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:

• 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.

• Cyber security. This proposed rule would provide that the cyber security requirements in § 73.54, “Protection of digital computer and communication systems and networks,” continue to apply to a nuclear power reactor after the licensee's permanent cessation of operations, until all the fuel has been removed from the reactor vessel and there has been sufficient decay of the fuel in the SFP such that it would not reach ignition temperature within 10 hours under adiabatic heatup conditions, at which point no digital computer and communications systems would be required to meet the criteria of § 73.54. This proposed rule would also provide for the removal of the cyber security license condition for 10 CFR part 50 nuclear power reactor licensees after the spent fuel decay period.

• Drug and alcohol testing. This proposed rule would correct inconsistencies in the NRC's regulations for fitness-for-duty (FFD) programs and clarify provisions regarding a nuclear power reactor licensee's insider mitigation program (IMP).

• Certified fuel handler definition and elimination of the shift technical advisor. This proposed rule would retain the existing definition for “certified fuel handler” and add an alternative that would eliminate the need for nuclear power reactor licensees to seek the Commission's approval of a fuel handler training program. The proposed provision would require the training program to address the safe conduct of decommissioning activities, safe handling and storage of spent fuel, and appropriate response to plant emergencies. The proposed alternative specifies that a CFH must be qualified in accordance with a fuel handler training program that meets the same requirements as training programs for non-licensed operators required by § 50.120, “Training and qualification of nuclear power plant personnel.” This proposed rule would also clarify that a Shift Technical Advisor (STA) is not required for decommissioning nuclear power reactors.

• Decommissioning funding assurance. This proposed rule recommends several changes regarding decommissioning funding for nuclear power reactors. It would modify the reporting frequency in § 50.75 to be consistent with the decommissioning funding assurance reporting frequency for ISFSIs in § 72.30(c). For ISFSI funding reports, this proposed rule would allow licensees to combine the reports that are required by § 50.82(a)(8)(v), § 50.82(a)(8)(vii), and § 72.30 and remove the requirement for NRC approval of ISFSI reports filed under § 72.30(c). It also would clarify that although the regulations establish a continuing obligation to provide reasonable assurance of decommissioning funding, when a licensee identifies a shortfall in the report required by § 50.75(f)(1), the licensee must obtain additional financial assurance to cover the shortfall and discuss that information in the next report. In addition, this proposed rule would make administrative changes to ensure consistency with § 50.4, “Written communications,” regarding the submission of notifications and to eliminate § 50.75(f)(2) because § 50.75(f)(1) fully encompasses paragraph (f)(2). Besides proposing conforming changes to 10 CFR part 52, the NRC is asking whether the NRC should maintain identical requirements in § 52.110 and § 50.82.

• Offsite and onsite financial protection requirements and indemnity agreements. This proposed rule would allow certain nuclear power reactor licensees in decommissioning to reduce the insurance amounts that they are required to maintain without obtaining exemptions from the NRC's regulations.

• Environmental considerations. This proposed rule would clarify that licensees must evaluate the environmental impacts of decommissioning and whether they are bounded by previous environmental reviews in the PSDAR. The proposed rule would also clarify environmental reporting requirements.

• Record retention requirements. This proposed rule would remove certain record retention requirements for structures, systems, and components (SSCs) that no longer remain in service during decommissioning and would remove requirements to keep multiple copies of certain spent fuel storage records. The NRC is also asking a specific question concerning the recordkeeping requirements for facilities licensed under 10 CFR part 52.

• Low-level waste transportation. This proposed rule would allow a 45-day window for notification of receipt of shipments of low-level radioactive waste (LLW). This increase from the current 20-day notification window is based on operating experience that shows that 45 days is an appropriate amount of time for notification of LLW shipments.

• Spent fuel management planning. This proposed rule would clarify requirements that the decommissioning documents contain information on spent fuel management planning in accordance with the regulatory requirements in § 72.218, “Termination of licenses.”

• Backfit rule. This proposed rule would clarify how the NRC applies § 50.109, “Backfitting,” to nuclear power reactor licensees in decommissioning and would make conforming changes to § 72.62.

• Foreign ownership, control, or domination. This proposed rule would specify the criteria for when a facility is no longer a production or utilization facility and that the foreign ownership, control, or domination (FOCD) prohibition found in § 50.38, “Ineligibility of certain applicants,” no longer applies to a person seeking a license for such a facility.

• Clarification of scope of license termination plan requirement. This proposed rule would clarify that the requirement for a license termination plan in §§ 50.82(a)(9) and 52.110(i) applies only to nuclear power reactor licensees that have loaded fuel into the reactor.

• Removal of license conditions and withdrawal of orders made redundant 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.

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.”

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) 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.

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).

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” sites where the associated reactor has already been decommissioned in comparison with “ISFSI-only” sites. As part of this rulemaking effort, the NRC recommends standardizing the terms “ISFSI-only” and “standalone ISFSI/Decommissioned Reactor” as follows:

• “ISFSI-only” sites contain nuclear power reactor facilities that are still involved in decommissioning activities, but the spent fuel has been completely transferred from the SFPs to dry storage in an onsite ISFSI. For these facilities, the remaining decommissioning activities are primarily related to remediation of any remaining residual radioactivity at the site to meet the license termination and decommissioning criteria in 10 CFR part 20, subpart E. The “ISFSI-only” term refers to the location of the spent fuel; the term reflects that no spent fuel is stored in the SFP, and all of the spent fuel is in dry storage in an onsite ISFSI.
• “Standalone ISFSI/Decommissioned Reactor” sites are those former nuclear power reactor facilities where the license termination and decommissioning criteria in 10 CFR part 20, subpart E, have already been met, with the exception of the ISFSI area. The licensee's 10 CFR part 50 license for the site has been reduced to an area that only encompasses the ISFSI facility (unless the facility ISFSI is licensed under a 10 CFR part 72 specific license, in which case the 10 CFR part 50 license is wholly terminated). The remaining activities at these facilities that are regulated by the NRC are spent fuel storage and the eventual decommissioning of the ISFSI itself, once the spent fuel has been permanently removed from the site. A 10 CFR part 72 specific license ISFSI is decommissioned in accordance with 10 CFR 72.54, “Expiration and termination of licenses and decommissioning of sites and separate buildings or outdoor areas.”

Accordingly, the proposed requirements would not apply to standalone ISFSI/Decommissioned Reactor sites because those licensees have already decommissioned their 10 CFR part 50 facilities and met the decommissioning and license termination criteria in 10 CFR part 20, subpart E, with the exception of the area encompassed by the remaining ISFSI. The proposed requirements are consistent with the licensing actions that the NRC has already approved for these licensees. In addition, the proposed requirements of this rulemaking provide an alternative to the existing decommissioning regulations and would not impose new requirements on ISFSI-only licensees.

F. Graded Approach

As the NRC reviewed the exemption and license amendment requests related to the recent nuclear power reactor decommissionings and noted the growing list of future planned permanent shutdowns, as discussed in the “ Background ” section of this document, the NRC realized that the existing regulatory framework could and should be revised to provide for a more efficient decommissioning process. As early as the late 1990's, the NRC contemplated an integrated rulemaking to provide an appropriate graded approach to the decommissioning process. A graded approach is a process by which the safety requirements and criteria adjust during the decommissioning process commensurate with several factors. These factors include the magnitude of any credible hazard involved, the particular characteristics of a facility, and the balance between radiological hazards and non-radiological hazards ( e.g., fire, flood, chemical spill) as applicable to specific points in time within the decommissioning process. This approach would be a risk-informed process.

Currently, no explicit regulatory provisions distinguish requirements in several technical areas for a nuclear power reactor that has permanently ceased operations from those for an operating nuclear power reactor. To address this, the NRC is proposing to amend its regulations to provide an efficient regulatory framework for the transition to decommissioning. Under this proposed rule, the NRC would adopt an optional graded approach for several technical areas that provides a set of requirements commensurate with the reductions in radiological risk at each of the following four levels of decommissioning: (1) Permanent cessation of operations and permanent removal of all fuel from the reactor vessel, (2) sufficient decay of fuel in the SFP such that it would not reach ignition temperature for the zirconium alloy cladding of the fuel within 10 hours under adiabatic heatup conditions ( i.e., a complete loss of SFP water inventory with no heat loss), (3) transfer of all fuel to dry storage, and (4) removal of all fuel from the site. Four technical areas of this proposed rule (Emergency Preparedness, Physical Security, Cyber Security, and Offsite and Onsite Insurance) use all or some of this graded approach.

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.

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 adequately protecting public health and safety. Proposed § 50.200(c)(1)(v)(H) would require licensees with PDEPs to establish a facility from which effective direction can be given and effective control can be exercised during an emergency. Because of the low probability of DBAs or other credible events that would be expected to exceed the limits of EPA PAGs offsite and the available time for event mitigation, the significantly reduced staff, and the minimal expected response required, offsite response would not be required at an EOF. Onsite actions may be directed from the control room or other location, without the requirements imposed on a TSC or EOF. Proposed § 50.200(b)(3) would remove reference to the EOF as a location for response. Additionally, under this proposed rule, a separate OSC would no longer be required to meet its original purpose of an assembly area for plant logistical support during an emergency. The OSC function could be incorporated into another facility. The NRC is issuing DG-1346 for public comment in conjunction with this proposed rule; DG-1346 provides one acceptable method for meeting the proposed emergency response facility requirements for PDEPs.

Appendix E to 10 CFR part 50, paragraph IV.E.9 addresses requirements for emergency communications systems, plans, and arrangements, including communications with OROs and between the control room, TSC, and EOF. Proposed § 50.200(c)(1)(v)(I) would require licensees with PDEPs to continue to maintain an onsite and an offsite communications system with backup power and communication plans with arrangements for emergencies. These arrangements would need to include provisions for communications with contiguous State and local governments, Federal emergency response organizations, NRC Headquarters, and the appropriate NRC Regional Office Operations Center. Because licensees with PDEPs may combine emergency response facilities, the current requirements for communication between emergency response facilities would not apply to these licensees. Under the proposed rule, communications with State and local emergency operations centers would be maintained to allow coordination of assistance onsite if required.

PDEP Hostile Action

Under this proposed rule, hostile action requirements would not apply to licensees with PDEPs. The definition of “hostile action” in appendix E to 10 CFR part 50, paragraph IV.A.7 applies here to the capability of implementing EP during hostile action events. However, in the statement of considerations (SOC) for the 2011 EP Final Rule, the NRC excluded non-power reactors from the definition of “hostile action” because a non-power reactor as defined in § 50.2, “Definitions,” is not a nuclear power plant, and a regulatory basis had not been developed to support the inclusion of non-power reactors in the definition of “hostile action.” A licensee with a PDEP would be similar to a non-power reactor in that both have a low likelihood of a credible accident resulting in radiological releases requiring response actions offsite. Additionally, regardless of how a disruption to the SFP cooling occurs, the spent fuel would take longer than 10 hours to heat up to ignition temperature, providing adequate time to coordinate a response between the ERO and law enforcement officials. As such, licensees with PDEPs would not fall within the scope of “hostile action,” and enhancements to EP in response to hostile action, such as alternative facilities for the staging of ERO personnel, protection of onsite personnel, and challenging drills and exercises involving hostile action, would not be warranted.

Although this rationale justifies the exclusion of licensees with PDEPs from the definition of “hostile action” and its related requirements (including conducting hostile action exercises) as they apply to EP, elements for security-based events would still be maintained for these facilities, including EALs for security-based events. Under the proposed rule, licensees with PDEPs would be required to identify ORO resources that would respond to a security event, and the assistance licensees expect from those resources would be maintained in PDEPs. For physical security, the objective for these facilities relates to protection of the spent fuel against sabotage. A level of security commensurate with the consequences of a sabotage event is required and is evaluated on a site-specific basis. The severity of the consequences declines as fuel ages and thereby removes over time the underlying concern that a sabotage attack, under the current definition, could cause offsite radiological consequences.

PDEP Drills and Exercises

Section 50.47(b)(14) and appendix E to 10 CFR part 50, paragraph IV.F provide training and drill and exercise requirements for nuclear power reactor licensees. Consistent with the language of § 50.47(b)(14), the proposed PDEP planning standard under § 50.200(b)(14) would require licensees with PDEPs to conduct periodic exercises to evaluate major portions of emergency response capabilities, to conduct periodic drills to develop and maintain key skills, and to correct deficiencies identified as a result of exercises and drills. The NRC is proposing new drill and exercise requirements for licensees with PDEPs under § 50.200(c)(1)(vi) that differ from the existing requirements under appendix E to 10 CFR part 50, paragraph IV.F to account for changes in principal functional areas, offsite radiological emergency response requirements, offsite PAR requirements, and the spectrum of possible accidents.

Similar to the requirements in appendix E to 10 CFR part 50, paragraph IV.F.1, proposed § 50.200(c)(1)(vi)(A) would require licensees with PDEPs to describe in their emergency plan provisions for the training of employees, exercising the emergency plan by conducting periodic drills, and including other individuals in training and drills when those individuals may provide assistance in the event of a radiological emergency. Under the proposed rule, the emergency plan would be required to describe the training to be provided to several categories of emergency personnel, with the exception of licensees' headquarters support personnel. Headquarters support personnel would no longer be required to augment the ERO for licensees with PDEPs. Licensees with PDEPs would need to continue to make available a radiological orientation training program for local services personnel expected to provide support onsite. Because of the time available to coordinate offsite agency notification to the public, licensees with PDEPs would not be required to provide radiological orientation training to local news media persons. Similar to the requirements in appendix E to 10 CFR part 50, paragraph IV.F.2, proposed § 50.200(c)(1)(vi)(B) would require licensees with PDEPs to continue to describe provisions for the conduct of EP exercises that test the adequacy of timing and content of implementing procedures and methods, test emergency equipment and communications networks, and ensure emergency organization personnel are familiar with their duties. Licensees with PDEPs would not be required to test the public alert and notification system during their exercises because the system would no longer be required, as discussed previously in this document.

Proposed § 50.200(c)(1)(vi)(B)( 1 ) and ( 2 ) would require licensees with PDEPs to conduct an exercise within two years of the last exercise of the onsite emergency plan conducted under paragraph IV.F.2.b of appendix E to 10 CFR part 50 and to continue to conduct subsequent biennial exercises of onsite emergency plans. Licensees with PDEPs would need to continue to conduct drills during the intervals between biennial exercises involving a combination of principal functional areas. The principal functional areas of emergency response for licensees with PDEPs would include all of the areas currently listed under appendix E to 10 CFR part 50, paragraph IV.F.2.b, with the exception of protective action development and protective action decision making (see discussion on protective action recommendations in the section “ PDEP Offsite Radiological Protective Action Recommendations” in this document).

Similar to the requirements in appendix E to 10 CFR part 50, paragraph IV.F.2.f, proposed § 50.200(c)(1)(vi)(B)( 4 ) would require licensees with PDEPs to conduct remedial exercises if the emergency plan is not satisfactorily tested during the biennial exercise. Like appendix E to 10 CFR part 50, paragraph IV.F.2.g, proposed § 50.200(c)(1)(vi)(B)( 5 ) would require licensees with PDEPs to provide for formal critiques of exercises, drills, and training that provide performance opportunities to develop, maintain, or demonstrate key skills and to correct weaknesses or deficiencies identified in a critique.

Proposed § 50.200(c)(1)(vi)(B)( 6 ) would require licensees with PDEPs to continue to use drills and exercise scenarios that provide reasonable assurance that anticipatory responses will not result from preconditioning of participants and that emphasize coordination among onsite and offsite response organizations. Unlike the current requirements under appendix E to 10 CFR part 50, paragraphs IV.F.2.b, IV.F.2.i, and IV.F.2.j, licensees with PDEPs would not be required to submit exercise scenarios 60 days before use in an exercise, demonstrate that exercise scenarios include a wide spectrum of radiological releases and events, or vary exercise scenarios across an eight calendar year exercise cycle to allow for the demonstration of responses to specified scenario elements, respectively. These requirements would no longer apply due to the limited types of events that could occur. The previously routine progression to a General Emergency, or even a Site Area Emergency, in nuclear power reactor site scenarios is not applicable for licensees with PDEPs.

The NRC is issuing DG-1346 for public comment in conjunction with this proposed rule; DG-1346 provides one method acceptable to the NRC for licensees with PDEPs to comply with the proposed drill and exercise requirements.

PDEP Offsite Response Organization Participation in Drills and Exercises

Appendix E to 10 CFR part 50, paragraph IV.F and § 50.47(b)(14) include requirements for periodic EP drills and exercises for licensees. Appendix E to 10 CFR part 50, paragraphs IV.F.2.c and IV.F.2.d requires offsite radiological emergency plans for each site to be exercised biennially with full participation by offsite authorities having a role under the radiological response plan. Appendix E to 10 CFR part 50, paragraphs IV.F.2.f and IV.F.2.h address State and local participation in remedial exercises and refusal of State and local governments to participate. Because no action is required from State and local government organizations in response to events other than firefighting, law enforcement, and ambulance/medical services, the requirements related to ORO participation in radiological drills and exercises would no longer be applicable to licensees with PDEPs. Proposed § 50.200(c)(1)(vi)(B) would remove the requirement to exercise offsite emergency plans once the NRC has docketed the licensee's certifications required under § 50.82(a)(1) or § 52.110(a) and the licensee elects under § 50.54(q)(7)(ii) to transition to a PDEP. For facilities that are located either on the same site or on adjacent contiguous sites to reactors that continue to operate, the offsite emergency plans would continue to be exercised as required under appendix E to 10 CFR part 50, paragraph IV.2.f, until all reactors at the site cease operation and transition to a PDEP. Similar to the requirements under appendix E to 10 CFR part 50, paragraph IV.2.f.e, under proposed § 50.200(c)(1)(vi)(B)( 3 ), a licensee with a PDEP would be required to enable any State or local government to participate in the licensee's drills and exercises when requested.

6. Independent Spent Fuel Storage Installation-Only Emergency Plans

In order to transition to an IOEP, the NRC is proposing under § 50.54(q)(7)(iii) that licensees must have all spent fuel in dry cask storage. Licensees with an IOEP must follow and maintain the effectiveness of an emergency plan that meets the requirements in § 72.32(a).

Licensees with 10 CFR part 72 specific licenses or under the 10 CFR part 72 general license may hold an IOEP. A licensee with all of its spent fuel in dry cask storage that terminates its 10 CFR part 50 or 10 CFR part 52 license must first obtain a 10 CFR part 72 specific license before transitioning to the EP requirements already provided in § 72.32(a). A licensee maintaining its 10 CFR part 50 or 10 CFR part 52 license, and thus its 10 CFR part 72 general license authorized under § 72.210, “General license issued,” may opt to change its EP program to align it with the requirements of § 72.32 once all spent fuel is transferred to dry cask storage. In addition, licensees under the 10 CFR part 72 general license would need to continue to comply with all applicable 10 CFR part 50 and 10 CFR part 52 requirements until the 10 CFR part 50 or 10 CFR part 52 license is terminated consistent with § 50.82 or § 52.110, respectively.

Under proposed § 50.54(q)(7)(iii), a licensee may choose not to comply with the EP requirements under § 72.32 and may instead maintain a PSEP or PDEP. Licensees with dry cask storage must ensure that the emergency plan includes an appropriate EAL scheme.

The NRC is issuing DG-1346 for public comment in conjunction with this proposed rule; DG-1346 provides guidance on transitioning to and maintaining an IOEP.

7. All Spent Fuel Removed From Site

During the fourth level of decommissioning, the proposed rule would allow a licensee to terminate its EP program under proposed § 50.54(q)(7)(iv) or proposed § 72.44(f). Once all spent fuel has been permanently removed from the site, the site no longer poses any risk of a radiological release. The licensee must then continue to follow its PSDAR submitted under § 50.82 until decommissioning is completed.

8. Changes to Emergency Plans

Existing § 50.54(q)(2) requires nuclear power reactor licensees to follow and maintain the effectiveness of an emergency plan that meets the planning standards in § 50.47(b) and the requirements in appendix E to 10 CFR part 50. In addition, § 50.54(q)(3) contains the conditions under which the licensee may make changes to its emergency plan without prior application to and approval by the NRC, provided that the changes do not reduce the effectiveness of the plan and that the plan, as changed, continues to meet the standards in § 50.47(b) and the requirements in appendix E to 10 CFR part 50. The NRC is proposing to add several new paragraphs that, similar to § 50.54(q)(2) and (3), would reference the requirements that emergency plans for decommissioning nuclear power reactors must meet and the process for making these plan changes. In particular, proposed § 50.54(q)(7) would reference the applicable emergency plan requirements after the NRC dockets a licensee's certifications under § 50.82(a)(1) or § 52.110(a), and proposed § 50.54(q)(8) would stipulate the conditions under which decommissioning nuclear power reactor licensees may make changes to their emergency plans without prior approval by the NRC. The NRC also would revise § 50.54(q)(1) to clarify that the definitions in paragraph (q) apply to only paragraph (q).

The existing change process under § 50.54(q) does not establish whether a proposed change would impact the agency's determination that there is reasonable assurance that a licensee can and will take adequate protective measures in the event of a radiological emergency; the change process establishes only whether the licensee has the authority to implement the proposed change without prior NRC approval. The change process uses the characteristic “reduction in effectiveness” to exclude from the requirement to seek prior NRC approval those changes that would likely not reduce the effectiveness of the licensee's emergency plan. Because these changes would not reduce the effectiveness of the plan, the NRC expects that the changes should not have an impact on the agency's reasonable assurance determination. A licensee's determination that a proposed change would reduce the effectiveness of the emergency plan does not mean that the licensee could not or would not implement adequate protective measures to protect public health and safety in the event of a radiological accident, but only that prior NRC review is required to evaluate the impact of the change on the reasonable assurance determination. As part of routine oversight, the NRC screens emergency plan changes, including EAL changes, and reviews a sample of changes documented in reports submitted under § 50.54(q)(5) that could potentially reduce effectiveness. These reviews do not constitute the NRC's approval of the plan changes, and all such changes remain subject to future inspection and enforcement actions. The NRC documents its approval of plan changes under § 50.54(q)(4) in its decisions to grant license amendment requests.

The licensee cannot properly evaluate a proposed change to the emergency plan if it has not considered the basis for the NRC's approval of the original plan or the basis for any subsequent changes to the plan—whether those changes were approved by the NRC or implemented by the licensee without prior NRC approval under § 50.54(q). Regulatory Guide (RG) 1.219, Revision 1, “Guidance on Making Changes to Emergency Plans for Nuclear Power Reactors” (ADAMS Accession No. ML16061A104), describes a method that the NRC considers acceptable to implement the requirements in § 50.54(q) as they relate to EP and specifically to making changes to emergency response plans. As provided in RG 1.219, the licensee should consider its licensing basis to inform a § 50.54(q) evaluation, and, principally, applicable regulatory requirements, which are binding on the licensee unless the NRC explicitly exempts the licensee from them. The NRC is issuing DG-1346 for public comment in conjunction with this proposed rule to provide guidance for decommissioning nuclear power reactors in evaluating changes to emergency plans under proposed § 50.54(q).

The change process is meant to ensure that emergency plans are maintained up to date and that the level of planning does not fall below the standards to which the licensee has committed. The regulations in § 50.54(q) define “reduction in effectiveness” as a change in an emergency plan that results in reducing the licensee's capability to perform an emergency planning function in the event of a radiological emergency. “Emergency planning function” is currently defined as a capability or resource necessary to prepare for, and respond to, a radiological emergency, as established in the planning standards of § 50.47(b) and the elements of appendix E to 10 CFR part 50, section IV. The NRC is proposing to remove the references to the planning standards of § 50.47(b) and appendix E to 10 CFR part 50 from this definition because this proposed rule would establish alternative emergency planning standards under proposed § 50.200, and the NRC does not consider the references essential to the definition.

When the NRC considers exemptions from EP requirements for a decommissioning nuclear power reactor licensee, the NRC considers whether there are special circumstances present as defined in § 50.12(a)(2). In particular, the NRC determines whether application of the EP regulations for which exemptions are under consideration in the particular circumstances would not serve their underlying purpose or are not necessary to achieve their underlying purpose, which is to provide reasonable assurance that adequate protective measures can and will be taken in the event of a radiological emergency. Once the NRC grants a licensee exemptions from EP requirements, the exempted regulations no longer apply to the licensee. The licensee may need to submit a separate license amendment request if the planned changes conflict with an element of the current licensing basis. If not, the licensee need not submit a separate license amendment request for NRC approval of the emergency plan changes unless the plan changes go beyond those resulting from the exemptions granted. The NRC intends that this proposed rule would establish clear regulatory requirements for EP, reducing the need to request certain exemptions. As such, the NRC is proposing to add § 50.54(q)(8) to establish the process for: (1) Transitions from one decommissioning level's EP planning standards and requirements to the next level's EP planning standards and requirements, and (2) changes to emergency plans within a decommissioning level.

In considering a graded approach to EP, the NRC recognizes that a transition between the EP planning standards and requirements of each decommissioning level is not equivalent to making changes to the emergency plan within a level. The transition between the EP planning standards and requirements of each decommissioning level is fundamentally a licensee's commitment to a different set of EP standards and associated emergency planning functions, and the change process should facilitate this transition.

For transitions from one decommissioning level to the next, the NRC would require licensees to establish emergency plans that meet the EP planning standards and requirements of the next level. The transition is optional, and a licensee that maintains its current level of emergency planning would satisfy the requirements of the next level; however, doing so would mean maintaining emergency planning functions above the commensurate level of planning for the risk involved. Under the proposed § 50.54(q)(8), a licensee would be able to make changes to the emergency plan to commit to the EP planning standards and requirements of the next decommissioning level ( i.e., PSEP, PDEP, or IOEP) using the § 50.54(q)(3) change process, but would only need to consider whether the changes meet the next level's planning standards and requirements. Licensees making changes to their emergency plans to commit to the EP planning standards and requirements of a decommissioning level would not be required to determine if the changes are reductions in effectiveness. Instead, the NRC would have already made this determination through its issuance of the regulations promulgating the EP planning standards and requirements of the decommissioning levels. The NRC's proposed regulatory approach to transitions between EP decommissioning levels does not go beyond the authority currently granted to licensees to make changes to their emergency plan under § 50.54(q)(3). Additionally, any change to the emergency plan that is not made to comply with the EP planning standards and requirements of the next decommissioning level would require a licensee to make a determination as to whether the change would be a reduction in effectiveness.

After the Three Mile Island accident in 1979, the NRC issued a final rule (45 FR 55402; August 19, 1980) (1980 EP Final Rule) that included § 50.54(u), which required licensees to upgrade their emergency plans to meet the then-new planning standards of § 50.47(b) and requirements in appendix E to 10 CFR part 50 and to submit those plans to the NRC. In the 2011 EP Final Rule, the NRC removed and reserved § 50.54(u). The NRC's proposed approach to transitions between EP planning standards and requirements of decommissioning levels is analogous to the approach taken by the NRC when the 16 EP planning standards went into effect in 1980 (see “Reasonable Assurance and Offsite Radiological Emergency Preparedness” section in this document). Under this approach, the NRC would not be relinquishing its oversight authority, as some commenters on the ANPR and draft regulatory basis supposed. As proposed, § 50.54(q)(8)(i) would require initial emergency plan changes made to transition between EP decommissioning levels to be submitted to the NRC at least 60 days prior to implementation, and emergency plans would remain subject to future inspection and enforcement. The proposed submittal is not intended to be a licensing action. It would provide a current copy of the emergency plan to the NRC prior to implementation in support of future inspection activities. This submittal would provide an opportunity for the NRC to assure that the licensee maintains the effectiveness of its emergency plan. Subsequent emergency plan changes would need to follow the existing change control process under § 50.54(q)(3) and (4). Hearing rights would not attach to transitions between EP decommissioning levels; however, the public has the opportunity to comment on the graded EP planning standards and requirements themselves in response to this proposed rule and the drafts of the supporting guidance documents. In addition, all emergency plan changes submitted under § 50.54(q)(5) and proposed § 50.54(q)(8) would be publicly available.

In addition to the general requirements in proposed § 50.54(q)(8)(i) governing transitions between EP decommissioning levels, proposed § 50.54(q)(8) would address changes specific to SSCs and EALs. Proposed § 50.54(q)(8)(ii) would specify that, for SSCs that are no longer needed to provide support for an emergency planning function (as defined under proposed § 50.54(q)(1)(iii)), a licensee may make a determination under § 50.54(q)(3) that emergency plan changes are not a reduction in effectiveness if the updated FSAR demonstrates that these SSCs are no longer required to be in service due to the decommissioning status of the facility. Proposed § 50.54(q)(8)(iii) would state that changes to EALs based on plant conditions that are not physically achievable or instrumentation that is no longer in service due to the transition to decommissioning are not reductions in effectiveness provided that a § 50.54(q)(3) evaluation demonstrates that the change does not reduce the capability of taking timely and appropriate protective actions. The NRC is proposing these requirements to provide clarity on § 50.54(q)(3) evaluations and alleviate the burden on licensees from submitting emergency plan changes that result from SSCs and instrumentation that are no longer required to be in service due to decommissioning.

After the implementation of a PSEP, PDEP, or IOEP, licensees would be required by proposed § 50.54(q)(7)(i) to continue to follow and maintain the effectiveness of the plan and by proposed § 50.54(q)(8)(i)-(iii) to comply with the change process described under existing § 50.54(q)(3) and (q)(4). Therefore, licensees would be allowed to make changes to these emergency plans without prior application to and approval by the NRC, provided that the changes would not reduce the effectiveness of the plan and that the plan, as changed, would continue to meet the EP planning standards and requirements for the applicable decommissioning level. Current § 50.54(q)(5) would require decommissioning licensees to submit to the NRC a report of each such change within 30 days after the change is put into effect. And, consistent with current requirements, decommissioning licensees would have to submit changes that would reduce the effectiveness of the plan for prior NRC review and approval in accordance with § 50.54(q)(4) so that the NRC could make the requisite reasonable assurance determination. For subsequent emergency plan changes once all fuel is in dry cask storage ( i.e., for changes to an IOEP), proposed § 50.54(q)(8)(i) would allow licensees to follow the change process under § 72.44(f).

The proposed amendments to the regulatory change process are necessary because:

• The regulation in existing § 50.54(q)(2), which provides that a licensee must follow and maintain the effectiveness of the emergency plan, should continue to apply in order to ensure that emergency plans are followed and kept up to date.
• The existing § 50.54(q) change process and the associated regulatory guidance currently do not address how a licensee could change its emergency plans to comply with the emergency plan standards as the licensee transitions to each level of decommissioning.
• This proposed rule would allow the NRC to maintain, through a regulatory change process, reasonable assurance that a licensee can and will take adequate protective measures in the event of a radiological emergency.

The proposed amendments to § 50.54(q), and related regulatory guidance, would ensure that licensees would maintain the effectiveness of the emergency plans. Emergency plans that comply with the proposed graded EP planning standards and requirements would continue to provide reasonable assurance that adequate protective measures can and will be taken in the event of a radiological emergency. Any plan that did not meet these standards and requirements and, if applicable, the reduction in effectiveness criterion, would be subject to inspection and enforcement actions. The proposed approaches to transitioning between EP decommissioning levels and to making emergency plan changes within decommissioning levels would provide an efficient and effective regulatory change process and would promote consistent and predictable implementation and enforcement.

9. Program Element Review Under § 50.54(t)

Under current § 50.54(t), licensees must conduct reviews of EP program elements either: (1) At intervals not to exceed 12 months or (2) as necessary, based on an assessment by the licensee against performance indicators and as soon as reasonably practicable after a change occurs in personnel, procedures, equipment, or facilities that potentially could adversely affect EP. If a licensee chooses the second option, it must still review all program elements at least once every 24 months. For several reasons, the proposed rule would provide decommissioning licensees with an alternative approach to reviewing EP program elements.

First, the NRC expects licensees to remain in the first level of decommissioning ( i.e., with a PSEP) for less than 24 months, and the scope of a PSEP is largely unchanged from the scope of an operating reactor's emergency plan. Conversely, the second level of decommissioning ( i.e., licensees with a PDEP) will involve more significant changes, and the NRC anticipates that licensees would remain in the second level of decommissioning for a longer period of time. Therefore, in order to support program continuity and minimize changes during the transition to a PDEP, the NRC is proposing to amend § 50.54(t) such that, starting after licensees enter the second level of decommissioning, licensees would be able to conduct program element reviews under § 50.54(t) at intervals not to exceed 24 months (rather than 12 months) without conducting an assessment against performance indicators. The NRC is proposing to add new § 50.54(t)(3) to remove the requirement to conduct periodic EP program element reviews once all fuel is in dry cask storage ( i.e., the third/IOEP level of decommissioning), consistent with the EP requirements for ISFSIs under 10 CFR part 72.

10. Reasonable Assurance and Offsite Radiological Emergency Preparedness

The regulations in §§ 50.47 and 50.54, “Conditions of licenses,” prescribe how the NRC will make licensing decisions or take appropriate enforcement actions by using findings of reasonable assurance that adequate protective measures can and will be taken to protect public health and safety in the event of a radiological emergency. Every 10 CFR part 50 or 10 CFR part 52 license includes as a condition of the license the requirements of § 50.54(s)(2)(ii) and (s)(3) regarding findings and determinations of reasonable assurance. The NRC has the authority and responsibility to make licensing findings on the overall adequacy of onsite and offsite emergency planning and preparedness. Commensurate with the NRC's responsibility to make such findings, the NRC has the authority to collect, review, and evaluate any information it needs to support its findings on EP. If available, the NRC must consider FEMA findings and determinations regarding the status of offsite EP. The relationship between the NRC and FEMA concerning findings of reasonable assurance of offsite EP is based on the Atomic Energy Act of 1954, as amended (AEA); the Energy Reorganization Act of 1974, as amended; the NRC Authorization Act for Fiscal Year 1980, the NRC's regulations; a memorandum of understanding between the two agencies (“Memorandum of Understanding Between the Department of Homeland Security/Federal Emergency Management Agency and Nuclear Regulatory Commission Regarding Radiological Emergency Response, Planning, and Preparedness”) first established in 1980 and last updated in 2015 (ADAMS Accession No. ML15344A371); and case law ( e.g., Massachusetts v. United States, 856 F.2d 378, 382 (1st Cir. 1988); State of Ohio ex rel. Celebrezze v. NRC, 868 F.2d 810, 815-16 (6th Cir. 1989)).

