-rm/ 

adams.html. To begin the search, select 

‘‘Begin Web-based ADAMS Search.’’ For 

problems with ADAMS, please contact 

the NRC’s Public Document Room (PDR) 

reference staff at 1–800–397–4209, at 

301–415–4737, or by email to 

pdr.resource@nrc.gov. For the 

convenience of the reader, instructions 

about obtaining materials referenced in 

this document are provided in the 

‘‘Availability of Documents’’ section of 

this document. 

• NRC’s PDR: You may examine and 

purchase copies of public documents, 

by appointment, at the NRC’s PDR, 

Room P1 B35, One White Flint North, 

11555 Rockville Pike, Rockville, 

Maryland 20852. To make an 

appointment to visit the PDR, please 

send an email to PDR.Resource@nrc.gov 

or call 1–800–397–4209 or 301–415– 

4737, between 8:00 a.m. and 4:00 p.m. 

(ET), Monday through Friday, except 

Federal holidays. 

B. Submitting Comments 

The NRC encourages electronic 

comment submission through the 

Federal rulemaking website (https://

www.regulations.gov). Please include 

Docket ID NRC–2015–0070 in your 

comment submission. 

The NRC cautions you not to include 

identifying or contact information that 

you do not want to be publicly 

disclosed in your comment submission. 

The NRC will post all comment 

submissions at https://

www.regulations.gov as well as enter the 

comment submissions into ADAMS. 

The NRC does not routinely edit 

comment submissions to remove 

identifying or contact information. 

If you are requesting or aggregating 

comments from other persons for 

submission to the NRC, then you should 

inform those persons not to include 

identifying or contact information that 

they do not want to be publicly 

disclosed in their comment submission. 

Your request should state that the NRC 

does not routinely edit comment 

submissions to remove such information 

before making the comment 

submissions available to the public or 

entering the comment into ADAMS. 

II. Background 

Under 10 CFR part 50 and 10 CFR 

part 52, the NRC requires current and 

future holders of operating licenses and 

current and future holders of combined 

licenses, respectively, to comply with a 

variety of regulatory requirements 

related to decommissioning. This 

section discusses previous rules that set 

out the NRC’s requirements for 

production and utilization facility 

decommissioning and activities that 

have led to the development of this 

proposed rule. 

A. 1988 Decommissioning Rule 

On June 27, 1988, the NRC published 

a final rule titled, ‘‘General 

Requirements for Decommissioning 

Nuclear Facilities’’ (53 FR 24018) 

(referred to herein as the ‘‘1988 Final 

Rule’’), which established 

decommissioning requirements for 

various types of licensees. In this rule, 

the NRC amended its regulations to 

provide specific requirements for the 

decommissioning of nuclear facilities. 

Specifically, the final rule established 

regulations on acceptable 

decommissioning alternatives, planning 

for decommissioning, decommissioning 

timeliness, assurance of the availability 

of funds for decommissioning, and 

environmental review requirements 

related to decommissioning. The 1988 

Final Rule amended the regulations that 

applied to applicants and licensees 

under 10 CFR part 30, ‘‘Rules of General 

Applicability to Domestic Licensing of 

Byproduct Material’’; 10 CFR part 40, 

‘‘Domestic Licensing of Source 

Material’’; 10 CFR part 50; 10 CFR part 

70, ‘‘Domestic Licensing of Special 

Nuclear Material’’; and 10 CFR part 72, 

‘‘Licensing Requirements for the 

Independent Storage of Spent Nuclear 

Fuel, High-Level Radioactive Waste, and 

Reactor-Related Greater than Class C 

Waste.’’ 

In the 1988 Final Rule, the NRC 

defined decommissioning as the 

‘‘removal of nuclear facilities safely 

from service and reduction of residual 

radioactivity to a level that permits 

release of the property for unrestricted 

use and termination of the license.’’ The 

NRC also stated in the 1988 Final Rule 

that decommissioning activities do not 

include the removal and disposal of 

spent fuel, which is considered to be an 

operational activity, or the removal and 

disposal of nonradioactive structures 

and materials beyond that necessary to 

terminate the NRC license. 

The purpose of the 1988 Final Rule, 

in part, was to ensure that reactor 

decommissioning would be carried out 

with minimal impact on public and 

occupational health and safety and the 

environment. The NRC’s objective was 

that decommissioned facility sites 

would ultimately be available for 

unrestricted use for any public or 

private purpose. The amended 

regulations provided a regulatory 

framework for efficient and consistent 

licensing actions related to 

decommissioning. 

The NRC noted in the 1988 Final Rule 

that, although decommissioning was not 

an imminent health and safety problem, 

the number and complexity of facilities 

that would require decommissioning 


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1License termination based upon a facility 

meeting the unrestricted use criteria under 

§ 20.1402 is the most common license termination 

scenario. The NRC may also terminate a facility 

license under restricted conditions (§ 20.1403, 

‘‘Criteria for license termination under restricted 

conditions’’) and under alternative criteria 

(§ 20.1404, ‘‘Alternative criteria for license 

termination’’). 

was expected to increase, and 

inadequate or untimely consideration of 

decommissioning, specifically in the 

areas of planning and financial 

assurance, could result in significant 

adverse health, safety, and 

environmental impacts. The 1988 Final 

Rule clearly states that the licensee is 

responsible for the funding and 

completion of decommissioning in a 

manner that protects public health and 

safety. The NRC stated, ‘‘With the 

increased number of decommissionings 

expected, case-by-case procedures 

would make licensing difficult and 

increase NRC and licensee staff 

resources needed for these activities’’ 

(53 FR 24019). 

The 1988 Final Rule required that, 

within 2 years after a licensee 

permanently ceases operation of a 

licensed nuclear facility, the licensee 

must submit a detailed 

decommissioning plan to the NRC for 

approval along with a supplemental 

environmental report that addresses 

environmental issues that have not 

already been considered. Based on these 

submittals, the NRC reviewed the 

licensee’s planned activities, prepared a 

safety evaluation report and an 

environmental assessment (EA), and 

either made a finding of no significant 

impact (the usual case) or prepared an 

environmental impact statement. Upon 

approval of the decommissioning plan, 

the NRC issued an order under § 2.202, 

‘‘Orders,’’ permitting the licensee to 

decommission its facility in accordance 

with the approved plan. As part of the 

approval process for the 

decommissioning plan, the public had 

the opportunity to request a hearing 

under 10 CFR part 2, ‘‘Agency Rules of 

Practice and Procedure.’’ The NRC 

would terminate the license once the 

decommissioning process was 

completed and the NRC was satisfied 

that the facility had been radioactively 

decontaminated to an unrestricted 

release level under § 20.1402, 

‘‘Radiological criteria for unrestricted 

use.’’ 1

If the licensee chose to place the 

reactor in storage and dismantle it at a 

later time, the initial decommissioning 

plan submittal was not required to be as 

detailed as a plan for prompt 

dismantlement. However, before the 

licensee could begin dismantlement, the 

regulations required that the licensee 

submit a detailed plan and 

environmental report to the NRC for 

approval. Before the decommissioning 

plan was approved, the licensee could 

not perform any major decommissioning 

activities. If a licensee desired a 

reduction in requirements because of 

the permanent cessation of operations, it 

had to obtain a license amendment for 

possession-only status. This possessiononly license amendment was usually 

granted after the licensee indicated that 

the reactor had permanently ceased 

operations and that fuel had been 

permanently removed from the reactor 

vessel. Three examples of licensees that 

were granted possession-only status are 

Yankee Atomic Electric Company for 

the Yankee Nuclear Power Station 

(Yankee Rowe) (August 5, 1992; 

ADAMS Accession No. ML17283A069), 

Portland General Electric Company for 

the Trojan Nuclear Power Plant (May 5, 

1993; ADAMS Accession No. 

ML18095A126), and Sacramento 

Municipal Utility District for the 

Rancho Seco Nuclear Generating Station 

(March 17, 1992; ADAMS Accession No. 

ML17283A071). 

The 1988 Final Rule required 

licensees to provide assurance that, at 

any time during the life of the facility 

through termination of the license, 

adequate funds will be available to 

complete decommissioning. For 

operating reactors, the 1988 Final Rule 

prescribed the required amount of 

decommissioning funding in § 50.75. 

The 1988 Final Rule also imposed the 

requirement that, 5 years before license 

expiration or cessation of operations, 

licensees must submit a preliminary 

decommissioning plan containing a sitespecific decommissioning cost estimate 

and appropriately adjust the financial 

assurance mechanism. In addition, the 

1988 Final Rule required licensees to 

submit a decommissioning plan, 

including a site-specific cost estimate 

for decommissioning and a 

correspondingly adjusted financial 

assurance mechanism, within 2 years 

after permanent cessation of operations. 

For delayed dismantlement of a nuclear 

facility, the 1988 Final Rule required 

licensees to submit an updated 

decommissioning plan with the 

estimated cost covering the delay of 

decommissioning and to appropriately 

adjust the financial assurance 

mechanism. Before approval of the 

decommissioning plan, the 1988 Final 

Rule specified that licensee use of the 

decommissioning funds would be 

determined on a case-specific basis for 

premature closure, when the accrual of 

required decommissioning funds may 

be incomplete. 

B. 1996 Decommissioning Rule 

On July 29, 1996, the NRC amended 

its regulations for reactor 

decommissioning to clarify ambiguities, 

codify procedures that reduced 

regulatory burden, provide greater 

flexibility, and allow for greater public 

participation in the decommissioning 

process in a final rule titled, 

‘‘Decommissioning of Nuclear Power 

Reactors’’ (61 FR 39278) (referred to 

herein as the ‘‘1996 Final Rule’’). The 

1996 Final Rule made fundamental 

changes to nuclear power reactor 

decommissioning by streamlining the 

process and reducing both licensee and 

NRC resource expenditures while 

maintaining safety, protecting the 

environment, and encouraging public 

involvement. 

In the 1996 Final Rule, the NRC 

explained that the degree of regulatory 

oversight required for a nuclear power 

reactor in decommissioning is 

considerably less than that required for 

a facility during its operating stage. 

During the operating stage of the reactor, 

fuel in the reactor core undergoes a 

controlled nuclear fission reaction that 

generates a high neutron flux and large 

amounts of heat. Safe control of the 

nuclear reaction involves the use and 

operation of many complex systems. 

First, the nuclear reaction must be 

carefully controlled through neutronabsorbing mechanisms. Second, the heat 

generated must be removed so that the 

fuel and its supporting structure do not 

overheat. Third, the confining structure 

and ancillary systems must be 

maintained and degradation caused by 

radiation and mechanical and thermal 

stress ameliorated. Fourth, the 

radioactivity resulting from the nuclear 

reaction in the form of direct radiation 

(especially near the high neutron flux 

areas around the reactor vessel) and any 

radiologically contaminated materials 

and radiological effluents (gaseous and 

liquid) must be minimized and 

controlled. Moreover, proper operating 

procedures must be established and 

maintained, with appropriately trained 

staff to ensure that the reactor system is 

properly operated and maintained, and 

that operating personnel minimize their 

exposure to radiation when performing 

their duties. Finally, emergency 

response procedures must be 

established and maintained to protect 

the public in the event of an accident. 

Decommissioning of a nuclear power 

reactor begins when the nuclear fission 

reaction is stopped and the fuel (in the 

form of spent fuel assemblies) is 

permanently removed from the reactor 


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vessel and placed in the SFP until 

transferred to interim storage in an 

onsite ISFSI or transported offsite for 

storage or disposal. While the spent fuel 

is still highly radioactive and generates 

heat caused by radioactive decay, the 

fuel slowly cools as its energetic decay 

products diminish. The SFP, which 

contains circulating water, removes the 

decay heat and filters out any small 

radioactive contaminants escaping the 

spent fuel assemblies. The SFP system 

is relatively simple to operate and 

maintain compared to an operating 

nuclear power reactor. The remainder of 

the facility may contain radioactive 

contamination in areas that were 

directly impacted by reactor operation, 

and will be more highly contaminated 

in the area of the reactor vessel. 

However, no new radioactivity can be 

generated because the spent fuel is 

stored in a configuration that precludes 

the nuclear fission reaction. Once the 

nuclear fission process has permanently 

ceased and the fuel assemblies have 

been removed from the reactor vessel, 

safety concerns for an SFP are greatly 

reduced because there is no longer 

generation of large amounts of heat, 

high neutron flux and related materials 

degradation, and other related stresses 

that result from the functioning of an 

operating reactor system. 

Contaminated areas of the facility 

must still be controlled to minimize 

radiation exposure to personnel and 

control the spread of radioactive 

material. This situation is now similar 

to a contaminated materials facility and 

does not require the oversight that an 

operating reactor would require. 

The amendments issued in the 1996 

Final Rule provided licensees with 

simplicity and flexibility in 

implementing the decommissioning 

process, especially with regard to 

premature closure. The amendments 

clarified ambiguities in the regulations 

existing at the time, codified procedures 

and terminology that had been used in 

a number of specific cases, and 

increased opportunities for the public to 

become informed about the licensee’s 

decommissioning activities. The 

amendments established a level of NRC 

oversight commensurate with the level 

of safety concerns expected during 

decommissioning activities. 

Specifically, the 1996 Final Rule 

established or modified requirements 

with regard to initial decommissioning 

activities, major decommissioning 

activities, and license termination 

procedures. 

With regard to initial 

decommissioning activities, the 1996 

Final Rule mandated that, once a 

licensee permanently ceases operation 

of the nuclear power reactor and 

removes the fuel assemblies from the 

reactor vessel, it could not undertake 

any major decommissioning activities 

until it provided the public and the NRC 

with additional information about the 

proposed decommissioning approach. 

The NRC required that the licensee 

submit this information in the form of 

a PSDAR, which consists of the 

licensee’s proposed decommissioning 

activities and schedule through license 

termination, a discussion of the reasons 

for concluding that the environmental 

impacts associated with the proposed 

site-specific decommissioning activities 

will be bounded by appropriate 

previously issued environmental impact 

statements, and a decommissioning cost 

estimate for the proposed activities. The 

NRC makes the PSDAR available to the 

public for comment and holds a public 

meeting concerning the PSDAR in the 

vicinity of the plant. The NRC, however, 

does not approve the PSDAR and the 

submission of the PSDAR and its review 

by the NRC does not require the licensee 

to request a license amendment or any 

other approval. 

The 1996 Final Rule also established 

that the licensee may not begin 

performing major decommissioning 

activities until 90 days after the NRC 

receives the PSDAR submittal and until 

the licensee submits the certifications 

under § 50.82(a)(1) that operations have 

permanently ceased and that fuel has 

been permanently removed from the 

reactor vessel. The 1996 Final Rule also 

amended certain 10 CFR part 50 

technical requirements to cover the 

transition of the facility from operating 

to permanently shutdown status. 

Specifically, the 1996 Final Rule 

removed the requirement for a licensee 

that has permanently ceased operations 

and removed fuel from the reactor 

vessel to obtain a license amendment 

before proceeding with certain 

decommissioning activities within 

established regulatory constraints (i.e., 

in accordance with § 50.59, ‘‘Changes, 

tests and experiments’’). These changes 

to the decommissioning requirements 

increased the flexibility in the type of 

actions that licensees could undertake 

without prior NRC approval. 

With regard to major 

decommissioning activities, the 1996 

Final Rule implemented a major change 

from the 1988 Final Rule in that nuclear 

power reactor licensees would no longer 

be required to have an approved 

decommissioning plan before being 

permitted to perform major 

decommissioning activities. The 1996 

Final Rule allowed licensees to perform 

activities that meet the criteria in 

§ 50.59, which the NRC amended to 

include additional criteria to ensure that 

licensees consider concerns specific to 

decommissioning. Based on NRC 

experience with licensee 

decommissioning activities at the time, 

the NRC recognized that the § 50.59 

process used by the licensee during 

reactor operations encompassed routine 

activities that were similar to those 

undertaken during the decommissioning 

process. The NRC concluded that the 

licensee could use the § 50.59 process to 

perform major decommissioning 

activities if licensing conditions and the 

level of NRC oversight required during 

reactor operations continued during 

decommissioning, commensurate with 

the risk profile of the facility being 

decommissioned. The 1996 Final Rule 

also required the licensee to provide 

written notification to the NRC before 

performing any decommissioning 

activity that is inconsistent with, or 

makes significant schedule changes 

from, the actions and schedules 

described in the PSDAR. 

With regard to license termination, 

the 1996 Final Rule required that a 

licensee wishing to terminate its license 

submit a license termination plan for 

NRC approval. The approval process for 

the termination plan provides for a 

hearing opportunity under 10 CFR part 

2. The licensee must submit a 

supplemental environmental report that 

considers new and significant 

environmental changes associated with 

license termination activities. The 1996 

Final Rule imposed an additional 

requirement for the purpose of keeping 

the public informed. A public meeting, 

similar to the one held after the PSDAR 

submittal, must take place after the 

licensee submits its license termination 

plan to the NRC. 

The 1996 Final Rule continued the 

same degree of decommissioning 

financial assurance that was previously 

required but provided more flexibility 

by allowing licensees to have limited, 

early use of decommissioning funds. 

The NRC presented this provision in a 

February 3, 1994, draft policy statement 

titled, ‘‘Use of Decommissioning Trust 

Funds before Decommissioning Plan 

Approval’’ (59 FR 5216), which was 

published for comment and eventually 

incorporated into the 1996 Final Rule. 

Before issuance of the 1996 Final Rule, 

licensee use of these funds was 

determined on a case-specific basis for 

prematurely shutdown plants. However, 

the 1996 Final Rule eliminated the 

requirement for a decommissioning plan 

and instead required a PSDAR 

submittal, which requires a 

decommissioning cost estimate. The 

1996 Final Rule permitted 3 percent of 

the decommissioning funds generically 


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required by § 50.75 to be available to the 

licensee for decommissioning planning 

purposes. Moreover, to allow the 

licensee to accomplish major 

decommissioning activities promptly, 

an additional 20 percent of the generic 

funding amount would be made 

available 90 days after the NRC had 

received the PSDAR if the licensee had 

also submitted the certifications 

required by § 50.82(a)(1). The use of any 

funds above those amounts required the 

licensee to submit a site-specific 

decommissioning cost estimate to the 

NRC prior to the use of those funds. 

C. Post-1996 Final Rule 

Decommissioning Activity 

In a series of Commission papers 

issued between 1997 and 2001, the NRC 

staff provided options and 

recommendations to the Commission to 

address regulatory improvements 

related to nuclear power reactor 

decommissioning. To consolidate these 

recommendations, in the Staff 

Requirements Memorandum (SRM) for 

SECY–99–168, ‘‘Staff Requirements— 

SECY–99–168—Improving 

Decommissioning Regulations for 

Nuclear Power Plants,’’ dated December 

21, 1999 (ADAMS Accession No. 

ML003752190), the Commission 

directed the NRC staff to proceed with 

a single, integrated, and risk-informed 

decommissioning rule addressing the 

areas of EP, insurance, safeguards, 

staffing and training, and backfitting for 

decommissioning nuclear power 

reactors. The objective of the 

rulemaking was to clarify and remove 

certain regulations for decommissioning 

nuclear power reactors based in large 

part on the reduction in radiological risk 

compared to operating reactors. 

On June 28, 2000, the NRC staff 

submitted SECY–00–0145, ‘‘Integrated 

Rulemaking Plan for Nuclear Power 

Plant Decommissioning,’’ to the 

Commission (ADAMS Accession No. 

ML003721626). In this paper, the NRC 

staff proposed an integrated 

decommissioning rulemaking plan and 

requested Commission approval to 

proceed with developing an integrated 

rulemaking for nuclear power plant 

decommissioning in accordance with 

the recommendations detailed in the 

rulemaking plan. The paper addressed 

the regulatory areas of EP, insurance, 

safeguards, staffing and training, and 

backfitting for decommissioning nuclear 

power reactors. The rulemaking plan 

was contingent on the completion of a 

SFP zirconium fire risk study. The 

Commission responded to SECY–00– 

0145 in an SRM dated September 27, 

2000 (ADAMS Accession No. 

ML003754381). The Commission 

returned that SECY to the staff without 

a vote on the rulemaking plan pending 

further developments in the area and 

requested that the staff submit a revised 

paper to the Commission. 

D. Spent Fuel Pool Studies 

In the late 1990s and early 2000s, the 

NRC was assessing the risk of an SFP 

accident at a nuclear power reactor site 

in decommissioning. Following the 

removal of spent fuel from the reactor, 

the principal radiological risks are 

associated with the storage of spent fuel 

on site. Generally, a few months after 

the reactor has been permanently shut 

down and defueled, there are no 

possible design-basis accidents that 

could result in a radiological release 

exceeding the limits established by the 

U.S. Environmental Protection Agency 

(EPA) early-phase Protective Action 

Guides (PAGs) at the exclusion area 

boundary (EPA–400–R–92–001, 

‘‘Manual of Protective Action Guides 

And Protective Actions For Nuclear 

Incidents,’’ issued May 1992, and final 

revision EPA–400/R–17/001, ‘‘PAG 

Manual: Protective Action Guides and 

Planning Guidance for Radiological 

Incidents,’’ issued January 2017). The 

only SFP accident scenario that might 

lead to a release with offsite 

consequences exceeding the PAGs at a 

decommissioning reactor is a zirconium 

fire. The zirconium fire scenario is a 

postulated, but highly unlikely, beyonddesign-basis accident scenario that 

involves a major loss of water inventory 

from the SFP, resulting in a significant 

heatup of the spent fuel, and 

culminating in substantial zirconium 

cladding oxidation, fire, and fuel 

damage. The significance of spent fuel 

heatup scenarios that might result in a 

zirconium fire depends on the decay 

heat of the irradiated fuel stored in the 

SFP. Therefore, the probability of a 

zirconium fire scenario continues to 

decrease as a function of the time that 

the decommissioning reactor has been 

permanently shut down and defueled. 

In the 1980s, the NRC examined the 

risk of an SFP accident as Generic 

Safety Issue 82, ‘‘Beyond Design Basis 

Accidents in Spent Fuel Pools,’’ because 

of the increased use of high-density 

storage racks and laboratory studies that 

indicated the possibility of a zirconium 

fire spreading between assemblies in an 

air-cooled environment (see Section 3 of 

NUREG–0933, ‘‘Resolution of Generic 

Safety Issues,’’ issued December 2011 

(available at https://www.nrc.gov/ 

sr0933/Section%203.%20New

%20Generic%20Issues/082r3.html)). 

The risk assessment and cost benefit 

analyses developed through this effort 

(Section 6.2 of NUREG–1353, 

‘‘Regulatory Analysis for the Resolution 

of Generic Issue 82, ‘Beyond Design 

Basis Accidents in Spent Fuel Pools,’ ’’ 

issued April 1989 (ADAMS Accession 

No. ML082330232)) concluded that the 

risk of a severe accident in the SFP was 

low and appeared to meet the public 

health objectives of the Commission’s 

Safety Goal Policy Statement (51 FR 

30028; August 21, 1986) and that no 

new regulatory requirements were 

warranted. 

To support the rulemaking for 

decommissioning nuclear power plants 

in the late 1990s, the NRC reevaluated 

the risk of an SFP accident. The NRC’s 

assessment in NUREG–1738, ‘‘Technical 

Study of Spent Fuel Pool Accident Risk 

at Decommissioning Nuclear Power 

Plants,’’ issued February 2001 (ADAMS 

Accession No. ML010430066), 

conservatively assumed that if the water 

level in the SFP dropped below the top 

of the spent fuel, an SFP zirconium fire 

involving all of the spent fuel would 

occur and thereby bounded those 

conditions associated with air cooling of 

the fuel (including partial draindown 

scenarios) and fire propagation. Even 

with this conservative assumption, the 

study found the risk of an SFP fire to be 

low and well within the Commission’s 

safety goals. 

Although NUREG–1738 did not 

completely rule out the possibility of a 

zirconium fire, it did demonstrate that 

storage of spent fuel in a high-density 

configuration in SFPs is safe and that 

the risk of accidental release of a 

significant amount of radioactive 

material to the environment is low. The 

study used simplified and sometimes 

bounding assumptions and models to 

characterize the likelihood and 

consequences of beyond-design-basis 

SFP accidents. Subsequent NRC 

regulatory activities and studies 

(described in more detail in this section) 

have reaffirmed the safety and security 

of spent fuel stored in pools and have 

demonstrated that SFPs are effectively 

designed to prevent accidents and 

minimize damage from malevolent 

attacks. 

In the wake of the terrorist attacks of 

September 11, 2001, the NRC took 

several actions to further reduce the 

possibility of an SFP fire. The NRC 

issued immediately effective nonpublic 

orders (see the cover letter at ADAMS 

Accession No. ML020510637) that 

required licensees to implement 

additional security measures, including 

increased patrols, augmented security 

forces and capabilities, and more 

restrictive site-access controls to reduce 

the likelihood of an SFP accident 

resulting from a terrorist-initiated event. 

A memorandum to the Commission 


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titled, ‘‘Documentation of Evolution of 

Security Requirements at Commercial 

Nuclear Power Plants with Respect to 

Mitigation Measures for Large Fires and 

Explosions,’’ dated February 4, 2010 

(ADAMS Accession No. ML092990438), 

provides a comprehensive discussion of 

these actions, some of which 

specifically address SFP safety and 

security. 

New requirements to mitigate a 

postulated loss of SFP water inventory 

were also implemented following the 

terrorist attacks of September 11, 2001; 

these requirements resulted in enhanced 

spent fuel coolability and the potential 

to recover SFP water level and cooling 

prior to a postulated SFP zirconium fire. 

Based on the implementation of these 

additional strategies, the probability 

and, accordingly, the risk to the public 

health and safety of an SFP zirconium 

fire scenario has decreased and is 

expected to be less than previously 

analyzed in NUREG–1738 and previous 

studies. 

After the events of September 11, 

2001, the NRC also addressed by order 

the issue of potential aircraft impacts to 

the SFP by requiring licensees to have 

in place mitigating strategies for large 

fires or explosions at nuclear power 

plants. The Nuclear Energy Institute 

(NEI) provided detailed guidance in NEI 

06–12, Revision 2, ‘‘B.5.b Phase 2 & 3 

Submittal Guideline,’’ dated December 

2006 (ADAMS Accession No. 

ML070090060). The NRC found this 

guidance acceptable for use as 

documented in NUREG–0800, 

‘‘Standard Review Plan for the Review 

of Safety Analysis Reports for Nuclear 

Power Plants: LWR Edition,’’ Section 

19.4, ‘‘Strategies and Guidance to 

Address Loss of Large Areas of the Plant 

Due to Explosions and Fires,’’ Revision 

0, dated June 2015 (ADAMS Accession 

No. ML13316B202). The NRC’s issuance 

of the final rule titled, ‘‘Power Reactor 

Security Requirements,’’ on March 27, 

2009 (74 FR 13926), made the 

requirements of the order generically 

applicable. In that final rule, the NRC 

added § 50.54(hh)(2) to require licensees 

to develop and implement guidance and 

strategies to, among other things, 

maintain or restore SFP cooling 

capability in the event of loss of large 

areas of the plant resulting from fires or 

explosions, which further decreases the 

probability of an SFP fire. 

Under § 50.54(hh)(2), nuclear power 

reactor licensees are required to 

implement strategies such as those 

provided in NEI 06–12. The NEI 

guidance specifies that portable, power 

independent pumping capabilities must 

be able to provide at least 500 gallons 

per minute of bulk water makeup to the 

SFP and at least 200 gallons per minute 

of water spray to the SFP. Recognizing 

that the SFP is more susceptible to a 

release when the spent fuel is in a 

nondispersed configuration (i.e., fuel 

assemblies with more decay heat are not 

dispersed among fuel assemblies with 

less decay heat), the guidance also 

specifies that the portable equipment 

should be capable of being deployed 

within 2 hours for a nondispersed 

configuration. 

Further, other organizations, such as 

Sandia National Laboratories (SNL), 

have confirmed the effectiveness of the 

additional mitigation strategies to 

maintain spent fuel cooling in the event 

that the pool is damaged and its initial 

water inventory is reduced or lost 

entirely. The analyses conducted by 

SNL (collectively referred to as the 

‘‘Sandia studies’’) are sensitive securityrelated information and are not 

available to the public. The Sandia 

studies considered spent fuel loading 

patterns and other aspects of a 

pressurized water reactor SFP and a 

boiling water reactor SFP, including the 

role that the circulation of air plays in 

the cooling of spent fuel when there is 

a partial or complete loss of water. The 

Sandia studies indicated that there is a 

significant amount of time between the 

initiating event (i.e., the event that 

causes the SFP water level to drop) and 

the point at which the spent fuel 

assemblies become partially or 

completely uncovered. In addition, the 

Sandia studies indicated that for those 

hypothetical conditions in which air 

cooling may not be effective in 

preventing a zirconium fire, there is a 

significant amount of time between the 

spent fuel becoming uncovered and the 

possible onset of such a zirconium fire, 

thereby providing a substantial 

opportunity for event mitigation. The 

Sandia studies, which account for 

relevant heat transfer and fluid flow 

mechanisms, also indicated that air 

cooling spent fuel could be sufficient to 

prevent SFP zirconium fires at a point 

much earlier following fuel offload from 

the reactor than previously considered 

in NUREG–1738. 

In NUREG–2161, ‘‘Consequence 

Study of a Beyond-Design-Basis 

Earthquake Affecting the Spent Fuel 

Pool for a U.S. Mark I Boiling Water 

Reactor,’’ issued September 2014 

(ADAMS Accession No. ML14255A365), 

the NRC evaluated the potential benefits 

of strategies required in § 50.54(hh)(2). 

The report explains that successful 

implementation of mitigation strategies 

significantly reduces the likelihood of a 

release from the SFP in the event of a 

loss of cooling water. Additionally, the 

NRC found that the placement of spent 

fuel in a dispersed configuration in the 

SFP would have a positive effect in 

promoting natural circulation, which 

enhances air coolability and thereby 

reduces the likelihood of a release from 

a completely drained SFP. The NRC 

issued Information Notice 2014–14, 

‘‘Potential Safety Enhancements to 

Spent Fuel Pool Storage,’’ dated 

November 14, 2014 (ADAMS Accession 

No. ML14218A493), to all nuclear 

power reactor and ISFSI licensees to 

inform them of the insights from 

NUREG–2161. This information notice 

describes the benefits of storing spent 

fuel in more favorable configurations, 

placing spent fuel in dispersed patterns 

immediately after core offload, and 

taking action to improve mitigation 

strategies. 

In 2013, the NRC documented a 

regulatory analysis in COMSECY–13– 

0030, ‘‘Staff Evaluation and 

Recommendation for Japan Lessons 

Learned Tier 3 Issue on Expedited 

Transfer of Spent Fuel’’ (ADAMS 

Accession No. ML13329A918), which 

considered a broad history of the NRC’s 

oversight of spent fuel storage and SFP 

operating experience (domestic and 

international) and relied on information 

compiled in NUREG–2161. In 

COMSECY–13–0030, the NRC staff 

concluded that SFPs are robust 

structures with large safety margins and 

recommended to the Commission that 

further regulatory actions to require the 

expedited transfer of spent fuel from 

SFPs to dry cask storage were not 

warranted. The Commission 

subsequently approved the staff’s 

recommendation in SRM–COMSECY– 

13–0030, dated May 23, 2014 (ADAMS 

Accession No. ML14143A360). 

In addition, in response to the 

Fukushima Dai-ichi accident, the NRC 

implemented additional regulatory 

actions to further enhance reactor and 

SFP safety. On March 12, 2012, the NRC 

issued two orders: Order EA–12–051, 

‘‘Order Modifying Licenses with Regard 

to Reliable Spent Fuel Pool 

Instrumentation’’ (ADAMS Accession 

No. ML12054A679), and Order EA–12– 

049, ‘‘Order Modifying Licenses with 

Regard to Requirements for Mitigation 

Strategies for Beyond-Design-Basis 

External Events’’ (ADAMS Accession 

No. ML12054A735). Order EA–12–051 

required licensees to install reliable 

means of remotely monitoring widerange SFP levels to support effective 

prioritization of event mitigation and 

recovery actions in the event of a 

beyond-design—basis external event. 

Although the primary purpose of the 

order was to ensure that operators were 

not distracted by uncertainties related to 

SFP conditions during the accident 


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response, the improved monitoring 

capabilities would help in the diagnosis 

and response to potential losses of SFP 

integrity. Order EA–12–049 required 

licensees to, among other actions, 

develop, implement, and maintain 

guidance and strategies to maintain or 

restore SFP cooling capabilities 

independent of normal alternating 

current power systems following a 

beyond-design-basis external event. 

Further, the NRC issued the Mitigation 

of Beyond-Design-Basis Events final rule 

on August 9, 2019 (84 FR 39684), which 

made these two orders generically 

applicable and moved the requirements 

of § 50.54(hh)(2) to paragraph (b)(2) of 

the new § 50.155, ‘‘Mitigation of 

beyond-design-basis events.’’ These 

requirements ensure that a more reliable 

and robust mitigation capability is in 

place to address degrading conditions in 

SFPs resulting from certain significant, 

but unlikely, events. 

The additional mitigation strategies 

implemented after the terrorist attacks 

of September 11, 2001, such as the 

issuance of § 50.54(hh)(2) (now 

§ 50.155(b)(2)) and the NRC’s review 

and approval of NEI 06–12, and the 

issuance of Orders EA–12–049 and EA– 

12–051, made generically applicable as 

§ 50.155(b)(1) and § 50.155(e), following 

the Fukushima Dai-ichi accident 

enhance spent fuel coolability and the 

potential to recover SFP water level and 

cooling before the initiation of a 

potential SFP zirconium fire. The 

Sandia studies also confirmed the 

effectiveness of additional mitigation 

strategies to maintain spent fuel cooling 

in the event that the pool is drained. 

Based on this information and the 

implementation of additional strategies, 

the probability of an SFP zirconium fire 

initiation in a draindown event is 

expected to be less than that reported in 

NUREG–1738 and previous studies and 

therefore well within the Commission’s 

expressed safety goals, as described 

previously. 

E. Changes in Nuclear Power Reactor 

Decommissioning at the NRC and 

Within the Nuclear Power Industry 

On June 4, 2001, the NRC staff 

submitted SECY–01–0100, ‘‘Policy 

Issues Related to Safeguards, Insurance, 

and Emergency Preparedness 

Regulations at Decommissioning 

Nuclear Power Plants Storing Fuel in 

Spent Fuel Pools’’ (ADAMS Accession 

No. ML011450420), to the Commission. 

Before the Commission responded to 

SECY–01–0100, the terrorist attacks of 

September 11, 2001, occurred. Given the 

security implications of those events 

and the results of the NUREG–1738 

zirconium fire risk study that showed 

the risk of an SFP fire to be low and well 

within the Commission’s safety goals, 

the NRC later redirected its rulemaking 

priorities and resources to focus on 

programmatic regulatory changes 

related to safeguards and security. In a 

memorandum to the Commission titled, 

‘‘Status of Regulatory Exemptions for 

Decommissioning Plants,’’ dated August 

16, 2002 (ADAMS Accession No. 

ML030550706), the NRC staff justified 

this redirection in part by observing that 

no additional permanent nuclear power 

reactor shutdowns were anticipated in 

the foreseeable future and that no 

immediate need existed to proceed with 

the decommissioning regulatory 

improvement work that was planned. 

The NRC staff concluded that, if any 

additional nuclear power reactors 

permanently shut down after the 

rulemaking effort was suspended, 

establishment of the decommissioning 

regulatory framework would continue to 

be addressed for each facility through 

the license amendment and exemption 

processes. 

Between 1998 and 2013, no nuclear 

power reactors permanently ceased 

operation. Between 2013 and 2021, 

however, 12 nuclear power reactors 

permanently shut down, defueled, and 

entered decommissioning. Notably, in 

2013, four nuclear power reactor units 

permanently shut down without 

significant advance notice or 

preplanning: Crystal River Unit 3 

Nuclear Generating Plant (Duke Energy 

Florida); Kewaunee Power Station 

(Dominion Energy); and San Onofre 

Nuclear Generating Station (SONGS), 

Units 2 and 3 (Southern California 

Edison). In addition, on December 29, 

2014, Entergy Nuclear Operations, Inc. 

(Entergy) permanently ceased 

operations at the Vermont Yankee 

Nuclear Power Station (VY); on October 

24, 2016, the Omaha Public Power 

District permanently ceased operations 

at Fort Calhoun Station, Unit 1; on 

September 17, 2018, Exelon Generation 

Company, LLC (Exelon) permanently 

ceased operations at Oyster Creek 

Nuclear Generating Station; on May 31, 

2019, Entergy permanently ceased 

operations at Pilgrim Nuclear Power 

Station; on September 20, 2019, Exelon 

permanently ceased operations at Three 

Mile Island, Unit 1; on April 30, 2020, 

and April 30, 2021, respectively, 

Entergy permanently ceased operations 

at Indian Point Nuclear Generating, Unit 

Nos. 2 and 3; and on August 10, 2020, 

NextEra Energy Duane Arnold, LLC 

(NextEra) permanently ceased 

operations of Duane Arnold Energy 

Center. Licensees have also announced 

plans for additional near-term 

permanent shutdowns, including 

Palisades Nuclear Plant (Entergy) and 

Diablo Canyon Power Plant, Units 1 and 

2 (Pacific Gas & Electric Co.). 

Decommissioning reactor licensees 

and the NRC have expended substantial 

resources processing licensing actions 

for nuclear power reactors during their 

transition period to decommissioning 

status. Consistent with the nuclear 

power reactors that permanently shut 

down in the 1990s, the licensees that are 

currently transitioning to 

decommissioning have been requesting 

NRC review and approval of licensing 

actions, informed by the low risk of an 

offsite radiological release posed by a 

decommissioning reactor. Specifically, 

the licensees are seeking NRC approvals 

of exemptions from requirements and 

license amendments to reflect the 

reduced operations and radiological 

risks posed by a permanently shutdown 

and defueled nuclear power reactor. 

F. Decommissioning Lessons Learned 

Report 

In October 2016, the NRC published 

the ‘‘Power Reactor Transition from 

Operations to Decommissioning: 

Lessons Learned Report’’ (ADAMS 

Accession No. ML16085A029). The 

report documents the lessons learned by 

the NRC and stakeholders associated 

with permanent nuclear power reactor 

shutdowns during the period from 2013 

to 2016. In particular, the report focuses 

on the transition from reactor operations 

to decommissioning for Kewaunee, 

Crystal River Unit 3, SONGS Units 2 

and 3, and VY. The transition process 

includes the NRC’s review and approval 

of certain requests for exemptions from 

the NRC’s regulations and for license 

amendments to modify the operating 

reactors’ licensing bases to reflect those 

of decommissioning reactors. After 

these actions are complete, the NRC 

then transfers the project management 

and oversight responsibility from its 

Office of Nuclear Reactor Regulation to 

its Office of Nuclear Material Safety and 

Safeguards (NMSS). Project 

management support is provided by 

NMSS for these decommissioning 

reactors until license termination. The 

report also provides a number of best 

practices identified from recent 

experience with reactor shutdowns and 

the transition to decommissioning. 

The report highlights some of the 

challenges experienced by the NRC 

during the decommissioning transition 

licensing reviews from 2013 to 2016 and 

the NRC’s actions to address those 

challenges. The report also discusses 

external stakeholders’ interest in the 

NRC’s review of the decommissioning 

transition licensing activities, especially 


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2Additional information about the existing 

options for decommissioning is available in 

NUREG/BR–0521, Rev. 1, ‘‘Decommissioning 

Nuclear Power Plants,’’ dated June 2017 (ADAMS 

Accession No. ML17177A253). 

3At the time of publication of the regulatory 

basis, the rulemaking title was ‘‘Regulatory 

Improvements for Power Reactors Transitioning to 

Decommissioning.’’ During the development of the 

proposed rule, the scope of the rulemaking 

expanded to include all production and utilization 

facilities licensed under 10 CFR parts 50 and 52. In 

order to reflect this change, the NRC has changed 

the title of the rulemaking to ‘‘Regulatory 

Improvements for Production and Utilization 

Facilities Transitioning to Decommissioning.’’ 

those associated with SONGS Units 2 

and 3 and VY, as represented by 

requests for hearings, public meetings, 

and questions to the NRC staff. 

In addition to the lessons learned and 

best practices, the report provides 

detailed project management guidance, 

recommendations, and documentation 

of precedent related to the reviews and 

evaluations specific to the types of 

licensing actions that the NRC expects 

to be processed during the 

decommissioning transition period, 

including oversight activities and 

communications. The NRC considered 

many of the lessons learned and 

recommendations described in this 

report during the development of this 

proposed rule. 

G. Initiation of This Proposed Rule 

In light of the number of licensees 

deciding to permanently shut down 

their nuclear power reactors, the 

Commission directed the NRC staff to 

proceed with an integrated rulemaking 

on nuclear power reactor 

decommissioning in an SRM dated 

December 30, 2014 (ADAMS Accession 

No. ML14364A111), associated with 

SECY–14–0118, ‘‘Request by Duke 

Energy Florida, Inc., for Exemptions 

from Certain Emergency Planning 

Requirements,’’ dated October 29, 2014 

(ADAMS Accession No. ML14219A444). 

The Commission further stated that this 

rulemaking should address: 

• Issues discussed in SECY–00–0145 

such as the graded approach to EP; 

• Lessons learned from the plants that 

have already gone through (or are 

currently going through) the 

decommissioning process; 

• The advisability of requiring a 

licensee’s PSDAR to be approved by the 

NRC; 

• The appropriateness of maintaining 

the three existing options for 

decommissioning (DECON, SAFSTOR, 

and ENTOMB) 2 and the timeframes 

associated with those options; 

• The appropriate role of State and 

local governments and nongovernmental stakeholders in the 

decommissioning process; and 

• Any other issues deemed relevant 

by the NRC staff. 

In SECY–15–0014, ‘‘Anticipated 

Schedule and Estimated Resources for a 

Power Reactor Decommissioning 

Rulemaking,’’ dated January 30, 2015 

(ADAMS Accession No. ML15082A089, 

redacted), the NRC staff committed to 

proceed with a rulemaking on nuclear 

power reactor decommissioning and 

provided an anticipated schedule and 

estimate of the resources required for 

the completion of a decommissioning 

rulemaking. 

H. Advance Notice of Proposed 

Rulemaking 

To begin the nuclear power reactor 

decommissioning rulemaking process, 

the NRC published an advance notice of 

proposed rulemaking (ANPR) in the 

Federal Register on November 19, 2015 

(80 FR 72358). In the ANPR, the NRC 

sought public comment on specific 

questions and issues with respect to 

possible revisions of the NRC’s 

decommissioning requirements. The 

NRC staff considered the comments 

received on the ANPR in its formulation 

of a draft regulatory basis for further 

regulatory action. Section 5 of the draft 

regulatory basis (ADAMS Accession No. 

ML17047A413) summarizes the public 

comments received on the ANPR. 

I. Regulatory Basis 

The NRC published the draft 

regulatory basis in the Federal Register 

on March 15, 2017 (82 FR 13778). In the 

draft regulatory basis, the NRC staff 

presented draft recommendations for 

amendments to the NRC’s regulations 

and guidance development to provide 

regulatory improvements for nuclear 

power reactors transitioning to 

decommissioning. The NRC requested 

public comment on these 

recommendations and asked specific 

questions regarding other possible 

revisions of the NRC’s requirements. In 

addition, the NRC published a 

preliminary draft regulatory analysis on 

May 9, 2017 (82 FR 21481). The NRC 

held a public meeting from May 8–10, 

2017, to discuss the draft regulatory 

basis and the associated preliminary 

draft regulatory analysis and issued a 

summary of the meeting on November 

15, 2017 (ADAMS Accession No. 

ML17157B211). 

The NRC received 40 public comment 

submissions on the draft regulatory 

basis and preliminary draft regulatory 

analysis, which it considered in its 

formulation of the revised regulatory 

basis. The NRC published a Federal 

Register notice announcing the public 

availability of the regulatory basis on 

November 27, 2017 (82 FR 55954).3

III. Discussion 

A. Current Regulatory Process 

Decommissioning requirements for 

production and utilization facilities are 

codified in §§ 50.82 and 52.110. 

Associated decommissioning funding 

requirements are codified in §§ 50.75, 

50.82, and 52.110. A nuclear power 

reactor licensee formally begins the 

decommissioning process when it 

certifies its permanent cessation of 

operations and permanent removal of 

fuel from the reactor vessel under 

§§ 50.82(a)(1) or 52.110(a). Once the 

NRC dockets these certifications, under 

§ 50.82(a)(2) or § 52.110(b), the 10 CFR 

part 50 or 10 CFR part 52 license no 

longer authorizes operation of the 

reactor or emplacement or retention of 

fuel in the reactor vessel. Despite this 

withdrawal of authority to operate the 

reactor, a decommissioning nuclear 

power plant continues to retain a 

license under 10 CFR part 50 or 10 CFR 

part 52. For this reason, the 

decommissioning plant continues to be 

subject to many of the requirements that 

apply to plants authorized to operate 

under 10 CFR part 50 or 10 CFR part 52. 

Regulations that are designed to 

protect the public against reactor 

operation related design-basis events 

that include conditions of normal 

operation, anticipated operational 

occurrences, and design-basis accidents 

(DBAs) are no longer applicable at a 

permanently shutdown and defueled 

reactor. For example, certain accident 

sequences for a nuclear power reactor 

that is operating, such as loss of coolant 

accidents and anticipated transients 

without scram, are no longer relevant to 

a permanently shutdown and defueled 

reactor. In addition, some regulations 

may not be relevant to certain SSCs 

because the SSCs are no longer required 

to be maintained, to operate, or to 

mitigate certain accidents, events, or 

transients, regardless of whether they 

are safety-related or security-related 

SSCs. Other regulations, although based 

on power operation of the plant, may 

continue to be applicable to the 

permanently defueled facility for a 

limited time, such as the standards for 

offsite radiological emergency 

preparedness (REP) plans under 10 CFR 

part 50 or 10 CFR part 52. Typically, the 

scope of NRC requirements can be 

reduced to those regulations and 

requirements that primarily pertain to 

the safe storage of the spent fuel in the 

SFP, as described in the site’s final 

safety analysis report (FSAR). 


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Upon permanent cessation of reactor 

operations and removal of fuel from the 

reactor vessel, the licensee is likely to 

submit a significant number of licensing 

actions (license amendment and 

exemption requests) to the NRC for 

review and approval based primarily on 

the reduced radiological risk to public 

health and safety. As discussed 

previously in this document, the types 

of potential accidents at 

decommissioning reactors are fewer, 

and the risks of radiological releases are 

reduced, when compared to those at an 

operating reactor. Therefore, to reflect 

this reduction in risk, licensees of 

decommissioning reactors typically 

request certain amendments to their 

licenses and certain exemptions from 

the NRC’s regulations. These licensing 

actions, which are processed by the 

NRC during licensees’ transition from 

operating to decommissioning status, 

establish the regulatory framework for 

reactors that have permanently shut 

down and defueled. 

For non-power reactor facilities, 

§ 50.82(b) requires that the licensee 

apply for license termination within two 

years following permanent cessation of 

operation. Each application for 

termination of a license must be 

accompanied, or preceded, by a 

proposed decommissioning plan (DP). 

In addition to the DP required by 

§ 50.82, § 50.75(f)(4) requires each 

licensee to submit a preliminary DP. 

The preliminary DP must be submitted 

at or about 2 years before the projected 

end of operation. In addition to the DP, 

§ 51.53(d) requires each applicant for a 

license amendment approving a DP to 

submit a supplement to its 

environmental report (ER). 

The decommissioning process for 

non-power reactor licensees begins with 

the removal of fuel as soon as possible 

after reactor operations permanently 

cease and the shipment of the fuel 

offsite in accordance with the U.S. 

Department of Energy, NRC, and U.S. 

Department of Transportation 

regulations. Under some circumstances, 

the licensee can apply for a possessiononly license amendment under § 50.90, 

‘‘Application for amendment of license, 

construction permit, or early site 

permit,’’ after operations have ended 

and before decommissioning starts. The 

possession-only license amendment 

limits the licensee’s authority to 

possessing specific nuclear material but 

does not authorize its use or the 

operation of a nuclear facility. If 

granted, a possession-only license 

amendment provides regulatory relief 

from the license and technical 

specification (TS) requirements for a 

non-power reactor in decommissioning. 

Further, the possession-only 

amendment permits the licensee to 

retain the facility, related radioactive 

byproduct material, and, in some cases, 

special nuclear material, pending 

approval of the DP. 

In addition to requesting license 

amendments and exemptions, nuclear 

power reactor licensees can make 

certain changes without prior NRC 

approval if the changes are permitted by 

an NRC regulation. Licensees primarily 

use an evaluation process with criteria 

in § 50.59 to make changes in a facility 

(or procedures) as described in the 

FSAR (as updated), including changes 

to the PSDAR, without prior NRC 

approval. The licensee’s updated FSAR 

should reflect changes to the 

decommissioning design-basis analyses, 

SSCs, and the licensee’s organizations, 

processes, and procedures. Licensees 

can also make changes without prior 

NRC approval as described in § 50.54(p) 

and § 50.54(q). In the case of non-power 

reactor facilities, the DP, which is put 

into effect with an order, provides for 

accommodation of any necessary 

changes in the DP and procedures 

through a process similar to the one in 

§ 50.59. 

The timing and implementation for 

some decommissioning licensing 

actions rely on an approach that 

recognizes the reduction in radiological 

risk after permanent cessation of power 

operation and removal of fuel from the 

reactor vessel. These risk reductions can 

be tied to several factors, including, but 

not limited to: (1) Reduction of the 

radiological source term after cessation 

of power operation and removal of fuel 

from the reactor vessel, (2) elapsed time 

after permanent shutdown, and (3) type 

of long-term onsite fuel storage. The two 

areas where these additional risk 

reductions are considered in the early 

decommissioning process are EP and 

facility insurance and indemnity. The 

NRC will not approve exemptions from 

EP and insurance coverage requirements 

until analyses confirm that there are no 

DBAs that would require protective 

actions for the public resulting from a 

release of radioactive material with a 

dose exceeding the EPA’s PAGs at the 

exclusion area boundary. The analyses 

also must assess a postulated beyonddesign-basis zirconium fire scenario. 

B. Objectives of This Proposed Rule 

This proposed rule would amend the 

current requirements for production and 

utilization facility licensees during 

decommissioning. Experience has 

demonstrated that licensees for 

decommissioning nuclear power 

reactors seek several exemptions and 

license amendments per site to establish 

a long-term licensing basis for 

decommissioning. Non-power 

production or utilization facility 

licensees typically seek license 

amendments in decommissioning to 

change their 10 CFR part 50 operating 

licenses to possession-only licenses. By 

issuing this rule, the NRC would 

establish regulations that would 

maintain safety and security at sites 

transitioning to decommissioning 

without the need to grant specific 

exemptions or license amendments in 

certain regulatory areas. Specifically, 

the decommissioning rulemaking 

would: (1) Propose a regulatory regime 

that continues to provide reasonable 

assurance of adequate protection of 

public health and safety and the 

common defense and security at 

decommissioning sites; (2) ensure that 

the requirements for decommissioning 

are clear and appropriate; (3) adopt 

regulations to address generic issues 

applicable to all decommissioning 

nuclear power reactors that have 

historically been addressed through 

similarly worded exemptions or license 

amendments; and (4) identify, define, 

and resolve additional areas of concern 

related to the regulation of 

decommissioning licensees under 10 

CFR parts 50 and 52. 

Given that the current regulatory 

framework regarding decommissioning 

is adequate to protect public health and 

safety and the common defense and 

security, many of the new requirements 

proposed by this rulemaking are 

alternatives to the current requirements. 

C. Applicability 

This proposed rule would apply to 

the following categories of license 

holders: 

• Nuclear power reactors currently 

licensed under 10 CFR part 50 

• Future nuclear power reactors 

licensed under 10 CFR part 50 

• Nuclear power reactors currently 

licensed under 10 CFR part 52 

• Future nuclear power reactors 

licensed under 10 CFR part 52 

• Non-power production or utilization 

facilities and fuel reprocessing plants 

currently licensed under 10 CFR part 

50 

• Future non-power production or 

utilization facilities and fuel 

reprocessing plants licensed under 10 

CFR part 50 

D. Applicability to NRC Licensees 

During Operations 

The proposed rule includes changes 

in three areas that would apply to NRC 

licensees during operations: (1) The 

process to change a licensee’s security 

plan, (2) the timing of decommissioning 


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4 Given that the public comments referred to 

‘‘standalone ISFSIs,’’ this proposed rule uses that 

same terminology. However, in accordance with 

Inspection Manual Chapter 2690, ‘‘Inspection 

Program for Dry Storage of Spent Reactor Fuel at 

Independent Spent Fuel Storage Installations and 

for 10 CFR part 71 Transportation Packagings,’’ 

dated March 9, 2012, the NRC uses the term ‘‘awayfrom-reactor (AFR) ISFSI’’ to refer to ‘‘any general 

licensed ISFSI where decommissioning and final 

survey activities related to reactor operations are 

completed and the only remaining operation 

conducted under the 10 CFR part 50 license is the 

operation of the general licensed ISFSI.’’ 

funding assurance reporting 

requirements, and (3) identification of 

10 CFR 26.3, ‘‘Scope,’’ as a regulation 

with substantive requirements that 

could result in criminal penalties if 

violated. 

The NRC’s regulations in § 50.54(p) 

establish processes that allow licensees 

to make changes to their security plans. 

The NRC is proposing that all nuclear 

power reactor licensees making a 

change under § 50.54(p)(2) submit in 

their report of the change a summary of 

any analysis that was completed to 

make the determination that the change 

does not decrease the safeguards 

effectiveness of the security plan. 

Additionally, the NRC is proposing to 

revise § 50.54(p) to include definitions 

of the terms ‘‘change’’ and ‘‘decrease in 

safeguards effectiveness.’’ The 

application of these definitions is 

limited to use with the revised 

§ 50.54(p) and will apply to all holders 

of 10 CFR part 50 operating licenses and 

10 CFR part 52 combined licenses. 

The proposed rule would change the 

timing of the decommissioning funding 

assurance reporting requirements in 

§ 50.75(f)(1) to coordinate them with the 

ISFSI decommissioning reporting 

requirements in § 72.30, ‘‘Financial 

assurance and recordkeeping for 

decommissioning.’’ This change would 

convert the biennial decommissioning 

funding status report required for 10 

CFR part 50 and 10 CFR part 52 nuclear 

power reactor licensees to a triennial 

decommissioning funding status report 

as currently required for 10 CFR part 72 

ISFSI licensees. 

Current § 26.3 includes a substantive 

requirement and violations of this 

regulation should be subject to criminal 

penalties. Therefore, this proposed rule 

would remove § 26.3 from the list of 

provisions that are not subject to 

criminal penalties if violated in 

§ 26.825(b). 

E. Applicability to ISFSI-Only and 

Standalone ISFSI/Decommissioned 

Reactor Sites 

During the public comment period for 

the draft regulatory basis, the NRC 

received many comments on the 

applicability of the decommissioning 

rulemaking to ‘‘standalone ISFSI’’ 4 sites 

where the associated reactor has already 

been decommissioned in comparison 

with ‘‘ISFSI-only’’ sites. As part of this 

rulemaking effort, the NRC recommends 

standardizing the terms ‘‘ISFSI-only’’ 

and ‘‘standalone ISFSI/Decommissioned 

Reactor’’ as follows: 

• ‘‘ISFSI-only’’ sites contain nuclear 

power reactor facilities that are still 

involved in decommissioning activities, 

but the spent fuel has been completely 

transferred from the SFPs to dry storage 

in an onsite ISFSI. For these facilities, 

the remaining decommissioning 

activities are primarily related to 

remediation of any remaining residual 

radioactivity at the site to meet the 

license termination and 

decommissioning criteria in 10 CFR part 

20, subpart E. The ‘‘ISFSI-only’’ term 

refers to the location of the spent fuel; 

the term reflects that no spent fuel is 

stored in the SFP, and all of the spent 

fuel is in dry storage in an onsite ISFSI. 

• ‘‘Standalone ISFSI/ 

Decommissioned Reactor’’ sites are 

those former nuclear power reactor 

facilities where the license termination 

and decommissioning criteria in 10 CFR 

part 20, subpart E, have already been 

met, with the exception of the ISFSI 

area. The licensee’s 10 CFR part 50 

license for the site has been reduced to 

an area that only encompasses the ISFSI 

facility (unless the facility ISFSI is 

licensed under a 10 CFR part 72 specific 

license, in which case the 10 CFR part 

50 license is wholly terminated). The 

remaining activities at these facilities 

that are regulated by the NRC are spent 

fuel storage and the eventual 

decommissioning of the ISFSI itself, 

once the spent fuel has been 

permanently removed from the site. A 

10 CFR part 72 specific license ISFSI is 

decommissioned in accordance with 10 

CFR 72.54, ‘‘Expiration and termination 

of licenses and decommissioning of 

sites and separate buildings or outdoor 

areas.’’ 

Accordingly, the proposed 

requirements would not apply to 

standalone ISFSI/Decommissioned 

Reactor sites because those licensees 

have already decommissioned their 10 

CFR part 50 facilities and met the 

decommissioning and license 

termination criteria in 10 CFR part 20, 

subpart E, with the exception of the area 

encompassed by the remaining ISFSI. 

The proposed requirements are 

consistent with the licensing actions 

that the NRC has already approved for 

these licensees. In addition, the 

proposed requirements of this 

rulemaking provide an alternative to the 

existing decommissioning regulations 

and would not impose new 

requirements on ISFSI-only licensees. 

F. Graded Approach 

As the NRC reviewed the exemption 

and license amendment requests related 

to the recent nuclear power reactor 

decommissionings and noted the 

growing list of future planned 

permanent shutdowns, as discussed in 

the ‘‘Background’’ section of this 

document, the NRC realized that the 

existing regulatory framework could and 

should be revised to provide for a more 

efficient decommissioning process. As 

early as the late 1990’s, the NRC 

contemplated an integrated rulemaking 

to provide an appropriate graded 

approach to the decommissioning 

process. A graded approach is a process 

by which the safety requirements and 

criteria adjust during the 

decommissioning process 

commensurate with several factors. 

These factors include the magnitude of 

any credible hazard involved, the 

particular characteristics of a facility, 

and the balance between radiological 

hazards and non-radiological hazards 

(e.g., fire, flood, chemical spill) as 

applicable to specific points in time 

within the decommissioning process. 

This approach would be a risk-informed 

process. 

Currently, no explicit regulatory 

provisions distinguish requirements in 

several technical areas for a nuclear 

power reactor that has permanently 

ceased operations from those for an 

operating nuclear power reactor. To 

address this, the NRC is proposing to 

amend its regulations to provide an 

efficient regulatory framework for the 

transition to decommissioning. Under 

this proposed rule, the NRC would 

adopt an optional graded approach for 

several technical areas that provides a 

set of requirements commensurate with 

the reductions in radiological risk at 

each of the following four levels of 

decommissioning: (1) Permanent 

cessation of operations and permanent 

removal of all fuel from the reactor 

vessel, (2) sufficient decay of fuel in the 

SFP such that it would not reach 

ignition temperature for the zirconium 

alloy cladding of the fuel within 10 

hours under adiabatic heatup conditions 

(i.e., a complete loss of SFP water 

inventory with no heat loss), (3) transfer 

of all fuel to dry storage, and (4) removal 

of all fuel from the site. Four technical 

areas of this proposed rule (Emergency 

Preparedness, Physical Security, Cyber 

Security, and Offsite and Onsite 

Insurance) use all or some of this graded 

approach. 


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G. Technical Basis for Graded Approach 

The NRC has approved exemptions 

from the emergency planning 

regulations in § 50.47, ‘‘Emergency 

plans,’’ and appendix E, ‘‘Emergency 

Planning and Preparedness for 

Production and Utilization Facilities,’’ 

to 10 CFR part 50 at several 

permanently shutdown and defueled 

nuclear power reactor sites. Licensees 

that have been granted EP exemptions 

must maintain an onsite emergency plan 

addressing the classification of an 

emergency, notification of emergencies 

to licensee personnel and offsite 

authorities, and coordination with 

designated offsite government officials 

following an event declaration so that, 

if needed, offsite authorities may initiate 

appropriate response actions. At the 

appropriate points in decommissioning, 

the EP exemptions may also relieve the 

licensee from certain requirements of 

§ 50.47 and appendix E to 10 CFR part 

50 as they pertain to offsite radiological 

EP, including the requirement to 

maintain the 10-mile plume exposure 

pathway and the 50-mile ingestion 

pathway emergency planning zones 

(EPZs). The NRC granted these 

exemptions based, in part, on its 

determination that there are no 

applicable design-basis accidents at a 

decommissioning licensee’s facility that 

could result in an offsite radiological 

release exceeding the limits established 

by the EPA’s early-phase PAGs at the 

exclusion area boundary. 

The NRC also relied on analyses from 

NUREG–1738 that showed that 

emergency planning would be of 

marginal benefit in reducing the risk of 

a beyond-design-basis zirconium fire in 

the SFP if the accident evolved slowly 

enough to allow mitigative measures 

and, if necessary, to allow offsite 

protective actions to be implemented 

without preplanning. This conclusion 

was based, in part, on the assumption 

that it would take at least 10 hours for 

spent fuel to heat up to the temperature 

at which the onset of fission product 

release is expected during an SFP rapid 

draindown event. This 10-hour period 

would provide a substantial amount of 

time for the licensee to take onsite 

mitigation measures and, if necessary, 

for offsite authorities to take appropriate 

response actions to protect the public. 

To support the approval of exemptions 

from portions of the EP regulations, 

licensees had to demonstrate through 

site-specific analyses that in a 

draindown event at their SFP the fuel 

would not reach the zirconium fuel 

cladding ignition temperature for at 

least 10 hours under adiabatic heatup 

conditions. 

A 10-hour timeframe has been 

justified in the past for similar purposes. 

In the Low Power Rule (47 FR 30232; 

July 13, 1982), the NRC amended its 

regulations to clarify that no NRC or 

Federal Emergency Management Agency 

(FEMA) review, findings, and 

determinations concerning the state or 

adequacy of offsite emergency 

preparedness were necessary for 

issuance of operating licenses 

authorizing fuel loading and low power 

operation (i.e., up to 5 percent of rated 

power). The NRC determined that 

several factors contributed to a 

substantial reduction in risk and 

potential accident consequences for low 

power testing as compared to the higher 

risks in continuous full power 

operation. These factors included 

consideration of the reduced source 

term, the capability of mitigation 

systems, and the time scale for taking 

actions to identify and mitigate an 

accident. Even for a postulated lowlikelihood, design-basis accident during 

low power operations, which eventually 

results in release of fission products into 

the containment, at least 10 hours 

would be available to allow adequate 

precautionary actions to be taken to 

protect the public near the site. 

To support a graded approach during 

decommissioning, the NRC further 

examined the certainty and margin 

provided by a 10-hour timeframe for the 

fuel to heat up in relation to the time for 

taking mitigating actions and 

appropriate EP response actions. The 

NRC conducted an applied research 

study (‘‘Transmittal of Reports to Inform 

Decommissioning Plant Rulemaking for 

User Need Request NSIR–2015–001,’’ 

dated May 31, 2016 (ADAMS Accession 

No. ML16110A416)) with three tasks: (1) 

To perform a task analysis that includes 

a timeline of responder actions at 

representative SFP configurations to 

mitigate a draindown event and 

determine its likelihood of success, (2) 

to analyze representative spent fuel to 

determine the decay time necessary for 

the fuel to remain below zirconium clad 

ignition temperature for at least 10 

hours assuming adiabatic heatup 

conditions, and (3) to analyze the offsite 

dose rate from the radionuclides 

released during a hypothetical spent 

fuel zirconium clad ignition accident. 

As demonstrated in these analyses, for 

many initiating events at 

decommissioning reactors, mitigative 

actions would have a high likelihood of 

preventing uncontrolled spent fuel 

heatup. In cases where an uncontrolled 

heatup is not prevented, the heatup 

would be relatively slow, providing 

significant time before a radiological 

release. In the case of a radiological 

release, dose rates would be low enough 

such that significant additional time is 

available to take offsite actions to 

protect the public. 

The NRC’s analysis of spent fuel 

decay times provided information on 

the time required for fuel to heat up to 

900 degrees Celsius (C) (i.e., the 

temperature at which the onset of 

fission product release is expected for a 

zirconium fuel cladding fire) as a 

function of decay time for both 

pressurized water reactor (PWR) and 

boiling water reactor (BWR) assemblies. 

