Regulatory Improvements for Production and Utilization Facilities Transitioning to Decommissioning

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 non-governmental 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:

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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

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.

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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.

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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.”

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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.

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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

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, low-level 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

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

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

1

License 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”).

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 possession-only 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 site-specific 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 neutron-absorbing 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

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

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, beyond-design-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

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 security-related 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 wide-range 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

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

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;

2

Additional 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).

• The appropriate role of State and local governments and non-governmental 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

3

At 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.”

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).

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 possession-only 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 beyond-design-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

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:

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 “away-from-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.”

• “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.

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 low-likelihood, 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

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 beyond-design-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.

EP03MR22.000

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 light-water 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

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-design-basis 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-by-case 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

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 beyond-design-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

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

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 post-shutdown 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 post-shutdown 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

implementation requirements for the emergency response data system (ERDS). These systems transmit near-real-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 beyond-design-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

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 non-power 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

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

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 owner-controlled area during a radiological emergency.

For licensees with PDEPs, pre-planned 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 pre-planned 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 post-shutdown).

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 adequate

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