Not all licensing decisions involving EP require findings and determinations on the adequacy of offsite plans. In the EP regulations for research and test reactors, fuel cycle facilities, and ISFSIs, there are no regulatory standards or requirements for offsite radiological emergency plans. As such, FEMA findings and determinations are not needed to support NRC licensing decisions for such facilities. The absence of NRC regulatory standards for offsite radiological EP at those facilities does not imply that offsite emergency planning, in general, is not adequate to protect the public health and safety. In addition, the support provided by offsite organizations does not automatically necessitate the need for findings and determinations. In the Low Power Rule (47 FR 30232; July 13, 1982), the NRC concluded that findings and determinations on the state of offsite EP were not needed to support issuance of a license for fuel loading and low-power testing because there was sufficient time (at least 10 hours) in which to take action to protect the public in even the worst-case accident. Additionally, the NRC has concluded in its review of several EP exemption requests for permanently shutdown and defueled nuclear power reactor licensees that formal offsite radiological emergency plans are not necessary after the spent fuel in the SFP has sufficiently decayed such that it would not reach zirconium fuel cladding ignition temperature within 10 hours under adiabatic heatup conditions. As a result, continued consultation with FEMA regarding the adequacy of the offsite plans was also no longer necessary.

For decommissioning nuclear power reactors, the NRC is proposing that if regulatory standards for offsite radiological EP are not required, then findings and determinations on the adequacy of offsite plans would not be needed in order for the NRC to make determinations regarding reasonable assurance under § 50.54(s)(2)(ii). Therefore, the NRC is proposing changes to § 50.54(s)(3) to clarify that FEMA findings and determinations are only necessary when the NRC's planning standards apply to offsite radiological emergency response plans. Additionally, the NRC staff is proposing to add a new § 50.47(f) to clarify when the 16 planning standards apply to offsite radiological emergency plans. A licensee must follow and maintain the effectiveness of its emergency plan if the NRC is to continue to find, under § 50.54(s)(2)(ii), that there is reasonable assurance that adequate protective measures can and will be taken in the event of a radiological emergency, and § 50.54(s)(2)(ii) would continue to apply to licensees as a condition of the license during decommissioning.

In 1979, the NRC predicated the rationale for the EP proposed rule (44 FR 75167; December 19, 1979) on the Commission's considered judgment in the aftermath of the accident at Three Mile Island. At the time, the Commission concluded that it must be in a position to know that offsite governmental plans had been reviewed and found adequate. However, the Commission also noted that the proposed rule was considered an interim upgrade of NRC emergency planning regulations based on past experience, and that further changes to emergency planning regulations may be proposed as more experience is gained. The NRC viewed the 1979 proposed rule as a first step in improving emergency planning.

The NRC recognizes the experience gained from implementing its regulations and also that significant advances in emergency planning have occurred over the decades following the accident at Three Mile Island. In particular, the terrorist attacks on September 11, 2001, led to the establishment of the U.S. Department of Homeland Security, and lessons learned from disasters such as Hurricane Katrina have resulted in a national effort to prepare for and respond to all hazards and disasters. Homeland Security Presidential Directive 5, “Management of Domestic Incidents” (February 28, 2003), and Presidential Policy Directive (PPD)-8, “National Preparedness” (issued March 30, 2011), established national initiatives for a common approach to preparedness and response. These initiatives include the National Incident Management System, National Preparedness Goal, Core Capabilities, National Preparedness System, National Planning Frameworks, and the development of comprehensive preparedness guides and exercise methodologies.

The PPD-8 directed the development of a national preparedness goal that identifies the core capabilities necessary for preparedness and a national preparedness system to guide activities that will enable the nation to achieve the goal. Core capabilities are intended to help coordinate and unify efforts, improve training and exercise programs, promote innovation, and ensure that the administrative, finance, and logistics systems are in place to support these capabilities. The PPD-8 is aimed at facilitating an integrated, all-of-nation, capabilities-based approach to preparedness, under the assumption that national preparedness is the shared responsibility of the “whole community,” which includes all levels of government, the private and nonprofit sectors, and individual citizens. Acknowledging the national preparedness goal, the NRC maintains the sole legal authority to establish any regulations it deems necessary to ensure the adequate protection of public health and safety from radiological events.

For a decommissioning site, the licensee, as part of the whole community, will maintain radiological EP capabilities. Only in the highly unlikely event of a zirconium fire—in which mitigation actions were not successful—would there be a potential need to initiate response actions offsite. But unlike the EP planning basis for an operating reactor, within a few months of cessation of operations, there is no longer a potential need to provide for prompt protective actions in the event of an accident. Additionally, protective actions such as evacuation are not unique to radiological events and occur in response to other unique hazards such as chemical spills, fires, and natural disasters, and are often initiated without any pre-planning. In NUREG-0396, the NRC states that “It has been, and continues to be the Federal position that it is possible (but exceedingly improbable) that accidents could occur calling for additional resources beyond those that are identified in specific emergency plans developed to support specific individual nuclear facilities. Further, the NRC and Federal position has been and continues to be, that as in other disaster situations, additional resources would be mobilized by State and Federal agencies.”

State and local governments are responsible for the protection of public health and safety (including at industrial sites like decommissioning reactors), and the NRC has high confidence in the ability of OROs to implement appropriate response actions when necessary. This confidence is further strengthened by the NRC's recognition of national-level efforts, in which the NRC participates, to improve the state of emergency planning at all levels of government and within the whole community. Consequently, for facilities licensed by the NRC where radiological hazards are unlikely to have an offsite impact, the risk posed by the remaining low-level hazard is somewhat analogous to that posed by non-nuclear hazards ( e.g., train derailments or oil spills) that are addressed by all-hazards planning and not by a separate radiological emergency plan. In such conditions, there is reasonable assurance that appropriate response actions can and will be taken in the event of a radiological emergency, without the need for regulatory standards for offsite radiological emergency response plans and the associated FEMA findings and determinations that offsite plans are adequate and can be implemented.

11. Clean-Up of Regulations

The NRC is proposing to remove obsolete dates for certain one-time actions that were required as part of the 2011 EP Final Rule and other obsolete dates. These actions are complete, and the requirements are no longer binding on any current licensee. The dates of requirements proposed to be removed are:

(1) Section 50.54(s)(2)(ii), which allows the NRC to shut down nuclear power reactors that did not provide reasonable assurance that adequate protective measures would be taken in the event of a radiological emergency after April 1, 1981. There is no longer a need for the date requirement of this provision because any future determinations made under § 50.54(s) will be after April 1, 1981. The NRC is proposing to delete “after April 1, 1981” and retain the remainder of the provision.

(2) Paragraph 6 of appendix E to 10 CFR part 50, section I, which was used to promulgate specific compliance dates for the Tennessee Valley Authority Watts Bar Nuclear Plant that was under construction at the time of the 2011 EP Final Rule. Because the Watts Bar Nuclear Plant is now operational and subject to all current requirements for operating reactors, the NRC is proposing to delete this provision.

(3) Appendix E to 10 CFR part 50, paragraph IV.4, which required nuclear power licensees to develop an ETE analysis using decennial data published within 365 days of the later date of the most recent decennial data or December 23, 2011. There is no longer a need for the date requirement of this provision because the date has expired. The NRC is proposing to delete “of the later of the date of” and “or December 23, 2011” from this provision.

(4) Appendix E to 10 CFR part 50, paragraph IV.A.7, which required licensees to identify and describe the expected assistance from appropriate local, State, and Federal agencies during an emergency, including a hostile act, by June 23, 2014. The NRC is proposing to delete “by June 23, 2014” from this provision because the date has expired.

(5) Appendix E to 10 CFR part 50, paragraph IV.A.9, which required licensees to conduct a detailed analysis by December 24, 2012, demonstrating that on-shift personnel are not assigned responsibilities that would prevent the timely performance of assigned functions in the emergency plan. The NRC is proposing to delete “By December 24, 2012” from this provision because the date has expired.

(6) Appendix E to 10 CFR part 50, paragraph IV.B.1, which required licensees, by June 20, 2012, to establish EALs that include hostile action that may adversely affect the nuclear power plant. There is no longer a need for the date requirement of this provision because the date has expired. The NRC is proposing to remove “By June 20, 2012” and retain the remainder of the provision.

(7) Appendix E to 10 CFR part 50, paragraph IV.C.2, which required licensees, by June 20, 2012, to establish and maintain capability to assess, classify, and declare an emergency condition within 15 minutes after indications that an EAL had been exceeded. There is no longer a need for the date requirement of this provision as the date has expired. The NRC is proposing to delete “By June 20, 2012” and retain the remainder of the provision.

(8) Appendix E to 10 CFR part 50, paragraph D.4, which included compliance periods for the backup alert and notification capability requirements under appendix E to 10 CFR part 50, paragraph D.3, including a final deadline of June 22, 2015. The NRC is proposing to remove this paragraph because the dates in the paragraph have expired, and any future applicants required to comply with appendix E to 10 CFR part 50 would be required to comply with the requirements of appendix E to 10 CFR part 50, paragraph D.3.

(9) Appendix E to 10 CFR part 50, paragraph IV.E.8.c, which required licensees' EOFs to have the capabilities required under the section by June 20, 2012. Because the date requirement of this provision has expired, the NRC is proposing to delete “By June 20, 2012” from this provision.

(10) Appendix E to 10 CFR part 50, paragraph IV.E.8.d, which required licensees to identify an alternative facility that would be accessible in the event of hostile action by December 23, 2014, with the exception of the capability for staging ERO personnel at the alternative facility and communications capabilities with emergency responses facilities, which had to be implemented by June 20, 2012. There is no longer a need for the date requirements of this provision as the dates have expired. The NRC is proposing to delete the deadlines for the implementation of this provision.

(11) Appendix E to 10 CFR part 50, paragraph IV.F.2.d, which required licensees to fully participate in one hostile action by December 31, 2015. Because the date requirement of this provision has expired, the NRC is proposing to delete “and should fully participate in one hostile action exercise by December 31, 2015” from this provision.

(12) Appendix E to 10 CFR part 50, paragraph IV.F.2.j, which required licensees to conduct a hostile action exercise for each of their sites no later than December 31, 2015. Because the date requirement of this provision has expired, the NRC is proposing to delete the requirement from this provision.

(13) Appendix E to 10 CFR part 50, paragraph IV.I, which required licensees, by June 20, 2012, to provide a range of protective actions to protect onsite personnel during hostile action. Because the date requirement of this provision has expired, the NRC is proposing to delete “By June 20, 2012” from this provision.

(14) Appendix E to 10 CFR part 50, paragraph VI.4.a, which required licensees to develop and submit an ERDS implementation plan to the NRC by October 28, 1991. There is no longer a need for the date requirement of this provision because the date has expired. The NRC is proposing to delete “by October 28, 1991” from this provision.

(15) Appendix E to 10 CFR part 50, paragraph VI.4.d, which required licensees to complete the implementation of the ERDS by February 13, 1993, or before escalation to full power, whichever comes later. There is no longer a need for the date requirement of this provision because the date has expired. The NRC is proposing to delete “by February 13, 1993, or” and “whichever comes later” from this provision and to continue to require licensees to submit an ERDS implementation plan to NRC before escalation to full power.

The NRC is proposing to eliminate these completed one-time requirements in the interest of regulatory clarity. Eliminating these requirements would not relax any currently effective regulatory requirement or cause any regulatory burden for current or future licensees or applicants.

12. Revisions to § 72.32

The NRC proposes to amend § 72.32(a) to address the applicability of that provision's requirement that an application for a specific license ISFSI must include an emergency plan that includes the information in § 72.32(a)(1) through (16). The proposed amendment would clarify that the requirement applies when the proposed ISFSI would not be located on the site or within the exclusion area of a nuclear power reactor licensed under 10 CFR parts 50 or 52. A nuclear power reactor licensed under 10 CFR parts 50 or 52 could be under construction, operating, or in decommissioning. The proposed revisions would consolidate the current language and remove redundancies by using standardized language consistent with other amendments in this proposed rule.

The NRC proposes to amend § 72.32(c) to clarify that the nuclear power reactor referenced in that provision need not be authorized to operate for the ISFSI licensee to use the emergency plan requirements in § 50.47 to meet the requirements of § 72.32. Currently, § 72.32(c) applies to ISFSI licensees located on the site or within the exclusion area of a nuclear power reactor that is licensed to operate. Because a nuclear power reactor licensee is not authorized to operate once the NRC dockets the certifications required under § 50.82(a)(1) or § 52.110(a), § 72.32(c) could be read not to apply to an ISFSI licensee at a decommissioning reactor site. However, the current language of § 72.32 allows an ISFSI licensee with a reactor emergency plan to use that emergency plan to meet the applicable requirements for an ISFSI emergency plan. Therefore, the proposed rule would clarify that, when the nuclear power reactor is under construction, operating, or in decommissioning, the ISFSI licensee could rely on the emergency plan requirements in appendix E to part 50 of this chapter and 10 CFR 50.47(b), or the requirements of 10 CFR 50.200(a) or 10 CFR 50.200(b), to meet the requirements of § 72.32.

B. Physical Security

The NRC's regulations governing physical security at a nuclear power reactor typically do not distinguish between an operating nuclear power reactor and a nuclear power reactor that is in a decommissioning status. However, the security risk profile presented by a decommissioning reactor decreases significantly from that of an operating nuclear power reactor due to the reduction in the number of target sets and the reduced consequences of radiological sabotage. The radiological consequences of a security event decrease as reactors transition through each of the following four levels of decommissioning: (1) Permanent cessation of operations and permanent removal of all fuel from the reactor vessel, (2) sufficient decay of fuel in the SFP such that it would not reach the zirconium fuel cladding ignition temperature within 10 hours under adiabatic heatup conditions, (3) transfer of all fuel to dry storage, and (4) removal of all fuel from the site. Decommissioning nuclear power reactor licensees have sought NRC approval of exemptions from, license amendments for, and alternative measures to, certain physical security regulatory requirements because of the reduction in the number of target sets and the reduced consequences of radiological sabotage as the nuclear power reactor site transitions through these levels. The NRC is proposing options to allow nuclear power reactor licensees to make certain commonly-requested changes to their physical security plans based on these decommissioning levels without requesting exemptions, alternative measures, or license amendments.

1. Security Plans

Upon the cessation of operations and removal of all fuel from the reactor vessel, licensees typically seek to modify their security plans to reflect changes in site conditions. The NRC's regulations in § 50.54(p) establish processes that allow licensees to make changes to their security plans. Section 50.54(p)(1) requires licensees to seek NRC review and approval of any changes that result in a decrease in safeguards effectiveness of their security plans. Section 50.54(p)(2) allows licensees to make changes to their security plans without prior NRC approval provided that the changes do not decrease the safeguards effectiveness of the plan.

The current regulations do not define the term “decrease in safeguards effectiveness” nor do they include examples of the types of changes that would constitute a decrease in safeguards effectiveness. Additionally, there is no definition of the term “change.” This lack of clear definitions has resulted in difficulties for licensees implementing security plan changes. For example, some licensees have implemented changes under § 50.54(p)(2) that the NRC later determined decreased the safeguards effectiveness of their security plan. Similarly, some licensees have unnecessarily requested NRC review and approval of changes that did not decrease the safeguards effectiveness of their security plan.

The NRC is proposing to revise § 50.54(p) to include definitions of the terms “change” and “decrease in safeguards effectiveness.” The application of these definitions would be limited to the revised § 50.54(p) and would apply to all 10 CFR part 50 and 10 CFR part 52 licensees with operating, decommissioning, and/or decommissioned reactor units. The term “change” would be defined in a new § 50.54(p)(1)(i) to mean an action that results in a modification of, addition to, or removal from, the licensee's security plans. The term “decrease in safeguards effectiveness” would be defined in a new § 50.54(p)(1)(ii) to mean a change or series of changes to an element or component of the security plans referenced in § 50.54(p)(2) that reduces or eliminates the licensee's ability to perform or maintain the capabilities established in § 73.55(b)(3)(i) without compensating changes to other security plan elements or components.

Currently, decommissioning (and operating) reactor licensees use the § 50.54(p)(2) process to implement changes that they have determined do not decrease the safeguards effectiveness of their security plans. The § 50.54(p)(2) process requires that licensees submit a report of these changes to the NRC. In addition to a description of these changes, reactor licensees have typically included in their report supplemental information demonstrating that such changes do not constitute a decrease in safeguards effectiveness. The submittal of this supplemental information in the reports has been voluntary. The NRC's practice is to review these reports to confirm that the licensee properly concluded that the changes would not decrease the safeguards effectiveness of their Commission-approved security plan. The submittal of supplemental information in the reports allows the NRC to verify in a timely manner that the change does not result in a decrease in the safeguards effectiveness of the plan. Without this supplemental information, the NRC could only make this determination through the inspection process. The NRC is proposing to require that reactor licensees include with the required § 50.54(p)(2) report a summary of the analysis performed to determine that the change does not decrease safeguards effectiveness of the security plan. The summary must be sufficient to demonstrate that the change does not decrease the safeguards effectiveness of the plan.

2. Dry Cask Storage

An ISFSI located at a nuclear power reactor site is typically licensed under a general license issued pursuant to subpart K of 10 CFR part 72. Under a general license, licensees are required to protect the SNF in the ISFSI in accordance with the physical security requirements in § 73.55, “Requirements for physical protection of licensed activities in nuclear power reactors against radiological sabotage,” with the additional conditions and exceptions noted in § 72.212, “Conditions of general license issued under § 72.210.” The NRC also licenses certain ISFSIs under a 10 CFR part 72 specific license. Consistent with § 72.180, “Physical protection plan,” licensees holding a specific license are required to protect the SNF in the ISFSI in accordance with the physical security requirements in § 73.51, “Requirements for the physical protection of stored spent nuclear fuel and high-level radioactive waste.” Although the physical security requirements that apply to general license ISFSIs and specific license ISFSIs provide equivalent levels of protection, there are differences. For instance, § 73.55 requires licensees to ensure they maintain the capability to detect, assess, interdict, and neutralize threats. Section 73.51 requires licensees to detect and assess threats and communicate with an appropriate response organization. The additional requirements in § 73.55 that support interdiction and neutralization of threats is only one example of differences that lead to licensee requests for exemptions once all fuel has been placed in dry cask storage.

As stated at the beginning of this section, decommissioning reactors typically transition through four distinct levels during decommissioning. Many decommissioning licensees have submitted license amendment requests, requests for exemptions, and requests for approval of alternative measures to remove § 73.55 physical security requirements that are no longer applicable once the licensee enters the third decommissioning level when all SNF has been moved to a dry cask storage system.

The need for license amendments, exemptions, and approvals of alternative measures imposes a regulatory burden upon both licensees and the NRC. Accordingly, the NRC is proposing that once all SNF has been placed in dry cask storage, licensees may elect to follow the proposed § 72.212(b)(9)(vii) and protect a general license ISFSI in accordance with the physical security requirements in § 73.51. The applicability section of § 73.51 would also be amended to reflect this change. A licensee would be able to use the process established in the revised and renumbered § 50.54(p)(3) to make this change and submit its revised physical security plan to the NRC. These security plans would have to continue to address the applicable security-related orders associated with an ISFSI that are conditions of the license. The NRC is also proposing conforming changes to § 72.13, “Applicability,” to reflect the requirements that would apply to a licensee that elects to follow the proposed § 72.212(b)(9)(vii).

3. Significant Core Damage

The prevention of significant core damage and spent fuel sabotage is a general performance objective of the reactor licensee physical protection program required by § 73.55. During the first level of decommissioning, when the NRC has docketed a licensee's certifications that the reactor has permanently ceased operating and all fuel has been removed from the reactor vessel and placed in the SFP, there is no longer fuel in the core and therefore the risk to public health and safety from significant core damage has been removed. This reduced risk allows licensees to eliminate requirements to protect against significant core damage or train security and operational personnel to protect and respond to core damage events.

The NRC is proposing that a licensee of a decommissioning nuclear power reactor no longer be required to meet the requirement in § 73.55(b)(3) to protect against significant core damage once the NRC has docketed a licensee's certifications that the reactor has permanently ceased operating and all fuel has been removed from the reactor vessel. The requirement in § 73.55(b)(3) to protect against spent fuel sabotage remains in effect as long as spent fuel remains in the spent fuel pool.

4. Vital Areas

A vital area (VA) is defined in § 73.2, “Definitions,” as any area that contains vital equipment. Under § 73.2, vital equipment means any equipment, system, device, or material, the failure, destruction, or release of which could directly or indirectly endanger public health and safety by exposure to radiation. The NRC also considers the equipment or systems that would be required to function to protect public health and safety following such a failure, destruction, or release to be vital. There are specific physical security requirements for the protection of VAs and vital equipment. The current regulation in § 73.55(e)(9)(v) specifies that the reactor control room shall be considered a VA.

The role of the reactor control room at an operating plant, as described in Criterion 19, “Control room,” of appendix A, “General Design Criteria for Nuclear Power Plants,” to 10 CFR part 50, is to provide a protected space from which actions can be taken to operate the nuclear power plant safely without interruption under normal or accident conditions. For a permanently shutdown and defueled nuclear power reactor, the vital equipment associated with operating the reactor vessel is no longer needed. The remaining vital equipment ( e.g., associated with SFP cooling) may no longer be needed or may be relocated to a VA separate from the reactor control room. Once a reactor has permanently ceased operations, the need for a reactor control room is eliminated if all of the vital equipment is removed and if the area does not serve as the VA boundary for other VAs. The proposed rule would revise § 73.55(e)(9)(v) to provide that a licensee of a decommissioning nuclear power reactor would no longer need to designate the reactor control room as a VA if it does not otherwise meet the definition of a VA in § 73.2.

5. Communications

Currently § 73.55(j)(4)(ii) requires continuous and redundant communications between the reactor control room and the central alarm station (CAS). Once a nuclear power reactor has permanently ceased operations, a licensee may no longer have a reactor control room or a licensed senior operator present in a reactor control room. Therefore, it would not be feasible for a licensee of a decommissioning nuclear power reactor to comply with the current regulatory requirement. Licensees typically request an exemption from this requirement and request that the CAS be allowed to establish continuous and redundant communications with the senior on-site licensee representative.

The NRC is proposing to amend § 73.55(j) to require continuous and redundant communications be maintained between the CAS and the CFH or senior on-shift licensee representative once the reactor has ceased operations and the licensee no longer has licensed senior operators in the control room. The intention of this change is to allow licensees flexibility in maintaining communications with one or both of these individuals.

Communication requirements will continue to include all the conditions currently required: Continuous communication capability with onsite and offsite resources; radio or microwave transmitted two-way voice communication, in addition to conventional telephone service, between the alarm stations and local law enforcement authorities; and alternative communication measures in place in areas where communication could be interrupted or cannot be maintained.

6. Suspension of Security Measures

Current regulations in § 73.55(p) allow for the suspension of security measures in an emergency or during severe weather. A senior licensed operator must approve the suspension of security measures. Once a nuclear power reactor has entered decommissioning status and all fuel has been removed from the reactor, there may no longer be a licensed senior operator on site. Therefore, it may not be feasible for a licensee of a decommissioning nuclear power reactor to implement this requirement in the event of an emergency or severe weather.

The NRC is proposing to amend the requirements in § 73.55(p) to allow a CFH to suspend security measures in the event of an emergency or severe weather once the reactor has shutdown and all fuel has been removed from the reactor core.

These proposed changes to § 73.55(p) would be consistent with the existing regulations in § 50.54(x) and (y) that govern approvals for reasonable actions that a licensee may take to depart from a license condition or a technical specification in an emergency. In accordance with the provisions of § 50.54(y), licensee actions permitted by § 50.54(x) must be approved (at a minimum) by a licensed senior operator or, at a decommissioning nuclear power reactor after submittal of the certifications required under § 50.82(a)(1) or § 52.110(a), by either a licensed senior operator or a CFH, before taking the action.

C. Cyber Security

The NRC is proposing to update cyber security requirements in § 73.54, “Protection of digital computer and communication systems and networks” for nuclear power reactor licensees. This update would clarify the cyber security requirements applicable to a nuclear power reactor during each stage of the decommissioning process.

As stated in § 73.54, applicants and licensees must provide high assurance that their digital computer and communication systems and networks associated with safety and important-to-safety, security, and emergency preparedness (SSEP) functions are adequately protected against cyber attacks, up to and including the design basis threat described in § 73.1, “Purpose and scope.” To accomplish this, each holder of a nuclear power reactor operating license under 10 CFR part 50 has submitted a cyber security plan (CSP) to the NRC that has been approved by the NRC. Further, each combined license (COL) applicant is required to submit its CSP as part of its COL application for review and approval. Each approved CSP is referenced in a license condition in each 10 CFR part 50 license, and this license condition requires a licensee to maintain its CSP until the license is terminated or the license condition is removed by license amendment. A COL holder does not have an equivalent cyber security license condition.

The cyber security requirements in § 73.54 apply to licensees currently licensed to operate a nuclear power plant. Once the NRC has docketed a licensee's § 50.82(a)(1) or § 52.110(a) certifications, that licensee is no longer authorized to operate a nuclear power plant. Therefore, the requirements in § 73.54 would no longer apply to such a licensee. However, each 10 CFR part 50 licensee has a license condition requiring the licensee to maintain its CSP, and this license condition remains in effect during decommissioning. A COL holder, without the license condition, is not required to maintain its CSP when it begins decommissioning.

Although a licensee that has submitted its § 50.82(a)(1) or § 52.110(a) certifications is no longer operating, such a licensee may still have fuel recently removed from the reactor vessel in its SFP. As discussed in the “ Technical Basis for Graded Approach” section of this document, if the spent fuel in the SFP has not sufficiently decayed, there is a risk that the spent fuel could heat up to clad ignition temperature and lead to a zirconium fire for postulated draindown scenarios in a timeframe that is too short to reliably implement mitigation measures or to take other appropriate response actions.

As discussed in the “ Technical Basis for Graded Approach” section of this document, in Level 2 there is little chance that the spent fuel in the SFP could heat up to clad ignition temperature within 10 hours. Accordingly, the NRC is proposing that the cyber security requirements in § 73.54 continue to apply to licensees through Level 1. This continuation of the cyber security requirements would ensure that a compromise of digital systems cannot adversely impact the effective operation of the licensees' physical security programs and emergency preparedness functions prior to the time at which the spent fuel cannot reasonably heat up to clad ignition temperature within 10 hours after a draindown event. Although the cyber security requirements would continue to apply through Level 1, the number of critical digital assets would decrease as systems are removed from service, which in turn reduces the number of critical digital assets that must be protected by the CSP.

To clarify the applicability of the cyber security rule to decommissioning nuclear power reactor licensees, the NRC is proposing to add two paragraphs to § 73.54. A new § 73.54(i) would state that the requirements of § 73.54 will remain in effect until: (1) The NRC has docketed the licensee's § 50.82(a)(1) or § 52.110(a) certifications, and (2) at least 10 months for a BWR or 16 months for a PWR have elapsed since the date of permanent cessation of operations or an NRC-approved alternative to the 10 or 16 month spent fuel decay period, submitted under proposed § 50.54(q)(7)(ii)(A) or (B), has elapsed. A new § 73.54(j) would state that, after both requirements of § 73.54(i) have been met, the licensee's license condition that requires implementation and maintenance of a cyber security plan would be removed from the license. The NRC is also proposing the removal of the introductory paragraph of § 73.54 in its entirety and revising the language of § 73.54(a), (b), and (c). These are conforming changes to clarify that the applicability of § 73.54 is not limited to “operating” reactors ( i.e., that § 73.54 would still be applicable after the NRC has docketed a licensee's § 50.82(a)(1) or § 52.110(a) certifications), to remove language that is no longer needed concerning the initial submission of cyber security plans by existing licensees, and to add clarifying language to § 73.54(b) and (c). Further, the NRC is proposing a change to § 73.55(c)(6), which requires the licensee to establish, maintain, and implement a cyber security plan. This is a conforming change to reflect the scenario in which a decommissioning nuclear power reactor licensee is no longer required to maintain a cyber security plan ( i.e., the NRC has docketed the certifications of permanent cessation of operations and permanent removal of fuel from the reactor vessel, and the fuel in the SFP has sufficiently decayed), but is still required to comply with § 73.55(c).

The proposed revision to § 73.54(a) would not constitute backfitting for 10 CFR part 50 licensees. The proposed revision would constitute a change affecting the issue finality of COL holders; extending the requirement to maintain a CSP during decommissioning would be a new requirement imposed on COL holders. The NRC's proposed backfit analysis is located in the “Backfitting and Issue Finality” section of this document.

D. Drug and Alcohol Testing

1. Scope of 10 CFR Part 26

The NRC is proposing to amend § 26.3, “Scope,” to correct an inconsistency within § 26.3(a) where the FFD requirements in 10 CFR part 26 apply differently to 10 CFR part 50 and 10 CFR part 52 licensees with decommissioning nuclear power reactors. The § 26.3(a) provision lists those licensees that are required to comply with designated subparts of 10 CFR part 26, including licensees who are authorized to operate a nuclear power reactor under § 50.57 and holders of a combined license under 10 CFR part 52 after the Commission has made the finding under § 52.103(g). In accordance with this requirement, 10 CFR part 26 does not apply to a holder of a nuclear power reactor license issued under 10 CFR part 50 that is no longer authorized to operate a nuclear power reactor because the NRC has docketed the certifications required under § 50.82(a)(1) ( i.e., a decommissioning 10 CFR part 50 nuclear power reactor licensee). However, 10 CFR part 26 continues to apply to holders of combined licenses issued under 10 CFR part 52 throughout decommissioning. Therefore, there is an inconsistency in the application of FFD requirements to nuclear power reactor licensees during decommissioning.

The NRC has determined that there is no technical basis for this inconsistency. In the 1989 10 CFR part 26 final rule (54 FR 24468; June 7, 1989) (1989 FFD Final Rule), the Commission explained that the intent of that rule was to address the potential for worker impairment of any kind, including substance abuse that could affect the safe operation of nuclear power plants. The emphasis throughout the 1989 FFD Final Rule is that the rule is necessary to promote public health and safety when the plant is operational. The wording for 10 CFR part 52 licensees described in the scope of the 2008 10 CFR part 26 final rule (73 FR 16966; March 31, 2008) (2008 FFD Final Rule), specifically § 26.3(a), was an oversight. The emphasis of the 1989 FFD final rule that FFD need only apply to operating 10 CFR part 50 sites should be the same for 10 CFR part 52 licensees. Due to the decreased risk to public health and safety during decommissioning, 10 CFR part 26 should not apply to these licensees during decommissioning.

Therefore, the NRC proposes to clarify that 10 CFR part 26 does not apply to 10 CFR part 52 licensees once the NRC has docketed their § 52.110(a) certifications. Section 26.3(a) of the proposed rule would specify that each holder of an operating license for a nuclear power reactor under 10 CFR part 50 and each holder of a COL under 10 CFR part 52 for which the Commission has made the finding under § 52.103(g) must comply with the requirements of 10 CFR part 26, except for subpart K of 10 CFR part 26, until the NRC's docketing of the license holder's certifications described in §§ 50.82(a)(1) or 52.110(a).

For clarity, the NRC proposes to divide the current paragraph of § 26.3(a) into two paragraphs. Paragraph (a)(1) would retain the requirement in the second sentence of current § 26.3(a) to state the deadline by which licensees must implement their FFD program. Paragraph (a)(2) would retain the requirement in the first sentence of current § 26.3(a) that these licensees must comply with the requirements of 10 CFR part 26, except subpart K, but clarify that this requirement ends when the NRC dockets the licensee's §§ 50.82(a)(1) or 52.110(a) certifications.

2. Fitness-for-Duty Elements for Insider Mitigation Program

Under § 73.55(b)(9), a licensee is required to establish, maintain, and implement an IMP to monitor the initial and continuing trustworthiness and reliability of individuals granted unescorted access authorization (UAA) or unescorted access (UA) to a protected area (PA) or vital area (VA).

Section 73.55(b)(9)(ii)(B) requires that an IMP must contain elements of an FFD program described in 10 CFR part 26. However, the regulations do not identify which FFD program elements must be included in the IMP. Section 73.55(b)(9)(ii)(B)( 1 ) and ( 2 ) of this proposed rule would amend § 73.55(b)(9)(ii)(B) to establish an appropriate set of FFD provisions to be incorporated into the IMP of operating and decommissioning 10 CFR part 50 and 10 CFR part 52 licensees to provide reasonable assurance that individuals granted UAA or UA to the PA or VA are trustworthy and reliable.

Section 73.55(b)(9)(ii)(B)( 1 ) of this proposed rule would clarify § 73.55(b)(9)(ii)(B) that licensees implementing 10 CFR part 26, regardless of whether they are required to do so, are in compliance with § 73.55(b)(9)(ii)(B). A licensee's full 10 CFR part 26 FFD program ( i.e., an FFD program that complies with all applicable 10 CFR part 26 requirements) would contain FFD elements appropriate for inclusion in the licensee's IMP. This would apply to both operating and decommissioning licensees.