The analysis also included sensitivities 

to the mass of the racks and the fuel 

configuration in the SFP. The NRC notes 

that the decay periods provided for 

PWRs and BWRs are based on studies 

that consider current operating 

parameters in the nuclear power 

industry (e.g., fuel types, enrichment, 

and fuel burnup levels). Based on this 

analysis, the NRC concluded that after 

a decay period of 10 months for BWRs 

or 16 months for PWRs, beginning when 

the reactor permanently shuts down, the 

spent fuel cannot reasonably heat up to 

clad ignition temperature within 10 

hours after a draindown event. These 

decay periods are based on an adiabatic 

heatup to 900 degrees C assuming the 

decay heat value for the hottest 

assembly (as opposed to an average 

assembly), a burnup of 60 gigawatt days 

per metric ton of heavy metal (GWd/ 

MTHM), and accounting for the mass of 

the racks. The analysis assumption of 60 

GWd/MTHM conservatively bounds 

current industry burnups and 

enrichments for zirconium clad fuel and 

provides margin for potentially higher 

burnup rates, up to 72 GWd/MTHM. 

This analysis does not account for the 

additional time margin that would be 

provided if additional cooling 

mechanisms were available or would be 

provided by a more favorable SFP 

configuration such that the heat load is 

more uniformly distributed. 

The NRC’s analysis of dose rates 

shows that even in the event of a 

beyond-design-basis accident leading to 

a rapid draindown of the SFP and 

subsequent zirconium fire, there would 

be additional time margin on the order 

of several hours beyond the 10-hour 

heatup time during which protective 

actions could be taken to protect the 

public before the dose levels associated 

with EPA PAGs would be exceeded 

offsite. 

In addition to the analyses performed 

by the NRC to support this rulemaking, 

as discussed in the ‘‘Background’’ 

section of this document, the 

conclusions of NUREG–2161 and 

NUREG–1738 support the technical 


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basis for a graded approach during 

decommissioning as they provide 

insight into the risk of an offsite release 

and the effectiveness of mitigation 

measures. 

• In NUREG–2161, the NRC 

considered various spent fuel cooling 

mechanisms and additional heat from 

oxidation. Because previous studies 

found that earthquakes present the 

dominant risk for SFPs, this analysis 

considered a severe earthquake with 

ground motion stronger than the 

maximum earthquake reasonably 

expected to occur for the reference 

plant, which would challenge the SFP 

integrity. The study considered two 

spent fuel configurations: High-density 

and low-density loading. The study also 

analyzed two cases for each scenario: 

One that credited the mitigation 

measures of § 50.54(hh)(2) (i.e., the 

strategies to maintain or restore SFP 

cooling in the event of a loss of large 

areas of the plant as a result of fire or 

explosion), and one in which those 

measures were not used or were 

unsuccessful. The study results showed 

that successful mitigation reduces the 

likelihood of a release and that the 

likelihood of a release was equally low 

for both high- and low-density loading 

in the SFP. The study found that a 

release is not expected to occur at the 

nuclear power reactor site studied for at 

least 72 hours following a beyonddesign-basis seismic event that occurs 

more than 60 days after shutdown. 

• In NUREG–1738, the NRC 

presented the results of its evaluation of 

the potential accident risk for an SFP at 

a decommissioning nuclear power 

reactor in the United States. NUREG– 

1738 identified a zirconium cladding 

fire resulting from a substantial loss of 

water from the SFP as the only 

postulated scenario at a 

decommissioning nuclear power reactor 

that could result in a significant 

radiological release. While highly 

unlikely, the consequences of such an 

accident could lead to an offsite dose in 

excess of the EPA PAGs. Based on spent 

fuel storage design characteristics and 

operating practices considered in the 

analysis, the scenarios that lead to this 

condition have very low probabilities of 

occurrence. Accordingly, these 

scenarios are considered to be beyond 

the facility’s design basis. Furthermore, 

as the spent fuel ages, the generation of 

decay heat decreases. After a certain 

amount of time, the overall risk of a 

zirconium fire becomes extremely low 

because of: (1) The large amount of time 

available for preventive and mitigating 

actions and (2) the increased probability 

that the decay heat will be low enough 

that the fuel will be air-coolable in the 

post-event configuration. 

H. Levels of Decommissioning 

Using the aforementioned analyses as 

its technical basis, the NRC is proposing 

to amend its regulations to provide an 

efficient regulatory framework during 

decommissioning using a graded 

approach in several technical areas. 

This graded approach is commensurate 

with the reductions in radiological risk 

at four levels of decommissioning: 

(Level 1) permanent cessation of 

operations and permanent removal of all 

fuel from the reactor vessel, (Level 2) 

sufficient decay of fuel in the SFP such 

that it would not reach ignition 

temperature within 10 hours under 

adiabatic heatup conditions, (Level 3) 

transfer of all spent fuel to dry storage, 

and (Level 4) removal of all fuel from 

the site. These levels are discussed 

further as follows: 

1. Level 1 

Licensees in Level 1 include nuclear 

power reactor licensees that have 

docketed certifications of permanent 

cessation of operations and permanent 

removal of fuel from the reactor vessel 

pursuant to § 50.82, ‘‘Termination of 

license,’’ or § 52.110, ‘‘Termination of 

license.’’ In this level, a 

decommissioning nuclear power reactor 

is defueled and permanently shut down, 

but the spent fuel in the SFP is still 

susceptible to a zirconium fuel cladding 

fire within 10 hours under adiabatic 

heatup conditions. 

2. Level 2 

In Level 2, the reactor is defueled and 

permanently shut down, and spent fuel 

in the SFP has decayed and cooled 

sufficiently such that it cannot heat up 

to the zirconium cladding ignition 

temperature within 10 hours under 

adiabatic conditions. The NRC has 

determined that this condition is 

reached after spent fuel has decayed for 

a minimum of either 10 months for a 

BWR or 16 months for a PWR or an 

alternative site-specific timeframe to be 

approved by the NRC. The decay period 

could begin when the fuel is still in the 

reactor vessel but the reactor has 

permanently ceased operations. In order 

to verify that a licensee has met the 

condition, the NRC would rely upon the 

date of permanent cessation of operation 

provided by a licensee under 

§ 50.4(b)(8) or § 52.3(b)(8), updated as 

necessary under § 50.9 or § 52.6, both 

entitled ‘‘Completeness and accuracy of 

information.’’ Because the identified 

date of permanent cessation of 

operations would determine transition 

from Level 1 to Level 2, the NRC would 

consider a change in the planned date 

initially certified to the NRC for 

permanent cessation of operations to the 

actual date as information ‘‘having a 

significant implication for public health 

and safety or common defense and 

security’’ under § 50.9 or § 52.6. At this 

point, the site may also possess a 

radioactive inventory of liquid 

radiological waste, radioactive reactor 

components, and contaminated 

structural materials. The radioactive 

inventory may change, depending on 

the licensee’s proposed shutdown 

activities and schedule. 

3. Level 3 

In Level 3, all spent nuclear fuel 

(SNF) is in dry cask storage pursuant to 

the terms and conditions of a license 

granted under 10 CFR part 72, including 

the general license issued in § 72.210. 

However, the licensee may still hold a 

10 CFR part 50 or 10 CFR part 52 

license, and the site may contain a 

radioactive inventory of liquid 

radiological waste, radioactive reactor 

components, and contaminated 

structural materials. 

4. Level 4 

At this point in the facility’s life 

cycle, all SNF has been removed from 

the site. The site may possess a 

radioactive inventory of liquid 

radiological waste, radioactive reactor 

components, and contaminated 

structural materials. The radioactive 

inventory during this configuration may 

change, depending on the licensee’s 

proposed decommissioning activities 

and schedule. 

As a facility transitions from being 

operational to having all SNF in dry 

cask storage, the proposed rule’s 

regulatory requirements are graded to 

provide for reasonable assurance of the 

health and safety of the public 

commensurate with the risk profile of 

the facility. Table 2 summarizes the 

proposed changes to decommissioning 

requirements in the technical areas that 

use aspects of this graded approach. 


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IV. Scope of the Proposal 

This rulemaking proposes revising 

requirements in 16 technical areas. 

A. Emergency Preparedness 

1. Introduction 

In 1978, an NRC and EPA task force 

established the planning basis for EP for 

nuclear power reactor accidents in 

NUREG–0396, ‘‘Planning Basis for the 

Development of State and Local 

Government Radiological Emergency 

Response Plans in Support of Light 

Water Nuclear Power Plants’’ (ADAMS 

Accession No. ML051390356). This 

guidance provides a basis for offsite 

radiological EP efforts for large lightwater nuclear power reactor facilities. In 

NUREG–0396, the task force determined 

that no single accident sequence should 

be identified as a planning basis and 

chose to provide recommendations in 

terms of the consequences and 

characteristics of accidents that would 

be important in determining the extent 

of the planning effort. The task force 

concluded that the EP planning basis 

requires consideration of a spectrum of 

accidents, informed by probability 

considerations. The scope of the 

planning effort was based on three key 

planning elements: (1) The distance to 

which planning for the initiation of 

predetermined protective actions is 

warranted, (2) the time-dependent 

characteristics of potential releases and 

exposures, and (3) the kinds of 

radioactive materials that can 

potentially be released to the 

environment. The risk-informed 

planning basis for EP, established in 

NUREG–0396, was endorsed for use in 

the NRC’s policy statement, ‘‘Planning 

Basis for Emergency Responses to 

Nuclear Power Reactor Accidents,’’ 

dated October 23, 1979 (44 FR 61123). 

This planning basis results in 

emergency plans that are effective, 

regardless of the accident probability. 

The rationale in NUREG–0396 and the 

planning basis elements can also be 

applied to light water nuclear power 

reactors in decommissioning to scope 

the planning effort. The NRC applied 

the NUREG–0396 methodology (i.e., 

consideration of a spectrum of accident 

consequences and the three key 

planning elements) to establish a graded 

approach to EP for decommissioning 

nuclear power reactors that maintains 

public health and safety. As discussed 

in NUREG–0396, no single specific 

accident sequence should be isolated as 

the one for which to plan because each 

accident could have different 

consequences, both in nature and 

degree. Further, the range of possible 

selections for a planning basis is very 

large, starting with a zero point of 

requiring no planning at all, because 

significant offsite radiological accident 

consequences are unlikely to occur to 

planning for the worst possible accident 

regardless of its extremely low 

likelihood. Fundamentally, the 

spectrum of possible accidents is 

significantly smaller and the risk of an 

offsite radiological release is 

significantly lower at a nuclear power 

facility that has permanently shut down 


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Table 2-Summary of Graded Approach 
Federal Register / Vol. 87, No. 42 / Thursday, March 3, 2022 / Proposed Rules 12269 

and removed fuel from the reactor 

vessel than at an operating nuclear 

power reactor. All such accidents would 

be associated with hazards based on the 

storage of spent fuel, either in the SFP 

or in dry cask storage, until its 

permanent removal from the site. In 

NUREG–1738, the NRC found that the 

event sequences important to risk at 

decommissioning sites are limited to 

large earthquakes and cask drop events. 

For EP assessments, this is an important 

difference relative to operating nuclear 

power reactors, where typically a large 

number of different sequences make 

significant contributions to risk. 

Although the NRC considered the full 

spectrum of accidents applicable to a 

decommissioning nuclear power 

reactor, the number of events that can 

have significant offsite consequences is 

greatly reduced, and the events are 

dominated by the zirconium fire 

scenario—a postulated, but highly 

unlikely, beyond-design-basis accident 

that involves a major loss of water 

inventory from the SFP, resulting in a 

significant heatup of the spent fuel and 

culminating in substantial zirconium 

cladding oxidation, fire, and fuel 

damage. The guidance in NUREG–0396 

states that while it is not appropriate to 

develop specific plans for the most 

severe and most improbable events, the 

characteristics of these events should be 

considered ‘‘in judging whether 

emergency plans based primarily on 

smaller accidents can be expanded to 

cope with larger events.’’ This approach 

provides reasonable assurance that 

capabilities exist to minimize the 

impacts of even the most severe events. 

Consistent with this guidance, the NRC 

considered the potential impacts of a 

zirconium fire, even with the assurance 

that mitigating strategies are in place to 

prevent an offsite release from occurring 

for this highly unlikely beyond-designbasis event. 

In addition to the three analyses 

performed by the NRC to support this 

rulemaking (ADAMS Accession No. 

ML16110A416), the NRC has previously 

conducted SFP studies, including 

NUREG–2161 and NUREG–1738, the 

conclusions of which support the 

technical basis for a graded approach to 

EP. Overall, these analyses: (1) 

Demonstrate that a period of 10 hours 

provides sufficient time to implement 

mitigation measures for design-basis 

events at decommissioning sites, (2) 

provide a conservative basis for a spent 

fuel decay time beyond which the fuel 

in the SFP can reasonably be expected 

to take longer than 10 hours to heat up 

to ignition temperature, and (3) provide 

additional understanding of the amount 

of time available for taking action in 

response to beyond-design-basis events, 

including the margin of time that offsite 

agencies have to decide upon and 

initiate actions to protect public health 

and safety. The NRC applied these 

analyses and the considerations from 

previous studies of SFP risk to the 

planning basis elements from NUREG– 

0396 to develop the proposed 

regulations for EP at various levels 

during decommissioning. 

2. Graded Approach for Emergency 

Preparedness 

A graded approach to EP has a 

longstanding regulatory history. The 16 

planning standards for operating 

reactors, outlined in § 50.47(b), and the 

associated evaluation criteria in 

NUREG–0654/FEMA–REP–1, Revision 

1, ‘‘Criteria for Preparation and 

Evaluation of Radiological Emergency 

Response Plans and Preparedness in 

Support of Nuclear Power Plants,’’ 

issued November 1980 (ADAMS 

Accession No. ML040420012) or 

Revision 2 issued December 2019 

(ADAMS Accession No. ML19347D139), 

are one part of a continuum of planning 

standards for radiological EP. The 

regulations in § 50.47(c)(2) for case-bycase EPZ size determinations; the EP 

regulations for research and test reactors 

and other non-power production or 

utilization facilities, fuel cycle facilities, 

and ISFSIs; and the EP considerations 

for small modular reactors and other 

new technologies (see the Proposed 

Rule for ‘‘Emergency Preparedness for 

Small Modular Reactors and Other New 

Technologies’’ (85 FR 28436 and 85 FR 

32308)), are also part of a graded 

approach to EP that is commensurate 

with the relative radiological risk, 

source term, and potential hazards, 

among other considerations. 

Consistent with the concept of a 

graded approach, the NRC is proposing 

four levels of emergency planning 

standards that coincide with the same 

milestones as the graded approach: 

• Post-Shutdown Emergency Plan 

(PSEP) (Level 1) 

• Permanently Defueled Emergency 

Plan (PDEP) (Level 2) 

• ISFSI-Only Emergency Plan (IOEP) 

(Level 3) 

• No emergency planning (Level 4) 

In developing this proposed rule, the 

NRC considered the appropriateness of 

the EP requirements in 10 CFR part 50 

and 10 CFR part 72 for 

decommissioning sites, including those 

requirements that have historically been 

addressed in approved exemptions and 

those that have not. The proposed 

planning standards within the levels are 

based on the current set of operating 

reactor EP standards informed by the 

analyses and considerations supporting 

a graded approach to EP as previously 

described, as well as public comments 

on the ANPR and on the draft regulatory 

basis for this rulemaking. The NRC also 

considered the criteria of safety, 

implementation costs, efficiency, 

transparency, flexibility, and 

responsiveness. The following 

discussion describes the proposed 

graded approach to EP. 

Post-Shutdown Emergency Plan 

For a decommissioning site, once all 

the fuel is in the SFP, the spectrum of 

accidents that can have significant 

offsite consequences is greatly reduced 

and is dominated by the highly unlikely 

occurrence of a zirconium fire. The 

primary consideration for the planning 

basis for a PSEP is the potential 

consequences and timing of this narrow 

spectrum of accidents in relation to the 

time needed to initiate protective 

actions. 

From a regulatory perspective, the 

purpose of a PSEP is to provide a 

transition period to ensure that an 

appropriate level of EP is maintained 

onsite and offsite to respond to 

applicable DBAs and to ensure a prompt 

response to the highly unlikely rapid 

draindown of the SFP and subsequent 

zirconium fire and release occurring in 

less than 10 hours. A nuclear power 

reactor licensee would be permitted to 

transition to a PSEP after the NRC’s 

docketing of the licensee’s certifications 

of permanent cessation of operations 

and permanent removal of fuel from the 

reactor vessel pursuant to §§ 50.82 or 

52.110. The NRC anticipates that 

licensees will maintain a PSEP from the 

date that the NRC dockets the licensee’s 

certifications of permanent cessation of 

operations and permanent removal of 

fuel from the reactor vessel, until the 

spent fuel has decayed for a period of 

at least 10 months (for BWRs) or 16 

months (for PWRs) from the date of 

permanent cessation of operations, 

unless a different period is justified. 

During this time, the licensee would be 

relieved of the regulatory burden of 

requirements that are not needed to 

support an appropriate level of EP as 

preparations are made to implement a 

PDEP. The PSEP is a transition period 

for both onsite and offsite emergency 

planning in which the regulatory 

requirements for periodic updates, 

reviews, and audits that were necessary 

to support operating reactor EP 

programs should not interfere with 

efforts to establish an appropriate level 

of EP for a PDEP. The NRC does not 

intend for many significant changes to 


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occur to the emergency plan while the 

PSEP is used. 

Permanently Defueled Emergency Plan 

For plants that have permanently shut 

down and defueled, the proposed EP 

approach is based primarily on 

conditions that: (1) A postulated 

radiological release would not exceed 

the EPA early-phase PAGs at the 

exclusion area boundary for DBAs 

applicable to a permanently shutdown 

and defueled reactor, and (2) sufficient 

time would exist to implement 

mitigative actions in response to a 

postulated zirconium fire beyonddesign-basis accident scenario in the 

SFP and, if warranted, for offsite 

officials to initiate appropriate response 

actions using all-hazards planning to 

protect public health and safety. 

Because of the additional time available 

to take mitigative actions and, if 

necessary, to initiate protective actions, 

many requirements applicable under an 

operating reactor emergency plan or a 

PSEP would not be required to protect 

public health and safety and, therefore, 

would not be applicable to licensees 

with sufficiently decayed spent fuel 

under a PDEP. 

The NRC is proposing two regulatory 

alternatives to specify when the 

transition to a PDEP may occur: (1) After 

a specified amount of spent fuel decay 

time that starts from the date of 

permanent cessation of operations, or 

(2) after an alternative timeframe based 

on a site-specific analysis that shows 

that the fuel in the SFP cannot heat up 

to zirconium fuel cladding ignition 

temperature (900 degrees C) within 10 

hours under adiabatic conditions. In 

either case, a licensee would be 

permitted to transition to a PDEP only 

after the NRC’s docketing of the 

licensee’s certifications of permanent 

cessation of operations and permanent 

removal of fuel from the reactor vessel 

pursuant to § 50.82 or § 52.110. This 

proposed rule specifies an acceptable 

decay time to remove the requirement 

for licensees to provide a site-specific 

analysis. Licensees are provided the 

option to submit a site-specific analysis 

proposing an alternative decay period, 

but such an analysis would be subject 

to NRC review and approval before a 

transition to a PDEP. 

Independent Spent Fuel Storage 

Installation-Only Emergency Plan 

The third level of decommissioning 

under the proposed rule would occur 

when all spent fuel is removed from the 

SFP and placed in dry cask storage. At 

this point, the licensee would have an 

ISFSI-only emergency plan, or IOEP. A 

licensee with all of its spent fuel in dry 

cask storage that terminates its 10 CFR 

part 50 or 10 CFR part 52 license must 

first obtain a specific 10 CFR part 72 

license. Accordingly, the licensee would 

then transition to the EP requirements 

for dry cask storage in § 72.32, 

‘‘Emergency Plan.’’ A licensee 

maintaining its 10 CFR part 50 or 10 

CFR part 52 license may opt to change 

its EP program to align it with the 

requirements of § 72.32 once all spent 

fuel is transferred to dry cask storage. 

These two categories of licensees (i.e., 

10 CFR part 72 specific licensees and 10 

CFR part 50 or 10 CFR part 52 licensees 

with ISFSIs licensed under the 10 CFR 

part 72 general license) would be 

permitted to adopt an IOEP, consistent 

with the EP requirements that currently 

exist under § 72.32(a). 

All Spent Fuel Removed From Site 

This proposed rule would allow a 

licensee to terminate its EP program 

once all the spent fuel has been 

permanently removed from the site, 

because the site no longer poses any risk 

of a radiological release from the spent 

fuel. 

3. Licensee Supporting Analyses 

Decommissioning nuclear power 

reactor licensees submitting requests for 

exemptions under § 50.12, ‘‘Specific 

exemptions,’’ from EP regulations have 

performed a series of supporting 

analyses for NRC review, as described in 

NSIR/DPR–ISG–02, ‘‘Interim Staff 

Guidance: Emergency Planning 

Exemption Requests for 

Decommissioning Nuclear Power 

Plants’’ (ADAMS Accession No. 

ML14106A057). To support the 

exemption requests, these analyses must 

demonstrate that: (1) Any radiological 

release for applicable DBAs (e.g., fuel 

handling accident in the spent fuel 

storage facility, waste gas system 

release, and cask handling accident if 

the cask handling system is not licensed 

as single-failure-proof) would not 

exceed the limits of EPA PAGs at the 

exclusion area boundary, and (2) 

mitigation strategies and guidelines 

exist to provide an integrated response 

capability for beyond-design-basis 

events. In addition, licensees are 

required to demonstrate that, in the 

event of a complete loss of SFP water 

inventory with no heat loss (adiabatic 

heatup), a period of at least 10 hours 

would be available from the time all 

cooling is lost until any zirconium fuel 

cladding temperature reaches 900 

degrees C. 

Under this proposed rule, the NRC 

would not require licensees to submit 

these analyses to the NRC for review 

and approval (separately from existing 

NRC oversight processes described later 

in this document) or to certify that these 

analyses have been completed to 

support a change between EP levels. 

The NRC anticipates that a licensee 

would analyze applicable DBAs using 

the process under § 50.59 and reflect the 

analysis in the licensee’s updated FSAR. 

The NRC expects that licensees have 

developed and maintained mitigation 

strategies for beyond-design-basis events 

as required by NRC Order EA–12–049. 

For the heatup analysis, the NRC has 

already performed analyses of 

representative PWR and BWR spent fuel 

to determine the decay time necessary 

for the fuel to remain below clad 

ignition temperature for at least 10 

hours assuming adiabatic heatup 

conditions. These analyses contain 

numerous conservatisms, such that the 

decay times specified in the rule would 

bound the decay time required for 

plants with fuel assemblies from the 

final offload to the spent fuel pool with 

burnup less than 72 GWd/MTHM and 

zirconium cladding to attain the 10-hour 

criterion. This particular analysis 

supports a transition to PDEP 

requirements, as previously described. 

The NRC is proposing an option to 

allow licensees to develop their own 

site-specific analysis for this transition 

time; however, licensees would need to 

submit such analyses to the NRC for 

review and approval. This proposed 

rule details that process. 

The following sections describe the 

proposed EP planning standards and 

requirements for each graded level of EP 

(i.e., PSEP, PDEP, and IOEP) under 

proposed §§ 50.54(q) and 50.200, 

‘‘Power reactor decommissioning 

emergency plans.’’ The NRC is issuing 

draft Regulatory Guide (DG) DG–1346, 

‘‘Emergency Planning for 

Decommissioning Nuclear Power 

Reactors’’ (ADAMS Accession No. 

ML21347A046), for public comment 

with this proposed rule that includes 

guidance on one method acceptable to 

the NRC for complying with these 

proposed requirements. This regulatory 

guide will supersede NSIR/DPR–ISG–02 

upon publication of the final rule. This 

proposed rule contains a risk-informed, 

consequence-oriented, graded approach 

to EP for decommissioning sites that 

maintains the defense-in-depth 

philosophy and provides reasonable 

assurance that adequate protective 

measures can and will be taken in the 

event of a radiological emergency. 

4. Post-Shutdown Emergency Plans 

The NRC is proposing in § 50.54(q)(7) 

that a licensee can transition to a PSEP 

after the NRC’s docketing of the 

licensee’s certifications of permanent 


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cessation of operations and permanent 

removal of all fuel from the reactor 

vessel pursuant to §§ 50.82(a)(1) or 

52.110(a). A PSEP provides a transition 

period from the EP requirements for an 

operating reactor to the PDEP 

requirements under proposed 

§ 50.200(b) and (c). The NRC is 

proposing regulations under new 

§ 50.200(a) that would clarify how the 

planning standards in § 50.47(b) and 

requirements in appendix E to 10 CFR 

part 50 apply to a nuclear power reactor 

licensee’s PSEP. 

PSEP Staffing and Emergency Response 

Organization 

Currently, the following regulations 

govern the staffing of the emergency 

response organization (ERO): 

• Section 50.47(b)(1), which states, in 

part, ‘‘Primary responsibilities for 

emergency response by the nuclear 

facility . . . have been assigned . . . 

and each principal response 

organization has staff to respond and to 

augment its initial response on a 

continuous basis.’’ 

• Section 50.47(b)(2), which states, in 

part, ‘‘[A]dequate staffing to provide 

initial facility accident response in key 

functional areas is maintained at all 

times, timely augmentation of response 

capabilities is available. . . .’’ 

• Appendix E to 10 CFR part 50, 

paragraph IV.A, which states, in part, 

‘‘The organization for coping with 

radiological emergencies shall be 

described, including definition of 

authorities, responsibilities, and duties 

of individuals assigned to the licensee’s 

emergency organization. . . .’’ 

This proposed rule would allow a 

licensee transitioning to a PSEP to 

revisit staffing levels and the staffing 

analysis for the ERO performed under 

paragraph IV.A.9 of appendix E to 10 

CFR part 50 to align staffing with the 

reduced spectrum of credible accidents 

for a permanently shutdown and 

defueled nuclear power reactor facility. 

The proposed requirement in § 50.200(a) 

would acknowledge that the spectrum 

of credible accidents requiring a 

response from the ERO at a facility that 

is permanently shutdown and defueled 

is reduced as compared to that for an 

operating plant. The principal public 

safety concern involves the potential 

radiological risks associated with the 

storage of spent fuel on site in the SFP. 

For example, the reactor, reactor coolant 

system, and reactor support systems are 

no longer in operation and have no 

function related to the storage of spent 

fuel. Therefore, postulated accidents 

involving a failure or malfunction of 

these systems are no longer applicable. 

As such, certain ERO positions and 

emergency functions as detailed in 

NUREG–0654/FEMA–REP–1, Revision 

2, Table B–1, ‘‘Emergency Response 

Organization (ERO) Staffing and 

Augmentation Plan,’’ may not be 

applicable or necessary under a PSEP. 

Commensurate with the reduced 

spectrum of credible accidents, 

proposed § 50.200(a) would allow 

licensees to change ERO staffing levels 

required by existing § 50.47(b)(2) within 

their PSEPs. Reductions in facility 

staffing may be made as long as the 

facility operates with no loss of 

necessary EP functions and the 

reductions have no impact on the formal 

offsite radiological emergency response 

plans that are in effect. In conjunction 

with this proposed rule, the NRC is 

issuing for public comment DG–1346, 

which provides guidance on ERO 

capabilities to be maintained at facilities 

with PSEPs when reducing staffing 

levels. 

PSEP Emergency Action Levels 

Currently, appendix E to 10 CFR part 

50, paragraph IV.C requires licensees to 

develop a set of emergency action levels 

(EALs) based not only on onsite and 

offsite radiation monitoring information 

but also on readings from a number of 

sensors that indicate a potential 

emergency, such as the pressure in 

containment and the response of the 

emergency core cooling system. This 

proposed rule would allow licensees 

transitioning to a PSEP to revise EALs 

consistent with the profile of a 

permanently shutdown and defueled 

nuclear power reactor. Proposed 

§ 50.54(q)(8)(iii) would state that 

changes to EALs resulting from changes 

in plant conditions due to the transition 

to decommissioning would not be 

reductions in effectiveness provided 

that the evaluation under § 50.54(q)(3) 

demonstrates that the changes do not 

reduce the capability of the licensee to 

take timely and appropriate protective 

actions. Given the defueled nature of 

facilities in decommissioning, EALs 

associated with nuclear power reactor 

operations (e.g., reactor vessel water 

level, core temperature, and 

containment radiation levels) and EALs 

for mitigation systems not associated 

with the SFP would no longer contain 

applicable initiating conditions. 

Containment parameters do not indicate 

the conditions relevant to EP at a 

defueled facility, and emergency core 

cooling systems would no longer be 

required. Other indications such as SFP 

level or temperature can be used at sites 

that have spent fuel in the SFPs. 

Consistent with existing requirements, 

licensees transitioning to a PSEP would 

still be required to maintain a set of 

EALs based on onsite radiation 

monitoring information and in-plant 

conditions and instrumentation 

applicable to EP for a defueled reactor. 

Guidance document NEI 99–01, 

Revision 6, ‘‘Development of Emergency 

Action Levels for Non-Passive Reactors’’ 

(ADAMS Accession No. ML12326A805), 

provides EALs for non-passive operating 

nuclear power reactors, permanently 

defueled reactors, and ISFSIs. The NRC 

found NEI 99–01, Revision 6, acceptable 

for use in a letter dated March 28, 2013 

(ADAMS Accession No. ML12346A463). 

To accompany this proposed rule, the 

NRC drafted guidance in Attachment 1 

of Appendix A in DG–1346, for how a 

permanently shutdown and defueled 

nuclear power reactor facility could 

make a partial EAL scheme change. 

Notwithstanding the proposed changes 

to § 50.54(q), a licensee desiring to 

change its entire EAL scheme must 

receive prior NRC approval in 

accordance with appendix E to 10 CFR 

part 50, paragraph IV.B.2. 

PSEP Evacuation Time Estimate Studies 

Appendix E to 10 CFR part 50, 

paragraph IV.3 requires licensees to use 

evacuation time estimates (ETEs) in the 

formulation of protective action 

recommendations (PARs) and to provide 

the ETEs to State and local 

governmental authorities for use in 

developing offsite protective action 

strategies. Licensees must update ETEs 

on a periodic basis in accordance with 

the requirements in § 50.47(b)(10) and 

appendix E to 10 CFR part 50, 

paragraphs IV.4, IV.5, and IV.6. The 

periodicity of these updates together 

with time needed to develop and 

implement the resulting protective 

action strategies may exceed the 

expected transition period covered by 

PSEPs. Therefore, the NRC is proposing 

to add a new paragraph IV.8 to 

appendix E to 10 CFR part 50 to clarify 

that the ETE requirements of paragraphs 

IV.4, IV.5, and IV.6 would no longer be 

applicable to licensees after permanent 

cessation of operations and permanent 

removal of fuel from the reactor vessel. 

Existing ETE analyses would remain 

effective within the emergency plan 

until no longer required for licensees 

with PDEPs. 

Under proposed § 50.54(q)(7)(ii), a 

licensee transitioning to a PSEP would 

need to maintain a PSEP from the date 

that the NRC dockets the licensee’s 

certifications of permanent cessation of 

operations and permanent removal of 

fuel from the reactor vessel, until the 

spent fuel has decayed for a period of 

at least 10 months (for BWRs) or 16 

months (for PWRs) from the date of 

permanent cessation of operations for 


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burnups less than 72 GWd/MTHM, 

unless an alternative spent fuel decay 

period is proposed by the licensee and 

approved by the NRC. For fuel with 

burnups greater than 72 GWd/MTHM or 

non-zirconium cladding, an alternative 

spent fuel decay period would be 

proposed by the licensee for approval by 

the NRC under § 50.54(q)(7)(ii). Updates 

to the ETE during this level of 

decommissioning would provide 

limited benefit for the enhancement of 

protective action strategies or offsite 

evacuation planning. Even if the criteria 

for updating the ETE analysis were met 

within the timeframe for a PSEP, 

updating an ETE report may take several 

months of analysis. After the ETE is 

updated, the regulations in appendix E 

to 10 CFR part 50, paragraph IV.6 

require an additional 180 days before an 

updated ETE can be used to inform 

PARs and offsite protective action 

strategies. The additional time and effort 

needed to develop and implement a 

revised protective action strategy may 

exceed the time that a facility would 

spend with a PSEP before transitioning 

to a PDEP. Based on the NRC’s review 

of submitted ETEs, population changes 

within a period comparable to the postshutdown timeframe are unlikely to 

impact ETEs enough to affect the 

formulation of protective action 

strategies. In addition, because licensees 

with PDEPs would not be required to 

have preplanned PARs to provide for a 

prompt response to a radiological 

emergency, updates to the ETE postshutdown would provide no significant 

benefit. 

PSEP Annual Dissemination of Public 

Information 

Appendix E to 10 CFR part 50, 

paragraph IV.D.2 currently requires 

licensees to make an annual 

dissemination of basic emergency 

planning information to the public 

within the plume exposure pathway 

EPZ. Section II.G of NUREG–0654/ 

FEMA–REP–1, Revision 2, contains 

criteria for the information that should 

be included in the annual dissemination 

of public information, including 

educational information on radiation, 

points of contact, protective measures, 

and information for special needs 

populations. The NRC is not proposing 

changes related to the requirement for 

an annual dissemination of public 

information for a PSEP because the 

change in the plant’s operating status 

and the ensuing changes to the EP 

program would be appropriate 

information to communicate to the 

public. However, consistent with the 

removal of regulatory standards for 

offsite radiological emergency plans for 

decommissioning sites (including the 

removal of EPZ requirements) as 

discussed later in this document, 

licensees with PDEPs would not be 

required to provide annual 

disseminations of information to the 

public. In DG–1346, the NRC provides 

guidance on one method acceptable to 

the NRC for a final dissemination of 

information to the public for licensees 

with PSEPs. 

PSEP Hostile Action 

In the 2011 final rule, ‘‘Enhancements 

to Emergency Preparedness 

Regulations’’ (76 FR 72559; November 

23, 2011) (2011 EP Final Rule), the NRC 

amended its regulations to include 

enhancements to EP in response to a 

hostile action event. Appendix E to 10 

CFR part 50, paragraph IV.A.7 defines 

‘‘hostile action’’ as an act directed 

toward a nuclear power plant or its 

personnel that includes the use of 

violent force to destroy equipment, take 

hostages, and/or intimidate the licensee 

to achieve an end. Appendix E to 10 

CFR part 50, paragraph IV.B.1 requires 

nuclear power reactor licensees to have 

EALs for hostile action, paragraph 

IV.E.8.d requires nuclear power reactor 

licensees to have alternative facilities 

that would be accessible even if the site 

is under threat of or experiencing 

hostile action for the staging of ERO 

personnel, paragraph IV.l requires 

nuclear power reactor licensees to 

develop protective actions to protect 

onsite personnel during hostile action, 

and paragraph IV.F.2.c.4 and paragraph 

IV.F.2.i require nuclear power reactor 

licensees to have hostile action 

scenarios in drills and exercises. These 

EP requirements related to hostile action 

are separate and distinct from the 

physical protection regulations in 10 

CFR part 73, ‘‘Physical Protection of 

Plants and Materials.’’ 

The NRC is proposing to maintain EP 

requirements related to hostile action 

for nuclear power reactor licensees 

transitioning to a PSEP. Spent fuel at a 

nuclear power reactor facility that has a 

PSEP has not yet undergone a 

significant period of decay, 

necessitating the maintenance of formal 

offsite radiological emergency planning. 

The potential consequences and timing 

of an accident are the primary 

considerations for the EP planning basis 

at nuclear power reactor facilities 

transitioning to a PSEP. Although 

NUREG–1738 did not evaluate the 

potential consequences of a sabotage 

event that could directly cause offsite 

fission production dispersion, the NRC 

did study the potential consequences of 

the zirconium fire event at different 

spent fuel decay times. Within the 

timeframe proposed for nuclear power 

reactor facilities transitioning to a PSEP, 

the study in NUREG–1738 shows that 

decay time is significant when 

considering short-term radiological 

consequences. Additionally, 

maintaining EP requirements related to 

hostile action during this transitional 

(and time-limited) level of 

decommissioning would help both the 

licensee and offsite response 

organizations (OROs) avoid immediate 

significant changes to the onsite and 

offsite emergency plans. 

PSEP Drills and Exercises 

Current regulations in appendix E to 

10 CFR part 50, paragraph IV.F and 

§ 50.47(b)(14) include requirements for 

periodic drills and exercises for nuclear 

power reactor licensees. Proposed 

paragraph IV.F.2.k would require 

licensees to follow the biennial exercise 

requirements of appendix E, paragraph 

IV.F.2 once the NRC dockets the 

licensee’s certifications required under 

§ 50.82(a)(1) or § 52.110(a). After the 

NRC dockets this certification, exercise 

scenarios would be reduced 

commensurate with the permanent 

cessation of operations and permanent 

removal of fuel from the reactor vessel 

to reflect a smaller suite of potential 

accident scenarios. 

Current regulations in appendix E to 

10 CFR part 50, paragraph IV.F.2.c 

require that offsite radiological 

emergency plans for each site be 

exercised biennially with full 

participation by each offsite authority 

having a role under the radiological 

emergency plan. Proposed paragraph 

IV.F.2.k would provide that biennial 

exercises of offsite emergency plans 

would be required after the NRC dockets 

a licensee’s certifications under 

§ 50.82(a)(1) or § 52.110(a) until 

transition to a PDEP. 

However, a licensee that conducts a 

full participation biennial exercise just 

prior to the NRC docketing the 

licensee’s certifications required under 

§ 50.82(a)(1) or § 52.110(a) may not be 

required to conduct another exercise 

before transitioning to a PDEP. If an 

exercise is conducted as part of the 8- 

year exercise cycle, as required under 

appendix E to 10 CFR part 50, paragraph 

IV.F.2.j, after the NRC dockets the 

licensee’s certifications required under 

§ 50.82(a)(1) or § 52.110(a), but prior to 

transitioning to a PDEP, the scenario 

would reflect actual plant conditions. 

PSEP Emergency Response Data 

Systems 

Appendix E to 10 CFR part 50, section 

VI, ‘‘Emergency Response Data System,’’ 

outlines a set of system, testing, and 


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implementation requirements for the 

emergency response data system 

(ERDS). These systems transmit nearreal-time electronic data directly 

between the licensee’s onsite computer 

system and the NRC Operations Center. 

Nuclear power facilities that are 

shutdown permanently or indefinitely 

are currently not required to provide 

hardware to interface with the NRC 

receiving system under appendix E to 

10 CFR part 50, paragraph VI.2, and the 

NRC is not proposing any regulatory 

changes to section VI beyond minor 

corrections (see ‘‘Clean-up of 

Regulations’’ section in this document). 

Under § 50.72, ‘‘Immediate notification 

requirements for operating nuclear 

power reactors,’’ licensees with PSEPs 

would maintain a capability to provide 

meteorological, radiological, and SFP 

data (e.g., level, flow, and temperature 

data) to the NRC within a reasonable 

timeframe following an event. 

5. Permanently Defueled Emergency 

Plans 

Proposed § 50.54(q)(7)(ii) describes 

the timeframe after which a licensee 

would be permitted to transition to a 

PDEP. As discussed in the ‘‘Technical 

Basis for Graded Approach’’ section of 

this document, the NRC concluded that 

after a decay period of 10 months (for 

BWRs) or 16 months (for PWRs), the 

spent fuel cannot reasonably heat up to 

the zirconium fuel cladding ignition 

temperature within 10 hours. Therefore, 

the NRC is proposing that a licensee can 

transition to a PDEP after the NRC’s 

docketing of the licensee’s certifications 

of permanent cessation of operations 

and permanent removal of all fuel from 

the reactor vessel pursuant to 

§§ 50.82(a)(1) or 52.110(a) and when at 

least 10 months (for BWR) or 16 months 

(for PWR) have elapsed since the date of 

permanent cessation of operations. 

Proposed § 50.54(q)(7)(ii) would also 

allow licensees to submit an analysis for 

NRC approval demonstrating that an 

alternative spent fuel decay period 

would ensure that spent fuel would not 

heat up to 900 degrees C in less than 10 

hours under adiabatic conditions. Under 

the proposed rule, licensees would be 

required to submit this analysis under 

§ 50.90 and the analysis would need to 

be approved by the NRC in order for a 

licensee to transition to a PDEP in less 

than 10 months (for a BWR) or 16 

months (for a PWR). While the NRC’s 

research conducted to inform this 

proposed rule supports a required decay 

period of 10 months (for BWRs) or 16 

months (for PWRs), it is possible that a 

licensee may be able to demonstrate, 

based on site-specific conditions, that a 

shorter decay period would still ensure 

that spent fuel cannot reasonably heat 

up to the zirconium fuel cladding 

ignition temperature within 10 hours; 

therefore, the NRC is allowing for the 

flexibility to submit an alternative decay 

period under proposed § 50.54(q)(7)(ii). 

The NRC is issuing DG–1346 for public 

comment in conjunction with this 

proposed rule; DG–1346 provides one 

method acceptable to the NRC for 

conducting the spent fuel heatup 

analysis. 

As demonstrated in the results of the 

NRC’s task analysis of mitigation 

actions, ‘‘A Human Reliability Analysis 

of the Spent Fuel in the Spent Fuel Pool 

of Decommissioning Nuclear Plants’’ 

(ADAMS Accession No. ML16110A432), 

a period of 10 hours will provide 

sufficient time for plant staff to 

implement mitigation strategies to 

prevent spent fuel heatup damage. 

Additionally, as noted in the NRC’s 

analysis, ‘‘Offsite Dose Accumulation 

Rates Following a Hypothetical Spent 

Fuel Pool Accident’’ (ADAMS 

Accession No. ML16110A430), even in 

the event of a highly unlikely beyonddesign-basis accident leading to a rapid 

draindown of the SFP and subsequent 

zirconium fire, there may be an 

additional time margin of several hours 

beyond the 10-hour heatup time during 

which protective actions can be taken to 

protect the public before the dose levels 

associated with EPA PAGs would be 

exceeded offsite. Because of the 

additional time available to take 

mitigation actions and, if necessary, to 

initiate protective actions, many 

requirements applicable to licensees 

with PSEPs would not be applicable to 

licensees with sufficiently decayed 

spent fuel (i.e., licensees with PDEPs). 

The following discussion addresses the 

planning standards under proposed 

§ 50.200(b) and requirements under 

proposed § 50.200(c) that would be 

necessary to adequately protect public 

health and safety at facilities with 

PDEPs. The proposed requirements for 

facilities with PDEPs are consistent with 

the guidance contained in NSIR/DPR– 

ISG–02. 

Offsite Radiological Emergency 

Response Plans 

Currently, § 50.47(b) applies to both 

onsite and offsite radiological 

emergency response plans, and 

appendix E to 10 CFR part 50 includes 

requirements for emergency plans to 

address offsite emergency response 

capabilities (e.g., public alert and 

notification systems, offsite PAR 

development, ETEs, and exercises of 

offsite emergency plans). Under this 

proposed rule, NRC planning standards 

would no longer be applied to offsite 

radiological emergency response plans 

for plants with PDEPs. 

In its review of several exemption 

requests, the NRC concluded that as 

long as a period of at least 10 hours is 

available to implement mitigation 

measures or initiate appropriate 

response actions offsite, formal offsite 

radiological emergency plans, required 

under 10 CFR part 50, are not necessary 

for permanently shutdown and defueled 

nuclear power reactor licensees with a 

PDEP. In a hypothetical SFP accident 

scenario, 10 hours is a conservative 

estimate of the amount of time available 

to implement mitigation measures or to 

take other appropriate response actions. 

The 10 hours assumes that the spent 

fuel begins to heat up immediately after 

the initiating event occurs and does not 

include the expected amount of time it 

would take for water to drain from the 

pool. A beyond-design-basis accident 

that results in the water draining from 

the pool (whether a full or partial 

draindown) would likely take much 

longer than 10 hours because of the 

robust construction of the SFP and the 

large volume of water in the SFP, 

delaying the onset of heatup. 

Additionally, 10 hours is a conservative 

period of time during which preplanned 

mitigation measures to provide makeup 

water or spray to the SFP can be 

implemented reliably before the onset of 

a zirconium cladding ignition. 

If a release is projected to occur, 10 

hours would be sufficient time for 

licensees to notify offsite agencies and 

for these agencies to initiate appropriate 

action to protect public health and 

safety. The NRC concludes that 10 hours 

provides ample time to take appropriate 

actions without the extensive 

preplanning and other requirements of 

the EP framework for operating plants, 

and, therefore, regulatory standards for 

offsite radiological emergency plans 

would no longer be necessary for the 

adequate protection of public health and 

safety. Licensees with PDEPs would still 

maintain a variety of onsite capabilities 

that may be available to support OROs 

in EP and response, including 

radiological training; regular 

coordination with OROs; radiological 

assessment capabilities; memoranda of 

understanding for firefighting, law 

enforcement, and ambulance/medical 

services; and the ability to make PARs 

upon request. For licensees with PDEPs, 

no action would be expected or required 

from State or local government 

organizations in response to an event at 

a decommissioning site other than 

firefighting, law enforcement, and 

ambulance/medical services. 

Requirements for licensees to maintain 

agreements for these services also exist 


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outside of radiological EP, including the 

requirement for licensees to maintain a 

fire protection plan in § 50.48, ‘‘Fire 

protection,’’ and physical security 

requirements in 10 CFR part 73. Since 

the requirements of § 50.47(b) continue 

to apply to offsite radiological 

emergency plans during 

decommissioning, the NRC is proposing 

to add § 50.47(f) to clarify when the 16 

planning standards in § 50.47(b) no 

longer apply to offsite radiological 

emergency plans. 

PDEP Staffing and Emergency Response 

Organization 

Currently, § 50.47(b)(1) and (2) and 

paragraph IV.A of appendix E to 10 CFR 

part 50 require licensees to maintain 

adequate staffing for initial and 

augmented response in the case of an 

emergency and to describe ERO 

responsibilities in their emergency 

plans. Further, appendix E to 10 CFR 

part 50, paragraph IV.A.9 requires 

licensees to conduct a detailed staffing 

analysis demonstrating that on-shift 

personnel assigned emergency plan 

implementation functions are not 

assigned responsibilities that would 

prevent the timely performance of their 

assigned functions as specified in the 

emergency plan. 

Proposed § 50.200(b)(1), (b)(2), and 

(c)(1)(i) would include similar staffing 

requirements for licensees with PDEPs, 

with the exception of changes made to 

reflect the small staffing levels required 

at a decommissioning facility and the 

removal of formal offsite radiological 

emergency response requirements for 

licensees with PDEPs. For example, 

licensees with PDEPs would not have to 

comply with the requirement under 

appendix E to 10 CFR part 50, paragraph 

IV.A.3 to augment the ERO with staff 

from licensee headquarters. Because of 

the much lower risk and much slower 

progression of events as compared to 

operating plants, decommissioning sites 

typically have a level of emergency 

response that does not require response 

by headquarters personnel. Licensees 

would not have to identify State and/or 

local officials responsible for protective 

actions, as currently required under 

appendix E to 10 CFR part 50, paragraph 

IV.A.8 because offsite emergency 

measures are limited to onsite support 

provided by local police, fire 

departments, and ambulance and 

hospital services, as appropriate. 

Proposed § 50.200(c)(1)(i) would require 

licensees with PDEPs to include in their 

emergency plans plant staff emergency 

assignments. 

In addition, the staffing analysis 

required under appendix E to 10 CFR 

part 50, paragraph IV.A.9 would no 

longer apply to licensees with PDEPs. In 

the 2011 EP Final Rule, the NRC 

concluded that the staffing analysis 

requirement was not necessary for nonpower reactor licensees because of the 

small staffing levels required for those 

facilities. For this same reason, licensees 

with PDEPs would no longer be 

required to perform this analysis under 

the proposed rule. 

As licensees transition to a PDEP, 

staffing levels may be reduced but must 

remain commensurate with the need to 

safely store spent fuel at the facility in 

a manner that is protective of public 

health and safety. The NRC is issuing 

DG–1346 for public comment in 

conjunction with this proposed rule; 

DG–1346 provides guidance on ERO 

staffing levels for a PDEP. Licensees 

with PDEPs would need to be able to 

augment on-shift capabilities within two 

hours after declaration of an emergency. 

The augmented staff would need to 

include engineering capability 

appropriate for SFP accident mitigation, 

but may otherwise be reduced. 

Currently, a licensee is required to 

maintain staffing levels at its technical 

support center (TSC), operational 

support center (OSC), and emergency 

operations facility (EOF). In accordance 

with NUREG–0696, ‘‘Functional Criteria 

for Emergency Response Facilities’’ 

(ADAMS Accession No. ML051390358), 

a TSC is an onsite facility located close 

to the control room that provides plant 

management and technical support to 

the reactor operating personnel located 

in the control room during emergency 

conditions; the OSC is an onsite area 

separate from the control room and the 

TSC where licensee operations support 

personnel will assemble in an 

emergency; and an EOF is an offsite 

support facility for the management of 

overall licensee emergency response 

(including coordination with Federal, 

State, and local officials), coordination 

of radiological and environmental 

assessments, and determination of 

recommended public protective actions. 

Because of the low probability of DBAs 

or other credible events that would be 

expected to exceed the EPA PAGs offsite 

and the available time to implement 

mitigation measures consistent with 

plant conditions and, if necessary, to 

initiate response actions, licensees with 

PDEPs would not need to maintain the 

TSC, OSC, and EOF designated staff or 

dedicated offsite dose assessment field 

teams. 

PDEP Emergency Classification Levels 

and Emergency Action Levels 

Currently, § 50.47(b)(4) and appendix 

E to 10 CFR part 50, paragraphs IV.B 

and IV.C specify the EAL and 

emergency classification level (ECL) 

requirements for operating reactors. 

Similar to § 50.47(b)(4), the proposed 

PDEP planning standard under 

§ 50.200(b)(4) would require licensees 

with PDEPs to establish a standard ECL 

and EAL scheme, the bases of which 

would include facility system and 

effluent parameters. The NRC is 

proposing EAL and ECL requirements 

for licensees with PDEPs that are 

analogous to appendix E to 10 CFR part 

50, paragraphs IV.B and IV.C with the 

exceptions of the requirements to base 

EALs on offsite monitoring information 

and the appendix E to 10 CFR part 50 

paragraph IV.B.1 requirement to include 

hostile action-based EALs. Because 

licensees with PDEPs would not be 

required to maintain formal offsite 

radiological emergency response plans 

and ‘‘hostile action’’ does not apply (see 

discussion in ‘‘PDEP Hostile Action’’ 

and ‘‘Offsite Radiological Emergency 

Response Plans’’ sections in this 

document), these requirements are no 

longer relevant to these facilities. 

However, EALs for security-based 

events would still be required. 

Under proposed § 50.200(c)(1)(ii)(A), 

licensees with PDEPs would continue to 

be required to describe in their 

emergency plans the EALs that are used 

as a criterion for determining the need 

for notification and participation of 

governmental agencies and the EALs 

that are used for determining when and 

what protective measures should be 

considered within the site boundary to 

protect public health and safety. In 

addition, licensees with PDEPs would 

be required to review EALs with State 

and local governmental authorities on 

an annual basis. Under proposed 

§ 50.200(c)(1)(iii)(A), licensees with 

PDEPs would continue to be required to 

describe in their emergency plans the 

spectrum of emergency conditions that 

involve the alerting or activating of the 

total emergency organization, the 

communication steps to be taken to alert 

or activate personnel, EALs for 

notification of offsite agencies, and the 

existence of a message authentication 

scheme. Under proposed 

§ 50.200(c)(1)(ii)(B), a licensee desiring 

to make an EAL scheme change as part 

of the PDEP must follow the 

requirements of appendix E to 10 CFR 

part 50, paragraph IV.B.2. 

For facilities with PDEPs, proposed 

§ 50.200(c)(1)(iii)(A) would specify that 

only the ECLs of Notification of Unusual 

Event and Alert would apply (and not 

the ECLs of Site Area Emergency and 

General Emergency, which apply to 

operating reactors). For these facilities, 

the probability of a condition reaching 

the level above an emergency 


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classification of Alert is very low. In the 

event of an accident at a facility with a 

PDEP, time will be available to 

implement mitigation measures 

consistent with plant conditions. As 

stated in NUREG–1738, small SFP leaks 

or loss of cooling scenarios evolve very 

slowly and generally leave many days 

for recovery efforts. Offsite radiation 

monitoring would be performed as the 

need arises. Because of the low 

probability of DBAs or other credible 

events that would reasonably be 

expected to exceed the EPA PAGs and 

the available time to implement 

mitigation measures consistent with 

plant conditions and, if necessary, to 

initiate appropriate response actions 

offsite, facilities with PDEPs would not 

require declarations of Site Area 

Emergency and General Emergency and 

the associated offsite radiation 

monitoring systems. The results from 

the NRC’s analyses previously discussed 

support this conclusion. 

Consistent with the discussion on 

PSEPs, EALs for nuclear power reactor 

operations (e.g., reactor vessel water 

level, core temperature, and 

containment radiation levels) and EALs 

related to mitigation systems not 

associated with the SFP would no 

longer be applicable for facilities with 

PDEPs. The NRC is issuing DG–1346 for 

public comment in conjunction with 

this proposed rule; DG–1346 provides 

one method acceptable to the NRC for 

EALs for facilities with PDEPs. As 

discussed previously, proposed 

§ 50.54(q)(8)(iii) describes requirements 

for decommissioning licensees to 

conduct reduction in effectiveness 

determinations for EAL schemes. 

PDEP Emergency Assessment, 

Classification, and Declaration 

Currently, appendix E to 10 CFR part 

50, paragraph IV.C.2 requires licensees 

to maintain the capability to assess, 

classify, and declare an emergency 

condition within 15 minutes. A 

decommissioning nuclear power reactor 

has a low likelihood of a design-basis 

accident or other credible event 

resulting in radiological releases 

requiring offsite protective measures, 

and the event progression is much 

slower compared to that for operating 

reactors. For these reasons, under this 

proposed rule licensees with PDEPs 

would not be required to assess, 

classify, and declare an emergency 

condition within 15 minutes. Instead, 

the NRC is proposing under 

§ 50.200(c)(1)(iii)(B) that licensees with 

PDEPs must document and maintain the 

capability to assess, classify, and declare 

an emergency condition as soon as 

possible and within 60 minutes after the 

availability of indications that an EAL 

has been exceeded and must promptly 

declare the emergency condition as soon 

as possible following identification of 

the appropriate ECL. Similar to the 

requirements in appendix E to 10 CFR 

part 50, paragraph IV.C, proposed 

§ 50.200(c)(1)(iii)(B) would clarify that 

PDEP licensees must not treat the 

timeframe as a grace period or delay the 

implementation of response actions. 

The 60-minute timeframe is 

commensurate with the slower 

progression of a credible event resulting 

in a radiological release requiring offsite 

protective measures (see discussion of 

the timeframe for potential releases and 

mitigation actions at decommissioning 

sites in the section ‘‘Permanently 

Defueled Emergency Plans’’ in this 

document). 

PDEP Notification Requirement to State 

and Local Governmental Agencies 

Currently, appendix E to 10 CFR part 

50, paragraph IV.D.3 requires licensees 

to have the capability to notify OROs of 

an emergency declaration within 15 

minutes. Under proposed 

§ 50.200(c)(1)(iv)(B), licensees with 

PDEPs would be required to promptly 

notify State and local governmental 

agencies and to make this notification as 

soon as possible and within 60 minutes 

after declaring an emergency. The NRC’s 

research and analysis shows that 

licensees with PDEPs would have 

sufficient time to implement mitigation 

measures consistent with plant 

conditions and, if necessary, for OROs 

to initiate protective actions offsite. 

Notifying OROs as soon as possible and 

within 60 minutes after declaring an 

emergency would not significantly 

impact the time available for OROs to 

initiate appropriate response actions. 

PDEP Public Alert and Notification 

Systems 

Currently, appendix E to 10 CFR part 

50, paragraph IV.D.3 requires licensees 

to demonstrate that appropriate 

governmental authorities have the 

capability to make a decision on alerting 

and notifying the public promptly on 

being informed of an emergency 

condition. Because of the low 

probability of DBAs or other credible 

events that would be expected to exceed 

the limits of EPA PAGs offsite and the 

available time for event mitigation, 

under this proposed rule, the public 

alert and notification system specified 

in appendix E to 10 CFR part 50, 

paragraph IV.D.3 would not be required 

for licensees with PDEPs. Similarly, 

exercises of this system, as required 

under appendix E to 10 CFR part 50, 

paragraph IV.F.2, would no longer be 

required for licensees with PDEPs. As 

previously discussed, licensees with 

PDEPs would still be required to 

maintain the capability to notify 

responsible State and local 

governmental agencies within 60 

minutes after declaring an emergency, 

and, based on research and analysis 

showing that there would be at least 10 

hours prior to a zirconium fuel cladding 

fire for licensees with PDEPs, sufficient 

time would be available for appropriate 

governmental authorities to inform the 

public and initiate protective actions, if 

necessary. Such actions would be 

within the capabilities of offsite 

response organizations and would be 

similar to actions required for other 

hazards that do not require a dedicated 

hazard-specific offsite response 

capability as is the case for operating 

reactors. 

PDEP Emergency Planning Zones 

Currently, § 50.47(b) and (c)(2) require 

licensees to conduct emergency 

planning for both the shorter-term 

plume exposure pathway EPZ (generally 

10 miles) and the longer-term ingestion 

exposure pathway EPZ (generally 50 

miles). Appendix E to 10 CFR part 50 

contains additional emergency planning 

requirements for these two types of 

EPZs. However, the maintenance of the 

plume exposure pathway and ingestion 

exposure pathway EPZs for licensees 

with PDEPs is not warranted because of 

the low probability of DBAs or other 

credible events that would be expected 

to exceed the EPA PAGs off site and the 

available time to implement mitigation 

measures. Additionally, if necessary, 

sufficient time would be available for 

OROs to initiate appropriate response 

actions even for a highly unlikely severe 

accident. Therefore, consistent with the 

NRC’s determination to not require the 

establishment of formal offsite 

radiological emergency response plans 

for licensees with PDEPs, the NRC is 

proposing to eliminate the requirements 

that EPZs be maintained for licensees 

with PDEPs. In other words, the plume 

exposure pathway EPZ for licensees 

with PDEPs does not exceed the site 

area boundary. Consequently, the 

planning standards for PDEPs under 

proposed § 50.200(b) and the 

requirements under proposed 

§ 50.200(c) do not include references to 

the EPZs. 

The NRC is also proposing to add a 

new paragraph (f) to § 50.47 that would 

clarify that the planning standards of 

§ 50.47(b) do not apply to offsite 

radiological emergency response plans 

if the licensee’s emergency plan is not 

required to meet these planning 

standards or if the plume exposure 


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pathway EPZ does not exceed the site 

area boundary. 

PDEP Offsite Radiological Protective 

Action Recommendations 

Currently, § 50.47(b) requires 

licensees to develop a range of 

protective actions for the plume 

exposure pathway EPZ for emergency 

workers and the public and to give 

consideration to evacuation, sheltering, 

and the use of potassium iodide. 

Licensees also must develop and put in 

place guidelines for the choice of 

protective actions during an emergency 

and develop protective actions for the 

ingestion exposure pathway EPZ. 

Proposed § 50.200(b)(10) would require 

licensees with PDEPs to continue to 

develop a range of protective actions for 

emergency workers and the public but, 

consistent with the removal of 

regulatory standards for offsite 

radiological EP for these licensees, 

would not reference specific offsite 

protective actions or pre-planned 

activities for the public in the EPZs. The 

proposed requirement would call for 

protective actions directed at emergency 

workers who may have to respond to the 

decommissioning site for firefighting, 

law enforcement, and ambulance/ 

medical services and members of the 

public present within the ownercontrolled area during a radiological 

emergency. 

For licensees with PDEPs, preplanned offsite protective actions to 

ensure a prompt response to a 

radiological emergency on site are not 

necessary given the time available for 

OROs to initiate appropriate response 

actions. Although the likelihood is low 

for events that would result in doses in 

excess of the EPA PAGs to the public 

beyond the owner-controlled area 

boundary based on the permanently 

shutdown and defueled status of the 

reactor, the proposed rule would require 

licensees with PDEPs to determine the 

magnitude of and continually assess the 

impact of a radiological release under 

proposed § 50.200(c)(1)(ii)(A), and, if a 

release is occurring, the licensee would 

be required to communicate that 

information to offsite authorities as soon 

as possible for their consideration in 

taking appropriate response actions 

under proposed § 50.200(c)(1)(iv)(B). 

In 2001, the NRC revised its EP 

regulations through the ‘‘Consideration 

of Potassium Iodide in Emergency 

Plans’’ (66 FR 5427; January 19, 2001) 

final rule to include the consideration of 

potassium iodide as a protective 

measure for the general public to 

supplement sheltering and evacuation 

in the unlikely event of a severe nuclear 

power plant accident with an offsite 

radioactive plume that would include 

radioactive iodine. For licensees with 

PDEPs, in addition to not needing preplanned protective action strategies, the 

iodine in the spent fuel has decayed 

sufficiently such that there is no need to 

consider a supplemental potassium 

iodide program to counteract the effects 

of radioactive iodine on the thyroid. 

PDEP Evacuation Time Estimate Studies 

Currently, licensees are required to 

develop and update ETEs in accordance 

with the requirements in § 50.47(b) and 

appendix E to 10 CFR part 50, paragraph 

IV.3. Paragraph IV.3 requires licensees 

to use ETEs in the formulation of PARs 

and to provide ETEs to State and local 

governmental authorities for use in 

developing offsite protective action 

strategies. Because of the low 

probability of DBAs or other credible 

events that would be expected to exceed 

the limits of EPA PAGs offsite and the 

available time for event mitigation, as 

well as the minimal expected offsite 

response required, the proposed rule 

would not require licensees with PDEPs 

to maintain ETEs (see section ‘‘PSEP 

Evacuation Time Estimate Studies’’ in 

this document for additional discussion 

regarding the need for ETEs postshutdown). 

PDEP Emergency Facilities and 

Equipment 

Currently, appendix E to 10 CFR part 

50, paragraph IV.E requires licensees to 

maintain and describe adequate 

provisions for emergency facilities and 

equipment, including equipment at the 

site for personnel monitoring, 

equipment for radiological assessment, 

facilities and supplies for 

decontaminating onsite individuals, 

first aid facilities and medical supplies, 

arrangements for qualified medical 

service providers and the transportation 

of contaminated injured individuals, 

and arrangements for the treatment of 

individuals injured in support of 

licensed activities. Decommissioning 

licensees have not received exemptions 

or license amendments for these 

requirements to date, and the NRC has 

determined that licensees with PSEPs 

and PDEPs would still need to maintain 

these capabilities under proposed 

§ 50.200(c)(1)(v). Appendix E to 10 CFR 

part 50, paragraph VI.E.8 further 

includes emergency response facility 

requirements for a TSC, OSC, and EOF. 

For licensees with PDEPs, there is no 

longer a need for separate, dedicated 

facilities. The functions of the control 

room, TSC, OSC, and EOF could be 

combined into one or more locations 

while still adequately protecting public 

health and safety. Proposed 

§ 50.200(c)(1)(v)(H) would require 

licensees with PDEPs to establish a 

facility from which effective direction 

can be given and effective control can be 

exercised during an emergency. Because 

of the low probability of DBAs or other 

credible events that would be expected 

to exceed the limits of EPA PAGs offsite 

and the available time for event 

mitigation, the significantly reduced 

staff, and the minimal expected 

response required, offsite response 

would not be required at an EOF. Onsite 

actions may be directed from the control 

room or other location, without the 

requirements imposed on a TSC or EOF. 

Proposed § 50.200(b)(3) would remove 

reference to the EOF as a location for 

response. Additionally, under this 

proposed rule, a separate OSC would no 

longer be required to meet its original 

purpose of an assembly area for plant 

logistical support during an emergency. 

The OSC function could be incorporated 

into another facility. The NRC is issuing 

DG–1346 for public comment in 

conjunction with this proposed rule; 

DG–1346 provides one acceptable 

method for meeting the proposed 

emergency response facility 

requirements for PDEPs. 