Section 73.55(b)(9)(ii)(B)( 2 )( i ) and ( ii ) of this proposed rule describes the minimum 10 CFR part 26 elements necessary for a 10 CFR part 50 and 10 CFR part 52 decommissioning licensee's IMP. Section 73.55(b)(9)(ii)(B)( 2 )( i ) of the proposed rule states that individuals who have unescorted access to the VAs at a decommissioning site, perform certified fuel handler functions ( i.e., individuals covered by § 50.2) prior to all spent nuclear fuel at a site being placed in dry cask storage, perform security—related functions ( i.e., individuals covered by § 26.4(a)(5)), or administer the drug testing program ( i.e., individuals covered by § 26.4(g)) are subject to the requirements in 10 CFR part 26 except for subparts I and K. Individuals who have fuel handler certifications are essential to the safe movement of spent nuclear fuel. Individuals who have security-related responsibilities or perform work around the spent fuel pool may have knowledge of value to an adversary. In addition, security personnel generally carry weapons on site and would pose a significant challenge to site security if they were to perform as an active insider during an attack. Testing of individuals who administer a drug testing program is viewed as essential to the integrity of the program.

Proposed § 73.55(b)(9)(ii)(B)( 2 )( ii ) states that individuals who have UA to the protected area, but do not perform certified fuel handling or security-related functions or administer the drug testing program would still be subject to pre-access and for-cause testing (§ 26.31(c)(1) and (2)) and behavior observation (§ 26.33), but would not be subject to random testing (§ 26.31(c)(5)). The NRC proposes to relax these requirements because while the reactor is in decommissioning the potential contribution of certain personnel to support an adversary as an insider is greatly reduced. Individuals who do not have any security-related responsibilities or regular SFP area UA will have less potential contribution as an insider threat.

The NRC has determined that the FFD elements necessary for an IMP under this proposed rule are commensurate with the hazard and potential event consequences associated with a facility's operational status. Section 73.55(b)(3) states that the physical protection program must be designed to prevent significant core damage and spent fuel sabotage. Operating nuclear power reactor facilities contain many target sets located throughout the PA of potential interest to an adversary seeking to affect core damage or spent fuel sabotage, thus anyone who has UAA or UA to the PA could contribute significantly to an adversary.

The hazard and potential event consequences associated with decommissioning facilities significantly decrease in comparison to those associated with the operating facilities. During decommissioning, the SFP becomes the primary focus of the licensee's obligation to protect against the radiological sabotage design basis threat, as it becomes the location where all spent fuel is located when a nuclear power reactor is no longer operating and prior to transitioning to an ISFSI. With this perspective, this proposed rule tailors applicability of the FFD elements commensurate with the duties and access of personnel who have been granted UAA and maintain UA to the PA or VA.

3. Criminal Penalties

The NRC proposes to amend the criminal penalties section of 10 CFR part 26 by including § 26.3 within § 26.825(a). Existing § 26.825(a) applies the NRC's authority under the AEA to impose criminal penalties for willful violations of, attempts to violate, or conspiracies to violate NRC regulations. Section 26.825(b) lists § 26.3 as one of the 10 CFR part 26 provisions that is excluded from § 26.825(a). In general, the criminal penalties sections of NRC regulations apply to substantive requirements, and administrative or procedural regulatory provisions are excluded from criminal penalties sections. The current § 26.3 is entitled “Scope” and identifies which entities are within the scope of 10 CFR part 26. Scoping provisions typically do not contain substantive requirements, which may explain why § 26.825(b) includes § 26.3. However, the current § 26.3(a) not only describes the entities that are subject to the requirements of 10 CFR part 26 but also includes a substantive requirement for certain entities to comply with requirements in 10 CFR part 26 by a specific deadline. This requirement was added to § 26.3(a) in the 2008 FFD Final Rule, but § 26.825(b) was not updated to reflect this change, which was an oversight. This proposed rule would not change the substantive requirement in § 26.3(a). Because proposed § 26.3(a) would continue to impose a substantive requirement, the NRC proposes to remove § 26.3 from § 26.825(b), thereby including § 26.3 in § 26.825(a).

E. Certified Fuel Handler Definition and Elimination of the Shift Technical Advisor

The NRC is proposing two revisions to its regulations. The first change would be to amend the definition of a CFH in § 50.2 to provide an alternative that would eliminate the need for licensees to seek NRC approval for fuel handler training programs by adding a provision that requires the training program to address the safe conduct of decommissioning activities, safe handling and storage of spent fuel, and appropriate response to plant emergencies, and specifies that a CFH must be qualified in accordance with a fuel handler training program that meets the same requirements as training programs for non-licensed operators required by § 50.120. This proposal would provide consistency in the regulatory treatment of the training programs for non-licensed operators (which do not require NRC approval) and fuel handler training programs to qualify a non-licensed operator as a CFH (which do require NRC approval). The second change would clarify that an STA is not required for decommissioning reactors. These changes would provide clarity to the CFH's responsibilities and functions and the role of an STA by codifying current licensing practices. This proposed rule would also clarify the management role of the CFH in a manner that is consistent with § 50.54(y) as discussed in section “ B. Physical Security” in this document.

1. Alternative Definition for Certified Fuel Handler

The current definition of a CFH in § 50.2 does not specify what is in an NRC-approved fuel training program. Licensees have submitted requests for the approval of CFH training and retraining programs in connection with their decommissioning. After receiving NRC approval of a CFH training program, the licensee typically submits a license amendment request to propose changes to the Administrative Controls section of its Technical Specifications (TS) to include a CFH, among other applicable changes based on the approval of the CFH training program.

For example, on May 12, 2014, the NRC approved the Shift Manager/Certified Fuel Handler training program for Kewaunee Power Station (ADAMS Accessions No. ML14104A046). The NRC's safety evaluation supporting approval of the CFH training program used criteria that focused on whether the licensee trained CFHs on the following three objectives: (1) Safe conduct of decommissioning activities; (2) safe handling and storage of spent fuel; and (3) appropriate response to plant emergencies. These three objectives have subsequently been the basis for other NRC approvals of CFH training programs for licensees entering or planning to enter the decommissioning process: Entergy for VY (ADAMS Accession No. ML14162A209); Exelon for Oyster Creek Nuclear Generation Station, Clinton Power Station, and Quad Cities Nuclear Power Station (ADAMS Accession No. ML16222A787); and Entergy for FitzPatrick Nuclear Power Plant (ADAMS Accession No. ML16259A347).

In the safety evaluations for those approved CFH training programs, the NRC discusses the 1996 Final Rule and its role in the development of the objectives for an acceptable CFH training program. The NRC recognized that the risks posed at decommissioning reactors are significantly less than those posed by operating reactors. The NRC noted specifically that:

• While the spent fuel is still highly radioactive and generates heat caused by radioactive decay, no neutron flux is generated and the fuel slowly cools as its energetic decay products diminish.

• The systems required for maintaining the spent fuel in the spent fuel pool as well as the operations required to contain the remaining residual contamination in the facility and spent fuel pool are relatively simple.
• Because the spent fuel is stored in a configuration that precludes a nuclear fission reaction, no generation of new radioactivity can occur and the potential for consequences that could result from an inadvertent nuclear reaction are highly unlikely.

Because of the reduced risks and relative simplicity of the systems needed for safe storage of the spent fuel, the NRC explained in the 1996 Final Rule that the degree of regulatory oversight required for a nuclear power reactor during its decommissioning stage is considerably less than that required for the facility during its operating stage. In the 1995 decommissioning proposed rule (60 FR 37374; July 20, 1995), the NRC provided insights as to the responsibilities of the proposed new position of the CFH. Specifically, the NRC stated that a CFH is an individual who has the requisite knowledge and experience to evaluate plant conditions and make judgments about emergency action decisions necessary to protect the public health and safety.

In addition to using the three objectives to evaluate the fuel handler training programs for licensees entering or planning to enter decommissioning, the NRC applied the criteria in § 50.120, “Training and qualification of nuclear power plant personnel,” and assessed the proposed fuel handler training programs against the elements of a systems approach to training (SAT) as defined in § 55.4, “Definitions.” Section 50.120 identifies individuals required to be subject to an SAT, including non-licensed operators such as CFHs, and necessary elements for training programs. These elements include the requirement to periodically evaluate and revise the training program, as appropriate, to reflect changes to the facility ( e.g., decommissioning), procedures, regulations, and quality assurance requirements.

Because it has developed succinct criteria to approve fuel handler training programs, the NRC proposes to include this criterion in its regulations as an alternative definition of a CFH to eliminate the need for licensees to submit requests for NRC approval of CFH training programs. Specifically, the NRC would codify current approval practices by amending § 50.2 to add the three broad-scope objectives as responsibilities for which a CFH must be trained: (1) Safe conduct of decommissioning activities; (2) safe handling and storage of spent fuel; and (3) appropriate response to plant emergencies. In addition, the CFH would have to qualify in accordance with a fuel handler training program that meets the same requirements as training programs for non-licensed operators required by § 50.120. Should a licensee not exercise the alternative definition, it would need to submit a request for approval of a fuel handler training program.

2. Elimination of the Shift Technical Advisor

The STA is a position identified in licensees' TSs. The STA provides engineering expertise in the diagnosis of complex problems with SSCs during reactor operation. Once a licensee enters the decommissioning process, the STA function is no longer needed. The current regulations do not address the acceptability of discontinuing the STA position for a decommissioning reactor. Licensees have been removing the STA position and replacing that position with a CFH in their TSs through license amendments (see Duke Energy Florida for Crystal River Unit 3 Nuclear Generating Plant (ADAMS Accession No. ML14097A145); Exelon for Oyster Creek Nuclear Generating Station (ADAMS Accession No. ML16235A413); and Entergy for VY (ADAMS Accession No. ML14217A072)). The NRC proposes to revise a footnote to the table titled “Minimum Requirements Per Shift for On-Site Staffing of Nuclear Power Units by Operators and Senior Operators Licensed Under 10 CFR part 55” in § 50.54(m)(2)(i) to state that an STA is not required upon the NRC's docketing of the license holder's certifications required under §§ 50.82(a)(1) or 52.110(a).

F. Decommissioning Funding Assurance

The NRC proposes to amend its regulations to modify decommissioning funding reporting requirements, clarify decommissioning funding assurance requirements, and eliminate duplicative regulations.

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

The NRC is proposing to amend the regulations in § 50.82(a)(8)(i)(A) and § 52.110(h)(1)(i) to remove the term “legitimate.” This term does not add any substance to the regulations and is potentially confusing. The intent of the regulation is to ensure that expenses fall within the NRC definition of decommission. Whether an expense falls within the definition of decommission would continue to be determined on a case-by-case basis by the licensee when considering whether to make a withdrawal from the decommissioning trust fund. Since this term is non-substantive, its removal would not change any of the existing requirements regarding the use of decommissioning funds.

2. Changes to Reporting Requirements

In the “Financial Assurance Requirements for Decommissioning Nuclear Power Reactors” final rule (63 FR 50465; September 22, 1998), the NRC added the provisions currently in § 50.75(f)(1) and (2) that require each nuclear power reactor licensee to file a report with the NRC on the status of its decommissioning funding for each reactor that it owns, by March 31st of every odd-numbered year or annually for plants that are within five years of their projected end of operation. This report must specify: (1) The amount of decommissioning funds estimated to be required pursuant to § 50.75(b) and (c); (2) the amount of decommissioning funds accumulated to the end of the calendar year preceding the date of the report; (3) a schedule of the annual amounts remaining to be collected; (4) the assumptions used regarding rates of escalation in decommissioning costs, rates of earnings on decommissioning funds, and rates of other factors used in funding projections; (5) any contracts upon which the licensee is relying; (6) any modifications occurring to a licensee's current method of providing financial assurance since the last submitted report; and (7) any material changes to trust agreements.

The NRC is proposing to change the reporting frequency in § 50.75(f)(1) to coordinate the reporting frequency with the ISFSI decommissioning reporting frequency in § 72.30. This change would convert the biennial decommissioning funding status report required for 10 CFR part 50 and 10 CFR part 52 nuclear power reactor licensees to a triennial decommissioning funding status report as currently is required for 10 CFR part 72 ISFSI licensees. This revision would not change the annual reporting frequency for a reactor licensee that is within 5 years of its projected end of operations, whether that projection is based on the license's expiration date or on a premature shutdown, and would not change the annual reporting frequency for a reactor that has permanently ceased operations. Also, the change in reporting frequency would not relieve the licensee from calculating annual adjustments as required under § 50.75(a)(2) and would not affect the Table of Minimum Amounts in § 50.75(c) or its escalation factors. Therefore, a licensee would be required to continue to monitor its decommissioning funding on an annual basis but instead of reporting at least once every 2 years to the NRC, it would report at least once every 3 years.

Since 1999, the NRC's regulations have mandated that licensees report to the NRC the status of their decommissioning funding. Under § 50.75(f)(1), the biennial decommissioning funding status report requires the disclosure of seven items, including the balance of the decommissioning trust fund as of December 31st of the prior year. The NRC conducted spot checks of licensee records related to this information. The NRC did not identify any major discrepancies related to this information, as explained in SECY-15-0005 (ADAMS Accession No. ML14210A554), dated January 15, 2015. Therefore, the NRC has confidence that changing from a biennial to a triennial reporting frequency will not subject the public to any additional risks associated with decommissioning funding assurance. In addition, even with a triennial reporting frequency, there would be ample time to resolve any decommissioning funding issue. Furthermore, the proposed revision does not change the requirement for more frequent reporting as a licensee approaches the permanent cessation of operations and while the licensee is in decommissioning or the requirement for a site-specific decommissioning cost estimate during this period.

The NRC proposes a rule change in § 50.75(h) in order to be consistent with the requirements of § 50.4. Specifically, notifications would be sent directly to the Document Control Desk, and not to the Director, Office of Nuclear Reactor Regulation, or Director, Office of Nuclear Material Safety and Safeguards, as applicable. This change would provide one consistent location for licensees to docket all notifications to the NRC.

The NRC proposes to delete § 50.75(f)(2). The language of existing § 50.75(f)(1) fully encompasses the language of paragraph (f)(2), and, therefore, paragraph (f)(2) is unnecessary and potentially confusing. By removing paragraph (f)(2) the NRC would not be removing the requirement on licensees to continue submitting decommissioning funding assurance status reports. Existing paragraphs (f)(3) through (5) would be redesignated as paragraphs (f)(2) through (4).

3. Shortfalls in Decommissioning Funding Assurance

The requirement in § 50.75 that the licensee provide reasonable assurance that sufficient funds will be available for radiological decommissioning is a continuing obligation. However, economic factors can cause the amount of a licensee's financial assurance to fall below the amount required (either by the NRC minimum formula in § 50.75(c), or by a licensee's site-specific decommissioning cost estimate), thereby creating a shortfall. The regulations do not explicitly discuss what to do when a licensee faces a funding shortfall, regardless of its cause. Instead, the NRC addressed the scenario in its guidance in RG 1.159, “Assuring the Availability of Funds for Decommissioning Nuclear Reactors” (ADAMS Accession No. ML003740066). This guidance provides that non-rate-regulated licensees should make up shortfalls in decommissioning funding within 2 years and electric utility licensees within 5 years.

The NRC is proposing to amend its regulations in § 50.75(f)(1) to clarify that, although the regulations establish a continuing obligation to provide reasonable assurance of decommissioning funding, when a licensee identifies a shortfall in the report required by § 50.75(f)(1), the licensee must identify additional financial assurance to cover the shortfall in the next report. Specifically, the proposed rule would require licensees to remedy shortfalls before permanent cessation of operations consistent with the methods identified in § 50.75(e) in the next § 50.75(f) report. The proposed rule would clarify the expectations for how reasonable assurance of funds will be available for the decommissioning process. For electric utilities that currently submit biennial reports but correct their shortfalls within 5 years, the NRC proposes that they would submit their decommissioning funding status reports triennially and explain in their reports how they plan to correct any existing shortfall. Electric utilities should continue to correct shortfalls within 5 years as explained in RG 1.159. For non-rate-regulated licensees that currently submit biennial reports and should correct shortfalls within a 2 year period, the NRC proposes that they correct any shortfalls within the 3 year reporting period. The NRC proposes to clarify the last sentence of current § 50.75(f)(1) to reduce the number of clauses and enhance readability.

The NRC proposes to revise § 50.82(a)(9)(ii)(F) to require licensees to identify the specific sources of funds for “remaining decommissioning costs,” including sources of funds for license termination, spent fuel management, and ISFSI decommissioning.

4. Conforming Changes to 10 CFR Part 52

The NRC proposes to revise § 52.110 to make the same changes proposed in § 50.82 for the reasons previously discussed and for consistency. In addition, the NRC proposes to add paragraphs (h)(5) through (h)(7) with site-specific decommissioning cost estimate reporting requirements that are identical to the requirements in § 50.82(a)(8)(v) through (vii). Consistent with proposed § 52.110(h)(7), a report on irradiated fuel should only be submitted if irradiated fuel is on site.

5. Change to 10 CFR Part 72

The NRC proposes to revise § 72.30 so that the submittals subsequent to the initial decommissioning funding plan would no longer require NRC approval. The NRC found little benefit in approving subsequent decommissioning funding plans for ISFSIs because the financial assurance mechanisms employed are very similar to those used for nuclear power reactors. The experience to date is that decommissioning funding plans have not changed substantively because of the passive nature of the ISFSI design, the static nature of ISFSI operations after loading, and the fact that there are no liquids or liquid effluents present in dry cask storage facilities. In addition, the NRC expects that the frequency of events that could potentially impact the decommissioning funding plan ( i.e., due to spills, facility modifications, or changes in possession limits that are cited in § 72.30(c)) would continue to be low. However, if they were to occur, it is important that these events be factored into the cost of decommissioning. This change would make the processes under § 72.30(c) more efficient and less burdensome to the licensee and the NRC, while still maintaining reasonable assurance of adequate funding for the decommissioning of ISFSIs.

G. Offsite and Onsite Financial Protection Requirements and Indemnity Agreements

The NRC proposes to amend its financial protection regulations under 10 CFR part 140, “Financial Protection Requirements and Indemnity Agreements,” and § 50.54(w) to address instances where a decommissioning reactor licensee may not need to maintain its full amounts of offsite liability insurance and onsite property insurance. Reductions in insurance amounts may be warranted commensurate with the reduction in probability of an incident at a reactor in decommissioning and also a reduction in the offsite and onsite consequences from this event. The proposed financial protection requirements would codify the approach currently used by the NRC to approve exemptions from the financial protection requirements for decommissioning 10 CFR part 50 and 10 CFR part 52 nuclear power reactor licensees. The proposed changes would also increase efficiency and transparency in this area by clarifying the requirements for financial protection of decommissioning plants, providing for regulatory certainty, and reducing regulatory burden without affecting public health and safety. Specifically, these proposed requirements would represent a graded approach, including the criteria to be considered, where the financial protection requirements for decommissioning sites are adjusted commensurate with the level of risk posed at two stages of the decommissioning process.

Proposed revisions to 10 CFR part 140 and § 50.54(w) would also address other regulatory topics including, for example, the applicability of procedures regarding extraordinary nuclear occurrences and a proposed new notification requirement for licensees when they make changes to the amount of onsite insurance.

1. Proposed Revisions to Offsite Liability and Onsite Property Insurance Requirements

The NRC proposes to allow 10 CFR part 50 and 10 CFR part 52 nuclear power reactor licensees in decommissioning to reduce the offsite liability and onsite property insurance amounts that they are required to maintain under §§ 140.11 and 50.54(w), respectively, without obtaining exemptions from the NRC's regulations. Instead, as proposed under §§ 140.11(a)(5) and 50.54(w)(5), once certain criteria are satisfied, licensees could reduce their financial protection to the amounts in Level 2 in Table 3:

Licensees in Level 1 of the graded approach would be required to maintain the full amounts of offsite liability and onsite property insurance currently required in §§ 140.11(a)(4) and 50.54(w), respectively, until the probability of a zirconium fuel cladding fire in the spent fuel pool is minimized. Maintaining the full level of insurance recognizes the potential for liability insurance claims following an accident of this type and the need for available resources to clean up the site.

The transition to Level 2 financial protection amounts for licensees would be optional and could occur after the passage of a specified amount of time ( i.e., 10 months for BWRs or 16 months for PWRs, beginning on the date of permanent cessation of operations, plus the NRC's docketing of the certifications required by § 50.82(a)(1) or § 52.110(a) or after the lapse of an NRC-approved alternative time period to the 10 or 16 month spent fuel decay period that is submitted under § 50.54(q)(7)(ii)(A) or (B)). For the latter option, licensees would need to submit an analysis that demonstrates a reduced risk of a zirconium fuel cladding fire in the SFP. The reduction in the financial protection amounts as identified in Table 3 ( i.e., $100 million in offsite liability insurance and withdrawal from the industry retrospective rating plan) was modeled on the offsite liability claims experience from the accident at Three Mile Island Unit 2 as documented in SECY-93-127, “Financial Protection Required of Licensees of Large Nuclear Power Plants During Decommissioning” (ADAMS Accession No. ML12257A628). SECY-93-127 provides a reasonable basis for using the Three Mile Island Unit 2 experience as a model for determining the appropriate liability insurance coverage level for a permanently shutdown reactor that has completed its respective spent fuel cooling period. Additionally, as documented in SECY-93-127, the reduced onsite financial protection amount in Table 3 ( i.e., $50 million in onsite property insurance coverage) was modeled on the potential onsite cleanup costs from a radiological incident involving the rupture of a large liquid radioactive waste storage tank (~450,000 gallons) containing slightly radioactive water. This event was selected as conceivable and a bounding scenario having negligible radiological consequences offsite.

The spent fuel heat-up analysis performed by the licensee for purposes of reducing its insurance amounts to those in Level 2 could be the same analysis that the licensee performs to relax the offsite emergency planning requirements under proposed § 50.54(q)(7)(ii)(A) or (B). The transition to Level 2 would prompt the licensee to notify the NRC under § 140.15(e) of a material change in financial protection—a reduction in offsite primary financial protection from $450 million to $100 million and withdrawal from the industry retrospective rating plan. The NRC proposes a conforming change to § 50.54(w) for a similar notification of a material change to onsite property insurance amounts.

The NRC is also proposing to periodically adjust the offsite and onsite financial protection amounts for decommissioned reactors to account for inflation. These adjustments would be in accordance with the aggregate percentage change in the Consumer Price Index and performed at intervals that coincide with the inflation adjustments for the retrospective premium under Section 170t of the AEA.

2. Proposed Revision to Extraordinary Nuclear Occurrences Requirements

The NRC proposes to amend its regulations in § 140.81, “Scope and purpose,” to clarify the applicability of the requirements for an Extraordinary Nuclear Occurrence (ENO) to reactors in decommissioning. Under Sections 11 and 170 of the AEA, and NRC regulations at subpart E, “Extraordinary Nuclear Occurrences,” to 10 CFR part 140, the NRC is authorized to make a determination as to whether an event at a production or utilization facility causing a discharge or dispersal of source, special nuclear, or byproduct material that has resulted or will result in substantial damages to offsite members of the public or property is an ENO. An event will qualify as an ENO if the NRC determines that the criteria in § 140.84, “Criterion I—Substantial discharge of radioactive material or substantial radiation levels offsite,” and § 140.85, “Criterion II—Substantial damages to persons offsite or property offsite,” have been met.

The NRC recognizes that the radiological consequences resulting from an accident at a decommissioning reactor in Level 1 can be similar to those from an accident at an operating reactor. As presented in NUREG-1738, in the timeframe beginning immediately after the reactor is defueled and the fuel is placed in the SFP, the radiological consequences of a zirconium fire may be comparable to those from operating reactor postulated severe accidents. The existing potential consequences from a zirconium fire, until the fuel in the SFP has sufficiently decayed, provides the basis for the NRC's proposal to amend its regulations to include plants in decommissioning within the scope of § 140.81.

3. Proposed New Rule Language in § 50.54(w)(6)

The NRC proposes to amend § 50.54(w) to require a prompt notification to the Commission of any material change in proof of onsite property insurance filed with the Commission under 10 CFR part 50. Specifically, the transition to Level 2 as proposed by the NRC would prompt the licensee to notify the NRC under § 50.54(w)(6) of a reduction in onsite property insurance from $1.06 billion to $50 million. This proposed amendment to § 50.54(w)(6) would be a conforming change, for consistency, with the existing offsite financial protection requirements under § 140.15(e).

H. Environmental Considerations

1. Clarifying Changes to 10 CFR Parts 50 and 52

A nuclear power reactor licensee's transition from operating to decommissioning status does not involve an agency action that would trigger NRC responsibilities under environmental statutes, such as the National Environmental Policy Act (NEPA), the Endangered Species Act (ESA), or the National Historic Preservation Act (NHPA). However, § 50.82(a)(4)(i) (for nuclear power reactors licensed under 10 CFR part 50) and § 52.110(d)(1) (for nuclear power reactors licensed under 10 CFR part 52) require that PSDARs provide the reasons for concluding that appropriate previously issued environmental impact statements (EIS) will bound the environmental impacts associated with site-specific decommissioning activities. After the PSDAR is submitted, the licensee must remain in compliance with § 50.82(a)(6)(ii) or § 52.110(f)(2), as applicable. These regulations state that licensees may not perform any decommissioning activities, as defined in § 50.2, that result in significant environmental impacts not previously reviewed. As explained in the 1996 Final Rule, the requirement in § 50.82(a)(6)(ii) functions as a prohibition against the licensee performing a decommissioning activity that would result in a significant impact “not previously reviewed” (61 FR 39283, 39286, and 39291; July 29, 1996). The NRC may develop updates to IMC 2561, “Decommissioning Power Reactor Inspection Program,” and the related Inspection Procedure (IP) 71801, “Decommissioning Performance and Status Review at Permanently Shutdown Reactors,” dated August 11, 1997, to provide guidance on inspections for compliance with § 50.82(a)(6)(ii) or § 52.110(f)(2) with respect to environmental reviews.

In certain circumstances, licensees may be unable to satisfy the requirement that licensees conclude in the PSDAR that all environmental impacts associated with site-specific decommissioning activities will be bounded by previous EISs. For example, NUREG-0586, Supplement 1, Volumes 1 and 2, “Generic Environmental Impact Statement on Decommissioning of Nuclear Facilities: Regarding the Decommissioning of Nuclear Power Reactors” (Decommissioning GEIS) (ADAMS Accession No. ML023470327), identified several resource areas that were not generically resolved. If the EISs previously prepared for the construction and initial operation of the plant, for license renewal, or for another licensing action did not include site-specific analyses for those resource areas not generically resolved under the Decommissioning GEIS, then the licensee would be unable to make the determination in the PSDAR that all impacts will be bounded. Therefore, the licensee would have to either change its planned decommissioning activities so that the impacts would be bounded or submit and have approved a license amendment request or an exemption request to satisfy § 50.82(a)(4)(i) or § 52.110(d)(1) prior to conducting the subject decommissioning activity.

The NRC proposes to change the PSDAR requirements in § 50.82(a)(4)(i) and § 52.110(d)(1) to require that licensees provide the basis for determining whether the environmental impacts from site-specific decommissioning activities are bounded by previous environmental reviews. This proposed rule change would clarify that licensees, at the PSDAR stage, are required to evaluate the environmental impacts and provide in the PSDAR the basis for whether the proposed decommissioning activities are bounded by previously issued, site-specific or generic environmental reviews. Given that some decommissioning activities will occur well in the future, licensees might not be able to make the definitive conclusion that impacts will be bounded at the PSDAR stage. Therefore, the proposed change would provide licensees flexibility to address any unbounded environmental impacts closer to, but still prior to, the decommissioning activity being undertaken that could cause the unbounded impact. In that case, the licensee should identify in the PSDAR the decommissioning activities that are not bounded by previous environmental reviews and will be addressed in the future. This proposed change would be consistent with the purpose of the PSDAR, as noted in RG 1.185, Revision 1, “Standard Format and Content for Post-Shutdown Decommissioning Activities Report” (ADAMS Accession No. ML13140A038), as a mechanism for NRC oversight because it would alert the NRC to any potentially unbounded environmental impacts associated with planned site-specific decommissioning activities. If a licensee were to consider a proposed decommissioning activity that would otherwise be prohibited by § 50.82(a)(6)(ii) or § 52.110(f)(2), then prior to undertaking that activity, the licensee could submit a request for a license amendment or an exemption request, decide not to perform the proposed activity, or modify the proposed activity so that the unbounded environmental impact does not occur. If the licensee chose to submit a license amendment or exemption request, then the request would trigger NRC responsibilities under environmental statutes. In addition, prior to performing a decommissioning activity that is inconsistent with the PSDAR but permitted by § 50.59, the licensee must notify the NRC in writing, with a copy to the affected States, in accordance with § 50.82(a)(7). This § 50.82(a)(7) requirement is in the current regulation and would not be changed in this proposed rule.

The NRC also proposes to change the § 50.82(a)(4)(i) and § 52.110(d)(1) regulations to allow licensees to use appropriate federally issued environmental review documents prepared in compliance with NEPA, ESA, NHPA, or other environmental statutes instead of only EISs. One reason for replacing the phrase “previously issued environmental impact statements” with “federally issued environmental review documents” is the NRC can, in many instances, satisfy its NEPA compliance obligations by the preparation of an environmental assessment or through a categorical exclusion finding rather than preparing an EIS. A second reason is that this change allows licensees to use a wider range of documents that address various resources. Examples of appropriate federally issued environmental review documents include environmental assessments prepared for license amendments such as extended power uprates; documents prepared during Section 7 consultations under the ESA such as biological opinions and biological assessments; or programmatic agreements prepared through Section 106 consultations under the NHPA to resolve impacts to historic properties. Environmental review documents prepared by other Federal agencies could also be used if they were relevant to the impacts associated with the site-specific decommissioning activities.

The regulations in § 50.82(a)(6)(ii) and § 52.110(f)(2) prohibit a licensee from undertaking a decommissioning activity that would result in a significant environmental impact not previously reviewed. The NRC is also proposing to change § 50.82(a)(6)(ii) and § 52.110(f)(2) to clarify that the previous review of any potentially significant environmental impact must be bounded by appropriate federally issued environmental review documents prepared in compliance with NEPA, ESA, NHPA, or other environmental statutes. In this regard, the determination of significance should be made in terms of the appropriate federal environmental resource protection statute. For example, if a proposed decommissioning activity were likely to result in a potential adverse effect upon a historic property, as the term “adverse effect” is described in the Advisory Council on Historic Preservation regulation, 36 CFR 800.5, “Typical classes of action,” then that potential adverse effect would most likely be equivalent to a potential significant impact under § 50.82(a)(6)(ii) or § 52.110(f)(2). Similarly, for species listed under the ESA, the equivalent threshold would be a proposed decommissioning activity that could result in a “take,” as that term is defined in 16 U.S.C. 1532(19), of any listed species at the time of the proposed decommissioning activity.

These proposed changes would reduce the regulatory burden on the licensee by removing the duplicative requirement to address unbounded environmental impacts at the PSDAR stage. Instead, licensees would only prepare an environmental report or provide other information as requested by the NRC under § 51.41, “Requirement to submit environmental information,” before performing any decommissioning activity that is likely to result in a significant impact not previously bounded.

2. Consistency Changes to 10 CFR Part 51

Currently, § 51.53(d) requires that an applicant for a license amendment authorizing decommissioning activities for a production or utilization facility either for unrestricted use or continuing use restrictions submit an environmental report. The regulation at § 51.95(d) states that the NRC will prepare a supplemental EIS or an environmental assessment in connection with an amendment of a license to authorize decommissioning activities.

The 1996 Final Rule eliminated the requirement for nuclear power reactor licensees to seek NRC authorization for decommissioning. Therefore, there was no need for licensees to submit a license amendment or to prepare and submit a supporting environmental report, and thus no federal action that would require the NRC to prepare a NEPA document. In response to the 1995 decommissioning proposed rule, commenters suggested that revisions should be made to then-§ 51.53, “Supplement to environmental report,” and then-§ 51.95, “Supplement to final environmental impact statement,” to reflect the rule change. However, the NRC at that time decided not to amend the 10 CFR part 51 regulations because non-power reactor facilities were still required to submit a decommissioning plan.

The NRC proposes to revise 10 CFR part 51 to reflect the changes made in the 1996 Final Rule that nuclear power reactor licensees are not required to submit license amendment requests for authorization to perform decommissioning activities. In § 51.53(d), the NRC is proposing to remove language referencing an amendment for authorizing decommissioning activities and the requirement to prepare an environmental report for nuclear power reactors only. In § 51.95(d), the NRC is similarly proposing to remove language referencing an amendment for authorizing decommissioning activities. The NRC further proposes to revise § 51.95(d) to indicate that the NRC would prepare the necessary NEPA document upon the submittal of a license amendment requesting approval of a license termination plan. The NRC also proposes to add a cross-reference to § 52.110 in § 51.53, “Postconstruction environmental reports,” as reactors licensed under 10 CFR part 52 will perform decommissioning under § 52.110, not § 50.82.

The NRC is not proposing to make any changes in 10 CFR part 51 that would impact non-power production or utilization facilities ( e.g., research and test reactors) or fuel reprocessing plants. Non-power production or utilization facility and fuel reprocessing plant licensees must continue to submit a license amendment requesting approval for a decommissioning plan and to prepare and submit the appropriate supporting environmental report, and the NRC would continue to prepare the appropriate NEPA documentation.

I. Record Retention Requirements

The NRC's regulations require nuclear power reactor licensees to retain the records associated with certain SSCs until the license is terminated and sometimes require that these records be kept in duplicate. To decrease the burden associated with long-term record storage and increase the overall efficiency of the decommissioning process, licensees that are transitioning to decommissioning frequently request exemptions from these requirements. Although this approach continues to meet the underlying purpose of the recordkeeping regulations, the process of preparing, submitting, and reviewing exemptions from the record retention requirements is not an efficient use of NRC or licensee resources given the fact that the subject records are no longer needed to support any NRC-regulated function. In addition, maintaining the current regulations with respect to record retention during decommissioning can create a situation wherein the facilities used to store records are ready to be dismantled in support of site decommissioning before the necessary exemptions can be processed. The NRC proposes to resolve these issues by amending its regulations in this rulemaking.

The recordkeeping requirements at issue include the following:

• Criterion XVII, “Quality Assurance Records,” of appendix B, “Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants,” to 10 CFR part 50 requires licensees to retain certain records consistent with regulatory requirements for a duration established by the licensees.

• Sections 50.59(d)(3) and 52.63(b)(2) require licensees to maintain certain records until termination of a license issued under 10 CFR part 50 or 10 CFR part 52.
• Section 50.71(c) requires licensees to maintain certain records consistent with various elements of the NRC regulations, facility TSs, and other licensing basis documents.
• Section 72.72(d) requires licensees to duplicate certain records of spent fuel and high-level radioactive waste and store them in a separate location sufficiently remote from the original records so that a single event would not destroy both sets.

Licensees that have previously requested exemptions from these requirements used the justification that, when the SSCs associated with these records are removed from service and the licensing basis documents, the SSCs will no longer serve any NRC-regulated function. Therefore, it would no longer be necessary to retain the records. In addition, several licensees requesting an exemption from the requirements of § 72.72(d) used the justification that they will store the ISFSI spent fuel records using the same procedures and processes used for the facility spent fuel (and other) records, which are typically stored in accordance with the NRC-approved quality assurance program (QAP).

The NRC granted the previous record retention exemptions based on a finding of reasonable assurance that the licensee would continue to meet the underlying purpose of the recordkeeping regulations, which is to establish the minimum retention periods necessary for the NRC to ensure compliance with the safety and health aspects of the nuclear environment and for the NRC to accomplish its mission to protect the public health and safety. In “Retention Periods for Records; Final Rule” (53 FR 19240; May 27, 1988), the Commission explained that requiring licensees to maintain adequate records assists the NRC in judging compliance and noncompliance, to act on possible noncompliance, and to examine facts as necessary following any incident. Because the SSCs that were safety-related or important to safety during reactor operations or operation of the SFP are removed from the licensing basis, and subsequently removed from the plant during the decommissioning process, the records associated with those SSCs are no longer required to achieve the purpose of the recordkeeping and record retention regulations.

Records associated with SSCs that maintain compliance with requirements or that protect public health and safety during the decommissioning process have been excluded from these exemptions. Examples include those SSCs associated with programmatic controls pertaining to residual radioactivity, security, and quality assurance (QA), and those SSCs associated with spent fuel assemblies or the SFP (while assemblies are still in the pool) and ISFSIs. These exemptions do not affect the record retention requirements of § 50.75 or any other requirements of 10 CFR part 50 that apply to decommissioning.

Based on these exemptions, the NRC proposes to change the recordkeeping and record retention requirements such that once the NRC dockets a licensee's notifications of permanent cessation of operation and permanent removal of fuel from the reactor vessel under § 50.82(a)(2) or § 52.110(a), licensees can then eliminate records associated with SSCs that no longer serve any NRC-regulated function. The NRC would allow this record disposal as long as appropriate change mechanisms, such as the § 50.59 evaluation process or NRC-approved TS changes, are used to assess the removal of those records to determine that elimination of the records would have no adverse impact on public health and safety.

The records that would be subject to removal are associated with SSCs that had been important to safety during reactor operation or operation of the SFP, but that are no longer capable of causing an event, incident, or condition that would adversely impact public health and safety, as evidenced by their appropriate removal from the licensing basis documents. Since the SSCs no longer have the potential to cause these scenarios, it is reasonable to conclude that the records associated with these SSCs would not reasonably be necessary to assist the NRC in determining compliance, taking action on possible noncompliance, and examining facts following an incident. Therefore, retention of such records would not serve the underlying purpose of the recordkeeping regulations.

The NRC proposes to make the following four changes to the recordkeeping and record retention requirements and regulatory guidance to enhance the efficiency of the decommissioning regulations:

1. Clarify in RG 1.184, “Decommissioning of Nuclear Power Reactors,” that the requirements in appendix B to 10 CFR part 50, Criterion XVII, concerning record retention, such as duration, location, and assigned responsibility, continue to be met with the recommended changes to the recordkeeping and record retention requirements.

2. Amend § 50.71(c) to specify that licensees for which the NRC has docketed the certifications required under § 50.82(a)(1) or § 52.110(a) are not required to retain records associated with SSCs that have been removed from service using an NRC-approved change process. However, § 50.71(c) would require licensees to retain records important to decommissioning as specified under § 50.75(g).

3. Amend §§ 50.59(d)(3) and 52.63(b)(2) to clarify that records of changes in the facility must be maintained until the termination of the license except for records associated with SSCs removed from service using an NRC-approved change process after the NRC has docketed the certifications required under § 50.82(a)(1) or § 52.110(a).

4. Amend § 72.72(d) to allow that records of spent fuel, high-level radioactive waste, and reactor-related greater than Class C (GTCC) waste containing special nuclear material no longer be kept in duplicate, as long as the licensee can demonstrate that it will store the records in the same manner as it would for other QA records using a single storage facility subject to the same procedures and processes outlined in an NRC-approved QAP.

In most cases, an NRC-approved QAP involves document storage requirements that meet American National Standards Institute (ANSI) standard N45 2.91974, “Requirements for Collection, Storage, and Maintenance of Quality Assurance Records,” which specifies, in part, the design requirements for use in the construction of record storage facilities when the use of a single storage facility is desired. In approving the associated QAP, the NRC typically approves the single facility location used for the storage and maintenance of QA records at the facility, and the licensee typically affirms in the QAP that the record storage facility was constructed and is being maintained to meet the requirements of the NRC-approved QAP.

Records for an ISFSI at a specific facility are typically classified as QA records and include all documents and records associated with the operation, maintenance, installation, repair, and modification of SSCs covered by the QAP. An ISFSI's records also include historical records that have been gathered and collected during plant and ISFSI operations. These records are either required in support of the dry cask storage systems used at the ISFSI or for ultimate shipment of the fuel to a Federal repository. The QAP typically allows the storage of QA records, including ISFSI records, to be done in accordance with ANSI N45 2.9-1974 in a single storage facility designed and maintained to minimize the risk of damage from adverse conditions.

The retention of records required by § 50.59(d)(3); § 52.63(b)(2); § 50.71(c); and appendix B to 10 CFR part 50, Criterion XVII provides assurance that records associated with SSCs will be captured, indexed, and stored in an environmentally suitable and retrievable condition. Although licensees retain the records required by their license as the plant transitions from operating conditions to a fully decommissioned state, plant dismantlement obviates the regulatory need for maintenance of most records. As the SSCs already removed from the licensing basis are subsequently dismantled and the need for the associated records is, on a practical basis, eliminated, the proposed rule changes would allow disposal of the records associated with SSCs and historical activities that are no longer relevant and thereby eliminate the associated regulatory and economic burdens of creating alternative storage locations, relocating records, or retaining irrelevant records. The proposed recordkeeping and record retention changes only expedite the schedule for disposition of the specified records. Considering the content of these records, their elimination on an advanced timetable has no reasonable potential of presenting any undue risk to public health and safety. In addition, upon dismantlement of the affected SSCs, the records have no functional purpose relative to maintaining the safe operation of the SSCs, maintaining conditions that would affect the ongoing health and safety of workers or the public, or informing decisions related to nuclear safety and security.

In addition, the proposed change to the portion of § 72.72(d) to eliminate the requirement for ISFSI licensees to keep a duplicate set of records for spent fuel in storage, would continue to meet the recordkeeping requirements of appendix B to 10 CFR part 50 and other applicable 10 CFR part 72 requirements for the storage and maintenance of spent fuel records in accordance with an NRC-approved QAP. Specifically, § 72.140(d) states that a QA program that the NRC has approved as meeting the applicable requirements of appendix B to 10 CFR part 50, will be accepted as satisfying the requirements of § 72.140(b) for establishing an ISFSI QA program. However, the licensee must also meet the recordkeeping provisions of § 72.174, “Quality assurance records.” In addition, the proposed rule change would not affect the record content, retrievability, or retention requirements specified in § 72.72, “Material balance, inventory, and records requirements for stored materials,” or § 72.174, such that the licensee will continue to meet all other applicable recordkeeping requirements for the ISFSI and associated special nuclear materials.

In proposing these rule changes, the NRC determined that the process and procedures used to store the ISFSI records ( i.e., in accordance with the QAP at a facility designed for protection against degradation mechanisms such as fire, humidity, and condensation) would help ensure that the licensee will adequately maintain the required spent fuel information. Therefore, changes to the duplicate record requirement of § 72.72(d) would not affect public health and safety. In addition, allowing the ISFSI spent fuel records to be stored in the same manner as that of other QA records for the nuclear facility would provide for greater efficiency in the storage of all records once the facility enters the final stages of decommissioning, where only the ISFSI facility would remain after license termination.

J. Low-Level Waste Transportation

Paragraph III.E of appendix G, “Requirements for Transfers of Low-Level Radioactive Waste Intended for Disposal at Licensed Land Disposal Facilities and Manifests,” to 10 CFR part 20, “Standards for Protection Against Radiation,” contains requirements for investigating, tracing, and reporting shipments of low-level radioactive waste (LLW) if the shipper has not received notification of receipt within 20 days after transfer. In addition, paragraph III.E requires the shipper to report such missing shipments to the NRC. Licensees, primarily those that are involved in the decommissioning process, frequently request an exemption from the requirement related to the 20-day receipt notification window. The NRC proposes to amend this requirement to extend the receipt notification window because such an extension would provide licensees with flexibility while not impacting public health and safety or the common defense and security.

Licensees that have previously been granted these exemptions typically requested extension of the investigation notification window to 45 days using the justification that operating experience indicates that, while the 20-day receipt notification window is adequate for waste shipments by truck, waste shipments using other modes of shipment such as rail, barge, or mixed-mode shipments, such as combinations of truck and rail, barge and rail, and barge and truck shipments, may take more than 20 days to reach their destination due to delays in the route that are outside the shipper's control ( e.g., rail cars in switchyards waiting to be included in a complete train to the disposal facility). The NRC granted the previous transportation investigation requirement exemptions based on a finding of reasonable assurance that the shipper would continue to meet the underlying purpose of the LLW transportation regulations—to require the shipper to investigate, trace, and report radioactive shipments that have not reached their destination, as scheduled, for unknown reasons.

Under the current regulations, the shipper must investigate, trace, and report to the NRC any shipments of LLW for which the shipper has not received a notification of receipt within 20 days after transfer unless the shipper receives an exemption from the 20-day receipt notification requirement. The NRC has found that exempting licensees from this requirement does not undermine public health and safety, nor does it increase any security risk. Further, the preparation and submission of the exemption request, and its review, evaluation, and approval by the NRC, are not efficient uses of NRC or licensee resources. Specifically, the NRC notes that allowing the receipt notification to be made past 20 days would not impact public health, safety, or security even if the LLW transportation package was situated in a publicly accessible area and waiting for continuing transport to the waste disposal site because: (1) Individuals in the vicinity of the LLW transportation package would receive no additional radiological dose above background levels resulting from the disposal container; and (2) the LLW would remain secured in the transportation package until the package can be delivered to the waste disposal site. The NRC also notes that, for LLW waste shipments, most shippers will use an electronic data tracking system interchange or similar tracking systems that allow the carrier to monitor the progress of the shipments daily. Because of the oversight and monitoring of radioactive waste shipments throughout the journey from the nuclear facility to the disposal site, the loss, misdirection, or diversion of a shipment without the knowledge of the carrier or the shipper is unlikely.

Therefore, the NRC proposes to change the requirement for the investigation, tracing, and reporting timeframe for LLW transportation to extend the receipt notification window to 45 days after the shipper transfers LLW from a licensed facility to a disposal site. This change would continue to meet the underlying purpose of appendix G to 10 CFR part 20, paragraph III.E, which requires the shipper to investigate, trace, and report LLW shipments that have not reached their destination, as scheduled, for unknown reasons. Furthermore, extending the time period for notification of receipt to 45 days before requiring investigation, tracing, and reporting, would maintain a reasonable upper limit on shipment duration if a breakdown of normal tracking systems were to occur, based on operating experience.

In addition, the NRC proposes correcting a typographical error in the current version of appendix G to 10 CFR part 20, paragraph III.E. Specifically, that paragraph states that LLW shipments must “be investigated by the shipper if the shipper has not received notification or receipt within 20 days after transfer . . .” (emphasis added). The “or” should be an “of,” consistent with the subsequent discussions in 10 CFR part 20 regarding notifications of receipt and the associated exemptions granted in this area. Therefore, the NRC proposes correcting this error as part of this proposed rule for consistency and clarity within 10 CFR part 20.

K. Spent Fuel Management Planning

The regulation in § 72.218(a) states that the § 50.54(bb) spent fuel management program ( i.e., the irradiated fuel management plan or IFMP) must include a plan for removing from the reactor site the spent fuel stored under the 10 CFR part 72 general license. The IFMP must show how the spent fuel will be managed before starting to decommission systems and components needed for moving, unloading, and shipping this spent fuel. Section 72.218(b) requires that an application for termination of a reactor operating license submitted under § 50.82 or § 52.110 must also describe how the spent fuel stored under the 10 CFR part 72 general license will be removed from the reactor site. Although § 72.218 states what information the § 50.54(bb) IFMP and the § 50.82 and § 52.110 application for termination of a reactor operating license must include, the regulations in §§ 50.54(bb), 50.82, and 52.110 do not contain this information.

As §§ 50.54(bb), 50.82, and 52.110 do not reflect or otherwise reference the provisions in § 72.218, this causes regulatory uncertainty. The NRC proposes to clarify and align the regulations in §§ 50.54(bb), 50.82, 52.110, and 72.218 to provide regulatory clarity and enhance overall regulatory transparency and openness regarding decommissioning and spent fuel management planning.

1. Requirements for the IFMP in § 50.54(bb) and the PSDAR in § 50.82 and § 52.110

The PSDAR and IFMP are planning documents for decommissioning and spent fuel management, respectively. The current requirements for the timing of the submittal of the PSDAR and IFMP are similar, as the NRC's regulations recognize that a licensee's ability to plan properly and safely for decommissioning is closely related to the licensee's ability to manage its spent fuel. Actions to manage spent fuel include activities taken prior to and subsequent to decommissioning. Therefore, a licensee's spent fuel management plans and its decommissioning plans should be consistent.

Because § 50.54(bb) already addresses the topic of spent fuel management planning, the NRC proposes including the § 72.218 provisions in § 50.54(bb) to clarify that the § 50.54(bb) IFMP must be submitted by the licensee and approved by the NRC before the licensee starts to decommission SSCs needed for moving, unloading, and shipping the spent fuel. Additionally, the NRC proposes that the IFMP must be submitted prior to or within 2 years following permanent cessation of operations.

The NRC proposes to further restructure § 50.54(bb) to clarify that the IFMP addresses both the safety and financial aspects of managing spent fuel. The IFMP would describe the licensee's planned actions for managing spent fuel, how those actions would be consistent with the NRC requirements for possession of spent fuel, and any actions related to spent fuel management that would require amendments to the license or certificate of compliance or exemptions from applicable regulations, which is consistent with the current rule language. The IFMP would also describe the projected cost of managing spent fuel and how the licensee would provide funding for the management of the spent fuel, until title to, and possession of, the spent fuel is transferred to the Department of Energy (DOE), which is also consistent with the current rule language. The regulation in § 50.54(bb) would also continue to require licensees to retain a copy of the IFMP as a record, and the NRC proposes to clarify that the IFMP must be retained until termination of the 10 CFR part 50 or 10 CFR part 52 license.

The NRC proposes to clarify the current IFMP approval process and the § 50.54(bb) provisions regarding preliminary approval and final NRC review of the IFMP as part of any proceeding for continued licensing under 10 CFR part 50 or 10 CFR part 72. With regard to the NRC's final review of the IFMP “as part of any proceeding for continued licensing under 10 CFR part 50 or 10 CFR part 72,” these proceedings no longer exist as they did when § 50.54(bb) was first promulgated in 1984. In the 1984 Final Rule, the Commission discussed the “proceeding for continued licensing under part 50” as the pre-1996 reactor decommissioning process, where licensees were required to submit a license amendment request for approval of the decommissioning plan and to change the license from an operating license to a possession-only license before licensees could begin decommissioning. The NRC noted in the 1984 Final Rule that the IFMP would become part of the conditions of an amended 10 CFR part 50 license for a shutdown reactor facility. After the 1996 rulemaking, the NRC no longer requires submittal of a license amendment when a reactor ceases operations, and thus, there is no longer a “proceeding for continued licensing under part 50” for the NRC to review and approve the IFMP.

The 1984 Final Rule discusses the “proceeding for continued licensing under part 72” as the application for, and NRC issuance of, a 10 CFR part 72 specific license for storage of spent fuel in an ISFSI. The 1984 issuance of § 50.54(bb) preceded the general license ISFSI provisions, which were added to 10 CFR part 72 in 1990. Regarding the 10 CFR part 72 general license, storage of spent fuel in a general license ISFSI is authorized by operation of law via § 72.210, so there is no NRC “licensing proceeding” or approval needed for the 10 CFR part 72 general license. As most reactor licensees use the 10 CFR part 72 general license for storage of spent fuel in an ISFSI, there would be no “proceeding for continued licensing under part 72” for the NRC to review and approve the IFMP. Therefore, the NRC proposes to require submittal of the IFMP to the NRC as a license amendment request. The NRC also proposes to require licensees to submit to the NRC any changes to the IFMP as an application for an amendment to its license.

2. Requirements in § 72.218 for Termination of the General License for Spent Fuel Storage

Because the current spent fuel management planning provisions of § 72.218 are initiated by reactor shutdown and are related to reactor decommissioning, the requirements fit best in 10 CFR part 50 and are not necessarily needed in 10 CFR part 72. Therefore, as the NRC proposes adding the spent fuel management provisions from § 72.218 into § 50.54(bb), the NRC also proposes deleting those provisions from § 72.218. In addition, the NRC proposes revising § 72.218 to address requirements related to termination of the 10 CFR part 72 general license, as the current title of § 72.218, “Termination of licenses,” suggests.

The 10 CFR part 72 general license is issued to 10 CFR part 50 or 10 CFR part 52 licensees, per the regulation in § 72.210. It follows that the 10 CFR part 72 general license would terminate coincident with the termination of the 10 CFR part 50 or 10 CFR part 52 license. In addition, since the general license ISFSI is part of the 10 CFR part 50 or 10 CFR part 52 licensed site, decommissioning of the general license ISFSI would follow the reactor decommissioning process in § 50.82 or § 52.110, respectively. This approach would also be consistent with the NRC's approach to ISFSI decommissioning funding as discussed in the “ Decommissioning Funding Assurance” section of this document.

However, to provide regulatory clarity between 10 CFR parts 50, 52, and 72 in terms of decommissioning and termination of the 10 CFR part 72 general license, the NRC proposes to revise § 72.218 to include the following provisions: (1) The general license ISFSI must be decommissioned consistent with the requirements in § 50.82 or § 52.110; and (2) the general license is terminated upon termination of the 10 CFR part 50 or 10 CFR part 52 license. This proposed change would provide regulatory clarity among 10 CFR parts 50, 52, and 72 in terms of decommissioning and termination of the 10 CFR part 72 general license, analogous to the provision in § 72.210 that ties the issuance of the 10 CFR part 72 general license to the existence of the 10 CFR part 50 or 10 CFR part 52 license.

L. Backfit Rule

For nuclear power reactor licensees, the NRC's backfitting provisions are located in § 50.109, “Backfitting,” and the issue finality provisions are in 10 CFR part 52 (hereinafter collectively referred to as the “Backfit Rule”). The language of the Backfit Rule clearly applies to a licensee designing, constructing, or operating a nuclear power facility. For example, § 50.109(a)(1) defines “backfitting” to mean changes to, among other things, the procedures or organization required to design, construct or operate a facility. The application of the Backfit Rule to decommissioning plants is not as clear. In SECY-98-253, “Applicability of Plant-Specific Backfit Requirements to Plants Undergoing Decommissioning,” dated November 4, 1998 (ADAMS Accession No. ML992870107), the NRC staff presented the Commission with a list of reasons underlying this uncertainty:

• The Backfit Rule has no end point when the rule no longer applies, “thereby implying that backfit protection continues into decommissioning and up to the point of license termination.”
• The term “operate” could reasonably be interpreted as including activities to decommission the reactor.
• The Backfit Rule was developed when the decommissioning of plants was not an active area of regulatory concern.
• The Backfit Rule's definition of “backfitting” uses terms associated with the design, construction, and operation of a facility rather than with its decommissioning, although the staff noted in SECY-98-253 that “prior to the 1996 decommissioning rule, the Commission regarded decommissioning as a phase of the plant's life cycle which is different from the operational phase.”
• Two of the factors used in evaluating a backfit—costs of construction delay/facility downtime, and changes in plant/operational complexity—are targeted to power operation and are “conceptually inappropriate in evaluating the impacts of a backfit on a decommissioning plant.”
• The SOC for the 1970 (35 FR 5317; March 31, 1970), 1985 (50 FR 38097; September 20, 1985), and 1988 (53 FR 20603; June 6, 1988) final Backfit Rules did not discuss any aspect of decommissioning, focusing instead on construction and operation.
• Proposed changes to decommissioning requirements usually focused on relaxing a requirement or on whether a requirement applicable to an operating reactor continued to be applicable to a decommissioning plant. Thus, “the notion of a `substantial increase' in protection to public health and safety from a backfit does not appear to be particularly useful [in decommissioning].”
• The 1996 Final Rule did not directly respond to questions from the public on the applicability of the Backfit Rule to a decommissioning plant.

Over the years, the NRC has tried to clarify the applicability of the Backfit Rule to nuclear power reactor licensees in decommissioning. In SECY-98-253, the NRC staff requested Commission approval to amend § 50.109, among other regulations, so that the Backfit Rule would clearly apply to licensees in decommissioning. In that paper, the NRC staff also proposed that, until the rulemaking was finished, the staff would apply the Backfit Rule to plants undergoing decommissioning “to the extent practical.”

In the February 12, 1999, SRM for SECY-98-253 (ADAMS Accession No. ML003753746), the Commission approved development of a Backfit Rule for plants undergoing decommissioning. The Commission directed the NRC staff to continue to apply the then-current Backfit Rule to plants undergoing decommissioning until issuance of the final rule. The Commission directed the staff to develop a rulemaking plan, which the staff transmitted to the Commission in SECY-00-0145. In SECY-00-0145, the NRC staff proposed, among other decommissioning-related amendments to its regulations, amendments to § 50.109 to show clearly that the Backfit Rule applies during decommissioning and to remove factors that are not applicable to nuclear power plants in decommissioning. As explained in the section titled “ Actions Leading to this 2018 Proposed Rule” in this document, the NRC ultimately did not conduct that rulemaking. Therefore, the NRC has continued to apply the Backfit Rule to licensee facilities undergoing decommissioning to the extent practical.

In addition to the Commission direction to clarify the application of the Backfit Rule for decommissioning nuclear power reactor licensees, the NRC's regulatory framework also supports application of the Backfit Rule to nuclear power reactor licensees in decommissioning. Under sections 101 and 103a. of the AEA (42 U.S.C. 2131 and 2133a.), the NRC's issuance of a nuclear power reactor operating license under 10 CFR part 50 or a combined license under 10 CFR part 52 grants the holder a license to, among other things, own, possess, and operate a “production facility” or “utilization facility,” as those terms are defined in section 11 of the AEA. Once the licensee under 10 CFR part 50 or 10 CFR part 52 submits its certifications of permanent cessation of reactor operations and permanent removal of fuel from the reactor vessel and the NRC dockets those certifications, the licensee is no longer authorized to operate the reactor under § 50.82(a)(2) or § 52.110(b), respectively. The license is no longer an “operating license” for the reactor because the licensee is not operating a production or utilization facility pursuant to sections 101 and 103a. of the AEA. Instead, as described in § 50.51(b) for 10 CFR part 50 licenses and § 52.109, “Continuation of combined license,” for 10 CFR part 52 combined licenses, when the reactor has permanently ceased operations, the license continues in effect beyond the expiration date and authorizes ownership and possession of the facility until the Commission terminates the license. Thus, when the licensee is no longer authorized to operate the reactor, it retains its possession and ownership authority under its 10 CFR part 50 or 10 CFR part 52 facility license.

Although a decommissioning licensee's license no longer authorizes operation of the reactor because the licensee is not operating a production or utilization facility, the licensee still must “operate” certain SSCs at the site. Under § 50.51(b) (with a similar requirement in § 52.109 for combined license holders), when the licensee has only a possession and ownership license for the reactor, the licensee must not only decommission and decontaminate the facility, but also continue to maintain the facility, including storing, controlling and maintaining the spent fuel in a safe condition. Therefore, nuclear power reactor licensees store, control, and maintain spent fuel after permanent cessation of reactor operations through the “operation” of an SFP and ISFSI.

Although § 50.109(a)(1) defines “backfitting” as changes to, among other things, the procedures or organization required to design, construct, or operate a facility, indicating that the Backfit Rule applies only to a holder of a license to “operate a facility,” the language of § 50.51(b) shows that “operating a facility” can be interpreted to mean more than just operating a reactor. This is supported by the Commission direction in the SRM for SECY-98-253 that the NRC staff develop a Backfit Rule for plants undergoing decommissioning ( i.e., when the licensee no longer operates a reactor) and continue to apply the then-current Backfit Rule to plants undergoing decommissioning until issuance of the final rule. Thus, the Backfit Rule still applies to a licensee that has a license to only possess and own a facility. For a facility in decommissioning, the phrase “operate a facility” in § 50.109(a)(1) is read to encompass operating the SFP and associated SSCs necessary for compliance with § 50.51(b).

As the Commission and the NRC staff recognized in the 1990s, certain provisions of the Backfit Rule do not clearly apply to nuclear power reactor licensees in decommissioning. In this proposed rule, the NRC proposes to complete the process begun two decades ago to clarify the application of the Backfit Rule to nuclear power reactor licensees in decommissioning.

The NRC proposes to amend § 50.109 so that nuclear power reactor licensees, which have had their § 50.82(a)(1) or § 52.110(a) certifications docketed by the NRC, are the subject of similar backfitting provisions as they were during their operating phase. A new backfitting provision for licensees in decommissioning would eliminate any confusion with the meaning of the words “operate a facility” in § 50.109(a)(1), as compared to other uses of the term “operate” in 10 CFR Chapter I.

The NRC would make other revisions to § 50.109. To make the section easier to read, the NRC proposes to insert paragraph headings. The NRC would remove current § 50.109(b) regarding backfits imposed prior to October 21, 1985, because the language is obsolete and no longer needed. In the current § 50.109(a)(6), the NRC proposes to insert a sentence explaining that a documented evaluation, which is used by the NRC to justify not performing a backfit analysis, must include a consideration of the costs of imposing the backfit if the basis for backfitting is bringing a facility into compliance with a license or the rules or orders of the Commission, or into conformance with the licensee's written commitments.

Further, the NRC proposes to make conforming changes to § 72.62 to clarify that the corresponding backfit regulations in part 72 apply during the decommissioning of an ISFSI or a Monitored Retrievable Storage facility subject to those provisions.

M. Foreign Ownership, Control, or Domination

The NRC is proposing to amend its regulations to address the circumstances when a facility licensed under 10 CFR part 50 or 10 CFR part 52 no longer meets the definition of a utilization facility or a production facility. The AEA has certain requirements specific to utilization or production facilities. By clarifying when a 10 CFR part 50 or 10 CFR part 52 licensed facility is no longer a utilization or a production facility, the NRC can then specify whether these AEA requirements still apply to the licensee for that facility. For instance, the AEA prohibits the issuance of a license for a utilization or a production facility to an entity that the Commission knows or has reason to believe is foreign owned, controlled, or dominated. The Commission's regulations that implement this prohibition, however, are unclear as to when a facility undergoing decommissioning is no longer a utilization or a production facility. Given this uncertainty, licensees have requested exemptions from § 50.38, “Ineligibility of certain applicants,” to transfer 10 CFR part 50 licenses for facilities that no longer meet the definition of utilization facility. The NRC proposes to amend its regulations to clarify when a facility licensed under 10 CFR part 50 or part 52 is not considered a production or utilization facility and therefore, the FOCD prohibition no longer applies.

The NRC's regulations in 10 CFR parts 50 and 52 provide for the issuance of a 10 CFR part 50 license for a utilization or a production facility and a 10 CFR part 52 license for a utilization facility. The AEA defines “utilization facility” as:

(1) Any equipment or device, except an atomic weapon, determined by rule of the Commission to be capable of making use of special nuclear material in such quantity as to be of significance to the common defense and security, or in such manner as to affect the health and safety of the public, or peculiarly adapted for making use of atomic energy in such quantity as to be of significance to the common defense and security, or in such manner as to affect the health and safety of the public; or (2) any important component part especially designed for such equipment or device as determined by the Commission.

The AEA defines “production facility,” in part, as:

(1) Any equipment or device determined by rule of the Commission to be capable of the production of special nuclear material in such quantity as to be of significance to the common defense and security, or in such manner as to affect the health and safety of the public; or (2) any important component part especially designed for such equipment or device as determined by the Commission.

As authorized by the AEA, the Commission has a rule defining utilization facility and production facility. In § 50.2, a utilization facility is defined as either (1) any nuclear reactor other than one designed or used primarily for the formation of plutonium or U-233; or (2) an accelerator-driven subcritical operating assembly used for the irradiation of materials containing special nuclear material and described in the application for the SHINE Medical Isotope Production Facility. A production facility is defined as a nuclear reactor designed or used primarily for the formation of plutonium or uranium-233; with certain exceptions not relevant here, a facility designed or used for the separation of the isotopes of plutonium; or, with certain exceptions not relevant here, a facility designed or used for the processing of irradiated materials containing special nuclear material.

NRC case law provides insight as to when a facility licensed under 10 CFR part 50 or 10 CFR part 52 is no longer a utilization or a production facility. In LBP-84-33, Cincinnati Gas & Electric Co. (Wm. H. Zimmer Nuclear Power Station, Unit 1), 20 NRC 765 (1984), an Atomic Safety and Licensing Board granted the licensee's motion to withdraw its application for a 10 CFR part 50 operating license for a nuclear power reactor, despite the fact that the facility was almost completely built. One of the conditions for granting the motion was that the nuclear steam supply system be modified to prevent the facility's operation as a utilization facility. The Board determined that because a utilization facility under the AEA is a facility that is capable of making use of special nuclear material, the facility must be modified to eliminate that capability for it to no longer be categorized as a utilization facility. The Board observed that this can be achieved, for example, by severing and welding caps on main feedwater lines and main steam lines and removing the fuel and the control rod drive mechanisms.

The NRC proposes to add to its regulations language similar to the Zimmer decision to establish the criteria for when a facility licensed under 10 CFR part 50 or 10 CFR part 52 no longer meets the statutory or regulatory definition of a utilization or a production facility ( i.e., is no longer capable of making use of special nuclear material or of the production of special nuclear material, separation of the isotopes of plutonium, or processing of irradiated materials containing special nuclear material (hereinafter collectively referred to as production-facility activities)). The first criterion is that the facility must not be legally authorized to operate. The second criterion is the physical modification of the licensed facility to be incapable of making use of special nuclear material and of production-facility activities, without significant facility alterations necessary to restore the capability to make use of special nuclear material or to engage in production-facility activities. When a utilization facility is physically modified to be incapable of making use of special nuclear material, it is no longer designed or used to sustain nuclear fission in a self-supporting chain reaction.

Sections 50.82(a)(2) and 52.110(b) already provide for the first criterion for nuclear power reactor licensees—that the facility is no longer legally authorized to operate. Sections 50.82(a)(2) and 52.110(b) state, respectively, that a 10 CFR part 50 license and a 10 CFR part 52 license no longer authorize operation of the reactor or emplacement or retention of fuel into the reactor vessel once the NRC has docketed the certifications for permanent cessation of operations and permanent removal of fuel from the reactor vessel, or when a final legally effective order to permanently cease operations has come into effect. The NRC would amend these regulations to add the second criterion—that the facility licensed under 10 CFR part 50 or 10 CFR part 52 is no longer a utilization facility once the licensee modifies the facility to be incapable of making use of special nuclear material without significant facility alterations.

Because the NRC's regulations do not state when a non-power production or utilization facility or fuel reprocessing plant licensee is no longer authorized to operate (other than at license termination), the NRC proposes to amend § 50.82(b) to add the criteria for when a non-power production or utilization facility or fuel reprocessing plant is no longer a production or utilization facility. The NRC would renumber current paragraph (b)(6) in § 50.82 as paragraph (b)(8) and add new paragraphs (b)(6) and (b)(7). New paragraph (b)(6) would provide that a non-power production or utilization facility or fuel reprocessing plant is not legally capable of operating when the NRC removes the licensee's authority to operate the facility through a license amendment. The NRC can remove a non-power production or utilization facility or fuel reprocessing plant licensee's authority to operate by issuing a possession-only license amendment or by approving the licensee's decommissioning plan through a license amendment, either of which would explicitly remove the licensee's authority to operate. Licensees typically request a possession-only license amendment first and then submit a decommissioning plan via a second license amendment request. This proposed rule would offer licensees the option to request only one licensing action—the decommissioning plan license amendment—that also would address the licensee's operating authority, rendering a separate “possession-only license amendment” unnecessary. To address those instances when the licensee is still operating the facility when the licensee submits its decommissioning plan license amendment request, the decommissioning plan license amendment would itself identify the date on which the authority to operate is removed.