Appendix E to 10 CFR part 50, 

paragraph IV.E.9 addresses 

requirements for emergency 

communications systems, plans, and 

arrangements, including 

communications with OROs and 

between the control room, TSC, and 

EOF. Proposed § 50.200(c)(1)(v)(I) 

would require licensees with PDEPs to 

continue to maintain an onsite and an 

offsite communications system with 

backup power and communication 

plans with arrangements for 

emergencies. These arrangements would 

need to include provisions for 

communications with contiguous State 

and local governments, Federal 

emergency response organizations, NRC 

Headquarters, and the appropriate NRC 

Regional Office Operations Center. 

Because licensees with PDEPs may 

combine emergency response facilities, 

the current requirements for 

communication between emergency 

response facilities would not apply to 

these licensees. Under the proposed 

rule, communications with State and 

local emergency operations centers 

would be maintained to allow 

coordination of assistance onsite if 

required. 

PDEP Hostile Action 

Under this proposed rule, hostile 

action requirements would not apply to 

licensees with PDEPs. The definition of 

‘‘hostile action’’ in appendix E to 10 

CFR part 50, paragraph IV.A.7 applies 


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here to the capability of implementing 

EP during hostile action events. 

However, in the statement of 

considerations (SOC) for the 2011 EP 

Final Rule, the NRC excluded nonpower reactors from the definition of 

‘‘hostile action’’ because a non-power 

reactor as defined in § 50.2, 

‘‘Definitions,’’ is not a nuclear power 

plant, and a regulatory basis had not 

been developed to support the inclusion 

of non-power reactors in the definition 

of ‘‘hostile action.’’ A licensee with a 

PDEP would be similar to a non-power 

reactor in that both have a low 

likelihood of a credible accident 

resulting in radiological releases 

requiring response actions offsite. 

Additionally, regardless of how a 

disruption to the SFP cooling occurs, 

the spent fuel would take longer than 10 

hours to heat up to ignition temperature, 

providing adequate time to coordinate a 

response between the ERO and law 

enforcement officials. As such, licensees 

with PDEPs would not fall within the 

scope of ‘‘hostile action,’’ and 

enhancements to EP in response to 

hostile action, such as alternative 

facilities for the staging of ERO 

personnel, protection of onsite 

personnel, and challenging drills and 

exercises involving hostile action, 

would not be warranted. 

Although this rationale justifies the 

exclusion of licensees with PDEPs from 

the definition of ‘‘hostile action’’ and its 

related requirements (including 

conducting hostile action exercises) as 

they apply to EP, elements for securitybased events would still be maintained 

for these facilities, including EALs for 

security-based events. Under the 

proposed rule, licensees with PDEPs 

would be required to identify ORO 

resources that would respond to a 

security event, and the assistance 

licensees expect from those resources 

would be maintained in PDEPs. For 

physical security, the objective for these 

facilities relates to protection of the 

spent fuel against sabotage. A level of 

security commensurate with the 

consequences of a sabotage event is 

required and is evaluated on a sitespecific basis. The severity of the 

consequences declines as fuel ages and 

thereby removes over time the 

underlying concern that a sabotage 

attack, under the current definition, 

could cause offsite radiological 

consequences. 

PDEP Drills and Exercises 

Section 50.47(b)(14) and appendix E 

to 10 CFR part 50, paragraph IV.F 

provide training and drill and exercise 

requirements for nuclear power reactor 

licensees. Consistent with the language 

of § 50.47(b)(14), the proposed PDEP 

planning standard under § 50.200(b)(14) 

would require licensees with PDEPs to 

conduct periodic exercises to evaluate 

major portions of emergency response 

capabilities, to conduct periodic drills 

to develop and maintain key skills, and 

to correct deficiencies identified as a 

result of exercises and drills. The NRC 

is proposing new drill and exercise 

requirements for licensees with PDEPs 

under § 50.200(c)(1)(vi) that differ from 

the existing requirements under 

appendix E to 10 CFR part 50, paragraph 

IV.F to account for changes in principal 

functional areas, offsite radiological 

emergency response requirements, 

offsite PAR requirements, and the 

spectrum of possible accidents. 

Similar to the requirements in 

appendix E to 10 CFR part 50, paragraph 

IV.F.1, proposed § 50.200(c)(1)(vi)(A) 

would require licensees with PDEPs to 

describe in their emergency plan 

provisions for the training of employees, 

exercising the emergency plan by 

conducting periodic drills, and 

including other individuals in training 

and drills when those individuals may 

provide assistance in the event of a 

radiological emergency. Under the 

proposed rule, the emergency plan 

would be required to describe the 

training to be provided to several 

categories of emergency personnel, with 

the exception of licensees’ headquarters 

support personnel. Headquarters 

support personnel would no longer be 

required to augment the ERO for 

licensees with PDEPs. Licensees with 

PDEPs would need to continue to make 

available a radiological orientation 

training program for local services 

personnel expected to provide support 

onsite. Because of the time available to 

coordinate offsite agency notification to 

the public, licensees with PDEPs would 

not be required to provide radiological 

orientation training to local news media 

persons. Similar to the requirements in 

appendix E to 10 CFR part 50, paragraph 

IV.F.2, proposed § 50.200(c)(1)(vi)(B) 

would require licensees with PDEPs to 

continue to describe provisions for the 

conduct of EP exercises that test the 

adequacy of timing and content of 

implementing procedures and methods, 

test emergency equipment and 

communications networks, and ensure 

emergency organization personnel are 

familiar with their duties. Licensees 

with PDEPs would not be required to 

test the public alert and notification 

system during their exercises because 

the system would no longer be required, 

as discussed previously in this 

document. 

Proposed § 50.200(c)(1)(vi)(B)(1) and 

(2) would require licensees with PDEPs 

to conduct an exercise within two years 

of the last exercise of the onsite 

emergency plan conducted under 

paragraph IV.F.2.b of appendix E to 10 

CFR part 50 and to continue to conduct 

subsequent biennial exercises of onsite 

emergency plans. Licensees with PDEPs 

would need to continue to conduct 

drills during the intervals between 

biennial exercises involving a 

combination of principal functional 

areas. The principal functional areas of 

emergency response for licensees with 

PDEPs would include all of the areas 

currently listed under appendix E to 10 

CFR part 50, paragraph IV.F.2.b, with 

the exception of protective action 

development and protective action 

decision making (see discussion on 

protective action recommendations in 

the section ‘‘PDEP Offsite Radiological 

Protective Action Recommendations’’ in 

this document). 

Similar to the requirements in 

appendix E to 10 CFR part 50, paragraph 

IV.F.2.f, proposed 

§ 50.200(c)(1)(vi)(B)(4) would require 

licensees with PDEPs to conduct 

remedial exercises if the emergency 

plan is not satisfactorily tested during 

the biennial exercise. Like appendix E 

to 10 CFR part 50, paragraph IV.F.2.g, 

proposed § 50.200(c)(1)(vi)(B)(5) would 

require licensees with PDEPs to provide 

for formal critiques of exercises, drills, 

and training that provide performance 

opportunities to develop, maintain, or 

demonstrate key skills and to correct 

weaknesses or deficiencies identified in 

a critique. 

Proposed § 50.200(c)(1)(vi)(B)(6) 

would require licensees with PDEPs to 

continue to use drills and exercise 

scenarios that provide reasonable 

assurance that anticipatory responses 

will not result from preconditioning of 

participants and that emphasize 

coordination among onsite and offsite 

response organizations. Unlike the 

current requirements under appendix E 

to 10 CFR part 50, paragraphs IV.F.2.b, 

IV.F.2.i, and IV.F.2.j, licensees with 

PDEPs would not be required to submit 

exercise scenarios 60 days before use in 

an exercise, demonstrate that exercise 

scenarios include a wide spectrum of 

radiological releases and events, or vary 

exercise scenarios across an eight 

calendar year exercise cycle to allow for 

the demonstration of responses to 

specified scenario elements, 

respectively. These requirements would 

no longer apply due to the limited types 

of events that could occur. The 

previously routine progression to a 

General Emergency, or even a Site Area 

Emergency, in nuclear power reactor 

site scenarios is not applicable for 

licensees with PDEPs. 


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The NRC is issuing DG–1346 for 

public comment in conjunction with 

this proposed rule; DG–1346 provides 

one method acceptable to the NRC for 

licensees with PDEPs to comply with 

the proposed drill and exercise 

requirements. 

PDEP Offsite Response Organization 

Participation in Drills and Exercises 

Appendix E to 10 CFR part 50, 

paragraph IV.F and § 50.47(b)(14) 

include requirements for periodic EP 

drills and exercises for licensees. 

Appendix E to 10 CFR part 50, 

paragraphs IV.F.2.c and IV.F.2.d 

requires offsite radiological emergency 

plans for each site to be exercised 

biennially with full participation by 

offsite authorities having a role under 

the radiological response plan. 

Appendix E to 10 CFR part 50, 

paragraphs IV.F.2.f and IV.F.2.h address 

State and local participation in remedial 

exercises and refusal of State and local 

governments to participate. Because no 

action is required from State and local 

government organizations in response to 

events other than firefighting, law 

enforcement, and ambulance/medical 

services, the requirements related to 

ORO participation in radiological drills 

and exercises would no longer be 

applicable to licensees with PDEPs. 

Proposed § 50.200(c)(1)(vi)(B) would 

remove the requirement to exercise 

offsite emergency plans once the NRC 

has docketed the licensee’s 

certifications required under 

§ 50.82(a)(1) or § 52.110(a) and the 

licensee elects under § 50.54(q)(7)(ii) to 

transition to a PDEP. For facilities that 

are located either on the same site or on 

adjacent contiguous sites to reactors that 

continue to operate, the offsite 

emergency plans would continue to be 

exercised as required under appendix E 

to 10 CFR part 50, paragraph IV.2.f, 

until all reactors at the site cease 

operation and transition to a PDEP. 

Similar to the requirements under 

appendix E to 10 CFR part 50, paragraph 

IV.2.f.e, under proposed 

§ 50.200(c)(1)(vi)(B)(3), a licensee with a 

PDEP would be required to enable any 

State or local government to participate 

in the licensee’s drills and exercises 

when requested. 

6. Independent Spent Fuel Storage 

Installation-Only Emergency Plans 

In order to transition to an IOEP, the 

NRC is proposing under § 50.54(q)(7)(iii) 

that licensees must have all spent fuel 

in dry cask storage. Licensees with an 

IOEP must follow and maintain the 

effectiveness of an emergency plan that 

meets the requirements in § 72.32(a). 

Licensees with 10 CFR part 72 

specific licenses or under the 10 CFR 

part 72 general license may hold an 

IOEP. A licensee with all of its spent 

fuel in dry cask storage that terminates 

its 10 CFR part 50 or 10 CFR part 52 

license must first obtain a 10 CFR part 

72 specific license before transitioning 

to the EP requirements already provided 

in § 72.32(a). A licensee maintaining its 

10 CFR part 50 or 10 CFR part 52 

license, and thus its 10 CFR part 72 

general license authorized under 

§ 72.210, ‘‘General license issued,’’ may 

opt to change its EP program to align it 

with the requirements of § 72.32 once 

all spent fuel is transferred to dry cask 

storage. In addition, licensees under the 

10 CFR part 72 general license would 

need to continue to comply with all 

applicable 10 CFR part 50 and 10 CFR 

part 52 requirements until the 10 CFR 

part 50 or 10 CFR part 52 license is 

terminated consistent with § 50.82 or 

§ 52.110, respectively. 

Under proposed § 50.54(q)(7)(iii), a 

licensee may choose not to comply with 

the EP requirements under § 72.32 and 

may instead maintain a PSEP or PDEP. 

Licensees with dry cask storage must 

ensure that the emergency plan includes 

an appropriate EAL scheme. 

The NRC is issuing DG–1346 for 

public comment in conjunction with 

this proposed rule; DG–1346 provides 

guidance on transitioning to and 

maintaining an IOEP. 

7. All Spent Fuel Removed From Site 

During the fourth level of 

decommissioning, the proposed rule 

would allow a licensee to terminate its 

EP program under proposed 

§ 50.54(q)(7)(iv) or proposed § 72.44(f). 

Once all spent fuel has been 

permanently removed from the site, the 

site no longer poses any risk of a 

radiological release. The licensee must 

then continue to follow its PSDAR 

submitted under § 50.82 until 

decommissioning is completed. 

8. Changes to Emergency Plans 

Existing § 50.54(q)(2) requires nuclear 

power reactor licensees to follow and 

maintain the effectiveness of an 

emergency plan that meets the planning 

standards in § 50.47(b) and the 

requirements in appendix E to 10 CFR 

part 50. In addition, § 50.54(q)(3) 

contains the conditions under which the 

licensee may make changes to its 

emergency plan without prior 

application to and approval by the NRC, 

provided that the changes do not reduce 

the effectiveness of the plan and that the 

plan, as changed, continues to meet the 

standards in § 50.47(b) and the 

requirements in appendix E to 10 CFR 

part 50. The NRC is proposing to add 

several new paragraphs that, similar to 

§ 50.54(q)(2) and (3), would reference 

the requirements that emergency plans 

for decommissioning nuclear power 

reactors must meet and the process for 

making these plan changes. In 

particular, proposed § 50.54(q)(7) would 

reference the applicable emergency plan 

requirements after the NRC dockets a 

licensee’s certifications under 

§ 50.82(a)(1) or § 52.110(a), and 

proposed § 50.54(q)(8) would stipulate 

the conditions under which 

decommissioning nuclear power reactor 

licensees may make changes to their 

emergency plans without prior approval 

by the NRC. The NRC also would revise 

§ 50.54(q)(1) to clarify that the 

definitions in paragraph (q) apply to 

only paragraph (q). 

The existing change process under 

§ 50.54(q) does not establish whether a 

proposed change would impact the 

agency’s determination that there is 

reasonable assurance that a licensee can 

and will take adequate protective 

measures in the event of a radiological 

emergency; the change process 

establishes only whether the licensee 

has the authority to implement the 

proposed change without prior NRC 

approval. The change process uses the 

characteristic ‘‘reduction in 

effectiveness’’ to exclude from the 

requirement to seek prior NRC approval 

those changes that would likely not 

reduce the effectiveness of the licensee’s 

emergency plan. Because these changes 

would not reduce the effectiveness of 

the plan, the NRC expects that the 

changes should not have an impact on 

the agency’s reasonable assurance 

determination. A licensee’s 

determination that a proposed change 

would reduce the effectiveness of the 

emergency plan does not mean that the 

licensee could not or would not 

implement adequate protective 

measures to protect public health and 

safety in the event of a radiological 

accident, but only that prior NRC review 

is required to evaluate the impact of the 

change on the reasonable assurance 

determination. As part of routine 

oversight, the NRC screens emergency 

plan changes, including EAL changes, 

and reviews a sample of changes 

documented in reports submitted under 

§ 50.54(q)(5) that could potentially 

reduce effectiveness. These reviews do 

not constitute the NRC’s approval of the 

plan changes, and all such changes 

remain subject to future inspection and 

enforcement actions. The NRC 

documents its approval of plan changes 

under § 50.54(q)(4) in its decisions to 

grant license amendment requests. 


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The licensee cannot properly evaluate 

a proposed change to the emergency 

plan if it has not considered the basis 

for the NRC’s approval of the original 

plan or the basis for any subsequent 

changes to the plan—whether those 

changes were approved by the NRC or 

implemented by the licensee without 

prior NRC approval under § 50.54(q). 

Regulatory Guide (RG) 1.219, Revision 

1, ‘‘Guidance on Making Changes to 

Emergency Plans for Nuclear Power 

Reactors’’ (ADAMS Accession No. 

ML16061A104), describes a method that 

the NRC considers acceptable to 

implement the requirements in 

§ 50.54(q) as they relate to EP and 

specifically to making changes to 

emergency response plans. As provided 

in RG 1.219, the licensee should 

consider its licensing basis to inform a 

§ 50.54(q) evaluation, and, principally, 

applicable regulatory requirements, 

which are binding on the licensee 

unless the NRC explicitly exempts the 

licensee from them. The NRC is issuing 

DG–1346 for public comment in 

conjunction with this proposed rule to 

provide guidance for decommissioning 

nuclear power reactors in evaluating 

changes to emergency plans under 

proposed § 50.54(q). 

The change process is meant to ensure 

that emergency plans are maintained up 

to date and that the level of planning 

does not fall below the standards to 

which the licensee has committed. The 

regulations in § 50.54(q) define 

‘‘reduction in effectiveness’’ as a change 

in an emergency plan that results in 

reducing the licensee’s capability to 

perform an emergency planning 

function in the event of a radiological 

emergency. ‘‘Emergency planning 

function’’ is currently defined as a 

capability or resource necessary to 

prepare for, and respond to, a 

radiological emergency, as established 

in the planning standards of § 50.47(b) 

and the elements of appendix E to 10 

CFR part 50, section IV. The NRC is 

proposing to remove the references to 

the planning standards of § 50.47(b) and 

appendix E to 10 CFR part 50 from this 

definition because this proposed rule 

would establish alternative emergency 

planning standards under proposed 

§ 50.200, and the NRC does not consider 

the references essential to the definition. 

When the NRC considers exemptions 

from EP requirements for a 

decommissioning nuclear power reactor 

licensee, the NRC considers whether 

there are special circumstances present 

as defined in § 50.12(a)(2). In particular, 

the NRC determines whether 

application of the EP regulations for 

which exemptions are under 

consideration in the particular 

circumstances would not serve their 

underlying purpose or are not necessary 

to achieve their underlying purpose, 

which is to provide reasonable 

assurance that adequate protective 

measures can and will be taken in the 

event of a radiological emergency. Once 

the NRC grants a licensee exemptions 

from EP requirements, the exempted 

regulations no longer apply to the 

licensee. The licensee may need to 

submit a separate license amendment 

request if the planned changes conflict 

with an element of the current licensing 

basis. If not, the licensee need not 

submit a separate license amendment 

request for NRC approval of the 

emergency plan changes unless the plan 

changes go beyond those resulting from 

the exemptions granted. The NRC 

intends that this proposed rule would 

establish clear regulatory requirements 

for EP, reducing the need to request 

certain exemptions. As such, the NRC is 

proposing to add § 50.54(q)(8) to 

establish the process for: (1) Transitions 

from one decommissioning level’s EP 

planning standards and requirements to 

the next level’s EP planning standards 

and requirements, and (2) changes to 

emergency plans within a 

decommissioning level. 

In considering a graded approach to 

EP, the NRC recognizes that a transition 

between the EP planning standards and 

requirements of each decommissioning 

level is not equivalent to making 

changes to the emergency plan within a 

level. The transition between the EP 

planning standards and requirements of 

each decommissioning level is 

fundamentally a licensee’s commitment 

to a different set of EP standards and 

associated emergency planning 

functions, and the change process 

should facilitate this transition. 

For transitions from one 

decommissioning level to the next, the 

NRC would require licensees to 

establish emergency plans that meet the 

EP planning standards and requirements 

of the next level. The transition is 

optional, and a licensee that maintains 

its current level of emergency planning 

would satisfy the requirements of the 

next level; however, doing so would 

mean maintaining emergency planning 

functions above the commensurate level 

of planning for the risk involved. Under 

the proposed § 50.54(q)(8), a licensee 

would be able to make changes to the 

emergency plan to commit to the EP 

planning standards and requirements of 

the next decommissioning level (i.e., 

PSEP, PDEP, or IOEP) using the 

§ 50.54(q)(3) change process, but would 

only need to consider whether the 

changes meet the next level’s planning 

standards and requirements. Licensees 

making changes to their emergency 

plans to commit to the EP planning 

standards and requirements of a 

decommissioning level would not be 

required to determine if the changes are 

reductions in effectiveness. Instead, the 

NRC would have already made this 

determination through its issuance of 

the regulations promulgating the EP 

planning standards and requirements of 

the decommissioning levels. The NRC’s 

proposed regulatory approach to 

transitions between EP 

decommissioning levels does not go 

beyond the authority currently granted 

to licensees to make changes to their 

emergency plan under § 50.54(q)(3). 

Additionally, any change to the 

emergency plan that is not made to 

comply with the EP planning standards 

and requirements of the next 

decommissioning level would require a 

licensee to make a determination as to 

whether the change would be a 

reduction in effectiveness. 

After the Three Mile Island accident 

in 1979, the NRC issued a final rule (45 

FR 55402; August 19, 1980) (1980 EP 

Final Rule) that included § 50.54(u), 

which required licensees to upgrade 

their emergency plans to meet the thennew planning standards of § 50.47(b) 

and requirements in appendix E to 10 

CFR part 50 and to submit those plans 

to the NRC. In the 2011 EP Final Rule, 

the NRC removed and reserved 

§ 50.54(u). The NRC’s proposed 

approach to transitions between EP 

planning standards and requirements of 

decommissioning levels is analogous to 

the approach taken by the NRC when 

the 16 EP planning standards went into 

effect in 1980 (see ‘‘Reasonable 

Assurance and Offsite Radiological 

Emergency Preparedness’’ section in 

this document). Under this approach, 

the NRC would not be relinquishing its 

oversight authority, as some 

commenters on the ANPR and draft 

regulatory basis supposed. As proposed, 

§ 50.54(q)(8)(i) would require initial 

emergency plan changes made to 

transition between EP decommissioning 

levels to be submitted to the NRC at 

least 60 days prior to implementation, 

and emergency plans would remain 

subject to future inspection and 

enforcement. The proposed submittal is 

not intended to be a licensing action. It 

would provide a current copy of the 

emergency plan to the NRC prior to 

implementation in support of future 

inspection activities. This submittal 

would provide an opportunity for the 

NRC to assure that the licensee 

maintains the effectiveness of its 

emergency plan. Subsequent emergency 

plan changes would need to follow the 


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existing change control process under 

§ 50.54(q)(3) and (4). Hearing rights 

would not attach to transitions between 

EP decommissioning levels; however, 

the public has the opportunity to 

comment on the graded EP planning 

standards and requirements themselves 

in response to this proposed rule and 

the drafts of the supporting guidance 

documents. In addition, all emergency 

plan changes submitted under 

§ 50.54(q)(5) and proposed § 50.54(q)(8) 

would be publicly available. 

In addition to the general 

requirements in proposed 

§ 50.54(q)(8)(i) governing transitions 

between EP decommissioning levels, 

proposed § 50.54(q)(8) would address 

changes specific to SSCs and EALs. 

Proposed § 50.54(q)(8)(ii) would specify 

that, for SSCs that are no longer needed 

to provide support for an emergency 

planning function (as defined under 

proposed § 50.54(q)(1)(iii)), a licensee 

may make a determination under 

§ 50.54(q)(3) that emergency plan 

changes are not a reduction in 

effectiveness if the updated FSAR 

demonstrates that these SSCs are no 

longer required to be in service due to 

the decommissioning status of the 

facility. Proposed § 50.54(q)(8)(iii) 

would state that changes to EALs based 

on plant conditions that are not 

physically achievable or 

instrumentation that is no longer in 

service due to the transition to 

decommissioning are not reductions in 

effectiveness provided that a 

§ 50.54(q)(3) evaluation demonstrates 

that the change does not reduce the 

capability of taking timely and 

appropriate protective actions. The NRC 

is proposing these requirements to 

provide clarity on § 50.54(q)(3) 

evaluations and alleviate the burden on 

licensees from submitting emergency 

plan changes that result from SSCs and 

instrumentation that are no longer 

required to be in service due to 

decommissioning. 

After the implementation of a PSEP, 

PDEP, or IOEP, licensees would be 

required by proposed § 50.54(q)(7)(i) to 

continue to follow and maintain the 

effectiveness of the plan and by 

proposed § 50.54(q)(8)(i)–(iii) to comply 

with the change process described 

under existing § 50.54(q)(3) and (q)(4). 

Therefore, licensees would be allowed 

to make changes to these emergency 

plans without prior application to and 

approval by the NRC, provided that the 

changes would not reduce the 

effectiveness of the plan and that the 

plan, as changed, would continue to 

meet the EP planning standards and 

requirements for the applicable 

decommissioning level. Current 

§ 50.54(q)(5) would require 

decommissioning licensees to submit to 

the NRC a report of each such change 

within 30 days after the change is put 

into effect. And, consistent with current 

requirements, decommissioning 

licensees would have to submit changes 

that would reduce the effectiveness of 

the plan for prior NRC review and 

approval in accordance with 

§ 50.54(q)(4) so that the NRC could 

make the requisite reasonable assurance 

determination. For subsequent 

emergency plan changes once all fuel is 

in dry cask storage (i.e., for changes to 

an IOEP), proposed § 50.54(q)(8)(i) 

would allow licensees to follow the 

change process under § 72.44(f). 

The proposed amendments to the 

regulatory change process are necessary 

because: 

• The regulation in existing 

§ 50.54(q)(2), which provides that a 

licensee must follow and maintain the 

effectiveness of the emergency plan, 

should continue to apply in order to 

ensure that emergency plans are 

followed and kept up to date. 

• The existing § 50.54(q) change 

process and the associated regulatory 

guidance currently do not address how 

a licensee could change its emergency 

plans to comply with the emergency 

plan standards as the licensee 

transitions to each level of 

decommissioning. 

• This proposed rule would allow the 

NRC to maintain, through a regulatory 

change process, reasonable assurance 

that a licensee can and will take 

adequate protective measures in the 

event of a radiological emergency. 

The proposed amendments to 

§ 50.54(q), and related regulatory 

guidance, would ensure that licensees 

would maintain the effectiveness of the 

emergency plans. Emergency plans that 

comply with the proposed graded EP 

planning standards and requirements 

would continue to provide reasonable 

assurance that adequate protective 

measures can and will be taken in the 

event of a radiological emergency. Any 

plan that did not meet these standards 

and requirements and, if applicable, the 

reduction in effectiveness criterion, 

would be subject to inspection and 

enforcement actions. The proposed 

approaches to transitioning between EP 

decommissioning levels and to making 

emergency plan changes within 

decommissioning levels would provide 

an efficient and effective regulatory 

change process and would promote 

consistent and predictable 

implementation and enforcement. 

9. Program Element Review Under 

§ 50.54(t) 

Under current § 50.54(t), licensees 

must conduct reviews of EP program 

elements either: (1) At intervals not to 

exceed 12 months or (2) as necessary, 

based on an assessment by the licensee 

against performance indicators and as 

soon as reasonably practicable after a 

change occurs in personnel, procedures, 

equipment, or facilities that potentially 

could adversely affect EP. If a licensee 

chooses the second option, it must still 

review all program elements at least 

once every 24 months. For several 

reasons, the proposed rule would 

provide decommissioning licensees 

with an alternative approach to 

reviewing EP program elements. 

First, the NRC expects licensees to 

remain in the first level of 

decommissioning (i.e., with a PSEP) for 

less than 24 months, and the scope of 

a PSEP is largely unchanged from the 

scope of an operating reactor’s 

emergency plan. Conversely, the second 

level of decommissioning (i.e., licensees 

with a PDEP) will involve more 

significant changes, and the NRC 

anticipates that licensees would remain 

in the second level of decommissioning 

for a longer period of time. Therefore, in 

order to support program continuity and 

minimize changes during the transition 

to a PDEP, the NRC is proposing to 

amend § 50.54(t) such that, starting after 

licensees enter the second level of 

decommissioning, licensees would be 

able to conduct program element 

reviews under § 50.54(t) at intervals not 

to exceed 24 months (rather than 12 

months) without conducting an 

assessment against performance 

indicators. The NRC is proposing to add 

new § 50.54(t)(3) to remove the 

requirement to conduct periodic EP 

program element reviews once all fuel is 

in dry cask storage (i.e., the third/IOEP 

level of decommissioning), consistent 

with the EP requirements for ISFSIs 

under 10 CFR part 72. 

10. Reasonable Assurance and Offsite 

Radiological Emergency Preparedness 

The regulations in §§ 50.47 and 50.54, 

‘‘Conditions of licenses,’’ prescribe how 

the NRC will make licensing decisions 

or take appropriate enforcement actions 

by using findings of reasonable 

assurance that adequate protective 

measures can and will be taken to 

protect public health and safety in the 

event of a radiological emergency. Every 

10 CFR part 50 or 10 CFR part 52 license 

includes as a condition of the license 

the requirements of § 50.54(s)(2)(ii) and 

(s)(3) regarding findings and 

determinations of reasonable assurance. 


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The NRC has the authority and 

responsibility to make licensing 

findings on the overall adequacy of 

onsite and offsite emergency planning 

and preparedness. Commensurate with 

the NRC’s responsibility to make such 

findings, the NRC has the authority to 

collect, review, and evaluate any 

information it needs to support its 

findings on EP. If available, the NRC 

must consider FEMA findings and 

determinations regarding the status of 

offsite EP. The relationship between the 

NRC and FEMA concerning findings of 

reasonable assurance of offsite EP is 

based on the Atomic Energy Act of 

1954, as amended (AEA); the Energy 

Reorganization Act of 1974, as 

amended; the NRC Authorization Act 

for Fiscal Year 1980, the NRC’s 

regulations; a memorandum of 

understanding between the two agencies 

(‘‘Memorandum of Understanding 

Between the Department of Homeland 

Security/Federal Emergency 

Management Agency and Nuclear 

Regulatory Commission Regarding 

Radiological Emergency Response, 

Planning, and Preparedness’’) first 

established in 1980 and last updated in 

2015 (ADAMS Accession No. 

ML15344A371); and case law (e.g., 

Massachusetts v. United States, 856 

F.2d 378, 382 (1st Cir. 1988); State of 

Ohio ex rel. Celebrezze v. NRC, 868 F.2d 

810, 815–16 (6th Cir. 1989)). 

Not all licensing decisions involving 

EP require findings and determinations 

on the adequacy of offsite plans. In the 

EP regulations for research and test 

reactors, fuel cycle facilities, and ISFSIs, 

there are no regulatory standards or 

requirements for offsite radiological 

emergency plans. As such, FEMA 

findings and determinations are not 

needed to support NRC licensing 

decisions for such facilities. The 

absence of NRC regulatory standards for 

offsite radiological EP at those facilities 

does not imply that offsite emergency 

planning, in general, is not adequate to 

protect the public health and safety. In 

addition, the support provided by offsite 

organizations does not automatically 

necessitate the need for findings and 

determinations. In the Low Power Rule 

(47 FR 30232; July 13, 1982), the NRC 

concluded that findings and 

determinations on the state of offsite EP 

were not needed to support issuance of 

a license for fuel loading and low-power 

testing because there was sufficient time 

(at least 10 hours) in which to take 

action to protect the public in even the 

worst-case accident. Additionally, the 

NRC has concluded in its review of 

several EP exemption requests for 

permanently shutdown and defueled 

nuclear power reactor licensees that 

formal offsite radiological emergency 

plans are not necessary after the spent 

fuel in the SFP has sufficiently decayed 

such that it would not reach zirconium 

fuel cladding ignition temperature 

within 10 hours under adiabatic heatup 

conditions. As a result, continued 

consultation with FEMA regarding the 

adequacy of the offsite plans was also 

no longer necessary. 

For decommissioning nuclear power 

reactors, the NRC is proposing that if 

regulatory standards for offsite 

radiological EP are not required, then 

findings and determinations on the 

adequacy of offsite plans would not be 

needed in order for the NRC to make 

determinations regarding reasonable 

assurance under § 50.54(s)(2)(ii). 

Therefore, the NRC is proposing 

changes to § 50.54(s)(3) to clarify that 

FEMA findings and determinations are 

only necessary when the NRC’s 

planning standards apply to offsite 

radiological emergency response plans. 

Additionally, the NRC staff is proposing 

to add a new § 50.47(f) to clarify when 

the 16 planning standards apply to 

offsite radiological emergency plans. A 

licensee must follow and maintain the 

effectiveness of its emergency plan if the 

NRC is to continue to find, under 

§ 50.54(s)(2)(ii), that there is reasonable 

assurance that adequate protective 

measures can and will be taken in the 

event of a radiological emergency, and 

§ 50.54(s)(2)(ii) would continue to apply 

to licensees as a condition of the license 

during decommissioning. 

In 1979, the NRC predicated the 

rationale for the EP proposed rule (44 

FR 75167; December 19, 1979) on the 

Commission’s considered judgment in 

the aftermath of the accident at Three 

Mile Island. At the time, the 

Commission concluded that it must be 

in a position to know that offsite 

governmental plans had been reviewed 

and found adequate. However, the 

Commission also noted that the 

proposed rule was considered an 

interim upgrade of NRC emergency 

planning regulations based on past 

experience, and that further changes to 

emergency planning regulations may be 

proposed as more experience is gained. 

The NRC viewed the 1979 proposed rule 

as a first step in improving emergency 

planning. 

The NRC recognizes the experience 

gained from implementing its 

regulations and also that significant 

advances in emergency planning have 

occurred over the decades following the 

accident at Three Mile Island. In 

particular, the terrorist attacks on 

September 11, 2001, led to the 

establishment of the U.S. Department of 

Homeland Security, and lessons learned 

from disasters such as Hurricane Katrina 

have resulted in a national effort to 

prepare for and respond to all hazards 

and disasters. Homeland Security 

Presidential Directive 5, ‘‘Management 

of Domestic Incidents’’ (February 28, 

2003), and Presidential Policy Directive 

(PPD)–8, ‘‘National Preparedness’’ 

(issued March 30, 2011), established 

national initiatives for a common 

approach to preparedness and response. 

These initiatives include the National 

Incident Management System, National 

Preparedness Goal, Core Capabilities, 

National Preparedness System, National 

Planning Frameworks, and the 

development of comprehensive 

preparedness guides and exercise 

methodologies. 

The PPD–8 directed the development 

of a national preparedness goal that 

identifies the core capabilities necessary 

for preparedness and a national 

preparedness system to guide activities 

that will enable the nation to achieve 

the goal. Core capabilities are intended 

to help coordinate and unify efforts, 

improve training and exercise programs, 

promote innovation, and ensure that the 

administrative, finance, and logistics 

systems are in place to support these 

capabilities. The PPD–8 is aimed at 

facilitating an integrated, all-of-nation, 

capabilities-based approach to 

preparedness, under the assumption 

that national preparedness is the shared 

responsibility of the ‘‘whole 

community,’’ which includes all levels 

of government, the private and 

nonprofit sectors, and individual 

citizens. Acknowledging the national 

preparedness goal, the NRC maintains 

the sole legal authority to establish any 

regulations it deems necessary to ensure 

the adequate protection of public health 

and safety from radiological events. 

For a decommissioning site, the 

licensee, as part of the whole 

community, will maintain radiological 

EP capabilities. Only in the highly 

unlikely event of a zirconium fire—in 

which mitigation actions were not 

successful—would there be a potential 

need to initiate response actions offsite. 

But unlike the EP planning basis for an 

operating reactor, within a few months 

of cessation of operations, there is no 

longer a potential need to provide for 

prompt protective actions in the event 

of an accident. Additionally, protective 

actions such as evacuation are not 

unique to radiological events and occur 

in response to other unique hazards 

such as chemical spills, fires, and 

natural disasters, and are often initiated 

without any pre-planning. In NUREG– 

0396, the NRC states that ‘‘It has been, 

and continues to be the Federal position 


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that it is possible (but exceedingly 

improbable) that accidents could occur 

calling for additional resources beyond 

those that are identified in specific 

emergency plans developed to support 

specific individual nuclear facilities. 

Further, the NRC and Federal position 

has been and continues to be, that as in 

other disaster situations, additional 

resources would be mobilized by State 

and Federal agencies.’’ 

State and local governments are 

responsible for the protection of public 

health and safety (including at 

industrial sites like decommissioning 

reactors), and the NRC has high 

confidence in the ability of OROs to 

implement appropriate response actions 

when necessary. This confidence is 

further strengthened by the NRC’s 

recognition of national-level efforts, in 

which the NRC participates, to improve 

the state of emergency planning at all 

levels of government and within the 

whole community. Consequently, for 

facilities licensed by the NRC where 

radiological hazards are unlikely to have 

an offsite impact, the risk posed by the 

remaining low-level hazard is somewhat 

analogous to that posed by non-nuclear 

hazards (e.g., train derailments or oil 

spills) that are addressed by all-hazards 

planning and not by a separate 

radiological emergency plan. In such 

conditions, there is reasonable 

assurance that appropriate response 

actions can and will be taken in the 

event of a radiological emergency, 

without the need for regulatory 

standards for offsite radiological 

emergency response plans and the 

associated FEMA findings and 

determinations that offsite plans are 

adequate and can be implemented. 

11. Clean-Up of Regulations 

The NRC is proposing to remove 

obsolete dates for certain one-time 

actions that were required as part of the 

2011 EP Final Rule and other obsolete 

dates. These actions are complete, and 

the requirements are no longer binding 

on any current licensee. The dates of 

requirements proposed to be removed 

are: 

(1) Section 50.54(s)(2)(ii), which 

allows the NRC to shut down nuclear 

power reactors that did not provide 

reasonable assurance that adequate 

protective measures would be taken in 

the event of a radiological emergency 

after April 1, 1981. There is no longer 

a need for the date requirement of this 

provision because any future 

determinations made under § 50.54(s) 

will be after April 1, 1981. The NRC is 

proposing to delete ‘‘after April 1, 1981’’ 

and retain the remainder of the 

provision. 

(2) Paragraph 6 of appendix E to 10 

CFR part 50, section I, which was used 

to promulgate specific compliance dates 

for the Tennessee Valley Authority 

Watts Bar Nuclear Plant that was under 

construction at the time of the 2011 EP 

Final Rule. Because the Watts Bar 

Nuclear Plant is now operational and 

subject to all current requirements for 

operating reactors, the NRC is proposing 

to delete this provision. 

(3) Appendix E to 10 CFR part 50, 

paragraph IV.4, which required nuclear 

power licensees to develop an ETE 

analysis using decennial data published 

within 365 days of the later date of the 

most recent decennial data or December 

23, 2011. There is no longer a need for 

the date requirement of this provision 

because the date has expired. The NRC 

is proposing to delete ‘‘of the later of the 

date of’’ and ‘‘or December 23, 2011’’ 

from this provision. 

(4) Appendix E to 10 CFR part 50, 

paragraph IV.A.7, which required 

licensees to identify and describe the 

expected assistance from appropriate 

local, State, and Federal agencies during 

an emergency, including a hostile act, 

by June 23, 2014. The NRC is proposing 

to delete ‘‘by June 23, 2014’’ from this 

provision because the date has expired. 

(5) Appendix E to 10 CFR part 50, 

paragraph IV.A.9, which required 

licensees to conduct a detailed analysis 

by December 24, 2012, demonstrating 

that on-shift personnel are not assigned 

responsibilities that would prevent the 

timely performance of assigned 

functions in the emergency plan. The 

NRC is proposing to delete ‘‘By 

December 24, 2012’’ from this provision 

because the date has expired. 

(6) Appendix E to 10 CFR part 50, 

paragraph IV.B.1, which required 

licensees, by June 20, 2012, to establish 

EALs that include hostile action that 

may adversely affect the nuclear power 

plant. There is no longer a need for the 

date requirement of this provision 

because the date has expired. The NRC 

is proposing to remove ‘‘By June 20, 

2012’’ and retain the remainder of the 

provision. 

(7) Appendix E to 10 CFR part 50, 

paragraph IV.C.2, which required 

licensees, by June 20, 2012, to establish 

and maintain capability to assess, 

classify, and declare an emergency 

condition within 15 minutes after 

indications that an EAL had been 

exceeded. There is no longer a need for 

the date requirement of this provision as 

the date has expired. The NRC is 

proposing to delete ‘‘By June 20, 2012’’ 

and retain the remainder of the 

provision. 

(8) Appendix E to 10 CFR part 50, 

paragraph D.4, which included 

compliance periods for the backup alert 

and notification capability requirements 

under appendix E to 10 CFR part 50, 

paragraph D.3, including a final 

deadline of June 22, 2015. The NRC is 

proposing to remove this paragraph 

because the dates in the paragraph have 

expired, and any future applicants 

required to comply with appendix E to 

10 CFR part 50 would be required to 

comply with the requirements of 

appendix E to 10 CFR part 50, paragraph 

D.3. 

(9) Appendix E to 10 CFR part 50, 

paragraph IV.E.8.c, which required 

licensees’ EOFs to have the capabilities 

required under the section by June 20, 

2012. Because the date requirement of 

this provision has expired, the NRC is 

proposing to delete ‘‘By June 20, 2012’’ 

from this provision. 

(10) Appendix E to 10 CFR part 50, 

paragraph IV.E.8.d, which required 

licensees to identify an alternative 

facility that would be accessible in the 

event of hostile action by December 23, 

2014, with the exception of the 

capability for staging ERO personnel at 

the alternative facility and 

communications capabilities with 

emergency responses facilities, which 

had to be implemented by June 20, 

2012. There is no longer a need for the 

date requirements of this provision as 

the dates have expired. The NRC is 

proposing to delete the deadlines for the 

implementation of this provision. 

(11) Appendix E to 10 CFR part 50, 

paragraph IV.F.2.d, which required 

licensees to fully participate in one 

hostile action by December 31, 2015. 

Because the date requirement of this 

provision has expired, the NRC is 

proposing to delete ‘‘and should fully 

participate in one hostile action exercise 

by December 31, 2015’’ from this 

provision. 

(12) Appendix E to 10 CFR part 50, 

paragraph IV.F.2.j, which required 

licensees to conduct a hostile action 

exercise for each of their sites no later 

than December 31, 2015. Because the 

date requirement of this provision has 

expired, the NRC is proposing to delete 

the requirement from this provision. 

(13) Appendix E to 10 CFR part 50, 

paragraph IV.I, which required 

licensees, by June 20, 2012, to provide 

a range of protective actions to protect 

onsite personnel during hostile action. 

Because the date requirement of this 

provision has expired, the NRC is 

proposing to delete ‘‘By June 20, 2012’’ 

from this provision. 

(14) Appendix E to 10 CFR part 50, 

paragraph VI.4.a, which required 

licensees to develop and submit an 

ERDS implementation plan to the NRC 

by October 28, 1991. There is no longer 


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5A target set is the minimum combination of 

equipment or operator actions which, if all are 

prevented from performing their intended safety 

function or prevented from being accomplished, 

would likely result in radiological sabotage. 

a need for the date requirement of this 

provision because the date has expired. 

The NRC is proposing to delete ‘‘by 

October 28, 1991’’ from this provision. 

(15) Appendix E to 10 CFR part 50, 

paragraph VI.4.d, which required 

licensees to complete the 

implementation of the ERDS by 

February 13, 1993, or before escalation 

to full power, whichever comes later. 

There is no longer a need for the date 

requirement of this provision because 

the date has expired. The NRC is 

proposing to delete ‘‘by February 13, 

1993, or’’ and ‘‘whichever comes later’’ 

from this provision and to continue to 

require licensees to submit an ERDS 

implementation plan to NRC before 

escalation to full power. 

The NRC is proposing to eliminate 

these completed one-time requirements 

in the interest of regulatory clarity. 

Eliminating these requirements would 

not relax any currently effective 

regulatory requirement or cause any 

regulatory burden for current or future 

licensees or applicants. 

12. Revisions to § 72.32 

The NRC proposes to amend 

§ 72.32(a) to address the applicability of 

that provision’s requirement that an 

application for a specific license ISFSI 

must include an emergency plan that 

includes the information in § 72.32(a)(1) 

through (16). The proposed amendment 

would clarify that the requirement 

applies when the proposed ISFSI would 

not be located on the site or within the 

exclusion area of a nuclear power 

reactor licensed under 10 CFR parts 50 

or 52. A nuclear power reactor licensed 

under 10 CFR parts 50 or 52 could be 

under construction, operating, or in 

decommissioning. The proposed 

revisions would consolidate the current 

language and remove redundancies by 

using standardized language consistent 

with other amendments in this 

proposed rule. 

The NRC proposes to amend 

§ 72.32(c) to clarify that the nuclear 

power reactor referenced in that 

provision need not be authorized to 

operate for the ISFSI licensee to use the 

emergency plan requirements in § 50.47 

to meet the requirements of § 72.32. 

Currently, § 72.32(c) applies to ISFSI 

licensees located on the site or within 

the exclusion area of a nuclear power 

reactor that is licensed to operate. 

Because a nuclear power reactor 

licensee is not authorized to operate 

once the NRC dockets the certifications 

required under § 50.82(a)(1) or 

§ 52.110(a), § 72.32(c) could be read not 

to apply to an ISFSI licensee at a 

decommissioning reactor site. However, 

the current language of § 72.32 allows 

an ISFSI licensee with a reactor 

emergency plan to use that emergency 

plan to meet the applicable 

requirements for an ISFSI emergency 

plan. Therefore, the proposed rule 

would clarify that, when the nuclear 

power reactor is under construction, 

operating, or in decommissioning, the 

ISFSI licensee could rely on the 

emergency plan requirements in 

appendix E to part 50 of this chapter 

and 10 CFR 50.47(b), or the 

requirements of 10 CFR 50.200(a) or 10 

CFR 50.200(b), to meet the requirements 

of § 72.32. 

B. Physical Security 

The NRC’s regulations governing 

physical security at a nuclear power 

reactor typically do not distinguish 

between an operating nuclear power 

reactor and a nuclear power reactor that 

is in a decommissioning status. 

However, the security risk profile 

presented by a decommissioning reactor 

decreases significantly from that of an 

operating nuclear power reactor due to 

the reduction in the number of target 

sets 5 and the reduced consequences of 

radiological sabotage. The radiological 

consequences of a security event 

decrease as reactors transition through 

each of the following four levels of 

decommissioning: (1) Permanent 

cessation of operations and permanent 

removal of all fuel from the reactor 

vessel, (2) sufficient decay of fuel in the 

SFP such that it would not reach the 

zirconium fuel cladding ignition 

temperature within 10 hours under 

adiabatic heatup conditions, (3) transfer 

of all fuel to dry storage, and (4) removal 

of all fuel from the site. 

Decommissioning nuclear power reactor 

licensees have sought NRC approval of 

exemptions from, license amendments 

for, and alternative measures to, certain 

physical security regulatory 

requirements because of the reduction 

in the number of target sets and the 

reduced consequences of radiological 

sabotage as the nuclear power reactor 

site transitions through these levels. The 

NRC is proposing options to allow 

nuclear power reactor licensees to make 

certain commonly-requested changes to 

their physical security plans based on 

these decommissioning levels without 

requesting exemptions, alternative 

measures, or license amendments. 

1. Security Plans 

Upon the cessation of operations and 

removal of all fuel from the reactor 

vessel, licensees typically seek to 

modify their security plans to reflect 

changes in site conditions. The NRC’s 

regulations in § 50.54(p) establish 

processes that allow licensees to make 

changes to their security plans. Section 

50.54(p)(1) requires licensees to seek 

NRC review and approval of any 

changes that result in a decrease in 

safeguards effectiveness of their security 

plans. Section 50.54(p)(2) allows 

licensees to make changes to their 

security plans without prior NRC 

approval provided that the changes do 

not decrease the safeguards 

effectiveness of the plan. 

The current regulations do not define 

the term ‘‘decrease in safeguards 

effectiveness’’ nor do they include 

examples of the types of changes that 

would constitute a decrease in 

safeguards effectiveness. Additionally, 

there is no definition of the term 

‘‘change.’’ This lack of clear definitions 

has resulted in difficulties for licensees 

implementing security plan changes. 

For example, some licensees have 

implemented changes under 

§ 50.54(p)(2) that the NRC later 

determined decreased the safeguards 

effectiveness of their security plan. 

Similarly, some licensees have 

unnecessarily requested NRC review 

and approval of changes that did not 

decrease the safeguards effectiveness of 

their security plan. 

The NRC is proposing to revise 

§ 50.54(p) to include definitions of the 

terms ‘‘change’’ and ‘‘decrease in 

safeguards effectiveness.’’ The 

application of these definitions would 

be limited to the revised § 50.54(p) and 

would apply to all 10 CFR part 50 and 

10 CFR part 52 licensees with operating, 

decommissioning, and/or 

decommissioned reactor units. The term 

‘‘change’’ would be defined in a new 

§ 50.54(p)(1)(i) to mean an action that 

results in a modification of, addition to, 

or removal from, the licensee’s security 

plans. The term ‘‘decrease in safeguards 

effectiveness’’ would be defined in a 

new § 50.54(p)(1)(ii) to mean a change 

or series of changes to an element or 

component of the security plans 

referenced in § 50.54(p)(2) that reduces 

or eliminates the licensee’s ability to 

perform or maintain the capabilities 

established in § 73.55(b)(3)(i) without 

compensating changes to other security 

plan elements or components. 

Currently, decommissioning (and 

operating) reactor licensees use the 

§ 50.54(p)(2) process to implement 

changes that they have determined do 

not decrease the safeguards 

effectiveness of their security plans. The 

§ 50.54(p)(2) process requires that 

licensees submit a report of these 


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changes to the NRC. In addition to a 

description of these changes, reactor 

licensees have typically included in 

their report supplemental information 

demonstrating that such changes do not 

constitute a decrease in safeguards 

effectiveness. The submittal of this 

supplemental information in the reports 

has been voluntary. The NRC’s practice 

is to review these reports to confirm that 

the licensee properly concluded that the 

changes would not decrease the 

safeguards effectiveness of their 

Commission-approved security plan. 

The submittal of supplemental 

information in the reports allows the 

NRC to verify in a timely manner that 

the change does not result in a decrease 

in the safeguards effectiveness of the 

plan. Without this supplemental 

information, the NRC could only make 

this determination through the 

inspection process. The NRC is 

proposing to require that reactor 

licensees include with the required 

§ 50.54(p)(2) report a summary of the 

analysis performed to determine that the 

change does not decrease safeguards 

effectiveness of the security plan. The 

summary must be sufficient to 

demonstrate that the change does not 

decrease the safeguards effectiveness of 

the plan. 

2. Dry Cask Storage 

An ISFSI located at a nuclear power 

reactor site is typically licensed under a 

general license issued pursuant to 

subpart K of 10 CFR part 72. Under a 

general license, licensees are required to 

protect the SNF in the ISFSI in 

accordance with the physical security 

requirements in § 73.55, ‘‘Requirements 

for physical protection of licensed 

activities in nuclear power reactors 

against radiological sabotage,’’ with the 

additional conditions and exceptions 

noted in § 72.212, ‘‘Conditions of 

general license issued under § 72.210.’’ 

The NRC also licenses certain ISFSIs 

under a 10 CFR part 72 specific license. 

Consistent with § 72.180, ‘‘Physical 

protection plan,’’ licensees holding a 

specific license are required to protect 

the SNF in the ISFSI in accordance with 

the physical security requirements in 

§ 73.51, ‘‘Requirements for the physical 

protection of stored spent nuclear fuel 

and high-level radioactive waste.’’ 

Although the physical security 

requirements that apply to general 

license ISFSIs and specific license 

ISFSIs provide equivalent levels of 

protection, there are differences. For 

instance, § 73.55 requires licensees to 

ensure they maintain the capability to 

detect, assess, interdict, and neutralize 

threats. Section 73.51 requires licensees 

to detect and assess threats and 

communicate with an appropriate 

response organization. The additional 

requirements in § 73.55 that support 

interdiction and neutralization of 

threats is only one example of 

differences that lead to licensee requests 

for exemptions once all fuel has been 

placed in dry cask storage. 

As stated at the beginning of this 

section, decommissioning reactors 

typically transition through four distinct 

levels during decommissioning. Many 

decommissioning licensees have 

submitted license amendment requests, 

requests for exemptions, and requests 

for approval of alternative measures to 

remove § 73.55 physical security 

requirements that are no longer 

applicable once the licensee enters the 

third decommissioning level when all 

SNF has been moved to a dry cask 

storage system. 

The need for license amendments, 

exemptions, and approvals of 

alternative measures imposes a 

regulatory burden upon both licensees 

and the NRC. Accordingly, the NRC is 

proposing that once all SNF has been 

placed in dry cask storage, licensees 

may elect to follow the proposed 

§ 72.212(b)(9)(vii) and protect a general 

license ISFSI in accordance with the 

physical security requirements in 

§ 73.51. The applicability section of 

§ 73.51 would also be amended to 

reflect this change. A licensee would be 

able to use the process established in 

the revised and renumbered 

§ 50.54(p)(3) to make this change and 

submit its revised physical security plan 

to the NRC. These security plans would 

have to continue to address the 

applicable security-related orders 

associated with an ISFSI that are 

conditions of the license. The NRC is 

also proposing conforming changes to 

§ 72.13, ‘‘Applicability,’’ to reflect the 

requirements that would apply to a 

licensee that elects to follow the 

proposed § 72.212(b)(9)(vii). 

3. Significant Core Damage 

The prevention of significant core 

damage and spent fuel sabotage is a 

general performance objective of the 

reactor licensee physical protection 

program required by § 73.55. During the 

first level of decommissioning, when 

the NRC has docketed a licensee’s 

certifications that the reactor has 

permanently ceased operating and all 

fuel has been removed from the reactor 

vessel and placed in the SFP, there is no 

longer fuel in the core and therefore the 

risk to public health and safety from 

significant core damage has been 

removed. This reduced risk allows 

licensees to eliminate requirements to 

protect against significant core damage 

or train security and operational 

personnel to protect and respond to core 

damage events. 

The NRC is proposing that a licensee 

of a decommissioning nuclear power 

reactor no longer be required to meet the 

requirement in § 73.55(b)(3) to protect 

against significant core damage once the 

NRC has docketed a licensee’s 

certifications that the reactor has 

permanently ceased operating and all 

fuel has been removed from the reactor 

vessel. The requirement in § 73.55(b)(3) 

to protect against spent fuel sabotage 

remains in effect as long as spent fuel 

remains in the spent fuel pool. 

4. Vital Areas 

A vital area (VA) is defined in § 73.2, 

‘‘Definitions,’’ as any area that contains 

vital equipment. Under § 73.2, vital 

equipment means any equipment, 

system, device, or material, the failure, 

destruction, or release of which could 

directly or indirectly endanger public 

health and safety by exposure to 

radiation. The NRC also considers the 

equipment or systems that would be 

required to function to protect public 

health and safety following such a 

failure, destruction, or release to be 

vital. There are specific physical 

security requirements for the protection 

of VAs and vital equipment. The current 

regulation in § 73.55(e)(9)(v) specifies 

that the reactor control room shall be 

considered a VA. 

The role of the reactor control room 

at an operating plant, as described in 

Criterion 19, ‘‘Control room,’’ of 

appendix A, ‘‘General Design Criteria 

for Nuclear Power Plants,’’ to 10 CFR 

part 50, is to provide a protected space 

from which actions can be taken to 

operate the nuclear power plant safely 

without interruption under normal or 

accident conditions. For a permanently 

shutdown and defueled nuclear power 

reactor, the vital equipment associated 

with operating the reactor vessel is no 

longer needed. The remaining vital 

equipment (e.g., associated with SFP 

cooling) may no longer be needed or 

may be relocated to a VA separate from 

the reactor control room. Once a reactor 

has permanently ceased operations, the 

need for a reactor control room is 

eliminated if all of the vital equipment 

is removed and if the area does not 

serve as the VA boundary for other VAs. 

The proposed rule would revise 

§ 73.55(e)(9)(v) to provide that a licensee 

of a decommissioning nuclear power 

reactor would no longer need to 

designate the reactor control room as a 

VA if it does not otherwise meet the 

definition of a VA in § 73.2. 


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5. Communications 

Currently § 73.55(j)(4)(ii) requires 

continuous and redundant 

communications between the reactor 

control room and the central alarm 

station (CAS). Once a nuclear power 

reactor has permanently ceased 

operations, a licensee may no longer 

have a reactor control room or a 

licensed senior operator present in a 

reactor control room. Therefore, it 

would not be feasible for a licensee of 

a decommissioning nuclear power 

reactor to comply with the current 

regulatory requirement. Licensees 

typically request an exemption from this 

requirement and request that the CAS be 

allowed to establish continuous and 

redundant communications with the 

senior on-site licensee representative. 

The NRC is proposing to amend 

§ 73.55(j) to require continuous and 

redundant communications be 

maintained between the CAS and the 

CFH or senior on-shift licensee 

representative once the reactor has 

ceased operations and the licensee no 

longer has licensed senior operators in 

the control room. The intention of this 

change is to allow licensees flexibility 

in maintaining communications with 

one or both of these individuals. 

Communication requirements will 

continue to include all the conditions 

currently required: Continuous 

communication capability with onsite 

and offsite resources; radio or 

microwave transmitted two-way voice 

communication, in addition to 

conventional telephone service, 

between the alarm stations and local 

law enforcement authorities; and 

alternative communication measures in 

place in areas where communication 

could be interrupted or cannot be 

maintained. 

6. Suspension of Security Measures 

Current regulations in § 73.55(p) 

allow for the suspension of security 

measures in an emergency or during 

severe weather. A senior licensed 

operator must approve the suspension 

of security measures. Once a nuclear 

power reactor has entered 

decommissioning status and all fuel has 

been removed from the reactor, there 

may no longer be a licensed senior 

operator on site. Therefore, it may not 

be feasible for a licensee of a 

decommissioning nuclear power reactor 

to implement this requirement in the 

event of an emergency or severe 

weather. 

The NRC is proposing to amend the 

requirements in § 73.55(p) to allow a 

CFH to suspend security measures in 

the event of an emergency or severe 

weather once the reactor has shutdown 

and all fuel has been removed from the 

reactor core. 

These proposed changes to § 73.55(p) 

would be consistent with the existing 

regulations in § 50.54(x) and (y) that 

govern approvals for reasonable actions 

that a licensee may take to depart from 

a license condition or a technical 

specification in an emergency. In 

accordance with the provisions of 

§ 50.54(y), licensee actions permitted by 

§ 50.54(x) must be approved (at a 

minimum) by a licensed senior operator 

or, at a decommissioning nuclear power 

reactor after submittal of the 

certifications required under 

§ 50.82(a)(1) or § 52.110(a), by either a 

licensed senior operator or a CFH, 

before taking the action. 

C. Cyber Security 

The NRC is proposing to update cyber 

security requirements in § 73.54, 

‘‘Protection of digital computer and 

communication systems and networks’’ 

for nuclear power reactor licensees. This 

update would clarify the cyber security 

requirements applicable to a nuclear 

power reactor during each stage of the 

decommissioning process. 

As stated in § 73.54, applicants and 

licensees must provide high assurance 

that their digital computer and 

communication systems and networks 

associated with safety and important-tosafety, security, and emergency 

preparedness (SSEP) functions are 

adequately protected against cyber 

attacks, up to and including the design 

basis threat described in § 73.1, 

‘‘Purpose and scope.’’ To accomplish 

this, each holder of a nuclear power 

reactor operating license under 10 CFR 

part 50 has submitted a cyber security 

plan (CSP) to the NRC that has been 

approved by the NRC. Further, each 

combined license (COL) applicant is 

required to submit its CSP as part of its 

COL application for review and 

approval. Each approved CSP is 

referenced in a license condition in each 

10 CFR part 50 license, and this license 

condition requires a licensee to 

maintain its CSP until the license is 

terminated or the license condition is 

removed by license amendment. A COL 

holder does not have an equivalent 

cyber security license condition. 

The cyber security requirements in 

§ 73.54 apply to licensees currently 

licensed to operate a nuclear power 

plant. Once the NRC has docketed a 

licensee’s § 50.82(a)(1) or § 52.110(a) 

certifications, that licensee is no longer 

authorized to operate a nuclear power 

plant. Therefore, the requirements in 

§ 73.54 would no longer apply to such 

a licensee. However, each 10 CFR part 

50 licensee has a license condition 

requiring the licensee to maintain its 

CSP, and this license condition remains 

in effect during decommissioning. A 

COL holder, without the license 

condition, is not required to maintain its 

CSP when it begins decommissioning. 

Although a licensee that has 

submitted its § 50.82(a)(1) or § 52.110(a) 

certifications is no longer operating, 

such a licensee may still have fuel 

recently removed from the reactor vessel 

in its SFP. As discussed in the 

‘‘Technical Basis for Graded Approach’’ 

section of this document, if the spent 

fuel in the SFP has not sufficiently 

decayed, there is a risk that the spent 

fuel could heat up to clad ignition 

temperature and lead to a zirconium fire 

for postulated draindown scenarios in a 

timeframe that is too short to reliably 

implement mitigation measures or to 

take other appropriate response actions. 

As discussed in the ‘‘Technical Basis 

for Graded Approach’’ section of this 

document, in Level 2 there is little 

chance that the spent fuel in the SFP 

could heat up to clad ignition 

temperature within 10 hours. 

Accordingly, the NRC is proposing that 

the cyber security requirements in 

§ 73.54 continue to apply to licensees 

through Level 1. This continuation of 

the cyber security requirements would 

ensure that a compromise of digital 

systems cannot adversely impact the 

effective operation of the licensees’ 

physical security programs and 

emergency preparedness functions prior 

to the time at which the spent fuel 

cannot reasonably heat up to clad 

ignition temperature within 10 hours 

after a draindown event. Although the 

cyber security requirements would 

continue to apply through Level 1, the 

number of critical digital assets would 

decrease as systems are removed from 

service, which in turn reduces the 

number of critical digital assets that 

must be protected by the CSP. 

To clarify the applicability of the 

cyber security rule to decommissioning 

nuclear power reactor licensees, the 

NRC is proposing to add two paragraphs 

to § 73.54. A new § 73.54(i) would state 

that the requirements of § 73.54 will 

remain in effect until: (1) The NRC has 

docketed the licensee’s § 50.82(a)(1) or 

§ 52.110(a) certifications, and (2) at least 

10 months for a BWR or 16 months for 

a PWR have elapsed since the date of 

permanent cessation of operations or an 

NRC-approved alternative to the 10 or 

16 month spent fuel decay period, 

submitted under proposed 

§ 50.54(q)(7)(ii)(A) or (B), has elapsed. A 

new § 73.54(j) would state that, after 

both requirements of § 73.54(i) have 

been met, the licensee’s license 


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12286 Federal Register / Vol. 87, No. 42 / Thursday, March 3, 2022 / Proposed Rules 

condition that requires implementation 

and maintenance of a cyber security 

plan would be removed from the 

license. The NRC is also proposing the 

removal of the introductory paragraph 

of § 73.54 in its entirety and revising the 

language of § 73.54(a), (b), and (c). These 

are conforming changes to clarify that 

the applicability of § 73.54 is not limited 

to ‘‘operating’’ reactors (i.e., that § 73.54 

would still be applicable after the NRC 

has docketed a licensee’s § 50.82(a)(1) or 

§ 52.110(a) certifications), to remove 

language that is no longer needed 

concerning the initial submission of 

cyber security plans by existing 

licensees, and to add clarifying language 

to § 73.54(b) and (c). Further, the NRC 

is proposing a change to § 73.55(c)(6), 

which requires the licensee to establish, 

maintain, and implement a cyber 

security plan. This is a conforming 

change to reflect the scenario in which 

a decommissioning nuclear power 

reactor licensee is no longer required to 

maintain a cyber security plan (i.e., the 

NRC has docketed the certifications of 

permanent cessation of operations and 

permanent removal of fuel from the 

reactor vessel, and the fuel in the SFP 

has sufficiently decayed), but is still 

required to comply with § 73.55(c). 

The proposed revision to § 73.54(a) 

would not constitute backfitting for 10 

CFR part 50 licensees. The proposed 

revision would constitute a change 

affecting the issue finality of COL 

holders; extending the requirement to 

maintain a CSP during 

decommissioning would be a new 

requirement imposed on COL holders. 

The NRC’s proposed backfit analysis is 

located in the ‘‘Backfitting and Issue 

Finality’’ section of this document. 

D. Drug and Alcohol Testing 

1. Scope of 10 CFR Part 26 

The NRC is proposing to amend 

§ 26.3, ‘‘Scope,’’ to correct an 

inconsistency within § 26.3(a) where the 

FFD requirements in 10 CFR part 26 

apply differently to 10 CFR part 50 and 

10 CFR part 52 licensees with 

decommissioning nuclear power 

reactors. The § 26.3(a) provision lists 

those licensees that are required to 

comply with designated subparts of 10 

CFR part 26, including licensees who 

are authorized to operate a nuclear 

power reactor under § 50.57 and holders 

of a combined license under 10 CFR 

part 52 after the Commission has made 

the finding under § 52.103(g). In 

accordance with this requirement, 10 

CFR part 26 does not apply to a holder 

of a nuclear power reactor license 

issued under 10 CFR part 50 that is no 

longer authorized to operate a nuclear 

power reactor because the NRC has 

docketed the certifications required 

under § 50.82(a)(1) (i.e., a 

decommissioning 10 CFR part 50 

nuclear power reactor licensee). 