The NRC would also include in new § 50.82(b)(6) the second criterion for when the non-power production or utilization facility or fuel reprocessing plant is no longer a production or a utilization facility ( i.e., once the licensee modifies the facility to be incapable of production-facility activities and making use of special nuclear material without significant facility alterations).

The NRC would add new § 50.82(b)(7) and amend § 50.82(a)(2) and § 52.110(b) to affirm the continuation of the NRC's statutory authority over the existing 10 CFR part 50 or 10 CFR part 52 license after the performance of decommissioning activities that lead to the licensed facility no longer meeting the definition of a utilization or a production facility. This facility transition occurs with every licensee during decommissioning: Eventually, the facility will be dismantled to the point where it is incapable of making use of special nuclear material or of production-facility activities without significant facility alterations.

Although the facility licensed under 10 CFR part 50 or 10 CFR part 52 may no longer be a utilization or a production facility, the NRC maintains the authority to regulate the existing 10 CFR part 50 or 52 license. A 10 CFR part 50 operating license for a production or utilization facility is issued under AEA sections 103 or 104, and a 10 CFR part 52 combined license for a utilization facility is issued under AEA sections 103 and 185b. That license may contain authorities beyond those governed by 10 CFR parts 50 or 52. Under § 50.52, “Combining licenses,” the Commission may combine in a single license the activities that would otherwise be licensed under separate licenses. Accordingly, a typical 10 CFR part 50 or 52 nuclear power reactor license also includes in a single license the authority under 10 CFR parts 30, 40, and 70 of the NRC's regulations to perform activities or possess materials authorized by those parts. Parts 30, 40, and 70 of 10 CFR are authorized by sections 81, 63, and 53 of the AEA and concern the licensing of byproduct, source, and special nuclear materials, respectively. A typical 10 CFR part 50 non-power production or utilization facility license also includes the authority under 10 CFR parts 30 and 70 of the NRC's regulations to perform activities or possess materials authorized by those parts. When the facility is no longer a production or utilization facility, the NRC maintains the authority to regulate the facility and the 10 CFR part 50 or 52 license under a combination of AEA sections 53, 63, 81, and 161. Sections 50.51(b) and 52.109 of the NRC's regulations also establish that the 10 CFR part 50 or 52 license continues in effect until the NRC terminates the license, notwithstanding the fact that at some point in time during the dismantlement required for license termination, the licensed facility will be disassembled to such an extent that it no longer satisfies the definition of a utilization or a production facility. Therefore, the NRC would amend § 50.82(a)(2), § 50.82(b), and § 52.110(b) to explicitly cite these statutory provisions as the basis for its retention of the authority to regulate the existing 10 CFR parts 50 or 52 facility. The NRC proposes to make conforming changes to the authority citations for 10 CFR parts 50 and 52 to add sections 53, 63, and 81 of the AEA.

The NRC proposes to amend § 50.82(a)(2), § 50.82(b), and § 52.110(b) to state which requirements apply to the existing 10 CFR part 50 or 52 license after the licensed facility is no longer a utilization or a production facility. As provided by section 161b of the AEA, the Commission is authorized to establish by regulation such standards to govern the possession and use of special nuclear material, source material, and byproduct material as the Commission may deem necessary or desirable to promote the common defense and security or to protect health or to minimize danger to life or property. Consistent with this statutory authority, the proposed amendments to § 50.82(a)(2), § 50.82(b), and § 52.110(b) will make clear that, after the facility licensed under 10 CFR part 50 or 52 is no longer a utilization or a production facility and until the termination of the 10 CFR part 50 license pursuant to § 50.82(a)(11) or § 50.82(b)(8) or the 10 CFR part 52 license pursuant to § 52.110(k), the NRC regulations applicable to utilization or production facilities will continue to apply to the holder of the 10 CFR part 50 or 10 CFR part 52 license, as applicable, unless those regulations explicitly state otherwise. These proposed amendments would enable a licensee to maintain reasonable assurance of adequate protection of the common defense and security and the public health and safety by requiring the licensee to continue to comply with those regulations applicable to utilization or production facilities, as applicable to that licensee, unless stated otherwise.

The NRC has identified that § 50.38 should not apply to a facility that is no longer a utilization or a production facility. Specifically, the AEA prohibits the issuance of a license for a utilization or a production facility to an entity that the Commission knows or has reason to believe is foreign owned, controlled, or dominated. However, since the FOCD prohibition only applies to a utilization or production facility, it would not apply once a 10 CFR part 50 or part 52 facility is no longer a utilization or a production facility. Therefore, the NRC is proposing to amend § 50.38 such that its prohibition on transferring a license to an entity that the Commission knows or has reason to believe is owned, controlled, or dominated by an alien, a foreign corporation, or a foreign government, is not applicable if the license is a 10 CFR part 50 or 10 CFR part 52 license for a facility that no longer meets the definition of a utilization or a production facility.

Section 50.80 governs the transfers of 10 CFR part 50 and 10 CFR part 52 licenses for production and utilization facilities. It requires the written consent of the NRC before the transfer of a production or utilization facility. This section also requires applicants for a license transfer to provide the same identifying, technical, and financial information that an initial license applicant is required to provide under §§ 50.33 and 50.34. In particular, § 50.33 requires an application to state the citizenship of the applicant. Under § 50.38, the applicant is ineligible to apply for and obtain a license if it is a foreign entity.

Section 50.38 implements sections 103 and 104 of the AEA, which provide in part that a license for a utilization or production facility may not be issued to an alien or any corporation or other entity if the Commission knows or has reason to believe it is owned, controlled, or dominated by an alien, a foreign corporation, or a foreign government. Since sections 103 and 104 of the AEA apply to utilization and production facilities, the NRC is proposing to amend § 50.38 to clarify that this prohibition does not apply to a person, corporation, or other entity seeking a license for a facility that is no longer a utilization or a production facility, as would be provided under revised § 50.82(a)(2), § 50.82(b)(6), or § 52.110(b).

The proposed amendment to § 50.38 would maintain the common defense and security and public health and safety because, even though § 50.38 would not prohibit the transfer to foreign entities of 10 CFR part 50 and 10 CFR part 52 licenses for facilities that do not meet the definition of utilization or production facility, other regulations ensure that such transfers would not be inimical to the common defense and security or to the health and safety of the public. For instance, § 50.80(c) states that the Commission will approve an application for the transfer of a license if the Commission determines that the proposed transferee is qualified to be the holder of the license and that the transfer of the license is otherwise consistent with applicable provisions of law, regulations, and orders issued by the Commission. In turn, under § 50.57 or § 52.97, the Commission may issue a 10 CFR part 50 or 10 CFR part 52 license, respectively, only if the Commission finds that the issuance of the license will not be inimical to the common defense and security or to the health and safety of the public.

The proposed amendment to § 50.38 is consistent with how the NRC analyzed requests for exemptions from § 50.38 for Maine Yankee Atomic Power Station, Haddam Neck Plant, and Yankee Nuclear Power Station (78 FR 58571; September 24, 2013). Specifically, the NRC granted those exemptions because the reactor facilities had been dismantled and removed such that only ISFSIs remained on site; an ISFSI, whether licensed under 10 CFR parts 50 or 72, is not capable of making use of special nuclear material; and the AEA definition of a utilization facility does not include ISFSIs. The NRC found that the foreign ownership, control, or domination prohibition did not apply to ISFSIs and, thus, did not preclude the NRC from granting the exemptions.

The NRC is also proposing to amend §§ 50.1, 50.51, 52.0, and 52.109 in light of the proposed amendments to §§ 50.38, 50.82, and 52.110. The proposed amendments would make clear that the regulations in 10 CFR part 50, and the similar regulations in 10 CFR part 52, provide not only for the licensing of utilization and production facilities, but also for their decommissioning and the termination of their associated licenses. These changes are clarifications; 10 CFR part 50 has included decommissioning and license termination since 1961 (“Creditors' Rights; and Transfer, Surrender, and Termination of Licenses,” 26 FR 9546; October 10, 1961). The NRC proposes to delete the language in §§ 50.51 and 52.109 that discusses what 10 CFR parts 50 and 52 licenses authorize in lieu of the more complete discussion provided in the proposed amendments to § 50.82(a)(2), § 50.82(b)(6) and (7), and § 52.110(b).

The NRC is proposing to add a specific definition for “non-power production or utilization facility” to § 50.2 to establish a term that is flexible enough to capture all non-power facilities licensed under § 50.22, “Class 103 licensees; for commercial and industrial facilities,” and § 50.21(a) or (c), except fuel reprocessing facilities. This proposed rule would address inconsistencies in definitions and terminology associated with non-power production and utilization facilities in § 50.2 that result in challenges in determining the applicability of the regulations. Fuel reprocessing plants would be excluded from the definition because the consequences associated with the hazards at a fuel reprocessing plant would likely exceed those anticipated at the facilities within the “non-power production or utilization facility” definition, thereby affecting the applicability of the “non-power production or utilization facility” term.

The only NRC-licensed fuel reprocessing plant is the Western New York Nuclear Service Center. The technical specifications of its NRC license are currently suspended by license amendment. Under the West Valley Demonstration Project Act, Public Law 96-368, 94 Stat. 1347 (codified as a note to 42 U.S.C. 2021a), the Department of Energy (DOE) is currently decommissioning portions of the plant. The NRC licensee, the New York State Energy Research and Development Authority, will complete the decommissioning work after DOE has completed its work. There is currently no application for another fuel reprocessing plant and the NRC does not anticipate any application in the foreseeable future.

The NRC proposes to revise the introductory text of § 50.82(b) to replace the term “non-power reactor licensees” with “non-power production or utilization facility licensees and fuel reprocessing plants” to ensure that all non-power facilities licensed under § 50.22 or § 50.21(a) or (c) are subject to the relevant termination and decommissioning regulations.

N. Clarification of Scope of License Termination Plan Requirement

The NRC is proposing to amend its regulations to clarify that the requirement for a license termination plan in § 50.82(a)(9) and § 52.110(i) applies only to nuclear power reactor licensees that commenced operation. This clarification is being proposed in response to apparent confusion among combined license holders that have sought to surrender their licenses before operation. By letter dated November 1, 2017 (ADAMS Accession No. ML17311A143), Duke Energy Florida informed the NRC that it would seek termination of the 10 CFR part 52 combined licenses for Levy Nuclear Plant Units 1 and 2 and would submit a license termination plan in accordance with § 52.110(i). Subsequently, South Carolina Electric & Gas Company (SCE&G) submitted a letter dated December 27, 2017 (ADAMS Accession No. ML17361A088), seeking withdrawal of the 10 CFR part 52 combined licenses for Virgil C. Summer Nuclear Station Units 2 and 3. The SCE&G request neither cited § 52.110 nor indicated that it would submit a license termination plan. Instead, SCE&G cited the Commission's final “Policy Statement on Deferred Plants” (52 FR 38077; October 14, 1987) (Policy Statement) to support its request for NRC approval to withdraw its combined licenses. The Policy Statement addresses holders of construction permits that defer or terminate plant construction. The Policy Statement provides that a permit holder can request to withdraw its permit and does not cite to the license termination provisions in 10 CFR part 50. The Policy Statement was issued prior to the promulgation of 10 CFR part 52 and has not been updated since, but there is nothing to prevent holders of a combined license from following the applicable parts of the Policy Statement while continuing to comply with the Commission's regulations and the terms and conditions of the combined license.

The requirement for a license termination plan in § 52.110(i) does not apply to plants that have not begun operating. While § 52.110(i) does refer to “[a]ll power reactor licensees,” the regulatory history and context indicates that § 52.110 as a whole applies only to plants that have started operation:

• The organization of § 52.110 generally follows the license termination process for an operating plant, from permanent cessation of operations to permanent removal of fuel to decommissioning activities to license termination. The requirement for a license termination plan should be understood in this context.
• The vast majority of the requirements in § 52.110 (including § 52.110(i)) either explicitly refer to, or make sense only in the context of, a plant that has operated and is undergoing decommissioning.
• The “[a]ll power reactor licensees” language also appears in § 50.82(a)(9), the 10 CFR part 50 analogue to § 52.110(i). But the NRC does not apply the similar requirements in § 50.82 to holders of construction permits even though construction permits fall within the definition of “License” in § 50.2. For example, the following construction permit terminations do not cite or otherwise address § 50.82: “Washington Public Power Supply System, Washington Nuclear Project, Unit 3; Order Revoking Construction Permit No. CPPR-154” (64 FR 4725; January 29, 1999); “Bellefonte Nuclear Plant, Units 1 and 2—Withdrawal of Construction Permit Nos. CPPR-122 for Unit 1 and CPPR-123 for Unit 2” (September 14, 2006) (ADAMS Accession No. ML061810505); and “Energy Northwest Nuclear Project No. 1—Termination of Construction Permit CPPR-134” (February 8, 2007) (ADAMS Accession No. ML070220011). And the rule issuing the “[a]ll power reactor licensees” language in § 50.82(a)(9)—the 1996 Final Rule—was directed at holders of operating licenses, not construction permits.
• According to the final rule issuing § 52.110, “Licenses, Certifications, and Approvals for Nuclear Power Plants” (72 FR 49351; August 28, 2007), § 52.110 and its companion regulation § 52.109 were intended to be analogous to the requirements in § 50.51 and § 50.82 for permanent shutdown of a nuclear power plant, its decommissioning, and the termination of the operating license.

For these reasons, § 52.110 is best understood to apply only to plants that began operation. However, to avoid confusion over the license termination plan requirement, the NRC proposes to amend § 52.110(i) so that it explicitly applies only to “power reactor licensees that have loaded fuel into the reactor.” As stated in the “Final Procedures for Conducting Hearings on Conformance With the Acceptance Criteria in Combined Licenses” (81 FR 43266; July 1, 2016), the NRC has historically understood operation as beginning with the loading of fuel into the reactor. Therefore, § 52.110(i) would apply to 10 CFR part 52 nuclear power reactor licensees that have begun to load fuel into the reactor.

A conforming change is also proposed in § 50.82(a)(9) to clarify that the requirement in that provision—that all 10 CFR part 50 nuclear power reactor licensees must submit an application for termination of license—applies to only those 10 CFR part 50 nuclear power reactor licensees that have loaded fuel into the reactor.

O. Removal of License Conditions and Withdrawal of Orders

The NRC is proposing to withdraw orders and remove license conditions that are substantively redundant with provisions in 10 CFR. Although NRC orders generally provide for their relaxation or rescission on a licensee-specific basis, use of that process would be an inefficient and unnecessary administrative burden on licensees and the NRC—with no impact on public health and safety—when a subsequent rule replaces the orders in their entirety for all applicable licensees. Therefore, the NRC is proposing to find that good cause is shown to rescind Order EA-06-137, “Order Modifying Licenses” (ADAMS Accession No. ML061600076), concerning mitigation strategies for large fires or explosions at nuclear power plants. This order was issued to certain licensees who received Order EA-02-026, “Order for Interim Safeguards and Security Compensatory Measures” (ADAMS Accession No. ML020510635), which required licensees to take specific interim compensatory measures, including mitigation strategies for large fires or explosions at nuclear power plants, in light of the then-high-level threat environment. Order EA-06-137 required that licensees to incorporate key mitigation strategies for large fires or explosions into their security plans. The requirement that these strategies be incorporated in security plans was subsequently relaxed by letter dated August 28, 2006, which permitted licensees to consent to having their licenses amended to incorporate a license condition on the subject. Several licensees had these license conditions imposed by administrative license amendment ( e.g., “Browns Ferry Nuclear Plant, Units 1, 2, and 3—Conforming License Amendments To Incorporate the Mitigation Strategies Required by Section B.5.b. of Commission Order EA-02-026 and the Radiological Protection Mitigation Strategies Required by Commission Order EA-06-137,” dated August 16, 2007). In its Power Reactor Security Requirements final rule, the NRC established in § 50.54(hh)(2) a regulation that provides a performance-based requirement that encompasses the mitigation strategies required under Order EA-06-137 and its associated license condition. The Mitigation of Beyond-Design-Basis Events rule subsequently moved § 50.54(hh)(2) to § 50.155(b)(3). As a result, neither Order EA-06-137 nor the license condition is necessary. Accordingly, the NRC proposes finding that good cause is shown to rescind Order EA-06-137 for each licensee that received the order. In addition, because § 50.155(b)(3) provides the same requirements as the license condition associated with Order EA-06-0137, the NRC proposes deeming the license condition removed from each applicable nuclear power reactor license.

Order EA-02-026 included a section, numbered B.5.b, in its attachment 2, requiring mitigation strategies for large fires or explosions at nuclear power plants. Extensive interactions among the NRC, industry, and licensees refined the strategies required by the order. In 2007, the NRC issued to all then-operating nuclear power reactor licensees an administrative license amendment ( e.g., “Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2—Conforming License Amendments to Incorporate the Mitigation Strategies Required by Section B.5.b. of Commission Order EA-02-026,” dated July 11, 2007), containing a license condition entitled, “Mitigation Strategy License Condition,” which required licensees to use 14 mitigation strategies. In the Power Reactor Security Requirements final rule, the NRC established in §§ 50.54(hh), 50.34(i), and 52.80(d) regulations that made the requirements of Order EA-02-026 generically applicable to nuclear power reactor licensees and applicants. In the Power Reactor Security Requirements final rule, the Commission explained that operating nuclear power reactor licensees already had procedures in place that complied with the new § 50.54(hh)(2). Licensees used the same implementation guidance to comply with the Mitigation Strategy License Condition as they used to comply with § 50.54(hh)(2); consequently, compliance with § 50.54(hh)(2) is sufficient to comply with the Mitigation Strategy License Condition. Subsequently, the NRC rescinded Order EA-02-026, section B.5.b by letter dated November 28, 2011, based on the fact that the regulations encompassed the order requirements. Because licensees comply with both the regulations and Mitigation Strategy License Condition via the same guidance, such that the former § 50.54(hh)(2) requirements encompass the license condition requirements, the NRC proposes concluding that § 50.155(b)(3) fully replaces the requirements that exist in the Mitigation Strategy License Condition and deeming that the Mitigation Strategy License Conditions imposed in 2007 are removed from the licenses for those licensees that received that license condition.

As discussed in section I.C., “Cyber Security,” of this document, the NRC imposed a license condition referencing the approved CSP in each 10 CFR part 50 license in the course of review and approval of the CSP. This proposed rule would remove that license condition once sufficient time has passed since the permanent removal of fuel from the reactor vessel.

Because this proposed rule would remove certain license conditions without actually amending the associated licenses, the NRC would issue by letter an administrative license amendment to each applicable licensee that would remove the relevant license condition(s) from that licensee's license and include revised license pages.

P. Changes for Consistent Treatment of Holders of Combined Licenses and Operating Licenses

The NRC proposes to revise § 50.36(c)(6), § 50.44(b), § 50.46(a)(1)(i), § 50.48(f), § 50.54(y), § 50.60(a), § 50.61(b)(1), § 50.62(a), § 50.71(e)(4), and 10 CFR part 50, Appendix I, Section IV.C., to provide consistent treatment for COL (Part 52) and operating license (Part 50) holders. These changes have the purpose of aligning regulatory applicabilities for COL holders upon submittal of the § 52.110(a) certifications with regulatory applicabilities for operating license holders upon submittal of the § 50.82(a)(1) certifications. In each section listed, the NRC would insert “or § 52.110(a)” following each instance of “§ 50.82(a)(1).”

The NRC proposes to revise incorrect references to § 52.110 in § 50.49(a), § 50.54(o), § 50.65(a)(1), and § 52.110(e) by replacing “§ 52.110(a)(1)” with “§ 52.110(a).” The NRC proposes to insert a reference to § 52.110 following an existing reference to § 50.82 in § 50.54(w)(4)(ii), § 50.54(w)(4)(iii), § 50.75(e)(1)(ii)(B), § 50.75(e)(1)(v), § 50.75(h)(1)(iv), and § 50.75(h)(2). The NRC proposes to remove the words “under this part” from § 50.54(w) introductory text because paragraph (w) is also applicable to holders of combined licenses issued under 10 CFR part 52 as stated in the introductory text for § 50.54. Finally, the NRC proposes to revise an incorrect reference in § 50.59(b) by replacing “§ 50.110” with “§ 52.110(a).”

V. Specific Requests for Comments

The NRC is seeking public comments on this proposed rule. The agency is particularly interested in comments and supporting rationale from the public on the following:

• PSDAR Approval: The current decommissioning regulations establish that once a licensee permanently ceases operation of the nuclear power reactor, it cannot undertake any major decommissioning activities until it provides the public and the NRC with additional information. The NRC requires that the licensee submit this information in the form of a PSDAR, which consists of the licensee's proposed decommissioning activities and schedule through license termination, a discussion of the reasons for concluding that the proposed activities will be bounded by existing analyses of environmental impacts, and a site-specific cost estimate for the proposed activities. The PSDAR is made available to the public for comment and is subject to NRC review (but not approval). Additionally, the current decommissioning regulations prohibit, at any time, the performance of any decommissioning activity that may result in significant environmental impacts not previously reviewed. Under this regulatory framework, licensees are not required to have an NRC-approved decommissioning plan; instead, 90 days after the NRC has received the licensee's PSDAR, licensees may perform, under 10 CFR 50.59, those major decommissioning activities that are bounded by existing environmental analyses. Therefore, no site-specific NEPA review is required and there is no hearing opportunity under 10 CFR part 2 before these decommissioning activities begin. To perform decommissioning activities that are not bounded by existing environmental analyses, however, a licensee would have to submit a request for a license amendment or an exemption request, which would trigger a site-specific NEPA review and hearing opportunity under 10 CFR part 2. Additionally, at least two years before termination of the license, the licensee must submit an application for termination of license and a license termination plan, which must be approved by the NRC. The requirement to approve the license termination plan also triggers a site-specific NEPA review and hearing opportunity under 10 CFR part 2.

As part of the development of the proposed rule, the NRC staff evaluated whether the NRC should explicitly approve each licensee's PSDAR before allowing major decommissioning activities to begin. The staff concluded that based on lessons learned and experience, there is currently no indication that requiring approval of a PSDAR has any substantial impact on the public health and safety. However, the NRC is gathering additional feedback from the public.

As part of this rulemaking, should the NRC require approval of the PSDAR, a site-specific environmental review, and hearing opportunity before a licensee undertakes any decommissioning activity? Other than NRC review and approval of the PSDAR, are there other activities that could help to increase transparency and public trust in the NRC regulatory framework for decommissioning? Should the rule provide a role for the states or local governments in the process? What should that role be? What are the advantages or disadvantages of various roles? Please provide an explanation for your response.

• Timeframe for Decommissioning: For nuclear power reactor licensees, 10 CFR 50.82(a)(3) and 10 CFR 52.110(c) state that decommissioning must be completed within 60 years of permanent cessation of operations. In this proposed rule, the NRC is not proposing changes to the decommissioning timeframe requirements.

What are the advantages and disadvantages of requiring prompt decontamination rather than allowing up to 60 years to decommission a site? As part of its review of a PSDAR, what are the advantages and disadvantages of NRC evaluating and making a decision about the timeframe for decommissioning on a site-specific basis?

• Emergency Planning: As discussed in the “Technical Basis for the Graded Approach” and “Emergency Preparedness” sections of this document, although the spectrum of credible accidents and operational events requiring an emergency response is reduced at a decommissioning nuclear power reactor as compared to that for an operating nuclear power reactor, reliable emergency preparedness functions are still required to ensure public health and safety in the event of a zirconium fire scenario.

The NRC has concluded that dry cask storage and spent fuel pools are both very safe. What are the advantages and disadvantages of requiring dedicated radiological emergency planning, including a 10-mile EPZ, until all spent nuclear fuel at a site is removed from the spent fuel pool and placed in dry cask storage? Is there additional information the NRC should consider in evaluating whether all-hazards planning would be as effective as dedicated radiological emergency planning?

The NRC has determined that 10 hours would be a sufficient amount of time for an emergency response to a spent fuel pool accident based on an all-hazards plan. Is there additional information the NRC should consider in evaluating this issue?

• Emergency Response Data Systems: Nuclear power facilities that are shutdown permanently or indefinitely are currently not required to maintain ERDS. These systems transmit near-real-time electronic data between the licensee's onsite computer system and the NRC Operation Center. Licensees in Level 1 would maintain a capability to provide meteorological, radiological, and spent fuel pool data to the NRC within a reasonable timeframe following an event. What are the advantages and disadvantages of requiring nuclear power plant licensees to maintain those aspects of ERDS until all spent fuel is removed from the pool?

• Cyber Security: The proposed rule applies cyber security requirements to Level 1 plants. However, a licensee in Level 2 would not be required to maintain a cyber security plan because the NRC has determined that there is little chance that the spent fuel in the SFP could heat up to clad ignition temperature within 10 hours. What are the advantages and disadvantages of extending cyber security requirements to shutdown nuclear power plants until all spent fuel is transferred to dry cask storage?

• Insurance: The proposed rule would allow nuclear power reactor licensees in decommissioning to reduce the offsite liability and onsite property insurance amounts that they are required to maintain once a plant enters Level 2. The transition to Level 2 financial protection amounts would be optional for licensees and they would have to submit an analysis that demonstrates a reduced risk of a zirconium fuel cladding fire in the SFP. What are the advantages and disadvantages of requiring the existing level of insurance to be maintained until all spent fuel is in dry cask storage (Level 3)?

• Financial Assurance: Pursuant to § 50.75, “Reporting and recordkeeping for decommissioning planning,” specifically paragraph (b)(1), nuclear power reactor licensees and applicants must certify that reasonable assurance for radiological decommissioning funding has been (for licensees) or will be (for applicants) provided in an amount that may be more, but not less, than the generic amount provided by the Commission's regulations ( i.e., the table of minimum amounts under § 50.75(c)). Alternatively, under § 50.75(b)(4), the certified amount of funding may be based on a site-specific cost estimate for decommissioning the facility.

The current table of minimum amounts (also referred to as the minimum decommissioning formula) has not been updated for over 30 years. The NRC is considering updates to the generic decommissioning funding formula to make it more reflective of current cost considerations.

What are the advantages and disadvantages of updating the formula to reflect recent data and to cover all estimated radiological decommissioning costs rather than the bulk of the costs?

• Site-Specific Cost Analysis: Currently, licensees can use either the generic amount under 10 CFR 50.75(c) or a site-specific cost estimate under 10 CFR 50.75(b)(4) to determine the certified amount of radiological decommissioning funding. As provided in 10 CFR 50.82(a)(8)(ii) and 10 CFR 52.110(h)(2), a licensee may withdraw funds from the decommissioning trust fund up to a cumulative total of 3 percent of the generic amount calculated under 10 CFR 50.75(c) for decommissioning planning purposes at any time without prior notification to the NRC. After submittal of the certifications of permanent shutdown and fuel removal required under 10 CFR 50.82(a)(1) and 10 CFR 52.110(a) and commencing 90 days after the NRC has received the PSDAR, the licensee may use up to an additional 20 percent of the decommissioning funds prescribed in 10 CFR 50.75(c) for decommissioning purposes. The licensee is prohibited from using the remaining 77 percent of the generic decommissioning funds until a site-specific decommissioning cost estimate is submitted to the NRC. Requirements in 10 CFR 50.82(a)(8)(iii) and 10 CFR 52.110(h)(3) establish that a licensee shall provide a site-specific decommissioning cost estimate within 2 years following permanent cessation of operations. If the estimate of costs provided with the PSDAR is a site-specific cost estimate, this requirement can be satisfied with the PSDAR submittal.

What are the advantages and disadvantages of requiring a full site investigation and characterization at the time of shutdown? What are the advantages and disadvantages of eliminating the formula and requiring a site-specific cost estimate during operations?

• Decommissioning Trust Fund: Under the NRC's existing regulations and this proposed rule, the amounts set aside for radiological decommissioning should not be used for the maintenance and storage of spent fuel in the spent fuel pool, or for the design or construction of spent fuel dry storage facilities, or for other activities not directly related to the long-term storage, radiological decontamination or dismantlement of the facility, or decontamination of the site.

Should the NRC's regulations allow decommissioning trust fund assets to be used for spent fuel management if (1) there is a projected surplus in the fund based on a comparison to the expected costs identified in a site-specific cost estimate and (2) the assets are returned to the fund within an established period of time? What are the advantages and disadvantages of allowing decommissioning trust fund assets to be used for those purposes? What are the advantages and disadvantages of allowing decommissioning trust fund assets to be used for non-radiological site restoration prior to the completion of radiological decommissioning?

• Timing of Decommissioning Funding Assurance Reporting: This proposed rule would change the timing of the decommissioning funding assurance reporting requirements in § 50.75(f)(1) to coordinate them with the ISFSI decommissioning reporting requirements in § 72.30. Under this proposed rule, operating reactors would be permitted to submit decommissioning funding status reports triennially instead of biennially.

What are the advantages and disadvantages to extending the reporting frequency from two years to three years? Does this change affect the risk of insufficient decommissioning funding? Please provide an explanation for your response.

• Backfit Rule: For nuclear power reactor licensees, the NRC's backfitting provisions are located in § 50.109, “Backfitting,” and the issue finality provisions are in 10CFRpart52 (the “Backfit Rule”). The language of the Backfit Rule clearly applies to a licensee designing, constructing, or operating a nuclear power facility. For example, § 50.109(a)(1) defines “backfitting” to mean changes to, among other things, the procedures or organization required to design, construct, or operate a facility.

This proposed rule states that the Backfit Rule applies to decommissioning nuclear power plants. What are the advantages and disadvantages of applying the Backfit Rule to decommissioning nuclear power plants?

• Exemptions: As stated in this proposed rule, one of the goals of amending these regulations is to reduce the need for regulatory exemptions. 10 CFR 50.12 states that the Commission may grant exemptions from the requirements of the regulations in 10 CFR part 50 if the request will not present an undue risk to the public health and safety, and is consistent with the common defense and security. What are the advantages and disadvantages of the current 10 CFR 50.12 approach to decommissioning-related exemptions? What standard should the NRC apply in determining whether to grant exemptions from the new or amended regulations? What are the advantages and disadvantages of providing an opportunity for the public to weigh in on such exemption requests? Are there other process changes the NRC should consider in determining whether to grant exemptions from the new or amended regulations?

• Applicability: Section III of this document provides a discussion of the applicability of this proposed rule. Specifically, there is a discussion for the applicability to NRC licensees during operations and to ISFSI-Only and Standalone ISFSI/Decommissioned Reactor Sites. Permanently shutdown nuclear power plants will be at different stages of decommissioning when the new decommissioning regulations become effective and will have previously received varying regulatory exemptions.

Can you foresee any implementation issues with the proposed rule as it is currently written? For any new or amended requirement included in this proposed rule, how should the requirement apply to sites currently in different stages of decommissioning?

• Insurance for Specific License ISFSI: A 10 CFR part 50 or 10 CFR part 52 nuclear power reactor licensee with a 10 CFR part 72 general license ISFSI at the reactor site is subject to the financial protection requirements under 10 CFR part 140, whereas a specific license ISFSI under 10 CFR part 72 is not. In SECY-04-0176, “Exemption Requests to Reduce Liability Insurance Coverage for Decommissioning Reactors after Transfer of all Spent Fuel from a Spent Fuel Pool to Dry Cask Storage,” dated September 29, 2004 (ADAMS Accession No. ML040850518), the NRC staff noted that general license ISFSIs subject to the requirements under 10 CFR part 72 were also subject to the requirements of a 10 CFR part 50 license and by virtue of this license, they are required to maintain some level of liability insurance under section 170, “Indemnification and Limitation of Liability,” of the AEA (known as the Price-Anderson Act) and the NRC's implementing regulations at 10 CFR part 140. Further, the NRC staff acknowledged that there was little technical difference between a general license ISFSI and a specific license ISFSI.

The NRC recognizes that as a reactor site is decommissioned, eventually all that remains of the 10 CFR part 50 or part 52 licensed site is a general license ISFSI under 10 CFR part 72, which is essentially the same as a specific license ISFSI under 10 CFR part 72. Considering that 10 CFR part 72 specific license ISFSIs have no financial protection requirements, should the NRC address the disparity between specific license and general license ISFSIs as a part of this rulemaking? Please provide an explanation for your response.

• Recordkeeping Requirements for Facilities Licensed under 10 CFR part 52: The current appendices in 10 CFR part 52 contain section X, “Records and Reporting,” for all of the certified designs codified in 10 CFR part 52. Section X requires, in part, that all departures from the certified design be recorded and those records kept throughout the term of the license. However, as part of this rulemaking, the NRC is proposing to change the record retention requirements for nuclear power reactors in the decommissioning process such that they no longer need to retain certain records associated with SSCs that are no longer in service or necessary to keep the plant in a safe condition. The NRC is considering making conforming changes to section X of the applicable appendices to 10 CFR part 52 to allow this change to apply to records of departures from the certified design as well as the associated SSCs. Given the already existing change control procedures in the appendices to 10 CFR part 52, as well as the significant changes in recordkeeping technology since the NRC's record retention requirements were introduced ( i.e., digital media instead of paper copies), should additional changes be made to the 10 CFR part 52 appendices as a part of this rulemaking, and would such changes be beneficial to 10 CFR part 52 licensees or add efficiency to the decommissioning process for these facilities? Please provide an explanation for your response.