However, 10 CFR part 26 continues to 

apply to holders of combined licenses 

issued under 10 CFR part 52 throughout 

decommissioning. Therefore, there is an 

inconsistency in the application of FFD 

requirements to nuclear power reactor 

licensees during decommissioning. 

The NRC has determined that there is 

no technical basis for this inconsistency. 

In the 1989 10 CFR part 26 final rule (54 

FR 24468; June 7, 1989) (1989 FFD Final 

Rule), the Commission explained that 

the intent of that rule was to address the 

potential for worker impairment of any 

kind, including substance abuse that 

could affect the safe operation of 

nuclear power plants. The emphasis 

throughout the 1989 FFD Final Rule is 

that the rule is necessary to promote 

public health and safety when the plant 

is operational. The wording for 10 CFR 

part 52 licensees described in the scope 

of the 2008 10 CFR part 26 final rule (73 

FR 16966; March 31, 2008) (2008 FFD 

Final Rule), specifically § 26.3(a), was 

an oversight. The emphasis of the 1989 

FFD final rule that FFD need only apply 

to operating 10 CFR part 50 sites should 

be the same for 10 CFR part 52 

licensees. Due to the decreased risk to 

public health and safety during 

decommissioning, 10 CFR part 26 

should not apply to these licensees 

during decommissioning. 

Therefore, the NRC proposes to clarify 

that 10 CFR part 26 does not apply to 

10 CFR part 52 licensees once the NRC 

has docketed their § 52.110(a) 

certifications. Section 26.3(a) of the 

proposed rule would specify that each 

holder of an operating license for a 

nuclear power reactor under 10 CFR 

part 50 and each holder of a COL under 

10 CFR part 52 for which the 

Commission has made the finding under 

§ 52.103(g) must comply with the 

requirements of 10 CFR part 26, except 

for subpart K of 10 CFR part 26, until 

the NRC’s docketing of the license 

holder’s certifications described in 

§§ 50.82(a)(1) or 52.110(a). 

For clarity, the NRC proposes to 

divide the current paragraph of § 26.3(a) 

into two paragraphs. Paragraph (a)(1) 

would retain the requirement in the 

second sentence of current § 26.3(a) to 

state the deadline by which licensees 

must implement their FFD program. 

Paragraph (a)(2) would retain the 

requirement in the first sentence of 

current § 26.3(a) that these licensees 

must comply with the requirements of 

10 CFR part 26, except subpart K, but 

clarify that this requirement ends when 

the NRC dockets the licensee’s 

§§ 50.82(a)(1) or 52.110(a) certifications. 

2. Fitness-for-Duty Elements for Insider 

Mitigation Program 

Under § 73.55(b)(9), a licensee is 

required to establish, maintain, and 

implement an IMP to monitor the initial 

and continuing trustworthiness and 

reliability of individuals granted 

unescorted access authorization (UAA) 

or unescorted access (UA) to a protected 

area (PA) or vital area (VA). 

Section 73.55(b)(9)(ii)(B) requires that 

an IMP must contain elements of an 

FFD program described in 10 CFR part 

26. However, the regulations do not 

identify which FFD program elements 

must be included in the IMP. Section 

73.55(b)(9)(ii)(B)(1) and (2) of this 

proposed rule would amend 

§ 73.55(b)(9)(ii)(B) to establish an 

appropriate set of FFD provisions to be 

incorporated into the IMP of operating 

and decommissioning 10 CFR part 50 

and 10 CFR part 52 licensees to provide 

reasonable assurance that individuals 

granted UAA or UA to the PA or VA are 

trustworthy and reliable. 

Section 73.55(b)(9)(ii)(B)(1) of this 

proposed rule would clarify 

§ 73.55(b)(9)(ii)(B) that licensees 

implementing 10 CFR part 26, 

regardless of whether they are required 

to do so, are in compliance with 

§ 73.55(b)(9)(ii)(B). A licensee’s full 10 

CFR part 26 FFD program (i.e., an FFD 

program that complies with all 

applicable 10 CFR part 26 requirements) 

would contain FFD elements 

appropriate for inclusion in the 

licensee’s IMP. This would apply to 

both operating and decommissioning 

licensees. 

Section 73.55(b)(9)(ii)(B)(2)(i) and (ii) 

of this proposed rule describes the 

minimum 10 CFR part 26 elements 

necessary for a 10 CFR part 50 and 10 

CFR part 52 decommissioning licensee’s 

IMP. Section 73.55(b)(9)(ii)(B)(2)(i) of 

the proposed rule states that individuals 

who have unescorted access to the VAs 

at a decommissioning site, perform 

certified fuel handler functions (i.e., 

individuals covered by § 50.2) prior to 

all spent nuclear fuel at a site being 

placed in dry cask storage, perform 

security—related functions (i.e., 

individuals covered by § 26.4(a)(5)), or 

administer the drug testing program 

(i.e., individuals covered by § 26.4(g)) 

are subject to the requirements in 10 

CFR part 26 except for subparts I and K. 

Individuals who have fuel handler 

certifications are essential to the safe 

movement of spent nuclear fuel. 

Individuals who have security-related 

responsibilities or perform work around 

the spent fuel pool may have knowledge 


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of value to an adversary. In addition, 

security personnel generally carry 

weapons on site and would pose a 

significant challenge to site security if 

they were to perform as an active 

insider during an attack. Testing of 

individuals who administer a drug 

testing program is viewed as essential to 

the integrity of the program. 

Proposed § 73.55(b)(9)(ii)(B)(2)(ii) 

states that individuals who have UA to 

the protected area, but do not perform 

certified fuel handling or securityrelated functions or administer the drug 

testing program would still be subject to 

pre-access and for-cause testing 

(§ 26.31(c)(1) and (2)) and behavior 

observation (§ 26.33), but would not be 

subject to random testing (§ 26.31(c)(5)). 

The NRC proposes to relax these 

requirements because while the reactor 

is in decommissioning the potential 

contribution of certain personnel to 

support an adversary as an insider is 

greatly reduced. Individuals who do not 

have any security-related 

responsibilities or regular SFP area UA 

will have less potential contribution as 

an insider threat. 

The NRC has determined that the FFD 

elements necessary for an IMP under 

this proposed rule are commensurate 

with the hazard and potential event 

consequences associated with a 

facility’s operational status. Section 

73.55(b)(3) states that the physical 

protection program must be designed to 

prevent significant core damage and 

spent fuel sabotage. Operating nuclear 

power reactor facilities contain many 

target sets located throughout the PA of 

potential interest to an adversary 

seeking to affect core damage or spent 

fuel sabotage, thus anyone who has 

UAA or UA to the PA could contribute 

significantly to an adversary. 

The hazard and potential event 

consequences associated with 

decommissioning facilities significantly 

decrease in comparison to those 

associated with the operating facilities. 

During decommissioning, the SFP 

becomes the primary focus of the 

licensee’s obligation to protect against 

the radiological sabotage design basis 

threat, as it becomes the location where 

all spent fuel is located when a nuclear 

power reactor is no longer operating and 

prior to transitioning to an ISFSI. With 

this perspective, this proposed rule 

tailors applicability of the FFD elements 

commensurate with the duties and 

access of personnel who have been 

granted UAA and maintain UA to the 

PA or VA. 

3. Criminal Penalties 

The NRC proposes to amend the 

criminal penalties section of 10 CFR 

part 26 by including § 26.3 within 

§ 26.825(a). Existing § 26.825(a) applies 

the NRC’s authority under the AEA to 

impose criminal penalties for willful 

violations of, attempts to violate, or 

conspiracies to violate NRC regulations. 

Section 26.825(b) lists § 26.3 as one of 

the 10 CFR part 26 provisions that is 

excluded from § 26.825(a). In general, 

the criminal penalties sections of NRC 

regulations apply to substantive 

requirements, and administrative or 

procedural regulatory provisions are 

excluded from criminal penalties 

sections. The current § 26.3 is entitled 

‘‘Scope’’ and identifies which entities 

are within the scope of 10 CFR part 26. 

Scoping provisions typically do not 

contain substantive requirements, 

which may explain why § 26.825(b) 

includes § 26.3. However, the current 

§ 26.3(a) not only describes the entities 

that are subject to the requirements of 

10 CFR part 26 but also includes a 

substantive requirement for certain 

entities to comply with requirements in 

10 CFR part 26 by a specific deadline. 

This requirement was added to § 26.3(a) 

in the 2008 FFD Final Rule, but 

§ 26.825(b) was not updated to reflect 

this change, which was an oversight. 

This proposed rule would not change 

the substantive requirement in § 26.3(a). 

Because proposed § 26.3(a) would 

continue to impose a substantive 

requirement, the NRC proposes to 

remove § 26.3 from § 26.825(b), thereby 

including § 26.3 in § 26.825(a). 

E. Certified Fuel Handler Definition and 

Elimination of the Shift Technical 

Advisor 

The NRC is proposing two revisions 

to its regulations. The first change 

would be to amend the definition of a 

CFH in § 50.2 to provide an alternative 

that would eliminate the need for 

licensees to seek NRC approval for fuel 

handler training programs by adding a 

provision that requires the training 

program to address the safe conduct of 

decommissioning activities, safe 

handling and storage of spent fuel, and 

appropriate response to plant 

emergencies, and specifies that a CFH 

must be qualified in accordance with a 

fuel handler training program that meets 

the same requirements as training 

programs for non-licensed operators 

required by § 50.120. This proposal 

would provide consistency in the 

regulatory treatment of the training 

programs for non-licensed operators 

(which do not require NRC approval) 

and fuel handler training programs to 

qualify a non-licensed operator as a CFH 

(which do require NRC approval). The 

second change would clarify that an 

STA is not required for 

decommissioning reactors. These 

changes would provide clarity to the 

CFH’s responsibilities and functions 

and the role of an STA by codifying 

current licensing practices. This 

proposed rule would also clarify the 

management role of the CFH in a 

manner that is consistent with § 50.54(y) 

as discussed in section ‘‘B. Physical 

Security’’ in this document. 

1. Alternative Definition for Certified 

Fuel Handler 

The current definition of a CFH in 

§ 50.2 does not specify what is in an 

NRC-approved fuel training program. 

Licensees have submitted requests for 

the approval of CFH training and 

retraining programs in connection with 

their decommissioning. After receiving 

NRC approval of a CFH training 

program, the licensee typically submits 

a license amendment request to propose 

changes to the Administrative Controls 

section of its Technical Specifications 

(TS) to include a CFH, among other 

applicable changes based on the 

approval of the CFH training program. 

For example, on May 12, 2014, the 

NRC approved the Shift Manager/ 

Certified Fuel Handler training program 

for Kewaunee Power Station (ADAMS 

Accessions No. ML14104A046). The 

NRC’s safety evaluation supporting 

approval of the CFH training program 

used criteria that focused on whether 

the licensee trained CFHs on the 

following three objectives: (1) Safe 

conduct of decommissioning activities; 

(2) safe handling and storage of spent 

fuel; and (3) appropriate response to 

plant emergencies. These three 

objectives have subsequently been the 

basis for other NRC approvals of CFH 

training programs for licensees entering 

or planning to enter the 

decommissioning process: Entergy for 

VY (ADAMS Accession No. 

ML14162A209); Exelon for Oyster Creek 

Nuclear Generation Station, Clinton 

Power Station, and Quad Cities Nuclear 

Power Station (ADAMS Accession No. 

ML16222A787); and Entergy for 

FitzPatrick Nuclear Power Plant 

(ADAMS Accession No. ML16259A347). 

In the safety evaluations for those 

approved CFH training programs, the 

NRC discusses the 1996 Final Rule and 

its role in the development of the 

objectives for an acceptable CFH 

training program. The NRC recognized 

that the risks posed at decommissioning 

reactors are significantly less than those 

posed by operating reactors. The NRC 

noted specifically that: 

• While the spent fuel is still highly 

radioactive and generates heat caused 

by radioactive decay, no neutron flux is 


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generated and the fuel slowly cools as 

its energetic decay products diminish. 

• The systems required for 

maintaining the spent fuel in the spent 

fuel pool as well as the operations 

required to contain the remaining 

residual contamination in the facility 

and spent fuel pool are relatively 

simple. 

• Because the spent fuel is stored in 

a configuration that precludes a nuclear 

fission reaction, no generation of new 

radioactivity can occur and the potential 

for consequences that could result from 

an inadvertent nuclear reaction are 

highly unlikely. 

Because of the reduced risks and 

relative simplicity of the systems 

needed for safe storage of the spent fuel, 

the NRC explained in the 1996 Final 

Rule that the degree of regulatory 

oversight required for a nuclear power 

reactor during its decommissioning 

stage is considerably less than that 

required for the facility during its 

operating stage. In the 1995 

decommissioning proposed rule (60 FR 

37374; July 20, 1995), the NRC provided 

insights as to the responsibilities of the 

proposed new position of the CFH. 

Specifically, the NRC stated that a CFH 

is an individual who has the requisite 

knowledge and experience to evaluate 

plant conditions and make judgments 

about emergency action decisions 

necessary to protect the public health 

and safety. 

In addition to using the three 

objectives to evaluate the fuel handler 

training programs for licensees entering 

or planning to enter decommissioning, 

the NRC applied the criteria in § 50.120, 

‘‘Training and qualification of nuclear 

power plant personnel,’’ and assessed 

the proposed fuel handler training 

programs against the elements of a 

systems approach to training (SAT) as 

defined in § 55.4, ‘‘Definitions.’’ Section 

50.120 identifies individuals required to 

be subject to an SAT, including nonlicensed operators such as CFHs, and 

necessary elements for training 

programs. These elements include the 

requirement to periodically evaluate 

and revise the training program, as 

appropriate, to reflect changes to the 

facility (e.g., decommissioning), 

procedures, regulations, and quality 

assurance requirements. 

Because it has developed succinct 

criteria to approve fuel handler training 

programs, the NRC proposes to include 

this criterion in its regulations as an 

alternative definition of a CFH to 

eliminate the need for licensees to 

submit requests for NRC approval of 

CFH training programs. Specifically, the 

NRC would codify current approval 

practices by amending § 50.2 to add the 

three broad-scope objectives as 

responsibilities for which a CFH must 

be trained: (1) Safe conduct of 

decommissioning activities; (2) safe 

handling and storage of spent fuel; and 

(3) appropriate response to plant 

emergencies. In addition, the CFH 

would have to qualify in accordance 

with a fuel handler training program 

that meets the same requirements as 

training programs for non-licensed 

operators required by § 50.120. Should a 

licensee not exercise the alternative 

definition, it would need to submit a 

request for approval of a fuel handler 

training program. 

2. Elimination of the Shift Technical 

Advisor 

The STA is a position identified in 

licensees’ TSs. The STA provides 

engineering expertise in the diagnosis of 

complex problems with SSCs during 

reactor operation. Once a licensee enters 

the decommissioning process, the STA 

function is no longer needed. The 

current regulations do not address the 

acceptability of discontinuing the STA 

position for a decommissioning reactor. 

Licensees have been removing the STA 

position and replacing that position 

with a CFH in their TSs through license 

amendments (see Duke Energy Florida 

for Crystal River Unit 3 Nuclear 

Generating Plant (ADAMS Accession 

No. ML14097A145); Exelon for Oyster 

Creek Nuclear Generating Station 

(ADAMS Accession No. ML16235A413); 

and Entergy for VY (ADAMS Accession 

No. ML14217A072)). The NRC proposes 

to revise a footnote to the table titled 

‘‘Minimum Requirements Per Shift for 

On-Site Staffing of Nuclear Power Units 

by Operators and Senior Operators 

Licensed Under 10 CFR part 55’’ in 

§ 50.54(m)(2)(i) to state that an STA is 

not required upon the NRC’s docketing 

of the license holder’s certifications 

required under §§ 50.82(a)(1) or 

52.110(a). 

F. Decommissioning Funding Assurance 

The NRC proposes to amend its 

regulations to modify decommissioning 

funding reporting requirements, clarify 

decommissioning funding assurance 

requirements, and eliminate duplicative 

regulations. 

1. Clarification of § 50.82(a) and 

§ 52.110(h) 

The NRC is proposing to amend the 

regulations in § 50.82(a)(8)(i)(A) and 

§ 52.110(h)(1)(i) to remove the term 

‘‘legitimate.’’ This term does not add 

any substance to the regulations and is 

potentially confusing. The intent of the 

regulation is to ensure that expenses fall 

within the NRC definition of 

decommission. Whether an expense 

falls within the definition of 

decommission would continue to be 

determined on a case-by-case basis by 

the licensee when considering whether 

to make a withdrawal from the 

decommissioning trust fund. Since this 

term is non-substantive, its removal 

would not change any of the existing 

requirements regarding the use of 

decommissioning funds. 

2. Changes to Reporting Requirements 

In the ‘‘Financial Assurance 

Requirements for Decommissioning 

Nuclear Power Reactors’’ final rule (63 

FR 50465; September 22, 1998), the NRC 

added the provisions currently in 

§ 50.75(f)(1) and (2) that require each 

nuclear power reactor licensee to file a 

report with the NRC on the status of its 

decommissioning funding for each 

reactor that it owns, by March 31st of 

every odd-numbered year or annually 

for plants that are within five years of 

their projected end of operation. This 

report must specify: (1) The amount of 

decommissioning funds estimated to be 

required pursuant to § 50.75(b) and (c); 

(2) the amount of decommissioning 

funds accumulated to the end of the 

calendar year preceding the date of the 

report; (3) a schedule of the annual 

amounts remaining to be collected; (4) 

the assumptions used regarding rates of 

escalation in decommissioning costs, 

rates of earnings on decommissioning 

funds, and rates of other factors used in 

funding projections; (5) any contracts 

upon which the licensee is relying; (6) 

any modifications occurring to a 

licensee’s current method of providing 

financial assurance since the last 

submitted report; and (7) any material 

changes to trust agreements. 

The NRC is proposing to change the 

reporting frequency in § 50.75(f)(1) to 

coordinate the reporting frequency with 

the ISFSI decommissioning reporting 

frequency in § 72.30. This change would 

convert the biennial decommissioning 

funding status report required for 10 

CFR part 50 and 10 CFR part 52 nuclear 

power reactor licensees to a triennial 

decommissioning funding status report 

as currently is required for 10 CFR part 

72 ISFSI licensees. This revision would 

not change the annual reporting 

frequency for a reactor licensee that is 

within 5 years of its projected end of 

operations, whether that projection is 

based on the license’s expiration date or 

on a premature shutdown, and would 

not change the annual reporting 

frequency for a reactor that has 

permanently ceased operations. Also, 

the change in reporting frequency 

would not relieve the licensee from 

calculating annual adjustments as 


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required under § 50.75(a)(2) and would 

not affect the Table of Minimum 

Amounts in § 50.75(c) or its escalation 

factors. Therefore, a licensee would be 

required to continue to monitor its 

decommissioning funding on an annual 

basis but instead of reporting at least 

once every 2 years to the NRC, it would 

report at least once every 3 years. 

Since 1999, the NRC’s regulations 

have mandated that licensees report to 

the NRC the status of their 

decommissioning funding. Under 

§ 50.75(f)(1), the biennial 

decommissioning funding status report 

requires the disclosure of seven items, 

including the balance of the 

decommissioning trust fund as of 

December 31st of the prior year. The 

NRC conducted spot checks of licensee 

records related to this information. The 

NRC did not identify any major 

discrepancies related to this 

information, as explained in SECY–15– 

0005 (ADAMS Accession No. 

ML14210A554), dated January 15, 2015. 

Therefore, the NRC has confidence that 

changing from a biennial to a triennial 

reporting frequency will not subject the 

public to any additional risks associated 

with decommissioning funding 

assurance. In addition, even with a 

triennial reporting frequency, there 

would be ample time to resolve any 

decommissioning funding issue. 

Furthermore, the proposed revision 

does not change the requirement for 

more frequent reporting as a licensee 

approaches the permanent cessation of 

operations and while the licensee is in 

decommissioning or the requirement for 

a site-specific decommissioning cost 

estimate during this period. 

The NRC proposes a rule change in 

§ 50.75(h) in order to be consistent with 

the requirements of § 50.4. Specifically, 

notifications would be sent directly to 

the Document Control Desk, and not to 

the Director, Office of Nuclear Reactor 

Regulation, or Director, Office of 

Nuclear Material Safety and Safeguards, 

as applicable. This change would 

provide one consistent location for 

licensees to docket all notifications to 

the NRC. 

The NRC proposes to delete 

§ 50.75(f)(2). The language of existing 

§ 50.75(f)(1) fully encompasses the 

language of paragraph (f)(2), and, 

therefore, paragraph (f)(2) is 

unnecessary and potentially confusing. 

By removing paragraph (f)(2) the NRC 

would not be removing the requirement 

on licensees to continue submitting 

decommissioning funding assurance 

status reports. Existing paragraphs (f)(3) 

through (5) would be redesignated as 

paragraphs (f)(2) through (4). 

3. Shortfalls in Decommissioning 

Funding Assurance 

The requirement in § 50.75 that the 

licensee provide reasonable assurance 

that sufficient funds will be available for 

radiological decommissioning is a 

continuing obligation. However, 

economic factors can cause the amount 

of a licensee’s financial assurance to fall 

below the amount required (either by 

the NRC minimum formula in 

§ 50.75(c), or by a licensee’s site-specific 

decommissioning cost estimate), thereby 

creating a shortfall. The regulations do 

not explicitly discuss what to do when 

a licensee faces a funding shortfall, 

regardless of its cause. Instead, the NRC 

addressed the scenario in its guidance 

in RG 1.159, ‘‘Assuring the Availability 

of Funds for Decommissioning Nuclear 

Reactors’’ (ADAMS Accession No. 

ML003740066). This guidance provides 

that non-rate-regulated licensees should 

make up shortfalls in decommissioning 

funding within 2 years and electric 

utility licensees within 5 years. 

The NRC is proposing to amend its 

regulations in § 50.75(f)(1) to clarify 

that, although the regulations establish 

a continuing obligation to provide 

reasonable assurance of 

decommissioning funding, when a 

licensee identifies a shortfall in the 

report required by § 50.75(f)(1), the 

licensee must identify additional 

financial assurance to cover the shortfall 

in the next report. Specifically, the 

proposed rule would require licensees 

to remedy shortfalls before permanent 

cessation of operations consistent with 

the methods identified in § 50.75(e) in 

the next § 50.75(f) report. The proposed 

rule would clarify the expectations for 

how reasonable assurance of funds will 

be available for the decommissioning 

process. For electric utilities that 

currently submit biennial reports but 

correct their shortfalls within 5 years, 

the NRC proposes that they would 

submit their decommissioning funding 

status reports triennially and explain in 

their reports how they plan to correct 

any existing shortfall. Electric utilities 

should continue to correct shortfalls 

within 5 years as explained in RG 1.159. 

For non-rate-regulated licensees that 

currently submit biennial reports and 

should correct shortfalls within a 2 year 

period, the NRC proposes that they 

correct any shortfalls within the 3 year 

reporting period. The NRC proposes to 

clarify the last sentence of current 

§ 50.75(f)(1) to reduce the number of 

clauses and enhance readability. 

The NRC proposes to revise 

§ 50.82(a)(9)(ii)(F) to require licensees to 

identify the specific sources of funds for 

‘‘remaining decommissioning costs,’’ 

including sources of funds for license 

termination, spent fuel management, 

and ISFSI decommissioning. 

4. Conforming Changes to 10 CFR Part 

52 

The NRC proposes to revise § 52.110 

to make the same changes proposed in 

§ 50.82 for the reasons previously 

discussed and for consistency. In 

addition, the NRC proposes to add 

paragraphs (h)(5) through (h)(7) with 

site-specific decommissioning cost 

estimate reporting requirements that are 

identical to the requirements in 

§ 50.82(a)(8)(v) through (vii). Consistent 

with proposed § 52.110(h)(7), a report 

on irradiated fuel should only be 

submitted if irradiated fuel is on site. 

5. Change to 10 CFR Part 72 

The NRC proposes to revise § 72.30 so 

that the submittals subsequent to the 

initial decommissioning funding plan 

would no longer require NRC approval. 

The NRC found little benefit in 

approving subsequent decommissioning 

funding plans for ISFSIs because the 

financial assurance mechanisms 

employed are very similar to those used 

for nuclear power reactors. The 

experience to date is that 

decommissioning funding plans have 

not changed substantively because of 

the passive nature of the ISFSI design, 

the static nature of ISFSI operations 

after loading, and the fact that there are 

no liquids or liquid effluents present in 

dry cask storage facilities. In addition, 

the NRC expects that the frequency of 

events that could potentially impact the 

decommissioning funding plan (i.e., due 

to spills, facility modifications, or 

changes in possession limits that are 

cited in § 72.30(c)) would continue to be 

low. However, if they were to occur, it 

is important that these events be 

factored into the cost of 

decommissioning. This change would 

make the processes under § 72.30(c) 

more efficient and less burdensome to 

the licensee and the NRC, while still 

maintaining reasonable assurance of 

adequate funding for the 

decommissioning of ISFSIs. 

G. Offsite and Onsite Financial 

Protection Requirements and Indemnity 

Agreements 

The NRC proposes to amend its 

financial protection regulations under 

10 CFR part 140, ‘‘Financial Protection 

Requirements and Indemnity 

Agreements,’’ and § 50.54(w) to address 

instances where a decommissioning 

reactor licensee may not need to 

maintain its full amounts of offsite 

liability insurance and onsite property 

insurance. Reductions in insurance 


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amounts may be warranted 

commensurate with the reduction in 

probability of an incident at a reactor in 

decommissioning and also a reduction 

in the offsite and onsite consequences 

from this event. The proposed financial 

protection requirements would codify 

the approach currently used by the NRC 

to approve exemptions from the 

financial protection requirements for 

decommissioning 10 CFR part 50 and 10 

CFR part 52 nuclear power reactor 

licensees. The proposed changes would 

also increase efficiency and 

transparency in this area by clarifying 

the requirements for financial protection 

of decommissioning plants, providing 

for regulatory certainty, and reducing 

regulatory burden without affecting 

public health and safety. Specifically, 

these proposed requirements would 

represent a graded approach, including 

the criteria to be considered, where the 

financial protection requirements for 

decommissioning sites are adjusted 

commensurate with the level of risk 

posed at two stages of the 

decommissioning process. 

Proposed revisions to 10 CFR part 140 

and § 50.54(w) would also address other 

regulatory topics including, for 

example, the applicability of procedures 

regarding extraordinary nuclear 

occurrences and a proposed new 

notification requirement for licensees 

when they make changes to the amount 

of onsite insurance. 

1. Proposed Revisions to Offsite 

Liability and Onsite Property Insurance 

Requirements 

The NRC proposes to allow 10 CFR 

part 50 and 10 CFR part 52 nuclear 

power reactor licensees in 

decommissioning to reduce the offsite 

liability and onsite property insurance 

amounts that they are required to 

maintain under §§ 140.11 and 50.54(w), 

respectively, without obtaining 

exemptions from the NRC’s regulations. 

Instead, as proposed under 

§§ 140.11(a)(5) and 50.54(w)(5), once 

certain criteria are satisfied, licensees 

could reduce their financial protection 

to the amounts in Level 2 in Table 3: 

TABLE 3—TWO-STEP GRADED APPROACH

Level Reactor site description Offsite requirement 

(§ 140.11) 

Onsite requirement 

(§ 50.54(w)) 

1 ........................ Operating or Permanently Ceased Operations and 

Permanently Defueled.

$450 million; participation in the industry retrospective rating plan.$1.06 billion. 

2 ........................ Sufficiently Decayed Fuel; ≥1,000 gallons of radioactive waste.$100 million; withdrawal from the industry retrospective rating plan.$50 million. 

Licensees in Level 1 of the graded 

approach would be required to maintain 

the full amounts of offsite liability and 

onsite property insurance currently 

required in §§ 140.11(a)(4) and 50.54(w), 

respectively, until the probability of a 

zirconium fuel cladding fire in the spent 

fuel pool is minimized. Maintaining the 

full level of insurance recognizes the 

potential for liability insurance claims 

following an accident of this type and 

the need for available resources to clean 

up the site. 

The transition to Level 2 financial 

protection amounts for licensees would 

be optional and could occur after the 

passage of a specified amount of time 

(i.e., 10 months for BWRs or 16 months 

for PWRs, beginning on the date of 

permanent cessation of operations, plus 

the NRC’s docketing of the certifications 

required by § 50.82(a)(1) or § 52.110(a) 

or after the lapse of an NRC-approved 

alternative time period to the 10 or 16 

month spent fuel decay period that is 

submitted under § 50.54(q)(7)(ii)(A) or 

(B)). For the latter option, licensees 

would need to submit an analysis that 

demonstrates a reduced risk of a 

zirconium fuel cladding fire in the SFP. 

The reduction in the financial 

protection amounts as identified in 

Table 3 (i.e., $100 million in offsite 

liability insurance and withdrawal from 

the industry retrospective rating plan) 

was modeled on the offsite liability 

claims experience from the accident at 

Three Mile Island Unit 2 as documented 

in SECY–93–127, ‘‘Financial Protection 

Required of Licensees of Large Nuclear 

Power Plants During Decommissioning’’ 

(ADAMS Accession No. ML12257A628). 

SECY–93–127 provides a reasonable 

basis for using the Three Mile Island 

Unit 2 experience as a model for 

determining the appropriate liability 

insurance coverage level for a 

permanently shutdown reactor that has 

completed its respective spent fuel 

cooling period. Additionally, as 

documented in SECY–93–127, the 

reduced onsite financial protection 

amount in Table 3 (i.e., $50 million in 

onsite property insurance coverage) was 

modeled on the potential onsite cleanup 

costs from a radiological incident 

involving the rupture of a large liquid 

radioactive waste storage tank (∼450,000 

gallons) containing slightly radioactive 

water. This event was selected as 

conceivable and a bounding scenario 

having negligible radiological 

consequences offsite. 

The spent fuel heat-up analysis 

performed by the licensee for purposes 

of reducing its insurance amounts to 

those in Level 2 could be the same 

analysis that the licensee performs to 

relax the offsite emergency planning 

requirements under proposed 

§ 50.54(q)(7)(ii)(A) or (B). The transition 

to Level 2 would prompt the licensee to 

notify the NRC under § 140.15(e) of a 

material change in financial 

protection—a reduction in offsite 

primary financial protection from $450 

million to $100 million and withdrawal 

from the industry retrospective rating 

plan. The NRC proposes a conforming 

change to § 50.54(w) for a similar 

notification of a material change to 

onsite property insurance amounts. 

The NRC is also proposing to 

periodically adjust the offsite and onsite 

financial protection amounts for 

decommissioned reactors to account for 

inflation. These adjustments would be 

in accordance with the aggregate 

percentage change in the Consumer 

Price Index and performed at intervals 

that coincide with the inflation 

adjustments for the retrospective 

premium under Section 170t of the 

AEA. 

2. Proposed Revision to Extraordinary 

Nuclear Occurrences Requirements 

The NRC proposes to amend its 

regulations in § 140.81, ‘‘Scope and 

purpose,’’ to clarify the applicability of 

the requirements for an Extraordinary 

Nuclear Occurrence (ENO) to reactors in 

decommissioning. Under Sections 11 

and 170 of the AEA, and NRC 

regulations at subpart E, ‘‘Extraordinary 

Nuclear Occurrences,’’ to 10 CFR part 

140, the NRC is authorized to make a 

determination as to whether an event at 

a production or utilization facility 

causing a discharge or dispersal of 

source, special nuclear, or byproduct 

material that has resulted or will result 

in substantial damages to offsite 

members of the public or property is an 


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ENO. An event will qualify as an ENO 

if the NRC determines that the criteria 

in § 140.84, ‘‘Criterion I—Substantial 

discharge of radioactive material or 

substantial radiation levels offsite,’’ and 

§ 140.85, ‘‘Criterion II—Substantial 

damages to persons offsite or property 

offsite,’’ have been met. 

The NRC recognizes that the 

radiological consequences resulting 

from an accident at a decommissioning 

reactor in Level 1 can be similar to those 

from an accident at an operating reactor. 

As presented in NUREG–1738, in the 

timeframe beginning immediately after 

the reactor is defueled and the fuel is 

placed in the SFP, the radiological 

consequences of a zirconium fire may be 

comparable to those from operating 

reactor postulated severe accidents. The 

existing potential consequences from a 

zirconium fire, until the fuel in the SFP 

has sufficiently decayed, provides the 

basis for the NRC’s proposal to amend 

its regulations to include plants in 

decommissioning within the scope of 

§ 140.81. 

3. Proposed New Rule Language in 

§ 50.54(w)(6) 

The NRC proposes to amend 

§ 50.54(w) to require a prompt 

notification to the Commission of any 

material change in proof of onsite 

property insurance filed with the 

Commission under 10 CFR part 50. 

Specifically, the transition to Level 2 as 

proposed by the NRC would prompt the 

licensee to notify the NRC under 

§ 50.54(w)(6) of a reduction in onsite 

property insurance from $1.06 billion to 

$50 million. This proposed amendment 

to § 50.54(w)(6) would be a conforming 

change, for consistency, with the 

existing offsite financial protection 

requirements under § 140.15(e). 

H. Environmental Considerations 

1. Clarifying Changes to 10 CFR Parts 50 

and 52 

A nuclear power reactor licensee’s 

transition from operating to 

decommissioning status does not 

involve an agency action that would 

trigger NRC responsibilities under 

environmental statutes, such as the 

National Environmental Policy Act 

(NEPA), the Endangered Species Act 

(ESA), or the National Historic 

Preservation Act (NHPA). However, 

§ 50.82(a)(4)(i) (for nuclear power 

reactors licensed under 10 CFR part 50) 

and § 52.110(d)(1) (for nuclear power 

reactors licensed under 10 CFR part 52) 

require that PSDARs provide the 

reasons for concluding that appropriate 

previously issued environmental impact 

statements (EIS) will bound the 

environmental impacts associated with 

site-specific decommissioning activities. 

After the PSDAR is submitted, the 

licensee must remain in compliance 

with § 50.82(a)(6)(ii) or § 52.110(f)(2), as 

applicable. These regulations state that 

licensees may not perform any 

decommissioning activities, as defined 

in § 50.2, that result in significant 

environmental impacts not previously 

reviewed. As explained in the 1996 

Final Rule, the requirement in 

§ 50.82(a)(6)(ii) functions as a 

prohibition against the licensee 

performing a decommissioning activity 

that would result in a significant impact 

‘‘not previously reviewed’’ (61 FR 

39283, 39286, and 39291; July 29, 1996). 

The NRC may develop updates to IMC 

2561, ‘‘Decommissioning Power Reactor 

Inspection Program,’’ and the related 

Inspection Procedure (IP) 71801, 

‘‘Decommissioning Performance and 

Status Review at Permanently 

Shutdown Reactors,’’ dated August 11, 

1997, to provide guidance on 

inspections for compliance with 

§ 50.82(a)(6)(ii) or § 52.110(f)(2) with 

respect to environmental reviews. 

In certain circumstances, licensees 

may be unable to satisfy the requirement 

that licensees conclude in the PSDAR 

that all environmental impacts 

associated with site-specific 

decommissioning activities will be 

bounded by previous EISs. For example, 

NUREG–0586, Supplement 1, Volumes 

1 and 2, ‘‘Generic Environmental Impact 

Statement on Decommissioning of 

Nuclear Facilities: Regarding the 

Decommissioning of Nuclear Power 

Reactors’’ (Decommissioning GEIS) 

(ADAMS Accession No. ML023470327), 

identified several resource areas that 

were not generically resolved. If the 

EISs previously prepared for the 

construction and initial operation of the 

plant, for license renewal, or for another 

licensing action did not include sitespecific analyses for those resource 

areas not generically resolved under the 

Decommissioning GEIS, then the 

licensee would be unable to make the 

determination in the PSDAR that all 

impacts will be bounded. Therefore, the 

licensee would have to either change its 

planned decommissioning activities so 

that the impacts would be bounded or 

submit and have approved a license 

amendment request or an exemption 

request to satisfy § 50.82(a)(4)(i) or 

§ 52.110(d)(1) prior to conducting the 

subject decommissioning activity. 

The NRC proposes to change the 

PSDAR requirements in § 50.82(a)(4)(i) 

and § 52.110(d)(1) to require that 

licensees provide the basis for 

determining whether the environmental 

impacts from site-specific 

decommissioning activities are bounded 

by previous environmental reviews. 

This proposed rule change would clarify 

that licensees, at the PSDAR stage, are 

required to evaluate the environmental 

impacts and provide in the PSDAR the 

basis for whether the proposed 

decommissioning activities are bounded 

by previously issued, site-specific or 

generic environmental reviews. Given 

that some decommissioning activities 

will occur well in the future, licensees 

might not be able to make the definitive 

conclusion that impacts will be 

bounded at the PSDAR stage. Therefore, 

the proposed change would provide 

licensees flexibility to address any 

unbounded environmental impacts 

closer to, but still prior to, the 

decommissioning activity being 

undertaken that could cause the 

unbounded impact. In that case, the 

licensee should identify in the PSDAR 

the decommissioning activities that are 

not bounded by previous environmental 

reviews and will be addressed in the 

future. This proposed change would be 

consistent with the purpose of the 

PSDAR, as noted in RG 1.185, Revision 

1, ‘‘Standard Format and Content for 

Post-Shutdown Decommissioning 

Activities Report’’ (ADAMS Accession 

No. ML13140A038), as a mechanism for 

NRC oversight because it would alert 

the NRC to any potentially unbounded 

environmental impacts associated with 

planned site-specific decommissioning 

activities. If a licensee were to consider 

a proposed decommissioning activity 

that would otherwise be prohibited by 

§ 50.82(a)(6)(ii) or § 52.110(f)(2), then 

prior to undertaking that activity, the 

licensee could submit a request for a 

license amendment or an exemption 

request, decide not to perform the 

proposed activity, or modify the 

proposed activity so that the unbounded 

environmental impact does not occur. If 

the licensee chose to submit a license 

amendment or exemption request, then 

the request would trigger NRC 

responsibilities under environmental 

statutes. In addition, prior to performing 

a decommissioning activity that is 

inconsistent with the PSDAR but 

permitted by § 50.59, the licensee must 

notify the NRC in writing, with a copy 

to the affected States, in accordance 

with § 50.82(a)(7). This § 50.82(a)(7) 

requirement is in the current regulation 

and would not be changed in this 

proposed rule. 

The NRC also proposes to change the 

§ 50.82(a)(4)(i) and § 52.110(d)(1) 

regulations to allow licensees to use 

appropriate federally issued 

environmental review documents 

prepared in compliance with NEPA, 


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ESA, NHPA, or other environmental 

statutes instead of only EISs. One reason 

for replacing the phrase ‘‘previously 

issued environmental impact 

statements’’ with ‘‘federally issued 

environmental review documents’’ is 

the NRC can, in many instances, satisfy 

its NEPA compliance obligations by the 

preparation of an environmental 

assessment or through a categorical 

exclusion finding rather than preparing 

an EIS. A second reason is that this 

change allows licensees to use a wider 

range of documents that address various 

resources. Examples of appropriate 

federally issued environmental review 

documents include environmental 

assessments prepared for license 

amendments such as extended power 

uprates; documents prepared during 

Section 7 consultations under the ESA 

such as biological opinions and 

biological assessments; or programmatic 

agreements prepared through Section 

106 consultations under the NHPA to 

resolve impacts to historic properties. 

Environmental review documents 

prepared by other Federal agencies 

could also be used if they were relevant 

to the impacts associated with the sitespecific decommissioning activities. 

The regulations in § 50.82(a)(6)(ii) and 

§ 52.110(f)(2) prohibit a licensee from 

undertaking a decommissioning activity 

that would result in a significant 

environmental impact not previously 

reviewed. The NRC is also proposing to 

change § 50.82(a)(6)(ii) and 

§ 52.110(f)(2) to clarify that the previous 

review of any potentially significant 

environmental impact must be bounded 

by appropriate federally issued 

environmental review documents 

prepared in compliance with NEPA, 

ESA, NHPA, or other environmental 

statutes. In this regard, the 

determination of significance should be 

made in terms of the appropriate federal 

environmental resource protection 

statute. For example, if a proposed 

decommissioning activity were likely to 

result in a potential adverse effect upon 

a historic property, as the term ‘‘adverse 

effect’’ is described in the Advisory 

Council on Historic Preservation 

regulation, 36 CFR 800.5, ‘‘Typical 

classes of action,’’ then that potential 

adverse effect would most likely be 

equivalent to a potential significant 

impact under § 50.82(a)(6)(ii) or 

§ 52.110(f)(2). Similarly, for species 

listed under the ESA, the equivalent 

threshold would be a proposed 

decommissioning activity that could 

result in a ‘‘take,’’ as that term is defined 

in 16 U.S.C. 1532(19), of any listed 

species at the time of the proposed 

decommissioning activity. 

These proposed changes would 

reduce the regulatory burden on the 

licensee by removing the duplicative 

requirement to address unbounded 

environmental impacts at the PSDAR 

stage. Instead, licensees would only 

prepare an environmental report or 

provide other information as requested 

by the NRC under § 51.41, 

‘‘Requirement to submit environmental 

information,’’ before performing any 

decommissioning activity that is likely 

to result in a significant impact not 

previously bounded. 

2. Consistency Changes to 10 CFR Part 

51 

Currently, § 51.53(d) requires that an 

applicant for a license amendment 

authorizing decommissioning activities 

for a production or utilization facility 

either for unrestricted use or continuing 

use restrictions submit an 

environmental report. The regulation at 

§ 51.95(d) states that the NRC will 

prepare a supplemental EIS or an 

environmental assessment in 

connection with an amendment of a 

license to authorize decommissioning 

activities. 

The 1996 Final Rule eliminated the 

requirement for nuclear power reactor 

licensees to seek NRC authorization for 

decommissioning. Therefore, there was 

no need for licensees to submit a license 

amendment or to prepare and submit a 

supporting environmental report, and 

thus no federal action that would 

require the NRC to prepare a NEPA 

document. In response to the 1995 

decommissioning proposed rule, 

commenters suggested that revisions 

should be made to then-§ 51.53, 

‘‘Supplement to environmental report,’’ 

and then-§ 51.95, ‘‘Supplement to final 

environmental impact statement,’’ to 

reflect the rule change. However, the 

NRC at that time decided not to amend 

the 10 CFR part 51 regulations because 

non-power reactor facilities were still 

required to submit a decommissioning 

plan. 

The NRC proposes to revise 10 CFR 

part 51 to reflect the changes made in 

the 1996 Final Rule that nuclear power 

reactor licensees are not required to 

submit license amendment requests for 

authorization to perform 

decommissioning activities. In 

§ 51.53(d), the NRC is proposing to 

remove language referencing an 

amendment for authorizing 

decommissioning activities and the 

requirement to prepare an 

environmental report for nuclear power 

reactors only. In § 51.95(d), the NRC is 

similarly proposing to remove language 

referencing an amendment for 

authorizing decommissioning activities. 

The NRC further proposes to revise 

§ 51.95(d) to indicate that the NRC 

would prepare the necessary NEPA 

document upon the submittal of a 

license amendment requesting approval 

of a license termination plan. The NRC 

also proposes to add a cross-reference to 

§ 52.110 in § 51.53, ‘‘Postconstruction 

environmental reports,’’ as reactors 

licensed under 10 CFR part 52 will 

perform decommissioning under 

§ 52.110, not § 50.82. 

The NRC is not proposing to make 

any changes in 10 CFR part 51 that 

would impact non-power production or 

utilization facilities (e.g., research and 

test reactors) or fuel reprocessing plants. 

Non-power production or utilization 

facility and fuel reprocessing plant 

licensees must continue to submit a 

license amendment requesting approval 

for a decommissioning plan and to 

prepare and submit the appropriate 

supporting environmental report, and 

the NRC would continue to prepare the 

appropriate NEPA documentation. 

I. Record Retention Requirements 

The NRC’s regulations require nuclear 

power reactor licensees to retain the 

records associated with certain SSCs 

until the license is terminated and 

sometimes require that these records be 

kept in duplicate. To decrease the 

burden associated with long-term record 

storage and increase the overall 

efficiency of the decommissioning 

process, licensees that are transitioning 

to decommissioning frequently request 

exemptions from these requirements. 

Although this approach continues to 

meet the underlying purpose of the 

recordkeeping regulations, the process 

of preparing, submitting, and reviewing 

exemptions from the record retention 

requirements is not an efficient use of 

NRC or licensee resources given the fact 

that the subject records are no longer 

needed to support any NRC-regulated 

function. In addition, maintaining the 

current regulations with respect to 

record retention during 

decommissioning can create a situation 

wherein the facilities used to store 

records are ready to be dismantled in 

support of site decommissioning before 

the necessary exemptions can be 

processed. The NRC proposes to resolve 

these issues by amending its regulations 

in this rulemaking. 

The recordkeeping requirements at 

issue include the following: 

• Criterion XVII, ‘‘Quality Assurance 

Records,’’ of appendix B, ‘‘Quality 

Assurance Criteria for Nuclear Power 

Plants and Fuel Reprocessing Plants,’’ to 

10 CFR part 50 requires licensees to 

retain certain records consistent with 


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regulatory requirements for a duration 

established by the licensees. 

• Sections 50.59(d)(3) and 52.63(b)(2) 

require licensees to maintain certain 

records until termination of a license 

issued under 10 CFR part 50 or 10 CFR 

part 52. 

• Section 50.71(c) requires licensees 

to maintain certain records consistent 

with various elements of the NRC 

regulations, facility TSs, and other 

licensing basis documents. 

• Section 72.72(d) requires licensees 

to duplicate certain records of spent fuel 

and high-level radioactive waste and 

store them in a separate location 

sufficiently remote from the original 

records so that a single event would not 

destroy both sets. 

Licensees that have previously 

requested exemptions from these 

requirements used the justification that, 

when the SSCs associated with these 

records are removed from service and 

the licensing basis documents, the SSCs 

will no longer serve any NRC-regulated 

function. Therefore, it would no longer 

be necessary to retain the records. In 

addition, several licensees requesting an 

exemption from the requirements of 

§ 72.72(d) used the justification that 

they will store the ISFSI spent fuel 

records using the same procedures and 

processes used for the facility spent fuel 

(and other) records, which are typically 

stored in accordance with the NRCapproved quality assurance program 

(QAP). 

The NRC granted the previous record 

retention exemptions based on a finding 

of reasonable assurance that the licensee 

would continue to meet the underlying 

purpose of the recordkeeping 

regulations, which is to establish the 

minimum retention periods necessary 

for the NRC to ensure compliance with 

the safety and health aspects of the 

nuclear environment and for the NRC to 

accomplish its mission to protect the 

public health and safety. In ‘‘Retention 

Periods for Records; Final Rule’’ (53 FR 

19240; May 27, 1988), the Commission 

explained that requiring licensees to 

maintain adequate records assists the 

NRC in judging compliance and 

noncompliance, to act on possible 

noncompliance, and to examine facts as 

necessary following any incident. 

Because the SSCs that were safetyrelated or important to safety during 

reactor operations or operation of the 

SFP are removed from the licensing 

basis, and subsequently removed from 

the plant during the decommissioning 

process, the records associated with 

those SSCs are no longer required to 

achieve the purpose of the 

recordkeeping and record retention 

regulations. 

Records associated with SSCs that 

maintain compliance with requirements 

or that protect public health and safety 

during the decommissioning process 

have been excluded from these 

exemptions. Examples include those 

SSCs associated with programmatic 

controls pertaining to residual 

radioactivity, security, and quality 

assurance (QA), and those SSCs 

associated with spent fuel assemblies or 

the SFP (while assemblies are still in the 

pool) and ISFSIs. These exemptions do 

not affect the record retention 

requirements of § 50.75 or any other 

requirements of 10 CFR part 50 that 

apply to decommissioning. 

Based on these exemptions, the NRC 

proposes to change the recordkeeping 

and record retention requirements such 

that once the NRC dockets a licensee’s 

notifications of permanent cessation of 

operation and permanent removal of 

fuel from the reactor vessel under 

§ 50.82(a)(2) or § 52.110(a), licensees can 

then eliminate records associated with 

SSCs that no longer serve any NRCregulated function. The NRC would 

allow this record disposal as long as 

appropriate change mechanisms, such 

as the § 50.59 evaluation process or 

NRC-approved TS changes, are used to 

assess the removal of those records to 

determine that elimination of the 

records would have no adverse impact 

on public health and safety. 

The records that would be subject to 

removal are associated with SSCs that 

had been important to safety during 

reactor operation or operation of the 

SFP, but that are no longer capable of 

causing an event, incident, or condition 

that would adversely impact public 

health and safety, as evidenced by their 

appropriate removal from the licensing 

basis documents. Since the SSCs no 

longer have the potential to cause these 

scenarios, it is reasonable to conclude 

that the records associated with these 

SSCs would not reasonably be necessary 

to assist the NRC in determining 

compliance, taking action on possible 

noncompliance, and examining facts 

following an incident. Therefore, 

retention of such records would not 

serve the underlying purpose of the 

recordkeeping regulations. 

The NRC proposes to make the 

following four changes to the 

recordkeeping and record retention 

requirements and regulatory guidance to 

enhance the efficiency of the 

decommissioning regulations: 

1. Clarify in RG 1.184, 

‘‘Decommissioning of Nuclear Power 

Reactors,’’ that the requirements in 

appendix B to 10 CFR part 50, Criterion 

XVII, concerning record retention, such 

as duration, location, and assigned 

responsibility, continue to be met with 

the recommended changes to the 

recordkeeping and record retention 

requirements. 

2. Amend § 50.71(c) to specify that 

licensees for which the NRC has 

docketed the certifications required 

under § 50.82(a)(1) or § 52.110(a) are not 

required to retain records associated 

with SSCs that have been removed from 

service using an NRC-approved change 

process. However, § 50.71(c) would 

require licensees to retain records 

important to decommissioning as 

specified under § 50.75(g). 

3. Amend §§ 50.59(d)(3) and 

52.63(b)(2) to clarify that records of 

changes in the facility must be 

maintained until the termination of the 

license except for records associated 

with SSCs removed from service using 

an NRC-approved change process after 

the NRC has docketed the certifications 

required under § 50.82(a)(1) or 

§ 52.110(a). 

4. Amend § 72.72(d) to allow that 

records of spent fuel, high-level 

radioactive waste, and reactor-related 

greater than Class C (GTCC) waste 

containing special nuclear material no 

longer be kept in duplicate, as long as 

the licensee can demonstrate that it will 

store the records in the same manner as 

it would for other QA records using a 

single storage facility subject to the 

same procedures and processes outlined 

in an NRC-approved QAP. 

In most cases, an NRC-approved QAP 

involves document storage requirements 

that meet American National Standards 

Institute (ANSI) standard N45 2.91974, 

‘‘Requirements for Collection, Storage, 

and Maintenance of Quality Assurance 

Records,’’ which specifies, in part, the 

design requirements for use in the 

construction of record storage facilities 

when the use of a single storage facility 

is desired. In approving the associated 

QAP, the NRC typically approves the 

single facility location used for the 

storage and maintenance of QA records 

at the facility, and the licensee typically 

affirms in the QAP that the record 

storage facility was constructed and is 

being maintained to meet the 

requirements of the NRC-approved 

QAP. 

Records for an ISFSI at a specific 

facility are typically classified as QA 

records and include all documents and 

records associated with the operation, 

maintenance, installation, repair, and 

modification of SSCs covered by the 

QAP. An ISFSI’s records also include 

historical records that have been 

gathered and collected during plant and 

ISFSI operations. These records are 

either required in support of the dry 

cask storage systems used at the ISFSI 


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6Paragraph III.E of appendix G to 10 CFR part 20 

uses the term ‘‘shipper,’’ which the regulation 

defines to mean ‘‘the licensed entity (i.e., the waste 

generator, waste collector, or waste processor) who 

offers low-level radioactive waste for transportation, 

typically consigning this type of waste to a licensed 

waste collector, waste processor, or land disposal 

facility operator.’’ 

or for ultimate shipment of the fuel to 

a Federal repository. The QAP typically 

allows the storage of QA records, 

including ISFSI records, to be done in 

accordance with ANSI N45 2.9–1974 in 

a single storage facility designed and 

maintained to minimize the risk of 

damage from adverse conditions. 

The retention of records required by 

§ 50.59(d)(3); § 52.63(b)(2); § 50.71(c); 

and appendix B to 10 CFR part 50, 

Criterion XVII provides assurance that 

records associated with SSCs will be 

captured, indexed, and stored in an 

environmentally suitable and retrievable 

condition. Although licensees retain the 

records required by their license as the 

plant transitions from operating 

conditions to a fully decommissioned 

state, plant dismantlement obviates the 

regulatory need for maintenance of most 

records. As the SSCs already removed 

from the licensing basis are 

subsequently dismantled and the need 

for the associated records is, on a 

practical basis, eliminated, the proposed 

rule changes would allow disposal of 

the records associated with SSCs and 

historical activities that are no longer 

relevant and thereby eliminate the 

associated regulatory and economic 

burdens of creating alternative storage 

locations, relocating records, or 

retaining irrelevant records. The 

proposed recordkeeping and record 

retention changes only expedite the 

schedule for disposition of the specified 

records. Considering the content of 

these records, their elimination on an 

advanced timetable has no reasonable 

potential of presenting any undue risk 

to public health and safety. In addition, 

upon dismantlement of the affected 

SSCs, the records have no functional 

purpose relative to maintaining the safe 

operation of the SSCs, maintaining 

conditions that would affect the ongoing 

health and safety of workers or the 

public, or informing decisions related to 

nuclear safety and security. 

In addition, the proposed change to 

the portion of § 72.72(d) to eliminate the 

requirement for ISFSI licensees to keep 

a duplicate set of records for spent fuel 

in storage, would continue to meet the 

recordkeeping requirements of appendix 

B to 10 CFR part 50 and other applicable 

10 CFR part 72 requirements for the 

storage and maintenance of spent fuel 

records in accordance with an NRCapproved QAP. Specifically, § 72.140(d) 

states that a QA program that the NRC 

has approved as meeting the applicable 

requirements of appendix B to 10 CFR 

part 50, will be accepted as satisfying 

the requirements of § 72.140(b) for 

establishing an ISFSI QA program. 

However, the licensee must also meet 

the recordkeeping provisions of 

§ 72.174, ‘‘Quality assurance records.’’ 

In addition, the proposed rule change 

would not affect the record content, 

retrievability, or retention requirements 

specified in § 72.72, ‘‘Material balance, 

inventory, and records requirements for 

stored materials,’’ or § 72.174, such that 

the licensee will continue to meet all 

other applicable recordkeeping 

requirements for the ISFSI and 

associated special nuclear materials. 

In proposing these rule changes, the 

NRC determined that the process and 

procedures used to store the ISFSI 

records (i.e., in accordance with the 

QAP at a facility designed for protection 

against degradation mechanisms such as 

fire, humidity, and condensation) would 

help ensure that the licensee will 

adequately maintain the required spent 

fuel information. Therefore, changes to 

the duplicate record requirement of 

§ 72.72(d) would not affect public health 

and safety. In addition, allowing the 

ISFSI spent fuel records to be stored in 

the same manner as that of other QA 

records for the nuclear facility would 

provide for greater efficiency in the 

storage of all records once the facility 

enters the final stages of 

decommissioning, where only the ISFSI 

facility would remain after license 

termination. 

J. Low-Level Waste Transportation 

Paragraph III.E of appendix G, 

‘‘Requirements for Transfers of LowLevel Radioactive Waste Intended for 

Disposal at Licensed Land Disposal 

Facilities and Manifests,’’ to 10 CFR part 

20, ‘‘Standards for Protection Against 

Radiation,’’ contains requirements for 

investigating, tracing, and reporting 

shipments of low-level radioactive 

waste (LLW) if the shipper 6 has not 

received notification of receipt within 

20 days after transfer. In addition, 

paragraph III.E requires the shipper to 

report such missing shipments to the 

NRC. Licensees, primarily those that are 

involved in the decommissioning 

process, frequently request an 

exemption from the requirement related 

to the 20-day receipt notification 

window. The NRC proposes to amend 

this requirement to extend the receipt 

notification window because such an 

extension would provide licensees with 

flexibility while not impacting public 

health and safety or the common 

defense and security. 

Licensees that have previously been 

granted these exemptions typically 

requested extension of the investigation 

notification window to 45 days using 

the justification that operating 

experience indicates that, while the 20- 

day receipt notification window is 

adequate for waste shipments by truck, 

waste shipments using other modes of 

shipment such as rail, barge, or mixedmode shipments, such as combinations 

of truck and rail, barge and rail, and 

barge and truck shipments, may take 

more than 20 days to reach their 

destination due to delays in the route 

that are outside the shipper’s control 

(e.g., rail cars in switchyards waiting to 

be included in a complete train to the 

disposal facility). The NRC granted the 

previous transportation investigation 

requirement exemptions based on a 

finding of reasonable assurance that the 

shipper would continue to meet the 

underlying purpose of the LLW 

transportation regulations—to require 

the shipper to investigate, trace, and 

report radioactive shipments that have 

not reached their destination, as 

scheduled, for unknown reasons. 

Under the current regulations, the 

shipper must investigate, trace, and 

report to the NRC any shipments of 

LLW for which the shipper has not 

received a notification of receipt within 

20 days after transfer unless the shipper 

receives an exemption from the 20-day 

receipt notification requirement. The 

NRC has found that exempting licensees 

from this requirement does not 

undermine public health and safety, nor 

does it increase any security risk. 

Further, the preparation and submission 

of the exemption request, and its 

review, evaluation, and approval by the 

NRC, are not efficient uses of NRC or 

licensee resources. Specifically, the 

NRC notes that allowing the receipt 

notification to be made past 20 days 

would not impact public health, safety, 

or security even if the LLW 

transportation package was situated in a 

publicly accessible area and waiting for 

continuing transport to the waste 

disposal site because: (1) Individuals in 

the vicinity of the LLW transportation 

package would receive no additional 

radiological dose above background 

levels resulting from the disposal 

container; and (2) the LLW would 

remain secured in the transportation 

package until the package can be 

delivered to the waste disposal site. The 

NRC also notes that, for LLW waste 

shipments, most shippers will use an 

electronic data tracking system 

interchange or similar tracking systems 

that allow the carrier to monitor the 


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progress of the shipments daily. Because 

of the oversight and monitoring of 

radioactive waste shipments throughout 

the journey from the nuclear facility to 

the disposal site, the loss, misdirection, 

or diversion of a shipment without the 

knowledge of the carrier or the shipper 

is unlikely. 

Therefore, the NRC proposes to 

change the requirement for the 

investigation, tracing, and reporting 

timeframe for LLW transportation to 

extend the receipt notification window 

to 45 days after the shipper transfers 

LLW from a licensed facility to a 

disposal site. This change would 

continue to meet the underlying 

purpose of appendix G to 10 CFR part 

20, paragraph III.E, which requires the 

shipper to investigate, trace, and report 

LLW shipments that have not reached 

their destination, as scheduled, for 

unknown reasons. Furthermore, 

extending the time period for 

notification of receipt to 45 days before 

requiring investigation, tracing, and 

reporting, would maintain a reasonable 

upper limit on shipment duration if a 

breakdown of normal tracking systems 

were to occur, based on operating 

experience. 

In addition, the NRC proposes 

correcting a typographical error in the 

current version of appendix G to 10 CFR 

part 20, paragraph III.E. Specifically, 

that paragraph states that LLW 

shipments must ‘‘be investigated by the 

shipper if the shipper has not received 

notification or receipt within 20 days 

after transfer . . .’’ (emphasis added). 

The ‘‘or’’ should be an ‘‘of,’’ consistent 

with the subsequent discussions in 10 

CFR part 20 regarding notifications of 

receipt and the associated exemptions 

granted in this area. Therefore, the NRC 

proposes correcting this error as part of 

this proposed rule for consistency and 

clarity within 10 CFR part 20. 

K. Spent Fuel Management Planning 

The regulation in § 72.218(a) states 

that the § 50.54(bb) spent fuel 

management program (i.e., the 

irradiated fuel management plan or 

IFMP) must include a plan for removing 

from the reactor site the spent fuel 

stored under the 10 CFR part 72 general 

license. The IFMP must show how the 

spent fuel will be managed before 

starting to decommission systems and 

components needed for moving, 

unloading, and shipping this spent fuel. 

Section 72.218(b) requires that an 

application for termination of a reactor 

operating license submitted under 

§ 50.82 or § 52.110 must also describe 

how the spent fuel stored under the 10 

CFR part 72 general license will be 

removed from the reactor site. Although 

§ 72.218 states what information the 

§ 50.54(bb) IFMP and the § 50.82 and 

§ 52.110 application for termination of a 

reactor operating license must include, 

the regulations in §§ 50.54(bb), 50.82, 

and 52.110 do not contain this 

information. 

As §§ 50.54(bb), 50.82, and 52.110 do 

not reflect or otherwise reference the 

provisions in § 72.218, this causes 

regulatory uncertainty. The NRC 

proposes to clarify and align the 

regulations in §§ 50.54(bb), 50.82, 

52.110, and 72.218 to provide regulatory 

clarity and enhance overall regulatory 

transparency and openness regarding 

decommissioning and spent fuel 

management planning. 

1. Requirements for the IFMP in 

§ 50.54(bb) and the PSDAR in § 50.82 

and § 52.110 

The PSDAR and IFMP are planning 

documents for decommissioning and 

spent fuel management, respectively. 

The current requirements for the timing 

of the submittal of the PSDAR and IFMP 

are similar, as the NRC’s regulations 

recognize that a licensee’s ability to plan 

properly and safely for 

decommissioning is closely related to 

the licensee’s ability to manage its spent 

fuel. Actions to manage spent fuel 

include activities taken prior to and 

subsequent to decommissioning. 

Therefore, a licensee’s spent fuel 

management plans and its 

decommissioning plans should be 

consistent. 

Because § 50.54(bb) already addresses 

the topic of spent fuel management 

planning, the NRC proposes including 

the § 72.218 provisions in § 50.54(bb) to 

clarify that the § 50.54(bb) IFMP must be 

submitted by the licensee and approved 

by the NRC before the licensee starts to 

decommission SSCs needed for moving, 

unloading, and shipping the spent fuel. 

Additionally, the NRC proposes that the 

IFMP must be submitted prior to or 

within 2 years following permanent 

cessation of operations. 

The NRC proposes to further 

restructure § 50.54(bb) to clarify that the 

IFMP addresses both the safety and 

financial aspects of managing spent fuel. 

The IFMP would describe the licensee’s 

planned actions for managing spent 

fuel, how those actions would be 

consistent with the NRC requirements 

for possession of spent fuel, and any 

actions related to spent fuel 

management that would require 

amendments to the license or certificate 

of compliance or exemptions from 

applicable regulations, which is 

consistent with the current rule 

language. The IFMP would also describe 

the projected cost of managing spent 

fuel and how the licensee would 

provide funding for the management of 

the spent fuel, until title to, and 

possession of, the spent fuel is 

transferred to the Department of Energy 

(DOE), which is also consistent with the 

current rule language. The regulation in 

§ 50.54(bb) would also continue to 

require licensees to retain a copy of the 

IFMP as a record, and the NRC proposes 

to clarify that the IFMP must be retained 

until termination of the 10 CFR part 50 

or 10 CFR part 52 license. 

The NRC proposes to clarify the 

current IFMP approval process and the 

§ 50.54(bb) provisions regarding 

preliminary approval and final NRC 

review of the IFMP as part of any 

proceeding for continued licensing 

under 10 CFR part 50 or 10 CFR part 72. 

With regard to the NRC’s final review of 

the IFMP ‘‘as part of any proceeding for 

continued licensing under 10 CFR part 

50 or 10 CFR part 72,’’ these 

proceedings no longer exist as they did 

when § 50.54(bb) was first promulgated 

in 1984. In the 1984 Final Rule, the 

Commission discussed the ‘‘proceeding 

for continued licensing under part 50’’ 

as the pre-1996 reactor 

decommissioning process, where 

licensees were required to submit a 

license amendment request for approval 

of the decommissioning plan and to 

change the license from an operating 

license to a possession-only license 

before licensees could begin 

decommissioning. The NRC noted in the 

1984 Final Rule that the IFMP would 

become part of the conditions of an 

amended 10 CFR part 50 license for a 

shutdown reactor facility. After the 1996 

rulemaking, the NRC no longer requires 

submittal of a license amendment when 

a reactor ceases operations, and thus, 

there is no longer a ‘‘proceeding for 

continued licensing under part 50’’ for 

the NRC to review and approve the 

IFMP. 

The 1984 Final Rule discusses the 

‘‘proceeding for continued licensing 

under part 72’’ as the application for, 

and NRC issuance of, a 10 CFR part 72 

specific license for storage of spent fuel 

in an ISFSI. The 1984 issuance of 

§ 50.54(bb) preceded the general license 

ISFSI provisions, which were added to 

10 CFR part 72 in 1990. Regarding the 

10 CFR part 72 general license, storage 

of spent fuel in a general license ISFSI 

is authorized by operation of law via 

§ 72.210, so there is no NRC ‘‘licensing 

proceeding’’ or approval needed for the 

10 CFR part 72 general license. As most 

reactor licensees use the 10 CFR part 72 

general license for storage of spent fuel 

in an ISFSI, there would be no 

‘‘proceeding for continued licensing 

under part 72’’ for the NRC to review 


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and approve the IFMP. Therefore, the 

NRC proposes to require submittal of 

the IFMP to the NRC as a license 

amendment request. The NRC also 

proposes to require licensees to submit 

to the NRC any changes to the IFMP as 

an application for an amendment to its 

license. 

2. Requirements in § 72.218 for 

Termination of the General License for 

Spent Fuel Storage 

Because the current spent fuel 

management planning provisions of 

§ 72.218 are initiated by reactor 

shutdown and are related to reactor 

decommissioning, the requirements fit 

best in 10 CFR part 50 and are not 

necessarily needed in 10 CFR part 72. 

Therefore, as the NRC proposes adding 

the spent fuel management provisions 

from § 72.218 into § 50.54(bb), the NRC 

also proposes deleting those provisions 

from § 72.218. In addition, the NRC 

proposes revising § 72.218 to address 

requirements related to termination of 

the 10 CFR part 72 general license, as 

the current title of § 72.218, 

‘‘Termination of licenses,’’ suggests. 

The 10 CFR part 72 general license is 

issued to 10 CFR part 50 or 10 CFR part 

52 licensees, per the regulation in 

§ 72.210. It follows that the 10 CFR part 

72 general license would terminate 

coincident with the termination of the 

10 CFR part 50 or 10 CFR part 52 

license. In addition, since the general 

license ISFSI is part of the 10 CFR part 

50 or 10 CFR part 52 licensed site, 

decommissioning of the general license 

ISFSI would follow the reactor 

decommissioning process in § 50.82 or 

§ 52.110, respectively. This approach 

would also be consistent with the NRC’s 

approach to ISFSI decommissioning 

funding as discussed in the 

‘‘Decommissioning Funding Assurance’’ 

section of this document. 

However, to provide regulatory clarity 

between 10 CFR parts 50, 52, and 72 in 

terms of decommissioning and 

termination of the 10 CFR part 72 

general license, the NRC proposes to 

revise § 72.218 to include the following 

provisions: (1) The general license ISFSI 

must be decommissioned consistent 

with the requirements in § 50.82 or 

§ 52.110; and (2) the general license is 

terminated upon termination of the 10 

CFR part 50 or 10 CFR part 52 license. 

This proposed change would provide 

regulatory clarity among 10 CFR parts 

50, 52, and 72 in terms of 

decommissioning and termination of the 

10 CFR part 72 general license, 

analogous to the provision in § 72.210 

that ties the issuance of the 10 CFR part 

72 general license to the existence of the 

10 CFR part 50 or 10 CFR part 52 

license. 

L. Backfit Rule 

For nuclear power reactor licensees, 

the NRC’s backfitting provisions are 

located in § 50.109, ‘‘Backfitting,’’ and 

the issue finality provisions are in 10 

CFR part 52 (hereinafter collectively 

referred to as the ‘‘Backfit Rule’’). The 

language of the Backfit Rule clearly 

applies to a licensee designing, 

constructing, or operating a nuclear 

power facility. For example, 

§ 50.109(a)(1) defines ‘‘backfitting’’ to 

mean changes to, among other things, 

the procedures or organization required 

to design, construct or operate a facility. 

The application of the Backfit Rule to 

decommissioning plants is not as clear. 

In SECY–98–253, ‘‘Applicability of 

Plant-Specific Backfit Requirements to 

Plants Undergoing Decommissioning,’’ 

dated November 4, 1998 (ADAMS 

Accession No. ML992870107), the NRC 

staff presented the Commission with a 

list of reasons underlying this 

uncertainty: 

• The Backfit Rule has no end point 

when the rule no longer applies, 

‘‘thereby implying that backfit 

protection continues into 

decommissioning and up to the point of 

license termination.’’ 

• The term ‘‘operate’’ could 

reasonably be interpreted as including 

activities to decommission the reactor. 

• The Backfit Rule was developed 

when the decommissioning of plants 

was not an active area of regulatory 

concern. 

• The Backfit Rule’s definition of 

‘‘backfitting’’ uses terms associated with 

the design, construction, and operation 

of a facility rather than with its 

decommissioning, although the staff 

noted in SECY–98–253 that ‘‘prior to the 

1996 decommissioning rule, the 

Commission regarded decommissioning 

as a phase of the plant’s life cycle which 

is different from the operational phase.’’ 

• Two of the factors used in 

evaluating a backfit—costs of 

construction delay/facility downtime, 

and changes in plant/operational 

complexity—are targeted to power 

operation and are ‘‘conceptually 

inappropriate in evaluating the impacts 

of a backfit on a decommissioning 

plant.’’ 

• The SOC for the 1970 (35 FR 5317; 

March 31, 1970), 1985 (50 FR 38097; 

September 20, 1985), and 1988 (53 FR 

20603; June 6, 1988) final Backfit Rules 

did not discuss any aspect of 

decommissioning, focusing instead on 

construction and operation. 

• Proposed changes to 

decommissioning requirements usually 

focused on relaxing a requirement or on 

whether a requirement applicable to an 

operating reactor continued to be 

applicable to a decommissioning plant. 

Thus, ‘‘the notion of a ‘substantial 

increase’ in protection to public health 

and safety from a backfit does not 

appear to be particularly useful [in 

decommissioning].’’ 

• The 1996 Final Rule did not 

directly respond to questions from the 

public on the applicability of the Backfit 

Rule to a decommissioning plant. 

Over the years, the NRC has tried to 

clarify the applicability of the Backfit 

Rule to nuclear power reactor licensees 

in decommissioning. In SECY–98–253, 

the NRC staff requested Commission 

approval to amend § 50.109, among 

other regulations, so that the Backfit 

Rule would clearly apply to licensees in 

decommissioning. In that paper, the 

NRC staff also proposed that, until the 

rulemaking was finished, the staff 

would apply the Backfit Rule to plants 

undergoing decommissioning ‘‘to the 

extent practical.’’ 

In the February 12, 1999, SRM for 

SECY–98–253 (ADAMS Accession No. 

ML003753746), the Commission 

approved development of a Backfit Rule 

for plants undergoing decommissioning. 

The Commission directed the NRC staff 

to continue to apply the then-current 

Backfit Rule to plants undergoing 

decommissioning until issuance of the 

final rule. The Commission directed the 

staff to develop a rulemaking plan, 

which the staff transmitted to the 

Commission in SECY–00–0145. In 

SECY–00–0145, the NRC staff proposed, 

among other decommissioning-related 

amendments to its regulations, 

amendments to § 50.109 to show clearly 

that the Backfit Rule applies during 

decommissioning and to remove factors 

that are not applicable to nuclear power 

plants in decommissioning. As 

explained in the section titled ‘‘Actions 

Leading to this 2018 Proposed Rule’’ in 

this document, the NRC ultimately did 

not conduct that rulemaking. Therefore, 

the NRC has continued to apply the 

Backfit Rule to licensee facilities 

undergoing decommissioning to the 

extent practical. 

In addition to the Commission 

direction to clarify the application of the 

Backfit Rule for decommissioning 

nuclear power reactor licensees, the 

NRC’s regulatory framework also 

supports application of the Backfit Rule 

to nuclear power reactor licensees in 

decommissioning. Under sections 101 

and 103a. of the AEA (42 U.S.C. 2131 

and 2133a.), the NRC’s issuance of a 

nuclear power reactor operating license 

under 10 CFR part 50 or a combined 

license under 10 CFR part 52 grants the 


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holder a license to, among other things, 

own, possess, and operate a ‘‘production 

facility’’ or ‘‘utilization facility,’’ as 

those terms are defined in section 11 of 

the AEA. Once the licensee under 10 

CFR part 50 or 10 CFR part 52 submits 

its certifications of permanent cessation 

of reactor operations and permanent 

removal of fuel from the reactor vessel 

and the NRC dockets those 

certifications, the licensee is no longer 

authorized to operate the reactor under 

§ 50.82(a)(2) or § 52.110(b), respectively. 

The license is no longer an ‘‘operating 

license’’ for the reactor because the 

licensee is not operating a production or 

utilization facility pursuant to sections 

101 and 103a. of the AEA. Instead, as 

described in § 50.51(b) for 10 CFR part 

50 licenses and § 52.109, ‘‘Continuation 

of combined license,’’ for 10 CFR part 

52 combined licenses, when the reactor 

has permanently ceased operations, the 

license continues in effect beyond the 

expiration date and authorizes 

ownership and possession of the facility 

until the Commission terminates the 

license. Thus, when the licensee is no 

longer authorized to operate the reactor, 

it retains its possession and ownership 

authority under its 10 CFR part 50 or 10 

CFR part 52 facility license. 

Although a decommissioning 

licensee’s license no longer authorizes 

operation of the reactor because the 

licensee is not operating a production or 

utilization facility, the licensee still 

must ‘‘operate’’ certain SSCs at the site. 

Under § 50.51(b) (with a similar 

requirement in § 52.109 for combined 

license holders), when the licensee has 

only a possession and ownership 

license for the reactor, the licensee must 

not only decommission and 

decontaminate the facility, but also 

continue to maintain the facility, 

including storing, controlling and 

maintaining the spent fuel in a safe 

condition. Therefore, nuclear power 

reactor licensees store, control, and 

maintain spent fuel after permanent 

cessation of reactor operations through 

the ‘‘operation’’ of an SFP and ISFSI. 

Although § 50.109(a)(1) defines 

‘‘backfitting’’ as changes to, among other 

things, the procedures or organization 

required to design, construct, or operate 

a facility, indicating that the Backfit 

Rule applies only to a holder of a 

license to ‘‘operate a facility,’’ the 

language of § 50.51(b) shows that 

‘‘operating a facility’’ can be interpreted 

to mean more than just operating a 

reactor. This is supported by the 

Commission direction in the SRM for 

SECY–98–253 that the NRC staff 

develop a Backfit Rule for plants 

undergoing decommissioning (i.e., 

when the licensee no longer operates a 

reactor) and continue to apply the thencurrent Backfit Rule to plants 

undergoing decommissioning until 

issuance of the final rule. Thus, the 

Backfit Rule still applies to a licensee 

that has a license to only possess and 

own a facility. For a facility in 

decommissioning, the phrase ‘‘operate a 

facility’’ in § 50.109(a)(1) is read to 

encompass operating the SFP and 

associated SSCs necessary for 

compliance with § 50.51(b). 

As the Commission and the NRC staff 

recognized in the 1990s, certain 

provisions of the Backfit Rule do not 

clearly apply to nuclear power reactor 

licensees in decommissioning. In this 

proposed rule, the NRC proposes to 

complete the process begun two decades 

ago to clarify the application of the 

Backfit Rule to nuclear power reactor 

licensees in decommissioning. 

The NRC proposes to amend § 50.109 

so that nuclear power reactor licensees, 

which have had their § 50.82(a)(1) or 

§ 52.110(a) certifications docketed by 

the NRC, are the subject of similar 

backfitting provisions as they were 

during their operating phase. A new 

backfitting provision for licensees in 

decommissioning would eliminate any 

confusion with the meaning of the 

words ‘‘operate a facility’’ in 

§ 50.109(a)(1), as compared to other uses 

of the term ‘‘operate’’ in 10 CFR Chapter 

I. 

The NRC would make other revisions 

to § 50.109. To make the section easier 

to read, the NRC proposes to insert 

paragraph headings. The NRC would 

remove current § 50.109(b) regarding 

backfits imposed prior to October 21, 

1985, because the language is obsolete 

and no longer needed. In the current 

§ 50.109(a)(6), the NRC proposes to 

insert a sentence explaining that a 

documented evaluation, which is used 

by the NRC to justify not performing a 

backfit analysis, must include a 

consideration of the costs of imposing 

the backfit if the basis for backfitting is 

bringing a facility into compliance with 

a license or the rules or orders of the 

Commission, or into conformance with 

the licensee’s written commitments. 

Further, the NRC proposes to make 

conforming changes to § 72.62 to clarify 

that the corresponding backfit 

regulations in part 72 apply during the 

decommissioning of an ISFSI or a 

Monitored Retrievable Storage facility 

subject to those provisions. 

M. Foreign Ownership, Control, or 

Domination 

The NRC is proposing to amend its 

regulations to address the circumstances 

when a facility licensed under 10 CFR 

part 50 or 10 CFR part 52 no longer 

meets the definition of a utilization 

facility or a production facility. The 

AEA has certain requirements specific 

to utilization or production facilities. By 

clarifying when a 10 CFR part 50 or 10 

CFR part 52 licensed facility is no 

longer a utilization or a production 

facility, the NRC can then specify 

whether these AEA requirements still 

apply to the licensee for that facility. 

For instance, the AEA prohibits the 

issuance of a license for a utilization or 

a production facility to an entity that the 

Commission knows or has reason to 

believe is foreign owned, controlled, or 

dominated. The Commission’s 

regulations that implement this 

prohibition, however, are unclear as to 

when a facility undergoing 

decommissioning is no longer a 

utilization or a production facility. 

Given this uncertainty, licensees have 

requested exemptions from § 50.38, 

‘‘Ineligibility of certain applicants,’’ to 

transfer 10 CFR part 50 licenses for 

facilities that no longer meet the 

definition of utilization facility. The 

NRC proposes to amend its regulations 

to clarify when a facility licensed under 

10 CFR part 50 or part 52 is not 

considered a production or utilization 

facility and therefore, the FOCD 

prohibition no longer applies. 

The NRC’s regulations in 10 CFR parts 

50 and 52 provide for the issuance of a 

10 CFR part 50 license for a utilization 

or a production facility and a 10 CFR 

part 52 license for a utilization facility. 

The AEA defines ‘‘utilization facility’’ 

as: 

(1) Any equipment or device, except an 

atomic weapon, determined by rule of the 

Commission to be capable of making use of 

special nuclear material in such quantity as 

to be of significance to the common defense 

and security, or in such manner as to affect 

the health and safety of the public, or 

peculiarly adapted for making use of atomic 

energy in such quantity as to be of 

significance to the common defense and 

security, or in such manner as to affect the 

health and safety of the public; or (2) any 

important component part especially 

designed for such equipment or device as 

determined by the Commission. 

The AEA defines ‘‘production 

facility,’’ in part, as: 

(1) Any equipment or device determined 

by rule of the Commission to be capable of 

the production of special nuclear material in 

such quantity as to be of significance to the 

common defense and security, or in such 

manner as to affect the health and safety of 

the public; or (2) any important component 

part especially designed for such equipment 

or device as determined by the Commission. 

As authorized by the AEA, the 

Commission has a rule defining 

utilization facility and production 


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facility. In § 50.2, a utilization facility is 

defined as either (1) any nuclear reactor 

other than one designed or used 

primarily for the formation of 

plutonium or U–233; or (2) an 

accelerator-driven subcritical operating 

assembly used for the irradiation of 

materials containing special nuclear 

material and described in the 

application for the SHINE Medical 

Isotope Production Facility. A 

production facility is defined as a 

nuclear reactor designed or used 

primarily for the formation of 

plutonium or uranium-233; with certain 

exceptions not relevant here, a facility 

designed or used for the separation of 

the isotopes of plutonium; or, with 

certain exceptions not relevant here, a 

facility designed or used for the 

processing of irradiated materials 

containing special nuclear material. 

NRC case law provides insight as to 

when a facility licensed under 10 CFR 

part 50 or 10 CFR part 52 is no longer 

a utilization or a production facility. In 

LBP–84–33, Cincinnati Gas & Electric 

Co. (Wm. H. Zimmer Nuclear Power 

Station, Unit 1), 20 NRC 765 (1984), an 

Atomic Safety and Licensing Board 

granted the licensee’s motion to 

withdraw its application for a 10 CFR 

part 50 operating license for a nuclear 

power reactor, despite the fact that the 

facility was almost completely built. 

One of the conditions for granting the 

motion was that the nuclear steam 

supply system be modified to prevent 

the facility’s operation as a utilization 

facility. The Board determined that 

because a utilization facility under the 

AEA is a facility that is capable of 

making use of special nuclear material, 

the facility must be modified to 

eliminate that capability for it to no 

longer be categorized as a utilization 

facility. The Board observed that this 

can be achieved, for example, by 

severing and welding caps on main 

feedwater lines and main steam lines 

and removing the fuel and the control 

rod drive mechanisms. 

The NRC proposes to add to its 

regulations language similar to the 

Zimmer decision to establish the criteria 

for when a facility licensed under 10 

CFR part 50 or 10 CFR part 52 no longer 

meets the statutory or regulatory 

definition of a utilization or a 

production facility (i.e., is no longer 

capable of making use of special nuclear 

material or of the production of special 

nuclear material, separation of the 

isotopes of plutonium, or processing of 

irradiated materials containing special 

nuclear material (hereinafter 

collectively referred to as productionfacility activities)). The first criterion is 

that the facility must not be legally 

authorized to operate. The second 

criterion is the physical modification of 

the licensed facility to be incapable of 

making use of special nuclear material 

and of production-facility activities, 

without significant facility alterations 

necessary to restore the capability to 

make use of special nuclear material or 

to engage in production-facility 

activities. When a utilization facility is 

physically modified to be incapable of 

making use of special nuclear material, 

it is no longer designed or used to 

sustain nuclear fission in a selfsupporting chain reaction. 

Sections 50.82(a)(2) and 52.110(b) 

already provide for the first criterion for 

nuclear power reactor licensees—that 

the facility is no longer legally 

authorized to operate. Sections 

50.82(a)(2) and 52.110(b) state, 

respectively, that a 10 CFR part 50 

license and a 10 CFR part 52 license no 

longer authorize operation of the reactor 

or emplacement or retention of fuel into 

the reactor vessel once the NRC has 

docketed the certifications for 

permanent cessation of operations and 

permanent removal of fuel from the 

reactor vessel, or when a final legally 

effective order to permanently cease 

operations has come into effect. The 

NRC would amend these regulations to 

add the second criterion—that the 

facility licensed under 10 CFR part 50 

or 10 CFR part 52 is no longer a 

utilization facility once the licensee 

modifies the facility to be incapable of 

making use of special nuclear material 

without significant facility alterations. 

Because the NRC’s regulations do not 

state when a non-power production or 

utilization facility or fuel reprocessing 

plant licensee is no longer authorized to 

operate (other than at license 

termination), the NRC proposes to 

amend § 50.82(b) to add the criteria for 

when a non-power production or 

utilization facility or fuel reprocessing 

plant is no longer a production or 

utilization facility. The NRC would 

renumber current paragraph (b)(6) in 

§ 50.82 as paragraph (b)(8) and add new 

paragraphs (b)(6) and (b)(7). New 

paragraph (b)(6) would provide that a 

non-power production or utilization 

facility or fuel reprocessing plant is not 

legally capable of operating when the 

NRC removes the licensee’s authority to 

operate the facility through a license 

amendment. The NRC can remove a 

non-power production or utilization 

facility or fuel reprocessing plant 

licensee’s authority to operate by 

issuing a possession-only license 

amendment or by approving the 

licensee’s decommissioning plan 

through a license amendment, either of 

which would explicitly remove the 

licensee’s authority to operate. 

Licensees typically request a 

possession-only license amendment first 

and then submit a decommissioning 

plan via a second license amendment 

request. This proposed rule would offer 

licensees the option to request only one 

licensing action—the decommissioning 

plan license amendment—that also 

would address the licensee’s operating 

authority, rendering a separate 

‘‘possession-only license amendment’’ 

unnecessary. To address those instances 

when the licensee is still operating the 

facility when the licensee submits its 

decommissioning plan license 

amendment request, the 

decommissioning plan license 

amendment would itself identify the 

date on which the authority to operate 

is removed. 

The NRC would also include in new 

§ 50.82(b)(6) the second criterion for 

when the non-power production or 

utilization facility or fuel reprocessing 

plant is no longer a production or a 

utilization facility (i.e., once the 

licensee modifies the facility to be 

incapable of production-facility 

activities and making use of special 

nuclear material without significant 

facility alterations). 

The NRC would add new § 50.82(b)(7) 

and amend § 50.82(a)(2) and § 52.110(b) 

to affirm the continuation of the NRC’s 

statutory authority over the existing 10 

CFR part 50 or 10 CFR part 52 license 

after the performance of 

decommissioning activities that lead to 

the licensed facility no longer meeting 

the definition of a utilization or a 

production facility. This facility 

transition occurs with every licensee 

during decommissioning: Eventually, 

the facility will be dismantled to the 

point where it is incapable of making 

use of special nuclear material or of 

production-facility activities without 

significant facility alterations. 

Although the facility licensed under 

10 CFR part 50 or 10 CFR part 52 may 

no longer be a utilization or a 

production facility, the NRC maintains 

the authority to regulate the existing 10 

CFR part 50 or 52 license. A 10 CFR part 

50 operating license for a production or 

utilization facility is issued under AEA 

sections 103 or 104, and a 10 CFR part 

52 combined license for a utilization 

facility is issued under AEA sections 

103 and 185b. That license may contain 

authorities beyond those governed by 10 

CFR parts 50 or 52. Under § 50.52, 

‘‘Combining licenses,’’ the Commission 

may combine in a single license the 

activities that would otherwise be 

licensed under separate licenses. 

Accordingly, a typical 10 CFR part 50 or 

52 nuclear power reactor license also 


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includes in a single license the authority 

under 10 CFR parts 30, 40, and 70 of the 

NRC’s regulations to perform activities 

or possess materials authorized by those 

parts. Parts 30, 40, and 70 of 10 CFR are 

authorized by sections 81, 63, and 53 of 

the AEA and concern the licensing of 

byproduct, source, and special nuclear 

materials, respectively. A typical 10 

CFR part 50 non-power production or 

utilization facility license also includes 

the authority under 10 CFR parts 30 and 

70 of the NRC’s regulations to perform 

activities or possess materials 

authorized by those parts. When the 

facility is no longer a production or 

utilization facility, the NRC maintains 

the authority to regulate the facility and 

the 10 CFR part 50 or 52 license under 

a combination of AEA sections 53, 63, 

81, and 161. Sections 50.51(b) and 

52.109 of the NRC’s regulations also 

establish that the 10 CFR part 50 or 52 

license continues in effect until the NRC 

terminates the license, notwithstanding 

the fact that at some point in time 

during the dismantlement required for 

license termination, the licensed facility 

will be disassembled to such an extent 

that it no longer satisfies the definition 

of a utilization or a production facility. 

Therefore, the NRC would amend 

§ 50.82(a)(2), § 50.82(b), and § 52.110(b) 

to explicitly cite these statutory 

provisions as the basis for its retention 

of the authority to regulate the existing 

10 CFR parts 50 or 52 facility. The NRC 

proposes to make conforming changes to 

the authority citations for 10 CFR parts 

50 and 52 to add sections 53, 63, and 

81 of the AEA. 

The NRC proposes to amend 

§ 50.82(a)(2), § 50.82(b), and § 52.110(b) 

to state which requirements apply to the 

existing 10 CFR part 50 or 52 license 

after the licensed facility is no longer a 

utilization or a production facility. As 

provided by section 161b of the AEA, 

the Commission is authorized to 

establish by regulation such standards 

to govern the possession and use of 

special nuclear material, source 

material, and byproduct material as the 

Commission may deem necessary or 

desirable to promote the common 

defense and security or to protect health 

or to minimize danger to life or 

property. Consistent with this statutory 

authority, the proposed amendments to 

§ 50.82(a)(2), § 50.82(b), and § 52.110(b) 

will make clear that, after the facility 

licensed under 10 CFR part 50 or 52 is 

no longer a utilization or a production 

facility and until the termination of the 

10 CFR part 50 license pursuant to 

§ 50.82(a)(11) or § 50.82(b)(8) or the 10 

CFR part 52 license pursuant to 

§ 52.110(k), the NRC regulations 

applicable to utilization or production 

facilities will continue to apply to the 

holder of the 10 CFR part 50 or 10 CFR 

part 52 license, as applicable, unless 

those regulations explicitly state 

otherwise. These proposed amendments 

would enable a licensee to maintain 

reasonable assurance of adequate 

protection of the common defense and 

security and the public health and 

safety by requiring the licensee to 

continue to comply with those 

regulations applicable to utilization or 

production facilities, as applicable to 

that licensee, unless stated otherwise. 

The NRC has identified that § 50.38 

should not apply to a facility that is no 

longer a utilization or a production 

facility. Specifically, the AEA prohibits 

the issuance of a license for a utilization 

or a production facility to an entity that 

the Commission knows or has reason to 

believe is foreign owned, controlled, or 

dominated. However, since the FOCD 

prohibition only applies to a utilization 

or production facility, it would not 

apply once a 10 CFR part 50 or part 52 

facility is no longer a utilization or a 

production facility. Therefore, the NRC 

is proposing to amend § 50.38 such that 

its prohibition on transferring a license 

to an entity that the Commission knows 

or has reason to believe is owned, 

controlled, or dominated by an alien, a 

foreign corporation, or a foreign 

government, is not applicable if the 

license is a 10 CFR part 50 or 10 CFR 

part 52 license for a facility that no 

longer meets the definition of a 

utilization or a production facility. 

Section 50.80 governs the transfers of 

10 CFR part 50 and 10 CFR part 52 

licenses for production and utilization 

facilities. It requires the written consent 

of the NRC before the transfer of a 

production or utilization facility. This 

section also requires applicants for a 

license transfer to provide the same 

identifying, technical, and financial 

information that an initial license 

applicant is required to provide under 

§§ 50.33 and 50.34. In particular, § 50.33 

requires an application to state the 

citizenship of the applicant. Under 

§ 50.38, the applicant is ineligible to 

apply for and obtain a license if it is a 

foreign entity. 

Section 50.38 implements sections 

103 and 104 of the AEA, which provide 

in part that a license for a utilization or 

production facility may not be issued to 

an alien or any corporation or other 

entity if the Commission knows or has 

reason to believe it is owned, controlled, 

or dominated by an alien, a foreign 

corporation, or a foreign government. 

Since sections 103 and 104 of the AEA 

apply to utilization and production 

facilities, the NRC is proposing to 

amend § 50.38 to clarify that this 

prohibition does not apply to a person, 

corporation, or other entity seeking a 

license for a facility that is no longer a 

utilization or a production facility, as 

would be provided under revised 

§ 50.82(a)(2), § 50.82(b)(6), or 

§ 52.110(b). 

The proposed amendment to § 50.38 

would maintain the common defense 

and security and public health and 

safety because, even though § 50.38 

would not prohibit the transfer to 

foreign entities of 10 CFR part 50 and 

10 CFR part 52 licenses for facilities that 

do not meet the definition of utilization 

or production facility, other regulations 

ensure that such transfers would not be 

inimical to the common defense and 

security or to the health and safety of 

the public. For instance, § 50.80(c) states 

that the Commission will approve an 

application for the transfer of a license 

if the Commission determines that the 

proposed transferee is qualified to be 

the holder of the license and that the 

transfer of the license is otherwise 

consistent with applicable provisions of 

law, regulations, and orders issued by 

the Commission. In turn, under § 50.57 

or § 52.97, the Commission may issue a 

10 CFR part 50 or 10 CFR part 52 

license, respectively, only if the 

Commission finds that the issuance of 

the license will not be inimical to the 

common defense and security or to the 

health and safety of the public. 

The proposed amendment to § 50.38 

is consistent with how the NRC 

analyzed requests for exemptions from 

§ 50.38 for Maine Yankee Atomic Power 

Station, Haddam Neck Plant, and 

Yankee Nuclear Power Station (78 FR 

58571; September 24, 2013). 

Specifically, the NRC granted those 

exemptions because the reactor facilities 

had been dismantled and removed such 

that only ISFSIs remained on site; an 

ISFSI, whether licensed under 10 CFR 

parts 50 or 72, is not capable of making 

use of special nuclear material; and the 

AEA definition of a utilization facility 

does not include ISFSIs. The NRC found 

that the foreign ownership, control, or 

domination prohibition did not apply to 

ISFSIs and, thus, did not preclude the 

NRC from granting the exemptions. 

The NRC is also proposing to amend 

§§ 50.1, 50.51, 52.0, and 52.109 in light 

of the proposed amendments to 

§§ 50.38, 50.82, and 52.110. The 

proposed amendments would make 

clear that the regulations in 10 CFR part 

50, and the similar regulations in 10 

CFR part 52, provide not only for the 

licensing of utilization and production 

facilities, but also for their 

decommissioning and the termination of 

their associated licenses. These changes 


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are clarifications; 10 CFR part 50 has 

included decommissioning and license 

termination since 1961 (‘‘Creditors’ 

Rights; and Transfer, Surrender, and 

Termination of Licenses,’’ 26 FR 9546; 

October 10, 1961). The NRC proposes to 

delete the language in §§ 50.51 and 

52.109 that discusses what 10 CFR parts 

50 and 52 licenses authorize in lieu of 

the more complete discussion provided 

in the proposed amendments to 

§ 50.82(a)(2), § 50.82(b)(6) and (7), and 

§ 52.110(b). 

The NRC is proposing to add a 

specific definition for ‘‘non-power 

production or utilization facility’’ to 

§ 50.2 to establish a term that is flexible 

enough to capture all non-power 

facilities licensed under § 50.22, ‘‘Class 

103 licensees; for commercial and 

industrial facilities,’’ and § 50.21(a) or 

(c), except fuel reprocessing facilities. 

This proposed rule would address 

inconsistencies in definitions and 

terminology associated with non-power 

production and utilization facilities in 

§ 50.2 that result in challenges in 

determining the applicability of the 

regulations. Fuel reprocessing plants 

would be excluded from the definition 

because the consequences associated 

with the hazards at a fuel reprocessing 

plant would likely exceed those 

anticipated at the facilities within the 

‘‘non-power production or utilization 

facility’’ definition, thereby affecting the 

applicability of the ‘‘non-power 

production or utilization facility’’ term. 

The only NRC-licensed fuel 

reprocessing plant is the Western New 

York Nuclear Service Center. The 

technical specifications of its NRC 

license are currently suspended by 

license amendment. Under the West 

Valley Demonstration Project Act, 

Public Law 96–368, 94 Stat. 1347 

(codified as a note to 42 U.S.C. 2021a), 

the Department of Energy (DOE) is 

currently decommissioning portions of 

the plant. The NRC licensee, the New 

York State Energy Research and 

Development Authority, will complete 

the decommissioning work after DOE 

has completed its work. There is 

currently no application for another fuel 

reprocessing plant and the NRC does 

not anticipate any application in the 

foreseeable future. 

The NRC proposes to revise the 

introductory text of § 50.82(b) to replace 

the term ‘‘non-power reactor licensees’’ 

with ‘‘non-power production or 

utilization facility licensees and fuel 

reprocessing plants’’ to ensure that all 

non-power facilities licensed under 

§ 50.22 or § 50.21(a) or (c) are subject to 

the relevant termination and 

decommissioning regulations. 

N. Clarification of Scope of License 

Termination Plan Requirement 

The NRC is proposing to amend its 

regulations to clarify that the 

requirement for a license termination 

plan in § 50.82(a)(9) and § 52.110(i) 

applies only to nuclear power reactor 

licensees that commenced operation. 

This clarification is being proposed in 

response to apparent confusion among 

combined license holders that have 

sought to surrender their licenses before 

operation. By letter dated November 1, 

2017 (ADAMS Accession No. 

ML17311A143), Duke Energy Florida 

informed the NRC that it would seek 

termination of the 10 CFR part 52 

combined licenses for Levy Nuclear 

Plant Units 1 and 2 and would submit 

a license termination plan in accordance 

with § 52.110(i). Subsequently, South 

Carolina Electric & Gas Company 

(SCE&G) submitted a letter dated 

December 27, 2017 (ADAMS Accession 

No. ML17361A088), seeking withdrawal 

of the 10 CFR part 52 combined licenses 

for Virgil C. Summer Nuclear Station 

Units 2 and 3. The SCE&G request 

neither cited § 52.110 nor indicated that 

it would submit a license termination 

plan. Instead, SCE&G cited the 

Commission’s final ‘‘Policy Statement 

on Deferred Plants’’ (52 FR 38077; 

October 14, 1987) (Policy Statement) to 

support its request for NRC approval to 

withdraw its combined licenses. The 

Policy Statement addresses holders of 

construction permits that defer or 

terminate plant construction. The Policy 

Statement provides that a permit holder 

can request to withdraw its permit and 

does not cite to the license termination 

provisions in 10 CFR part 50. The Policy 

Statement was issued prior to the 

promulgation of 10 CFR part 52 and has 

not been updated since, but there is 

nothing to prevent holders of a 

combined license from following the 

applicable parts of the Policy Statement 

while continuing to comply with the 

Commission’s regulations and the terms 

and conditions of the combined license. 

The requirement for a license 

termination plan in § 52.110(i) does not 

apply to plants that have not begun 

operating. While § 52.110(i) does refer to 

‘‘[a]ll power reactor licensees,’’ the 

regulatory history and context indicates 

that § 52.110 as a whole applies only to 

plants that have started operation: 

• The organization of § 52.110 

generally follows the license 

termination process for an operating 

plant, from permanent cessation of 

operations to permanent removal of fuel 

to decommissioning activities to license 

termination. The requirement for a 

license termination plan should be 

understood in this context. 

• The vast majority of the 

requirements in § 52.110 (including 

§ 52.110(i)) either explicitly refer to, or 

make sense only in the context of, a 

plant that has operated and is 

undergoing decommissioning. 

• The ‘‘[a]ll power reactor licensees’’ 

language also appears in § 50.82(a)(9), 

the 10 CFR part 50 analogue to 

§ 52.110(i). But the NRC does not apply 

the similar requirements in § 50.82 to 

holders of construction permits even 

though construction permits fall within 

the definition of ‘‘License’’ in § 50.2. For 

example, the following construction 

permit terminations do not cite or 

otherwise address § 50.82: ‘‘Washington 

Public Power Supply System, 

Washington Nuclear Project, Unit 3; 

Order Revoking Construction Permit No. 

CPPR–154’’ (64 FR 4725; January 29, 

1999); ‘‘Bellefonte Nuclear Plant, Units 

1 and 2—Withdrawal of Construction 

Permit Nos. CPPR–122 for Unit 1 and 

CPPR–123 for Unit 2’’ (September 14, 

2006) (ADAMS Accession No. 

ML061810505); and ‘‘Energy Northwest 

Nuclear Project No. 1—Termination of 

Construction Permit CPPR–134’’ 

(February 8, 2007) (ADAMS Accession 

No. ML070220011). And the rule 

issuing the ‘‘[a]ll power reactor 

licensees’’ language in § 50.82(a)(9)—the 

1996 Final Rule—was directed at 

holders of operating licenses, not 

construction permits. 

• According to the final rule issuing 

§ 52.110, ‘‘Licenses, Certifications, and 

Approvals for Nuclear Power Plants’’ 

(72 FR 49351; August 28, 2007), 

§ 52.110 and its companion regulation 

§ 52.109 were intended to be analogous 

to the requirements in § 50.51 and 

§ 50.82 for permanent shutdown of a 

nuclear power plant, its 

decommissioning, and the termination 

of the operating license. 

For these reasons, § 52.110 is best 

understood to apply only to plants that 

began operation. However, to avoid 

confusion over the license termination 

plan requirement, the NRC proposes to 

amend § 52.110(i) so that it explicitly 

applies only to ‘‘power reactor licensees 

that have loaded fuel into the reactor.’’ 

As stated in the ‘‘Final Procedures for 

Conducting Hearings on Conformance 

With the Acceptance Criteria in 

Combined Licenses’’ (81 FR 43266; July 

1, 2016), the NRC has historically 

understood operation as beginning with 

the loading of fuel into the reactor. 

Therefore, § 52.110(i) would apply to 10 

CFR part 52 nuclear power reactor 

licensees that have begun to load fuel 

into the reactor. 


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A conforming change is also proposed 

in § 50.82(a)(9) to clarify that the 

requirement in that provision—that all 

10 CFR part 50 nuclear power reactor 

licensees must submit an application for 

termination of license—applies to only 

those 10 CFR part 50 nuclear power 

reactor licensees that have loaded fuel 

into the reactor. 

O. Removal of License Conditions and 

Withdrawal of Orders 

The NRC is proposing to withdraw 

orders and remove license conditions 

that are substantively redundant with 

provisions in 10 CFR. Although NRC 

orders generally provide for their 

relaxation or rescission on a licenseespecific basis, use of that process would 

be an inefficient and unnecessary 

administrative burden on licensees and 

the NRC—with no impact on public 

health and safety—when a subsequent 

rule replaces the orders in their entirety 

for all applicable licensees. Therefore, 

the NRC is proposing to find that good 

cause is shown to rescind Order EA–06– 

137, ‘‘Order Modifying Licenses’’ 

(ADAMS Accession No. ML061600076), 

concerning mitigation strategies for 

large fires or explosions at nuclear 

power plants. This order was issued to 

certain licensees who received Order 

EA–02–026, ‘‘Order for Interim 

Safeguards and Security Compensatory 

Measures’’ (ADAMS Accession No. 

ML020510635), which required 

licensees to take specific interim 

compensatory measures, including 

mitigation strategies for large fires or 

explosions at nuclear power plants, in 

light of the then-high-level threat 

environment. Order EA–06–137 

required that licensees to incorporate 

key mitigation strategies for large fires 

or explosions into their security plans. 

The requirement that these strategies be 

incorporated in security plans was 

subsequently relaxed by letter dated 

August 28, 2006, which permitted 

licensees to consent to having their 

licenses amended to incorporate a 

license condition on the subject. Several 

licensees had these license conditions 

imposed by administrative license 

amendment (e.g., ‘‘Browns Ferry 

Nuclear Plant, Units 1, 2, and 3— 

Conforming License Amendments To 

Incorporate the Mitigation Strategies 

Required by Section B.5.b. of 

Commission Order EA–02–026 and the 

Radiological Protection Mitigation 

Strategies Required by Commission 

Order EA–06–137,’’ dated August 16, 

2007). In its Power Reactor Security 

Requirements final rule, the NRC 

established in § 50.54(hh)(2) a 

regulation that provides a performancebased requirement that encompasses the 

mitigation strategies required under 

Order EA–06–137 and its associated 

license condition. The Mitigation of 

Beyond-Design-Basis Events rule 

subsequently moved § 50.54(hh)(2) to 

§ 50.155(b)(3). As a result, neither Order 

EA–06–137 nor the license condition is 

necessary. Accordingly, the NRC 

proposes finding that good cause is 

shown to rescind Order EA–06–137 for 

each licensee that received the order. In 

addition, because § 50.155(b)(3) 

provides the same requirements as the 

license condition associated with Order 

EA–06–0137, the NRC proposes 

deeming the license condition removed 

from each applicable nuclear power 

reactor license. 

Order EA–02–026 included a section, 

numbered B.5.b, in its attachment 2, 

requiring mitigation strategies for large 

fires or explosions at nuclear power 

plants. Extensive interactions among the 

NRC, industry, and licensees refined the 

strategies required by the order. In 2007, 

the NRC issued to all then-operating 

nuclear power reactor licensees an 

administrative license amendment (e.g., 

‘‘Calvert Cliffs Nuclear Power Plant, 

Unit Nos. 1 and 2—Conforming License 

Amendments to Incorporate the 

Mitigation Strategies Required by 

Section B.5.b. of Commission Order EA– 

02–026,’’ dated July 11, 2007), 

containing a license condition entitled, 

‘‘Mitigation Strategy License 

Condition,’’ which required licensees to 

use 14 mitigation strategies. In the 

Power Reactor Security Requirements 

final rule, the NRC established in 

§§ 50.54(hh), 50.34(i), and 52.80(d) 

regulations that made the requirements 

of Order EA–02–026 generically 

applicable to nuclear power reactor 

licensees and applicants. In the Power 

Reactor Security Requirements final 

rule, the Commission explained that 

operating nuclear power reactor 

licensees already had procedures in 

place that complied with the new 

§ 50.54(hh)(2). Licensees used the same 

implementation guidance to comply 

with the Mitigation Strategy License 

Condition as they used to comply with 

§ 50.54(hh)(2); consequently, 

compliance with § 50.54(hh)(2) is 

sufficient to comply with the Mitigation 

Strategy License Condition. 

Subsequently, the NRC rescinded Order 

EA–02–026, section B.5.b by letter dated 

November 28, 2011, based on the fact 

that the regulations encompassed the 

order requirements. Because licensees 

comply with both the regulations and 

Mitigation Strategy License Condition 

via the same guidance, such that the 

former § 50.54(hh)(2) requirements 

encompass the license condition 

requirements, the NRC proposes 

concluding that § 50.155(b)(3) fully 

replaces the requirements that exist in 

the Mitigation Strategy License 

Condition and deeming that the 

Mitigation Strategy License Conditions 

imposed in 2007 are removed from the 

licenses for those licensees that received 

that license condition. 

As discussed in section I.C., ‘‘Cyber 

Security,’’ of this document, the NRC 

imposed a license condition referencing 

the approved CSP in each 10 CFR part 

50 license in the course of review and 

approval of the CSP. This proposed rule 

would remove that license condition 

once sufficient time has passed since 

the permanent removal of fuel from the 

reactor vessel. 

Because this proposed rule would 

remove certain license conditions 

without actually amending the 

associated licenses, the NRC would 

issue by letter an administrative license 

amendment to each applicable licensee 

that would remove the relevant license 

condition(s) from that licensee’s license 

and include revised license pages. 

P. Changes for Consistent Treatment of 

Holders of Combined Licenses and 

Operating Licenses 

The NRC proposes to revise 

§ 50.36(c)(6), § 50.44(b), § 50.46(a)(1)(i), 

§ 50.48(f), § 50.54(y), § 50.60(a), 

§ 50.61(b)(1), § 50.62(a), § 50.71(e)(4), 

and 10 CFR part 50, Appendix I, Section 

IV.C., to provide consistent treatment 

for COL (Part 52) and operating license 

(Part 50) holders. These changes have 

the purpose of aligning regulatory 

applicabilities for COL holders upon 

submittal of the § 52.110(a) 

certifications with regulatory 

applicabilities for operating license 

holders upon submittal of the 

§ 50.82(a)(1) certifications. In each 

section listed, the NRC would insert ‘‘or 

§ 52.110(a)’’ following each instance of 

‘‘§ 50.82(a)(1).’’ 

The NRC proposes to revise incorrect 

references to § 52.110 in § 50.49(a), 

§ 50.54(o), § 50.65(a)(1), and § 52.110(e) 

by replacing ‘‘§ 52.110(a)(1)’’ with 

‘‘§ 52.110(a).’’ The NRC proposes to 

insert a reference to § 52.110 following 

an existing reference to § 50.82 in 

§ 50.54(w)(4)(ii), § 50.54(w)(4)(iii), 

§ 50.75(e)(1)(ii)(B), § 50.75(e)(1)(v), 

§ 50.75(h)(1)(iv), and § 50.75(h)(2). The 

NRC proposes to remove the words 

‘‘under this part’’ from § 50.54(w) 

introductory text because paragraph (w) 

is also applicable to holders of 

combined licenses issued under 10 CFR 

part 52 as stated in the introductory text 

for § 50.54. Finally, the NRC proposes to 

revise an incorrect reference in 


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§ 50.59(b) by replacing ‘‘§ 50.110’’ with 

‘‘§ 52.110(a).’’ 

V. Specific Requests for Comments 

The NRC is seeking public comments 

on this proposed rule. The agency is 

particularly interested in comments and 

supporting rationale from the public on 

the following: 

• PSDAR Approval: The current 

decommissioning regulations establish 

that once a licensee permanently ceases 

operation of the nuclear power reactor, 

it cannot undertake any major 

decommissioning activities until it 

provides the public and the NRC with 

additional information. The NRC 

requires that the licensee submit this 

information in the form of a PSDAR, 

which consists of the licensee’s 

proposed decommissioning activities 

and schedule through license 

termination, a discussion of the reasons 

for concluding that the proposed 

activities will be bounded by existing 

analyses of environmental impacts, and 

a site-specific cost estimate for the 

proposed activities. The PSDAR is made 

available to the public for comment and 

is subject to NRC review (but not 

approval). Additionally, the current 

decommissioning regulations prohibit, 

at any time, the performance of any 

decommissioning activity that may 

result in significant environmental 

impacts not previously reviewed. Under 

this regulatory framework, licensees are 

not required to have an NRC-approved 

decommissioning plan; instead, 90 days 

after the NRC has received the licensee’s 

PSDAR, licensees may perform, under 

10 CFR 50.59, those major 

decommissioning activities that are 

bounded by existing environmental 

analyses. Therefore, no site-specific 

NEPA review is required and there is no 

hearing opportunity under 10 CFR part 

2 before these decommissioning 

activities begin. To perform 

decommissioning activities that are not 

bounded by existing environmental 

analyses, however, a licensee would 

have to submit a request for a license 

amendment or an exemption request, 

which would trigger a site-specific 

NEPA review and hearing opportunity 

under 10 CFR part 2. Additionally, at 

least two years before termination of the 

license, the licensee must submit an 

application for termination of license 

and a license termination plan, which 

must be approved by the NRC. The 

requirement to approve the license 

termination plan also triggers a sitespecific NEPA review and hearing 

opportunity under 10 CFR part 2. 

As part of the development of the 

proposed rule, the NRC staff evaluated 

whether the NRC should explicitly 

approve each licensee’s PSDAR before 

allowing major decommissioning 

activities to begin. The staff concluded 

that based on lessons learned and 

experience, there is currently no 

indication that requiring approval of a 

PSDAR has any substantial impact on 

the public health and safety. However, 

the NRC is gathering additional 

feedback from the public. 

As part of this rulemaking, should the 

NRC require approval of the PSDAR, a 

site-specific environmental review, and 

hearing opportunity before a licensee 

undertakes any decommissioning 

activity? Other than NRC review and 

approval of the PSDAR, are there other 

activities that could help to increase 

transparency and public trust in the 

NRC regulatory framework for 

decommissioning? Should the rule 

provide a role for the states or local 

governments in the process? What 

should that role be? What are the 

advantages or disadvantages of various 

roles? Please provide an explanation for 

your response. 

• Timeframe for Decommissioning: 

For nuclear power reactor licensees, 10 

CFR 50.82(a)(3) and 10 CFR 52.110(c) 

state that decommissioning must be 

completed within 60 years of permanent 

cessation of operations. In this proposed 

rule, the NRC is not proposing changes 

to the decommissioning timeframe 

requirements. 

What are the advantages and 

disadvantages of requiring prompt 

decontamination rather than allowing 

up to 60 years to decommission a site? 

As part of its review of a PSDAR, what 

are the advantages and disadvantages of 

NRC evaluating and making a decision 

about the timeframe for 

decommissioning on a site-specific 

basis? 

• Emergency Planning: As discussed 

in the ‘‘Technical Basis for the Graded 

Approach’’ and ‘‘Emergency 

Preparedness’’ sections of this 

document, although the spectrum of 

credible accidents and operational 

events requiring an emergency response 

is reduced at a decommissioning 

nuclear power reactor as compared to 

that for an operating nuclear power 

reactor, reliable emergency 

preparedness functions are still required 

to ensure public health and safety in the 

event of a zirconium fire scenario. 

The NRC has concluded that dry cask 

storage and spent fuel pools are both 

very safe. What are the advantages and 

disadvantages of requiring dedicated 

radiological emergency planning, 

including a 10-mile EPZ, until all spent 

nuclear fuel at a site is removed from 

the spent fuel pool and placed in dry 

cask storage? Is there additional 

information the NRC should consider in 

evaluating whether all-hazards planning 

would be as effective as dedicated 

radiological emergency planning? 

The NRC has determined that 10 

hours would be a sufficient amount of 

time for an emergency response to a 

spent fuel pool accident based on an allhazards plan. Is there additional 

information the NRC should consider in 

evaluating this issue? 

• Emergency Response Data Systems: 

Nuclear power facilities that are 

shutdown permanently or indefinitely 

are currently not required to maintain 

ERDS. These systems transmit near-realtime electronic data between the 

licensee’s onsite computer system and 

the NRC Operation Center. Licensees in 

Level 1 would maintain a capability to 

provide meteorological, radiological, 

and spent fuel pool data to the NRC 

within a reasonable timeframe following 

an event. What are the advantages and 

disadvantages of requiring nuclear 

power plant licensees to maintain those 

aspects of ERDS until all spent fuel is 

removed from the pool? 

• Cyber Security: The proposed rule 

applies cyber security requirements to 

Level 1 plants. However, a licensee in 

Level 2 would not be required to 

maintain a cyber security plan because 

the NRC has determined that there is 

little chance that the spent fuel in the 

SFP could heat up to clad ignition 

temperature within 10 hours. What are 

the advantages and disadvantages of 

extending cyber security requirements 

to shutdown nuclear power plants until 

all spent fuel is transferred to dry cask 

storage? 

• Insurance: The proposed rule 

would allow nuclear power reactor 

licensees in decommissioning to reduce 

the offsite liability and onsite property 

insurance amounts that they are 

required to maintain once a plant enters 

Level 2. The transition to Level 2 

financial protection amounts would be 

optional for licensees and they would 

have to submit an analysis that 

demonstrates a reduced risk of a 

zirconium fuel cladding fire in the SFP. 

What are the advantages and 

disadvantages of requiring the existing 

level of insurance to be maintained until 

all spent fuel is in dry cask storage 

(Level 3)? 

• Financial Assurance: Pursuant to 

§ 50.75, ‘‘Reporting and recordkeeping 

for decommissioning planning,’’ 

specifically paragraph (b)(1), nuclear 

power reactor licensees and applicants 

must certify that reasonable assurance 

for radiological decommissioning 

funding has been (for licensees) or will 

be (for applicants) provided in an 

amount that may be more, but not less, 


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than the generic amount provided by 

the Commission’s regulations (i.e., the 

table of minimum amounts under 

§ 50.75(c)). Alternatively, under 

§ 50.75(b)(4), the certified amount of 

funding may be based on a site-specific 

cost estimate for decommissioning the 

facility. 

The current table of minimum 

amounts (also referred to as the 

minimum decommissioning formula) 

has not been updated for over 30 years. 

The NRC is considering updates to the 

generic decommissioning funding 

formula to make it more reflective of 

current cost considerations. 

What are the advantages and 

disadvantages of updating the formula 

to reflect recent data and to cover all 

estimated radiological decommissioning 

costs rather than the bulk of the costs? 

• Site-Specific Cost Analysis: 

Currently, licensees can use either the 

generic amount under 10 CFR 50.75(c) 

or a site-specific cost estimate under 10 

CFR 50.75(b)(4) to determine the 

certified amount of radiological 

decommissioning funding. As provided 

in 10 CFR 50.82(a)(8)(ii) and 10 CFR 

52.110(h)(2), a licensee may withdraw 

funds from the decommissioning trust 

fund up to a cumulative total of 3 

percent of the generic amount 

calculated under 10 CFR 50.75(c) for 

decommissioning planning purposes at 

any time without prior notification to 

the NRC. After submittal of the 

certifications of permanent shutdown 

and fuel removal required under 10 CFR 

50.82(a)(1) and 10 CFR 52.110(a) and 

commencing 90 days after the NRC has 

received the PSDAR, the licensee may 

use up to an additional 20 percent of the 

decommissioning funds prescribed in 

10 CFR 50.75(c) for decommissioning 

purposes. The licensee is prohibited 

from using the remaining 77 percent of 

the generic decommissioning funds 

until a site-specific decommissioning 

cost estimate is submitted to the NRC. 

Requirements in 10 CFR 50.82(a)(8)(iii) 

and 10 CFR 52.110(h)(3) establish that a 

licensee shall provide a site-specific 

decommissioning cost estimate within 2 

years following permanent cessation of 

operations. If the estimate of costs 

provided with the PSDAR is a sitespecific cost estimate, this requirement 

can be satisfied with the PSDAR 

submittal. 

What are the advantages and 

disadvantages of requiring a full site 

investigation and characterization at the 

time of shutdown? What are the 

advantages and disadvantages of 

eliminating the formula and requiring a 

site-specific cost estimate during 

operations? 

• Decommissioning Trust Fund: 

Under the NRC’s existing regulations 

and this proposed rule, the amounts set 

aside for radiological decommissioning 

should not be used for the maintenance 

and storage of spent fuel in the spent 

fuel pool, or for the design or 

construction of spent fuel dry storage 

facilities, or for other activities not 

directly related to the long-term storage, 

radiological decontamination or 

dismantlement of the facility, or 

decontamination of the site. 

Should the NRC’s regulations allow 

decommissioning trust fund assets to be 

used for spent fuel management if (1) 

there is a projected surplus in the fund 

based on a comparison to the expected 

costs identified in a site-specific cost 

estimate and (2) the assets are returned 

to the fund within an established period 

of time? What are the advantages and 

disadvantages of allowing 

decommissioning trust fund assets to be 

used for those purposes? What are the 

advantages and disadvantages of 

allowing decommissioning trust fund 

assets to be used for non-radiological 

site restoration prior to the completion 

of radiological decommissioning? 

• Timing of Decommissioning 

Funding Assurance Reporting: This 

proposed rule would change the timing 

of the decommissioning funding 

assurance reporting requirements in 

§ 50.75(f)(1) to coordinate them with the 

ISFSI decommissioning reporting 

requirements in § 72.30. Under this 

proposed rule, operating reactors would 

be permitted to submit 

decommissioning funding status reports 

triennially instead of biennially. 

What are the advantages and 

disadvantages to extending the reporting 

frequency from two years to three years? 

Does this change affect the risk of 

insufficient decommissioning funding? 

Please provide an explanation for your 

response. 

• Backfit Rule: For nuclear power 

reactor licensees, the NRC’s backfitting 

provisions are located in § 50.109, 

‘‘Backfitting,’’ and the issue finality 

provisions are in 10CFRpart52 (the 

‘‘Backfit Rule’’). The language of the 

Backfit Rule clearly applies to a licensee 

designing, constructing, or operating a 

nuclear power facility. For example, 

§ 50.109(a)(1) defines ‘‘backfitting’’ to 

mean changes to, among other things, 

the procedures or organization required 

to design, construct, or operate a 

facility. 

This proposed rule states that the 

Backfit Rule applies to 

decommissioning nuclear power plants. 

What are the advantages and 

disadvantages of applying the Backfit 

Rule to decommissioning nuclear power 

plants? 

• Exemptions: As stated in this 

proposed rule, one of the goals of 

amending these regulations is to reduce 

the need for regulatory exemptions. 10 

CFR 50.12 states that the Commission 

may grant exemptions from the 

requirements of the regulations in 10 

CFR part 50 if the request will not 

present an undue risk to the public 

health and safety, and is consistent with 

the common defense and security. What 

are the advantages and disadvantages of 

the current 10 CFR 50.12 approach to 

decommissioning-related exemptions? 

What standard should the NRC apply in 

determining whether to grant 

exemptions from the new or amended 

regulations? What are the advantages 

and disadvantages of providing an 

opportunity for the public to weigh in 

on such exemption requests? Are there 

other process changes the NRC should 

consider in determining whether to 

grant exemptions from the new or 

amended regulations? 

• Applicability: Section III of this 

document provides a discussion of the 

applicability of this proposed rule. 

Specifically, there is a discussion for the 

applicability to NRC licensees during 

operations and to ISFSI-Only and 

Standalone ISFSI/Decommissioned 

Reactor Sites. Permanently shutdown 

nuclear power plants will be at different 

stages of decommissioning when the 

new decommissioning regulations 

become effective and will have 

previously received varying regulatory 

exemptions. 

Can you foresee any implementation 

issues with the proposed rule as it is 

currently written? For any new or 

amended requirement included in this 

proposed rule, how should the 

requirement apply to sites currently in 

different stages of decommissioning? 

• Insurance for Specific License 

ISFSI: A 10 CFR part 50 or 10 CFR part 

52 nuclear power reactor licensee with 

a 10 CFR part 72 general license ISFSI 

at the reactor site is subject to the 

financial protection requirements under 

10 CFR part 140, whereas a specific 

license ISFSI under 10 CFR part 72 is 

not. In SECY–04–0176, ‘‘Exemption 

Requests to Reduce Liability Insurance 

Coverage for Decommissioning Reactors 

after Transfer of all Spent Fuel from a 

Spent Fuel Pool to Dry Cask Storage,’’ 

dated September 29, 2004 (ADAMS 

Accession No. ML040850518), the NRC 

staff noted that general license ISFSIs 

subject to the requirements under 10 

CFR part 72 were also subject to the 

requirements of a 10 CFR part 50 license 

and by virtue of this license, they are 

required to maintain some level of 


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liability insurance under section 170, 

‘‘Indemnification and Limitation of 

Liability,’’ of the AEA (known as the 

Price-Anderson Act) and the NRC’s 

implementing regulations at 10 CFR part 

140. Further, the NRC staff 

acknowledged that there was little 

technical difference between a general 

license ISFSI and a specific license 

ISFSI. 

The NRC recognizes that as a reactor 

site is decommissioned, eventually all 

that remains of the 10 CFR part 50 or 

part 52 licensed site is a general license 

ISFSI under 10 CFR part 72, which is 

essentially the same as a specific license 

ISFSI under 10 CFR part 72. 

Considering that 10 CFR part 72 specific 

license ISFSIs have no financial 

protection requirements, should the 

NRC address the disparity between 

specific license and general license 

ISFSIs as a part of this rulemaking? 

Please provide an explanation for your 

response. 

• Recordkeeping Requirements for 

Facilities Licensed under 10 CFR part 

52: The current appendices in 10 CFR 

part 52 contain section X, ‘‘Records and 

Reporting,’’ for all of the certified 

designs codified in 10 CFR part 52. 

Section X requires, in part, that all 

departures from the certified design be 

recorded and those records kept 

throughout the term of the license. 

However, as part of this rulemaking, the 

NRC is proposing to change the record 

retention requirements for nuclear 

power reactors in the decommissioning 

process such that they no longer need to 

retain certain records associated with 

SSCs that are no longer in service or 

necessary to keep the plant in a safe 

condition. The NRC is considering 

making conforming changes to section X 

of the applicable appendices to 10 CFR 

part 52 to allow this change to apply to 

records of departures from the certified 

design as well as the associated SSCs. 

Given the already existing change 

control procedures in the appendices to 

10 CFR part 52, as well as the significant 

changes in recordkeeping technology 

since the NRC’s record retention 

requirements were introduced (i.e., 

digital media instead of paper copies), 

should additional changes be made to 

the 10 CFR part 52 appendices as a part 

of this rulemaking, and would such 

changes be beneficial to 10 CFR part 52 

licensees or add efficiency to the 

decommissioning process for these 

facilities? Please provide an explanation 

for your response. 

• Identical Requirements under 

§ 50.82 and § 52.110: As part of this 

rulemaking, the NRC proposes to revise 

§ 52.110 to make the same changes 

proposed in § 50.82 for the reasons 

previously discussed and for 

consistency. The NRC also proposes to 

add paragraphs (h)(5) through (h)(7) to 

§ 52.110 with site-specific 

decommissioning cost estimate 

reporting requirements that are identical 

to the requirements in § 50.82(a)(8)(v) 

through (vii). Given that the 

decommissioning financial assurance 

requirements in § 52.110 are identical to 

the requirements in § 50.82, should the 

NRC consider removing the specific 

requirements from § 52.110(f)–(h) and 

instead add a reference in § 52.110 to 

the identical regulations in 

§ 50.82(a)(6)–(8)? Are there any other 

provisions in § 52.110 that the NRC 

should consider removing and replacing 

with a reference to an identical 

requirement in § 50.82 (e.g., the 

decommissioning requirements under 

§ 52.110(c)–(e))? Please provide an 

explanation for your response. 

• Removal of License Conditions and 

Withdrawal of Orders: This rulemaking 

seeks to improve regulatory efficiency 

by removing license conditions and 

withdrawing an order for which 

substantively identical requirements 

have been imposed by rulemaking. This 

would avoid the future administrative 

expenditures by licensees and the NRC 

to accomplish the removal of these 

requirements on a license-specific basis 

through a generic regulatory action 

either upon the effective date of the 

final rule or when conditions permit the 

removal during the decommissioning 

process. The NRC has identified certain 

orders that were issued following the 

terrorist events of September 11, 2001, 

license conditions regarding these 

orders, and license conditions regarding 

cyber security implementation as having 

substantively identical requirements 

made generically applicable through 

rulemaking. Because these licensespecific requirements are duplicative 

with other generic requirements, the 

NRC concludes there would be no 

reduction in safety. Please provide any 

comments you may have on rescinding 

Order EA–06–137 and the related 

license conditions. As part of this 

rulemaking, are there other licensespecific requirements in license 

conditions or orders that have 

substantively identical generic 

requirements that should be addressed 

in this rulemaking? Please provide an 

explanation for your response. 

• Spent Fuel Management Planning: 

Section IV.K of this document discusses 

spent fuel management planning in the 

§ 50.54(bb) regulation. The § 50.54(bb) 

current rule language requires NRC 

preliminary approval and final review, 

as part of any proceeding for continued 

licensing under part 50 or part 72, of the 

IFMP. The discussion in Section IV.K 

points out that the proceedings for 

continued licensing under part 50 or 

part 72 no longer exist. Therefore, the 

proposed rule includes language 

intended to clarify the current IFMP 

approval process by requiring submittal 

of the IFMP for NRC review and 

approval by license amendment. What, 

if any, challenges do you foresee with 

implementing this part of the proposed 

rule? Please provide an explanation for 

your response. 

The § 50.54(bb) current rule language 

requires licensees to notify the NRC of 

any significant changes to the IFMP. As 

discussed in section IV.K, the NRC 

proposes to revise this requirement to 

require licensees to submit to the NRC 

any changes to the IFMP as an 

application for an amendment to its 

license. The NRC is also considering 

replacing the notification requirement 

with a change control provision to 

specify what changes a licensee can 

make to the IFMP without NRC 

approval. Examples of change control 

provisions in the current NRC 

regulations include § 50.54(a) for quality 

assurance programs and § 50.54(q) for 

emergency plans. If the NRC includes a 

similar change control provision in 

§ 50.54(bb), what should the safety and 

environmental criteria be for 

determining whether a licensee could 

make a change to its IFMP without 

seeking NRC approval? For example, the 

NRC could permit changes that are not 

considered to be reductions in the 

commitments, including (1) changes to 

the planned actions for managing spent 

fuel that result in an addition of one or 

more SSCs that the licensee relies on for 

irradiated fuel management, and (2) 

changes to the projected cost or funding 

for managing irradiated fuel that is 

already included in the report required 

by 10 CFR 50.82(a)(8)(vii) or 10 CFR 

52.110(h)(7). Should the NRC also 

include recordkeeping and reporting 

provisions for a licensee to retain a 

record of each change to the IFMP made 

without prior NRC approval and submit 

a report to the NRC of those changes? If 

so, what should be the timeframe for the 

records to be retained and the timeframe 

for reporting to the NRC after the change 

is made, taking into consideration the 

estimated frequency of performing IFMP 

changes? Please provide an explanation 

for your response. 

VI. Section-by-Section Analysis 

The following paragraphs describe the 

specific changes proposed by this 

rulemaking. 


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Appendix G to 10 CFR Part 20, 

Requirements for Transfers of Low-Level 

Radioactive Waste Intended for Disposal 

at Licensed Land Disposal Facilities and 

Manifests 

In section III, paragraph E.1., this 

proposed rule would remove the word 

‘‘or’’ and add in its place the word ‘‘of’’ 

and it would also remove the phrase ‘‘20 

days’’, and add in its place the phrase 

‘‘45 days’’. 

Section 26.3 Scope 

In § 26.3, this proposed rule would 

revise paragraph (a) by subdividing it 

into two subparagraphs, (a)(1) and (2), to 

include the NRC’s docketing of a license 

holder’s certifications required under 

§§ 50.82 and 52.110(a). 