• Identical Requirements under § 50.82 and § 52.110: As part of this rulemaking, the NRC proposes to revise § 52.110 to make the same changes proposed in § 50.82 for the reasons previously discussed and for consistency. The NRC also proposes to add paragraphs (h)(5) through (h)(7) to § 52.110 with site-specific decommissioning cost estimate reporting requirements that are identical to the requirements in § 50.82(a)(8)(v) through (vii). Given that the decommissioning financial assurance requirements in § 52.110 are identical to the requirements in § 50.82, should the NRC consider removing the specific requirements from § 52.110(f)-(h) and instead add a reference in § 52.110 to the identical regulations in § 50.82(a)(6)-(8)? Are there any other provisions in § 52.110 that the NRC should consider removing and replacing with a reference to an identical requirement in § 50.82 ( e.g., the decommissioning requirements under § 52.110(c)-(e))? Please provide an explanation for your response.

• Removal of License Conditions and Withdrawal of Orders: This rulemaking seeks to improve regulatory efficiency by removing license conditions and withdrawing an order for which substantively identical requirements have been imposed by rulemaking. This would avoid the future administrative expenditures by licensees and the NRC to accomplish the removal of these requirements on a license-specific basis through a generic regulatory action either upon the effective date of the final rule or when conditions permit the removal during the decommissioning process. The NRC has identified certain orders that were issued following the terrorist events of September 11, 2001, license conditions regarding these orders, and license conditions regarding cyber security implementation as having substantively identical requirements made generically applicable through rulemaking. Because these license-specific requirements are duplicative with other generic requirements, the NRC concludes there would be no reduction in safety. Please provide any comments you may have on rescinding Order EA-06-137 and the related license conditions. As part of this rulemaking, are there other license-specific requirements in license conditions or orders that have substantively identical generic requirements that should be addressed in this rulemaking? Please provide an explanation for your response.

• Spent Fuel Management Planning: Section IV.K of this document discusses spent fuel management planning in the § 50.54(bb) regulation. The § 50.54(bb) current rule language requires NRC preliminary approval and final review, as part of any proceeding for continued licensing under part 50 or part 72, of the IFMP. The discussion in Section IV.K points out that the proceedings for continued licensing under part 50 or part 72 no longer exist. Therefore, the proposed rule includes language intended to clarify the current IFMP approval process by requiring submittal of the IFMP for NRC review and approval by license amendment. What, if any, challenges do you foresee with implementing this part of the proposed rule? Please provide an explanation for your response.

The § 50.54(bb) current rule language requires licensees to notify the NRC of any significant changes to the IFMP. As discussed in section IV.K, the NRC proposes to revise this requirement to require licensees to submit to the NRC any changes to the IFMP as an application for an amendment to its license. The NRC is also considering replacing the notification requirement with a change control provision to specify what changes a licensee can make to the IFMP without NRC approval. Examples of change control provisions in the current NRC regulations include § 50.54(a) for quality assurance programs and § 50.54(q) for emergency plans. If the NRC includes a similar change control provision in § 50.54(bb), what should the safety and environmental criteria be for determining whether a licensee could make a change to its IFMP without seeking NRC approval? For example, the NRC could permit changes that are not considered to be reductions in the commitments, including (1) changes to the planned actions for managing spent fuel that result in an addition of one or more SSCs that the licensee relies on for irradiated fuel management, and (2) changes to the projected cost or funding for managing irradiated fuel that is already included in the report required by 10 CFR 50.82(a)(8)(vii) or 10 CFR 52.110(h)(7). Should the NRC also include recordkeeping and reporting provisions for a licensee to retain a record of each change to the IFMP made without prior NRC approval and submit a report to the NRC of those changes? If so, what should be the timeframe for the records to be retained and the timeframe for reporting to the NRC after the change is made, taking into consideration the estimated frequency of performing IFMP changes? Please provide an explanation for your response.

VI. Section-by-Section Analysis

The following paragraphs describe the specific changes proposed by this rulemaking.

Appendix G to 10 CFR Part 20, Requirements for Transfers of Low-Level Radioactive Waste Intended for Disposal at Licensed Land Disposal Facilities and Manifests

In section III, paragraph E.1., this proposed rule would remove the word “or” and add in its place the word “of” and it would also remove the phrase “20 days”, and add in its place the phrase “45 days”.

Section 26.3 Scope

In § 26.3, this proposed rule would revise paragraph (a) by subdividing it into two subparagraphs, (a)(1) and (2), to include the NRC's docketing of a license holder's certifications required under §§ 50.82 and 52.110(a).

Section 26.825 Criminal Penalties

In § 26.825, this proposed rule would revise paragraph (b) to remove the number “26.3” from the list of regulations in 10 CFR part 26 that are excluded from § 26.825(a).

Section 50.1 Basis, Purpose, and Procedures Applicable

In § 50.1, this proposed rule would add language clarifying that the regulations in 10 CFR part 50 provide for the licensing of production and utilization facilities through the termination of the associated 10 CFR part 50 licenses.

Section 50.2 Definitions

In § 50.2, this proposed rule would retain the existing definition of certified fuel handler and add an alternative definition for the purposes explained elsewhere in this document. This proposed rule also would add a definition for a non-power production or utilization facility.

Section 50.36 Technical Specifications

In § 50.36, this proposed rule would revise paragraph (c)(6) to insert “or § 52.110(a)” following “§ 50.82(a)(1).”

Section 50.38 Ineligibility of Certain Applicants

This proposed rule would revise § 50.38 by including the current text as paragraph (a) and by adding paragraph (b) to state that the prohibition in paragraph (a) of this section does not apply to a person, corporation, or other entity seeking a license for a facility that is not a production or utilization facility.

Section 50.44 Combustible Gas Control for Nuclear Power Reactors

In § 50.44, this proposed rule would revise paragraph (b) to insert “or § 52.110(a)” following “§ 50.82(a)(1).”

Section 50.46 Acceptance Criteria for Emergency Core Cooling Systems for Light-Water Nuclear Power Plants

In § 50.46, this proposed rule would revise paragraph (a)(1)(i) to insert “or § 52.110(a)” following “§ 50.82(a)(1).”

Section 50.47 Emergency Plans

This proposed rule would make conforming changes to paragraph (b) in § 50.47 and would add paragraph (f) denoting when the planning standards for offsite emergency plans in paragraph (b) of this section do not apply.

Section 50.48 Fire Protection

In § 50.48, this proposed rule would revise paragraph (f) to insert “or § 52.110(a)” following “§ 50.82(a)(1).”

Section 50.49 Environmental Qualification of Electric Equipment Important to Safety for Nuclear Power Plants

In § 50.49, this proposed rule would revise paragraph (a) by replacing “§ 52.110(a)(1)” with “§ 52.110(a).”

Section 50.51 Continuation of License

In § 50.51, this proposed rule would remove the phrase, “to authorize ownership and possession of the production or utilization facility,” for reasons discussed elsewhere in this document.

Section 50.54 Conditions of Licenses

In § 50.54, this proposed rule would revise footnote 2 to the table in paragraph (m)(2)(i) to indicate when a Shift Technical Advisor is not required. Paragraph (o) would be revised by replacing “52.110(a)(1)” with “52.110(a).” The NRC also would revise § 50.54(p) to include the definitions for change and decrease in safeguards effectiveness for use in paragraph (p), would revise and redesignate existing paragraphs (p)(1) and (2) as (p)(2) and (3), would redesignate paragraphs (p)(3) and (4) as paragraphs (p)(5) and (6), and would add new paragraphs (p)(1) and (4). A portion of the existing text in paragraphs (p)(1) and (2) would be used to create new paragraph (p)(4).

This proposed rule would revise: Paragraph (q)(1) to clarify that the definitions are for use in paragraph (q), paragraph (q)(1)(iii) to remove the reference to appendix E to 10 CFR part 50, paragraph (q)(2) to add clarification to the applicability, paragraph (q)(3) to add applicable emergency planning requirements, paragraphs (q)(4) and (5) to remove the phrase “after February 21, 2012,” and add new paragraphs (q)(7) and (8) to add the requirements for licensees after the NRC dockets their certifications required for decommissioning under § 50.82(a)(1) or § 52.110(a).

Paragraph (s)(2)(ii) would be revised by removing the phrase “after April 1, 1981,” and paragraph (s)(3) would be revised by adding clarification at the beginning of the sentence that if the standards apply to offsite radiological response plans then the NRC will base its findings on a review of FEMA findings and determinations.

Paragraph (t) would be revised by replacing “.” with “or” in the second sentence of paragraph (t)(1)(ii), adding new subparagraph (t)(1)(iii) to clarify the interval at which the licensee's emergency preparedness plan must be reviewed after the NRC has docketed the certifications required for decommissioning, and by adding new paragraph (t)(3) to state that the review requirement is no longer required once all fuel is in dry cask storage.

Paragraph (w) would be revised by removing the words “under this part” from the introductory text, adding a reference to § 52.110 in paragraphs (w)(4)(ii) and (w)(4)(iii), and adding new paragraphs (w)(5) and (6) to include the financial protection requirements for production or utilization facilities undergoing decommissioning.

Paragraph (y) would be revised to insert “or § 52.110(a)” following “§ 50.82(a)(1).”

Paragraph (bb) would be revised by restructuring the paragraph and revising the requirements of an irradiated fuel management plan.

Section 50.59 Changes, Tests, and Experiments

In § 50.59, this proposed rule would revise paragraph (b) to correct a reference to § 52.110(a). It would also revise paragraph (d)(3) to include the exception for when the records of changes requirement in paragraph (d)(3) applies.

Section 50.60 Acceptance Criteria for Fracture Prevention Measures for Lightwater Nuclear Power Reactors for Normal Operation

In § 50.60, this proposed rule would revise paragraph (a) to insert “or § 52.110(a)” following “§ 50.82(a)(1).”

Section 50.61 Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events

In § 50.61, this proposed rule would revise paragraph (b)(1) to insert “or § 52.110(a)” following “§ 50.82(a)(1).”

Section 50.62 Requirements for Reduction of Risk From Anticipated Transients Without Scram (ATWS) Events for Light-Water-Cooled Nuclear Power Plants

In § 50.62, this proposed rule would revise paragraph (a) to insert “or § 52.110(a)” following “§ 50.82(a)(1).”

Section 50.65 Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants

In § 50.65, this proposed rule would revise paragraph (a)(1) by replacing “§ 52.110(a)(1)” with “§ 52.110(a).”

Section 50.71 Maintenance of Records, Making of Reports

In § 50.71, this proposed rule would revise paragraph (c) by including the current text as paragraph (c)(1) and it would add new paragraph (c)(2) to add records requirements for licensees for whom the NRC has docketed the certifications required for decommissioning.

Paragraph (e)(4) would be revised to insert “or § 52.110(a)” following “§ 50.82(a)(1).”

Section 50.75 Reporting and Recordkeeping for Decommissioning Planning

In § 50.75, this proposed rule would revise paragraph (a) by clarifying the availability of funds to decommission a facility as defined in § 50.2.

Paragraph (b)(1) would be revised by replacing “financial” with “reasonable” assurance and other conforming changes; paragraph (b)(3) would be revised by removing the phrase “as acceptable to the NRC” from the end of the paragraph; paragraph (b)(4) would be revised to include a site-specific decommissioning cost estimate and the second sentence of current paragraph (b)(4) would be moved to become a new paragraph (b)(5).

Paragraph (e)(1) would be revised to include the term “reasonable assurance of funds to decommission,” and paragraphs (e)(1)(i) and (ii) would be revised to include the description of “decommissioning cost” before the word estimate throughout each paragraph. Paragraphs (e)(1)(ii)(B) and (e)(1)(v) would be revised to add a reference to § 52.110.

Paragraph (f) would be amended by revising (f)(1) to include the requirement for a report to include information regarding any potential decommissioning shortfall, it would be further amended by removing paragraph (f)(2) and redesignating (f)(3) through (5) as (f)(2) through (4) with minor revisions.

Paragraphs (h)(1)(iii) and (iv) and (h)(2) would be revised to remove the reference to three office directors within the NRC for the submission of written notice of the intention to make a payment or disbursement of funds and replace it with the Document Control Desk. Paragraphs (h)(1)(iv) and (h)(2) would be revised to add a reference to § 52.110.

Section 50.82 Termination of License

In § 50.82, this proposed rule would revise paragraph (a)(2) to provide clarification as to when a licensed nuclear power reactor is no longer considered to be a utilization facility. It also would revise paragraph (a)(4)(i) to clarify that licensees provide the basis for whether the environmental impacts from site-specific decommissioning activities are bounded by federally issued environmental review documents. The phrase “including the projected cost of managing irradiated fuel” would be removed at the end of the last sentence. Paragraph (a)(4)(ii) would be revised to include the requirement for the NRC to include the irradiated fuel management plan in the notice of the receipt of the PSDAR in the Federal Register and to allow the public to comment.

Paragraph (a)(6)(ii) would be revised to provide clarification.

Paragraph (a)(8)(i)(A) would be revised to remove the words “legitimate decommissioning” and to replace the word “decommissioning” with “decommission.” Paragraph (a)(8)(ii) would be revised to clarify paragraph (c) to § 50.75 is where the specified amount is located.

Paragraph (a)(8)(v) would be revised to spell out the acronym DCE, decommissioning cost estimate, and to include the ability for the licensee to combine the reporting requirements of 10 CFR part 72, § 50.82(a)(8)(v), and § 50.82(a)(8)(vii).

Paragraph (a)(8)(vii) would be revised to spell out the acronym DCE, decommissioning cost estimate.

Paragraph (a)(9) would be revised to clarify that all nuclear power reactors that have loaded fuel into the reactor must submit an application for termination of a license and paragraph (a)(9)(ii)(F) would be revised to include the requirement to identify funding sources for license termination, spent fuel management, and ISFSI decommissioning.

The introductory text of paragraph (b) would be revised to replace the term “non-power reactor licensees” with “non-power production or utilization facilities and fuel reprocessing plants.”

Paragraph (b)(6) would be redesignated as (b)(8) and new paragraphs (b)(6) and (7) would be added to include the criteria for when a non-power production or utilization facility or fuel reprocessing plant licensed under 10 CFR part 50 is no longer considered a production or utilization facility.

Section 50.109 Backfitting

This proposed rule would revise § 50.109 in its entirety to provide backfitting provisions for reactors both before and during decommissioning and to require that a documented evaluation for a modification necessary to bring a facility into compliance with a license or the rules or orders of the Commission, or into conformance with the licensee's written commitments, must include a consideration of the costs of imposing the modification.

Section 50.155 Mitigation of Beyond-Design-Basis Events

This proposed rule would add new paragraphs (h)(6), (h)(7) and (h)(8) that would deem removed certain license conditions and withdraw certain orders made redundant by regulations imposing substantively identical requirements.

Section 50.200 Power Reactor Decommissioning Emergency Plans

This proposed rule would add new § 50.200 that would contain alternate emergency preparedness requirements for nuclear power reactor facilities in decommissioning.

Appendix E to 10 CFR Part 50, Emergency Planning and Preparedness for Production and Utilization Facilities

This proposed rule would revise section I. Introduction of appendix E to 10 CFR part 50 by removing paragraph 6.

Section IV. Content of Emergency Plans of appendix E to 10 CFR part 50 would be revised by removing from paragraph 4 the phrases “of the later of the date” and “or December 23, 2011,” from the first sentence; new paragraph 8 would be added to inform licensees that the requirements of paragraphs 4, 5, and 6 of this section are no longer required once the NRC dockets the licensee's certifications required for decommissioning; paragraphs A.7., A.9., B.1., C.2., E.8.c., and I. would all be revised by removing the “by date” phrases; paragraph D.4. would be removed; the last sentence of paragraph E.8.d. would be removed; in paragraph F.2.d., the end of the 3rd sentence beginning with the word “and” would be removed; paragraph F.2.j(v) would be removed and reserved; and new paragraph F.2.k would be added to require licensees to follow the biennial exercise requirements in paragraph F.2 of appendix E to 10 CFR part 50 after the NRC dockets the certifications required for decommissioning.

This proposed rule would revise section VI. Emergency Response Data System of appendix E to 10 CFR part 50 by removing the date in paragraph 4.a. and the date in paragraph 4.d., also in paragraph 4.d. it would remove the phrase “, whichever comes later” from the first sentence.

Appendix I to 10 CFR Part 50, Numerical Guides for Design Objectives and Limiting Conditions for Operation To Meet the Criterion “As Low as Is Reasonably Achievable” for Radioactive Material in Light-Water-Cooled Nuclear Power Reactor Effluents

This proposed rule would revise section IV.C of appendix I to 10 CFR part 50 by inserting “or § 52.110(a)” following “§ 50.82(a)(1).”

Section 51.53 Postconstruction Environmental Reports

This proposed rule would revise the first sentence in paragraph (d) to include applicants for a license amendment approving an irradiated fuel management plan under § 50.54(bb). The proposed rule would also add references to § 50.82 and § 52.110 after “license termination plan.”

Section 51.95 Postconstruction Environmental Impact Statements

This proposed rule would revise the first sentence in paragraph (d) to refer to an amendment approving an irradiated fuel management plan under § 50.54(bb), the license termination plan under § 50.82 or § 52.110, or a decommissioning plan under § 50.82.

Section 52.0 Scope

In § 52.0, this proposed rule would add language clarifying that the regulations in 10 CFR part 52 remain effective through the termination of the associated 10 CFR part 52 licenses.

Section 52.63 Finality of Standard Design Certifications

This proposed rule would revise paragraph (b)(2) by removing the last sentence and by adding new paragraphs (b)(2)(i) and (ii) regarding the recordkeeping and retention requirements for departures from the design of a facility.

Section 52.109 Continuation of Combined License

In § 52.109, this proposed rule would remove the phrase, “to authorize ownership and possession of the production or utilization facility,” for reasons discussed elsewhere in this document.

Section 52.110 Termination of License

This proposed rule would revise paragraph (b) as paragraph (b)(1) and would add paragraph (b)(2) to provide clarification as to when a facility licensed under 10 CFR part 52 is no longer considered to be a production or utilization facility. Paragraph (d)(1) would be revised to clarify that licensees provide the basis for whether the environmental impacts from site-specific decommissioning activities are bounded by federally issued environmental review documents, and the phrase “site-specific decommissioning cost estimate” would be added at the end of the last sentence. Paragraph (d)(2) would be revised to include the requirement for the NRC to include the irradiated fuel management plan in the notice of the receipt of the PSDAR in the Federal Register and to allow the public to comment.

Paragraph (e) would be revised by replacing “§ 52.110(a)(1)” with “§ 52.110(a).”

Paragraph (f)(2) would be revised to clarify the decommissioning activities licensees shall not perform. Paragraph (h)(1)(i) would be revised to remove the phrase “legitimate decommissioning,” paragraph (h)(2) would be revised to include a more specific regulatory reference, and paragraphs (h)(5) through (8) would be added with requirements for the submission of financial status reports. Paragraph (i) would be revised to clarify that all nuclear power reactor licensees that have loaded fuel into the reactor must submit an application for termination of a license. Paragraph (i)(2)(vi) would be revised to include identification of sources of funds for license termination, spent fuel management, and ISFSI decommissioning, as applicable.

Section 72.13 Applicability

This proposed rule would revise § 72.13 by adding a new paragraph (e) to incorporate conforming changes to match technical changes elsewhere in the rule.

Section 72.30 Financial Assurance and Recordkeeping for Decommissioning

This proposed rule would revise § 72.30 by removing the second sentence in paragraph (c). The proposed revisions would create new paragraphs (b)(1) through(3) and redesignate the existing paragraphs (b)(1) through(6) as new (b)(3)(i) through(vi).

Section 72.32 Emergency Plan

In § 72.32, this proposed rule would clarify that the requirement for having an emergency plan applies when the proposed ISFSI would not be located on the site or within the exclusion area of a nuclear power reactor licensed under 10 CFR parts 50 or 52. The proposed revisions would consolidate the current language and remove redundancies by using standardized language consistent with other proposed rule provisions.

Section 72.44 License Conditions

This proposed rule would revise § 72.44 by adding a sentence to paragraph (f) to indicate that licensees need not comply with the requirements of paragraph (f) once all spent fuel has been removed from the site.

Section 72.62 Backfitting

This proposed rule would revise paragraph (a)(2) to clarify that the backfitting provisions under this part continue to apply during decommissioning.

Section 72.72 Material Balance, Inventory, and Records Requirements for Stored Material

This proposed rule would revise paragraph (d) by breaking it into three paragraphs. The last sentence of the current paragraph (d) would become paragraph (d)(3). New text is proposed for paragraph (d)(2) and minor revisions are proposed for paragraph (d)(1).

Section 72.212 Conditions of General License Issued Under § 72.210

This proposed rule would revise § 72.212 by adding new paragraphs (b)(9)(vii)(A) and (B) regarding the protection of spent fuel after the NRC dockets the decommissioning certifications. Paragraph (b)(9)(vii)(A) would allow a licensee to voluntarily provide for physical protection of the spent fuel under Subpart H of this part and § 73.51 of this chapter. Paragraph (b)(9)(vii)(B) would require a licensee who elects to provide physical protection under Subpart H of this part and § 73.51 of this chapter to notify the NRC of this decision using the provisions of § 50.54(p)(2).

Section 72.218 Termination of License

This proposed rule would revise § 72.218 by revising paragraphs (a) and (b) and removing paragraph (c). Paragraph (a) is revised to reference the decommissioning requirements in § 50.82 or § 52.110 that apply to the general license and paragraph (b) is revised to state when the general license is considered terminated.

Section 73.51 Requirements for the Physical Protection of Stored Spent Nuclear Fuel and High-Level Radioactive Waste

This proposed rule would revise § 73.51 by removing text from paragraph (a), (a)(1), (a)(2), and adding new paragraph (a)(3). Paragraph (a)(3) would be added to require notification to the NRC under the provisions of § 72.212(b)(9)(vii) of this chapter by a licensee who elects to provide physical protection under Subpart H of 10 CFR part 72.

Section 73.54 Protection of Digital Computer and Communications Systems and Networks

This proposed rule would revise § 73.54 by removing the introductory text of the section and revising the introductory text of paragraphs (a), (b), and (c), and adding new paragraphs (i), and (j). The introductory text of paragraph (a) would be revised to capture that the rule applies during operation and decommissioning. Minor edits would be made to paragraphs (b) and (c). Paragraph (i) states that the requirements of § 73.54 no longer apply once the criteria in (i)(1) and (2) are met. Paragraph (j) provides for the removal of the cyber security license condition.

Section 73.55 Requirements for Physical Protection of Licensed Activities in Nuclear Power Reactors Against Radiological Sabotage

This proposed rule would revise § 73.55 by clarifying in paragraph (b)(3) that a licensee's physical protection program must be designed to prevent significant core damage until the NRC dockets the certifications required for decommissioning.

New paragraphs (b)(9)(ii)(B)( 1 ), ( 2 ), ( 2 )( i ), and ( 2 )( ii ) would be added to provide additional clarification for licensees implementing fitness for duty programs.

Paragraph (c)(6) would be revised by replacing the text beginning with the words “that describes” through the end of the sentence with the phrase, “in accordance with the requirements of § 73.54 of this part.”

Paragraph (e)(9)(v)(A) would be revised to provide clarification for when the reactor control room would not be considered a vital area.

Paragraph (j)(4)(ii) would be revised to include a system for communication with certified fuel handlers if the NRC had docketed the certifications required for decommissioning.

Paragraph (p)(1)(i) and (ii) would be revised to allow a certified fuel handler or a licensed senior operator to approve the suspension of security measures if the NRC has docketed the certifications required for decommissioning.

Section 140.11 Amounts of Financial Protection Required for Certain Reactors

This proposed rule would revise § 140.11 by adding new paragraphs (a)(5), (a)(5)(i) and (ii) and by redesignating paragraph (b) as paragraph (c) and adding new paragraph (b) that would provide the requirements for the amounts of financial protection required for reactors in decommissioning.

Section 140.81 Scope and Purpose

This proposed rule would revise § 140.81 by clarifying the scope of who is subject to the requirements in this section and to further clarify that this section no longer applies once a licensee meets the requirements of § 140.11(a)(5)(i) and (ii).

VII. Regulatory Flexibility Certification

As required by the Regulatory Flexibility Act of 1980, 5 U.S.C. 605(b), the Commission certifies that this rule, if adopted, will not have a significant economic impact on a substantial number of small entities. This proposed rule affects only the licensing and operation of nuclear production and utilization facilities. The companies that own these plants do not fall within the scope of the definition of “small entities” set forth in the Regulatory Flexibility Act or the size standards established by the NRC (§ 2.810).

VIII. Regulatory Analysis

The NRC has prepared a draft regulatory analysis for this proposed rule. The analysis examines the costs and benefits of the alternatives considered by the NRC. The NRC requests public comment on the draft regulatory analysis. The draft regulatory analysis is available as indicated in the “ Availability of Documents ” section of this document. Comments on the draft analysis may be submitted to the NRC as indicated under the ADDRESSES section of this document.

IX. Backfitting and Issue Finality

The NRC's backfitting provisions for holders of construction permits and operating licenses appear in § 50.109, “Backfitting.” Issue finality provisions (analogous to the backfitting provisions in § 50.109) for applicants and holders of combined licenses are located in § 52.83, “Finality of referenced NRC approvals; partial initial decision on site suitability,” and § 52.98, “Finality of combined licenses; information requests.” This section describes the backfitting and issue finality implications of the draft guidance documents described in section XVI, “ Availability of Guidance,” in this document and this proposed rule as applied to applicants and holders of pertinent NRC approvals. As stated in section III, “ Discussion,” in this document, the proposed changes to 10 CFR part 72 would not impose requirements on ISFSI-only licensees. Accordingly, the proposed rule would not constitute “backfitting” as that term is defined in § 72.62, “Backfitting.”

A. Current and Future Applicants

Applicants and potential applicants (for licenses, permits, and regulatory approvals such as design certifications) are not, with certain exceptions, the subject of either the 10 CFR part 50 backfitting provisions or any issue finality provisions under 10 CFR part 52. The backfitting and issue finality regulations include language delineating when those provisions begin; in general, they begin after the issuance of a license, permit, or approval ( e.g., § 50.109(a)(1)(iii), § 52.98(a)). Furthermore, neither the 10 CFR part 50 backfitting provisions nor the issue finality provisions under 10 CFR part 52—with certain exclusions discussed below—were intended to apply to every NRC action that substantially changes the expectations of current and future applicants, and applicants have no reasonable expectation that future requirements will not change (“Early Site Permits; Standard Design Certifications; and Combined Licenses for Nuclear Power Plants; Final Rule,” 54 FR 15372, at 15385-15386; April 18, 1989).

The exceptions to this general principle are applicable whenever a combined license applicant references a 10 CFR part 52 license ( e.g., an early site permit) or NRC regulatory approval ( e.g., a design certification rule) with specified issue finality provisions. The issues that are resolved in an early site permit or a design certification and accorded issue finality do not include decommissioning matters that are the subject of this proposed rule and draft guidance, and the proposed rule and draft guidance do not contain design requirements. Therefore, the proposed rule and draft guidance would not affect the issue finality accorded early site permits and design certifications. For the same reasons, the issue finality provision applicable to combined license applicants (§ 52.83) would not apply to a combined license applicant referencing either an early site permit or a design certification with respect to compliance with this rule.

B. Existing Design Certifications

The issues that are resolved in a design certification and accorded issue finality do not include decommissioning matters that are the subject of this proposed rule and draft guidance. Because the decommissioning matters that are the subject of this proposed rule and draft guidance are limited to nuclear power reactor decommissioning, they would not be applied to existing or future design certifications.

C. Existing Licensees

Section IV.A of this document describes a proposed alternative approach to the current requirements for radiological emergency preparedness at a nuclear power reactor. The proposed addition of 10 CFR 50.200 would not constitute backfitting or affect the issue finality of a COL because the proposed change would provide a voluntary alternative set of requirements. Backfitting is defined in § 50.109(a)(1) as, in relevant part, a modification of or addition to the systems, structures, or components (SSCs) or design of a facility, or the procedures or organization required to design, construct, or operate a facility, which results from a new or amended provision in the Commission's regulations. This proposed rule would not require holders of operating licenses and COLs to use the alternative emergency preparedness requirements, so the proposed change would not result in a modification or addition that would be backfitting or affect the issue finality of a COL.

Section IV.A of this document also describes other proposed changes related to emergency preparedness. The NRC would revise § 50.47 to add a paragraph (f) to explain when the planning standards of § 50.47(b) would no longer apply. Removing a requirement would not create a new requirement or amend a requirement because amending means the requirement still exists in some form. Without creating or amending a regulation, this proposed change would not meet the definition of “backfitting” or affect the issue finality of a COL.

The proposed changes to § 50.54(q) would be made to allow a licensee using the emergency preparedness framework of 10 CFR 50.200 to also use § 50.54(q). The proposed changes would not require a licensee to use the § 50.54(q) emergency plan change process or result in a modification of or addition to SSCs or the design of a facility or the procedures or organization required to design, construct, or operate a facility. Therefore, the proposed changes would not meet the definition of “backfitting” or affect the issue finality of a COL.

The proposed change to § 50.54(s)(3) would clarify that FEMA findings and determinations are only necessary when the NRC's planning standards apply to offsite radiological emergency response plans. These changes to the NRC's and FEMA's review of emergency plans would not result in a modification of or addition to SSCs or the design of a facility or the procedures or organization required to design, construct, or operate a facility. Therefore, the proposed changes would not meet the definition of “backfitting” or affect the issue finality of a COL.

The NRC is proposing to amend § 50.54(t) so licensees in decommissioning would be able to conduct emergency preparedness program element reviews at intervals not to exceed 24 months (rather than the current requirement of 12 months) without conducting an assessment against performance indicators. This proposed change would not constitute backfitting or affect the issue finality of a COL because the proposed change would provide a voluntary alternative requirement.

The NRC would add new § 50.54(t)(3) to remove the requirement to conduct periodic emergency preparedness program element reviews once all fuel is in dry cask storage. This proposed change would not meet the definition of “backfitting” or affect the issue finality of a COL because the NRC would be removing a requirement.

The addition of a new paragraph IV.8 to appendix E to 10 CFR part 50 would clarify that the evacuation time estimate requirements of paragraphs IV.4, IV.5, and IV.6 would no longer be applicable to licensees after permanent cessation of operations and permanent removal of fuel from the reactor vessel. This proposed change would not meet the definition of “backfitting” or affect the issue finality of a COL because the NRC would be removing a requirement.

The NRC would add a new paragraph k to part 50, appendix E, section IV.F.2 to state that licensees in decommissioning need to follow the biennial exercise requirements of section IV.F.2. This is the current requirement for these licensees, so this change to the regulations would not change a requirement. Therefore, the proposed change would not meet the definition of “backfitting” or affect the issue finality of a COL.

The NRC is proposing to remove obsolete dates for certain one-time actions that were required as part of the 2011 emergency preparedness final rule and other obsolete dates. These actions are complete, and the requirements are no longer binding on any current licensee. These proposed changes would not meet the definition of “backfitting” or affect the issue finality of a COL because the NRC would be removing a requirement.

The proposed changes to 72.32(a) would clarify the emergency plan requirements for an applicant of a specific license under 10 CFR part 72. As discussed in section IX.A. of this document, applicants such as this one are outside the scope of the 10 CFR part 50 backfitting provisions and issue finality provisions.

The proposed changes to 72.32(c) would clarify that the ISFSI licensee can rely on its 10 CFR part 50 emergency plan to meet the requirements of § 72.32 when the nuclear power reactor is under construction, operating, or in decommissioning. Other provisions of § 72.32 allow an ISFSI licensee with a reactor emergency plan to use that emergency plan to meet the applicable requirements for an ISFSI emergency plan. Therefore, this clarification would not meet the definition of “backfitting” or affect the issue finality of a COL because it would not result in a modification of or addition to SSCs or the design of a facility or the procedures or organization required to design, construct, or operate a facility.

Section IV.B of this document describes proposed changes to physical security requirements. The NRC would permit a certified fuel handler to approve the temporary suspension of security measures once the reactor has shut down and all fuel has been removed from the reactor core. This proposed change would not constitute backfitting or affect the issue finality of a COL because the proposed change would provide a voluntary alternative requirement.

The proposed changes to § 50.54(p) would add definitions of “change” and “decrease in safeguard effectiveness” and require that reactor licensees include with the required § 50.54(p)(2) report a summary of the analysis performed to determine that the change does not decrease safeguards effectiveness of the security plan. The proposed changes would not require a licensee to use the § 50.54(p) security plan change process unless the licensee voluntarily seeks to change its security plan and would not result in a modification of or addition to SSCs or the design of a facility or the procedures or organization required to design, construct, or operate a facility. Therefore, the proposed changes would not meet the definition of “backfitting” or affect the issue finality of a COL.

The NRC would provide an option that, once all spent nuclear fuel has been placed in dry cask storage, licensees could protect a general license ISFSI under § 73.51 instead of § 73.55. This proposed change would not constitute backfitting or affect the issue finality of a COL because the proposed change would provide a voluntary alternative requirement.

Current § 73.55(b)(3) requires that a licensee's physical protection program be designed to prevent significant core damage. The NRC would remove this requirement once the NRC has docketed the licensee's certifications that its reactor has permanently ceased operating and all fuel has been removed from the reactor vessel. This proposed change would not constitute backfitting or affect the issue finality of a COL. The issue finality provision for COLs located in § 52.98 provides, in relevant part, that the Commission may not modify, add, or delete any term or condition of a COL except in accordance with the provisions of § 50.109. Under § 50.109, removing a requirement as proposed with § 73.55(b)(3) is not backfitting because removing a requirement does not create a new requirement and does not amend a requirement because amending means the requirement still exists in some form.

The proposed change to § 73.55(e)(9)(v) would remove the requirement that a licensee must designate the reactor control room as a “vital area” if the NRC has docketed the licensee's certifications that the reactor has permanently ceased operating and all fuel has been removed from the reactor vessel, and the licensee has documented that all vital equipment has been removed from the control room and the control room does not serve as the vital area boundary for other vital areas. This proposed change would not constitute backfitting or affect the issue finality of a COL because the proposed change would be a voluntary alternative requirement. Even if a licensee submitted and the NRC docketed the certifications that the reactor has permanently ceased operating and all fuel has been removed from the reactor vessel, the licensee could still designate the reactor control room as a vital area. If not all of the vital equipment has been removed from the control room or the control room still serves as the vital area boundary for other vital areas, then the licensee would not be required to, and in fact could not, document that all vital equipment has been removed from the control room or the control room does not serve as the vital area boundary for other vital areas, respectively.

The NRC would revise § 73.55(j)(4)(ii) to provide an alternative to the requirement for maintaining continuous communications between the alarm stations and the control room with a requirement for maintaining communications between alarm stations and the CFH or senior on shift licensee representative, once a licensee submits and the NRC dockets the certifications that the reactor has permanently ceased operating and all fuel has been removed from the reactor vessel. This proposed change would relax the requirement for these licensees. However, a licensee in decommissioning could maintain its control room such that its continuous communication system still communicates between the alarm stations and the control room. In this situation, the control room could redirect communications from the alarm stations to the certified fuel handler or the senior on-shift licensee representative as appropriate. Thus, a licensee could continue to comply with the current requirement to maintain continuous communications between the alarm stations and the control room and still satisfy the proposed rule. This makes the relaxation non-mandatory and, as explained in MD 8.4, non-mandatory relaxations of regulations generally do not meet the definition of “backfitting.” This proposed change would provide the voluntary relaxation of a current requirement and, thus, not constitute backfitting or affect the issue finality of a COL.

Section IV.C of this document describes proposed changes to cyber security requirements. The NRC would revise § 73.54 so the cyber security requirements in § 73.54 continue to apply to licensees through Level 1 of decommissioning. Each 10 CFR part 50 licensee has a license condition requiring the licensee to maintain its cyber security plan, and this license condition remains in effect during decommissioning. If the NRC issues an operating license for a nuclear power reactor before this final rule goes into effect, then the NRC can include a license condition similar to those issued to current holders of operating licenses for nuclear power reactors. Thus, this proposed change would not constitute backfitting for 10 CFR part 50 licensees.

A COL holder without the license condition is currently not required to maintain its cyber security plan when it begins decommissioning. The proposed revision to § 73.54 would constitute a change affecting the issue finality accorded these COL holders because extending the requirement to maintain a cyber security plan during decommissioning would modify the terms and conditions of a COL. Under § 52.98, the NRC must apply the provisions of § 50.109 to the proposed change. The proposed change would constitute backfitting under § 50.109. The NRC's backfit analysis justifying this backfitting action is presented in section IX.D of this document. If the NRC issues a COL before this final rule goes into effect, then the NRC can include a license condition similar to those issued to current holders of operating licenses for nuclear power reactors.

Section IV.D of this document describes proposed changes to fitness for duty requirements. The NRC proposes to amend § 26.3(a) so the requirements of 10 CFR part 26 would not apply to COL holders once the NRC has docketed their § 52.110(a) certifications. This proposed change would not affect the issue finality of a COL because the NRC would be removing a requirement.

The proposed changes to § 73.55(b)(9)(ii)(B) would provide minimum requirements for the fitness for duty elements of operating and decommissioning 10 CFR part 50 and 10 CFR part 52 licensees' insider mitigation programs. These licensees are already required to comply with the insider mitigation program requirements of § 73.55(b)(9), so the proposed rule changes would clarify existing requirements and would not constitute backfitting or affect the issue finality of a COL.

The NRC proposes to amend the criminal penalties section of 10 CFR part 26 by including § 26.3 within § 26.825(a) by removing § 26.3 from § 26.825(b). This proposed change would not revise § 26.3 in any way. Enabling the NRC to impose criminal penalties for willful violations of, attempts to violate, or conspiracies to violate § 26.3 would not result in a modification of or addition to SSCs or the design of a facility or the procedures or organization required to design, construct, or operate a facility. Therefore, the proposed change would not meet the definition of “backfitting” or affect the issue finality of a COL.

Section IV.E of this document describes proposed changes to the “certified fuel handler” definition and the elimination of the shift technical advisor. The NRC proposes to amend § 50.2 to provide an alternative definition of “certified fuel handler” to eliminate the need for licensees to submit requests for NRC approval of CFH training programs. This proposed change would not constitute backfitting or affect the issue finality of a COL because the proposed change would provide a voluntary alternative to submitting a request for approval of a fuel handler training program.

The proposed change to § 50.54(m)(2)(i) to state that a shift technical advisor is not required upon the NRC's docketing of the license holder's certifications required under §§ 50.82(a)(1) or 52.110(a) would not result in a modification of or addition to SSCs or the design of a facility or the procedures or organization required to design, construct, or operate a facility. Therefore, the proposed change would not meet the definition of “backfitting” or affect the issue finality of a COL.

Section IV.F of this document describes proposed changes to the NRC's decommissioning funding assurance requirements. The proposed change to § 50.75(f)(1) would modify the reporting frequency for reactor decommissioning funding reports from at least once every 2 years to at least once every 3 years. This reporting requirement would not result in a modification of or addition to SSCs or the design of a facility or the procedures or organization required to design, construct, or operate a facility. Therefore, the proposed change would not meet the definition of “backfitting” or affect the issue finality of a COL.

The NRC would revise § 50.75(h) to require certain notifications be sent directly to the NRC's Document Control Desk and not to the Director, Office of Nuclear Reactor Regulation, or Director, Office of Nuclear Material Safety and Safeguards, as applicable. This reporting requirement would not result in a modification of or addition to SSCs or the design of a facility or the procedures or organization required to design, construct, or operate a facility. Therefore, the proposed change would not meet the definition of “backfitting” or affect the issue finality of a COL.

The NRC proposes to delete § 50.75(f)(2). The language of existing § 50.75(f)(1) fully encompasses the language of paragraph (f)(2), and, therefore, paragraph (f)(2) is unnecessary and potentially confusing. This change would not result in a modification of or addition to SSCs or the design of a facility or the procedures or organization required to design, construct, or operate a facility. Therefore, the proposed change would not meet the definition of “backfitting” or affect the issue finality of a COL.

The NRC is proposing to amend its regulations in § 50.75(f)(1) to clarify that when a licensee identifies a shortfall in the decommissioning funding report required by § 50.75(f)(1), the licensee must identify additional financial assurance to cover the shortfall in the next report. Licensees are already required to provide reasonable assurance of decommissioning funding on an ongoing basis. The proposed change would not change this obligation; the proposed rule would clarify how reasonable assurance of funds will be available for the decommissioning process. This change would not result in a modification of or addition to SSCs or the design of a facility or the procedures or organization required to design, construct, or operate a facility. Therefore, the proposed change would not meet the definition of “backfitting” or affect the issue finality of a COL.

The proposed change to § 50.82(a)(9)(ii)(F) would require licensees to identify the specific sources of funds for “remaining decommissioning costs,” including sources of funds for license termination, spent fuel management, and ISFSI decommissioning. This reporting requirement would not result in a modification of or addition to SSCs or the design of a facility or the procedures or organization required to design, construct, or operate a facility. Therefore, the proposed change would not meet the definition of “backfitting” or affect the issue finality of a COL.

The NRC would revise § 50.82(a)(8)(v) to allow licensees to combine the reports that are required by §§ 50.82(a)(8)(v), 50.82(a)(8)(vii) and 72.30(c). This proposed change would not constitute backfitting or affect the issue finality of a COL because the proposed change would provide a voluntary alternative requirement.

The NRC proposes to revise § 52.110 to make the same changes proposed in § 50.82. For the reasons previously discussed, these proposed changes would not affect the issue finality of a COL. The NRC also proposes to add to § 52.110 paragraphs (h)(5) through (h)(7) with site-specific decommissioning cost estimate reporting requirements that are identical to the requirements in § 50.82(a)(8)(v) through (vii). These reporting requirements would not result in a modification of or addition to SSCs or the design of a facility or the procedures or organization required to design, construct, or operate a facility. Therefore, the proposed changes would not meet the definition of “backfitting” and, under § 52.98, would not affect the issue finality of a COL.

The NRC proposes to revise § 72.30 so that the submittals subsequent to the initial decommissioning funding plan would no longer require NRC approval. This proposed change would not meet the definition of “backfitting” or affect the issue finality of a COL because the NRC would be removing a requirement.

The proposed changes to § 72.30(b) would clarify the requirements for an applicant for a specific licensee and a holder of a general license to submit decommissioning funding plans for NRC review and approval. The current requirement requires applicants and holders of licenses under 10 CFR part 72 to submit decommissioning funding plans for NRC review and approval. These changes would not change any substantive requirement and would not result in a modification of or addition to SSCs or the design of a facility or the procedures or organization required to design, construct, or operate a facility. Accordingly, these proposed changes would not constitute backfitting or affect the issue finality of a COL.

Section IV.G of this document describes proposed changes to the NRC's onsite and offsite financial protection requirements and indemnity agreements. These changes would include revisions to the following regulations: §§ 140.11(a)(5) and 50.54(w)(5), to allow nuclear power reactor licensees in decommissioning to reduce the offsite liability and onsite property insurance amounts, respectively, that they are required to maintain; § 140.81, to include plants in decommissioning within the scope of § 140.81, thereby clarifying the applicability of the requirements for an Extraordinary Nuclear Occurrence ENO to reactors in decommissioning; and § 50.54(w), to require a prompt notification to the Commission of any material change in proof of onsite property insurance filed with the Commission under 10 CFR part 50.

Changes to 10 CFR part 140 are not subject to the 10 CFR part 50 backfitting provisions and the issue finality provisions in 10 CFR part 52 because the Price-Anderson Act requires licensees to have offsite financial protection. Even if they were subject to the 10 CFR part 50 backfitting provisions and the issue finality provisions in 10 CFR part 52, the proposed changes would not result in a modification of or addition to SSCs or the design of a facility or the procedures or organization required to design, construct, or operate a facility. Therefore, the proposed changes would not meet the definition of “backfitting” and would not affect the issue finality of a COL.

Similarly, the onsite insurance requirements in § 50.54(w) do not fall within the purview of the 10 CFR part 50 backfitting provisions or the issue finality provisions in 10 CFR part 52. In the backfitting discussion for the 1987 final rule, “Changes in Property Insurance Requirements for NRC Licensed Nuclear Power Plants” (52 FR 28963, 28972; August 5, 1987), the Commission stated that requiring an increase in property damage insurance does not meet the definition of “backfitting.” The Commission took similar positions on backfitting in subsequent rulemakings to amend § 50.54(w) ( e.g., 54 FR 11163, March 17, 1989; 55 FR 12163, April 2, 1990).

Section IV.H of this document describes proposed changes to the requirements concerning consideration of environmental effects of decommissioning activities. The NRC proposes to change § 50.82(a)(4)(i) and § 52.110(d)(1) to require that licensees provide the basis for determining whether the environmental impacts of decommissioning activities are bounded by previous environmental reviews and include a description in the PSDAR of any activities that will not be bounded. These reporting requirements would not result in a modification of or addition to SSCs or the design of a facility or the procedures or organization required to design, construct, or operate a facility. Therefore, the proposed changes would not meet the definition of “backfitting” and would not affect the issue finality of a COL.

The NRC also proposes to change § 50.82(a)(4)(i) and § 52.110(d)(1) to allow licensees to use appropriate federally issued environmental review documents prepared in compliance with NEPA, ESA, NHPA, or other environmental statutes instead of only environmental impact statements. These reporting requirements would not result in a modification of or addition to SSCs or the design of a facility or the procedures or organization required to design, construct, or operate a facility. Therefore, the proposed changes would not meet the definition of “backfitting” and would not affect the issue finality of a COL.

The NRC would change § 50.82(a)(6)(ii) and § 52.110(f)(2) to clarify that the previous review of any potentially significant environmental impact must be bounded by appropriate federally issued environmental review documents prepared in compliance with NEPA, ESA, NHPA, or other environmental statutes. These reporting requirements would not result in a modification of or addition to SSCs or the design of a facility or the procedures or organization required to design, construct, or operate a facility. Therefore, the proposed changes would not meet the definition of “backfitting” and would not affect the issue finality of a COL.

The NRC proposes to revise 10 CFR part 51 to reflect the changes made in the 1996 Final Rule that nuclear power reactor licensees are not required to submit license amendment requests for authorization to perform decommissioning activities. These changes would not change any substantive requirement and would not result in a modification of or addition to SSCs or the design of a facility or the procedures or organization required to design, construct, or operate a facility. Accordingly, these proposed changes would not constitute backfitting or affect the issue finality of a COL.

Section IV.I of this document describes proposed changes to record retention requirements. These changes would eliminate certain recordkeeping requirements and the requirement to keep certain duplicate records. These recordkeeping changes would not result in a modification of or addition to SSCs or the design of a facility or the procedures or organization required to design, construct, or operate a facility. Therefore, the proposed changes would not meet the definition of “backfitting” and would not affect the issue finality of a COL. The proposed changes also would not meet the definition of “backfitting” or affect the issue finality of a COL because the NRC would be removing these requirements.

Section IV.J of this document describes proposed changes to low-level radioactive waste transportation requirements. The NRC would revise Paragraph III.E of appendix G to 10 CFR part 20 to increase from 20 days to 45 days the window of time for notification of receipt of shipments of low-level waste before a shipper would be required to investigate, trace, and report to the NRC any shipments of low-level waste for which the shipper has not received a notification of receipt. This proposed change would relax the requirement. However, a shipper could still investigate, trace, and report shipments of low-level waste if the shipper has not received notification of receipt within 20 days. Thus, a shipper could continue to comply with the current 20-day requirement and still satisfy the proposed rule. This makes the relaxation non-mandatory and, as explained in MD 8.4, non-mandatory relaxations of regulations generally do not meet the definition of “backfitting.” This proposed change would provide the voluntary relaxation of a current requirement and, thus, not constitute backfitting or affect the issue finality of a COL.

Section IV.K of this document describes proposed changes to spent fuel management requirements. The NRC would revise §§ 50.54(bb) and 72.218 to clarify the contents of an irradiated fuel management plan, which licensees are already required to submit to the NRC for approval. This clarification of a reporting requirement would not result in a modification of or addition to SSCs or the design of a facility or the procedures or organization required to design, construct, or operate a facility. Therefore, the proposed changes would not meet the definition of “backfitting” and would not affect the issue finality of a COL.

The NRC proposes to change § 72.218 to remove spent fuel management provisions that the NRC would move to § 50.54(bb) and clarify provisions concerning termination of part 72 general licenses. The proposed changes would not result in a modification of or addition to SSCs or the design of a facility or the procedures or organization required to design, construct, or operate a facility. Therefore, the proposed changes would not meet the definition of “backfitting” and would not affect the issue finality of a COL.

Section IV.L of this document describes proposed changes to the NRC's backfitting provisions in 10 CFR part 50 and part 72. The NRC proposes to change § 50.109 to clarify application of the 10 CFR part 50 backfitting provisions to NRC actions constituting backfitting or affecting the issue finality of nuclear power reactor licensees in decommissioning. The NRC also would revise § 50.109 to require a documented evaluation to include a consideration of the costs of imposing the backfit if the basis for backfitting is bringing a facility into compliance with a license or the rules or orders of the Commission, or into conformance with the licensee's written commitments. The proposed change to § 72.62 would clarify that the backfit regulations in part 72 apply during the decommissioning of an independent spent fuel storage installation or a monitored retrievable storage facility. The proposed changes to backfitting provisions would be changes to requirements imposed on the NRC, not on a licensee, so the proposed changes would be outside the scope of backfitting and issue finality.

Section IV.M of this document describes proposed changes to the NRC's regulations related to foreign ownership, control, or domination of a production or utilization facility. The NRC would revise § 50.38 to clarify when a facility licensed under 10 CFR part 50 or part 52 is not considered a production or utilization facility and, therefore, the foreign ownership, control, or domination prohibition no longer applies. The proposed changes would not result in a modification of or addition to SSCs or the design of a facility or the procedures or organization required to design, construct, or operate a facility. Therefore, the proposed changes would not meet the definition of “backfitting” and would not affect the issue finality of a COL.

The NRC also would amend § 50.82(b) to add the criteria for when a non-power production or utilization facility or fuel reprocessing plant is no longer a production or utilization facility. The only part 50 licensees considered within the scope of the part 50 backfitting provision are nuclear power reactor licensees. Further, the proposed changes would not result in a modification of or addition to SSCs or the design of a facility or the procedures or organization required to design, construct, or operate a facility. Therefore, the proposed changes would not meet the definition of “backfitting.”

The NRC would revise § 50.82(a) and (b) and § 52.110(b) to affirm the continuation of the NRC's statutory authority over the existing 10 CFR part 50 or 10 CFR part 52 license, and to state which regulations would still apply to the licensee, after the performance of decommissioning activities that lead to the licensed facility no longer meeting the definition of a utilization or a production facility. The proposed changes would not result in a modification of or addition to SSCs or the design of a facility or the procedures or organization required to design, construct, or operate a facility. Therefore, the proposed changes would not meet the definition of “backfitting” and would not affect the issue finality of a COL.

In light of the proposed amendments to §§ 50.38, 50.82, and 52.110, the NRC would amend §§ 50.1, 50.51, 52.0, and 52.109 to clarify that the regulations in 10 CFR part 50, and the similar regulations in 10 CFR part 52, provide not only for the licensing of utilization and production facilities, but also for their decommissioning and the termination of their associated licenses. The proposed changes would not result in a modification of or addition to SSCs or the design of a facility or the procedures or organization required to design, construct, or operate a facility. Therefore, the proposed changes would not meet the definition of “backfitting” and would not affect the issue finality of a COL.

The NRC is proposing to add a definition for “non-power production or utilization facility” to § 50.2 that captures all non-power facilities licensed under § 50.22 and § 50.21(a) or (c), except fuel reprocessing facilities. The only part 50 licensees considered within the scope of the part 50 backfitting provision are nuclear power reactor licensees. Further, the proposed definition would not result in a modification of or addition to SSCs or the design of a facility or the procedures or organization required to design, construct, or operate a facility. Therefore, the proposed changes would not meet the definition of “backfitting.”

Section IV.N of this document describes proposed changes to license termination plan requirements. The NRC would revise § 50.82(a)(9) and § 52.110(i) to clarify that only nuclear power reactor licensees that have loaded fuel into their reactors must submit license termination plans. The proposed change would not change this requirement; the proposed rule would only clarify that nuclear power reactor licensees that have not loaded fuel into their reactors would not need to submit license termination plans. This change would not result in a modification of or addition to SSCs or the design of a facility or the procedures or organization required to design, construct, or operate a facility. Therefore, the proposed change would not meet the definition of “backfitting” or affect the issue finality of a COL.

Section IV.O of this document describes the proposed removal of license conditions and withdrawal of orders. These changes would not change any substantive requirement because the license conditions and orders are substantively redundant with NRC regulations issued after the license conditions and orders were issued. Because the NRC would not change a requirement, the proposed changes would not meet the definition of “backfitting” or affect the issue finality of a COL.

D. Backfit Analysis

1. Introduction and Background

As part of this proposed rule, the NRC is proposing a modification to the cyber security requirements in § 73.54. This proposed rule would ensure that these requirements continue to apply to nuclear power reactor licensees that have submitted their § 50.82(a)(1) or § 52.110(a) certifications until such time that all spent fuel in the SFP has sufficiently decayed ( i.e., at least 10 months for BWRs and 16 months for PWRs after the date of permanent cessation of operations, or an NRC-approved alternative spent fuel decay period).

This amendment would likely constitute a change affecting issue finality for 10 CFR part 52 COL holders, as defined in § 52.98. These licensees are not currently required to maintain their cyber security programs past the date that they are no longer authorized to operate the reactor. If the proposal to require these licensees to maintain their cyber security program into the decommissioning phase would extend the duration that a COL holder would be required to maintain a cyber security program, then that extension would constitute a new or changed requirement for that licensee and, thus, affect that COL's issue finality.

2. Detailed Description of the Proposed Change Affecting Issue Finality

The NRC sets forth the current cyber security requirements for nuclear power reactors in § 73.54. The NRC established these requirements as part of the 2009 Power Reactor Security Requirements final rule. The preamble to § 73.54 states, in part, that by November 23, 2009, each nuclear power reactor licensee “currently licensed to operate” must submit to the NRC a cyber security plan (CSP) for review and approval. The preamble further states that the requirements in § 73.54 are applicable to current “applicants for an operating license or combined license” and mandates such applicants to amend their applications to include a CSP. In addition, every 10 CFR part 50 license for a nuclear power reactor that was operating in 2009 contains a license condition to have and maintain a Commission-approved CSP. These license conditions were issued when the NRC approved each licensee's CSP that was submitted to the NRC as required by the Power Reactor Security Requirements final rule. The Tennessee Valley Authority's 10 CFR part 50 operating license for Watts Bar Nuclear Plant, Unit 2, issued in 2015, also contains a license condition to have and maintain a CSP.

As an initial step in the decommissioning process, a nuclear power reactor licensee must submit written certifications that it has decided to permanently cease operations and has permanently removed all fuel from its reactor vessel, in accordance with § 50.82(a)(1)(i) and (ii) for nuclear power reactor licensees under 10 CFR part 50, or § 52.110(a)(1) and (2) for 10 CFR part 52 combined license holders. As stated in § 50.82(a)(2) and § 52.110(b), upon the NRC's docketing of these certifications, the license no longer authorizes operation of the reactor or the placement or retention of fuel in the reactor vessel. In a December 5, 2016 memorandum to the Commission, the NRC staff explained that § 73.54 no longer applies to nuclear power reactor licensees once they have submitted, and the NRC has docketed, these certifications.

As discussed in the “ Technical Basis for Graded Approach ” section of this document, the NRC has concluded that after 10 months for BWRs and 16 months for PWRs, the spent fuel in the SFP will have decayed and cooled sufficiently such that the fuel cannot heat up to clad ignition temperature within 10 hours under adiabatic conditions. The NRC has determined that until the fuel has decayed and cooled sufficiently, nuclear power reactor licensees must maintain reasonable assurance that their critical digital assets remain protected against cyber attacks. As such, this proposed rule would modify the cyber security requirements in § 73.54 to ensure that they continue to apply to licensees of decommissioning nuclear power reactors until the spent fuel has decayed and cooled sufficiently (either through the application of a 10 month (BWR) or 16 month (PWR) decay period or an NRC-approved site-specific decay period). This proposed rule would also remove the CSP license condition from the 10 CFR part 50 licenses at the applicable 10 or 16 month interval.

This proposed rule would not constitute backfitting for currently operating or recently shutdown 10 CFR part 50 reactor licensees. Their CSP license condition remains in effect until the termination of the license or the NRC removes the condition from the license ( e.g., if the licensee submits a license amendment request and the NRC approves it). The NRC has determined that the requirements of the CSP license conditions are not necessary after the spent fuel in the SFP has sufficiently cooled. The proposed rule would codify, during Level 1 of decommissioning, the already-imposed requirements of the CSP license conditions. These requirements would continue to provide adequate protection of the public health and safety and common defense and security and continue to support the effective operation of licensees' security and emergency preparedness programs during the time when a draindown scenario can credibly lead to a zirconium fire. (See sections 3 and 4 of this backfit analysis for additional cost/benefit discussion.) Therefore, this proposed rule would not impact these licensees' overall requirement to maintain a cyber security program, but would instead enable the automatic removal of cyber security requirements once fuel in the SFP has sufficiently cooled. Thus, the decommissioning rulemaking would not impose a new or changed requirement as the licensees are already implementing the requirement as part of their cyber security program license conditions.

Conversely, this rulemaking would constitute a change affecting the issue finality for 10 CFR part 52 COL holders. Each currently approved COL includes a license condition to provide the NRC with the licensee's Operational Program Implementation Schedule. The operational programs (which include development and implementation of a security program, including a cyber security program) are requirements in the regulations and not separately identified as license conditions. As a result, a COL does not require the licensee to maintain the cyber security program throughout the duration of its license. COL holders are currently required to maintain a program only as long as § 73.54 is applicable to them. Because § 73.54 no longer applies to the licensee once it is not authorized to operate a nuclear power reactor, and a nuclear power reactor licensee is not authorized to operate a nuclear power reactor during decommissioning, COL holders are not required to maintain their CSP during decommissioning. This proposed rule, which would require licensees to maintain their cyber security program for 10 months (BWR) or 16 months (PWR) beyond the date of permanent cessation of operations (or for an NRC-approved alternative spent fuel decay period) could extend the duration over which a COL holder would be required to maintain a cyber security program. That extension would constitute a new or changed requirement for that licensee.

Under § 52.98, the Commission cannot modify any term or condition of an issued combined license except in accordance with the provisions of § 52.103 or § 50.109, as applicable. This proposed rule's amendment of the cyber security requirements would constitute a change affecting the issue finality of the COLs issued at the time of the final rule's effective date. The provisions of § 52.103 do not apply to this proposed rule, so the NRC must show that the amendment would meet the requirements of § 50.109 to justify proceeding with this amendment. Because none of the exceptions to the requirement to prepare a backfit analysis in § 50.109(a)(4) applies to this rulemaking, § 50.109(a)(3) requires the NRC to prepare a backfit analysis that demonstrates that the proposed amendment would result in a substantial increase in the overall protection of the public health and safety or the common defense and security, and that the direct and indirect costs of implementation are justified in view of this increased protection.

3. Benefits: Substantial Increase in Public Health and Safety and Common Defense and Security

The NRC identified qualitative (non-quantifiable) benefits that would occur if the proposed change affecting issue finality were implemented.

The NRC identified two qualitative benefits to the common defense and security and public health and safety that would be realized if the proposed rule is implemented. Specifically, the NRC finds that extending the duration over which the licensee must maintain cyber security requirements would:

• Constitute a substantial increase in protection to common defense and security by ensuring that a compromise of digital systems cannot adversely impact the effective operation of licensees' physical security programs; and
• Constitute a substantial increase in public health and safety by ensuring that a compromise of digital systems cannot adversely impact the effective operation of emergency preparedness systems in the event of a zirconium fire scenario.

Effective Operation of Physical Security Program

The NRC has previously determined that attacks on the SFP are credible and have the potential to lead to an unacceptable impact to common defense and security. Specifically, a physical attack by either an external force or malicious insiders could directly lead to a draindown scenario and subsequent zirconium fire.

As established in § 73.54, cyber security is an essential element of a licensee's physical security program that enables the licensee to effectively protect its site against the design basis threat of radiological sabotage defined in § 73.1, in accordance with § 73.55(b). Specifically, a physical attack that is augmented with a coincident cyber attack would, in many cases, have a higher chance of success over a purely physical attack. Thus, although there is no cyber attack that can directly lead to a draindown scenario, a cyber attack can be combined with a physical attack on the SFP to improve the physical attack's likelihood of success.

Given a facility without adequate cyber security controls in place, several mechanisms exist that could improve the effectiveness of a physical attack on the SFP. For example, a cyber attack could aid a physical assault on the SFP by an external attacker by:

• Disabling perimeter detection to delay or prevent onsite response to the physical assault prior to the attacker gaining entry to the SFP
• disrupting onsite and offsite security-related communication to reduce the effectiveness of the licensee's response to the physical assault
• disabling access control doors and gates to enable the attacker expedited physical access to the SFP

In addition, inadequate cyber security controls on facilities' access control systems could enable an attacker to inject information into a licensee's access control system in a manner that would allow unauthorized individuals to obtain unescorted access into the protected or vital areas of the facility. This could allow one or more attackers direct access to the SFP, which could then be exploited to sabotage the SFP in a manner that would result in a draindown scenario.

This factor, combined with the severity of the consequences of a draindown scenario and subsequent zirconium fire that could result from a successful physical attack, demonstrates that maintaining cyber security requirements during the period when a draindown scenario could reasonably result in a zirconium fire ( i.e., prior to the fuel in the SFP sufficiently cooling) represents a substantial increase in security.

Effective Operation of Emergency Preparedness Systems

As discussed in the “ Technical Basis for the Graded Approach” and “ Emergency Preparedness” sections of this document, although the spectrum of credible accidents and operational events requiring an emergency response is reduced at a decommissioning nuclear power reactor as compared to that for an operating nuclear power reactor, reliable emergency preparedness functions are still required to ensure public health and safety in the event of a zirconium fire scenario.

As established in § 73.54, cyber security is an essential element of a licensee's physical security program that, in part, ensures that a compromise of digital systems cannot adversely impact emergency preparedness functions. For example, in the event of a zirconium fire scenario, the licensee's cyber security program prevents a cyber attack from adversely impacting the ability to:

• Notify state, local, and Federal personnel of the emergency
• Request and communicate with offsite support
• Assess and classify the emergency conditions
• Disseminate information to the public during an emergency
• Conduct a radiological accident assessment

The NRC has determined that this factor demonstrates that maintaining cyber security requirements to ensure that a compromise of digital systems cannot adversely impact the operation of emergency preparedness functions until the time in which a SFP draindown would likely be mitigated prior to a zirconium fire scenario ( i.e., once the fuel in the SFP has sufficiently cooled) represents a substantial increase in public health and safety.

4. Costs

The NRC identified quantitative costs ( i.e., costs that are amenable to quantitative evaluation) that would be incurred if the proposed change affecting issue finality were implemented.

Based on a review of feedback received during recent inspections of the full implementation of licensees' cyber security programs, the NRC estimates that the cost to implement a cyber security program for a decommissioning nuclear power reactor is approximately $300,000 per site per year. As previously stated, this proposed change affecting issue finality would extend the duration that a licensee must maintain its cyber security program for 10 (BWR) or 16 (PWR) months. Thus, the cost associated with this extension is approximately $250,000 (BWR) or $400,000 (PWR).

COLs have been issued at a total of 3 sites that utilize BWR units, and 4 sites that utilize PWR units. Assuming that all units are constructed and the per-site costs from the previous paragraph, the total cost associated with this proposed change affecting issue finality if all reactors entered decommissioning today would be approximately $2.35 million. If it is assumed that all sites with units licensed under 10 CFR part 52 decommission their reactors 40 years after the effective date of the final rule, with a discount rate of 7 percent, then the total, combined cost for all affected licensees associated with this proposed change affecting issue finality would be approximately $157,000. Due to the potential that some of these facilities may not be constructed or that some licensees may have voluntarily chosen to maintain their cyber security programs during this timeframe, this estimate is expected to be an upper bound.

5. Determination of Substantial Benefits Justifying Costs of the Proposed Change Affecting Issue Finality

The NRC finds that the proposed change affecting issue finality would provide a substantial increase in protection to public health and safety and common defense and security for current 10 CFR part 52 COL holders by ensuring that a compromise of digital systems cannot adversely impact the effective operation of licensees' security and emergency preparedness programs during the time when a draindown scenario can credibly lead to a zirconium fire. The NRC finds that this substantial increase would justify the $157,000 in costs that would accrue to the licensees.

6. Conclusion

On the basis of this analysis, the NRC determines that the change affecting issue finality resulting from the cyber security portion of this proposed rule would be justified under § 50.109(a)(3).

7. Evaluation of Factors in § 50.109(c)(1) Through (9)

In performing this analysis, the NRC considered the nine factors in § 50.109(c), as follows:

Statement of the Specific Objectives That the Backfit Is Designed To Achieve

The two objectives for the cyber security portion of the “Regulatory Improvements for Production and Utilization Facilities Transitioning to Decommissioning” rulemaking are:

• To ensure the effectiveness of the physical protection program during the period over which a SFP draindown could realistically result in a zirconium fire scenario; and
• To ensure the effectiveness of emergency preparedness functions during the period over which a SFP draindown may not be mitigatable prior to the draindown resulting in a zirconium fire

Note that the change affecting issue finality is only applicable to nuclear power reactors licensed under 10 CFR part 52 as of the effective date of the final rule.

General Description of the Activity That Will Be Required by the Licensee or Applicant in Order To Complete the Backfit

The NRC is proposing a modification to the cyber security requirements in § 73.54 to ensure that these requirements continue to apply to licensees of decommissioning nuclear power reactors until such time that all spent fuel in the SFP has sufficiently decayed ( i.e., 10 months for BWRs and 16 months for PWRs since the date of permanent cessation of operations, or an NRC-approved alternative spent fuel decay period). The change affecting issue finality is only applicable to nuclear power reactors currently licensed under 10 CFR part 52 as of the effective date of the final rule.

Potential Change in the Risk to the Public From the Accidental Off-Site Release of Radioactive Material

The rulemaking is intended to reduce risk of offsite releases as a result of breaches in security at nuclear power plants, and to ensure the functionality of emergency preparedness functions in the case of a zirconium fire scenario. However, the reduction in risk to the public from offsite releases of radioactive materials has not been fully quantified because there is insufficient information and modeling to support such quantification.

Potential Impact on Radiological Exposure of Facility Employees

The rulemaking would provide added assurance that nuclear industry workers are not subjected to unnecessary radiological exposures as the result of a breach in security that causes a zirconium fire leading to a release of radiation that security personnel are exposed to as the result of their response activities. Further, the rulemaking would ensure that emergency preparedness functions, including evacuation procedures, are not adversely impacted by a cyber attack during the period when a draindown scenario could reasonably result in a zirconium fire, thus ensuring that nuclear industry workers are not subjected to unnecessary radiological exposures in the case of a zirconium fire scenario.