Section 26.825 Criminal Penalties 

In § 26.825, this proposed rule would 

revise paragraph (b) to remove the 

number ‘‘26.3’’ from the list of 

regulations in 10 CFR part 26 that are 

excluded from § 26.825(a). 

Section 50.1 Basis, Purpose, and 

Procedures Applicable 

In § 50.1, this proposed rule would 

add language clarifying that the 

regulations in 10 CFR part 50 provide 

for the licensing of production and 

utilization facilities through the 

termination of the associated 10 CFR 

part 50 licenses. 

Section 50.2 Definitions 

In § 50.2, this proposed rule would 

retain the existing definition of certified 

fuel handler and add an alternative 

definition for the purposes explained 

elsewhere in this document. This 

proposed rule also would add a 

definition for a non-power production or 

utilization facility. 

Section 50.36 Technical Specifications 

In § 50.36, this proposed rule would 

revise paragraph (c)(6) to insert ‘‘or 

§ 52.110(a)’’ following ‘‘§ 50.82(a)(1).’’ 

Section 50.38 Ineligibility of Certain 

Applicants 

This proposed rule would revise 

§ 50.38 by including the current text as 

paragraph (a) and by adding paragraph 

(b) to state that the prohibition in 

paragraph (a) of this section does not 

apply to a person, corporation, or other 

entity seeking a license for a facility that 

is not a production or utilization 

facility. 

Section 50.44 Combustible Gas Control 

for Nuclear Power Reactors 

In § 50.44, this proposed rule would 

revise paragraph (b) to insert ‘‘or 

§ 52.110(a)’’ following ‘‘§ 50.82(a)(1).’’ 

Section 50.46 Acceptance Criteria for 

Emergency Core Cooling Systems for 

Light-Water Nuclear Power Plants 

In § 50.46, this proposed rule would 

revise paragraph (a)(1)(i) to insert ‘‘or 

§ 52.110(a)’’ following ‘‘§ 50.82(a)(1).’’ 

Section 50.47 Emergency Plans 

This proposed rule would make 

conforming changes to paragraph (b) in 

§ 50.47 and would add paragraph (f) 

denoting when the planning standards 

for offsite emergency plans in paragraph 

(b) of this section do not apply. 

Section 50.48 Fire Protection 

In § 50.48, this proposed rule would 

revise paragraph (f) to insert ‘‘or 

§ 52.110(a)’’ following ‘‘§ 50.82(a)(1).’’ 

Section 50.49 Environmental 

Qualification of Electric Equipment 

Important to Safety for Nuclear Power 

Plants 

In § 50.49, this proposed rule would 

revise paragraph (a) by replacing 

‘‘§ 52.110(a)(1)’’ with ‘‘§ 52.110(a).’’ 

Section 50.51 Continuation of License 

In § 50.51, this proposed rule would 

remove the phrase, ‘‘to authorize 

ownership and possession of the 

production or utilization facility,’’ for 

reasons discussed elsewhere in this 

document. 

Section 50.54 Conditions of Licenses 

In § 50.54, this proposed rule would 

revise footnote 2 to the table in 

paragraph (m)(2)(i) to indicate when a 

Shift Technical Advisor is not required. 

Paragraph (o) would be revised by 

replacing ‘‘52.110(a)(1)’’ with 

‘‘52.110(a).’’ The NRC also would revise 

§ 50.54(p) to include the definitions for 

change and decrease in safeguards 

effectiveness for use in paragraph (p), 

would revise and redesignate existing 

paragraphs (p)(1) and (2) as (p)(2) and 

(3), would redesignate paragraphs (p)(3) 

and (4) as paragraphs (p)(5) and (6), and 

would add new paragraphs (p)(1) and 

(4). A portion of the existing text in 

paragraphs (p)(1) and (2) would be used 

to create new paragraph (p)(4). 

This proposed rule would revise: 

Paragraph (q)(1) to clarify that the 

definitions are for use in paragraph (q), 

paragraph (q)(1)(iii) to remove the 

reference to appendix E to 10 CFR part 

50, paragraph (q)(2) to add clarification 

to the applicability, paragraph (q)(3) to 

add applicable emergency planning 

requirements, paragraphs (q)(4) and (5) 

to remove the phrase ‘‘after February 21, 

2012,’’ and add new paragraphs (q)(7) 

and (8) to add the requirements for 

licensees after the NRC dockets their 

certifications required for 

decommissioning under § 50.82(a)(1) or 

§ 52.110(a). 

Paragraph (s)(2)(ii) would be revised 

by removing the phrase ‘‘after April 1, 

1981,’’ and paragraph (s)(3) would be 

revised by adding clarification at the 

beginning of the sentence that if the 

standards apply to offsite radiological 

response plans then the NRC will base 

its findings on a review of FEMA 

findings and determinations. 

Paragraph (t) would be revised by 

replacing ‘‘.’’ with ‘‘or’’ in the second 

sentence of paragraph (t)(1)(ii), adding 

new subparagraph (t)(1)(iii) to clarify 

the interval at which the licensee’s 

emergency preparedness plan must be 

reviewed after the NRC has docketed the 

certifications required for 

decommissioning, and by adding new 

paragraph (t)(3) to state that the review 

requirement is no longer required once 

all fuel is in dry cask storage. 

Paragraph (w) would be revised by 

removing the words ‘‘under this part’’ 

from the introductory text, adding a 

reference to § 52.110 in paragraphs 

(w)(4)(ii) and (w)(4)(iii), and adding new 

paragraphs (w)(5) and (6) to include the 

financial protection requirements for 

production or utilization facilities 

undergoing decommissioning. 

Paragraph (y) would be revised to 

insert ‘‘or § 52.110(a)’’ following 

‘‘§ 50.82(a)(1).’’ 

Paragraph (bb) would be revised by 

restructuring the paragraph and revising 

the requirements of an irradiated fuel 

management plan. 

Section 50.59 Changes, Tests, and 

Experiments 

In § 50.59, this proposed rule would 

revise paragraph (b) to correct a 

reference to § 52.110(a). It would also 

revise paragraph (d)(3) to include the 

exception for when the records of 

changes requirement in paragraph (d)(3) 

applies. 

Section 50.60 Acceptance Criteria for 

Fracture Prevention Measures for 

Lightwater Nuclear Power Reactors for 

Normal Operation 

In § 50.60, this proposed rule would 

revise paragraph (a) to insert ‘‘or 

§ 52.110(a)’’ following ‘‘§ 50.82(a)(1).’’ 

Section 50.61 Fracture Toughness 

Requirements for Protection Against 

Pressurized Thermal Shock Events 

In § 50.61, this proposed rule would 

revise paragraph (b)(1) to insert ‘‘or 

§ 52.110(a)’’ following ‘‘§ 50.82(a)(1).’’ 


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Section 50.62 Requirements for 

Reduction of Risk From Anticipated 

Transients Without Scram (ATWS) 

Events for Light-Water-Cooled Nuclear 

Power Plants 

In § 50.62, this proposed rule would 

revise paragraph (a) to insert ‘‘or 

§ 52.110(a)’’ following ‘‘§ 50.82(a)(1).’’ 

Section 50.65 Requirements for 

Monitoring the Effectiveness of 

Maintenance at Nuclear Power Plants 

In § 50.65, this proposed rule would 

revise paragraph (a)(1) by replacing 

‘‘§ 52.110(a)(1)’’ with ‘‘§ 52.110(a).’’ 

Section 50.71 Maintenance of Records, 

Making of Reports 

In § 50.71, this proposed rule would 

revise paragraph (c) by including the 

current text as paragraph (c)(1) and it 

would add new paragraph (c)(2) to add 

records requirements for licensees for 

whom the NRC has docketed the 

certifications required for 

decommissioning. 

Paragraph (e)(4) would be revised to 

insert ‘‘or § 52.110(a)’’ following 

‘‘§ 50.82(a)(1).’’ 

Section 50.75 Reporting and 

Recordkeeping for Decommissioning 

Planning 

In § 50.75, this proposed rule would 

revise paragraph (a) by clarifying the 

availability of funds to decommission a 

facility as defined in § 50.2. 

Paragraph (b)(1) would be revised by 

replacing ‘‘financial’’ with ‘‘reasonable’’ 

assurance and other conforming 

changes; paragraph (b)(3) would be 

revised by removing the phrase ‘‘as 

acceptable to the NRC’’ from the end of 

the paragraph; paragraph (b)(4) would 

be revised to include a site-specific 

decommissioning cost estimate and the 

second sentence of current paragraph 

(b)(4) would be moved to become a new 

paragraph (b)(5). 

Paragraph (e)(1) would be revised to 

include the term ‘‘reasonable assurance 

of funds to decommission,’’ and 

paragraphs (e)(1)(i) and (ii) would be 

revised to include the description of 

‘‘decommissioning cost’’ before the 

word estimate throughout each 

paragraph. Paragraphs (e)(1)(ii)(B) and 

(e)(1)(v) would be revised to add a 

reference to § 52.110. 

Paragraph (f) would be amended by 

revising (f)(1) to include the 

requirement for a report to include 

information regarding any potential 

decommissioning shortfall, it would be 

further amended by removing paragraph 

(f)(2) and redesignating (f)(3) through (5) 

as (f)(2) through (4) with minor 

revisions. 

Paragraphs (h)(1)(iii) and (iv) and 

(h)(2) would be revised to remove the 

reference to three office directors within 

the NRC for the submission of written 

notice of the intention to make a 

payment or disbursement of funds and 

replace it with the Document Control 

Desk. Paragraphs (h)(1)(iv) and (h)(2) 

would be revised to add a reference to 

§ 52.110. 

Section 50.82 Termination of License 

In § 50.82, this proposed rule would 

revise paragraph (a)(2) to provide 

clarification as to when a licensed 

nuclear power reactor is no longer 

considered to be a utilization facility. It 

also would revise paragraph (a)(4)(i) to 

clarify that licensees provide the basis 

for whether the environmental impacts 

from site-specific decommissioning 

activities are bounded by federally 

issued environmental review 

documents. The phrase ‘‘including the 

projected cost of managing irradiated 

fuel’’ would be removed at the end of 

the last sentence. Paragraph (a)(4)(ii) 

would be revised to include the 

requirement for the NRC to include the 

irradiated fuel management plan in the 

notice of the receipt of the PSDAR in the 

Federal Register and to allow the public 

to comment. 

Paragraph (a)(6)(ii) would be revised 

to provide clarification. 

Paragraph (a)(8)(i)(A) would be 

revised to remove the words ‘‘legitimate 

decommissioning’’ and to replace the 

word ‘‘decommissioning’’ with 

‘‘decommission.’’ Paragraph (a)(8)(ii) 

would be revised to clarify paragraph (c) 

to § 50.75 is where the specified amount 

is located. 

Paragraph (a)(8)(v) would be revised 

to spell out the acronym DCE, 

decommissioning cost estimate, and to 

include the ability for the licensee to 

combine the reporting requirements of 

10 CFR part 72, § 50.82(a)(8)(v), and 

§ 50.82(a)(8)(vii). 

Paragraph (a)(8)(vii) would be revised 

to spell out the acronym DCE, 

decommissioning cost estimate. 

Paragraph (a)(9) would be revised to 

clarify that all nuclear power reactors 

that have loaded fuel into the reactor 

must submit an application for 

termination of a license and paragraph 

(a)(9)(ii)(F) would be revised to include 

the requirement to identify funding 

sources for license termination, spent 

fuel management, and ISFSI 

decommissioning. 

The introductory text of paragraph (b) 

would be revised to replace the term 

‘‘non-power reactor licensees’’ with 

‘‘non-power production or utilization 

facilities and fuel reprocessing plants.’’ 

Paragraph (b)(6) would be 

redesignated as (b)(8) and new 

paragraphs (b)(6) and (7) would be 

added to include the criteria for when 

a non-power production or utilization 

facility or fuel reprocessing plant 

licensed under 10 CFR part 50 is no 

longer considered a production or 

utilization facility. 

Section 50.109 Backfitting 

This proposed rule would revise 

§ 50.109 in its entirety to provide 

backfitting provisions for reactors both 

before and during decommissioning and 

to require that a documented evaluation 

for a modification necessary to bring a 

facility into compliance with a license 

or the rules or orders of the 

Commission, or into conformance with 

the licensee’s written commitments, 

must include a consideration of the 

costs of imposing the modification. 

Section 50.155 Mitigation of BeyondDesign-Basis Events 

This proposed rule would add new 

paragraphs (h)(6), (h)(7) and (h)(8) that 

would deem removed certain license 

conditions and withdraw certain orders 

made redundant by regulations 

imposing substantively identical 

requirements. 

Section 50.200 Power Reactor 

Decommissioning Emergency Plans 

This proposed rule would add new 

§ 50.200 that would contain alternate 

emergency preparedness requirements 

for nuclear power reactor facilities in 

decommissioning. 

Appendix E to 10 CFR Part 50, 

Emergency Planning and Preparedness 

for Production and Utilization Facilities 

This proposed rule would revise 

section I. Introduction of appendix E to 

10 CFR part 50 by removing paragraph 

6. 

Section IV. Content of Emergency 

Plans of appendix E to 10 CFR part 50 

would be revised by removing from 

paragraph 4 the phrases ‘‘of the later of 

the date’’ and ‘‘or December 23, 2011,’’ 

from the first sentence; new paragraph 

8 would be added to inform licensees 

that the requirements of paragraphs 4, 5, 

and 6 of this section are no longer 

required once the NRC dockets the 

licensee’s certifications required for 

decommissioning; paragraphs A.7., A.9., 

B.1., C.2., E.8.c., and I. would all be 

revised by removing the ‘‘by date’’ 

phrases; paragraph D.4. would be 

removed; the last sentence of paragraph 

E.8.d. would be removed; in paragraph 

F.2.d., the end of the 3rd sentence 

beginning with the word ‘‘and’’ would 

be removed; paragraph F.2.j(v) would be 


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removed and reserved; and new 

paragraph F.2.k would be added to 

require licensees to follow the biennial 

exercise requirements in paragraph F.2 

of appendix E to 10 CFR part 50 after 

the NRC dockets the certifications 

required for decommissioning. 

This proposed rule would revise 

section VI. Emergency Response Data 

System of appendix E to 10 CFR part 50 

by removing the date in paragraph 4.a. 

and the date in paragraph 4.d., also in 

paragraph 4.d. it would remove the 

phrase ‘‘, whichever comes later’’ from 

the first sentence. 

Appendix I to 10 CFR Part 50, 

Numerical Guides for Design Objectives 

and Limiting Conditions for Operation 

To Meet the Criterion ‘‘As Low as Is 

Reasonably Achievable’’ for Radioactive 

Material in Light-Water-Cooled Nuclear 

Power Reactor Effluents 

This proposed rule would revise 

section IV.C of appendix I to 10 CFR 

part 50 by inserting ‘‘or § 52.110(a)’’ 

following ‘‘§ 50.82(a)(1).’’ 

Section 51.53 Postconstruction 

Environmental Reports 

This proposed rule would revise the 

first sentence in paragraph (d) to 

include applicants for a license 

amendment approving an irradiated fuel 

management plan under § 50.54(bb). 

The proposed rule would also add 

references to § 50.82 and § 52.110 after 

‘‘license termination plan.’’ 

Section 51.95 Postconstruction 

Environmental Impact Statements 

This proposed rule would revise the 

first sentence in paragraph (d) to refer to 

an amendment approving an irradiated 

fuel management plan under 

§ 50.54(bb), the license termination plan 

under § 50.82 or § 52.110, or a 

decommissioning plan under § 50.82. 

Section 52.0 Scope 

In § 52.0, this proposed rule would 

add language clarifying that the 

regulations in 10 CFR part 52 remain 

effective through the termination of the 

associated 10 CFR part 52 licenses. 

Section 52.63 Finality of Standard 

Design Certifications 

This proposed rule would revise 

paragraph (b)(2) by removing the last 

sentence and by adding new paragraphs 

(b)(2)(i) and (ii) regarding the 

recordkeeping and retention 

requirements for departures from the 

design of a facility. 

Section 52.109 Continuation of 

Combined License 

In § 52.109, this proposed rule would 

remove the phrase, ‘‘to authorize 

ownership and possession of the 

production or utilization facility,’’ for 

reasons discussed elsewhere in this 

document. 

Section 52.110 Termination of License 

This proposed rule would revise 

paragraph (b) as paragraph (b)(1) and 

would add paragraph (b)(2) to provide 

clarification as to when a facility 

licensed under 10 CFR part 52 is no 

longer considered to be a production or 

utilization facility. Paragraph (d)(1) 

would be revised to clarify that 

licensees provide the basis for whether 

the environmental impacts from sitespecific decommissioning activities are 

bounded by federally issued 

environmental review documents, and 

the phrase ‘‘site-specific 

decommissioning cost estimate’’ would 

be added at the end of the last sentence. 

Paragraph (d)(2) would be revised to 

include the requirement for the NRC to 

include the irradiated fuel management 

plan in the notice of the receipt of the 

PSDAR in the Federal Register and to 

allow the public to comment. 

Paragraph (e) would be revised by 

replacing ‘‘§ 52.110(a)(1)’’ with 

‘‘§ 52.110(a).’’ 

Paragraph (f)(2) would be revised to 

clarify the decommissioning activities 

licensees shall not perform. Paragraph 

(h)(1)(i) would be revised to remove the 

phrase ‘‘legitimate decommissioning,’’ 

paragraph (h)(2) would be revised to 

include a more specific regulatory 

reference, and paragraphs (h)(5) through 

(8) would be added with requirements 

for the submission of financial status 

reports. Paragraph (i) would be revised 

to clarify that all nuclear power reactor 

licensees that have loaded fuel into the 

reactor must submit an application for 

termination of a license. Paragraph 

(i)(2)(vi) would be revised to include 

identification of sources of funds for 

license termination, spent fuel 

management, and ISFSI 

decommissioning, as applicable. 

Section 72.13 Applicability 

This proposed rule would revise 

§ 72.13 by adding a new paragraph (e) 

to incorporate conforming changes to 

match technical changes elsewhere in 

the rule. 

Section 72.30 Financial Assurance 

and Recordkeeping for 

Decommissioning 

This proposed rule would revise 

§ 72.30 by removing the second 

sentence in paragraph (c). The proposed 

revisions would create new paragraphs 

(b)(1) through(3) and redesignate the 

existing paragraphs (b)(1) through(6) as 

new (b)(3)(i) through(vi). 

Section 72.32 Emergency Plan 

In § 72.32, this proposed rule would 

clarify that the requirement for having 

an emergency plan applies when the 

proposed ISFSI would not be located on 

the site or within the exclusion area of 

a nuclear power reactor licensed under 

10 CFR parts 50 or 52. The proposed 

revisions would consolidate the current 

language and remove redundancies by 

using standardized language consistent 

with other proposed rule provisions. 

Section 72.44 License Conditions 

This proposed rule would revise 

§ 72.44 by adding a sentence to 

paragraph (f) to indicate that licensees 

need not comply with the requirements 

of paragraph (f) once all spent fuel has 

been removed from the site. 

Section 72.62 Backfitting 

This proposed rule would revise 

paragraph (a)(2) to clarify that the 

backfitting provisions under this part 

continue to apply during 

decommissioning. 

Section 72.72 Material Balance, 

Inventory, and Records Requirements 

for Stored Material 

This proposed rule would revise 

paragraph (d) by breaking it into three 

paragraphs. The last sentence of the 

current paragraph (d) would become 

paragraph (d)(3). New text is proposed 

for paragraph (d)(2) and minor revisions 

are proposed for paragraph (d)(1). 

Section 72.212 Conditions of General 

License Issued Under § 72.210 

This proposed rule would revise 

§ 72.212 by adding new paragraphs 

(b)(9)(vii)(A) and (B) regarding the 

protection of spent fuel after the NRC 

dockets the decommissioning 

certifications. Paragraph (b)(9)(vii)(A) 

would allow a licensee to voluntarily 

provide for physical protection of the 

spent fuel under Subpart H of this part 

and § 73.51 of this chapter. Paragraph 

(b)(9)(vii)(B) would require a licensee 

who elects to provide physical 

protection under Subpart H of this part 

and § 73.51 of this chapter to notify the 

NRC of this decision using the 

provisions of § 50.54(p)(2). 

Section 72.218 Termination of License 

This proposed rule would revise 

§ 72.218 by revising paragraphs (a) and 

(b) and removing paragraph (c). 

Paragraph (a) is revised to reference the 

decommissioning requirements in 


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§ 50.82 or § 52.110 that apply to the 

general license and paragraph (b) is 

revised to state when the general license 

is considered terminated. 

Section 73.51 Requirements for the 

Physical Protection of Stored Spent 

Nuclear Fuel and High-Level 

Radioactive Waste 

This proposed rule would revise 

§ 73.51 by removing text from paragraph 

(a), (a)(1), (a)(2), and adding new 

paragraph (a)(3). Paragraph (a)(3) would 

be added to require notification to the 

NRC under the provisions of 

§ 72.212(b)(9)(vii) of this chapter by a 

licensee who elects to provide physical 

protection under Subpart H of 10 CFR 

part 72. 

Section 73.54 Protection of Digital 

Computer and Communications 

Systems and Networks 

This proposed rule would revise 

§ 73.54 by removing the introductory 

text of the section and revising the 

introductory text of paragraphs (a), (b), 

and (c), and adding new paragraphs (i), 

and (j). The introductory text of 

paragraph (a) would be revised to 

capture that the rule applies during 

operation and decommissioning. Minor 

edits would be made to paragraphs (b) 

and (c). Paragraph (i) states that the 

requirements of § 73.54 no longer apply 

once the criteria in (i)(1) and (2) are met. 

Paragraph (j) provides for the removal of 

the cyber security license condition. 

Section 73.55 Requirements for 

Physical Protection of Licensed 

Activities in Nuclear Power Reactors 

Against Radiological Sabotage 

This proposed rule would revise 

§ 73.55 by clarifying in paragraph (b)(3) 

that a licensee’s physical protection 

program must be designed to prevent 

significant core damage until the NRC 

dockets the certifications required for 

decommissioning. 

New paragraphs (b)(9)(ii)(B)(1), (2), 

(2)(i), and (2)(ii) would be added to 

provide additional clarification for 

licensees implementing fitness for duty 

programs. 

Paragraph (c)(6) would be revised by 

replacing the text beginning with the 

words ‘‘that describes’’ through the end 

of the sentence with the phrase, ‘‘in 

accordance with the requirements of 

§ 73.54 of this part.’’ 

Paragraph (e)(9)(v)(A) would be 

revised to provide clarification for when 

the reactor control room would not be 

considered a vital area. 

Paragraph (j)(4)(ii) would be revised 

to include a system for communication 

with certified fuel handlers if the NRC 

had docketed the certifications required 

for decommissioning. 

Paragraph (p)(1)(i) and (ii) would be 

revised to allow a certified fuel handler 

or a licensed senior operator to approve 

the suspension of security measures if 

the NRC has docketed the certifications 

required for decommissioning. 

Section 140.11 Amounts of Financial 

Protection Required for Certain Reactors 

This proposed rule would revise 

§ 140.11 by adding new paragraphs 

(a)(5), (a)(5)(i) and (ii) and by 

redesignating paragraph (b) as paragraph 

(c) and adding new paragraph (b) that 

would provide the requirements for the 

amounts of financial protection required 

for reactors in decommissioning. 

Section 140.81 Scope and Purpose 

This proposed rule would revise 

§ 140.81 by clarifying the scope of who 

is subject to the requirements in this 

section and to further clarify that this 

section no longer applies once a 

licensee meets the requirements of 

§ 140.11(a)(5)(i) and (ii). 

VII. Regulatory Flexibility Certification 

As required by the Regulatory 

Flexibility Act of 1980, 5 U.S.C. 605(b), 

the Commission certifies that this rule, 

if adopted, will not have a significant 

economic impact on a substantial 

number of small entities. This proposed 

rule affects only the licensing and 

operation of nuclear production and 

utilization facilities. The companies that 

own these plants do not fall within the 

scope of the definition of ‘‘small 

entities’’ set forth in the Regulatory 

Flexibility Act or the size standards 

established by the NRC (§ 2.810). 

VIII. Regulatory Analysis 

The NRC has prepared a draft 

regulatory analysis for this proposed 

rule. The analysis examines the costs 

and benefits of the alternatives 

considered by the NRC. The NRC 

requests public comment on the draft 

regulatory analysis. The draft regulatory 

analysis is available as indicated in the 

‘‘Availability of Documents’’ section of 

this document. Comments on the draft 

analysis may be submitted to the NRC 

as indicated under the ADDRESSES

section of this document. 

IX. Backfitting and Issue Finality 

The NRC’s backfitting provisions for 

holders of construction permits and 

operating licenses appear in § 50.109, 

‘‘Backfitting.’’ Issue finality provisions 

(analogous to the backfitting provisions 

in § 50.109) for applicants and holders 

of combined licenses are located in 

§ 52.83, ‘‘Finality of referenced NRC 

approvals; partial initial decision on site 

suitability,’’ and § 52.98, ‘‘Finality of 

combined licenses; information 

requests.’’ This section describes the 

backfitting and issue finality 

implications of the draft guidance 

documents described in section XVI, 

‘‘Availability of Guidance,’’ in this 

document and this proposed rule as 

applied to applicants and holders of 

pertinent NRC approvals. As stated in 

section III, ‘‘Discussion,’’ in this 

document, the proposed changes to 10 

CFR part 72 would not impose 

requirements on ISFSI-only licensees. 

Accordingly, the proposed rule would 

not constitute ‘‘backfitting’’ as that term 

is defined in § 72.62, ‘‘Backfitting.’’ 

A. Current and Future Applicants 

Applicants and potential applicants 

(for licenses, permits, and regulatory 

approvals such as design certifications) 

are not, with certain exceptions, the 

subject of either the 10 CFR part 50 

backfitting provisions or any issue 

finality provisions under 10 CFR part 

52. The backfitting and issue finality 

regulations include language delineating 

when those provisions begin; in general, 

they begin after the issuance of a 

license, permit, or approval (e.g., 

§ 50.109(a)(1)(iii), § 52.98(a)). 

Furthermore, neither the 10 CFR part 50 

backfitting provisions nor the issue 

finality provisions under 10 CFR part 

52—with certain exclusions discussed 

below—were intended to apply to every 

NRC action that substantially changes 

the expectations of current and future 

applicants, and applicants have no 

reasonable expectation that future 

requirements will not change (‘‘Early 

Site Permits; Standard Design 

Certifications; and Combined Licenses 

for Nuclear Power Plants; Final Rule,’’ 

54 FR 15372, at 15385–15386; April 18, 

1989). 

The exceptions to this general 

principle are applicable whenever a 

combined license applicant references a 

10 CFR part 52 license (e.g., an early site 

permit) or NRC regulatory approval 

(e.g., a design certification rule) with 

specified issue finality provisions. The 

issues that are resolved in an early site 

permit or a design certification and 

accorded issue finality do not include 

decommissioning matters that are the 

subject of this proposed rule and draft 

guidance, and the proposed rule and 

draft guidance do not contain design 

requirements. Therefore, the proposed 

rule and draft guidance would not affect 

the issue finality accorded early site 

permits and design certifications. For 

the same reasons, the issue finality 

provision applicable to combined 

license applicants (§ 52.83) would not 


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apply to a combined license applicant 

referencing either an early site permit or 

a design certification with respect to 

compliance with this rule. 

B. Existing Design Certifications 

The issues that are resolved in a 

design certification and accorded issue 

finality do not include 

decommissioning matters that are the 

subject of this proposed rule and draft 

guidance. Because the decommissioning 

matters that are the subject of this 

proposed rule and draft guidance are 

limited to nuclear power reactor 

decommissioning, they would not be 

applied to existing or future design 

certifications. 

C. Existing Licensees 

Section IV.A of this document 

describes a proposed alternative 

approach to the current requirements for 

radiological emergency preparedness at 

a nuclear power reactor. The proposed 

addition of 10 CFR 50.200 would not 

constitute backfitting or affect the issue 

finality of a COL because the proposed 

change would provide a voluntary 

alternative set of requirements. 

Backfitting is defined in § 50.109(a)(1) 

as, in relevant part, a modification of or 

addition to the systems, structures, or 

components (SSCs) or design of a 

facility, or the procedures or 

organization required to design, 

construct, or operate a facility, which 

results from a new or amended 

provision in the Commission’s 

regulations. This proposed rule would 

not require holders of operating licenses 

and COLs to use the alternative 

emergency preparedness requirements, 

so the proposed change would not result 

in a modification or addition that would 

be backfitting or affect the issue finality 

of a COL. 

Section IV.A of this document also 

describes other proposed changes 

related to emergency preparedness. The 

NRC would revise § 50.47 to add a 

paragraph (f) to explain when the 

planning standards of § 50.47(b) would 

no longer apply. Removing a 

requirement would not create a new 

requirement or amend a requirement 

because amending means the 

requirement still exists in some form. 

Without creating or amending a 

regulation, this proposed change would 

not meet the definition of ‘‘backfitting’’ 

or affect the issue finality of a COL. 

The proposed changes to § 50.54(q) 

would be made to allow a licensee using 

the emergency preparedness framework 

of 10 CFR 50.200 to also use § 50.54(q). 

The proposed changes would not 

require a licensee to use the § 50.54(q) 

emergency plan change process or result 

in a modification of or addition to SSCs 

or the design of a facility or the 

procedures or organization required to 

design, construct, or operate a facility. 

Therefore, the proposed changes would 

not meet the definition of ‘‘backfitting’’ 

or affect the issue finality of a COL. 

The proposed change to § 50.54(s)(3) 

would clarify that FEMA findings and 

determinations are only necessary when 

the NRC’s planning standards apply to 

offsite radiological emergency response 

plans. These changes to the NRC’s and 

FEMA’s review of emergency plans 

would not result in a modification of or 

addition to SSCs or the design of a 

facility or the procedures or 

organization required to design, 

construct, or operate a facility. 

Therefore, the proposed changes would 

not meet the definition of ‘‘backfitting’’ 

or affect the issue finality of a COL. 

The NRC is proposing to amend 

§ 50.54(t) so licensees in 

decommissioning would be able to 

conduct emergency preparedness 

program element reviews at intervals 

not to exceed 24 months (rather than the 

current requirement of 12 months) 

without conducting an assessment 

against performance indicators. This 

proposed change would not constitute 

backfitting or affect the issue finality of 

a COL because the proposed change 

would provide a voluntary alternative 

requirement. 

The NRC would add new § 50.54(t)(3) 

to remove the requirement to conduct 

periodic emergency preparedness 

program element reviews once all fuel is 

in dry cask storage. This proposed 

change would not meet the definition of 

‘‘backfitting’’ or affect the issue finality 

of a COL because the NRC would be 

removing a requirement. 

The addition of a new paragraph IV.8 

to appendix E to 10 CFR part 50 would 

clarify that the evacuation time estimate 

requirements of paragraphs IV.4, IV.5, 

and IV.6 would no longer be applicable 

to licensees after permanent cessation of 

operations and permanent removal of 

fuel from the reactor vessel. This 

proposed change would not meet the 

definition of ‘‘backfitting’’ or affect the 

issue finality of a COL because the NRC 

would be removing a requirement. 

The NRC would add a new paragraph 

k to part 50, appendix E, section IV.F.2 

to state that licensees in 

decommissioning need to follow the 

biennial exercise requirements of 

section IV.F.2. This is the current 

requirement for these licensees, so this 

change to the regulations would not 

change a requirement. Therefore, the 

proposed change would not meet the 

definition of ‘‘backfitting’’ or affect the 

issue finality of a COL. 

The NRC is proposing to remove 

obsolete dates for certain one-time 

actions that were required as part of the 

2011 emergency preparedness final rule 

and other obsolete dates. These actions 

are complete, and the requirements are 

no longer binding on any current 

licensee. These proposed changes 

would not meet the definition of 

‘‘backfitting’’ or affect the issue finality 

of a COL because the NRC would be 

removing a requirement. 

The proposed changes to 72.32(a) 

would clarify the emergency plan 

requirements for an applicant of a 

specific license under 10 CFR part 72. 

As discussed in section IX.A. of this 

document, applicants such as this one 

are outside the scope of the 10 CFR part 

50 backfitting provisions and issue 

finality provisions. 

The proposed changes to 72.32(c) 

would clarify that the ISFSI licensee can 

rely on its 10 CFR part 50 emergency 

plan to meet the requirements of § 72.32 

when the nuclear power reactor is under 

construction, operating, or in 

decommissioning. Other provisions of 

§ 72.32 allow an ISFSI licensee with a 

reactor emergency plan to use that 

emergency plan to meet the applicable 

requirements for an ISFSI emergency 

plan. Therefore, this clarification would 

not meet the definition of ‘‘backfitting’’ 

or affect the issue finality of a COL 

because it would not result in a 

modification of or addition to SSCs or 

the design of a facility or the procedures 

or organization required to design, 

construct, or operate a facility. 

Section IV.B of this document 

describes proposed changes to physical 

security requirements. The NRC would 

permit a certified fuel handler to 

approve the temporary suspension of 

security measures once the reactor has 

shut down and all fuel has been 

removed from the reactor core. This 

proposed change would not constitute 

backfitting or affect the issue finality of 

a COL because the proposed change 

would provide a voluntary alternative 

requirement. 

The proposed changes to § 50.54(p) 

would add definitions of ‘‘change’’ and 

‘‘decrease in safeguard effectiveness’’ 

and require that reactor licensees 

include with the required § 50.54(p)(2) 

report a summary of the analysis 

performed to determine that the change 

does not decrease safeguards 

effectiveness of the security plan. The 

proposed changes would not require a 

licensee to use the § 50.54(p) security 

plan change process unless the licensee 

voluntarily seeks to change its security 

plan and would not result in a 

modification of or addition to SSCs or 

the design of a facility or the procedures 


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12310 Federal Register / Vol. 87, No. 42 / Thursday, March 3, 2022 / Proposed Rules 

or organization required to design, 

construct, or operate a facility. 

Therefore, the proposed changes would 

not meet the definition of ‘‘backfitting’’ 

or affect the issue finality of a COL. 

The NRC would provide an option 

that, once all spent nuclear fuel has 

been placed in dry cask storage, 

licensees could protect a general license 

ISFSI under § 73.51 instead of § 73.55. 

This proposed change would not 

constitute backfitting or affect the issue 

finality of a COL because the proposed 

change would provide a voluntary 

alternative requirement. 

Current § 73.55(b)(3) requires that a 

licensee’s physical protection program 

be designed to prevent significant core 

damage. The NRC would remove this 

requirement once the NRC has docketed 

the licensee’s certifications that its 

reactor has permanently ceased 

operating and all fuel has been removed 

from the reactor vessel. This proposed 

change would not constitute backfitting 

or affect the issue finality of a COL. The 

issue finality provision for COLs located 

in § 52.98 provides, in relevant part, that 

the Commission may not modify, add, 

or delete any term or condition of a COL 

except in accordance with the 

provisions of § 50.109. Under § 50.109, 

removing a requirement as proposed 

with § 73.55(b)(3) is not backfitting 

because removing a requirement does 

not create a new requirement and does 

not amend a requirement because 

amending means the requirement still 

exists in some form. 

The proposed change to 

§ 73.55(e)(9)(v) would remove the 

requirement that a licensee must 

designate the reactor control room as a 

‘‘vital area’’ if the NRC has docketed the 

licensee’s certifications that the reactor 

has permanently ceased operating and 

all fuel has been removed from the 

reactor vessel, and the licensee has 

documented that all vital equipment has 

been removed from the control room 

and the control room does not serve as 

the vital area boundary for other vital 

areas. This proposed change would not 

constitute backfitting or affect the issue 

finality of a COL because the proposed 

change would be a voluntary alternative 

requirement. Even if a licensee 

submitted and the NRC docketed the 

certifications that the reactor has 

permanently ceased operating and all 

fuel has been removed from the reactor 

vessel, the licensee could still designate 

the reactor control room as a vital area. 

If not all of the vital equipment has been 

removed from the control room or the 

control room still serves as the vital area 

boundary for other vital areas, then the 

licensee would not be required to, and 

in fact could not, document that all vital 

equipment has been removed from the 

control room or the control room does 

not serve as the vital area boundary for 

other vital areas, respectively. 

The NRC would revise § 73.55(j)(4)(ii) 

to provide an alternative to the 

requirement for maintaining continuous 

communications between the alarm 

stations and the control room with a 

requirement for maintaining 

communications between alarm stations 

and the CFH or senior on shift licensee 

representative, once a licensee submits 

and the NRC dockets the certifications 

that the reactor has permanently ceased 

operating and all fuel has been removed 

from the reactor vessel. This proposed 

change would relax the requirement for 

these licensees. However, a licensee in 

decommissioning could maintain its 

control room such that its continuous 

communication system still 

communicates between the alarm 

stations and the control room. In this 

situation, the control room could 

redirect communications from the alarm 

stations to the certified fuel handler or 

the senior on-shift licensee 

representative as appropriate. Thus, a 

licensee could continue to comply with 

the current requirement to maintain 

continuous communications between 

the alarm stations and the control room 

and still satisfy the proposed rule. This 

makes the relaxation non-mandatory 

and, as explained in MD 8.4, nonmandatory relaxations of regulations 

generally do not meet the definition of 

‘‘backfitting.’’ This proposed change 

would provide the voluntary relaxation 

of a current requirement and, thus, not 

constitute backfitting or affect the issue 

finality of a COL. 

Section IV.C of this document 

describes proposed changes to cyber 

security requirements. The NRC would 

revise § 73.54 so the cyber security 

requirements in § 73.54 continue to 

apply to licensees through Level 1 of 

decommissioning. Each 10 CFR part 50 

licensee has a license condition 

requiring the licensee to maintain its 

cyber security plan, and this license 

condition remains in effect during 

decommissioning. If the NRC issues an 

operating license for a nuclear power 

reactor before this final rule goes into 

effect, then the NRC can include a 

license condition similar to those issued 

to current holders of operating licenses 

for nuclear power reactors. Thus, this 

proposed change would not constitute 

backfitting for 10 CFR part 50 licensees. 

A COL holder without the license 

condition is currently not required to 

maintain its cyber security plan when it 

begins decommissioning. The proposed 

revision to § 73.54 would constitute a 

change affecting the issue finality 

accorded these COL holders because 

extending the requirement to maintain a 

cyber security plan during 

decommissioning would modify the 

terms and conditions of a COL. Under 

§ 52.98, the NRC must apply the 

provisions of § 50.109 to the proposed 

change. The proposed change would 

constitute backfitting under § 50.109. 

The NRC’s backfit analysis justifying 

this backfitting action is presented in 

section IX.D of this document. If the 

NRC issues a COL before this final rule 

goes into effect, then the NRC can 

include a license condition similar to 

those issued to current holders of 

operating licenses for nuclear power 

reactors. 

Section IV.D of this document 

describes proposed changes to fitness 

for duty requirements. The NRC 

proposes to amend § 26.3(a) so the 

requirements of 10 CFR part 26 would 

not apply to COL holders once the NRC 

has docketed their § 52.110(a) 

certifications. This proposed change 

would not affect the issue finality of a 

COL because the NRC would be 

removing a requirement. 

The proposed changes to 

§ 73.55(b)(9)(ii)(B) would provide 

minimum requirements for the fitness 

for duty elements of operating and 

decommissioning 10 CFR part 50 and 10 

CFR part 52 licensees’ insider mitigation 

programs. These licensees are already 

required to comply with the insider 

mitigation program requirements of 

§ 73.55(b)(9), so the proposed rule 

changes would clarify existing 

requirements and would not constitute 

backfitting or affect the issue finality of 

a COL. 

The NRC proposes to amend the 

criminal penalties section of 10 CFR 

part 26 by including § 26.3 within 

§ 26.825(a) by removing § 26.3 from 

§ 26.825(b). This proposed change 

would not revise § 26.3 in any way. 

Enabling the NRC to impose criminal 

penalties for willful violations of, 

attempts to violate, or conspiracies to 

violate § 26.3 would not result in a 

modification of or addition to SSCs or 

the design of a facility or the procedures 

or organization required to design, 

construct, or operate a facility. 

Therefore, the proposed change would 

not meet the definition of ‘‘backfitting’’ 

or affect the issue finality of a COL. 

Section IV.E of this document 

describes proposed changes to the 

‘‘certified fuel handler’’ definition and 

the elimination of the shift technical 

advisor. The NRC proposes to amend 

§ 50.2 to provide an alternative 

definition of ‘‘certified fuel handler’’ to 

eliminate the need for licensees to 

submit requests for NRC approval of 


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CFH training programs. This proposed 

change would not constitute backfitting 

or affect the issue finality of a COL 

because the proposed change would 

provide a voluntary alternative to 

submitting a request for approval of a 

fuel handler training program. 

The proposed change to 

§ 50.54(m)(2)(i) to state that a shift 

technical advisor is not required upon 

the NRC’s docketing of the license 

holder’s certifications required under 

§§ 50.82(a)(1) or 52.110(a) would not 

result in a modification of or addition to 

SSCs or the design of a facility or the 

procedures or organization required to 

design, construct, or operate a facility. 

Therefore, the proposed change would 

not meet the definition of ‘‘backfitting’’ 

or affect the issue finality of a COL. 

Section IV.F of this document 

describes proposed changes to the 

NRC’s decommissioning funding 

assurance requirements. The proposed 

change to § 50.75(f)(1) would modify the 

reporting frequency for reactor 

decommissioning funding reports from 

at least once every 2 years to at least 

once every 3 years. This reporting 

requirement would not result in a 

modification of or addition to SSCs or 

the design of a facility or the procedures 

or organization required to design, 

construct, or operate a facility. 

Therefore, the proposed change would 

not meet the definition of ‘‘backfitting’’ 

or affect the issue finality of a COL. 

The NRC would revise § 50.75(h) to 

require certain notifications be sent 

directly to the NRC’s Document Control 

Desk and not to the Director, Office of 

Nuclear Reactor Regulation, or Director, 

Office of Nuclear Material Safety and 

Safeguards, as applicable. This reporting 

requirement would not result in a 

modification of or addition to SSCs or 

the design of a facility or the procedures 

or organization required to design, 

construct, or operate a facility. 

Therefore, the proposed change would 

not meet the definition of ‘‘backfitting’’ 

or affect the issue finality of a COL. 

The NRC proposes to delete 

§ 50.75(f)(2). The language of existing 

§ 50.75(f)(1) fully encompasses the 

language of paragraph (f)(2), and, 

therefore, paragraph (f)(2) is 

unnecessary and potentially confusing. 

This change would not result in a 

modification of or addition to SSCs or 

the design of a facility or the procedures 

or organization required to design, 

construct, or operate a facility. 

Therefore, the proposed change would 

not meet the definition of ‘‘backfitting’’ 

or affect the issue finality of a COL. 

The NRC is proposing to amend its 

regulations in § 50.75(f)(1) to clarify that 

when a licensee identifies a shortfall in 

the decommissioning funding report 

required by § 50.75(f)(1), the licensee 

must identify additional financial 

assurance to cover the shortfall in the 

next report. Licensees are already 

required to provide reasonable 

assurance of decommissioning funding 

on an ongoing basis. The proposed 

change would not change this 

obligation; the proposed rule would 

clarify how reasonable assurance of 

funds will be available for the 

decommissioning process. This change 

would not result in a modification of or 

addition to SSCs or the design of a 

facility or the procedures or 

organization required to design, 

construct, or operate a facility. 

Therefore, the proposed change would 

not meet the definition of ‘‘backfitting’’ 

or affect the issue finality of a COL. 

The proposed change to 

§ 50.82(a)(9)(ii)(F) would require 

licensees to identify the specific sources 

of funds for ‘‘remaining 

decommissioning costs,’’ including 

sources of funds for license termination, 

spent fuel management, and ISFSI 

decommissioning. This reporting 

requirement would not result in a 

modification of or addition to SSCs or 

the design of a facility or the procedures 

or organization required to design, 

construct, or operate a facility. 

Therefore, the proposed change would 

not meet the definition of ‘‘backfitting’’ 

or affect the issue finality of a COL. 

The NRC would revise § 50.82(a)(8)(v) 

to allow licensees to combine the 

reports that are required by 

§§ 50.82(a)(8)(v), 50.82(a)(8)(vii) and 

72.30(c). This proposed change would 

not constitute backfitting or affect the 

issue finality of a COL because the 

proposed change would provide a 

voluntary alternative requirement. 

The NRC proposes to revise § 52.110 

to make the same changes proposed in 

§ 50.82. For the reasons previously 

discussed, these proposed changes 

would not affect the issue finality of a 

COL. The NRC also proposes to add to 

§ 52.110 paragraphs (h)(5) through (h)(7) 

with site-specific decommissioning cost 

estimate reporting requirements that are 

identical to the requirements in 

§ 50.82(a)(8)(v) through (vii). These 

reporting requirements would not result 

in a modification of or addition to SSCs 

or the design of a facility or the 

procedures or organization required to 

design, construct, or operate a facility. 

Therefore, the proposed changes would 

not meet the definition of ‘‘backfitting’’ 

and, under § 52.98, would not affect the 

issue finality of a COL. 

The NRC proposes to revise § 72.30 so 

that the submittals subsequent to the 

initial decommissioning funding plan 

would no longer require NRC approval. 

This proposed change would not meet 

the definition of ‘‘backfitting’’ or affect 

the issue finality of a COL because the 

NRC would be removing a requirement. 

The proposed changes to § 72.30(b) 

would clarify the requirements for an 

applicant for a specific licensee and a 

holder of a general license to submit 

decommissioning funding plans for 

NRC review and approval. The current 

requirement requires applicants and 

holders of licenses under 10 CFR part 72 

to submit decommissioning funding 

plans for NRC review and approval. 

These changes would not change any 

substantive requirement and would not 

result in a modification of or addition to 

SSCs or the design of a facility or the 

procedures or organization required to 

design, construct, or operate a facility. 

Accordingly, these proposed changes 

would not constitute backfitting or 

affect the issue finality of a COL. 

Section IV.G of this document 

describes proposed changes to the 

NRC’s onsite and offsite financial 

protection requirements and indemnity 

agreements. These changes would 

include revisions to the following 

regulations: §§ 140.11(a)(5) and 

50.54(w)(5), to allow nuclear power 

reactor licensees in decommissioning to 

reduce the offsite liability and onsite 

property insurance amounts, 

respectively, that they are required to 

maintain; § 140.81, to include plants in 

decommissioning within the scope of 

§ 140.81, thereby clarifying the 

applicability of the requirements for an 

Extraordinary Nuclear Occurrence ENO 

to reactors in decommissioning; and 

§ 50.54(w), to require a prompt 

notification to the Commission of any 

material change in proof of onsite 

property insurance filed with the 

Commission under 10 CFR part 50. 

Changes to 10 CFR part 140 are not 

subject to the 10 CFR part 50 backfitting 

provisions and the issue finality 

provisions in 10 CFR part 52 because 

the Price-Anderson Act requires 

licensees to have offsite financial 

protection. Even if they were subject to 

the 10 CFR part 50 backfitting 

provisions and the issue finality 

provisions in 10 CFR part 52, the 

proposed changes would not result in a 

modification of or addition to SSCs or 

the design of a facility or the procedures 

or organization required to design, 

construct, or operate a facility. 

Therefore, the proposed changes would 

not meet the definition of ‘‘backfitting’’ 

and would not affect the issue finality 

of a COL. 

Similarly, the onsite insurance 

requirements in § 50.54(w) do not fall 

within the purview of the 10 CFR part 


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50 backfitting provisions or the issue 

finality provisions in 10 CFR part 52. In 

the backfitting discussion for the 1987 

final rule, ‘‘Changes in Property 

Insurance Requirements for NRC 

Licensed Nuclear Power Plants’’ (52 FR 

28963, 28972; August 5, 1987), the 

Commission stated that requiring an 

increase in property damage insurance 

does not meet the definition of 

‘‘backfitting.’’ The Commission took 

similar positions on backfitting in 

subsequent rulemakings to amend 

§ 50.54(w) (e.g., 54 FR 11163, March 17, 

1989; 55 FR 12163, April 2, 1990). 

Section IV.H of this document 

describes proposed changes to the 

requirements concerning consideration 

of environmental effects of 

decommissioning activities. The NRC 

proposes to change § 50.82(a)(4)(i) and 

§ 52.110(d)(1) to require that licensees 

provide the basis for determining 

whether the environmental impacts of 

decommissioning activities are bounded 

by previous environmental reviews and 

include a description in the PSDAR of 

any activities that will not be bounded. 

These reporting requirements would not 

result in a modification of or addition to 

SSCs or the design of a facility or the 

procedures or organization required to 

design, construct, or operate a facility. 

Therefore, the proposed changes would 

not meet the definition of ‘‘backfitting’’ 

and would not affect the issue finality 

of a COL. 

The NRC also proposes to change 

§ 50.82(a)(4)(i) and § 52.110(d)(1) to 

allow licensees to use appropriate 

federally issued environmental review 

documents prepared in compliance with 

NEPA, ESA, NHPA, or other 

environmental statutes instead of only 

environmental impact statements. These 

reporting requirements would not result 

in a modification of or addition to SSCs 

or the design of a facility or the 

procedures or organization required to 

design, construct, or operate a facility. 

Therefore, the proposed changes would 

not meet the definition of ‘‘backfitting’’ 

and would not affect the issue finality 

of a COL. 

The NRC would change 

§ 50.82(a)(6)(ii) and § 52.110(f)(2) to 

clarify that the previous review of any 

potentially significant environmental 

impact must be bounded by appropriate 

federally issued environmental review 

documents prepared in compliance with 

NEPA, ESA, NHPA, or other 

environmental statutes. These reporting 

requirements would not result in a 

modification of or addition to SSCs or 

the design of a facility or the procedures 

or organization required to design, 

construct, or operate a facility. 

Therefore, the proposed changes would 

not meet the definition of ‘‘backfitting’’ 

and would not affect the issue finality 

of a COL. 

The NRC proposes to revise 10 CFR 

part 51 to reflect the changes made in 

the 1996 Final Rule that nuclear power 

reactor licensees are not required to 

submit license amendment requests for 

authorization to perform 

decommissioning activities. These 

changes would not change any 

substantive requirement and would not 

result in a modification of or addition to 

SSCs or the design of a facility or the 

procedures or organization required to 

design, construct, or operate a facility. 

Accordingly, these proposed changes 

would not constitute backfitting or 

affect the issue finality of a COL. 

Section IV.I of this document 

describes proposed changes to record 

retention requirements. These changes 

would eliminate certain recordkeeping 

requirements and the requirement to 

keep certain duplicate records. These 

recordkeeping changes would not result 

in a modification of or addition to SSCs 

or the design of a facility or the 

procedures or organization required to 

design, construct, or operate a facility. 

Therefore, the proposed changes would 

not meet the definition of ‘‘backfitting’’ 

and would not affect the issue finality 

of a COL. The proposed changes also 

would not meet the definition of 

‘‘backfitting’’ or affect the issue finality 

of a COL because the NRC would be 

removing these requirements. 

Section IV.J of this document 

describes proposed changes to low-level 

radioactive waste transportation 

requirements. The NRC would revise 

Paragraph III.E of appendix G to 10 CFR 

part 20 to increase from 20 days to 45 

days the window of time for notification 

of receipt of shipments of low-level 

waste before a shipper would be 

required to investigate, trace, and report 

to the NRC any shipments of low-level 

waste for which the shipper has not 

received a notification of receipt. This 

proposed change would relax the 

requirement. However, a shipper could 

still investigate, trace, and report 

shipments of low-level waste if the 

shipper has not received notification of 

receipt within 20 days. Thus, a shipper 

could continue to comply with the 

current 20-day requirement and still 

satisfy the proposed rule. This makes 

the relaxation non-mandatory and, as 

explained in MD 8.4, non-mandatory 

relaxations of regulations generally do 

not meet the definition of ‘‘backfitting.’’ 

This proposed change would provide 

the voluntary relaxation of a current 

requirement and, thus, not constitute 

backfitting or affect the issue finality of 

a COL. 

Section IV.K of this document 

describes proposed changes to spent 

fuel management requirements. The 

NRC would revise §§ 50.54(bb) and 

72.218 to clarify the contents of an 

irradiated fuel management plan, which 

licensees are already required to submit 

to the NRC for approval. This 

clarification of a reporting requirement 

would not result in a modification of or 

addition to SSCs or the design of a 

facility or the procedures or 

organization required to design, 

construct, or operate a facility. 

Therefore, the proposed changes would 

not meet the definition of ‘‘backfitting’’ 

and would not affect the issue finality 

of a COL. 

The NRC proposes to change § 72.218 

to remove spent fuel management 

provisions that the NRC would move to 

§ 50.54(bb) and clarify provisions 

concerning termination of part 72 

general licenses. The proposed changes 

would not result in a modification of or 

addition to SSCs or the design of a 

facility or the procedures or 

organization required to design, 

construct, or operate a facility. 

Therefore, the proposed changes would 

not meet the definition of ‘‘backfitting’’ 

and would not affect the issue finality 

of a COL. 

Section IV.L of this document 

describes proposed changes to the 

NRC’s backfitting provisions in 10 CFR 

part 50 and part 72. The NRC proposes 

to change § 50.109 to clarify application 

of the 10 CFR part 50 backfitting 

provisions to NRC actions constituting 

backfitting or affecting the issue finality 

of nuclear power reactor licensees in 

decommissioning. The NRC also would 

revise § 50.109 to require a documented 

evaluation to include a consideration of 

the costs of imposing the backfit if the 

basis for backfitting is bringing a facility 

into compliance with a license or the 

rules or orders of the Commission, or 

into conformance with the licensee’s 

written commitments. The proposed 

change to § 72.62 would clarify that the 

backfit regulations in part 72 apply 

during the decommissioning of an 

independent spent fuel storage 

installation or a monitored retrievable 

storage facility. The proposed changes 

to backfitting provisions would be 

changes to requirements imposed on the 

NRC, not on a licensee, so the proposed 

changes would be outside the scope of 

backfitting and issue finality. 

Section IV.M of this document 

describes proposed changes to the 

NRC’s regulations related to foreign 

ownership, control, or domination of a 

production or utilization facility. The 

NRC would revise § 50.38 to clarify 

when a facility licensed under 10 CFR 


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part 50 or part 52 is not considered a 

production or utilization facility and, 

therefore, the foreign ownership, 

control, or domination prohibition no 

longer applies. The proposed changes 

would not result in a modification of or 

addition to SSCs or the design of a 

facility or the procedures or 

organization required to design, 

construct, or operate a facility. 

Therefore, the proposed changes would 

not meet the definition of ‘‘backfitting’’ 

and would not affect the issue finality 

of a COL. 

The NRC also would amend § 50.82(b) 

to add the criteria for when a non-power 

production or utilization facility or fuel 

reprocessing plant is no longer a 

production or utilization facility. The 

only part 50 licensees considered within 

the scope of the part 50 backfitting 

provision are nuclear power reactor 

licensees. Further, the proposed changes 

would not result in a modification of or 

addition to SSCs or the design of a 

facility or the procedures or 

organization required to design, 

construct, or operate a facility. 

Therefore, the proposed changes would 

not meet the definition of ‘‘backfitting.’’ 

The NRC would revise § 50.82(a) and 

(b) and § 52.110(b) to affirm the 

continuation of the NRC’s statutory 

authority over the existing 10 CFR part 

50 or 10 CFR part 52 license, and to 

state which regulations would still 

apply to the licensee, after the 

performance of decommissioning 

activities that lead to the licensed 

facility no longer meeting the definition 

of a utilization or a production facility. 

The proposed changes would not result 

in a modification of or addition to SSCs 

or the design of a facility or the 

procedures or organization required to 

design, construct, or operate a facility. 

Therefore, the proposed changes would 

not meet the definition of ‘‘backfitting’’ 

and would not affect the issue finality 

of a COL. 

In light of the proposed amendments 

to §§ 50.38, 50.82, and 52.110, the NRC 

would amend §§ 50.1, 50.51, 52.0, and 

52.109 to clarify that the regulations in 

10 CFR part 50, and the similar 

regulations in 10 CFR part 52, provide 

not only for the licensing of utilization 

and production facilities, but also for 

their decommissioning and the 

termination of their associated licenses. 

The proposed changes would not result 

in a modification of or addition to SSCs 

or the design of a facility or the 

procedures or organization required to 

design, construct, or operate a facility. 

Therefore, the proposed changes would 

not meet the definition of ‘‘backfitting’’ 

and would not affect the issue finality 

of a COL. 

The NRC is proposing to add a 

definition for ‘‘non-power production or 

utilization facility’’ to § 50.2 that 

captures all non-power facilities 

licensed under § 50.22 and § 50.21(a) or 

(c), except fuel reprocessing facilities. 

The only part 50 licensees considered 

within the scope of the part 50 

backfitting provision are nuclear power 

reactor licensees. Further, the proposed 

definition would not result in a 

modification of or addition to SSCs or 

the design of a facility or the procedures 

or organization required to design, 

construct, or operate a facility. 

Therefore, the proposed changes would 

not meet the definition of ‘‘backfitting.’’ 

Section IV.N of this document 

describes proposed changes to license 

termination plan requirements. The 

NRC would revise § 50.82(a)(9) and 

§ 52.110(i) to clarify that only nuclear 

power reactor licensees that have loaded 

fuel into their reactors must submit 

license termination plans. The proposed 

change would not change this 

requirement; the proposed rule would 

only clarify that nuclear power reactor 

licensees that have not loaded fuel into 

their reactors would not need to submit 

license termination plans. This change 

would not result in a modification of or 

addition to SSCs or the design of a 

facility or the procedures or 

organization required to design, 

construct, or operate a facility. 

Therefore, the proposed change would 

not meet the definition of ‘‘backfitting’’ 

or affect the issue finality of a COL. 

Section IV.O of this document 

describes the proposed removal of 

license conditions and withdrawal of 

orders. These changes would not change 

any substantive requirement because the 

license conditions and orders are 

substantively redundant with NRC 

regulations issued after the license 

conditions and orders were issued. 

Because the NRC would not change a 

requirement, the proposed changes 

would not meet the definition of 

‘‘backfitting’’ or affect the issue finality 

of a COL. 

D. Backfit Analysis 

1. Introduction and Background 

As part of this proposed rule, the NRC 

is proposing a modification to the cyber 

security requirements in § 73.54. This 

proposed rule would ensure that these 

requirements continue to apply to 

nuclear power reactor licensees that 

have submitted their § 50.82(a)(1) or 

§ 52.110(a) certifications until such time 

that all spent fuel in the SFP has 

sufficiently decayed (i.e., at least 10 

months for BWRs and 16 months for 

PWRs after the date of permanent 

cessation of operations, or an NRCapproved alternative spent fuel decay 

period). 

This amendment would likely 

constitute a change affecting issue 

finality for 10 CFR part 52 COL holders, 

as defined in § 52.98. These licensees 

are not currently required to maintain 

their cyber security programs past the 

date that they are no longer authorized 

to operate the reactor. If the proposal to 

require these licensees to maintain their 

cyber security program into the 

decommissioning phase would extend 

the duration that a COL holder would be 

required to maintain a cyber security 

program, then that extension would 

constitute a new or changed 

requirement for that licensee and, thus, 

affect that COL’s issue finality. 

2. Detailed Description of the Proposed 

Change Affecting Issue Finality 

The NRC sets forth the current cyber 

security requirements for nuclear power 

reactors in § 73.54. The NRC established 

these requirements as part of the 2009 

Power Reactor Security Requirements 

final rule. The preamble to § 73.54 

states, in part, that by November 23, 

2009, each nuclear power reactor 

licensee ‘‘currently licensed to operate’’ 

must submit to the NRC a cyber security 

plan (CSP) for review and approval. The 

preamble further states that the 

requirements in § 73.54 are applicable to 

current ‘‘applicants for an operating 

license or combined license’’ and 

mandates such applicants to amend 

their applications to include a CSP. In 

addition, every 10 CFR part 50 license 

for a nuclear power reactor that was 

operating in 2009 contains a license 

condition to have and maintain a 

Commission-approved CSP. These 

license conditions were issued when the 

NRC approved each licensee’s CSP that 

was submitted to the NRC as required 

by the Power Reactor Security 

Requirements final rule. The Tennessee 

Valley Authority’s 10 CFR part 50 

operating license for Watts Bar Nuclear 

Plant, Unit 2, issued in 2015, also 

contains a license condition to have and 

maintain a CSP. 

As an initial step in the 

decommissioning process, a nuclear 

power reactor licensee must submit 

written certifications that it has decided 

to permanently cease operations and has 

permanently removed all fuel from its 

reactor vessel, in accordance with 

§ 50.82(a)(1)(i) and (ii) for nuclear power 

reactor licensees under 10 CFR part 50, 

or § 52.110(a)(1) and (2) for 10 CFR part 

52 combined license holders. As stated 

in § 50.82(a)(2) and § 52.110(b), upon 

the NRC’s docketing of these 

certifications, the license no longer 


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authorizes operation of the reactor or 

the placement or retention of fuel in the 

reactor vessel. In a December 5, 2016 

memorandum to the Commission, the 

NRC staff explained that § 73.54 no 

longer applies to nuclear power reactor 

licensees once they have submitted, and 

the NRC has docketed, these 

certifications. 

As discussed in the ‘‘Technical Basis 

for Graded Approach’’ section of this 

document, the NRC has concluded that 

after 10 months for BWRs and 16 

months for PWRs, the spent fuel in the 

SFP will have decayed and cooled 

sufficiently such that the fuel cannot 

heat up to clad ignition temperature 

within 10 hours under adiabatic 

conditions. The NRC has determined 

that until the fuel has decayed and 

cooled sufficiently, nuclear power 

reactor licensees must maintain 

reasonable assurance that their critical 

digital assets remain protected against 

cyber attacks. As such, this proposed 

rule would modify the cyber security 

requirements in § 73.54 to ensure that 

they continue to apply to licensees of 

decommissioning nuclear power 

reactors until the spent fuel has decayed 

and cooled sufficiently (either through 

the application of a 10 month (BWR) or 

16 month (PWR) decay period or an 

NRC-approved site-specific decay 

period). This proposed rule would also 

remove the CSP license condition from 

the 10 CFR part 50 licenses at the 

applicable 10 or 16 month interval. 

This proposed rule would not 

constitute backfitting for currently 

operating or recently shutdown 10 CFR 

part 50 reactor licensees. Their CSP 

license condition remains in effect until 

the termination of the license or the 

NRC removes the condition from the 

license (e.g., if the licensee submits a 

license amendment request and the NRC 

approves it). The NRC has determined 

that the requirements of the CSP license 

conditions are not necessary after the 

spent fuel in the SFP has sufficiently 

cooled. The proposed rule would 

codify, during Level 1 of 

decommissioning, the already-imposed 

requirements of the CSP license 

conditions. These requirements would 

continue to provide adequate protection 

of the public health and safety and 

common defense and security and 

continue to support the effective 

operation of licensees’ security and 

emergency preparedness programs 

during the time when a draindown 

scenario can credibly lead to a 

zirconium fire. (See sections 3 and 4 of 

this backfit analysis for additional cost/ 

benefit discussion.) Therefore, this 

proposed rule would not impact these 

licensees’ overall requirement to 

maintain a cyber security program, but 

would instead enable the automatic 

removal of cyber security requirements 

once fuel in the SFP has sufficiently 

cooled. Thus, the decommissioning 

rulemaking would not impose a new or 

changed requirement as the licensees 

are already implementing the 

requirement as part of their cyber 

security program license conditions. 

Conversely, this rulemaking would 

constitute a change affecting the issue 

finality for 10 CFR part 52 COL holders. 

Each currently approved COL includes 

a license condition to provide the NRC 

with the licensee’s Operational Program 

Implementation Schedule. The 

operational programs (which include 

development and implementation of a 

security program, including a cyber 

security program) are requirements in 

the regulations and not separately 

identified as license conditions. As a 

result, a COL does not require the 

licensee to maintain the cyber security 

program throughout the duration of its 

license. COL holders are currently 

required to maintain a program only as 

long as § 73.54 is applicable to them. 

Because § 73.54 no longer applies to the 

licensee once it is not authorized to 

operate a nuclear power reactor, and a 

nuclear power reactor licensee is not 

authorized to operate a nuclear power 

reactor during decommissioning, COL 

holders are not required to maintain 

their CSP during decommissioning. This 

proposed rule, which would require 

licensees to maintain their cyber 

security program for 10 months (BWR) 

or 16 months (PWR) beyond the date of 

permanent cessation of operations (or 

for an NRC-approved alternative spent 

fuel decay period) could extend the 

duration over which a COL holder 

would be required to maintain a cyber 

security program. That extension would 

constitute a new or changed 

requirement for that licensee. 

Under § 52.98, the Commission 

cannot modify any term or condition of 

an issued combined license except in 

accordance with the provisions of 

§ 52.103 or § 50.109, as applicable. This 

proposed rule’s amendment of the cyber 

security requirements would constitute 

a change affecting the issue finality of 

the COLs issued at the time of the final 

rule’s effective date. The provisions of 

§ 52.103 do not apply to this proposed 

rule, so the NRC must show that the 

amendment would meet the 

requirements of § 50.109 to justify 

proceeding with this amendment. 

Because none of the exceptions to the 

requirement to prepare a backfit 

analysis in § 50.109(a)(4) applies to this 

rulemaking, § 50.109(a)(3) requires the 

NRC to prepare a backfit analysis that 

demonstrates that the proposed 

amendment would result in a 

substantial increase in the overall 

protection of the public health and 

safety or the common defense and 

security, and that the direct and indirect 

costs of implementation are justified in 

view of this increased protection. 

3. Benefits: Substantial Increase in 

Public Health and Safety and Common 

Defense and Security 

The NRC identified qualitative (nonquantifiable) benefits that would occur 

if the proposed change affecting issue 

finality were implemented. 

The NRC identified two qualitative 

benefits to the common defense and 

security and public health and safety 

that would be realized if the proposed 

rule is implemented. Specifically, the 

NRC finds that extending the duration 

over which the licensee must maintain 

cyber security requirements would: 

• Constitute a substantial increase in 

protection to common defense and 

security by ensuring that a compromise 

of digital systems cannot adversely 

impact the effective operation of 

licensees’ physical security programs; 

and 

• Constitute a substantial increase in 

public health and safety by ensuring 

that a compromise of digital systems 

cannot adversely impact the effective 

operation of emergency preparedness 

systems in the event of a zirconium 

fire scenario. 

Effective Operation of Physical Security 

Program 

The NRC has previously determined 

that attacks on the SFP are credible and 

have the potential to lead to an 

unacceptable impact to common 

defense and security. Specifically, a 

physical attack by either an external 

force or malicious insiders could 

directly lead to a draindown scenario 

and subsequent zirconium fire. 

As established in § 73.54, cyber 

security is an essential element of a 

licensee’s physical security program 

that enables the licensee to effectively 

protect its site against the design basis 

threat of radiological sabotage defined 

in § 73.1, in accordance with § 73.55(b). 

Specifically, a physical attack that is 

augmented with a coincident cyber 

attack would, in many cases, have a 

higher chance of success over a purely 

physical attack. Thus, although there is 

no cyber attack that can directly lead to 

a draindown scenario, a cyber attack can 

be combined with a physical attack on 

the SFP to improve the physical attack’s 

likelihood of success. 

Given a facility without adequate 

cyber security controls in place, several 


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mechanisms exist that could improve 

the effectiveness of a physical attack on 

the SFP. For example, a cyber attack 

could aid a physical assault on the SFP 

by an external attacker by: 

• Disabling perimeter detection to delay 

or prevent onsite response to the 

physical assault prior to the attacker 

gaining entry to the SFP 

• disrupting onsite and offsite securityrelated communication to reduce the 

effectiveness of the licensee’s 

response to the physical assault 

• disabling access control doors and 

gates to enable the attacker expedited 

physical access to the SFP 

In addition, inadequate cyber security 

controls on facilities’ access control 

systems could enable an attacker to 

inject information into a licensee’s 

access control system in a manner that 

would allow unauthorized individuals 

to obtain unescorted access into the 

protected or vital areas of the facility. 

This could allow one or more attackers 

direct access to the SFP, which could 

then be exploited to sabotage the SFP in 

a manner that would result in a 

draindown scenario. 

This factor, combined with the 

severity of the consequences of a 

draindown scenario and subsequent 

zirconium fire that could result from a 

successful physical attack, demonstrates 

that maintaining cyber security 

requirements during the period when a 

draindown scenario could reasonably 

result in a zirconium fire (i.e., prior to 

the fuel in the SFP sufficiently cooling) 

represents a substantial increase in 

security. 

Effective Operation of Emergency 

Preparedness Systems 

As discussed in the ‘‘Technical Basis 

for the Graded Approach’’ and 

‘‘Emergency Preparedness’’ sections of 

this document, although the spectrum of 

credible accidents and operational 

events requiring an emergency response 

is reduced at a decommissioning 

nuclear power reactor as compared to 

that for an operating nuclear power 

reactor, reliable emergency 

preparedness functions are still required 

to ensure public health and safety in the 

event of a zirconium fire scenario. 

As established in § 73.54, cyber 

security is an essential element of a 

licensee’s physical security program 

that, in part, ensures that a compromise 

of digital systems cannot adversely 

impact emergency preparedness 

functions. For example, in the event of 

a zirconium fire scenario, the licensee’s 

cyber security program prevents a cyber 

attack from adversely impacting the 

ability to: 

• Notify state, local, and Federal 

personnel of the emergency 

• Request and communicate with offsite 

support 

• Assess and classify the emergency 

conditions 

• Disseminate information to the public 

during an emergency 

• Conduct a radiological accident 

assessment 

The NRC has determined that this 

factor demonstrates that maintaining 

cyber security requirements to ensure 

that a compromise of digital systems 

cannot adversely impact the operation 

of emergency preparedness functions 

until the time in which a SFP 

draindown would likely be mitigated 

prior to a zirconium fire scenario (i.e., 

once the fuel in the SFP has sufficiently 

cooled) represents a substantial increase 

in public health and safety. 

4. Costs 

The NRC identified quantitative costs 

(i.e., costs that are amenable to 

quantitative evaluation) that would be 

incurred if the proposed change 

affecting issue finality were 

implemented. 

Based on a review of feedback 

received during recent inspections of 

the full implementation of licensees’ 

cyber security programs, the NRC 

estimates that the cost to implement a 

cyber security program for a 

decommissioning nuclear power reactor 

is approximately $300,000 per site per 

year. As previously stated, this 

proposed change affecting issue finality 

would extend the duration that a 

licensee must maintain its cyber 

security program for 10 (BWR) or 16 

(PWR) months. Thus, the cost associated 

with this extension is approximately 

$250,000 (BWR) or $400,000 (PWR). 

COLs have been issued at a total of 3 

sites that utilize BWR units, and 4 sites 

that utilize PWR units. Assuming that 

all units are constructed and the per-site 

costs from the previous paragraph, the 

total cost associated with this proposed 

change affecting issue finality if all 

reactors entered decommissioning today 

would be approximately $2.35 million. 

If it is assumed that all sites with units 

licensed under 10 CFR part 52 

decommission their reactors 40 years 

after the effective date of the final rule, 

with a discount rate of 7 percent, then 

the total, combined cost for all affected 

licensees associated with this proposed 

change affecting issue finality would be 

approximately $157,000. Due to the 

potential that some of these facilities 

may not be constructed or that some 

licensees may have voluntarily chosen 

to maintain their cyber security 

programs during this timeframe, this 

estimate is expected to be an upper 

bound. 

5. Determination of Substantial Benefits 

Justifying Costs of the Proposed Change 

Affecting Issue Finality 

The NRC finds that the proposed 

change affecting issue finality would 

provide a substantial increase in 

protection to public health and safety 

and common defense and security for 

current 10 CFR part 52 COL holders by 

ensuring that a compromise of digital 

systems cannot adversely impact the 

effective operation of licensees’ security 

and emergency preparedness programs 

during the time when a draindown 

scenario can credibly lead to a 

zirconium fire. The NRC finds that this 

substantial increase would justify the 

$157,000 in costs that would accrue to 

the licensees. 

6. Conclusion 

On the basis of this analysis, the NRC 

determines that the change affecting 

issue finality resulting from the cyber 

security portion of this proposed rule 

would be justified under § 50.109(a)(3). 

7. Evaluation of Factors in § 50.109(c)(1) 

Through (9) 

In performing this analysis, the NRC 

considered the nine factors in 

§ 50.109(c), as follows: 

Statement of the Specific Objectives 

That the Backfit Is Designed To Achieve 

The two objectives for the cyber 

security portion of the ‘‘Regulatory 

Improvements for Production and 

Utilization Facilities Transitioning to 

Decommissioning’’ rulemaking are: 

• To ensure the effectiveness of the 

physical protection program during 

the period over which a SFP 

draindown could realistically result 

in a zirconium fire scenario; and 

• To ensure the effectiveness of 

emergency preparedness functions 

during the period over which a SFP 

draindown may not be mitigatable 

prior to the draindown resulting in a 

zirconium fire 

Note that the change affecting issue 

finality is only applicable to nuclear 

power reactors licensed under 10 CFR 

part 52 as of the effective date of the 

final rule. 

General Description of the Activity That 

Will Be Required by the Licensee or 

Applicant in Order To Complete the 

Backfit 

The NRC is proposing a modification 

to the cyber security requirements in 

§ 73.54 to ensure that these 

requirements continue to apply to 


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licensees of decommissioning nuclear 

power reactors until such time that all 

spent fuel in the SFP has sufficiently 

decayed (i.e., 10 months for BWRs and 

16 months for PWRs since the date of 

permanent cessation of operations, or an 

NRC-approved alternative spent fuel 

decay period). The change affecting 

issue finality is only applicable to 

nuclear power reactors currently 

licensed under 10 CFR part 52 as of the 

effective date of the final rule. 

Potential Change in the Risk to the 

Public From the Accidental Off-Site 

Release of Radioactive Material 

The rulemaking is intended to reduce 

risk of offsite releases as a result of 

breaches in security at nuclear power 

plants, and to ensure the functionality 

of emergency preparedness functions in 

the case of a zirconium fire scenario. 

However, the reduction in risk to the 

public from offsite releases of 

radioactive materials has not been fully 

quantified because there is insufficient 

information and modeling to support 

such quantification. 

Potential Impact on Radiological 

Exposure of Facility Employees 

The rulemaking would provide added 

assurance that nuclear industry workers 

are not subjected to unnecessary 

radiological exposures as the result of a 

breach in security that causes a 

zirconium fire leading to a release of 

radiation that security personnel are 

exposed to as the result of their 

response activities. Further, the 

rulemaking would ensure that 

emergency preparedness functions, 

including evacuation procedures, are 

not adversely impacted by a cyber attack 

during the period when a draindown 

scenario could reasonably result in a 

zirconium fire, thus ensuring that 

nuclear industry workers are not 

subjected to unnecessary radiological 

exposures in the case of a zirconium fire 

scenario. 

Installation and Continuing Costs 

Associated With the Backfit, Including 

the Cost of Facility Downtime or the 

Cost of Construction Delay 

The backfit analysis to support the 

change affecting issue finality resulting 

from this proposed rule includes the 

NRC’s estimate of the total costs for 

maintaining a licensee’s cyber security 

program until the fuel in the SFP has 

sufficiently cooled to adequately ensure 

that a SFP draindown does not result in 

a zirconium fire scenario. The estimated 

one-time industry net cost associated 

with the change affecting issue finality 

would be approximately $157,000. 

The Potential Safety Impact of Changes 

in Plant or Operational Complexity, 

Including the Relationship to Final and 

Existing Regulatory Requirements 

The cyber security portion of this 

proposed rule would not impose any 

requirements beyond those in place 

while the nuclear power reactor is 

operational. As such, this rule is not 

expected to have an effect on facility 

complexity. 

The Estimated Resource Burden on the 

NRC Associated With the Backfit and 

the Availability of Such Resources 

The rulemaking may result in a minor 

increase in the expenditure of agency 

resources, due to the potential for cyber 

security inspections to be conducted 

after the licensee has ceased operations 

and before fuel in the SFP has 

sufficiently cooled. 

The Potential Impact of Differences in 

Facility Type, Design or Age on the 

Relevancy and Practicality of the Backfit 

The specific cost of this rulemaking to 

a facility does vary, depending on 

whether the facility utilizes BWR or 

PWR reactors. This is due to time 

required for fuel in the SFP to 

sufficiently cool for each type of reactor. 

Further, since the change affecting issue 

finality is only applicable to reactors 

licensed under 10 CFR part 52, the 

specific cost also depends on the 

percentage of reactors licensed under 10 

CFR part 52 at the licensee’s facility. 

Whether the Backfit is Interim or Final 

and, if Interim, the Justification for 

Imposing the Backfit on an Interim Basis 

The change affecting issue finality 

would be final. 

E. Draft Regulatory Guidance 

As described in Section XVI, 

‘‘Availability of Guidance,’’ in this 

document, the NRC is issuing four draft 

regulatory guides (DGs) that, if finalized, 

would provide guidance on the methods 

acceptable to the NRC for complying 

with aspects of this proposed rule. The 

DGs would apply to all current holders 

of operating licenses under 10 CFR part 

50 and COLs under 10 CFR part 52. 

Issuance of the DGs in final form would 

not constitute backfitting under § 50.109 

and would not otherwise constitute a 

change affecting issue finality under 10 

CFR part 52. As discussed in the 

‘‘Implementation’’ section of each DG, 

the NRC has no current intention to 

impose the DGs on current holders of an 

operating license or COL. 

For the same reasons provided under 

‘‘Current and Future Applicants’’ that 

explain why the proposed rule does not 

constitute backfitting or a change 

affecting issue finality for applicants, 

applying the DGs to applications for 

operating licenses or COLs would not 

constitute backfitting as defined in 

§ 50.109 and would not otherwise 

constitute a change affecting issue 

finality under 10 CFR part 52. 

X. Cumulative Effects of Regulation 

The NRC is following its Cumulative 

Effects of Regulation (CER) process by 

engaging extensively with external 

stakeholders throughout this rulemaking 

and related regulatory activities. Public 

involvement has included: (1) The 

publication of an ANPR for public 

comment (80 FR 72358) on November 

19, 2015, to inform the NRC’s efforts in 

drafting a proposed rule regulatory basis 

to address issues associated with 

nuclear power reactor 

decommissioning; (2) holding a public 

meeting on December 9, 2015, to afford 

external stakeholders an opportunity to 

ask the NRC staff clarifying questions 

regarding the ANPR; (3) the publication 

of the draft regulatory basis for public 

comment (82 FR 13778) on March 15, 

2017; (4) the publication of a 

preliminary draft of the regulatory 

analysis for public comment (82 FR 

21481) on May 9, 2017; and (5) holding 

a public meeting on May 8–10, 2017, to 

facilitate public comments on the 

development of the final regulatory 

basis and regulatory analysis. 

Another opportunity for comment is 

being provided to the public with this 

proposed rule. The NRC will be issuing 

the draft implementing guidance with 

this proposed rule to support more 

informed external stakeholder feedback. 

Further, the NRC will continue to hold 

public meetings throughout the 

rulemaking process. Section XVI, 

‘‘Availability of Guidance,’’ of this 

document describes how the public can 

access the draft implementing guidance 

for which the NRC seeks external 

stakeholder feedback. 

Finally, the NRC is requesting CER 

feedback on the following questions: 

1. In light of any current or projected 

CER challenges, does the proposed 

rule’s effective date provide sufficient 

time to implement the new proposed 

requirements, including changes to 

programs, procedures, and facilities? 

2. If CER challenges currently exist or 

are expected, what should be done to 

address them? For example, if more 

time is required for implementation of 

the new requirements, what period of 

time is sufficient? 

3. Do other (NRC or other agency) 

regulatory actions (e.g., orders, generic 

communications, license amendment 

requests, inspection findings of a 

generic nature) influence the 


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implementation of the proposed rule’s 

requirements? 

4. Are there unintended 

consequences? Does the proposed rule 

create conditions that would be contrary 

to the proposed rule’s purpose and 

objectives? If so, what are the 

unintended consequences, and how 

should they be addressed? 

5. Please comment on the NRC’s cost 

and benefit estimates in the draft 

regulatory analysis that supports the 

proposed rule. The draft regulatory 

analysis is available as indicated in the 

‘‘Availability of Documents’’ section of 

this document. 

XI. Plain Writing 

The Plain Writing Act of 2010 (Pub. 

L. 111–274) requires Federal agencies to 

write documents in a clear, concise, and 

well-organized manner. The NRC has 

written this document to be consistent 

with the Plain Writing Act as well as the 

Presidential Memorandum, ‘‘Plain 

Language in Government Writing,’’ 

published June 10, 1998 (63 FR 31885). 

The NRC requests comment on this 

document with respect to the clarity and 

effectiveness of the language used. 

XII. National Environmental Policy Act 

This proposed rule includes some 

actions that are of the types described in 

§ 51.22(c). The NRC has previously 

determined that these types of actions 

do not have a significant impact on the 

environment and has categorically 

excluded them from the requirement to 

prepare an environmental analysis. 

Specifically, the NRC has determined 

that some amendments in this proposed 

rule are the types of actions described 

in the § 51.22(c) exclusions noted in 

Table 4. Accordingly, the NRC has not 

developed an environmental impact 

statement or an environmental 

assessment for these portions of the 

proposed rule. 

TABLE 4—APPLICATION OF 10 CFR 

51.22 CATEGORICAL EXCLUSIONS TO

THE PROPOSED REQUIREMENTS

Regulation 

Applicable 10 

CFR 51.22 

paragraph 

10 CFR part 26 ....................... (c)(1), (c)(3). 

10 CFR 50.2 ........................... (c)(2), (c)(3). 

10 CFR 50.54(bb) ................... (c)(3). 

10 CFR 50.59(d) ..................... (c)(3). 

10 CFR 50.71(c) ..................... (c)(3). 

10 CFR 50.75(f) ...................... (c)(3). 

Elimination of 10 CFR 

50.75(f)(2).

(c)(2). 

10 CFR 50.82(a) ..................... (c)(2), (c)(3). 

10 CFR 50.109 ....................... (c)(2). 

10 CFR part 50, appendix A ... (c)(3). 

10 CFR part 20, appendix G .. (c)(3). 

10 CFR 51.53 ......................... (c)(3). 

TABLE 4—APPLICATION OF 10 CFR 

51.22 CATEGORICAL EXCLUSIONS TO

THE PROPOSED REQUIREMENTS— 

Continued 

Regulation 

Applicable 10 

CFR 51.22 

paragraph 

10 CFR 51.95 ......................... (c)(3). 

10 CFR 52.63 ......................... (c)(3). 

10 CFR 52.110 ....................... (c)(2). 

10 CFR 72.72 ......................... (c)(3). 

10 CFR 72.218 ....................... (c)(3). 

10 CFR part 140 ..................... (c)(1). 

Draft Finding of No Significant Impacts 

The NRC has prepared a draft 

environmental assessment (EA) for the 

portions of this proposed rule not 

categorically excluded under § 51.22. 

The draft EA is available in ADAMS at 

Accession No. ML22019A140. The NRC 

prepared the draft EA to determine 

environmental impacts of the proposed 

action: A rulemaking to update the 

NRC’s regulations related to production 

and utilization facilities transitioning to 

decommissioning. Based on the draft 

EA, the NRC concludes that this 

proposed rule would not have 

significant environmental impacts 

because the changes would be 

administrative or procedural in nature 

and would have no nexus to the 

physical environment or would have no 

significant impact on the environment. 

Therefore, this proposed rule does not 

warrant preparation of an 

environmental impact statement. 

Accordingly, the NRC has determined 

that a finding of no significant impact 

(FONSI) is appropriate. 

XIII. Paperwork Reduction Act 

This proposed rule contains new or 

amended collections of information 

subject to the Paperwork Reduction Act 

of 1995 (44 U.S.C. 3501–21). This 

proposed rule has been submitted to the 

Office of Management and Budget for 

review and approval of the information 

collections. 

Type of submission, new or revision: 

Revision. 

The title of the information collection: 

Regulatory Improvements for 

Production and Utilization Facilities 

Transitioning to Decommissioning, 

Proposed Rule. 

The form number if applicable: Not 

applicable. 

How often the collection is required or 

requested: Annually and on occasion. 

Who will be required or asked to 

respond: Production and utilization 

facility licensees. 

An estimate of the number of annual 

responses: 102 (1 response for 10 CFR 

part 20, 0 responses for 10 CFR part 26, 

97 responses for 10 CFR part 50, 0 

responses for 10 CFR part 52, 1 response 

for 10 CFR part 72, and 3 responses for 

10 CFR part 73). 

The estimated number of annual 

respondents: 62 (1 respondent for 10 

CFR part 20, 0 respondents for 10 CFR 

part 26, 62 respondents for 10 CFR part 

50, 0 respondents for 10 CFR part 52, 20 

respondents for 10 CFR part 72, and 1 

respondent for 10 CFR part 73). 

An estimate of the total number of 

hours needed annually to comply with 

the information collection requirement 

or request: ¥3,658 (¥77.5 hours for 10 

CFR part 20, 0 hours for 10 CFR part 26, 

¥3,114.5 hours for 10 CFR part 50, 0 

hours for 10 CFR part 52, ¥436 hours 

for 10 CFR part 72, and ¥30 hours for 

10 CFR part 73). 

Abstract: The proposed rule would 

result in changes in recordkeeping and 

reporting burden relative to existing 

rules by creating a regulatory framework 

for production and utilization facility 

licensees transitioning to 

decommissioning and amending 

existing regulations that relate to the 

decommissioning of production and 

utilization facilities. Decommissioning 

nuclear power reactor licensees and the 

NRC have expended substantial 

resources processing licensing actions 

for nuclear power reactors during their 

transition period to decommissioning 

status. Licensees that are currently 

transitioning to decommissioning have 

been requesting NRC review and 

approval of licensing actions, informed 

by the low risk of an offsite radiological 

release posed by a decommissioning 

reactor. Specifically, the licensees are 

seeking NRC approval of exemptions 

and license amendments to revise 

requirements to reflect the reduced 

operations and risks posed by a 

permanently shutdown and defueled 

reactor. The proposed rule would, on 

balance, reduce the paperwork burden 

imposed on production and utilization 

facility licensees transitioning to 

decommissioning by establishing a 

graded approach to the requirements 

imposed on these facilities. A graded 

approach would adjust the level of 

analysis, documentation, and actions 

necessary to comply with safety 

requirements and criteria commensurate 

with several factors, including 

magnitude of any credible hazard 

involved, and the balance between 

radiological and non-radiological 

hazards as applicable to the level within 

the decommissioning process. The NRC 

expects that these proposed changes 

would enhance the efficiency of the 

decommissioning process and reduce 

the overall burden on licensees. 


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The NRC is seeking public comment 

on the potential impact of the 

information collections contained in 

this proposed rule and on the following 

issues: 

1. Is the proposed information 

collection necessary for the proper 

performance of the functions of the 

NRC, including whether the information 

will have practical utility? 

2. Is the estimate of the burden of the 

proposed information collection 

accurate? 

3. Is there a way to enhance the 

quality, utility, and clarity of the 

information to be collected? 

4. How can the burden of the 

proposed information collection on 

respondents be minimized, including 

the use of automated collection 

techniques or other forms of information 

technology? 

A copy of the OMB clearance package 

and proposed rule is available in 

ADAMS under Accession No. 

ML18039A192 or can be obtained free of 

charge by contacting the NRC’s Public 

Document Room reference staff at 1– 

800–397–4209, at 301–415–4737, or by 

email to PDR.Resource@nrc.gov. You 

may obtain information and comment 

submissions related to the OMB 

clearance package by searching on 

https://www.regulations.gov under 

Docket ID NRC–2015–0070. 

You may submit comments on any 

aspect of these proposed information 

collections, including suggestions for 

reducing the burden and on the above 

issues, by the following methods: 

• Federal rulemaking website: Go to 

https://www.regulations.gov and search 

for Docket ID NRC–2015–0070. 

• Mail comments to: FOIA, Library, 

and Information Collections Branch, 

Office of the Chief Information Officer, 

Mail Stop: T6–A10M, U.S. Nuclear 

Regulatory Commission, Washington, 

DC 20555–0001 or to the OMB reviewer 

at: OMB Office of Information and 

Regulatory Affairs (3150–0014, –0146, 

–0011, –0151, –0132, –0002), Attn: Desk 

Officer for the Nuclear Regulatory 

Commission, 725 17th Street NW, 

Washington, DC 20503; email: oira_

submission@omb.eop.gov. 

Submit comments by April 4, 2022. 

Comments received after this date will 

be considered if it is practical to do so, 

but the NRC staff is able to ensure 

consideration only for comments 

received on or before this date. 

Public Protection Notification 

The NRC may not conduct or sponsor, 

and a person is not required to respond 

to, a collection of information unless the 

document requesting or requiring the 

collection displays a currently valid 

OMB control number. 

XIV. Criminal Penalties 

For the purposes of Section 223 of the 

Atomic Energy Act of 1954, as amended 

(AEA), the NRC is issuing this proposed 

rule that would amend or add §§ 26.3, 

50.47, 50.54, 50.59, 50.71, 50.75, 50.82, 

50.200, 52.110, 72.30, 72.72, 72.212, 

72.218, 73.51, 73.54, 73.55, and 140.11 

as well as appendix G to 10 CFR part 20, 

appendix A to 10 CFR part 50, and 

appendix E to 10 CFR part 50, under 

one or more of Sections 161b, 161i, or 

161o of the AEA. Willful violations of 

these provisions would be subject to 

criminal enforcement. Criminal 

penalties as they apply to regulations in 

10 CFR parts 20, 26, 50, 52, 72, 73 and 

140 are discussed in §§ 20.2402, 26.825, 

50.111, 52.303, 72.86, 73.81 and 140.89. 

XV. Voluntary Consensus Standards 

The National Technology Transfer 

and Advancement Act of 1995, Public 

Law 104–113, requires that Federal 

agencies use technical standards that are 

developed or adopted by voluntary 

consensus standards bodies unless the 

use of such a standard is inconsistent 

with applicable law or otherwise 

impractical. In this proposed rule, the 

NRC would revise regulations 

associated with decommissioning in 10 

CFR parts 20, 26, 50, 51, 52, 72, 73, and 

140. This action would not constitute 

the establishment of a standard that 

contains generally applicable 

requirements. 

XVI. Availability of Guidance 

The NRC is issuing for comment four 

draft regulatory guides to support the 

implementation of the proposed 

requirements in this proposed rule, as 

well as to support other 

recommendations made in the 

supporting regulatory bases regarding 

areas where the decommissioning 

guidance could be improved or 

enhanced. You may access information 

and comment submissions related to the 

Draft Guides (DGs) by searching on 

https://www.regulations.gov under 

Docket ID NRC–2015–0070. You may 

submit comments on this draft guidance 

by the methods outlined in the 

ADDRESSES section of this document. 

1. The DG–1346, ‘‘Emergency 

Planning for Decommissioning Nuclear 

Power Reactors’’ (ADAMS Accession 

No. ML21347A046), is a new regulatory 

guide. 

2. The DG–1347, ‘‘Decommissioning 

of Nuclear Power Reactors,’’ (ADAMS 

Accession No. ML21347A080), would 

be Revision 2 to the existing Regulatory 

Guide 1.184. 

3. The DG–1348, ‘‘Assuring the 

Availability of Funds for 

Decommissioning Production or 

Utilization Facilities,’’ (ADAMS 

Accession No. ML21347A081), would 

be Revision 2 to the existing Regulatory 

Guide 1.159. 

4. The DG–1349, ‘‘Standard Format 

and Content for Post-Shutdown 

Decommissioning Activities Report,’’ 

(ADAMS Accession No. ML21347A138), 

would be Revision 2 to the existing 

Regulatory Guide 1.185. 

XVII. Public Meeting 

The NRC will conduct a public 

meeting on this proposed rule for the 

purpose of describing this proposed rule 

to the public and facilitating 

development of public comments on 

this proposed rule. 

The NRC will publish a notice of the 

location, time, and agenda of the 

meeting in the Federal Register, on 

Regulations.gov, and on the NRC’s 

public meeting website at least 10 

calendar days before the meeting. 

Stakeholders should monitor the NRC’s 

public meeting website for information 

about the public meeting at: https://

www.nrc.gov/public-involve/publicmeetings/index.cfm. 

XVIII. Availability of Documents 

The documents identified in the 

following table are available to 

interested persons through one or more 

of the following methods, as indicated. 

Document ADAMS accession No./web link/ 

Federal Register citation 

Proposed Rule Documents 

Draft Regulatory Analysis ............................................................................................................................. ML22019A132. 

Draft Environmental Assessment and FONSI .............................................................................................. ML22019A140. 

Draft Information Collection Analysis ............................................................................................................ ML18039A192. 


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Document ADAMS accession No./web link/ 

Federal Register citation 

Draft Regulatory Guidance Documents 

Draft Regulatory Guide DG–1346, ‘‘Emergency Planning for Decommissioning Nuclear Power Reactors’’.ML21347A046. 

Draft Regulatory Guide DG–1347, ‘‘Decommissioning of Nuclear Power Reactors’’ .................................. ML21347A080. 

Draft Regulatory Guide DG–1348, ‘‘Assuring the Availability of Funds for Decommissioning Production 

or Utilization Facilities’’.

ML21347A081. 

Draft Regulatory Guide DG–1349, ‘‘Standard Format and Content for Post-Shutdown Decommissioning 

Activities Report’’.

ML21347A138. 

Other References 

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6, 1988.

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Federal Register / Vol. 87, No. 42 / Thursday, March 3, 2022 / Proposed Rules 12321 

Document ADAMS accession No./web link/ 

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Throughout the development of this 

rule, the NRC may post documents 

related to this rule, including public 

comments, on the Federal rulemaking 

website at https://www.regulations.gov 

under Docket ID NRC–2015–0070. 

List of Subjects 

10 CFR Part 20 

Byproduct material, Criminal 

penalties, Hazardous waste, Licensed 

material, Nuclear energy, Nuclear 

materials, Nuclear power plants and 

reactors, Occupational safety and 

health, Packaging and containers, 

Penalties, Radiation protection, 

Reporting and recordkeeping 

requirements, Source material, Special 

nuclear material, Waste treatment and 

disposal. 

10 CFR Part 26 

Administrative practice and 

procedure, Alcohol abuse, Alcohol 

testing, Appeals, Chemical testing, Drug 

abuse, Drug testing, Employee 

assistance programs, Fitness for duty, 

Management actions, Nuclear power 

plants and reactors, Privacy, Protection 

of information, Radiation protection, 

Reporting and recordkeeping 

requirements. 

10 CFR Part 50 

Administrative practice and 

procedure, Antitrust, Backfitting, 

Classified information, Criminal 

penalties, Education, Emergency 

planning, Fire prevention, Fire 

protection, Incorporation by reference, 

Intergovernmental relations, Nuclear 

power plants and reactors, Penalties, 

Radiation protection, Reactor siting 

criteria, Reporting and recordkeeping 

requirements, Whistleblowing. 


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12322 Federal Register / Vol. 87, No. 42 / Thursday, March 3, 2022 / Proposed Rules 

10 CFR Part 51 

Administrative practice and 

procedure, Environmental impact 

statements, Hazardous waste, Nuclear 

energy, Nuclear materials, Nuclear 

power plants and reactors, Reporting 

and recordkeeping requirements. 

10 CFR Part 52 

Administrative practice and 

procedure, Antitrust, Combined license, 

Early site permit, Emergency planning, 

Fees, Incorporation by reference, 

Inspection, Issue finality, Limited work 

authorization, Nuclear power plants and 

reactors, Probabilistic risk assessment, 

Prototype, Reactor siting criteria, 

Redress of site, Penalties, Reporting and 

recordkeeping requirements, Standard 

design, Standard design certification. 

10 CFR Part 72 

Administrative practice and 

procedure, Hazardous waste, Indians, 

Intergovernmental relations, Nuclear 

energy, Penalties, Radiation protection, 

Reporting and recordkeeping 

requirements, Security measures, Spent 

fuel, Whistleblowing. 

10 CFR Part 73 

Criminal penalties, Exports, 

Hazardous materials transportation, 

Imports, Incorporation by reference, 

Nuclear energy, Nuclear materials, 

Nuclear power plants and reactors, 

Penalties, Reporting and recordkeeping 

requirements, Security measures. 

10 CFR Part 140 

Criminal penalties, Extraordinary 

nuclear occurrence, Insurance, 

Intergovernmental relations, Nuclear 

materials, Nuclear power plants and 

reactors, Penalties, Reporting and 

recordkeeping requirements. 

For the reasons set out in the 

preamble and under the authority of the 

Atomic Energy Act of 1954, as amended; 

the Energy Reorganization Act of 1974, 

as amended; and 5 U.S.C. 552 and 553, 

the NRC is proposing to amend 10 CFR 

parts 20, 26, 50, 51, 52, 72, 73, and 140 

as follows: 

PART 20—STANDARDS FOR 

PROTECTION AGAINST RADIATION 

■ 1. The authority citation for part 20 

continues to read as follows: 

Authority: Atomic Energy Act of 1954, 

secs. 11, 53, 63, 65, 81, 103, 104, 161, 170H, 

182, 186, 223, 234, 274, 1701 (42 U.S.C. 2014, 

2073, 2093, 2095, 2111, 2133, 2134, 2201, 

2210h, 2232, 2236, 2273, 2282, 2021, 2297f), 

Energy Reorganization Act of 1974, secs. 201, 

202 (42 U.S.C. 5841, 5842); Low-Level 

Radioactive Waste Policy Amendments Act 

of 1985, sec. 2 (42 U.S.C. 2021b); 44 U.S.C. 

3504 note. 

Appendix G to Part 20 [Amended] 

■ 2. In appendix G to part 20, amend 

paragraph E.1. of section III by: 

■ a. Removing the word ‘‘or’’ and 

adding in its place the word ‘‘of’’; and 

■ b. Removing the phrase ‘‘20 days’’ and 

adding in its place the phrase, ‘‘45 

days’’. 

PART 26—FITNESS FOR DUTY 

PROGRAMS 

■ 3. The authority citation for part 26 

continues to read as follows: 

Authority: Atomic Energy Act of 1954, 

secs. 53, 103, 104, 107, 161, 223, 234, 1701 

(42 U.S.C. 2073, 2133, 2134, 2137, 2201, 

2273, 2282, 2297f); Energy Reorganization 

Act of 1974, secs. 201, 202 (42 U.S.C. 5841, 

5842); 44 U.S.C. 3504 note. 

■ 4. Amend § 26.3, by revising 

paragraph (a) to read as follows: 

§ 26.3 Scope. 

(a)(1) Each holder of an operating 

license for a nuclear power reactor 

under part 50 of this chapter that 

receives the license after March 31, 

2008, and holders of a combined license 

under part 52 of this chapter after the 

Commission has made the finding under 

§ 52.103(g) of this chapter must 

implement the FFD program before the 

receipt of special nuclear material in the 

form of fuel assemblies. 

(2) Each holder of an operating license 

for a nuclear power reactor under part 

50 of this chapter and each holder of a 

combined license under part 52 of this 

chapter for which the Commission has 

made the finding under § 52.103(g) of 

this chapter must comply with the 

requirements of this part, except for 

subpart K of this part, until the NRC’s 

docketing of the license holder’s 

certifications required under 

§ 50.82(a)(1) of this chapter or 

§ 52.110(a) of this chapter. 

* * * * * 

§ 26.825 [Amended] 

■ 5. In § 26.825(b), remove ‘‘26.3’’. 

PART 50—DOMESTIC LICENSING OF 

PRODUCTION AND UTILIZATION 

FACILITIES 

■ 6. Revise the authority citation for part 

50 to read as follows: 

Authority: Atomic Energy Act of 1954, 

secs. 11, 53, 63, 81, 101, 102, 103, 104, 105, 

108, 122, 147, 149, 161, 181, 182, 183, 184, 

185, 186, 187, 189, 223, 234 (42 U.S.C. 2014, 

2073, 2093, 2113, 2131, 2132, 2133, 2134, 

2135, 2138, 2152, 2167, 2169, 2201, 2231, 

2232, 2233, 2234, 2235, 2236, 2237, 2239, 

2273, 2282); Energy Reorganization Act of 

1974, secs. 201, 202, 206, 211 (42 U.S.C. 

5841, 5842, 5846, 5851); Nuclear Waste 

Policy Act of 1982, sec. 306 (42 U.S.C. 

10226); National Environmental Policy Act of 

1969 (42 U.S.C. 4332); 44 U.S.C. 3504 note; 

Sec. 109, Pub. L. 96–295, 94 Stat. 783. 

■ 7. Revise § 50.1 to read as follows: 

§ 50.1 Basis, purpose, and procedures 

applicable. 

The regulations in this part are 

promulgated by the Nuclear Regulatory 

Commission pursuant to the Atomic 

Energy Act of 1954, as amended (68 

Stat. 919), and Title II of the Energy 

Reorganization Act of 1974 (88 Stat. 

1242), to provide for the licensing of 

production and utilization facilities 

through the termination of the 

associated 10 CFR part 50 licenses. This 

part also gives notice to all persons who 

knowingly provide to any licensee, 

applicant, contractor, or subcontractor, 

components, equipment, materials, or 

other goods or services, that relate to a 

licensee’s or applicant’s activities 

subject to this part, that they may be 

individually subject to NRC 

enforcement action for violation of 

§ 50.5. 

■ 8. In § 50.2, revise the definition for 

Certified fuel handler and add a 

definition for Non-power production or 

utilization facility in alphabetical order 

to read as follows: 

§ 50.2 Definitions. 

* * * * * 

Certified fuel handler means, for a 

nuclear power reactor facility, either 

(1) A non-licensed operator who has 

qualified in accordance with a fuel 

handler training program approved by 

the Commission; or 

(2) A non-licensed operator who 

meets the following criteria: 

(i) Has qualified in accordance with a 

fuel handler training program that meets 

the same requirements as training 

programs for non-licensed operators 

required by § 50.120, and 

(ii) Is responsible for decisions on: 

(A) Safe conduct of decommissioning 

activities; 

(B) Safe handling and storage of spent 

fuel; and 

(C) Appropriate response to plant 

emergencies. 

* * * * * 

Non-power production or utilization 

facility means a non-power reactor, 

testing facility, or other production or 

utilization facility, licensed under 

§ 50.21(a), § 50.21(c), or § 50.22, that is 

not a nuclear power reactor or fuel 

reprocessing plant. 

* * * * * 

§ 50.36 [Amended] 

■ 9. In § 50.36(c)(6), add ‘‘or § 52.110(a) 

of this chapter’’ after ‘‘§ 50.82(a)(1)’’. 

■ 10. Revise § 50.38 to read as follows: 


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Federal Register / Vol. 87, No. 42 / Thursday, March 3, 2022 / Proposed Rules 12323 

§ 50.38 Ineligibility of certain applicants. 

(a) Any person who is a citizen, 

national, or agent of a foreign country, 

or any corporation, or other entity 

which the Commission knows or has 

reason to believe is owned, controlled, 

or dominated by an alien, a foreign 

corporation, or a foreign government, 

shall be ineligible to apply for and 

obtain a license. 

(b) The prohibition of paragraph (a) of 

this section does not apply to a person, 

corporation, or other entity seeking a 

license for a facility that meets the 

criteria of § 50.82(a)(2)(ii), § 50.82(b)(6), 

or § 52.110(b)(2) of this chapter. 

§ 50.44 [Amended] 

■ 11. In § 50.44(b) introductory text, add 

‘‘or § 52.110(a) of this chapter’’ after 

‘‘§ 50.82(a)(1)’’. 

§ 50.46 [Amended] 

■ 12. In § 50.46(a)(1)(i), add ‘‘or 

§ 52.110(a) of this chapter’’ after 

‘‘§ 50.82(a)(1)’’. 

■ 13. In § 50.47, revise paragraph (b) 

introductory text and add paragraph (f) 

to read as follows: 

§ 50.47 Emergency plans. 

* * * * * 

(b) The onsite and, except as provided 

in paragraphs (d) and (f) of this section, 

offsite emergency response plans for 

nuclear power reactors must meet the 

following standards: 

* * * * * 

(f) The planning standards of 

paragraph (b) of this section do not 

apply to offsite radiological emergency 

response plans if the licensee’s 

emergency plan is not required to meet 

these planning standards or if the plume 

exposure pathway EPZ does not extend 

beyond the site boundary. 

§ 50.48 [Amended] 

■ 14. In § 50.48(f) introductory text, add 

‘‘or § 52.110(a) of this chapter’’ after 

‘‘§ 50.82(a)(1)’’. 

§ 50.49 [Amended] 

■ 15. In § 50.49(a), remove 

‘‘§ 52.110(a)(1)’’ and add in its place 

‘‘§ 52.110(a)’’. 

§ 50.51 [Amended] 

■ 16. In § 50.51, in paragraph (b) 

introductory text, remove the words ‘‘to 

authorize ownership and possession of 

the production or utilization facility,’’. 

■ 17. In § 50.54: 

■ a. Amend paragraph (m)(2)(i) by: 

■ i. Designating the table; 

■ ii. Revising the heading of the newly 

designated table; and 

■ iii. Revising footnote 2 to the table; 

■ b. In paragraph (o), remove 

‘‘52.110(a)(1)’’ and add in its place 

‘‘52.110(a)’’; 

■ c. Redesignate paragraphs (p)(3) and 

(4) as paragraphs (p)(5) and (6); 

■ d. Redesignate paragraphs (p)(1) and 

(2) as paragraphs (p)(2) and (3) and 

revise newly redesignated paragraphs 

(p)(2) and (3); 

■ e. Add new paragraph (p) 

introductory text and paragraphs (p)(1) 

and (4); 

■ f. Revise paragraphs (q)(1) 

introductory text and (q)(1)(iii) and 

(q)(2) and (3); 

■ g. Remove the words ‘‘after February 

21, 2012’’ wherever they appear in 

paragraphs (q)(4) and (5); and 

■ h. Add paragraphs (q)(7) and (8); 

■ i. Remove the words ‘‘after April 1, 

1981,’’ in paragraph (s)(2)(ii); 

■ j. In paragraph (s)(3), remove the 

words ‘‘The NRC’’ and add in their 

place the words ‘‘If the planning 

standards for radiological emergency 

preparedness apply to offsite 

radiological emergency response plans, 

the NRC’’; 

■ k. In paragraph (t)(1)(ii), remove the 

period from the second sentence and 

add in its place the word ‘‘or,’’; 

■ l. Add paragraphs (t)(1)(iii) and (t)(3); 

■ m. In paragraph (w) introductory text, 

remove the words ‘‘under this part’’; 

■ n. In paragraphs (w)(4)(ii) and (iii), 

add the words ‘‘or § 52.110 of this 

chapter’’ after the words ‘‘§ 50.82’’ 

wherever they appear; 

■ o. Add paragraphs (w)(5) and (6); 

■ p. In paragraph (y), add ‘‘or 

§ 52.110(a) of this chapter’’ after 

‘‘§ 50.82(a)(1)’’; and 

■ q. Revise paragraph (bb). 

The revisions and additions read as 

follows: 

§ 50.54 Conditions of licenses. 

* * * * * 

(m) * * * 

(2) * * * 

(i) * * * 

Table 1 to paragraph (m)(2)(i)— 

Minimum Requirements 1 Per Shift for 

On-Site Staffing of Nuclear Power Units 

by Operators and Senior Operators 

Licensed Under 10 CFR part 55 

* * * * * 1 Temporary deviations from the numbers 

required by this table shall be in accordance 

with criteria established in the unit’s 

technical specifications. 2 For the purpose of this table, a nuclear 

power unit is considered to be operating 

when it is in a mode other than cold 

shutdown or refueling as defined by the 

unit’s technical specifications. A Shift 

Technical Advisor is not required upon the 

NRC’s docketing of the license holder’s 

certifications required under § 50.82(a)(1) or 

§ 52.110(a) of this chapter. 

* * * * * 

(p) Security plans—(1) Definitions for 

the purpose of this paragraph, (p): 

(i) Change means an action that 

results in modification of, addition to, 

or removal from, the licensee’s security 

plans. All changes are subject to the 

provisions of this section except where 

the applicable regulations establish 

specific criteria for accomplishing a 

particular change. 

(ii) Decrease in safeguards 

effectiveness means a change or series of 

changes to an element or component of 

the security plans referenced in 

paragraph (p)(2) of this section that 

reduces or eliminates the licensee’s 

ability to perform or maintain the 

capabilities set forth in § 73.55(b)(3)(i) of 

this chapter without compensating 

changes to other security plan elements 

or components. 

(2) The licensee may not make a 

change which would decrease the 

effectiveness of a physical security plan, 

or guard training and qualification plan, 

or cyber security plan prepared under 

§ 50.34(c) or § 52.79(a) of this chapter, or 

part 73 of this chapter, or of the first 

four categories of information 

(Background, Generic Planning Base, 

Licensee Planning Base, Responsibility 

Matrix) contained in a licensee 

safeguards contingency plan prepared 

under § 50.34(d) or § 52.79(a) of this 

chapter, or part 73 of this chapter, as 

applicable, without prior approval of 

the Commission. A licensee desiring to 

make such a change shall submit an 

application for amendment to the 

licensee’s license under § 50.90. 

(3) The licensee may make changes to 

the security plans referenced in 

paragraph (p)(2) of this section, without 

prior Commission approval if the 

changes do not decrease the safeguards 

effectiveness of the plan. The licensee 

shall maintain records of changes to the 

plans made without prior Commission 

approval for a period of 3 years from the 

date of the change, and shall submit, as 

specified in § 50.4 or § 52.3 of this 

chapter, a report containing a 

description of each change within 2 

months after the change is made. The 

licensee shall include a summary of the 

analysis completed to determine that 

the change does not decrease the 

safeguards effectiveness of the plan. 

(4) The licensee shall prepare and 

maintain safeguards contingency plan 

procedures in accordance with 

appendix C of part 73 of this chapter for 

effecting the actions and decisions 

contained in the Responsibility Matrix 

of the safeguards contingency plan. 

Prior to the safeguards contingency plan 

being put into effect, the licensee shall 

have: 


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(i) All safeguards capabilities 

specified in the safeguards contingency 

plan available and functional; 

(ii) Detailed procedures developed 

according to appendix C to part 73 of 

this chapter available at the licensee’s 

site; and 

(iii) All appropriate personnel trained 

to respond to safeguards incidents as 

outlined in the plan and specified in the 

detailed procedures. 

* * * * * 

(q) Emergency plans—(1) Definitions 

for the purpose of this paragraph (q): 

* * * * * 

(iii) Emergency planning function 

means a capability or resource necessary 

to prepare for and respond to a 

radiological emergency. 

* * * * * 

(2) Except as provided in paragraph 

(q)(7) of this section, a holder of a 

license under this part, or a combined 

license under part 52 of this chapter 

after the Commission makes the finding 

under § 52.103(g) of this chapter, shall 

follow and maintain the effectiveness of 

an emergency plan that meets the 

requirements in appendix E to this part 

and, for nuclear power reactor licensees, 

the planning standards of § 50.47(b). 

(3) The licensee may make changes to 

its emergency plan without NRC 

approval only if the licensee performs 

and retains an analysis demonstrating 

that the changes do not reduce the 

effectiveness of the plan and the plan, 

as changed, continues to meet the 

applicable requirements in appendix E 

to this part and, for nuclear power 

reactor licensees, the planning 

standards of § 50.47(b), or the applicable 

requirements of § 50.200 or § 72.32 of 

this chapter. 

* * * * * 

(7) Upon the NRC’s docketing of the 

nuclear power reactor licensee’s 

certifications required under 

§ 50.82(a)(1) or § 52.110(a) of this 

chapter: 

(i) Licensees must follow and 

maintain the effectiveness of an 

emergency plan that meets the 

requirements of § 50.200(a) or paragraph 

(q)(2) of this section. 

(ii) If the fuel assembly with the 

highest burnup from the final offload 

that is transferred to the spent fuel pool 

has a burnup of less than or equal to 72 

gigawatt days per metric ton of heavy 

metal (GWd/MTHM) and has zirconium 

cladding, then after at least 10 months 

(for a boiling water reactor) or 16 

months (for a pressurized water reactor) 

have elapsed since the date of 

permanent cessation of operations, 

licensees must follow and maintain the 

effectiveness of an emergency plan that 

meets the planning standards of 

§ 50.200(b) and the requirements in 

§ 50.200(c) or paragraph (q)(7)(i) of this 

section. 

(A) In lieu of the 10- or 16-month 

spent fuel decay period in paragraph 

(q)(7)(ii) of this section, a licensee may 

submit under § 50.90 a request for NRC 

approval of an alternative spent fuel 

decay period. 

(B) If the fuel assembly with the 

highest burnup transferred to the spent 

fuel pool at the time of shutdown 

exceeds a burnup of 72 GWd/MTHM or 

does not have zirconium cladding, then 

the licensee must submit under § 50.90 

a request for NRC approval of an 

alternative spent fuel decay period. 

(C) In support of the request 

submitted in paragraph (q)(7)(ii)(A) or 

(B) of this section, the licensee must 

include an analysis demonstrating that 

the alternative spent fuel decay period 

ensures that the spent fuel would not 

heat up to 900 °C in less than 10 hours 

under adiabatic heatup conditions. 

(iii) When all the spent fuel is in dry 

cask storage, licensees must follow and 

maintain the effectiveness of an 

emergency plan that meets the 

standards in § 72.32(a)(1) through (16) of 

this chapter, or paragraph (q)(7)(ii) of 

this section. 

(iv) Licensees need not comply with 

the requirements of this section when 

all spent fuel has been removed from 

the site. 

(8) The following provisions apply to 

emergency plan changes to be 

implemented after the NRC’s docketing 

of the nuclear power reactor licensee’s 

certifications required under 

§ 50.82(a)(1) or § 52.110(a) of this 

chapter: 

(i) Initial plan changes made under 

paragraph (q)(3) of this section to 

comply with the requirements of 

§ 50.200 or § 72.32(a) of this chapter as 

permitted by paragraph (q)(7)(i), (ii), or 

(iii) of this section are not reductions in 

effectiveness of the plan and do not 

need to be submitted to the NRC for 

prior approval. These plan changes 

must be submitted to the NRC at least 

60 days prior to implementation, as 

specified in § 50.4. Subsequent plan 

changes must be made under paragraph 

(q)(3) or (4) of this section, or licensees 

may follow the change process under 

§ 72.44(f) of this chapter if the 

emergency plan meets the requirements 

in § 72.32(a) of this chapter. 

(ii) For structures, systems, and 

components that are no longer needed 

to provide support for an emergency 

planning function as defined in 

paragraph (q)(1)(iii) of this section, 

licensees may make a determination 

under paragraph (q)(3) of this section 

that changes to the emergency plan 

related to these structures, systems, and 

components are not reductions in 

effectiveness if the Final Safety Analysis 

Report demonstrates that these 

structures, systems, and components are 

no longer required to be in service due 

to the decommissioning status of the 

facility. 

(iii) Changes to emergency action 

levels based on plant conditions that are 

not physically achievable or 

instrumentation that is no longer in 

service due to the decommissioning 

status of the facility, are not reductions 

in effectiveness provided that the 

evaluation under paragraph (q)(3) of this 

section demonstrates that these changes 

do not reduce the capability of the 

emergency plan to take timely and 

appropriate protective actions. 

* * * * * 

(t) * * * 

(1) * * * 

(iii) At intervals not to exceed 24 

months after the first required element 

review following transition to an 

emergency plan that meets the 

requirements of § 50.200(b). 

* * * * * 

(3) The review of the emergency 

preparedness program elements is no 

longer required once all fuel is in dry 

cask storage. 

* * * * * 

(w) * * * 

(5) Each power reactor licensee for a 

production or utilization facility of the 

type described in § 50.21(b) or § 50.22 

shall have and maintain financial 

protection in an amount of at least 

$50,000,000 for each reactor station site: 

(i) For which the NRC has docketed 

the certifications required under 

§ 50.82(a)(1) or § 52.110(a) of this 

chapter; and 

(ii) For which at least 10 months (for 

a boiling water reactor) or 16 months 

(for a pressurized water reactor) have 

elapsed since the date of permanent 

cessation of operations if the fuel meets 

the criteria of § 50.54(q)(7)(ii), or for 

which an NRC-approved alternative to 

the 10- or 16-month spent fuel decay 

period, submitted under 

§ 50.54(q)(7)(ii)(A) or (B), has elapsed. 

(6) The licensee shall promptly notify 

the Commission of any material change 

in the insurance or other financial 

security information reported to the 

Commission under paragraph (w)(3) of 

this section. 

* * * * * 

(bb) Irradiated Fuel Management Plan 

(1) Prior to or within 2 years following 

permanent cessation of operations, the 

licensee must submit an irradiated fuel 

management plan (IFMP) to the NRC as 


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an application for an amendment to its 

license. Licensees may not start to 

decommission structures, systems, and 

components needed for moving, 

unloading, and shipping the irradiated 

fuel until after the NRC approves the 

IFMP. 

(2) The IFMP must contain a 

discussion of the licensee’s planned 

actions for managing irradiated fuel and 

how those actions will be consistent 

with NRC requirements for licensed 

possession of irradiated fuel until title 

to, and possession of, the irradiated fuel 

is transferred to the Secretary of Energy. 

(3) If any planned actions for 

managing irradiated fuel would require 

exemptions from applicable regulations 

or amendments to the licensee’s license 

issued under this part or part 52 or 72 

of this chapter or the certificate of 

compliance issued under part 72 of this 

chapter being used by the licensee, then 

the licensee shall identify them in the 

IFMP and state that these requests have 

been or will be made to the NRC. 

(4) The IFMP must contain the 

projected cost of managing irradiated 

fuel and discuss how the licensee will 

provide funding for the management of 

the irradiated fuel following permanent 

cessation of operations until title to, and 

possession of, the irradiated fuel is 

transferred to the Secretary of Energy. 

(5) Licensees shall submit to the NRC 

any changes to the IFMP as an 

application for an amendment to its 

license. 

(6) The licensee shall retain a copy of 

the IFMP as a record until termination 

of the operating license issued under 

this part or combined license issued 

under part 52 of this chapter. 

* * * * * 

■ 18. In § 50.59: 

■ a. In paragraph (b), remove ‘‘§ 50.110’’ 

and add in its place ‘‘§ 52.110(a) of this 

chapter’’; and 

■ b. Revise paragraph (d)(3) to read as 

follows: 

§ 50.59 Changes, tests and experiments. 

* * * * * 

(d) * * * 

(3) Except as specified in § 50.71(c)(2), 

the records of changes in the facility 

must be maintained until the 

termination of an operating license 

issued under this part, a combined 

license issued under part 52 of this 

chapter, or a renewed license issued 

under part 54 of this chapter. Records of 

changes in procedures and records of 

tests and experiments must be 

maintained for a period of 5 years. 

§ 50.60 [Amended] 

■ 19. In § 50.60(a), add ‘‘or § 52.110(a) of 

this chapter’’ after ‘‘§ 50.82(a)(1)’’. 

§ 50.61 [Amended] 

■ 20. In § 50.61(b)(1), add ‘‘or 

§ 52.110(a) of this chapter’’ after 

‘‘§ 50.82(a)(1)’’. 

§ 50.62 [Amended] 

■ 21. In § 50.62(a), add ‘‘or § 52.110(a) of 

this chapter’’ after ‘‘§ 50.82(a)(1)’’. 

§ 50.65 [Amended] 

■ 22. In § 50.65(a)(1), remove 

‘‘52.110(a)(1)’’ and add in its place 

‘‘52.110(a)’’. 

■ 23. In § 50.71, revise paragraphs (c) 

and (e)(4) to read as follows: 

§ 50.71 Maintenance of records, making of 

reports. 

* * * * * 

(c)(1) Records that are required by the 

regulations in this part or part 52 of this 

chapter, by license condition, or by 

technical specifications must be 

retained for the period specified by the 

appropriate regulation, license 

condition, or technical specification. If 

a retention period is not otherwise 

specified, these records must be 

retained until the Commission 

terminates the facility license, except as 

specified in paragraph (c)(2) of this 

section, or, in the case of an early site 

permit, until the permit expires. 

(2) Licensees for which the NRC has 

docketed the certifications required 

under § 50.82(a)(1) or § 52.110(a) of this 

chapter are not required to retain 

records associated with structures, 

systems, and components that have 

been permanently removed from service 

under the NRC license using an NRCapproved change process. Licensees 

shall continue to retain records as 

specified under § 50.75(g). 

* * * * * 

(e) * * * 

(4) Subsequent revisions must be filed 

annually or 6 months after each 

refueling outage provided the interval 

between successive updates does not 

exceed 24 months. The revisions must 

reflect all changes up to a maximum of 

6 months prior to the date of filling. For 

nuclear power reactor facilities that 

have submitted the certifications 

required by § 50.82(a)(1) or § 52.110(a), 

subsequent revisions must be filed every 

24 months. 

* * * * * 

■ 24. In § 50.75: 

■ a. Revise the first sentence in 

paragraph (a); 

■ b. Revise paragraphs (b)(1), (3), and (4) 

and add paragraph (b)(5); 

■ c. Revise paragraph (e)(1) introductory 

text; 

■ d. In paragraph (e)(1)(i): 

■ i. Remove the phrase ‘‘formulas in 

§ 50.75(c)’’ and add in its place the 

phrase ‘‘table of minimum amounts in 

paragraph (c)’’; 

■ ii. Remove the phrase ‘‘site-specific 

estimate’’ wherever it appears and add 

in its place the phrase ‘‘site-specific 

decommissioning cost estimate’’; 

■ e. In paragraph (e)(1)(ii) introductory 

text, remove the phrase ‘‘site-specific 

estimate’’ wherever it appears and add 

in its place the phrase ‘‘site-specific 

decommissioning cost estimate’’; 

■ f. In paragraph (e)(1)(ii)(B), add ‘‘or 

§ 52.110 of this chapter’’ after ‘‘50.82 of 

this part’’; 

■ g. In paragraph (e)(1)(v), add ‘‘or 

§ 52.110 of this chapter’’ after ‘‘or 

§ 50.82’’; 

■ h. Amend paragraph (f) by: 

■ i. Revising paragraph (f)(1); 

■ ii. Removing paragraph (f)(2); 

■ iii. Redesignating paragraphs (f)(3) 

through (5) as (f)(2) through (4); and 

■ iv. Revising newly redesignated 

paragraph (f)(2) and paragraph (f)(3) 

introductory text; 

■ i. In paragraphs (h)(1)(iii) and (iv), 

remove the words ‘‘Director, Office of 

Nuclear Reactor Regulation, or Director, 

Office of Nuclear Material Safety and 

Safeguards, as applicable,’’ wherever 

they appear in the first sentence of each 

paragraph and add in their place the 

words, ‘‘Document Control Desk as 

specified in § 50.4’’. 

■ j. In paragraph (h)(1)(iv), add ‘‘or 

§ 52.110(h) of this chapter’’ after 

‘‘§ 50.82(a)(8)’’ wherever it appears. 

■ k. In paragraph (h)(2), remove the 

words ‘‘given the Director, Office of 

Nuclear Reactor Regulation, or Director, 

Office of Nuclear Material Safety and 

Safeguards, as applicable,’’ wherever 

they appear and add in their place, the 

words, ‘‘given to the Document Control 

Desk as specified in § 50.4’’. 

■ l. In paragraph (h)(2), add ‘‘or 

§ 52.110(h) of this chapter’’ after 

‘‘§ 50.82(a)(8)’’ wherever it appears. 

The revisions and addition read as 

follows: 

§ 50.75 Reporting and recordkeeping for 

decommissioning planning. 

(a) This section establishes 

requirements for indicating to NRC how 

a licensee will provide reasonable 

assurance that funds will be available to 

decommission the facility, as defined in 

§ 50.2. * * * 

* * * * * 

(b) * * * 

(1) For an applicant for or holder of 

an operating license under this part, the 

report must contain a certification that 

reasonable assurance that funds will be 

available to decommission will be (for a 

license applicant), or has been (for a 

license holder), provided in an amount 

which may be more, but not less, than 


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the amount stated in the table of 

minimum amounts in paragraph (c)(1) 

of this section, adjusted using a rate at 

least equal to that stated in paragraph 

(c)(2) of this section. For an applicant 

for a combined license under subpart C 

of part 52 of this chapter, the report 

must contain a certification that 

reasonable assurance of funds to 

decommission will be provided no later 

than 30 days after the Commission 

publishes notice in the Federal Register 

under § 52.103(a) of this chapter in an 

amount which may be more, but not 

less, than the amount stated in the table 

of minimum amounts in paragraph 

(c)(1) of this section, adjusted using a 

rate at least equal to that stated in 

paragraph (c)(2) of this section. 

* * * * * 

(3) The amount must be covered by 

one or more of the methods described in 

paragraph (e) of this section. 

(4) The amount stated in the 

applicant’s or licensee’s certification 

may be based on a site-specific 

decommissioning cost estimate for 

decommissioning the facility. The sitespecific decommissioning cost estimate 

may be more, but not less, than the 

amount stated in the table of minimum 

amounts in paragraph (c)(1) of this 

section, adjusted using a rate at least 

equal to that stated in paragraph (c)(2) 

of this section. 

(5) As part of the certification, a copy 

of the financial instrument obtained to 

satisfy the requirements of paragraph (e) 

of this section must be submitted to 

NRC; provided, however, that an 

applicant for or holder of a combined 

license need not obtain such financial 

instrument or submit a copy to the 

Commission except as provided in 

paragraph (e)(3) of this section. 

* * * * * 

(e)(1) Reasonable assurance of funds 

to decommission is to be provided by 

the following methods: 

* * * * * 

(f)(1) Each power reactor licensee 

shall report, on a calendar-year basis, to 

the NRC by March 31, 2023, and at least 

once every 3 years thereafter on the 

status of its decommissioning funding 

provided by the financial assurance 

methods described in paragraph (e)(1) of 

this section for each reactor or part of 

a reactor that it owns. However, each 

holder of a combined license under part 

52 of this chapter need not begin 

reporting until the date that the 

Commission has made the finding under 

§ 52.103(g) of this chapter. The 

information in this report must include, 

at a minimum, the amount of 

decommissioning funds estimated to be 

required pursuant to paragraphs (b) and 

(c) of this section; the amount of 

decommissioning funds accumulated to 

the end of the calendar year preceding 

the date of the report; a schedule of the 

annual amounts remaining to be 

collected; the assumptions used 

regarding rates of escalation in 

decommissioning costs, rates of 

earnings on decommissioning funds, 

and rates of other factors used in 

funding projections; any contracts upon 

which the licensee is relying pursuant 

to paragraph (e)(1)(v) of this section; any 

modifications occurring to a licensee’s 

current method of providing financial 

assurance since the last submitted 

report; and any material changes to trust 

agreements. If any of the preceding 

items is not applicable, the licensee 

should so state in its report. If the 

projected balance of any 

decommissioning funds does not cover 

the estimated cost of decommissioning, 

the licensee must include additional 

financial assurance to cover the shortfall 

by the time the next report is due. Once 

a licensee has determined that it is 

within 5 years of permanent cessation of 

operations, or if it is involved in a 

merger or an acquisition, it shall submit 

this report annually. Once the plant has 

permanently ceased operations, the 

reporting requirements of 

§ 50.82(a)(8)(v) (for 10 CFR part 50 

licensees) or § 52.110(h)(5) of this 

chapter (for 10 CFR part 52 licensees) 

shall apply. 

(2) Each power reactor licensee shall 

at or about 5 years prior to the projected 

end of operations submit a preliminary 

site-specific decommissioning cost 

estimate which includes an up-to-date 

assessment of the major factors that 

could affect the cost to decommission. 

(3) Each non-power reactor licensee 

shall at or about 2 years prior to the 

projected end of operations submit a 

preliminary decommissioning plan 

containing a site-specific 

decommissioning cost estimate and an 

up-to-date assessment of the major 

factors that could affect planning for 

decommissioning. Factors to be 

considered in submitting this 

preliminary decommissioning plan 

information include— 

* * * * * 

■ 25. In § 50.82: 

■ a. Revise paragraphs (a)(2), (a)(4), 

(a)(6)(ii), (a)(8)(i)(A), (a)(8)(ii), (a)(8)(v) 

introductory text and (a)(8)(vii) 

introductory text; 

■ b. Revise paragraphs (a)(9) 

introductory text and (a)(9)(ii)(F); 

■ c. Revise paragraph (b) introductory 

text; and 

■ d. Redesignate paragraph (b)(6) as 

(b)(8) and add new paragraphs (b)(6) 

and (7). 

The revisions and additions read as 

follows: 

§ 50.82 Termination of license. 

* * * * * 

(a) * * * 

(2)(i) Upon the NRC’s docketing of the 

licensee’s certifications required under 

paragraph (a)(1) of this section, or when 

a final legally effective order to 

permanently cease operations has come 

into effect, the 10 CFR part 50 license 

no longer authorizes operation of the 

reactor or emplacement or retention of 

fuel into the reactor vessel. 

(ii) The facility licensed under this 

part is no longer a utilization facility 

once the licensee meets the criteria of 

paragraph (a)(2)(i) of this section and 

modifies the facility to be incapable of 

making use of special nuclear material 

without significant facility alterations 

necessary to restore the capability to 

make use of special nuclear material. 

The NRC maintains the authority to 

regulate the 10 CFR part 50 license with 

respect to the possession of special 

nuclear material, source material, and 

byproduct material under sections 53, 

63, 81, and 161 of the Act, as applicable. 

Until the termination of the 10 CFR part 

50 license under paragraph (a)(11) of 

this section, the regulations of this 

chapter applicable to a utilization 

facility continue to apply to the holder 

of the license unless the regulations 

explicitly state otherwise. 

* * * * * 

(4)(i) Prior to or within 2 years 

following permanent cessation of 

operations, the licensee shall submit a 

post-shutdown decommissioning 

activities report (PSDAR) to the NRC, 

and a copy to the affected State(s). The 

PSDAR must contain a description of 

the planned decommissioning activities 

along with a schedule for their 

accomplishment, a discussion whether 

the environmental impacts associated 

with site-specific decommissioning 

activities will be bounded by 

appropriate federally issued 

environmental review documents, a 

description of any decommissioning 

activities whose environmental impacts 

will not be so bounded and will be 

evaluated prior to the performance of 

the activities, and a site-specific 

decommissioning cost estimate, 

including the projected cost of 

managing irradiated fuel. 

(ii) The NRC shall publish a notice in 

the Federal Register acknowledging the 

receipt of the PSDAR and the 

availability for public comment of the 

PSDAR. The NRC shall also schedule a 

public meeting in the vicinity of the 

licensee’s facility upon receipt of the 

PSDAR. The NRC shall include a notice 


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in a forum, such as local newspapers, 

that is readily accessible to individuals 

in the vicinity of the site, and in the 

Federal Register notice required by this 

paragraph, announcing the date, time 

and location of the meeting, along with 

a brief description of the purpose of the 

meeting. 

* * * * * 

(6) * * * 

(ii) Result in significant 

environmental impacts not bounded by 

appropriate federally issued 

environmental review documents; or 

* * * * * 

(8) * * * 

(i) * * * 

(A) The withdrawals are for expenses 

for activities consistent with the 

definition of decommission in § 50.2; 

* * * * * 

(ii) Initially, 3 percent of the generic 

amount specified in § 50.75(c) may be 

used for decommissioning planning. For 

licensees that have submitted the 

certifications required under 

§ 50.82(a)(1) and commencing 90 days 

after the NRC has received the PSDAR, 

an additional 20 percent may be used. 

A site-specific decommissioning cost 

estimate must be submitted to the NRC 

prior to the licensee using any funding 

in excess of these amounts. 

* * * * * 

(v) After submitting its site-specific 

decommissioning cost estimate required 

by paragraph (a)(4)(i) of this section, and 

until the licensee has completed its final 

radiation survey and demonstrated that 

residual radioactivity has been reduced 

to a level that permits termination of its 

license, the licensee must annually 

submit to the NRC, by March 31, a 

financial assurance status report. The 

report may combine the reporting 

requirements of § 72.30 of this chapter 

and § 50.82(a)(8)(vii). The report must 

include the following information, 

current through the end of the previous 

calendar year: 

* * * * * 

(vii) After submitting its site-specific 

decommissioning cost estimate required 

by paragraph (a)(4)(i) of this section, if 

spent fuel is on site, the licensee must 

annually submit to the NRC, by March 

31, a report on the status of its funding 

for managing irradiated fuel. The report 

must include the following information, 

current through the end of the previous 

calendar year: 

* * * * * 

(9) All power reactor licensees that 

have loaded fuel into the reactor must 

submit an application for termination of 

license. The application for termination 

of license must be accompanied or 

preceded by a license termination plan 

to be submitted for NRC approval. 

* * * * * 

(ii) * * * 

(F) An updated site-specific estimate 

of remaining decommissioning costs 

and identification of sources of funds 

for license termination, spent fuel 

management, and ISFSI 

decommissioning, as applicable; 

* * * * * 

(b) For non-power production or 

utilization facilities and fuel 

reprocessing plants— 

* * * * * 

(6) The facility licensed under this 

part is no longer a production or 

utilization facility once the following 

criteria are met: 

(i) The NRC removes the licensee’s 

authority to operate the facility through 

a license amendment; and 

(ii) The licensee modifies the facility 

to be incapable of the production of 

special nuclear material, separation of 

the isotopes of plutonium, processing of 

irradiated materials containing special 

nuclear material, or making use of 

special nuclear material, without 

significant facility alterations necessary 

to restore the capability to produce 

special nuclear material, separate the 

isotopes of plutonium, process 

irradiated materials containing special 

nuclear material, or make use of special 

nuclear material. 

(7) For a facility licensed under this 

part that is no longer a production or 

utilization facility under paragraph 

(b)(6) of this section, the NRC maintains 

the authority to regulate the 10 CFR part 

50 license with respect to the possession 

of special nuclear material, source 

material, and byproduct material under 

sections 53, 63, 81, and 161 of the Act, 

as applicable. Until the termination of 

the 10 CFR part 50 license under 

paragraph (b)(8) of this section, the 

regulations of this chapter applicable to 

a non-power production or utilization 

facility or fuel reprocessing plant 

continue to apply to the holder of the 

license unless the regulations explicitly 

state otherwise. 

* * * * * 

■ 26. Revise § 50.109 to read as follows: 

§ 50.109 Backfitting. 

(a) Backfitting for nuclear power 

reactor licensees prior to 

decommissioning. (1)(i) Definition. 

Backfitting is defined as the 

modification of or addition to systems, 

structures, components, or design of a 

facility; or the design approval or 

manufacturing license for a facility; or 

the procedures or organization required 

to design, construct or operate a facility; 

any of which may result from a new or 

amended provision in the Commission’s 

regulations or the imposition of a 

regulatory staff position interpreting the 

Commission’s regulations that is either 

new or different from a previously 

applicable staff position after: 

(A) The date of issuance of the 

construction permit for the facility for 

facilities having construction permits 

issued after October 21, 1985; 

(B) Six (6) months before the date of 

docketing of the operating license 

application for the facility for facilities 

having construction permits issued 

before October 21, 1985; 

(C) The date of issuance of the 

operating license for the facility for 

facilities having operating licenses; 

(D) The date of issuance of the design 

approval under subpart E of part 52 of 

this chapter; 

(E) The date of issuance of a 

manufacturing license under subpart F 

of part 52 of this chapter; 

(F) The date of issuance of the first 

construction permit issued for a 

duplicate design under appendix N to 

this part; or 

(G) The date of issuance of a 

combined license under subpart C of 

part 52 of this chapter, provided that if 

the combined license references an early 

site permit, the provisions in § 52.39 of 

this chapter apply with respect to the 

site characteristics, design parameters, 

and terms and conditions specified in 

the early site permit. If the combined 

license references a standard design 

certification rule under subpart B of 10 

CFR part 52, the provisions in § 52.63 of 

this chapter apply with respect to the 

design matters resolved in the standard 

design certification rule, provided 

however, that if any specific backfitting 

limitations are included in a referenced 

design certification rule, those 

limitations shall govern. If the combined 

license references a standard design 

approval under subpart E of 10 CFR part 

52, the provisions in § 52.145 of this 

chapter apply with respect to the design 

matters resolved in the standard design 

approval. If the combined license uses 

a reactor manufactured under a 

manufacturing license under subpart F 

of 10 CFR part 52, the provisions of 

§ 52.171 of this chapter apply with 

respect to matters resolved in the 

manufacturing license proceeding. 

(ii) Proposed backfitting. Except as 

provided in paragraph (a)(1)(iv) of this 

section, the Commission shall require a 

systematic and documented analysis 

pursuant to paragraph (a)(2) of this 

section for backfits which it seeks to 

impose. 

(iii) Backfit analysis. Except as 

provided in paragraph (a)(1)(iv) of this 


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section, the Commission shall require 

the backfitting of a facility only when it 

determines, based on the analysis 

described in paragraph (a)(2) of this 

section, that there is a substantial 

increase in the overall protection of the 

public health and safety or the common 

defense and security to be derived from 

the backfit and that the direct and 

indirect costs of implementation for that 

facility are justified in view of this 

increased protection. 

(iv) Exceptions. The provisions of 

paragraphs (a)(1)(ii) and (iii) of this 

section are inapplicable and, therefore, 

backfit analysis is not required and the 

standards in paragraph (a)(1)(iii) of this 

section do not apply where the 

Commission or staff, as appropriate, 

finds and declares, with appropriated 

documented evaluation for its finding, 

either: 

(A) That a modification is necessary 

to bring a facility into compliance with 

a license or the rules or orders of the 

Commission, or into conformance with 

written commitments by the licensee; or 

(B) That regulatory action is necessary 

to ensure that the facility provides 

adequate protection to the health and 

safety of the public and is in accord 

with the common defense and security; 

or 

(C) That the regulatory action involves 

defining or redefining what level of 

protection to the public health and 

safety or common defense and security 

should be regarded as adequate. 

(v) Mandatory backfitting. The 

Commission shall always require the 

backfitting of a facility if it determines 

that such regulatory action is necessary 

to ensure that the facility provides 

adequate protection to the health and 

safety of the public and is in accord 

with the common defense and security. 

(vi) Documented evaluation. The 

documented evaluation required by 

paragraph (a)(1)(iv) of this section shall 

include a statement of the objectives of 

and reasons for the modification and the 

basis for invoking the exception. If 

immediately effective regulatory action 

is required, then the documented 

evaluation may follow rather than 

precede the regulatory action. The 

documented evaluation required by 

paragraph (a)(1)(iv)(A) of this section 

must include a consideration of the 

costs of imposing the modification. 

(vii) Implementation. If there are two 

or more ways to achieve compliance 

with a license or the rules or orders of 

the Commission, or with written 

licensee commitments, or there are two 

or more ways to reach a level of 

protection which is adequate, then 

ordinarily the applicant or licensee is 

free to choose the way which best suits 

its purposes. However, should it be 

necessary or appropriate for the 

Commission to prescribe a specific way 

to comply with its requirements or to 

achieve adequate protection, then cost 

may be a factor in selecting the way, 

provided that the objective of 

compliance or adequate protection is 

met. 

(2) Backfit analysis factors. In 

reaching the determination required by 

paragraph (a)(1)(iii) of this section, the 

Commission will consider how the 

backfit should be scheduled in light of 

other ongoing regulatory activities at the 

facility and, in addition, will consider 

information available concerning any of 

the following factors as may be 

appropriate and any other information 

relevant and material to the proposed 

backfit: 

(i) Statement of the specific objectives 

that the proposed backfit is designed to 

achieve; 

(ii) General description of the activity 

that would be required by the licensee 

or applicant in order to complete the 

backfit; 

(iii) Potential change in the risk to the 

public from the accidental off-site 

release of radioactive material; 

(iv) Potential impact on radiological 

exposure of facility employees; 

(v) Installation and continuing costs 

associated with the backfit, including 

the cost of facility downtime or the cost 

of construction delay; 

(vi) The potential safety impact of 

changes in plant or operational 

complexity, including the relationship 

to proposed and existing regulatory 

requirements; 

(vii) The estimated resource burden 

on the NRC associated with the 

proposed backfit and the availability of 

such resources; 

(viii) The potential impact of 

differences in facility type, design or age 

on the relevancy and practicality of the 

proposed backfit; 

(ix) Whether the proposed backfit is 

interim or final and, if interim, the 

justification for imposing the proposed 

backfit on an interim basis. 

(3) Impact on licensing actions. No 

licensing action will be withheld during 

the pendency of backfit analyses 

required by the Commission’s rules. 

(b) Backfitting for decommissioning 

nuclear power reactor licensees. 

(1) Definition. Backfitting is defined 

as the modification of or addition to 

systems, structures, or components in 

use after permanent cessation of 

operations and certification of 

permanent removal of fuel from the 

reactor vessel has been docketed as 

required under § 50.82(a)(1) or 

§ 52.110(a) of this chapter, or the design 

of the licensee’s facility, or the 

procedures or organization required to 

decommission the facility, any of which 

may result from a new or amended 

provision in the Commission rules or 

the imposition of a regulatory staff 

position interpreting the Commission 

rules that is either new or different from 

a previously applicable staff position, 

after the date of issuance of the 

operating license issued under this part 

or combined license issued under 

subpart C of part 52 of this chapter. 

(2) Proposed backfits. Except as 

provided in paragraph (b)(4) of this 

section, the Commission shall require a 

systematic and documented analysis 

pursuant to paragraph (b)(8) of this 

section for backfits that it seeks to 

impose. 

(3) Backfit analysis. Except as 

provided in paragraph (b)(4) of this 

section, the Commission shall require 

the backfitting of a facility only when it 

determines, based on the analysis 

described in paragraph (b)(8) of this 

section, that there is a substantial 

increase in the overall protection of the 

public health and safety or the common 

defense and security to be derived from 

the backfit and that the direct and 

indirect costs of implementation for that 

facility are justified in view of this 

increased protection. 

(4) Exceptions. The provisions of 

paragraphs (b)(2) and (3) of this section 

are inapplicable and, therefore, backfit 

analysis is not required and the 

standards in paragraph (b)(3) of this 

section do not apply where the 

Commission or staff, as appropriate, 

finds and declares, with appropriated 

documented evaluation for its finding, 

either: 

(i) That a modification is necessary to 

bring a facility into compliance with a 

license or the rules or orders of the 

Commission, or into conformance with 

written commitments by the licensee; 

(ii) That regulatory action is necessary 

to ensure that the facility provides 

adequate protection to the health and 

safety of the public and is in accord 

with the common defense and security; 

or 

(iii) That the regulatory action 

involves defining or redefining what 

level of protection to the public health 

and safety or common defense and 

security should be regarded as adequate. 

(5) Mandatory backfitting. The 

Commission shall always require the 

backfitting of a facility if it determines 

that such regulatory action is necessary 

to ensure that the facility provides 

adequate protection to the health and 

safety of the public and is in accord 

with the common defense and security. 


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(6) Documented evaluation. The 

documented evaluation required by 

paragraph (b)(4) of this section shall 

include a statement of the objectives of 

and reasons for the modification and the 

basis for invoking the exception. If 

immediately effective regulatory action 

is required, then the documented 

evaluation may follow rather than 

precede the regulatory action. The 

documented evaluation required by 

paragraph (b)(4)(i) of this section must 

include a consideration of the costs of 

imposing the modification. 

(7) Implementation. If there are two or 

more ways to achieve compliance with 

a license or the rules or orders of the 

Commission, or with written licensee 

commitments, or there are two or more 

ways to reach a level of protection that 

is adequate, then ordinarily the licensee 

is free to choose the way that best suits 

its purposes. However, should it be 

necessary or appropriate for the 

Commission to prescribe a specific way 

to comply with its requirements or to 

achieve adequate protection, then cost 

may be a factor in selecting the way, 

provided that the objective of 

compliance or adequate protection is 

met. 

(8) Backfit analysis factors. In 

reaching the determination required by 

paragraph (b)(3) of this section, the 

Commission will consider how the 

backfit should be scheduled in light of 

other ongoing regulatory activities at the 

facility and, in addition, will consider 

information available concerning any of 

the following factors as may be 

appropriate and any other information 

relevant and material to the proposed 

backfit: 

(i) Statement of the specific objectives 

that the proposed backfit is designed to 

achieve; 

(ii) General description of the activity 

that would be required by the licensee 

in order to complete the backfit; 

(iii) Potential change in the risk to the 

public from the accidental off-site 

release of radioactive material; 

(iv) Potential impact on radiological 

exposure of facility employees; 

(v) Installation and continuing costs 

associated with the backfit, including 

the cost of decommissioning delay; 

(vi) The potential safety impact of 

changes in major decommissioning 

activities, including the relationship to 

proposed and existing regulatory 

requirements; 

(vii) The estimated resource burden 

on the NRC associated with the 

proposed backfit and the availability of 

such resources; 

(viii) The potential impact of 

differences in facility type and the 

percentage of decommissioning 

completed on the relevancy and 

practicality of the proposed backfit; and 

(ix) Whether the proposed backfit is 

interim or final and, if interim, the 

justification for imposing the proposed 

backfit on an interim basis. 

(9) Impact on licensing actions. No 

licensing action will be withheld during 

the pendency of backfit analyses 

required by the Commission’s rules. 

(c) Responsibility for implementation. 

The Executive Director for Operations 

shall be responsible for implementation 

of this section, and all analyses required 

by this section shall be approved by the 

Executive Director for Operations or his 

designee. 

■ 27. In § 50.155, add paragraphs (h)(6), 

(7), and (8) to read as follows: 

§ 50.155 Mitigation of beyond-design-basis 

events. 

* * * * * 

(h) * * * 

(6) On [EFFECTIVE DATE OF THE 

FINAL RULE], Order EA–06–137, 

‘‘Order Modifying Licenses,’’ is 

rescinded for each licensee that was 

issued Order EA–06–137. 

(7) On [EFFECTIVE DATE OF THE 

FINAL RULE], the Mitigation Strategies 

License Condition is deemed removed 

from the power reactor license of each 

licensee subject to this section. 

(8) On [EFFECTIVE DATE OF THE 

FINAL RULE], the license condition 

associated with Order EA–06–137 is 

deemed removed from the power reactor 

license of each applicable licensee 

subject to this section. 

* * * * * 

■ 28. Add § 50.200 to read as follows: 

§ 50.200 Power reactor decommissioning 

emergency plans. 

(a) Post-shutdown emergency plans 

(PSEP). If the licensee elects in 

§ 50.54(q)(7)(i) to comply with this 

section, then the licensee’s onsite 

emergency response plans must meet 

the planning standards of § 50.47(b) and 

the requirements in appendix E to this 

part. For a PSEP, emergency response 

organization (ERO) staffing required by 

§ 50.47(b)(2) and appendix E to this part 

may be commensurate with a reduced 

spectrum of credible accidents for a 

permanently shutdown and defueled 

power reactor facility. 

(b) Permanently defueled emergency 

plans (PDEP). If the licensee elects in 

§ 50.54(q)(7)(ii) to comply with this 

section, then the licensee’s onsite 

emergency response plans must meet 

the requirements in paragraph (c) of this 

section and the following planning 

standards: 

(1) Primary responsibilities for 

emergency response by the nuclear 

facility licensee and by State and local 

organizations have been assigned, the 

emergency responsibilities of the 

various supporting organizations have 

been specifically established, and each 

principal response organization has staff 

to respond and to augment its initial 

response on a continuous basis. 

(2) On-shift facility licensee 

responsibilities for emergency response 

are unambiguously defined, adequate 

staffing to provide initial facility 

accident response in key functional 

areas is maintained at all times, timely 

augmentation of response capabilities is 

available, and the interfaces among 

various onsite response activities and 

offsite support and response activities 

are specified. 

(3) Arrangements for requesting and 

effectively using assistance resources 

have been made, and other 

organizations capable of augmenting the 

planned response have been identified. 

(4) A standard emergency 

classification and action level scheme, 

the bases of which include facility 

system and effluent parameters, is in 

use by the nuclear facility licensee. 

(5) Procedures have been established 

for notification, by the licensee, of State 

and local response organizations and for 

notification of emergency personnel by 

all organizations; the content of initial 

and followup messages to response 

organizations has been established. 

(6) Provisions exist for prompt 

communications among principal 

response organizations to emergency 

personnel. 

(7) The principal points of contact 

with the news media for dissemination 

of information during an emergency are 

established in advance, and procedures 

for coordinated dissemination of 

information to the public are 

established. 

(8) Adequate emergency facilities and 

equipment to support the emergency 

response are provided and maintained. 

(9) Adequate methods, systems, and 

equipment for assessing and monitoring 

actual or potential consequences of a 

radiological emergency condition are in 

use. 

(10) A range of protective actions has 

been developed for emergency workers 

and the public. 

(11) Means for controlling radiological 

exposures in an emergency are 

established for emergency workers. 

(12) Arrangements are made for 

medical services for contaminated 

injured individuals. 

(13) General plans for recovery and 

reentry are developed. 

(14) Periodic exercises will be 

conducted to evaluate major portions of 

emergency response capabilities, 


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periodic drills will be conducted to 

develop and maintain key skills, and 

deficiencies identified as a result of 

exercises or drills will be corrected. 

(15) Radiological emergency response 

training is provided to those who may 

be called on to assist in an emergency. 

(16) Responsibilities for plan 

development and review and for 

distribution of emergency plans are 

established, and planners are properly 

trained. 

(c) Content of emergency plans. (1) 

Emergency plans must contain, but not 

necessarily be limited to, information 

needed to demonstrate compliance with 

the elements set forth in this paragraph, 

i.e., organization for coping with 

radiological emergencies, assessment 

actions, activation of emergency 

organization, notification procedures, 

emergency facilities and equipment, 

training, maintaining emergency 

preparedness, and recovery. 

(i) Organization. (A) The organization 

for coping with radiological 

emergencies must be described, 

including definition of authorities, 

responsibilities, and duties of 

individuals assigned to the licensee’s 

emergency organization and the means 

for notification of such individuals in 

the event of an emergency. Specifically, 

the following must be included: 

(1) A description of the normal plant 

organization. 

(2) A description of the onsite ERO 

with a detailed discussion of: 

(i) Authorities, responsibilities, and 

duties of the individual(s) who will take 

charge during an emergency; 

(ii) Plant staff emergency assignments; 

(iii) Authorities, responsibilities, and 

duties of an onsite emergency 

coordinator who shall be in charge of 

the exchange of information with offsite 

authorities responsible for coordinating 

and implementing offsite emergency 

measures. 

(3) Identification, by position and 

function to be performed, of persons 

within the licensee organization who 

will be responsible for making dose 

projections, and a description of how 

these projections will be made and the 

results transmitted to State and local 

authorities, NRC, and other appropriate 

governmental entities. 

(4) A description of the local offsite 

services to be provided in support of the 

licensee’s emergency organization. 

(5) Identification of assistance 

expected from appropriate State, local, 

and Federal agencies with 

responsibilities for coping with 

emergencies, including an act directed 

toward a nuclear power plant or its 

personnel that includes the use of 

violent force to destroy equipment, take 

hostages, and/or intimidate the licensee 

to achieve an end. This includes attack 

by air, land, or water using guns, 

explosives, projectiles, vehicles, or other 

devices used to deliver destructive 

force. 

(B) [Reserved] 

(ii) Assessment actions. (A) The 

means to be used for determining the 

magnitude of, and for continually 

assessing the impact of, the release of 

radioactive materials must be described, 

including emergency action levels that 

are to be used as criteria for determining 

the need for notification and 

participation of local and State agencies, 

the Commission, and other Federal 

agencies, and the emergency action 

levels that are to be used for 

determining when and what type of 

protective measures should be 

considered within the site boundary to 

protect health and safety. The 

emergency action levels must be based 

on in-plant conditions and 

instrumentation in addition to onsite 

monitoring. Emergency action levels 

must be reviewed with the State and 

local governmental authorities on an 

annual basis. 

(B) A licensee desiring to change its 

entire emergency action level scheme 

must submit an application for an 

amendment to its license and receive 

NRC approval before implementing the 

change. Licensees must follow the 

change process in § 50.54(q) for all other 

emergency action level changes. 

(iii) Activation of emergency 

organization. (A) The entire spectrum of 

emergency conditions that involve the 

alerting or activating of progressively 

larger segments of the total emergency 

organization must be described. The 

communication steps to be taken to alert 

or activate emergency personnel under 

each class of emergency must be 

described. Emergency action levels, 

based not only on onsite radiation 

monitoring information but also on 

readings from a number of sensors that 

indicate a potential emergency for 

notification of offsite agencies, must be 

described. The existence, but not the 

details, of a message authentication 

scheme must be noted for such agencies. 

The emergency classes defined must 

include: 

(1) Notification of unusual events; and 

(2) Alert. 

(B) Licensees must establish and 

maintain the capability to assess, 

classify, and declare an emergency 

condition as soon as possible and 

within 60 minutes after the availability 

of indications to plant operators that an 

emergency action level has been 

exceeded and must promptly declare 

the emergency condition as soon as 

possible following identification of the 

appropriate emergency classification 

level. Licensees must not construe these 

criteria as a grace period to attempt to 

restore plant conditions to avoid 

declaring an emergency action due to an 

emergency action level that has been 

exceeded. Licensees must not construe 

these criteria as preventing 

implementation of response actions 

deemed by the licensee to be necessary 

to protect public health and safety 

provided that any delay in declaration 

does not deny the State and local 

authorities the opportunity to 

implement measures necessary to 

protect the public health and safety. 

(iv) Notification procedures. (A) 

Administrative and physical means for 

notifying local, State, and Federal 

officials and agencies must be 

described. This description must 

include identification of the State and 

local government agencies. 

(B) A licensee must have the 

capability to notify responsible State 

and local governmental agencies as soon 

as possible and within 60 minutes after 

declaring an emergency. 

(v) Emergency facilities and 

equipment. Adequate provisions must 

be made and described for emergency 

facilities and equipment, including: 

(A) Equipment at the site for 

personnel monitoring; 

(B) Equipment for determining the 

magnitude of and for continuously 

assessing the impact of the release of 

radioactive materials to the 

environment; 

(C) Facilities and supplies at the site 

for decontamination of onsite 

individuals; 

(D) Facilities and medical supplies at 

the site for appropriate emergency first 

aid treatment; 

(E) Arrangements for medical service 

providers qualified to handle 

radiological emergencies onsite; 

(F) Arrangements for transportation of 

contaminated injured individuals from 

the site to specifically identified 

treatment facilities outside the site 

boundary; 

(G) Arrangements for treatment of 

individuals injured in support of 

licensed activities on the site at 

treatment facilities outside the site 

boundary; 

(H) A licensee facility from which 

effective direction can be given and 

effective control can be exercised during 

an emergency; 

(I) At least one onsite and one offsite 

communications system; each system 

must have a backup power source. All 

communication plans must have 

arrangements for emergencies, including 

titles and alternates for those in charge 


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1Use of site-specific simulators or computers is 

acceptable for any exercise. 

at both ends of the communication links 

and the primary and backup means of 

communication. Where consistent with 

the function of the governmental 

agency, these arrangements will 

include: 

(1) Provision for communications 

with contiguous State and local 

governments. Such communications 

must be tested monthly. 

(2) Provision for communications 

with Federal emergency response 

organizations. Such communications 

systems must be tested annually. 

(3) Provisions for communications by 

the licensee with NRC Headquarters and 

the appropriate NRC Regional Office 

Operations Center from the facility. 

Such communications must be tested 

monthly. 

(vi) Training. (A) The training 

program must provide for: 

(1) The training of employees and 

exercising, by periodic drills, of 

emergency plans to ensure that 

employees of the licensee are familiar 

with their specific emergency response 

duties, and 

(2) The participation in the training 

and drills by other persons whose 

assistance may be needed in the event 

of a radiological emergency. The plan 

must include a description of 

specialized initial training and periodic 

retraining programs to be provided to 

each of the following categories of 

emergency personnel: 

(i) Directors and/or coordinators of 

the plant emergency organization; 

(ii) Personnel responsible for accident 

assessment; 

(iii) Radiological monitoring teams; 

(iv) Fire control teams (fire brigades); 

(v) Repair and damage control teams; 

(vi) First aid and rescue teams; 

(vii) Medical support personnel; and 

(viii) Security personnel. 

(3) In addition, a radiological 

orientation training program must be 

made available to local services 

personnel, such as local emergency 

services and local law enforcement 

personnel. 

(B) The plan must describe provisions 

for the conduct of emergency 

preparedness exercises as follows: 

Exercises must test the adequacy of 

timing and content of implementing 

procedures and methods, test 

emergency equipment and 

communications networks, and ensure 

that emergency organization personnel 

are familiar with their duties.1

(1) Within two years of the last 

exercise of the onsite emergency plan 

performed under section IV.F.2.b of 

appendix E to this part, each licensee 

must conduct an exercise of its onsite 

emergency plan. 

(2) Each licensee at each site must 

conduct a subsequent exercise of its 

onsite emergency plan every 2 years. In 

addition, the licensee must take actions 

necessary to ensure that adequate 

emergency response capabilities are 

maintained during the interval between 

biennial exercises by conducting drills, 

including at least one drill involving a 

combination of some of the principal 

functional areas of the licensee’s onsite 

emergency response capabilities. The 

principal functional areas of emergency 

response include activities such as 

management and coordination of 

emergency response, accident 

assessment, event classification, 

notification of offsite authorities, 

assessment of the onsite impact of 

radiological releases, system repair, and 

mitigative action implementation. 

During these drills, activation of all of 

the licensee’s emergency response 

facilities is not necessary, licensees have 

the opportunity to consider accident 

management strategies, supervised 

instruction is permitted, operating staff 

in all participating facilities have the 

opportunity to resolve problems 

(success paths) rather than have 

controllers intervene, and the drills may 

focus on the onsite exercise training 

objectives. 

(3) Each licensee shall enable any 

State or local government to participate 

in the licensee’s drills and exercises 

when requested by such State or local 

government. 

(4) Remedial exercises will be 

required if the emergency plan is not 

satisfactorily tested during the biennial 

exercise, such that NRC cannot: 

(i) Find reasonable assurance that 

adequate protective measures can and 

will be taken in the event of a 

radiological emergency; or 

(ii) Determine that the ERO has 

maintained key skills specific to 

emergency response. 

(5) All exercises, drills, and training 

that provide performance opportunities 

to develop, maintain, or demonstrate 

key skills must provide for formal 

critiques in order to identify weak or 

deficient areas that need correction. Any 

weaknesses or deficiencies that are 

identified in a critique of exercises, 

drills, or training must be corrected. 

(6) Each licensee shall use drill and 

exercise scenarios that provide 

reasonable assurance that anticipatory 

responses will not result from 

preconditioning of participants. 

Exercise and drill scenarios as 

appropriate must emphasize 

coordination among onsite and offsite 

response organizations. 

(vii) Maintaining emergency 

preparedness. (A) Provisions to be 

employed to ensure that the emergency 

plan, its implementing procedures, and 

emergency equipment and supplies are 

maintained up to date must be 

described. 

(B) [Reserved] 

(viii) Recovery. (A) Criteria to be used 

to determine when, following an 

accident, reentry of the facility would be 

appropriate must be described. 

(B) [Reserved] 

(2) [Reserved] 

* * * * * 

■ 29. Amend appendix E to part 50 by: 

■ a. Removing paragraph I.6; 

■ b. In paragraph IV.4, removing the 

words ‘‘of the later of the date’’ and ‘‘or 

December 23, 2011,’’; 

■ c. Adding paragraph IV.8; 

■ d. In paragraph IV.A.7, removing the 

words, ‘‘By June 23, 2014, 

identification’’ and adding in their place 

the word, ‘‘Identification’’; 

■ e. In paragraph IV.A.9, removing the 

words, ‘‘By December 24, 2012, for’’ and 

adding in their place the word, ‘‘For’’; 

■ f. In paragraph IV.B.1, removing the 

words, ‘‘By June 20, 2012, for’’ and 

adding in their place the word, ‘‘For’’; 

■ g. In paragraph IV.C.2, removing the 

words, ‘‘By June 20, 2012, nuclear’’ and 

adding in their place the word, 

‘‘Nuclear’’; 

■ h. In paragraph IV.E.8.c introductory 

text, removing the words, ‘‘By June 20, 

2012, for’’ and adding in their place the 

word, ‘‘For’’; 

■ i. In paragraph IV.E.8.d, removing the 

last sentence; 

■ j. In paragraph IV.F.2.d removing the 

words ‘‘and should fully participate in 

one hostile action exercise by December 

31, 2015’’; 

■ k. Removing and reserving paragraph 

IV.F.2.j(v); 

■ l. Adding paragraph IV.F.2.k; 

■ m. In paragraph IV.I, removing the 

words, ‘‘By June 20, 2012, for’’ and 

adding in their place the word, ‘‘For’’; 

The revisions and addition read as 

follows: 

Appendix E to Part 50—Emergency 

Planning and Preparedness for 

Production and Utilization Facilities 

* * * * * 

IV. * * * 

8. A nuclear power reactor licensee is not 

subject to the requirements of paragraphs 4, 

5, and 6 of this section once the NRC dockets 

the licensee’s certifications required under 

§ 50.82(a)(1) or § 52.110(a) of this chapter. 

* * * * * 

F. * * * 

2. * * * 


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k. For each nuclear reactor for which the 

NRC has docketed the certifications required 

under § 50.82(a)(1) or § 52.110(a) of this 

chapter, the nuclear reactor’s licensee must 

follow the biennial exercise requirements of 

paragraph 2 of this section. 

* * * * * 

Appendix I to Part 50 [Amended] 

■ 30. In section IV.C, add ‘‘or § 52.110(a) 

of this chapter’’ after ‘‘§ 50.82(a)(1)’’. 

PART 51—ENVIRONMENTAL 

PROTECTION REGULATIONS FOR 

DOMESTIC LICENSING AND RELATED 

REGULATORY FUNCTIONS 

■ 31. The authority citation for part 51 

continues to read as follows: 

Authority: Atomic Energy Act of 1954, 

secs. 161, 193 (42 U.S.C. 2201, 2243) Energy 

Reorganization Act of 1974, secs. 201, 202 

(42 U.S.C. 5841, 5842); National 

Environmental Policy Act of 1969 (42 U.S.C. 

4332, 4334, 4335); Nuclear Waste Policy Act 

of 1982, sec. 144(f), 121, 135, 141, 148 (42 

U.S.C. 10134(f), 10141, 10155, 10161, 10168); 

44 U.S.C. 3504 note. 

§ 51.53 [Amended] 

■ 32. In § 51.53, in paragraph (d), 

remove the words ‘‘Each applicant for a 

license amendment authorizing 

decommissioning activities for a 

production or utilization facility either 

for unrestricted use or based on 

continuing use restrictions applicable to 

the site; and each applicant for a license 

amendment approving a license 

termination plan or decommissioning 

plan under § 50.82 of this chapter’’ and 

add in their place the words ‘‘Each 

applicant for a license amendment 

approving an irradiated fuel 

management plan under § 50.54(bb) of 

this chapter; each applicant for a license 

amendment approving a license 

termination plan under § 50.82 of this 

chapter or § 52.110 of this chapter or a 

decommissioning plan under § 50.82 of 

this chapter’’. 

§ 51.95 [Amended] 

■ 33. In § 51.95, in paragraph (d) remove 

the words ‘‘of an operating or combined 

license authorizing decommissioning 

activities at a production or utilization 

facility covered by § 51.20,’’ and add in 

their place the words ‘‘approving an 

irradiated fuel management plan under 

§ 50.54(bb) of this chapter, or the 

amendment approving a license 

termination plan under § 50.82 of this 

chapter or § 52.110 of this chapter or a 

decommissioning plan under § 50.82 of 

this chapter’’. 

PART 52—LICENSES, 

CERTIFICATIONS, AND APPROVALS 

FOR NUCLEAR POWER PLANTS 

■ 34. Revise the authority citation for 

part 52 to read as follows: 

Authority: Atomic Energy Act of 1954, 

secs. 53, 63, 81, 103, 104, 147, 149, 161, 181, 

182, 183, 185, 186, 189, 223, 234 (42 U.S.C. 

2073, 2093, 2113, 2133, 2134, 2167, 2169, 

2201, 2231, 2232, 2233, 2235, 2236, 2239, 

2273, 2282); Energy Reorganization Act of 

1974, secs. 201, 202, 206, 211 (42 U.S.C. 

5841, 5842, 5846, 5851); 44 U.S.C. 3504 note. 

■ 35. In § 52.0, revise paragraph (a) to 

read as follows: 

§ 52.0 Scope; applicability of 10 CFR 

Chapter I provisions. 

(a) This part governs the issuance of 

early site permits, standard design 

certifications, combined licenses, 

standard design approvals, and 

manufacturing licenses for nuclear 

power facilities licensed under Section 

103 of the Atomic Energy Act of 1954, 

as amended (68 Stat. 919), and Title II 

of the Energy Reorganization Act of 

1974 (88 Stat. 1242) through the 

termination of the associated 10 CFR 

part 52 licenses. This part also gives 

notice to all persons who knowingly 

provide to any holder of or applicant for 

an approval, certification, permit, or 

license, or to a contractor, 

subcontractor, or consultant of any of 

them, components, equipment, 

materials, or other goods or services that 

relate to the activities of a holder of or 

applicant for an approval, certification, 

permit, or license, subject to this part, 

that they may be individually subject to 

NRC enforcement action for violation of 

the provisions in § 52.4. 

* * * * * 

■ 36. In § 52.63, revise paragraph (b)(2) 

to read as follows: 

§ 52.63 Finality of standard design 

certifications. 

* * * * * 

(b) * * * 

(2) Subject to § 50.59 of this chapter, 

a licensee who references a design 

certification rule may make departures 

from the design of the nuclear power 

facility, without prior Commission 

approval, unless the proposed departure 

involves a change to the design as 

described in the rule certifying the 

design. 

(i) The licensee shall maintain records 

of all departures from the design of the 

facility and these records must be 

maintained and available for audit until 

the date of termination of the license. 

(ii) Licensees for which the NRC has 

docketed the certifications required 

under § 52.110(a) are not required to 

retain records of departures from the 

design of the facility associated solely 

with structures, systems, and 

components that have been permanently 

removed from service using an NRCapproved change process. 

* * * * * 

§ 52.109 [Amended] 

■ 37. In § 52.109, remove the words ‘‘to 

authorize ownership and possession of 

the production or utilization facility,’’. 

■ 38. In § 52.110, revise paragraphs (b), 

(d), (e), (f)(2), (h)(1)(i), and (h)(2), add 

paragraphs (h)(5) through (7), and revise 

paragraph (i) introductory text and 

paragraph (i)(2)(vi) to read as follows: 

§ 52.110 Termination of license. 

* * * * * 

(b)(1) Upon the NRC’s docketing of 

the licensee’s certifications required 

under paragraph (a) of this section, or 

when a final legally effective order to 

permanently cease operations has come 

into effect, the 10 CFR part 52 license 

no longer authorizes operation of the 

reactor or emplacement or retention of 

fuel into the reactor vessel. 

(2) The facility licensed under this 

part is no longer a utilization facility 

once the licensee meets the criteria of 

paragraph (b)(1) of this section and 

modifies the facility to be incapable of 

making use of special nuclear material 

without significant facility alterations 

necessary to restore the capability to 

make use of special nuclear material. 

The NRC maintains the authority to 

regulate the 10 CFR part 52 license with 

respect to the possession of special 

nuclear material, source material, and 

byproduct material under sections 53, 

63, 81, and 161 of the Act, as applicable. 

Until the termination of the 10 CFR part 

52 license under paragraph (k) of this 

section, the regulations of this chapter 

applicable to a utilization facility 

continue to apply to the holder of the 

license unless the regulations explicitly 

state otherwise. 

* * * * * 

(d)(1) Prior to or within 2 years 

following permanent cessation of 

operations, the licensee shall submit a 

post-shutdown decommissioning 

activities report (PSDAR) to the NRC, 

and a copy to the affected State(s). The 

PSDAR must contain a description of 

the planned decommissioning activities 

along with a schedule for their 

accomplishment, a discussion whether 

the environmental impacts associated 

with site-specific decommissioning 

activities will be bounded by 

appropriate federally issued 

environmental review documents, a 

description of any decommissioning 

activities whose environmental impacts 


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will not be so bounded and will be 

evaluated prior to the performance of 

the activities, and a site-specific 

decommissioning cost estimate, 

including the projected cost of 

managing irradiated fuel. 

(2) The NRC shall notice in the 

Federal Register the receipt of the 

PSDAR and the availability for public 

comment of the PSDAR. The NRC shall 

also schedule a public meeting in the 

vicinity of the licensee’s facility upon 

receipt of the PSDAR. The NRC shall 

include a notice in a forum, such as 

local newspapers, that is readily 

accessible to individuals in the vicinity 

of the site, and in the Federal Register 

notice required by this paragraph (d)(2), 

announcing the date, time and location 

of the meeting, along with a brief 

description of the purpose of the 

meeting. 

(e) Licensees shall not perform any 

major decommissioning activities, as 

defined in § 50.2 of this chapter, until 

90 days after the NRC has received the 

licensee’s PSDAR submittal and until 

certifications of permanent cessation of 

operations and permanent removal of 

fuel from the reactor vessel, as required 

under § 52.110(a), have been submitted. 

(f) * * * 

(2) Result in significant 

environmental impacts not bounded by 

appropriate federally issued 

environmental review documents; or 

* * * * * 

(h) * * * 

(1) * * * 

(i) The withdrawals are for expenses 

for activities consistent with the 

definition of decommission in § 52.1; 

* * * * * 

(2) Initially, 3 percent of the generic 

amount specified in § 50.75(c) of this 

chapter may be used for 

decommissioning planning. For 

licensees that have submitted the 

certifications required under paragraph 

(a) of this section and commencing 90 

days after the NRC has received the 

PSDAR, an additional 20 percent may 

be used. A site-specific 

decommissioning cost estimate must be 

submitted to the NRC before the 

licensee may use any funding in excess 

of these amounts. 

* * * * * 

(5) After submitting its site-specific 

decommissioning cost estimate required 

by paragraph (d)(1) of this section, and 

until the licensee has completed its final 

radiation survey and demonstrated that 

residual radioactivity has been reduced 

to a level that permits termination of its 

license, the licensee must annually 

submit to the NRC, by March 31, a 

financial assurance status report. The 

report may combine the reporting 

requirements of § 72.30 of this chapter 

and § 52.110(h)(7). The report must 

include the following information, 

current through the end of the previous 

calendar year: 

(i) The amount spent on 

decommissioning, both cumulative and 

over the previous calendar year, the 

remaining balance of any 

decommissioning funds, and the 

amount provided by other financial 

assurance methods being relied upon; 

(ii) An estimate of the costs to 

complete decommissioning, reflecting 

any difference between actual and 

estimated costs for work performed 

during the year, and the 

decommissioning criteria upon which 

the estimate is based; 

(iii) Any modifications occurring to a 

licensee’s current method of providing 

financial assurance since the last 

submitted report; and 

(iv) Any material changes to trust 

agreements or financial assurance 

contracts. 

(6) If the sum of the balance of any 

remaining decommissioning funds, plus 

earnings on such funds calculated at not 

greater than a 2 percent real rate of 

return, together with the amount 

provided by other financial assurance 

methods being relied upon, does not 

cover the estimated cost to complete the 

decommissioning, the financial 

assurance status report must include 

additional financial assurance to cover 

the estimated cost of completion. 

(7) After submitting its site-specific 

decommissioning cost estimate required 

by paragraph (d)(1) of this section, if 

spent fuel is on site, the licensee must 

annually submit to the NRC, by March 

31, a report on the status of its funding 

for managing irradiated fuel. The report 

must include the following information, 

current through the end of the previous 

calendar year: 

(i) The amount of funds accumulated 

to cover the cost of managing the 

irradiated fuel; 

(ii) The projected cost of managing 

irradiated fuel until title to the fuel and 

possession of the fuel is transferred to 

the Secretary of Energy; and 

(iii) If the funds accumulated do not 

cover the projected cost, a plan to obtain 

additional funds to cover the cost. 

(i) All power reactor licensees that 

have loaded fuel into the reactor must 

submit an application for termination of 

license. The application for termination 

of license must be accompanied or 

preceded by a license termination plan 

to be submitted for NRC approval. 

* * * * * 

(2) * * * 

(vi) An updated site-specific estimate 

of remaining decommissioning costs 

and identification of sources of funds 

for license termination, spent fuel 

management, and ISFSI 

decommissioning, as applicable; 

* * * * * 

PART 72—LICENSING 

REQUIREMENTS FOR THE 

INDEPENDENT STORAGE OF SPENT 

NUCLEAR FUEL, HIGH-LEVEL 

RADIOACTIVE WASTE, AND 

REACTOR-RELATED GREATER THAN 

CLASS C WASTE 

■ 39. The authority citation for part 72 

continues to read as follows: 

Authority: Atomic Energy Act of 1954, 

secs. 51, 53, 57, 62, 63, 65, 69, 81, 161, 182, 

183, 184, 186, 187, 189, 223, 234, 274 (42 

U.S.C. 2071, 2073, 2077, 2092, 2093, 2095, 

2099, 2111, 2201, 2210e, 2232, 2233, 2234, 

2236, 2237, 2238, 2273, 2282, 2021); Energy 

Reorganization Act of 1974, secs. 201, 202, 

206, 211 (42 U.S.C. 5841, 5842, 5846, 5851); 

National Environmental Policy Act of 1969 

(42 U.S.C. 4332); Nuclear Waste Policy Act 

of 1982, secs. 117(a), 132, 133, 134, 135, 137, 

141, 145(g), 148, 218(a) (42 U.S.C. 10137(a), 

10152, 10153, 10154, 10155, 10157, 10161, 

10165(g), 10168, 10198(a)); 44 U.S.C. 3504 

note. 

■ 40. In § 72.13, add paragraph (e) to 

read as follows: 

§ 72.13 Applicability. 

* * * * * 

(e) The following sections apply to 

activities associated with a general 

license, where the licensee has elected 

to provide for physical protection of the 

spent fuel in accordance with 

§ 72.212(b)(9)(vii)(A): § 72.1; § 72.2(a)(1), 

(b), (c), and (e); §§ 72.3 through 

72.6(c)(1); §§ 72.7 through § 72.13(a) and 

(e); § 72.30(b), (c), (d), (e), and (f); 

§ 72.32(c) and (d); § 72.44(b) and (f); 

§ 72.48; § 72.50(a); § 72.52(a), (b), (d), 

and (e); § 72.60; § 72.62; §§ 72.72 

through 72.80(f); §§ 72.82 through 72.86; 

§§ 72.104 through 72.106; §§ 72.122 

through 72.126; §§ 72.140 through 

72.176; §§ 72.180 through 72.186; 

§ 72.190; § 72.194; §§ 72.210 through 

72.220; and § 72.240(a). 

■ 41. In § 72.30, revise paragraph (b) and 

paragraph (c) introductory text to read 

as follows: 

§ 72.30 Financial assurance and 

recordkeeping for decommissioning. 

* * * * * 

(b)(1) Each applicant for a specific 

license under this part must submit, as 

part of its application, a 

decommissioning funding plan for NRC 

review and approval. 

(2) Each holder of a general license 

under this part must submit, prior to the 


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initial storage of spent fuel under 

§ 72.212(a)(3), a decommissioning 

funding plan for NRC review and 

approval. 

(3) The decommissioning funding 

plans required by paragraphs (b)(1) and 

(2) of this section must contain: 

(i) Information on how reasonable 

assurance will be provided that funds 

will be available to decommission the 

ISFSI or MRS. 

(ii) A detailed cost estimate for 

decommissioning, in an amount 

reflecting: 

(A) The cost of an independent 

contractor to perform all 

decommissioning activities; 

(B) An adequate contingency factor; 

and 

(C) The cost of meeting the § 20.1402 

of this chapter criteria for unrestricted 

use, provided that, if the applicant or 

licensee can demonstrate its ability to 

meet the provisions of § 20.1403 of this 

chapter, the cost estimate may be based 

on meeting the § 20.1403 criteria. 

(iii) Identification of and justification 

for using the key assumptions contained 

in the decommissioning cost estimate. 

(iv) A description of the method of 

assuring funds for decommissioning 

from paragraph (e) of this section, 

including means for adjusting cost 

estimates and associated funding levels 

periodically over the life of the facility. 

(v) The volume of onsite subsurface 

material containing residual 

radioactivity that will require 

remediation to meet the criteria for 

license termination. 

(vi) A certification that financial 

assurance for decommissioning has 

been provided in the amount of the cost 

estimate for decommissioning. 

(c) At the time of license renewal and 

at intervals not to exceed 3 years, the 

decommissioning funding plan must be 

resubmitted with adjustments as 

necessary to account for changes in 

costs and the extent of contamination. 

The decommissioning funding plan 

must update the information submitted 

with the original or prior plan and must 

specifically consider the effect of the 

following events on decommissioning 

costs: 

* * * * * 

■ 42. In § 72.32, revise paragraphs (a) 

introductory text and (c) to read as 

follows: 

§ 72.32 Emergency Plan. 

(a) Each application for an ISFSI that 

is licensed under this part which is not 

located on the site or within the 

exclusion area, as defined in 10 CFR 

part 100, of a nuclear power reactor 

licensed under part 50 of this chapter or 

part 52 of this chapter must be 

accompanied by an Emergency Plan that 

includes the following information: 

* * * * * 

(c) For an ISFSI that is located on the 

site or within the exclusion area, as 

defined in 10 CFR part 100, of a nuclear 

power reactor licensed under parts 50 or 

52 of this chapter, an emergency plan 

that meets the requirements in appendix 

E to part 50 of this chapter and 

§ 50.47(b) of this chapter, or the 

requirements of 10 CFR 50.200(a) or 10 

CFR 50.200(b) shall be deemed to satisfy 

the requirements of this section. 

■ 43. In § 72.44, revise paragraph (f) to 

read as follows: 

§ 72.44 License conditions. 

* * * * * 

(f) A licensee shall follow and 

maintain in effect an emergency plan 

that is approved by the Commission. 

The licensee may make changes to the 

approved plan without Commission 

approval only if such changes do not 

decrease the effectiveness of the plan. 

Within six months after any change is 

made, the licensee shall submit, in 

accordance with § 72.4, a report 

containing a description of any changes 

made in the plan addressed to Director, 

Division of Fuel Management, Office of 

Nuclear Material Safety and Safeguards, 

U.S. Nuclear Regulatory Commission, 

with a copy to the appropriate NRC 

Regional Office shown in appendix D to 

part 20 of this chapter. Proposed 

changes that decrease the effectiveness 

of the approved emergency plan must 

not be implemented unless the licensee 

has received prior approval of such 

changes from the Commission. 

Licensees need not comply with the 

requirements of this paragraph when all 

spent fuel has been removed from the 

site. 

■ 44. In § 72.62, revise paragraph (a)(2) 

to read as follows: 

§ 72.62 Backfitting. 

(a) * * * 

(2) Procedures or organization 

required to operate or decommission an 

ISFSI or MRS. 

* * * * * 

■ 45. In § 72.72, revise paragraph (d) to 

read as follows: 

§ 72.72 Material balance, inventory, and 

records requirements for stored materials. 

* * * * * 

(d)(1) Except as provided in paragraph 

(d)(2) of this section, records of spent 

fuel, high-level radioactive waste, and 

reactor-related GTCC waste containing 

special nuclear material meeting the 

requirements in paragraph (a) of this 

section must be kept in duplicate. The 

duplicate set of records must be kept at 

a separate location sufficiently remote 

from the original records that a single 

event would not destroy both sets of 

records. 

(2) A single copy of the records 

described in paragraph (d)(1) of this 

section may be maintained in a single 

storage facility provided the facility 

meets the requirements of an NRCapproved quality assurance program for 

the storage of records. 

(3) Records of spent fuel or reactorrelated GTCC waste containing special 

nuclear material transferred out of an 

ISFSI or records of spent fuel, high-level 

radioactive waste, or reactor-related 

GTCC waste containing special nuclear 

material transferred out of an MRS must 

be preserved for a period of five years 

after the date of transfer. 

■ 46. In § 72.212, add paragraph 

(b)(9)(vii) to read as follows: 

§ 72.212 Conditions of general license 

issued under § 72.210. 

* * * * * 

(b) * * * 

(9) * * * 

(vii)(A) Upon NRC docketing of the 

certifications required under 

§ 50.82(a)(1) of this chapter or 

§ 52.110(a) of this chapter, and when all 

spent fuel has been placed in dry cask 

storage at the facility, the licensee may, 

as an alternative to the requirements of 

§ 72.212(b)(9)(i) through (vi), provide for 

physical protection of the spent fuel 

under subpart H of this part and § 73.51 

of this chapter. 

(B) A licensee who elects to provide 

physical protection under subpart H of 

this part and § 73.51 of this chapter will 

submit their physical security plan to 

the NRC under § 50.54(p) of this 

chapter. 

* * * * * 

■ 47. Revise § 72.218 to read as follows: 

§ 72.218 Termination of licenses. 

(a) Upon removal of the spent fuel 

stored under this general license from 

the reactor site, the licensee must 

decommission the ISFSI consistent with 

requirements in § 50.82 of this chapter 

or § 52.110 of this chapter, as 

applicable. 

(b) The general license under this part 

is terminated upon termination of the 10 

CFR part 50 or 10 CFR part 52 license 

under § 50.82(a)(11) of this chapter or 

§ 52.110(k) of this chapter, respectively. 

PART 73—PHYSICAL PROTECTION OF 

PLANTS AND MATERIALS 

■ 48. The authority citation for part 73 

continues to read as follows: 

Authority: Atomic Energy Act of 1954, 

secs. 53, 147, 149, 161, 170D, 170E, 170H, 


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170I, 223, 229, 234, 1701 (42 U.S.C. 2073, 

2167, 2169, 2201, 2210d, 2210e, 2210h, 

2210i, 2273, 2278a, 2282, 2297f); Energy 

Reorganization Act of 1974, secs. 201, 202 

(42 U.S.C. 5841, 5842); Nuclear Waste Policy 

Act of 1982, secs. 135, 141 (42 U.S.C. 10155, 

10161); 44 U.S.C. 3504 note. 

Section 73.37(b)(2) also issued under sec. 

301, Pub. L. 96–295, 94 Stat. 789 (42 U.S.C. 

5841 note). 

■ 49. In § 73.51, revise paragraphs (a) 

introductory text, (a)(1) introductory 

text, and (a)(2) and add paragraph (a)(3) 

to read as follows: 

§ 73.51 Requirements for the physical 

protection of stored spent nuclear fuel and 

high-level radioactive waste. 

(a) Applicability. Notwithstanding the 

provisions of § 73.20, § 73.50, or § 73.67, 

the physical protection requirements of 

this section apply to each licensee that 

stores spent nuclear fuel and high-level 

radioactive waste: 

(1) Under a specific license issued 

pursuant to part 72 of this chapter: 

* * * * * 

(2) At a geologic repository operations 

area (GROA) licensed pursuant to part 

60 or 63 of this chapter; or 

(3) Under a general license issued 

pursuant to part 72 of this chapter and 

upon the NRC’s docketing of the 

certifications required under 

§ 50.82(a)(1) of this chapter or 

§ 52.110(a) of this chapter, when all 

spent fuel has been placed in dry cask 

storage at the facility, and notification 

has been made to the NRC under the 

provisions of § 72.212(b)(9)(vii) of this 

chapter. 

* * * * * 

■ 50. In § 73.54, remove the 

introductory text, revise the paragraph 

(a) introductory text, paragraph (b) 

introductory text, and paragraph (c) 

introductory text, and add paragraphs (i) 

and (j) to read as follows: 

§ 73.54 Protection of digital computer and 

communication systems and networks. 

(a) Each holder of an operating license 

for a nuclear power reactor under part 

50 of this chapter and each holder of a 

combined license under part 52 of this 

chapter for which the Commission has 

made the finding under § 52.103(g) of 

this chapter shall provide high 

assurance that its digital computer and 

communication systems and networks 

are adequately protected against cyber 

attacks, up to and including the design 

basis threat as described in § 73.1. 

* * * * * 

(b) To accomplish the objectives in 

paragraph (a) of this section, the 

licensee shall: 

* * * * * 

(c) The licensee’s cyber security 

program must be designed to: 

* * * * * 

(i) The requirements of this section no 

longer apply once the following criteria 

are satisfied: 

(1) The NRC has docketed the 

licensee’s certifications required under 

§ 50.82(a)(1) of this chapter or 

§ 52.110(a) of this chapter; and 

(2) At least 10 months (for a boiling 

water reactor) or at least 16 months (for 

a pressurized water reactor) have 

elapsed since the date of permanent 

cessation of operations if the fuel meets 

the criteria of § 50.54(q)(7)(ii) of this 

chapter, or an NRC-approved alternative 

spent fuel decay period, submitted 

under § 50.54(q)(7)(ii)(A) or (B) of this 

chapter, has elapsed. 

(j) Removal of cyber security license 

condition. The cyber security plan 

license condition, which requires the 

licensee to fully implement and 

maintain in effect all provisions of the 

Commission-approved cyber security 

plan including changes made pursuant 

to the authority of § 50.90 of this chapter 

and § 50.54(p) of this chapter, is 

removed from the license once the 

conditions in paragraph (i) of this 

section are satisfied. 

■ 51. In § 73.55: 

■ a. Revise paragraph (b)(3) introductory 

text; 

■ b. Add paragraphs (b)(9)(ii)(B)(1) and 

(2); 

■ c. Revise paragraphs (c)(6), 

(e)(9)(v)(A), (j)(4)(ii), and (p)(1)(i) and 

(ii). 

The revisions and additions read as 

follows: 

§ 73.55 Requirements for physical 

protection of licensed activities in nuclear 

power reactors against radiological 

sabotage. 

* * * * * 

(b) * * * 

(3) The physical protection program 

must be designed to prevent significant 

core damage until the NRC has docketed 

the certifications required under 

§ 50.82(a)(1) of this chapter or 

§ 52.110(a) of this chapter. The physical 

protection program must also be 

designed to prevent spent fuel sabotage. 

Specifically, the program must: 

* * * * * 

(9) * * * 

(ii) * * * 

(B) * * * 

(1) Licensees who are implementing 

10 CFR part 26, regardless of whether 

they are required to do so, are in 

compliance with paragraph (b)(9)(ii)(B) 

of this section. 

(2) Licensees, upon the NRC’s 

docketing of their certifications required 

under § 50.82(a)(1) of this chapter or 

§ 52.110(a) of this chapter, will be in 

compliance with paragraph (b)(9)(ii)(B) 

of this section by implementing the 

following: 

(i) A fitness for duty program in 

which individuals who maintain 

unescorted access authorization and 

have unescorted access to a vital area, 

individuals who perform certified fuel 

handler duties under § 50.2 of this 

chapter prior to all spent nuclear fuel at 

a site being placed in dry cask storage, 

individuals who perform the duties 

under § 26.4(a)(5) of this chapter, and 

individuals who perform duties under 

§ 26.4(g) of this chapter, are subject to 

the requirements in 10 CFR part 26 

except for subparts I and K; and 

(ii) A fitness for duty program in 

which those individuals who maintain 

unescorted access authorization and 

have unescorted access to the protected 

area who are not included in paragraph 

(b)(9)(ii)(B)(2)(i) of this section, are 

subject to the requirements of 

§§ 26.31(c)(1) and (2) and 26.33 of this 

chapter. 

* * * * * 

(c) * * * 

(6) Cyber Security Plan. The licensee 

shall establish, maintain, and 

implement a Cyber Security Plan in 

accordance with the requirements of 

§ 73.54. The licensee no longer needs to 

maintain and implement its Cyber 

Security Plan once the criteria in 

§ 73.54(i) have been satisfied. 

* * * * * 

(e) * * * 

(9) * * * 

(v) * * * 

(A) The reactor control room, unless 

the licensee has submitted and the NRC 

has docketed the certifications required 

under § 50.82(a)(1) of this chapter or 

§ 52.110(a) of this chapter, and the 

licensee has documented that all vital 

equipment has been removed from the 

control room and the control room does 

not serve as the vital area boundary for 

other vital areas; 

* * * * * 

(j) * * * 

(4) * * * 

(ii) A system for communication with 

the control room, or, if the NRC has 

docketed the certifications required 

under § 50.82(a)(1) of this chapter or 

§ 52.110(a) of this chapter, a system for 

communication with the certified fuel 

handler or the senior on-shift licensee 

representative responsible for overall 

safety and security of the permanently 

shutdown and defueled facility. 

* * * * * 

(p) * * * 

(1) * * * 


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12336 Federal Register / Vol. 87, No. 42 / Thursday, March 3, 2022 / Proposed Rules 

(i) In accordance with § 50.54(x) and 

(y) of this chapter, the licensee may 

suspend any security measures under 

this section in an emergency when this 

action is immediately needed to protect 

the public health and safety and no 

action consistent with license 

conditions and technical specifications 

that can provide adequate or equivalent 

protection is immediately apparent. 

This suspension of security measures 

must be approved as a minimum by a 

licensed senior operator, or, if the 

certifications required under 

§ 50.82(a)(1) of this chapter or 

§ 52.110(a) of this chapter have been 

docketed by the NRC, by either a 

licensed senior operator or a certified 

fuel handler, before taking this action. 

(ii) During severe weather when the 

suspension of affected security 

measures is immediately needed to 

protect the personal health and safety of 

security force personnel and no other 

immediately apparent action consistent 

with the license conditions and 

technical specifications can provide 

adequate or equivalent protection. This 

suspension of security measures must 

be approved, as a minimum, by a 

licensed senior operator, or, if the 

certifications required under 

§ 50.82(a)(1) of this chapter or 

§ 52.110(a) of this chapter have been 

docketed by the NRC, by either a 

licensed senior operator or a certified 

fuel handler, with input from the 

security supervisor or manager, before 

taking this action. 

* * * * * 

PART 140—FINANCIAL PROTECTION 

REQUIREMENTS AND INDEMNITY 

AGREEMENTS 

■ 52. The authority citation for part 140 

continues to read as follows: 

Authority: Atomic Energy Act of 1954, 

secs. 161, 170, 223, 234 (42 U.S.C. 2201, 

2210, 2273, 2282); Energy Reorganization Act 

of 1974, secs. 201, 202 (42 U.S.C. 5841, 

5842); 44 U.S.C. 3504 note. 

■ 53. In § 140.11, add paragraph (a)(5), 

redesignate paragraph (b) as paragraph 

(c), revise newly redesignated paragraph 

(c), and add new paragraph (b) to read 

as follows: 

§ 140.11 Amounts of financial protection 

for certain reactors. 

(a) * * * 

(5) In the amount of at least 

$100,000,000, for each nuclear reactor: 

(i) For which the NRC has docketed 

the certifications required under 

§ 50.82(a)(1) of this chapter or 

§ 52.110(a) of this chapter, and 

(ii) For which at least 10 months (for 

a boiling water reactor) or 16 months 

(for a pressurized water reactor) have 

elapsed since the date of permanent 

cessation of operations if the fuel meets 

the criteria of § 50.54(q)(7)(ii) of this 

chapter, or for which an NRC-approved 

alternative to the 10- or 16-month spent 

fuel decay period, submitted under 

§ 50.54(q)(7)(ii)(A) or (B) of this chapter, 

has elapsed. 

(b) Secondary financial protection (in 

the form of private liability insurance 

available under an industry 

retrospective rating plan providing for 

deferred premium charges) will no 

longer be required once the criteria in 

§ 140.11(a)(5)(i) and (ii) have been met. 

(c) In any case where two or more 

nuclear reactors at the same location are 

licensed under parts 50, 52, or 54 of this 

chapter, the total financial protection 

required of the licensee for all such 

reactors (excluding any applicable 

secondary financial protection) is the 

highest amount which would otherwise 

be required for any one of those 

reactors; provided, that such financial 

protection covers all reactors at the 

location. 

■ 54. In § 140.81, revise paragraph (a) to 

read as follows: 

§ 140.81 Scope and purpose. 

(a) Scope. This subpart applies to 

applicants for and holders of operating 

licenses issued under part 50 of this 

chapter, combined licenses issued 

under part 52 of this chapter, or 

renewed licenses issued under part 54 

of this chapter, authorizing operation of 

production facilities and utilization 

facilities, and to other persons 

indemnified with respect to such 

facilities. This subpart shall cease to 

apply to licensees under part 50, part 

52, and part 54 of this chapter once the 

licensee satisfies the criteria in 

§ 140.11(a)(5)(i) and (ii). 

* * * * * 

Dated: February 9, 2022. 

For the Nuclear Regulatory Commission. 

Annette L. Vietti-Cook, 

Secretary of the Commission. 

[FR Doc. 2022–03131 Filed 3–2–22; 8:45 am] 

BILLING CODE 7590–01–P 


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