Installation and Continuing Costs Associated With the Backfit, Including the Cost of Facility Downtime or the Cost of Construction Delay

The backfit analysis to support the change affecting issue finality resulting from this proposed rule includes the NRC's estimate of the total costs for maintaining a licensee's cyber security program until the fuel in the SFP has sufficiently cooled to adequately ensure that a SFP draindown does not result in a zirconium fire scenario. The estimated one-time industry net cost associated with the change affecting issue finality would be approximately $157,000.

The Potential Safety Impact of Changes in Plant or Operational Complexity, Including the Relationship to Final and Existing Regulatory Requirements

The cyber security portion of this proposed rule would not impose any requirements beyond those in place while the nuclear power reactor is operational. As such, this rule is not expected to have an effect on facility complexity.

The Estimated Resource Burden on the NRC Associated With the Backfit and the Availability of Such Resources

The rulemaking may result in a minor increase in the expenditure of agency resources, due to the potential for cyber security inspections to be conducted after the licensee has ceased operations and before fuel in the SFP has sufficiently cooled.

The Potential Impact of Differences in Facility Type, Design or Age on the Relevancy and Practicality of the Backfit

The specific cost of this rulemaking to a facility does vary, depending on whether the facility utilizes BWR or PWR reactors. This is due to time required for fuel in the SFP to sufficiently cool for each type of reactor. Further, since the change affecting issue finality is only applicable to reactors licensed under 10 CFR part 52, the specific cost also depends on the percentage of reactors licensed under 10 CFR part 52 at the licensee's facility.

Whether the Backfit is Interim or Final and, if Interim, the Justification for Imposing the Backfit on an Interim Basis

The change affecting issue finality would be final.

E. Draft Regulatory Guidance

As described in Section XVI, “Availability of Guidance,” in this document, the NRC is issuing four draft regulatory guides (DGs) that, if finalized, would provide guidance on the methods acceptable to the NRC for complying with aspects of this proposed rule. The DGs would apply to all current holders of operating licenses under 10 CFR part 50 and COLs under 10 CFR part 52. Issuance of the DGs in final form would not constitute backfitting under § 50.109 and would not otherwise constitute a change affecting issue finality under 10 CFR part 52. As discussed in the “Implementation” section of each DG, the NRC has no current intention to impose the DGs on current holders of an operating license or COL.

For the same reasons provided under “Current and Future Applicants” that explain why the proposed rule does not constitute backfitting or a change affecting issue finality for applicants, applying the DGs to applications for operating licenses or COLs would not constitute backfitting as defined in § 50.109 and would not otherwise constitute a change affecting issue finality under 10 CFR part 52.

X. Cumulative Effects of Regulation

The NRC is following its Cumulative Effects of Regulation (CER) process by engaging extensively with external stakeholders throughout this rulemaking and related regulatory activities. Public involvement has included: (1) The publication of an ANPR for public comment (80 FR 72358) on November 19, 2015, to inform the NRC's efforts in drafting a proposed rule regulatory basis to address issues associated with nuclear power reactor decommissioning; (2) holding a public meeting on December 9, 2015, to afford external stakeholders an opportunity to ask the NRC staff clarifying questions regarding the ANPR; (3) the publication of the draft regulatory basis for public comment (82 FR 13778) on March 15, 2017; (4) the publication of a preliminary draft of the regulatory analysis for public comment (82 FR 21481) on May 9, 2017; and (5) holding a public meeting on May 8-10, 2017, to facilitate public comments on the development of the final regulatory basis and regulatory analysis.

Another opportunity for comment is being provided to the public with this proposed rule. The NRC will be issuing the draft implementing guidance with this proposed rule to support more informed external stakeholder feedback. Further, the NRC will continue to hold public meetings throughout the rulemaking process. Section XVI, “Availability of Guidance,” of this document describes how the public can access the draft implementing guidance for which the NRC seeks external stakeholder feedback.

Finally, the NRC is requesting CER feedback on the following questions:

1. In light of any current or projected CER challenges, does the proposed rule's effective date provide sufficient time to implement the new proposed requirements, including changes to programs, procedures, and facilities?

2. If CER challenges currently exist or are expected, what should be done to address them? For example, if more time is required for implementation of the new requirements, what period of time is sufficient?

3. Do other (NRC or other agency) regulatory actions ( e.g., orders, generic communications, license amendment requests, inspection findings of a generic nature) influence the implementation of the proposed rule's requirements?

4. Are there unintended consequences? Does the proposed rule create conditions that would be contrary to the proposed rule's purpose and objectives? If so, what are the unintended consequences, and how should they be addressed?

5. Please comment on the NRC's cost and benefit estimates in the draft regulatory analysis that supports the proposed rule. The draft regulatory analysis is available as indicated in the “ Availability of Documents” section of this document.

XI. Plain Writing

The Plain Writing Act of 2010 (Pub. L. 111-274) requires Federal agencies to write documents in a clear, concise, and well-organized manner. The NRC has written this document to be consistent with the Plain Writing Act as well as the Presidential Memorandum, “Plain Language in Government Writing,” published June 10, 1998 (63 FR 31885). The NRC requests comment on this document with respect to the clarity and effectiveness of the language used.

XII. National Environmental Policy Act

This proposed rule includes some actions that are of the types described in § 51.22(c). The NRC has previously determined that these types of actions do not have a significant impact on the environment and has categorically excluded them from the requirement to prepare an environmental analysis. Specifically, the NRC has determined that some amendments in this proposed rule are the types of actions described in the § 51.22(c) exclusions noted in Table 4. Accordingly, the NRC has not developed an environmental impact statement or an environmental assessment for these portions of the proposed rule.

Draft Finding of No Significant Impacts

The NRC has prepared a draft environmental assessment (EA) for the portions of this proposed rule not categorically excluded under § 51.22. The draft EA is available in ADAMS at Accession No. ML22019A140. The NRC prepared the draft EA to determine environmental impacts of the proposed action: A rulemaking to update the NRC's regulations related to production and utilization facilities transitioning to decommissioning. Based on the draft EA, the NRC concludes that this proposed rule would not have significant environmental impacts because the changes would be administrative or procedural in nature and would have no nexus to the physical environment or would have no significant impact on the environment. Therefore, this proposed rule does not warrant preparation of an environmental impact statement. Accordingly, the NRC has determined that a finding of no significant impact (FONSI) is appropriate.

XIII. Paperwork Reduction Act

This proposed rule contains new or amended collections of information subject to the Paperwork Reduction Act of 1995 (44 U.S.C. 3501-21). This proposed rule has been submitted to the Office of Management and Budget for review and approval of the information collections.

Type of submission, new or revision: Revision.

The title of the information collection: Regulatory Improvements for Production and Utilization Facilities Transitioning to Decommissioning, Proposed Rule.

The form number if applicable: Not applicable.

How often the collection is required or requested: Annually and on occasion.

Who will be required or asked to respond: Production and utilization facility licensees.

An estimate of the number of annual responses: 102 (1 response for 10 CFR part 20, 0 responses for 10 CFR part 26, 97 responses for 10 CFR part 50, 0 responses for 10 CFR part 52, 1 response for 10 CFR part 72, and 3 responses for 10 CFR part 73).

The estimated number of annual respondents: 62 (1 respondent for 10 CFR part 20, 0 respondents for 10 CFR part 26, 62 respondents for 10 CFR part 50, 0 respondents for 10 CFR part 52, 20 respondents for 10 CFR part 72, and 1 respondent for 10 CFR part 73).

An estimate of the total number of hours needed annually to comply with the information collection requirement or request: −3,658 (−77.5 hours for 10 CFR part 20, 0 hours for 10 CFR part 26, −3,114.5 hours for 10 CFR part 50, 0 hours for 10 CFR part 52, −436 hours for 10 CFR part 72, and −30 hours for 10 CFR part 73).

Abstract: The proposed rule would result in changes in recordkeeping and reporting burden relative to existing rules by creating a regulatory framework for production and utilization facility licensees transitioning to decommissioning and amending existing regulations that relate to the decommissioning of production and utilization facilities. Decommissioning nuclear power reactor licensees and the NRC have expended substantial resources processing licensing actions for nuclear power reactors during their transition period to decommissioning status. Licensees that are currently transitioning to decommissioning have been requesting NRC review and approval of licensing actions, informed by the low risk of an offsite radiological release posed by a decommissioning reactor. Specifically, the licensees are seeking NRC approval of exemptions and license amendments to revise requirements to reflect the reduced operations and risks posed by a permanently shutdown and defueled reactor. The proposed rule would, on balance, reduce the paperwork burden imposed on production and utilization facility licensees transitioning to decommissioning by establishing a graded approach to the requirements imposed on these facilities. A graded approach would adjust the level of analysis, documentation, and actions necessary to comply with safety requirements and criteria commensurate with several factors, including magnitude of any credible hazard involved, and the balance between radiological and non-radiological hazards as applicable to the level within the decommissioning process. The NRC expects that these proposed changes would enhance the efficiency of the decommissioning process and reduce the overall burden on licensees.

The NRC is seeking public comment on the potential impact of the information collections contained in this proposed rule and on the following issues:

1. Is the proposed information collection necessary for the proper performance of the functions of the NRC, including whether the information will have practical utility?

2. Is the estimate of the burden of the proposed information collection accurate?

3. Is there a way to enhance the quality, utility, and clarity of the information to be collected?

4. How can the burden of the proposed information collection on respondents be minimized, including the use of automated collection techniques or other forms of information technology?

A copy of the OMB clearance package and proposed rule is available in ADAMS under Accession No. ML18039A192 or can be obtained free of charge by contacting the NRC's Public Document Room reference staff at 1-800-397-4209, at 301-415-4737, or by email to PDR.Resource@nrc.gov. You may obtain information and comment submissions related to the OMB clearance package by searching on https://www.regulations.gov under Docket ID NRC-2015-0070.

You may submit comments on any aspect of these proposed information collections, including suggestions for reducing the burden and on the above issues, by the following methods:

• Federal rulemaking website: Go to https://www.regulations.gov and search for Docket ID NRC-2015-0070.

• Mail comments to: FOIA, Library, and Information Collections Branch, Office of the Chief Information Officer, Mail Stop: T6-A10M, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001 or to the OMB reviewer at: OMB Office of Information and Regulatory Affairs (3150-0014, -0146, -0011, -0151, -0132, -0002), Attn: Desk Officer for the Nuclear Regulatory Commission, 725 17th Street NW, Washington, DC 20503; email: oira_submission@omb.eop.gov.

Submit comments by April 4, 2022. Comments received after this date will be considered if it is practical to do so, but the NRC staff is able to ensure consideration only for comments received on or before this date.

Public Protection Notification

The NRC may not conduct or sponsor, and a person is not required to respond to, a collection of information unless the document requesting or requiring the collection displays a currently valid OMB control number.

XIV. Criminal Penalties

For the purposes of Section 223 of the Atomic Energy Act of 1954, as amended (AEA), the NRC is issuing this proposed rule that would amend or add §§ 26.3, 50.47, 50.54, 50.59, 50.71, 50.75, 50.82, 50.200, 52.110, 72.30, 72.72, 72.212, 72.218, 73.51, 73.54, 73.55, and 140.11 as well as appendix G to 10 CFR part 20, appendix A to 10 CFR part 50, and appendix E to 10 CFR part 50, under one or more of Sections 161b, 161i, or 161o of the AEA. Willful violations of these provisions would be subject to criminal enforcement. Criminal penalties as they apply to regulations in 10 CFR parts 20, 26, 50, 52, 72, 73 and 140 are discussed in §§ 20.2402, 26.825, 50.111, 52.303, 72.86, 73.81 and 140.89.

XV. Voluntary Consensus Standards

The National Technology Transfer and Advancement Act of 1995, Public Law 104-113, requires that Federal agencies use technical standards that are developed or adopted by voluntary consensus standards bodies unless the use of such a standard is inconsistent with applicable law or otherwise impractical. In this proposed rule, the NRC would revise regulations associated with decommissioning in 10 CFR parts 20, 26, 50, 51, 52, 72, 73, and 140. This action would not constitute the establishment of a standard that contains generally applicable requirements.

XVI. Availability of Guidance

The NRC is issuing for comment four draft regulatory guides to support the implementation of the proposed requirements in this proposed rule, as well as to support other recommendations made in the supporting regulatory bases regarding areas where the decommissioning guidance could be improved or enhanced. You may access information and comment submissions related to the Draft Guides (DGs) by searching on https://www.regulations.gov under Docket ID NRC-2015-0070. You may submit comments on this draft guidance by the methods outlined in the ADDRESSES section of this document.

1. The DG-1346, “Emergency Planning for Decommissioning Nuclear Power Reactors” (ADAMS Accession No. ML21347A046), is a new regulatory guide.

2. The DG-1347, “Decommissioning of Nuclear Power Reactors,” (ADAMS Accession No. ML21347A080), would be Revision 2 to the existing Regulatory Guide 1.184.

3. The DG-1348, “Assuring the Availability of Funds for Decommissioning Production or Utilization Facilities,” (ADAMS Accession No. ML21347A081), would be Revision 2 to the existing Regulatory Guide 1.159.

4. The DG-1349, “Standard Format and Content for Post-Shutdown Decommissioning Activities Report,” (ADAMS Accession No. ML21347A138), would be Revision 2 to the existing Regulatory Guide 1.185.

XVII. Public Meeting

The NRC will conduct a public meeting on this proposed rule for the purpose of describing this proposed rule to the public and facilitating development of public comments on this proposed rule.

The NRC will publish a notice of the location, time, and agenda of the meeting in the Federal Register , on Regulations.gov , and on the NRC's public meeting website at least 10 calendar days before the meeting. Stakeholders should monitor the NRC's public meeting website for information about the public meeting at: https://www.nrc.gov/public-involve/public-meetings/index.cfm.

XVIII. Availability of Documents

The documents identified in the following table are available to interested persons through one or more of the following methods, as indicated.

Throughout the development of this rule, the NRC may post documents related to this rule, including public comments, on the Federal rulemaking website at https://www.regulations.gov under Docket ID NRC-2015-0070.

List of Subjects

10 CFR Part 20

• Byproduct material
• Criminal penalties
• Hazardous waste
• Licensed material
• Nuclear energy
• Nuclear materials
• Nuclear power plants and reactors
• Occupational safety and health
• Packaging and containers
• Penalties
• Radiation protection
• Reporting and recordkeeping requirements
• Source material
• Special nuclear material
• Waste treatment and disposal

10 CFR Part 26

• Administrative practice and procedure
• Alcohol abuse
• Alcohol testing
• Appeals
• Chemical testing
• Drug abuse
• Drug testing
• Employee assistance programs
• Fitness for duty
• Management actions
• Nuclear power plants and reactors
• Privacy
• Protection of information
• Radiation protection
• Reporting and recordkeeping requirements

10 CFR Part 50

• Administrative practice and procedure
• Antitrust
• Backfitting
• Classified information
• Criminal penalties
• Education
• Emergency planning
• Fire prevention
• Fire protection
• Incorporation by reference
• Intergovernmental relations
• Nuclear power plants and reactors
• Penalties
• Radiation protection
• Reactor siting criteria
• Reporting and recordkeeping requirements
• Whistleblowing

10 CFR Part 51

• Administrative practice and procedure
• Environmental impact statements
• Hazardous waste
• Nuclear energy
• Nuclear materials
• Nuclear power plants and reactors
• Reporting and recordkeeping requirements

10 CFR Part 52

• Administrative practice and procedure
• Antitrust
• Combined license
• Early site permit
• Emergency planning
• Fees
• Incorporation by reference
• Inspection
• Issue finality
• Limited work authorization
• Nuclear power plants and reactors
• Probabilistic risk assessment
• Prototype
• Reactor siting criteria
• Redress of site
• Penalties
• Reporting and recordkeeping requirements
• Standard design
• Standard design certification

10 CFR Part 72

• Administrative practice and procedure
• Hazardous waste
• Indians
• Intergovernmental relations
• Nuclear energy
• Penalties
• Radiation protection
• Reporting and recordkeeping requirements
• Security measures
• Spent fuel
• Whistleblowing

10 CFR Part 73

10 CFR Part 140

For the reasons set out in the preamble and under the authority of the Atomic Energy Act of 1954, as amended; the Energy Reorganization Act of 1974, as amended; and 5 U.S.C. 552 and 553, the NRC is proposing to amend 10 CFR parts 20, 26, 50, 51, 52, 72, 73, and 140 as follows:

PART 20—STANDARDS FOR PROTECTION AGAINST RADIATION

1. The authority citation for part 20 continues to read as follows:

Authority: Atomic Energy Act of 1954, secs. 11, 53, 63, 65, 81, 103, 104, 161, 170H, 182, 186, 223, 234, 274, 1701 (42 U.S.C. 2014, 2073, 2093, 2095, 2111, 2133, 2134, 2201, 2210h, 2232, 2236, 2273, 2282, 2021, 2297f), Energy Reorganization Act of 1974, secs. 201, 202 (42 U.S.C. 5841, 5842); Low-Level Radioactive Waste Policy Amendments Act of 1985, sec. 2 (42 U.S.C. 2021b); 44 U.S.C. 3504 note.

Appendix G to Part 20 [Amended]

2. In appendix G to part 20, amend paragraph E.1. of section III by:

a. Removing the word “or” and adding in its place the word “of”; and

b. Removing the phrase “20 days” and adding in its place the phrase, “45 days”.

PART 26—FITNESS FOR DUTY PROGRAMS

3. The authority citation for part 26 continues to read as follows:

Authority: Atomic Energy Act of 1954, secs. 53, 103, 104, 107, 161, 223, 234, 1701 (42 U.S.C. 2073, 2133, 2134, 2137, 2201, 2273, 2282, 2297f); Energy Reorganization Act of 1974, secs. 201, 202 (42 U.S.C. 5841, 5842); 44 U.S.C. 3504 note.

4. Amend § 26.3, by revising paragraph (a) to read as follows:

5. In § 26.825(b), remove “26.3”.

PART 50—DOMESTIC LICENSING OF PRODUCTION AND UTILIZATION FACILITIES

6. Revise the authority citation for part 50 to read as follows:

Authority: Atomic Energy Act of 1954, secs. 11, 53, 63, 81, 101, 102, 103, 104, 105, 108, 122, 147, 149, 161, 181, 182, 183, 184, 185, 186, 187, 189, 223, 234 (42 U.S.C. 2014, 2073, 2093, 2113, 2131, 2132, 2133, 2134, 2135, 2138, 2152, 2167, 2169, 2201, 2231, 2232, 2233, 2234, 2235, 2236, 2237, 2239, 2273, 2282); Energy Reorganization Act of 1974, secs. 201, 202, 206, 211 (42 U.S.C. 5841, 5842, 5846, 5851); Nuclear Waste Policy Act of 1982, sec. 306 (42 U.S.C. 10226); National Environmental Policy Act of 1969 (42 U.S.C. 4332); 44 U.S.C. 3504 note; Sec. 109, Pub. L. 96-295, 94 Stat. 783.

7. Revise § 50.1 to read as follows:

8. In § 50.2, revise the definition for Certified fuel handler and add a definition for Non-power production or utilization facility in alphabetical order to read as follows:

9. In § 50.36(c)(6), add “or § 52.110(a) of this chapter” after “§ 50.82(a)(1)”.

10. Revise § 50.38 to read as follows:

11. In § 50.44(b) introductory text, add “or § 52.110(a) of this chapter” after “§ 50.82(a)(1)”.

12. In § 50.46(a)(1)(i), add “or § 52.110(a) of this chapter” after “§ 50.82(a)(1)”.

13. In § 50.47, revise paragraph (b) introductory text and add paragraph (f) to read as follows:

14. In § 50.48(f) introductory text, add “or § 52.110(a) of this chapter” after “§ 50.82(a)(1)”.

15. In § 50.49(a), remove “§ 52.110(a)(1)” and add in its place “§ 52.110(a)”.

16. In § 50.51, in paragraph (b) introductory text, remove the words “to authorize ownership and possession of the production or utilization facility,”.

17. In § 50.54:

a. Amend paragraph (m)(2)(i) by:

i. Designating the table;

ii. Revising the heading of the newly designated table; and

iii. Revising footnote 2 to the table;

b. In paragraph (o), remove “52.110(a)(1)” and add in its place “52.110(a)”;

c. Redesignate paragraphs (p)(3) and (4) as paragraphs (p)(5) and (6);

d. Redesignate paragraphs (p)(1) and (2) as paragraphs (p)(2) and (3) and revise newly redesignated paragraphs (p)(2) and (3);

e. Add new paragraph (p) introductory text and paragraphs (p)(1) and (4);

f. Revise paragraphs (q)(1) introductory text and (q)(1)(iii) and (q)(2) and (3);

g. Remove the words “after February 21, 2012” wherever they appear in paragraphs (q)(4) and (5); and

h. Add paragraphs (q)(7) and (8);

i. Remove the words “after April 1, 1981,” in paragraph (s)(2)(ii);

j. In paragraph (s)(3), remove the words “The NRC” and add in their place the words “If the planning standards for radiological emergency preparedness apply to offsite radiological emergency response plans, the NRC”;

k. In paragraph (t)(1)(ii), remove the period from the second sentence and add in its place the word “or,”;

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

m. In paragraph (w) introductory text, remove the words “under this part”;

n. In paragraphs (w)(4)(ii) and (iii), add the words “or § 52.110 of this chapter” after the words “§ 50.82” wherever they appear;

o. Add paragraphs (w)(5) and (6);

p. In paragraph (y), add “or § 52.110(a) of this chapter” after “§ 50.82(a)(1)”; and

q. Revise paragraph (bb).

The revisions and additions read as follows:

18. In § 50.59:

a. In paragraph (b), remove “§ 50.110” and add in its place “§ 52.110(a) of this chapter”; and

b. Revise paragraph (d)(3) to read as follows:

19. In § 50.60(a), add “or § 52.110(a) of this chapter” after “§ 50.82(a)(1)”.

20. In § 50.61(b)(1), add “or § 52.110(a) of this chapter” after “§ 50.82(a)(1)”.

21. In § 50.62(a), add “or § 52.110(a) of this chapter” after “§ 50.82(a)(1)”.

22. In § 50.65(a)(1), remove “52.110(a)(1)” and add in its place “52.110(a)”.

23. In § 50.71, revise paragraphs (c) and (e)(4) to read as follows:

24. In § 50.75:

a. Revise the first sentence in paragraph (a);

b. Revise paragraphs (b)(1), (3), and (4) and add paragraph (b)(5);

c. Revise paragraph (e)(1) introductory text;

d. In paragraph (e)(1)(i):

i. Remove the phrase “formulas in § 50.75(c)” and add in its place the phrase “table of minimum amounts in paragraph (c)”;

ii. Remove the phrase “site-specific estimate” wherever it appears and add in its place the phrase “site-specific decommissioning cost estimate”;

e. In paragraph (e)(1)(ii) introductory text, remove the phrase “site-specific estimate” wherever it appears and add in its place the phrase “site-specific decommissioning cost estimate”;

f. In paragraph (e)(1)(ii)(B), add “or § 52.110 of this chapter” after “50.82 of this part”;

g. In paragraph (e)(1)(v), add “or § 52.110 of this chapter” after “or § 50.82”;

h. Amend paragraph (f) by:

i. Revising paragraph (f)(1);

ii. Removing paragraph (f)(2);

iii. Redesignating paragraphs (f)(3) through (5) as (f)(2) through (4); and

iv. Revising newly redesignated paragraph (f)(2) and paragraph (f)(3) introductory text;

i. In paragraphs (h)(1)(iii) and (iv), remove the words “Director, Office of Nuclear Reactor Regulation, or Director, Office of Nuclear Material Safety and Safeguards, as applicable,” wherever they appear in the first sentence of each paragraph and add in their place the words, “Document Control Desk as specified in § 50.4”.

j. In paragraph (h)(1)(iv), add “or § 52.110(h) of this chapter” after “§ 50.82(a)(8)” wherever it appears.

k. In paragraph (h)(2), remove the words “given the Director, Office of Nuclear Reactor Regulation, or Director, Office of Nuclear Material Safety and Safeguards, as applicable,” wherever they appear and add in their place, the words, “given to the Document Control Desk as specified in § 50.4”.

l. In paragraph (h)(2), add “or § 52.110(h) of this chapter” after “§ 50.82(a)(8)” wherever it appears.

The revisions and addition read as follows:

25. In § 50.82:

a. Revise paragraphs (a)(2), (a)(4), (a)(6)(ii), (a)(8)(i)(A), (a)(8)(ii), (a)(8)(v) introductory text and (a)(8)(vii) introductory text;

b. Revise paragraphs (a)(9) introductory text and (a)(9)(ii)(F);

c. Revise paragraph (b) introductory text; and

d. Redesignate paragraph (b)(6) as (b)(8) and add new paragraphs (b)(6) and (7).

The revisions and additions read as follows:

26. Revise § 50.109 to read as follows:

27. In § 50.155, add paragraphs (h)(6), (7), and (8) to read as follows:

28. Add § 50.200 to read as follows:

29. Amend appendix E to part 50 by:

a. Removing paragraph I.6;

b. In paragraph IV.4, removing the words “of the later of the date” and “or December 23, 2011,”;

c. Adding paragraph IV.8;

d. In paragraph IV.A.7, removing the words, “By June 23, 2014, identification” and adding in their place the word, “Identification”;

e. In paragraph IV.A.9, removing the words, “By December 24, 2012, for” and adding in their place the word, “For”;

f. In paragraph IV.B.1, removing the words, “By June 20, 2012, for” and adding in their place the word, “For”;

g. In paragraph IV.C.2, removing the words, “By June 20, 2012, nuclear” and adding in their place the word, “Nuclear”;

h. In paragraph IV.E.8.c introductory text, removing the words, “By June 20, 2012, for” and adding in their place the word, “For”;

i. In paragraph IV.E.8.d, removing the last sentence;

j. In paragraph IV.F.2.d removing the words “and should fully participate in one hostile action exercise by December 31, 2015”;

k. Removing and reserving paragraph IV.F.2.j(v);

l. Adding paragraph IV.F.2.k;

m. In paragraph IV.I, removing the words, “By June 20, 2012, for” and adding in their place the word, “For”;

The revisions and addition read as follows:

Appendix E to Part 50—Emergency Planning and Preparedness for Production and Utilization Facilities

IV. * * *

8. A nuclear power reactor licensee is not subject to the requirements of paragraphs 4, 5, and 6 of this section once the NRC dockets the licensee's certifications required under § 50.82(a)(1) or § 52.110(a) of this chapter.

F. * * *

2. * * *

k. For each nuclear reactor for which the NRC has docketed the certifications required under § 50.82(a)(1) or § 52.110(a) of this chapter, the nuclear reactor's licensee must follow the biennial exercise requirements of paragraph 2 of this section.

Appendix I to Part 50 [Amended]

30. In section IV.C, add “or § 52.110(a) of this chapter” after “§ 50.82(a)(1)”.

PART 51—ENVIRONMENTAL PROTECTION REGULATIONS FOR DOMESTIC LICENSING AND RELATED REGULATORY FUNCTIONS

31. The authority citation for part 51 continues to read as follows:

Authority: Atomic Energy Act of 1954, secs. 161, 193 (42 U.S.C. 2201, 2243) Energy Reorganization Act of 1974, secs. 201, 202 (42 U.S.C. 5841, 5842); National Environmental Policy Act of 1969 (42 U.S.C. 4332, 4334, 4335); Nuclear Waste Policy Act of 1982, sec. 144(f), 121, 135, 141, 148 (42 U.S.C. 10134(f), 10141, 10155, 10161, 10168); 44 U.S.C. 3504 note.

32. In § 51.53, in paragraph (d), remove the words “Each applicant for a license amendment authorizing decommissioning activities for a production or utilization facility either for unrestricted use or based on continuing use restrictions applicable to the site; and each applicant for a license amendment approving a license termination plan or decommissioning plan under § 50.82 of this chapter” and add in their place the words “Each applicant for a license amendment approving an irradiated fuel management plan under § 50.54(bb) of this chapter; each applicant for a license amendment approving a license termination plan under § 50.82 of this chapter or § 52.110 of this chapter or a decommissioning plan under § 50.82 of this chapter”.

33. In § 51.95, in paragraph (d) remove the words “of an operating or combined license authorizing decommissioning activities at a production or utilization facility covered by § 51.20,” and add in their place the words “approving an irradiated fuel management plan under § 50.54(bb) of this chapter, or the amendment approving a license termination plan under § 50.82 of this chapter or § 52.110 of this chapter or a decommissioning plan under § 50.82 of this chapter”.

PART 52—LICENSES, CERTIFICATIONS, AND APPROVALS FOR NUCLEAR POWER PLANTS

34. Revise the authority citation for part 52 to read as follows:

Authority: Atomic Energy Act of 1954, secs. 53, 63, 81, 103, 104, 147, 149, 161, 181, 182, 183, 185, 186, 189, 223, 234 (42 U.S.C. 2073, 2093, 2113, 2133, 2134, 2167, 2169, 2201, 2231, 2232, 2233, 2235, 2236, 2239, 2273, 2282); Energy Reorganization Act of 1974, secs. 201, 202, 206, 211 (42 U.S.C. 5841, 5842, 5846, 5851); 44 U.S.C. 3504 note.

35. In § 52.0, revise paragraph (a) to read as follows:

36. In § 52.63, revise paragraph (b)(2) to read as follows:

37. In § 52.109, remove the words “to authorize ownership and possession of the production or utilization facility,”.

38. In § 52.110, revise paragraphs (b), (d), (e), (f)(2), (h)(1)(i), and (h)(2), add paragraphs (h)(5) through (7), and revise paragraph (i) introductory text and paragraph (i)(2)(vi) to read as follows:

PART 72—LICENSING REQUIREMENTS FOR THE INDEPENDENT STORAGE OF SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C WASTE

39. The authority citation for part 72 continues to read as follows:

Authority: Atomic Energy Act of 1954, secs. 51, 53, 57, 62, 63, 65, 69, 81, 161, 182, 183, 184, 186, 187, 189, 223, 234, 274 (42 U.S.C. 2071, 2073, 2077, 2092, 2093, 2095, 2099, 2111, 2201, 2210e, 2232, 2233, 2234, 2236, 2237, 2238, 2273, 2282, 2021); Energy Reorganization Act of 1974, secs. 201, 202, 206, 211 (42 U.S.C. 5841, 5842, 5846, 5851); National Environmental Policy Act of 1969 (42 U.S.C. 4332); Nuclear Waste Policy Act of 1982, secs. 117(a), 132, 133, 134, 135, 137, 141, 145(g), 148, 218(a) (42 U.S.C. 10137(a), 10152, 10153, 10154, 10155, 10157, 10161, 10165(g), 10168, 10198(a)); 44 U.S.C. 3504 note.

40. In § 72.13, add paragraph (e) to read as follows:

41. In § 72.30, revise paragraph (b) and paragraph (c) introductory text to read as follows:

42. In § 72.32, revise paragraphs (a) introductory text and (c) to read as follows:

43. In § 72.44, revise paragraph (f) to read as follows:

44. In § 72.62, revise paragraph (a)(2) to read as follows:

45. In § 72.72, revise paragraph (d) to read as follows:

46. In § 72.212, add paragraph (b)(9)(vii) to read as follows:

47. Revise § 72.218 to read as follows:

PART 73—PHYSICAL PROTECTION OF PLANTS AND MATERIALS

48. The authority citation for part 73 continues to read as follows:

Authority: Atomic Energy Act of 1954, secs. 53, 147, 149, 161, 170D, 170E, 170H, 170I, 223, 229, 234, 1701 (42 U.S.C. 2073, 2167, 2169, 2201, 2210d, 2210e, 2210h, 2210i, 2273, 2278a, 2282, 2297f); Energy Reorganization Act of 1974, secs. 201, 202 (42 U.S.C. 5841, 5842); Nuclear Waste Policy Act of 1982, secs. 135, 141 (42 U.S.C. 10155, 10161); 44 U.S.C. 3504 note.

Section 73.37(b)(2) also issued under sec. 301, Pub. L. 96-295, 94 Stat. 789 (42 U.S.C. 5841 note).

49. In § 73.51, revise paragraphs (a) introductory text, (a)(1) introductory text, and (a)(2) and add paragraph (a)(3) to read as follows:

50. In § 73.54, remove the introductory text, revise the paragraph (a) introductory text, paragraph (b) introductory text, and paragraph (c) introductory text, and add paragraphs (i) and (j) to read as follows:

51. In § 73.55:

a. Revise paragraph (b)(3) introductory text;

b. Add paragraphs (b)(9)(ii)(B)( 1 ) and ( 2 );

c. Revise paragraphs (c)(6), (e)(9)(v)(A), (j)(4)(ii), and (p)(1)(i) and (ii).

The revisions and additions read as follows:

PART 140—FINANCIAL PROTECTION REQUIREMENTS AND INDEMNITY AGREEMENTS

52. The authority citation for part 140 continues to read as follows:

Authority: Atomic Energy Act of 1954, secs. 161, 170, 223, 234 (42 U.S.C. 2201, 2210, 2273, 2282); Energy Reorganization Act of 1974, secs. 201, 202 (42 U.S.C. 5841, 5842); 44 U.S.C. 3504 note.

53. In § 140.11, add paragraph (a)(5), redesignate paragraph (b) as paragraph (c), revise newly redesignated paragraph (c), and add new paragraph (b) to read as follows:

54. In § 140.81, revise paragraph (a) to read as follows: