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COMMUNICATION STRATEGY FOR THE ENHANCEMENT OF PUBLIC AWARENESS 

REGARDING POWER REACTORS TRANSITIONING TO DECOMMISSIONING 

February 2015 

Introduction 

The objective of this communication plan is to outline the U.S. Nuclear Regulatory Commission 

(NRC) strategy for communicating the key messages regarding the NRC process and practices 

for public and stakeholder engagement during the decommissioning of nuclear power plants, as 

well as to provide a resource for addressing frequently asked questions in this area. 

Background 

To date, the NRC has provided oversight for the decommissioning of 11 nuclear power plants. 

When a licensee decides to permanently cease operations at a nuclear power plant, the facility 

must be decommissioned by safely removing it from service and reducing residual radioactivity 

to a level that permits release of the property and termination of the operating license. The 

NRC has strict rules governing nuclear power plant decommissioning, involving cleanup of 

radioactively contaminated plant systems and structures, and removal of the radioactive fuel. 

These requirements protect workers and the public during the entire decommissioning process 

and the public after the license is terminated. With the recent increase in the number of power 

reactors beginning the decommissioning process and significant changes that have occurred in 

the regulations since 1996, there is an increased public interest in the decommissioning 

process. 

Public involvement in the NRC’s activities is a cornerstone of strong, transparent regulation of 

the nuclear industry. The NRC recognizes the public’s interest in the proper regulation of 

nuclear activities and provides various opportunities for citizens to make their opinions known. 

The NRC seeks to elicit public involvement early in any regulatory process, including 

decommissioning, so that safety concerns that may affect a community can be resolved in a 

timely and practical manner. This process is considered vital to assuring the public that the 

NRC is making sound, balanced decisions about nuclear safety. Consistent with this policy, the 

NRC frequently hosts meetings with interested stakeholders, including members of the public, 

non-government organizations, and local and state government officials in order to discuss any 

topics that may be of interest. 

The NRC recognizes the need and desire for community involvement in the decommissioning of 

a nuclear power plant. Decommissioning is a complex process and the NRC believes that 

decommissioning impacts need to be vetted within the local community. However, the NRC 

was created by the Congress to be an independent regulator charged with ensuring public 

health and safety and protecting the environment. As an independent regulator, the NRC 

ensures that all members of the public are given a fair and equal opportunity to comment on a 

licensee’s decommissioning and license termination plans. Therefore, the NRC does not 

officially recognize or endorse any specific special interest group, public or private organization, 

community group, coalition, or individual. This approach assures that one or more 

organizations does not dominate the public forum and allows members of the public to provide 

alternative and differing viewpoints and comments to the NRC. 

 
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Current Public Outreach Tools for Power Reactor Decommissioning 

Opportunities for the Public to Comment on Decommissioning Documents

The NRC regulations currently offer the public several opportunities to review and provide 

comments on licensee documents during the decommissioning process. Specifically, under the 

NRC regulations in Title 10 of the Code of Federal Regulations (10 CFR) Part 50.82, the NRC is 

required to publish a notice of the receipt of the licensee’s Post-Shutdown Decommissioning 

Activities Report (PSDAR) and the License Termination Plan (LTP), make the PSDAR and LTP 

available for public comment, schedule separate meetings in the vicinity of the location of the 

licensed facility to discuss both the PSDAR and LTP within 60 days of receipt, and publish a 

notice of the meetings in the Federal Register and another forum readily accessible to 

individuals in the vicinity of the site. An example of this type of Federal Register (FR) notice for 

a PSDAR is the NRC’s Notice of Public Meeting and Availability of Report published in the 

Federal Register on April 1, 2013, for the Kewaunee Power Station (78 FR 19540). 

The PSDAR serves as the main planning tool for the decommissioning process, including the 

estimated cost of the decommissioning activities. By regulation, a licensee has two years to 

submit the PSDAR from the time operations are permanently ceased at the nuclear power plant 

(i.e., the beginning of decommissioning process). The licensee must submit its LTP at least two 

years before the expected license termination request (i.e., the end of the decommissioning 

process). In addition, because the NRC approves the LTP by amending the license, a hearing 

opportunity occurs for the LTP and follows the normal amendment process. 

Community Outreach and Advisory Groups

For many years, the NRC has recommended that licensees involved in decommissioning 

activities form a community committee to obtain local citizen views and concerns regarding the 

decommissioning process and spent fuel storage issues. It has been the NRC view that those 

licensees who actively engage the community maintain better relations with the local citizens. 

NRC guidance related to creating a site-specific community advisory board can be found in 

NUREG-1757, “Consolidated Decommissioning Guidance,” Appendix M, “Overview of the 

Restricted Use and Alternate Criteria Provisions of 10 CFR Part 20, Subpart E,” Section M.6. 

However, it should be noted that Appendix M applies only to sites requesting to terminate a 

license with restricted use conditions, which has not occurred. To date, all nuclear power plant 

facilities have been released for unrestricted use at the conclusion of the decommissioning 

activities. In addition, Appendix M does not require licensees to create a community advisory 

board, but only provides recommendations for methods of soliciting public advice. Nonetheless, 

Section M.6 contains useful guidance and suggestions for effective public involvement in the 

decommissioning process that could be adopted by any licensee. 

Experience gained and lessons learned from prior decommissioning projects have been well 

documented by the nuclear industry. In 2005, the Electric Power Research Institute (EPRI) 

published the “Maine Yankee Decommissioning – Experience Report – Detailed Experience 

1997 – 2004” (EPRI 1011734). In this lessons learned report, industry recognized that 

engaging the local community and officially forming a Community Advisory Panel or Board 

(CAP/CAB) is a good practice. Specifically, the EPRI report states that “the Maine Yankee 

Community Advisory Panel was established in 1997 to enhance opportunities for public 

involvement in the decommissioning process of Maine Yankee. The CAP represents the local 

community. By thoroughly reviewing the decommissioning process, the CAP is in a position to 

advise Maine Yankee on key issues of concern to the local community.” 
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In addition, the NRC held a decommissioning status meeting in October 2005, during which 

members of the Maine Yankee CAP were invited to speak to the Commission regarding their 

experiences. At this meeting, the Vice Chairman of the Maine Yankee CAP concluded that 

CAPs “provide an important window for the public in the process of decommissioning, and 

provide the opportunity for issues of local concern to be addressed both within and without the 

strict process defined by the regulations. As a result, in our decommissioning, a level of trust 

was gained that had evaded Maine Yankee for the previous 24 years of operation.” The 

complete information presented during that meeting can be found at http://www.nrc.gov/readingrm/doc-collections/commission/slides/2005/20051018/hudson-material-decommissioning.pdf. 

Since the decommissioning of Maine Yankee, licensees have employed a CAP or CAB at many 

other sites, including Connecticut Yankee, Yankee Rowe, Big Rock Point, Millstone, Humboldt 

Bay, and others. For all of the nuclear plants currently entering into the decommissioning 

process, the NRC has strongly encouraged the licensees to establish a CAP/CAB for the 

decommissioning effort in order to enhance communications with the local communities and 

stakeholders. In addition, as part of our role as an independent regulator, the NRC frequently 

attends CAP/CAB meetings to address questions or concerns from the community members. 

Decommissioning Lessons Learned and Additional Guidance Documents

The NRC routinely documents experience gained and lessons learned. Generally, lessons 

learned are any items that could be of interest and benefit to many licensees. Lessons learned 

include positive or negative discussions that are worth sharing with NRC licensees and 

stakeholders. For example, lessons learned can include significant and recurring NRC 

comments on licensing documents, issues that have been formally documented and evaluated 

by the staff, case studies providing useful site-specific examples or best practices for some 

aspects of decommissioning, or unsuccessful experiences that should be shared to avoid 

recurrence. The NRC staff has historically collected lessons learned during the 

decommissioning process and attempted to incorporate these insights into ongoing activities. 

Additional References on Decommissioning Lessons Learned can be accessed on the NRC 

website. This is a preliminary collection of documents that the NRC, Fuel Cycle Facilities 

Forum, Nuclear Energy Institute, Electric Power Research Institute, and Organization of 

Agreement States have compiled to inform existing and future NRC licensees, NRC staff, 

States, and other interested stakeholders about potential lessons learned from past 

decommissioning actions. This list includes information such as: a) title of the document, b) list 

of author(s), c) summary of the content of each document, and d) details on the availability of 

each document in this bibliography. In addition, reports from the Decommissioning Lessons 

Learned Database can be accessed for the years 2002, 2004, 2005, 2006, and 2007. 

The NRC believes that adequate processes and the associated guidance documents are 

already in place to establish best practices regarding public and stakeholder engagement in the 

decommissioning of nuclear power plants within the bounds of the NRC’s role as an 

independent regulator. However, many of these guidance documents are being updated to 

reflect additional lessons learned in decommissioning and will continue to be revised as new 

insights are gained. In addition, the existing practices within the agency promote community 

involvement in the decommissioning process to the extent practicable and attempt to capture 

lessons learned during each decommissioning activity to improve the overall regulatory process. 

As such, the NRC does not currently plan to create a Regulatory Issue Summary or any other 

type of additional guidance document to further discuss public participation in the 
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decommissioning process, but will continue to implement the process and procedures already 

established for the decommissioning of nuclear power plants. 

Infographics Related to Decommissioning 

Numerous graphics related to decommissioning are available; two of the more useful examples 

are included below. 
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In addition, the Decommissioning Program Brochure and the Decommissioning Backgrounder, 

which are released by the NRC’s Office of Public Affairs and have been recently updated, give a 

high level overview of the decommissioning process and discuss involvement opportunities. 

Frequently Asked Questions on Reactor Decommissioning 

Because of the breadth and scope of questions regarding the decommissioning process, the 

information has been grouped into various topics that can be accessed by clicking on the 

appropriate link below to go to the associated appendix: 

Power Reactor Decommissioning Process

Decommissioning Funding

Spent Fuel Considerations

Emergency Preparedness During Decommissioning

Physical Security During Decommissioning

Note that several of the questions refer to specific decommissioning reactors; however, the 

information contained in the answers is generally applicable to all decommissioning facilities. 

Please use these links, as well as those provided throughout the document to other external 

information sources to promote consistent communication on these topics. 
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Power Reactor Decommissioning Process 

NUREG-1628 – This report contains a definition of decommissioning and a discussion of 

alternatives. It also provides a focus on decommissioning experiences in the United States and 

how the NRC regulates the decommissioning process. Questions related to spent fuel, lowlevel waste, and transportation related to decommissioning are answered. Questions related to 

license termination, the ultimate disposition of the facility, and finances for completing 

decommissioning and hazards associated with decommissioning are also addressed. This 

document provides responses to questions related to public involvement in decommissioning as 

well as providing the public with sources of obtaining additional information. 

1. What are the unrestricted and restricted release criteria for power reactor 

decommissioning? 

A: For Unrestricted Release

• Total Effective Dose Equivalent (TEDE) ≤ 25 mrem/yr (0.25 mSv/a) 

• As Low As is Reasonably Achievable (ALARA) 

• Average member of the critical group 

• All pathways 

• Period of performance = 1000 years 

For Restricted Release

• Total Effective Dose Equivalent (TEDE) ≤ 25 mrem/yr (0.25 mSv/a) 

• ALARA, with institutional controls in effect 

• Legally enforceable institutional controls 

• If institutional controls fail, doses do not exceed 100 mrem/yr (1 mSv/a), or 500 mrem/yr 

(5 mSv/a), under specific circumstances 

• Financial assurance via an independent third party 

• Licensee and NRC public input/outreach requirements 

2. Will the NRC allow the resale of non-radioactive equipment and secondary side 

components (e.g., turbines, moisture separator reheaters, heat exchangers, 

condensers, intake pumps, intake piping, outfall piping, all associated piping and 

electrical components, etc.) since some of these are worth substantial amounts? Will 

the proceeds go to offset the cost of decommissioning? 

A: The NRC regulates the safe use of radioactive materials and does not regulate commerce. 

As such, the NRC will ensure that the materials released from the site for unrestricted use 

meet radiological release requirements. If a licensee has non-radioactive assets, including 

plant components, equipment, and recyclable materials that can generate revenue, the use 

of this revenue is outside of the NRC’s authority and is under the oversight of the state 

public utility commission. The licensee may also transfer or sell contaminated parts and 

equipment to other licensees for use in their nuclear facilities, within the limitations of the 

applicable transportation requirements. 

3. Will there be public announcements when any "allowable" waste is to be released 

into the environment? What are the upper limits for releasing radiation and 
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chemicals into the environment during the decommissioning process? When were 

those limits established and what would trigger a process to reevaluate those limits? 

A: During decommissioning, both liquid and airborne radiological releases will be monitored 

and are required to be maintained below the same radiological limits as when the plant was 

in operation. The radiological effluent release criteria were established in 10 CFR Part 20 

many decades ago. The licensee will continue to provide the NRC with environmental 

effluent reports and the NRC will conduct inspections of this area throughout the 

decommissioning process. The results of the NRC inspections and any associated findings 

will be published in inspection reports that are publicly available. The NRC does not monitor 

or regulate the release of toxins or other forms of non-radiological waste as that is within the 

purview of the Environmental Protection Agency and other similar federal and state 

government organizations. 

4. What level of on-site staffing will NRC provide during the decommissioning process, 

and in which areas of technical expertise and oversight? 

A: Consistent with agency procedures, the NRC typically maintains a full time resident 

inspector onsite during part of the first year after permanent shutdown. The resident 

inspector oversees the plant transition from operation to permanent shutdown, in order to 

verify that the licensee complies with their license, technical specifications, and procedures. 

Generally, early in the first year after permanent shutdown, the Inspection Program is 

transferred from the Reactor Oversight Program to Inspection Manual Chapter (IMC) 2561, 

“Reactor Decommissioning.” The contents of this IMC are publicly available and outline the 

oversight activities that the NRC staff will be involved in throughout decommissioning. 

During the first year, the licensee prepares the plant for safe decommissioning. The actions 

taken by the licensee include the modification of systems, shipment of radioactive waste, 

emptying of tanks, draining of systems, and electrical isolation of components. As during 

plant operations, the resident inspection staff is supplemented with special inspection 

expertise as needed, which includes security, emergency response, health physics, 

environmental monitoring, and engineering. NRC inspections continue throughout 

decommissioning until the licensee demonstrates that the site meets the license termination 

requirements. The level of decommissioning inspections will be commensurate with the 

licensee’s planned decommissioning activities. 

5. Why doesn’t the NRC review and approve the Post Shutdown Decommissioning 

Activities Report (PSDAR)? 

The NRC does not approve the PSDAR. In Commission Paper, SECY 96-086, the 

Commission determined that decommissioning activities could be safely conducted under 

the current license conditions and restrictions. The Commission determined that a detailed 

decommissioning plan requiring NRC review and approval would be redundant to the 

activities already authorized by the Commission in the facility license. Any actions outside 

the license would require the licensee to request a license amendment and would ask them 

to appropriately justify the why the change was safe. 
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6. Will the NRC commence a rulemaking on decommissioning in order to add efficiency 

to the current process of requesting exemptions from operating plant requirements? 

A: The NRC staff added to its Common Prioritization of Rulemaking a line item for a rulemaking 

on power reactor decommissioning transition. The Common Prioritization of Rulemaking is 

a budgeting tool that the NRC uses to assign resources to rulemaking activities, in light of 

the NRC’s priorities and resource limitations. Separately, the NRC staff is in the process of 

determining the appropriate timeframe for the decommissioning transition rulemaking. 

 Even if this rulemaking were initiated immediately and implemented following the NRC 

rulemaking process of developing a regulatory basis; issuing a proposed rule; requesting, 

receiving, and addressing public comments on the proposed rule; and issuing a final rule; 

any improvements it may provide would have little impact on the plants that are currently 

transitioning to decommissioning. It would be available primarily for plants that may be 

considering permanently shutting down in the future. This consideration will be a factor in 

the NRC’s determination of how to schedule such a rulemaking effort. The Commission 

recently directed the staff to report to the Commission its views on the need for an 

integrated rulemaking for decommissioning, based on lessons learned from the most recent 

operating closures. 

7. What guidance or other regulatory documents does the NRC currently use to evaluate 

power reactors as they transition into decommissioning? 

A: The main NRC guidance documents for decommissioning are: 

NUREG-1496 Generic Environmental Impact Statement in Support of Rulemaking on 

Radiological Criteria for License Termination of NRC-Licensed Nuclear 

Facilities 

NUREG-1700 Standard Review Plan for Evaluating Nuclear Power Reactor License 

Termination Plans 

NUREG-1727 NMSS Decommissioning Standard Review Plan 

NUREG-1757 Consolidated Decommissioning Guidance 

Regulatory Guide (RG) 1.179 Standard Format and Content of License Termination 

Plans for Nuclear Power Reactors 

RG 1.184 Decommissioning of Nuclear Power Reactors 

RG 1.185 Standard Format and Content for Post-Shutdown Decommissioning 

Activities Report 

RG 4.21 Minimization of Contamination and Radioactive Waste Generation: LifeCycle Planning 

RG 4.22 Decommissioning Planning During Operations 

 
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Decommissioning Funding

1. What is the minimum decommissioning funding assurance amount required by NRC? 

How is this amount calculated? 

A: Licensees must estimate the minimum funding assurance amount needed for radiological 

decommissioning by using formulas in NRC regulations. As an alternative, licensees may 

also use a site-specific cost estimate to determine the minimum funding assurance amount 

needed, provided that this amount is greater than the amount derived from the NRC 

formulas. The minimum decommissioning funding assurance amount must be adjusted 

annually by applying the latest escalation factors for labor, energy and waste burial costs. 

2. What are NRC decommissioning trust funds used for? Will there be sufficient funds 

to bring the site back to greenfield status? 

A: NRC decommissioning trust funds are used for decommissioning as defined and regulated 

by the NRC. The NRC formulas address only those decommissioning costs needed to 

remove a facility or site safely from service and reduce radioactivity to safe levels to allow for 

termination of the license. However, the costs of removal of non-radiological systems and 

structures are not included in the NRC decommissioning cost formulas. In addition, the 

costs of managing and storing spent fuel on site until transfer to the Department of Energy 

for permanent disposal are not included in NRC decommissioning cost formulas. The NRC 

does not ensure that there are sufficient funds to bring a site to greenfield status. 

3. What will be the NRC’s response if a licensee’s present decommissioning fund 

balance does not appear to be adequate? 

A: At any time the NRC determines that unacceptable levels of decommissioning funding 

assurance exists, the NRC will determine a remedial course of action for any particular 

licensee on a case-by-case basis. 

4. What types of actions might the NRC implement to ensure that licensees make any 

necessary adjustments to their financial assurance? 

A: The NRC may require the licensee to take one or more actions, including but not limited to: 

• Produce a funding guarantee from a parent company 

• Develop a specific catch-up funding plan 

• Provide more frequent or additional funding reports 

• Make additional payments towards principal 

• Obtain a surety bond 

5. Does the NRC impose investment restrictions for decommissioning trust funds? 

A: At the end of 2003, new investment regulations applicable to non-electric utility licensees 

became effective. The new restrictions prohibit decommissioning trust fund investments in 

licensee and nuclear sector securities or obligations. Absent State or Federal law 

restrictions applicable to decommissioning trusts, the NRC regulations require that 

investments be made in accordance with the "prudent investor standard.” 
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6. Does the NRC regulate trustees for nuclear decommissioning trusts? 

A: Three federal bank regulatory agencies - the Federal Deposit Insurance Corporation, the 

Office of the Comptroller of the Currency, and the Federal Reserve Board (Banking 

Agencies) - have oversight of the trust departments of financial institutions that typically act 

as trustees for nuclear decommissioning trusts. The Banking Agencies' examination 

process evaluates a trustee organization's compliance with its fiduciary obligations, including 

complying with applicable investment restrictions, whether they are the NRC's or those of 

any other agency or those of a state. 

7. What happens if a licensee goes bankrupt? 

A: If the licensee of a nuclear power plant or its parent company declares bankruptcy, it must 

still fulfill its decommissioning obligations. The NRC takes an active role to monitor and 

possibly intervene in bankruptcy proceedings through the Department of Justice to assure 

protection of public health and safety. 

8. Are limited liability companies shielded in any way from decommissioning 

obligations? 

A: Licensees, whether they are corporations or limited liability companies, are fully responsible 

for fulfilling their decommissioning obligations. Normally, their parent companies or 

stockholders are not licensed entities and therefore are not responsible, absent certain 

special circumstances that might render them liable. 
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Spent Fuel Considerations

High Burnup Spent Fuel

Nuclear fuel rods are ceramic pellets of uranium oxide (UO2), about the size of a finger joint, 

stacked and sealed inside a long metal tube (cladding) about as big around as a Sharpie pen. 

The space between the pellets and cladding is filled with helium. 

“Spent fuel” refers to fuel used in a commercial nuclear reactor that has been removed because 

it can no longer economically sustain a nuclear reaction. Burnup refers to the uranium 

consumed in the nuclear reaction. It is expressed in gigawatt-days per metric ton of uranium 

(GWd/MTU)—a measure of how long a fuel rod is in the core and the power level it reaches. 

“High burnup fuel” is in the reactor core for longer than “low burnup fuel.” 

In a reactor core, nuclear fuel undergoes physical and chemical changes. The UO2 fissions—

splits apart and releases energy—producing fission gas and fission products. The pellets swell, 

crack and release a small amount of fission gas inside the rod. The cladding also reacts with the 

reactor cooling water. This reaction forms an oxide layer on the outside (similar to rust) and 

produces hydrogen. About 15% of the hydrogen enters into the cladding metal. These 

processes occur slowly at first, then start to accelerate at about the time the fuel reaches burnup 

of 45 GWd/MTU. At this burnup level, the fuel is about 75% through its currently useful life. 

The maximum burnup that NRC has allowed increased as technology advanced our ability to 

understand the changes the fuel undergoes in the reactor core. Our understanding of those 

changes is key to our ability to make safety decisions. Average fuel burn-ups have increased 

from around 35 GWd/MTU two decades ago, to over 45 GWd/MTU today. Anything over 45 

GWd/MTU is considered high burnup. 

Available information indicates that both low and high burnup spent fuel can be stored and 

transported safely. This information comes from operational experience with storage systems 

and short term tests. Considerable data is available on the properties of low burnup spent fuel 

and more confirmatory data is being obtained daily on high burnup fuel. That data is 

instrumental in enabling the NRC to make licensing decisions to allow spent fuel storage in 

specific dry cask designs. 

As utilities began discharging high burnup spent fuel, dry storage cask designers amended their 

designs to account for physical differences in the spent fuel. The NRC approved several spent 

fuel storage amendments based on research data sufficient to show that high burnup fuel could 

be stored safely. Additional high burnup storage amendments are under review. The NRC fully 

expects data from current testing will become available to confirm that high burnup spent fuel 

will behave as expected and remain in a safe condition even in a transportation accident. 

From the San Onofre Nuclear Generating Station (SONGS), Units 2 and 3, PSDAR Meeting

1. High burnup fuel has been used at San Onofre since 1996. Was notice given to the 

public? Were workers made aware that high burnup fuel is more radioactive? 

Answer

New fuel designs, including high burnup fuel (HBF), undergo an NRC technical review, and 

are generally approved for use via a Topical Report that provides the technical evaluation of 
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the new fuel design and lists any limitations for its use. Once the new fuel design is 

approved with an associated Topical Report, all NRC reactor licensees are permitted to use 

that fuel design within their reactor core without requesting specific NRC approval, as long 

as the core continues to meet all design and safety limits. In an operating nuclear reactor, 

burnup is one of the many parameters that are considered in designing the fuel and core for 

each operating cycle. Many parameters are evaluated throughout the operating cycle to 

verify that design specific limits are met. Data-based, predictive tools are used to evaluate 

these parameters over the cycle. Throughout the cycle, physics testing is also done to 

confirm key physics parameters are consistent with predictions. When a new fuel is 

designed its use is limited by the data available to support the associated predictive tools. 

As such, burnup is limited for a particular fuel by the supporting predictive tools, the data 

supporting the predictive tools, and the requirement to not exceed any design limit. In an 

operating reactor, the main distinction between HBF and low burnup fuel (LBF) is the 

amount of exposure to which the fuel is subjected. The fuel is required to meet all safety 

limits at all times during the operating cycle. 

a. High burnup fuel is hotter and "between 2 and 158 times more radioactive," 

requiring the waste to be cooled on-site in spent fuel pools for at least 12-15 years 

(rather than 5 years). Does the NRC agree with this statement? If not, how much 

more radioactive would the NRC say high burnup is? 

Answer

HBF is typically defined as fuel with a burnup (a measure of the time a fuel assembly 

stays in the reactor core) greater than 45,000 megawatt-days per metric ton of initial 

uranium. LBF means any fuel with a lower exposure than this value. Average fuel 

burnups have increased from around 35,000 megawatt days per metric ton uranium two 

decades ago, to over 45,000 megawatt days per metric ton uranium today. Higher 

burnup fuel is thermally hotter and more radioactive than lower burnup fuel for a given 

cooling time. The difference in decay heat (a function of the fuel transferring heat to 

decrease its temperature over time) and radioactive source term depends on the 

difference in the fuel burnup (i.e., how long the fuel was being used in the reactor), the 

initial enrichment of the fuel, and the irradiation environment that the fuel was exposed to 

in the reactor. Moving from 35,000 megawatt days per metric ton uranium to 45,000 and 

55,000 increases the overall heat source term or level by 35% and 78%, respectively, 

and increases the radiation source term or level by 33% and 72%, respectively. 

HBF must be cooled longer than LBF before it can be placed into a dry storage system. 

How much longer depends on the difference in burnup, the specific dry storage system 

design, and the decay heat loading pattern of the fuel being used. As an example, for a 

5.0 weight percent enriched (in Uranium-235) fuel assembly in one particular storage 

system, the required cooling time goes from 4.5, to 7, to 12 years, for a fuel burnup of 

35,000, 45,000, and 55,000 megawatt days per metric ton uranium, respectively. 

b. How does the presence of high burnup fuel affect the decommissioning process 

at San Onofre? What specific problems does this higher radioactive fuel present 

for waste storage in fuel pools and dry cask storage at San Onofre and how much 

longer will this radiation last? How will decommissioning be impacted by the 

current onsite storage of the spent fuel? 
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Answer: 

HBF is decommissioned in accordance with the same regulatory requirements as other 

fuel types. The higher heat load of HBF will require the water circulation pumps in the 

spent fuel pool to circulate more water in order to efficiently reject the added heat from 

the HBF and maintain water temperatures. Once the HBF spent fuel is sufficiently 

cooled in the pool, it may be transferred from the fuel pool to the dry storage canisters in 

the Interim Spent Fuel Storage Installation (ISFSI), after which the spent fuel pool can be 

decommissioned consistent with the licensee’s decommissioning strategy and schedule. 

A dry cask storage system that is designed to handle HBF has been reviewed and 

approved for use at SONGS. This review included determining if the dry cask storage 

system provides adequate margins to safely store HBF. SCE is only able to store fuel 

within the fuel specifications and other limits in the associated certificate of compliance, 

which provides reasonable assurance of the safety of the stored fuel. The impact on the 

overall decommissioning schedule for SONGS will be dependent on the spent fuel 

management plan that the licensee will determine. To date, SCE has not provided their 

decommissioning and spent fuel management plans for the NRC to evaluate. 

c. We understand the NRC staff is worried about short and long-term waste storage 

in dry casks of high burnup fuel and has initiated a new study to determine if it 

can safely be stored in dry casks. Is this report complete? Will it be released to 

the public, and when? 

Answer

It is not clear what NRC study the question is addressing. Currently HBF is licensed to 

be stored in approved dry cask storage systems for an initial period of 20 years, with a 

potential extension of one or more 40 year intervals. However, with the delay in 

availability of a final repository for spent nuclear fuel it was determined that the dry 

storage of spent fuel might have to account for a considerably longer period of time than 

originally planned. As such, the NRC staff is examining the regulatory framework and 

potential technical issues related to extended dry storage and subsequent transportation 

of spent nuclear fuel for periods beyond the initial licensing and single renewal period 

(i.e., beyond ~60 years of storage). This analysis has been undertaken in order to 

identify potential changes needed to the associated regulations or guidance, in 

accordance with direction from the Commission in Staff Requirements Memorandum 

COMSECY-10-0007, “Project Plan for the Regulatory Program Review to Support 

Extended Storage and Transportation (EST) of Spent Nuclear Fuel” (see Agencywide 

Documents Access and Management System (ADAMS) Accession No. ML103400287). 

As a first step in the EST Program, the NRC identified the technical information needs 

associated with extended dry fuel storage systems (see draft issued for public comment, 

at ADAMS Accession No. ML120580143) in order to understand potential degradation 

mechanisms, the level of knowledge about these mechanisms, and how that 

degradation would affect the ability of a dry storage component to fulfill its regulatory or 

safety function. The EST Program is expected to be complete in the next five to seven 

years, subject to any funding limitations that may occur. The report, once finalized, will 

be made available to the public. 
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d. One of the NRC concerns is that there is no way to monitor what is occurring 

inside the dry casks. How does the NRC propose to monitor the highly 

radioactive material inside of the dry casks? How many casks will be required to 

safely store all the high burnup fuel that is on site in both the spent fuel pools and 

dry casks at San Onofre? How much high burnup fuel is on site in fuel pools and 

dry cask at San Onofre? 

Answer

The NRC collects data on the total amount of spent fuel stored at commercial facilities, 

like SONGS, throughout the country. This information, and much more concerning 

spent nuclear fuel, is available on the NRC website at http://www.nrc.gov/waste/spentfuel-storage.html. The NRC does not independently maintain records showing how 

many fuel assemblies have been loaded into each cask that were specifically HBF. 

However, licensees and certificate holders are required to register each cask with the 

NRC pursuant to 10 CFR 72.212(b)(2). The cask registrations and information 

contained therein are subject to routine inspection by the NRC. When the last routine 

ISFSI inspection report was issued for SONGS on May 20, 2011, SCE had loaded 11 

canisters that contained at least one HBF assembly out of the 55 canisters loaded on the 

ISFSI pad at SONGS (see ADAMS Accession No. ML111430612). For any given 

facility, more specific spent fuel information is considered security-sensitive and is 

therefore not disclosed to the public. 

The NRC is actively monitoring the efforts of industry and the Department of Energy 

(DOE) to better understand fuel aging mechanisms and promote methods for monitoring 

the behavior of fuel inside a sealed dry cask. DOE is taking an active role in funding 

Nuclear Energy University Program (NEUP) projects to study this issue. In addition, 

DOE is sponsoring a demonstration test with a variety of HBFs to benchmark models of 

fuel behavior and obtain, through monitoring of the mockup test casks, the temperature 

and gases evolved which will indicate how the fuel is behaving. This demonstration will 

provide data that will allow the creation of enhanced models to predict the behavior of 

various types of spent fuel. 

 

In terms of the number of dry casks that will be necessary to store the spent fuel at 

SONGS, it will depend on which dry cask storage system is used for the 

decommissioning effort and when the spent fuel is actually moved from the fuel pool to 

dry cask storage due to the different sizes of casks available and the actual heat being 

given off by the assemblies the licensee wants to load into the dry casks. This 

information is not yet available from SCE, but should be provided as a part of the 

SONGS spent fuel management plan. 

e. We know that MOX fuel was used in San Onofre, Unit 1 and was removed to the 

GE Morris facility in Illinois. How and when was that done and under what permit 

was that done? If MOX fuel was transported away, can other high burnup fuel be 

moved from the site in the same way to the same or similar places? 

Answer

Between March 1972 and September 1980, 270 fuel assemblies were shipped from 

SONGS to the GE Morris facility. Mixed oxide (MOX) fuel was not included as a part of 

these shipments. Specifically, Appendix A of the Technical Specifications for GE Morris 
- 15 - 

(see ADAMS Accession No. ML042180413) identifies that 270 stainless steel clad fuel 

assemblies are allowed to be stored at the GE Morris facility. Page 12 of a 2004 NRC 

Inspection Report (see ADAMS Accession No. ML040070255) indicates that the MOX 

fuel irradiated at SONGS is zircaloy-clad. Consequently, GE Morris is not allowed to 

store the MOX fuel from SONGS. The NRC report also indicates that the MOX fuel was 

stored in the SONGS Unit 1 spent fuel pool, and Page 9 of the 2004 NRC Inspection 

Report identifies that irradiated MOX fuel can be loaded into the Advanced NUHOMS 

dry cask storage system employed by SONGS. In addition, a 2011 NRC Inspection 

Report (see ADAMS Accession No. ML111430612) identified that eighteen of the dry 

storage canisters located on the SONGS ISFSI pad contain either irradiated fuel or 

greater than Class C radioactive waste from SONGS Unit 1. 

SONGS can transfer spent fuel only to licensees/licensed facilities that are authorized to 

receive the spent fuel and have an agreement in place with SCE to accept the fuel. 

Information on SCE’s plans for the spent fuel is not yet available, but should be provided 

as a part of the SONGS spent fuel management plan. 

2. The NRC has not approved the transport of dry storage casks nor even short-term dry 

cask storage (beyond 20 years) for high burnup fuel. Will the NRC continue to allow 

high burnup fuel use even though they do NOT have an approved safe solution to 

store or transport this waste - even short-term? High burnup fuel is more difficult to 

store and transport. In addition, there is no transport cask design approved to store 

high burnup fuel. The NRC currently licenses dry cask storage for high burnup for 

only 20 years. The current expiration date for SONGS is February 5, 2023. 

 

Answer

While HBF does possess a higher initial heat load that LBF, it can be safely stored and 

transported if the potential fuel degradation mechanisms are addressed and the NRC 

regulatory requirements are satisfied. The assertion that the NRC has not licensed any 

casks for transport of HBF or short term storage of HBF is incorrect. HBF is in dry cask 

storage at a number of reactor sites across the country. In addition, there are a limited 

number of transportation packages certified for the transport of spent nuclear fuel. Of those, 

some have HBF included in the authorized contents. For example, the Model No. UMS 

Universal Transport Cask package and the Model No. Hi-Star 180 transportation package 

have current transportation certificates of compliance which include HBF contents. 

The NRC will continue to license the storage of HBF as long as the applicants can show that 

it can be done safely by meeting the applicable regulatory requirements. Currently HBF is 

licensed to be stored in approved dry cask storage systems for an initial period of 20 years, 

with a potential extension of one or more 40 year intervals. The NRC is developing a path 

forward for licensing the storage of HBF up to 60 years and beyond based on time limiting 

aging analysis, aging management plans, analysis of consequences of a variety of potential 

fuel behaviors, and test data provided by the applicants. 

There are three types of dry cask storage systems for spent nuclear fuel: 1) storage only 

which have not been approved for transportation, 2) dual purpose systems that are 

designed for both storage and transport (most of these have been approved for storage of 

HBF only but some have also been approved for transport of HBF), and 3) canisterized 

systems where the canister may or may not be put into a new overpack for transport. The 

current dry storage cask design in use at SONGS is the Transnuclear (TN) NUHOMS 
- 16 - 

system, Certificate of Compliance 072-1029, Amendment 1, with a 24PT4-DSC canister. 

The Technical Specifications for this canister allow fuel burnup of up to 60,000 megaawatt 

days per metric ton uranium. Note that higher burnup fuel requires a longer required cooling 

time (years after discharge) in the spent fuel pool before the heat load has decreased 

enough to meet the Certificate’s Technical Specifications for using the dry cask storage 

system. 

The Coalition cites the Regulatory Information Conference (RIC) talk by a senior NRC staff 

member on the storage and transportation of HBF. Specifically, the statement on Slide 7 of 

the Argonne National Laboratory (ANL) results that says there is “enough data to determine 

there is a regulatory issue. Insufficient data to support a licensing position.” In late 2012 the 

NRC completed an experimental program at ANL to determine at what temperature cladding 

goes from ductile (able to bend) to brittle condition. That program found that different types 

of cladding become brittle at different temperatures. Dry storage cask designers will 

therefore need to address the specific fuel cladding type and its temperature to show that 

HBF can stay intact during a transport accident. 

3. Is removing the spent fuel pool considered part of the decommissioning process? If 

so, how can dry casks be transported without use of a spent fuel pool in cases where 

that may be needed? What vulnerabilities are there in San Onofre’s spent fuel pools? 

What improvements could be made to improve safety? Will any of them be made? If 

so, when? If not, why not? 

Answer

There are two spent fuel pools at SONGS, both of which were built during construction of 

the plants. Both of the spent fuel pools meet all regulatory requirements for the storage of 

new and spent nuclear fuel. Decontamination of the spent fuel pools, which may involve 

removal and disposal, is a decommissioning activity. Before starting to decommission 

systems and components needed for moving, unloading, and shipping the spent fuel (like 

the spent fuel pool), the licensee must develop a plan for removal of the spent fuel from the 

reactor site, and a plan for how the spent fuel will be managed until the time that DOE takes 

title to, and possession of, the spent fuel (these requirements are in 10 CFR 72.218). This 

plan is part of the licensee’s program for managing and providing funding for management 

of spent fuel following permanent shutdown of the reactor until the time that DOE takes the 

fuel. The licensee is required (in 10 CFR 50.54(bb)) to submit this spent fuel management 

program to the NRC for review within 2 years following permanent shutdown. 

The licensee’s program for spent fuel management and its plan for removal of the spent fuel 

from the site would need to consider what equipment, systems, or facilities would be needed 

to place the spent fuel into an approved transportation package, before decommissioning 

these systems. For example, if the storage system is a canister-based system (the spent 

fuel is confined in a welded canister that is placed in a storage overpack that provides 

radiation shielding), the canister may be placed into a transportation overpack / package 

without the need for a spent fuel pool. If the storage system was not transportable, a spent 

fuel pool or dry transfer facility would be needed to repackage the spent fuel into an 

appropriate transportation package when necessary. 

SONGS currently uses a dry canister based storage system for its spent fuel in conjunction 

with the spent fuel pools. Regarding transportability of this system, the 24PT1 canister is 

certified for transport (under CoC 9255; MP187 transportation cask), as is the 24PT4 
- 17 - 

canister (under CoC 9302; MP197 transportation cask). As discussed earlier, the NRC is 

currently reviewing an application for amendment to CoC 9302 (MP197 transportation cask) 

to allow transportation of the 32PTH2 canister. 

4. Is the utility allowed to decommission the reactor and support facilities, including the 

spent fuel pool, while spent fuel is still being stored at the ISFSI? If so, what happens 

if a spent fuel cask has a problem and needs to be unloaded, but there is no spent 

fuel pool? 

 Answer 

 There are currently no NRC requirements that prohibit decommissioning of reactor support 

facilities while spent fuel is being stored at the ISFSI. NRC [in 10 CFR 50.54(bb)] requires 

licensees to submit, for NRC review and approval, a program for managing and providing 

funding for management of spent fuel following permanent cessation of operation of the 

reactor until the time that title to, and possession of, the spent fuel is transferred to the 

Department of Energy (DOE). In addition, NRC requires [in 10 CFR 72.218] licensees to 

include in this spent fuel management program, a plan for removal of the spent fuel from the 

reactor site, and a plan for how the spent fuel will be managed before starting to 

decommission systems and components needed for moving, unloading, and shipping the 

spent fuel. 

Dry spent fuel storage is safe and environmentally sound. Over the past 20 years, there 

have been no radiation releases which have affected the public, no radioactive 

contamination, and no known or suspected attempts to sabotage spent fuel casks or ISFSIs. 

Dry cask storage systems are designed to withstand earthquakes, floods, tsunamis, 

tornadoes, projectiles, temperature extremes, and other unusual or accident scenarios, and 

to continue to perform their safety functions without the need to repackage the fuel into a 

new system. 

However, because of the uncertainty regarding DOE’s program and planning for ultimate 

spent fuel disposition, the storage period may be longer than initially envisioned. The 

associated future aging-management inspection or repackaging needs at these sites will 

continue to be evaluated through the renewal process for licenses and storage Certificates 

of Compliance. Staff believes that appropriate fuel handling capability or safe movement of 

the fuel to a site where handling capability is available could be provided in an appropriate 

timeframe in the event that a problem is identified with a storage cask at a shutdown or 

decommissioned reactor. 

This issue was included in Petition for Rulemaking by the C-10 group. One of the 

Petitioner’s requests was that NRC require a safe and secure hot cell transfer station 

coupled with an auxiliary pool as part of an upgraded ISFSI design certification and licensing 

process. The staff is further considering the issue of fuel handling capability at ISFSIs at 

shutdown or decommissioned reactors, where there is no spent fuel pool for immediate 

examination or unloading of the storage casks, as part of our review of that Petition. 

5. How do the waste confidence hearings affect the probability and timing at San Onofre 

for shipment of nuclear waste to remote interim or permanent storage? 
- 18 - 

Answer

The proposed Waste Confidence rulemaking has no impact on the probability or timing of 

SONGS spent nuclear fuel being shipped to a mined geologic repository because Waste 

Confidence is the NRC's generic determination regarding the environmental impacts of 

storing spent nuclear fuel after the end of the licensed life for operations of a nuclear reactor 

and before final disposal in a repository. The proposed Waste Confidence rulemaking does 

not authorize the shipment to or disposal of spent nuclear fuel to a mined geologic 

repository. Similarly, Waste Confidence has no impact on SONGS spent nuclear fuel being 

shipped to a remote interim storage site because the proposed rulemaking does not 

authorize the shipment to or storage of spent nuclear fuel at an interim storage facility. 

 
- 19 - 

Emergency Preparedness During Decommissioning

Key Messages 

1. Emergency preparedness requirements are not being eliminated but aligned to the risk to the public, which i

with the spent fuel pool, rather than an operating power reactor. 

2. Licensese must maintain an onsite emergency plan providing for the classification of emergencies, notificati

government authorities, and coordination of offsite organizations responding onsite (i.e., firefighting, medica

3. Licensees will be required to maintain the capability to mitigate a spent fuel pool event. 

No. Question Response

1 Are the current NRC regulations for 

emergency preparedness for nuclear power plants designed for plants that are decommissioning? 

There are no explicit regulatory provisions distinguishing EPreparedness (EP) requirements for a nuclear power reacdown from those for an operating power reactor. The EP r10 CFR 50.47, “Emergency plans,” and Appendix E, “EmePreparedness for Production and Utilization Facilities,” to 1continue to apply to a nuclear power reactor after permaneoperations and removal of fuel from the reactor vessel. Comodify their emergency plans to reflect the reduced risk to safety from power reactors that have been permanently shreactor licensees transitioning to a decommissioning statufrom certain EP regulatory requirements before amending plans. 

2 What is the scope of exemptions to EP 

requirements being considered? 

The exemptions requested are consistent with those previodecommissioning power reactors and bring the EP requiredecommissioning facilities in line consistent with the currenindependent spent fuel storage installations (ISFSIs) – 10 exemptions include the requirement for formal offsite radiopreparedness (REP) plans and a scaling back of the licensemergency response commensurate with reduced risk. Threquired to maintain an onsite emergency plan addressing an emergency, notification of designated offsite governmecoordination onsite the response onsite of offsite response firefighting, medical assistance). 
- 20 - 

No. Question Response

3 What does the elimination of formal offsite 

radiological emergency preparedness (REP) plans mean? How would actions be taken to protect the public around the facility? 

A formal offsite radiological emergency plan requiring certinot required because the risk to public health and safety is significantly less than when the plant was operating. As sufor detailed pre-planning within 10-miles (plume exposure facility would no longer exist. In the highly unlikely case wnecessary, there should also be sufficient time for State or undertake offsite protective measures. State and local govcomprehensive emergency management (all hazard) plans a wide-range of non-radiological emergencies. There is amthat offsite response organizations can effectively carry ou(including evacuations) for a wide range of severe events. 

4 If not, how does the NRC determine what 

emergency preparedness requirements are appropriate to the risk presented by the facility? 

When a licensee submits a request for an exemption from an operating nuclear power plant, the NRC staff reviews thassociated justifications. The justifications include technictype and risk of accidents that could occur at the plant. Thused to provide the NRC reasonable assurance that in graexemption: (1) an offsite radiological release would not exEnvironmental Protection Agency protective action guides (as applicable to design basis accidents such as fuel handwaste processing accidents); and (2) in the unlikely event design-basis accident resulting in a loss of all cooling, suffto initiate appropriate mitigating actions, and if needed, impprotective actions using a comprehensive emergency manprotect the health and safety of the public. If the impacts were determined not to require offsite proteclicensee would not need to maintain compliance with certarequirements. For the onsite risks at the plant, the NRC wlicensee continue to comply with the appropriate EP requir
- 21 - 

No. Question Response

5 What is an exemption, and why is it 

required? 

An exemption applies to a licensee seeking regulatory reliecomply with regulation). Current EP regulations do not takreduced risks associated with potential accidents that may power reactor that has permanently ceased operation and the reactor vessel to a spent fuel pool, which presents less state than does an operating plant because certain types olonger possible. Historically, exemptions to EP requiremengrant regulatory relief on a case-by-case basis. Until an exonsite and offsite EP programs must be maintained and al

6 How does the NRC exemption process 

work, and why are exemptions allowed? 

The practice of considering exemptions is a well-establisheregulatory process that allows licensees to address site-spimplement alternative approaches for circumstances not ncontemplated in the regulations for operating reactors. This not unique to decommissioning licensees or to the specibut is an important tool that allows the agency to provide arelief, permitting licensees to make appropriate modificatiocommensurate with the site-specific risks that are present shutdown reactor during decommissioning. The NRC makexemption requests on a site-specific, case-by-case basis, established process that includes the NRC staff’s assessmtechnical safety analysis submitted by the licensee. The NRC may grant exemptions from various regulatory reresponse to a request from a licensee, applicant, or interesthe Commission to consider granting an exemption, the reinformation that demonstrates that specific criteria in the rethose in 10 CFR 50.12, “Specific exemptions,” will be met. review of requests for exemptions pursuant to 10 CFR 50.the exemptions are authorized by law, will not present undhealth and safety, and are consistent with the common defand whether any special circumstances, as defined in the The special circumstances listed in 10 CFR 50.12 reflect sthat exemptions may be necessary. Special circumstancewhenever application of the regulation conflicts with other 
- 22 - 

No. Question Response

application of the regulation in the particular circumstance underlying purpose of the rule, compliance with the regulatundue hardship, the exemption would result in benefit to thof the public, the exemption would provide temporary relief regulation and the licensee has made good faith efforts to present a material circumstance not considered when the adopted for which it would be in the public interest to grant The NRC will grant an exemption only if it concludes that aare met. If granted, the exemption, which contains the staand conclusions, is published in the Federal Register. 

7 Will the NRC ensure that adequate safety 

requirements are in place for power reactors during decommissioning? 

Yes, the NRC will continue to maintain appropriate safety rprovide appropriate oversight for all permanently shut dowreactors through all phases of the decommissioning procesrequired to comply with the existing conditions and technictheir licenses until and unless those requirements are chanamendment or exemption. During decommissioning, licensees typically request licensexemptions from certain regulations, accompanied by a techange the requirements based on a facility’s permanently defueled condition. The NRC staff applies the same rigor technical evaluation of amendment and exemption requestwhether a facility is in operation or a permanently shutdowcondition. The NRC will only approve a license amendmeexemption if the staff concludes its issuance will continue tassurance of adequate protection of the public health and In addition, the NRC’s oversight and inspection activities wdecommissioning sites throughout the decommissioning p
- 23 - 

No. Question Response

8 Why is Commission approval required? In a Staff Requirements Memorandum (SRM) dated May 1

SECY-08-0024, the Commission directed that “The staff shCommission approval for any reduction in the effectivenesemergency plan that requires an exemption from the requi50.47(b) and Appendix E to 10 CFR Part 50.” 

9 Does NRC have the ability to impose 

condition as part of any exemption request that it might approve? 

Yes, the NRC staff may grant an exemption based on a relicensee satisfy certain conditions. In this case, the conditdescribed and evaluated, and incorporated into the license is granted. 

10 The use of the license exemption process 

appears to undermine the NRC Safety Regulations in that it does not allow meaningful public participation, even though the result may be perceived as resulting in serious compromise to public safety. Why isn’t the license amendment process used, instead of the exemption process? 

The NRC regulations in 10 CFR Part 50.12 specify conditiolicensees can seek exemptions from certain regulatory reqsituations, licensees can also seek amendments. 

11 Why has the NRC never refused a request 

of a decommissioning reactor licensee to be exempted from off-site emergency preparedness requirements? 

The NRC staff has not approved all exemption requests as Some licensees have modified or withdrawn their exemptioaccordance with determinations made during the staff’s revThe NRC staff reviews requests for exemptions for decomusing the same process and standards it uses to review reexemptions submitted by licensees of operating plants. Th

exemption applications considers whether the exemption iwhether the exemption would present an undue risk to the safety, whether the exemption is consistent with common dand whether special circumstances, as defined in the NRCIf the NRC staff determines that a proposed exemption reqthese requirements, the NRC staff typically provides the lic

to supplement its application with additional information. If through interactions with the NRC staff that the exemption 
- 24 - 

No. Question Response

the licensee often chooses to withdraw the exemption requhaving NRC issue a denial. As a recent example, on August 27, 2013, Southern Califowithdrew its request for an exemption to defer performing tevaluated exercise of its onsite and offsite emergency planNuclear Generating Station (SONGS) Units 2 and 3, as reqPart 50, Appendix E, Sections IV.F.2.b and c. As indicated request was withdrawn “[b]ased on discussions with the Nconsulting with the offsite agencies.” SCE requested the eAugust 8, 2013, shortly after submitting certifications that it shut down and defueled SONGS Units 2 and 3. 

12 When does the FEMA oversight of offsite 

emergency preparedness, as established in 44 CFR 350, cease? 

When approval of the EP Exemption SECY Paper is grantCommission, and all required safety evaluation reports havthe NRC’s approval of exemption and license amendment proposed permanently defueled emergency plan (PDEP) awill notify FEMA that formal radiological emergency prepar44 CFR 350) is no longer required as a condition of the Pa

13 What is the NRC process for reviewing 

and making decisions on Permanently Defueled Emergency Plan (PDEP) submittals? Please provide the staff and Commissioner processes, as well as accurate and realistic timeframes under which the PDEP is reviewed, considered and available to the public, and voted on. 

The NRC staff uses the same process for reviewing and mpermanently defueled emergency plans as the staff uses fmaking decisions on an operating facility’s emergency planLicensees may make changes to the emergency plan withonly if the change does not reduce the effectiveness of the and the emergency plan, as changed, continues to meet thAppendix E to 10 CFR Part 50 and the planning standards If the licensee determines that the change involves a reduceffectiveness, the licensee must submit the proposed eme

for prior NRC approval as a license amendment request uThe NRC review process is initiated when a licensee subm

amendment request to the NRC Document Control Desk. review is performed by NRC staff, to: (1) determine if docuis sufficient to support the staff’s technical review; (2) coord
- 25 - 

No. Question Response

interfaces with various NRC offices; and (3) establish a schtechnical review of the request. As soon as practicable, following completion of the acceptawill publish a notice in the Federal Register. The Federal Rdescribe the proposed change, describe the procedure for on the proposed change, announce the opportunity to requprovide instructions for requesting a hearing. As part of thlicense amendment request, the NRC staff verifies that the informed the appropriate State officials of the proposed chDuring the review process, requests for additional informatthe licensee for the purpose of enabling the staff to obtain information needed to make a regulatory decision on a licerequest that is well-informed, technically correct, and legalthe NRC completes its technical review of the proposed chwritten safety evaluation report, it publishes a notice in the indicating the approval or denial of the request. The NRC has a general goal of completing action on a licerequest within one year; however, staff action on a license may be expedited to reduce time to approximately 6-9 monrequest’s safety significance, complexity, and assigned priIf the licensee determines that the plan as changed would requirements in Appendix E to 10 CFR Part 50 and the pla10 CFR 50.47(b), the licensee may submit an exemption rlicense amendment request, prior to implementing any chaPDEP changes submitted for prior NRC approval, including implement exemptions already approved by the Commissithe Director of the Office of Nuclear Reactor Regulation anseparate Commission approval. 
- 26 - 

No. Question Response

14 Under an approved PDEP, does the 

licensee retain responsibility for onsite emergency response and communication with offsite organizations? 

Yes. At all times, the licensee is primarily responsible for sactivities, including on-site emergency response and all asand communication with offsite organizations (including, burelevant law enforcement and emergency management orgFollowing approval of the PDEP, the requirement in 10 CFmaintain an onsite emergency plan that meets the plannin10 CFR 50.47(b) and the requirements of Appendix E to 1(unless exempted by the Commission), remains in effect. The onsite emergency plan will outline on-shift and augmestaffing capable of responding to emergency events approthat has permanently ceased operation and transferred fuevessel to a spent fuel pool or dry cask storage. The onsite also require the notification of designated offsite governmethe declaration of an emergency event and provide for the response organizations (e.g., firefighting, medical transpor

15 Do you believe a defueled reactor poses 

essentially the same off-site radiological emergency risk profile as an operating reactor? Why? 

No. When a nuclear power plant permanently ceases operlicensee defuels the reactor, the risk to the public from an asignificantly, since the accident sequences that dominated risk are no longer applicable. The primary remaining source of risk to the public is assocaccidents that involve the spent fuel stored in the spent fuepredominant design-basis accident for a defueled reactor iaccident. The risk of a radiological release from a spent fuel pool at plant would typically be lower than from a spent fuel pool aThis is because the heat generated by spent fuel significantime following transfer of the spent fuel from the reactor to products decay. In contrast, the amount of heat generated at an operating reactor does not significantly decrease with additional heat discharged from newly transferred spent furemoved from the reactor and placed in the spent fuel poomonths. 
- 27 - 

No. Question Response

16 Since the NRC staff studied the potential 

for fires to occur in spent fuel pools, what is their conclusion on the actual probability of such an event occurring? 

Over the past several decades, the NRC has periodically eof spent fuel pools, and has consistently concluded that sprobust structures that are likely to withstand severe earthqcredible challenges and thus provide for safe storage of spTwo recent evaluations, including NUREG-1738, “TechnicaFuel Pool Accident Risk at Decommissioning Nuclear PowNUREG-2161, “Consequence Study of a Beyond-Design-BAffecting the Spent Fuel Pool for a U.S. Mark I Boiling Watdemonstrated that the risks of a radiological release form a very low. 

17 Has the NRC performed an analysis of the 

risks of a spent fuel pool fire? If so, what did the NRC conclude? 

Over the past 35 years, the NRC has sponsored a number evaluate various aspects of spent fuel pool safety, securitysummary of the past studies (including NUREG-1738 for dplants) is provided in SECY-13-0112, “Consequence Study Design-Basis Earthquake Affecting the Spent Fuel Pool foWater Reactor,” which is commonly known as the Spent FA regulatory analysis that analyzed spent fuel pools across nuclear power plants was submitted to the Commission in These studies have shown that spent fuel storage is safe arelease of radionuclides due to an accident is low. The anreview of the risk of a release from a spent fuel pool fire. The results of NUREG-1738 indicated that the risk is low aCommission's quantitative health objectives. The risk was because of the very low likelihood of a zirconium fire, even consequences from a zirconium fire could be serious. The results of more recent studies, such as NUREG-2161, the earlier conclusions that spent fuel pools are robust struto withstand severe earthquakes without leaking and expocould lead to a fire. 
- 28 - 

No. Question Response

18 There are 12 sites where civilian nuclear 

power generating activities have permanently ceased and all that does, or will remain at these facilities is the spent fuel and reactor generated Greater-ThanClass C Wastes stored on site awaiting removal by the Department of Energy. After these facilities permanently ceased operations, each submitted a PDEP to modify parts of the Emergency Plan based on the significantly reduced risks of a radiological event at a decommissioning plant. After these facilities were given approval to make the changes authorized under the PDEP, has there been an event, emergency or threat including natural disasters or hostile acts that proved to be too significant or challenging to be handled by the NRC approved PDEP? 

There have been no events, emergencies, or threats, incluor hostile acts that have proved to be either too significant handled in accordance with the NRC-approved PDEP. The radiological risks are greatly reduced at a decommissiplant and have a very low probability of causing an event timpacts offsite. In addition, the sites are required to maintain an NRC-apprensure potential hostile threats will be safely and securely the plant. 
- 29 - 

No. Question Response

19 NUREG-1738 states that "SFP fires could 

have health effects comparable to those of a severe reactor accident.... Large seismic events that fail the SFP are the dominant contributor [to causing an SFP fire].” Please confirm the NRC position regarding this statement. 

The Commission stands by the finding that the offsite healfuel pool zirconium fire as evaluated in NUREG-1738 can those from a severe accident at an operating reactor as evNUREG-1150, “Severe Accident Risks: An Assessment foPower Plants.” More recent studies, such as NUREG-193Reactor Consequence Analyses” (SOARCA) Report, and N“Consequence Study of a Beyond-Design-Basis EarthquakFuel Pool for a U.S. Mark I Boiling Water Reactor,” have cotypes of accidents examined at both reactors and spent fuweighted health impacts (i.e., calculated risk) would be low Commission’s safety goals. In the unlikely event of an accident at either a reactor or spprobability-weighted offsite economic impacts could be combecause while the economic impacts of a SFP fire could bfuel pool accident is expected to be less likely than a reactthe development of NUREG-2161, the NRC staff did not fininformation to challenge the view expressed in NUREG-17that large seismic events are the largest contributor to the large radiological release from the spent fuel pool. 

20 [NUREG-1738] "Further, the analysis 

indicates that timely evacuation, implemented through either pre-planned or ad hoc measures, can significantly reduce the number of early fatalities due to a zirconium fire." Please confirm the NRC position regarding this statement. 

The Commission stands by the finding that an early evacudefined in NUREG-1738 as an evacuation that is initiated athe spent fuel pool release), is effective at reducing the nufatalities, because early fatalities arise from very high acutexposures. Large acute radiation exposures can be signifeliminated altogether, by an evacuation that is completed bpool release begins. As stated in this study, the effect of timely evacuation is thimplemented through pre-planned evacuation or whether it through effective ad-hoc measures. NUREG-1738 states trisk in conjunction with differences in dominant sequences reactors, results in a small change in risk at a decommissioplanning is relaxed.“ 
- 30 - 

No. Question Response

The NRC staff continues to believe that, in the unlikely evebeyond design-basis accident resulting in a loss of sufficieair cooling within the pool, there is sufficient time for offsite protective measures under a comprehensive emergency mhazards) plan to protect the health and safety of the public

21 [NUREG-1738] "[T]he long-term 

consequences of an SFP fire may be significant. These long-term consequences (and risk) decrease very slowly because Cesium-137 has a half-life of approximately 30 years." Please confirm the NRC position regarding this statement.

The Commission stands by the finding, as stated in NUREas a zirconium fire is possible, the long-term consequencefire may be significant. NUREG-1738 illustrates the changconsequences from a zirconium fire as a function of time sdemonstrates that latent fatality risks and long-term collectwhich are more sensitive to the inventory of the longer-livelife of two years) and cesium-137 (half-life of 30 years), droHowever, the NRC also considers calculated risk in its conthe probability of a release multiplied by the radiological corelease (i.e., probability-weighted). This consideration is ilfinding in NUREG-1738 that “the risk at decommissioning within the Commission's safety goals. The risk is low becalikelihood of a zirconium fire even though the consequencefire could be serious.” Although the staff cannot completely rule out large radiologspent fuel pool, the staff’s analysis shows that the probabilspent fuel pool at a plant in decommissioning decreases wtime due to the drop in decay heat from the spent fuel storthe fact that there would be no further additions to the invefuel because the plant is no longer operating. 

22 [NUREG-1738] "[T]he consequences 

from a zirconium fire could be serious." 

Please confirm the NRC position regarding this statement. 

This statement is correct. Over several decades of researconsistently found that spent fuel pool fires are very high-clow probability events. 
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No. Question Response

23 [NUREG-1738] "Insurance, security, and 

emergency planning requirement revisions need to be considered in light of other policy considerations, because a criterion of "sufficient cooling to preclude a fire" cannot be satisfied on a generic basis." 

Please confirm the NRC position regarding this statement. 

The Commission stands by the finding in NUREG-1738, wand notes that further staff review of safeguards provisions plants has occurred since that document was issued. The number of additional actions related to this finding. On June 4, 2001, the NRC staff submitted a policy paper “to Safeguards, Insurance and Emergency Preparedness RDecommissioning Nuclear Power Plants Storing Fuel in SpSECY-01-100, to the Commission for consideration. The pCommission with the staff’s assessment of the policy impliNUREG-1738 study related to decommissioning exemptioand safeguards. The paper also recommended the NRC iCommission policy direction in response to the paper in a fdecommissioning rulemaking. Although SECY-01-0100 did not result in a rulemaking regand safeguards requirements for decommissioning plants, assessment of the findings of NUREG-1738 at that time. Tusing this assessment, as well as additional information gasubsequent studies and Commission policy decisions, to insite-specific license amendment and exemption requests fdown power reactors on a case-by-case basis. On January 10, 2014, the staff issued for public comment a Guidance document, “Draft Interim Staff Guidance on EmeExemption Requests for Decommissioning Nuclear Power ISG-02. The guidance will assist NRC staff in processing exemption from EP requirements for nuclear power reactoundergoing decommissioning. It considers historical expewith previously issued exemptions and a number of related NUREG-1738, SECY-01-0100, and SECY-13-0112. The NRC staff is also considering insights from NUREG-170100 in conjunction with Commission policy decisions madSECY-93-127, “Financial Protection Required of Licensees 
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No. Question Response

Power Plants During Decommissioning,” and SECY-04-01Requests to Reduce Liability Insurance Coverage for DecoReactors after Transfer of All Spent Fuel From a Spent FuStorage,” in its review of financial protection and insurance from decommissioning licensees. In response to SECY-93-127, the Commission approved reamount of financial protection and insurance required of dereactor licensees. In response to SECY-04-0176, the Comthe denial of exemption requests from decommissioning rerequested additional reductions of insurance requirements spent fuel from the spent fuel pool to a dry cask storage. Iexemption requests would have lowered the level of liabilitbelow the minimum levels previously established by the Co

24 [NUREG-1738] Figure 2.1 shows that 

even after a reactor had been shut down for 1 year, it would take only about 3 hours for PWR reactor fuel to heat to 900 degrees Celsius and only about 7 hours for BWR reactor fuel to heat to 900 degrees Celsius, even when the spent fuel pool accident does not prevent the assemblies from being air cooled. Please confirm the NRC position regarding this statement.

The Commission stands by the analysis in NUREG-1738, the hypothetical assumptions upon which the analysis is bintended purpose (e.g., exemption requests from NRC reqemergency preparedness for decommissioning reactors), tintroduced conservative assumptions into the analysis. Thinclude simplified treatment of the thermal-hydraulic respofuel and temperature change) and the use of assumed and configurations that do not allow for thermal radiation betwebundles and low power bundles (as stated in NUREG-1738spent fuel assemblies to the spent fuel pool wall liner. In a more realistic calculation, as demonstrated in the receStudy (NUREG-2161), thermal radiation heat transfer (in acan play a significant role. For example, it could take more the fuel to heat up to 900°C even one month after being mto the pool if the assemblies most recently removed from thottest) are distributed among older, cooler, fuel assemblieNUREG-1738 makes simplifying assumptions regarding thrate that results in instantaneous drain down of the pool. IStudy, even for a moderate leak scenario, it took more thawater level to reach the top of the fuel. 
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No. Question Response

In summary, the analyses performed in NUREG-1738 werintentionally conservative assumptions, and resulted in conconsequences, which was an appropriate regulatory approunder consideration. More recent and realistic analyses shconsequences of these accident scenarios are not as seveNUREG-1738 and that they take place over a much longer 

25 [NUREG-1738] Figure 2.2 shows that for 

PWR reactor fuel that is subject to a spent fuel pool accident that does not allow for air cooling to occur (the so-called adiabatic case in which the pool would only partially drain and thus preclude air circulation), it would take only 6 hours for the fuel to heat up even one year after the reactor shuts down. Please confirm the NRC position regarding this statement.

The Commission stands by the analysis in NUREG-1738, the hypothetical assumptions upon which the analysis is bstated above, there are conservatisms associated with the document. In an adiabatic calculation, both the oxidation energy and rare not taken into account. While the oxidation energy tenfuel temperature, thermal radiation would limit the fuel heapartial drain down cases, the blocked airflow can limit the mopposed to steam) oxidation reaction, while thermal radiatthe temperature and would play an important role in limitinThe Spent Fuel Pool Study (NUREG-2161), which – unlike addressed these effects, showed that for small leak scenaflow at 107 days after shutdown for a specific site, it would hours to increase the fuel temperature to 650°C. For an athe actual time for the hottest fuel assemblies to reach 900specific calculation. 

26 What are the risks for a spent fuel pool 

accident compared to that for an operating power reactor? 

While a SFP would normally contain more fuel assemblies present in a single operating reactor core, without the divintemperatures and pressures present in an operating reactopotential accident scenarios which could lead to a significarelease that exceed the U.S. Environmental Protection AgeProtective Action Guidelines (PAGs). As the spent fuel aggenerate heat diminishes due to the decay of fission produunlike operating reactor accident sequences that could learelease, accident scenarios at decommissioning plant SFPwould provide adequate time to initiate mitigation or protec
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No. Question Response

27 A document entitled, “Environmental 

Impacts of Storing Nuclear Fuel and HighLevel Waste from Commercial Nuclear Reactors: A Critique of NRC’s Waste Confidence Decision and Environmental Impact Determination,” authored by Dr. Gordon Thompson, describes SFP fires that could start in as little as 2 or 3 hours after cooling is lost to uncovered fuel. This appears inconsistent with criteria used by the NRC. Please clarify. 

Based on the results of various studies, the NRC believes a zirconium fire in the SFP is very low and timely implemenmitigation measures are available to address initiating conto a zirconium fire in the SFP. In the unlikely event that micould not be implemented in a timely manner to preclude t

zirconium fire in the SFP, the staff believes that sufficient tthe implementation of offsite protective actions using compemergency management plans (CEMPs) to provide for the the public. Three such studies are: NUREG/CR-6451, “A Safety and RAssessment of Generic BWR and PWR Permanently ShutPlants” (ADAMS Accession No. ML082260098); NUREG-2Study of a Beyond-Design-Basis Earthquake Affecting the U.S. Mark I Boiling Water Reactor” (ADAMS Accession Noand NUREG-1738, “Technical Study of Spent Fuel AccideDecommissioning Nuclear Power Plants” (ADAMS AccessML010430066). Regarding a fire that could start in as little as 2 or 3 hours, that this could be true for spent fuel 1 to 3 months after remreactor. However, decommissioning reactors, with fuel in tolder, would take much longer to heat up. The exemption include site-specific analyses demonstrating that at least 1from the time there is a total loss of all cooling to the fuel ureaches 900 degrees Celsius. 

28 What basis is there for assuming 

reasonable assurance that there will be time for preventative and mitigation actions? 

The risk associated with a zirconium fire event is directly refrom the fuel, and therefore, the time since shutdown. NUconservatively estimated that greater than 100 hours would SFPs lowered to within 3 feet of the top of the fuel for loss when pressurized water reactor (PWR) fuel has decayed aThe NRC currently requires, and will continue to require fopower reactors, that the equipment, procedures, and trainenecessary be designated in the exemption request to prov
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No. Question Response

fuel cooling in the unlikely event of a SFP drain down. An the licensee’s retention of the license condition to maintain 10 CFR 50.54(hh)(2) pertaining to SFP mitigation actions. Additionally, the NRC has studied evacuations nationwide responders are capable of implementing protective actions study conducted by Sandia National Laboratories, NUREG“Identification and Analysis of Factors Affecting Emergency (ADAMS Accession No. ML050250245), concluded that “esuccessfully protect public health and safety over a broad circumstances and challenges” and “large-scale evacuatioStates, whether preplanned or ad hoc, are very effective alives and reduce the potential number of injuries associateThe SFP fire scenario would be a relatively slow developinto incidents many local responders routinely respond to. It recognize that the regulations will still require the licensee its onsite radiological emergency response plans and that governments continue to maintain the responsibility of protevent of emergencies and, if needed, would implement prothe CEMP process. 

29 How has the NRC demonstrated that a 

licensee of a decommissioning reactor would be able to mitigate the potential consequences of an accident at a spent nuclear fuel pool within ten hours if the accident was caused by a severe, unexpected initiating event, such as a massive and devastating earthquake or a terrorist attack? 

Decommissioning power reactor licensees are required to effective emergency plans that meet the requirements of 1Appendix E to 10 CFR Part 50. Licensees that have procedecommissioning an operating reactor have submitted reqthese regulatory requirements. The NRC staff evaluates site-specific analyses supporting 

request for exemptions from emergency plan requirementsbe used to provide the NRC reasonable assurance that in exemption: (1) an offsite radiological release would not excEnvironmental Protection Agency protective action guides for design basis accidents applicable to shutdown reactorstime would exist to initiate appropriate mitigating actions or actions, if needed, to protect public health and safety in the 
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No. Question Response

severe beyond design-basis accident resulting in loss of supool. In normal operations, the spent fuel pool is filled with watefuel cool. The staff has generally determined that 10 hours minimum time available to implement mitigation actions, anprotective offsite measures using a State and local governcomprehensive emergency management plan, if the water the pool and jeopardize the ability to keep it sufficiently cooHowever, in a hypothetical spent fuel pool accident scenarexpected amount of time it would take for water to drain frobeyond design-basis accident that results in the water drai(whether a full or partial drain-down) would likely take muchours because of the robust construction of the spent fuel volume of water in the pool. Furthermore, particularly for oand other heat removal mechanisms following loss of coolsufficient to keep the fuel cool indefinitely or would significaheat-up time. To be conservative, the exemption analysis and 10-hour cthe potential consequences pool do not credit natural air ccooling in the spent fuel pool after the event. Instead, it is aimmediately begins to heat up without any natural removal NRC staff accepts this simplified approach and time estimaregulatory decisions whether to grant the exemption and if exemption becomes effective (e.g., typically 12-24 months The NRC staff reviews the analyses provided by the licensminimum of 10 hours is still available to restore cooling, or protective measures, before the fuel slowly heats up to a tecause a zirconium fire. A licensee may in part rely upon thcapabilities it had for its operating reactor as required in 10 mitigate the potential consequences of a severe accident a
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No. Question Response

In addition to normal plant structures, systems and compowater levels and cooling to the spent fuel, there are redundindependently powered equipment to perform these functiooperating reactors. Decommissioning licensees may choothese capabilities to provide assurance they can implemenwithin 10 hours, or may commit to another strategy that proassurance. 

30 What is the purpose of the interim staff 

guidance on emergency planning exemptions issued by the NRC? 

The NRC has developed draft interim staff guidance (ISG) Planning Exemption Requests for Decommissioning Nucleaddress licensee exemption requests from certain NRC em(EP) requirements. The draft ISG provides a technical discoverview of existing guidance for reviewing emergency pla

requests, as discussed above. In developing this guidance, the NRC relied upon its previoexperience. Additionally, the guidance was informed by exspent fuel pool fires and risks. The staff is planning to publfollowing incorporation of public comments, as appropriategained from its review of current exemption requests. 

31 The NRC’s ISG on emergency planning 

would allow exemption from offsite emergency planning requirements for decommissioning reactors 12 months after permanent shutdown. This is inconsistent with previous NRC staff statements that EP exemptions are evaluated on their technical merits on a case-by-case basis. Please clarify. 

The staff has revised the proposed ISG to clearly state tharequest requires a site-specific analysis to define the date could be permitted. The required analysis must show that available from the time that both water and air cooling is lountil the hottest fuel assembly reaches 900°C to take mitigrequired, to take protective actions using a CEMP approacThe analysis contains the conservative assumption that thimmediately begins to heat up without any heat from the futo the surroundings. In reality, there would be heat loss to the fuel, to any remaining water in the pool or on fuel surfacooler fuel assemblies and to the SFP walls. As discussed “Consequence Study of a Beyond-Design-Basis EarthquakFuel Pool for a U.S. Mark I Boiling Water Reactor,” radiativaddition to air cooling) can play a significant role. For exam
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No. Question Response

more than 10 hours for boiling water reactor fuel to heat up month of being removed from the reactor, if the most recendistributed among older, cooler fuel assemblies in the SFP

32 The NRC’s draft ISG states that “after a 

certain amount of time, the overall risk of a zirconium fire becomes insignificant.” Why is there reasonable assurance of an increased probability that the fuel is coolable? 

Particularly for older fuel, air cooling and other heat removfollowing loss of cooling water may be sufficient to keep thor significantly extend the fuel heat up time. Although the completely rule out the possibility of a radiological release probability of a release from a SFP at a decommissioning decreases with the passage of time due to the drop in decstored in the SFP, and the fact that there would be no furthinventory of fresh spent fuel into the SFP, as the plant is nMore recent analyses have been performed with site-specdetermine whether a release from a SFP could occur, conspractice regarding the physical arrangement of the fuel in tFor example, SECY-13-0112 demonstrated that a release occur at the operating power reactor site studied for at leasa large, beyond design-basis seismic event that occurs moshutdown. In this study, the reference plant SFP contained removed from the reactor, which would be more susceptibrelease scenario than would a SFP at a decommissioning NRC studies also show that fuel that has aged sufficiently or water spray. The requirement for emergency response based upon the likelihood of an accident, but rather upon dprinciples that are part of all NRC regulatory actions. Emeis a defense-in-depth measure required to mitigate the conunlikely accidents, regardless of the probability of those acdepth is provided by mitigation procedures that provide coocapability to the SFP should there be a loss of water inven
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No. Question Response

33 On what specific basis (and using what 

specific analysis) has the NRC concluded that it would be possible, within ten hours after the initiating event, to: a) repair all key safety equipment; b) repair what could be a large breach in the spent fuel pool; and c) ensure the capability to continually refill the spent fuel pool with fresh water as it drains in the midst of a broader response to or consequences of a severe seismic event (or other natural hazard) or on-going terrorist attack? 

The NRC staff has determined that an accident caused by results in the water draining from the pool (whether a full ois too unlikely to have been considered in the initial plant dbeyond-design-basis accident). Such a severe event wouthan 10 hours for mitigation or protective measures becausconstruction of the spent fuel pool and other features of the intended to prevent a rapid loss of water. Furthermore, pafuel, air cooling and other heat removal mechanisms followwater may be sufficient to significantly extend the fuel heatfuel cool indefinitely. Although the staff cannot completely rule out large radiologspent fuel pool, the staff believes that the probability of a refuel pool at a plant in decommissioning decreases with the to the drop in decay heat from the spent fuel stored in the because the plant is no longer operating there would be no the inventory of fresh spent fuel. More recent analyses have been performed with site-specdetermine whether a release from a spent fuel pool could ostudied, considering that site’s practice regarding the physthe fuel in the spent fuel pool. In particular, the study transCommission with NUREG-2161, “Consequence Study of a Basis Earthquake Affecting the Spent Fuel Pool for a U.S. Reactor,” provided consequence estimates of a hypotheticaccident initiated by a low-likelihood seismic event at a refon the Peach Bottom BWR Mark I spent fuel pool. The study demonstrated that a release is not expected to ostudied for at least 72 hours following a large, beyond desievent that occurs more than 60 days after shutdown. The that it is, therefore, likely that there would be more than 10 potential consequences of a severe accident at a spent fueSFP water loss. 
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No. Question Response

For evaluating exemption requests from NRC requirementemergency preparedness for decommissioning reactors, thintroduced conservative assumptions into the analysis. Spanalysis and 10-hour criterion for mitigating the potential cbeyond design-basis accident at a spent fuel pool does nocooling and water cooling in the spent fuel pool after the evsimplification. Rather, the analysis assumes that the fuel iheat-up without any natural removal of its energy. These conservatisms include simplified treatment of the thresponse and the use of assumed and often bounding connot allow for thermal radiation between high power bundles bundles (as stated in NUREG-1738) and also from the spethe spent fuel pool wall liner. In a more realistic calculationthe recent Spent Fuel Pool Study (SECY-13-0112), thermatransfer (in addition to air cooling) can play a significant rolFor example, it could take more than 10 hours for the fuel only one month after being moved from the reactor to the pmost recently removed from the reactor are distributed amassemblies. The NRC staff uses the simplified approach to determine tpart of its regulatory decision whether to grant the exemptiwhen a granted exemption would become effective (i.e., afcooled for a certain period of time, typically 12-24 months aNRC staff reviews the analyses provided by the licensee to minimum of 10 hours is still available to restore cooling or protective measures, once it is assumed that all cooling is fuel heats up to a temperature where rapid oxidation of the occur, commonly referred to as a zirconium fire. The capabilities specified in 10 CFR 50.54(hh)(2), put in plSeptember 11, 2001, terrorist attacks, are one means by wprovide the capability to mitigate the potential consequenc
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No. Question Response

accident at a spent fuel pool. Regulatory standards under provide that, in addition to normal plant structures, systemmaintain water levels and cooling to spent fuel, redundant powered equipment needed to perform these functions are Decommissioning licensees requesting exemptions from epreparedness requirements may choose to retain these syanother strategy that provides similar assurance. The 10-hour time frame is not intended to be the time it wokey safety systems or to repair a large breach in the spent considering the very low probability of beyond-design-basispent fuel pool, in the staff’s professional judgment, 10 houreasonable time period to implement these pre-planned mprovide makeup or spray to the spent fuel pool before the ocladding ignition, or, if necessary, to initiate offsite protectivrepair of required safety systems or repair of a large breacpool would be assessed and performed by additional staffiresponding to the site as part of the licensee’s on-site emeensure long-term cooling of the spent fuel. 

34 What initiating event could be expected to 

result in an accident scenario resulting in a loss of SFP water inventory? 

The risk of a SFP accident is dominated by a beyond desigevent, which could challenge SFP liner integrity. Assuming configuration and heat decay times, spent fuel assemblies water inventory is lost. NUREG-2161, “Consequence Study of a Beyond-Design-BAffecting the Spent Fuel Pool for a U.S. Mark I Boiling Watthe likelihood of a radiological release from the spent fuel aearthquake at the reference plant (GE BWR-4 / Mark I Conabout one time in 10 million years. The study considered a severe earthquake with ground mothe maximum earthquake reasonably expected to occur, wexpects is more challenging for a spent fuel pool structure experienced at the Fukushima Daiichi Nuclear Power Planearthquake that occurred off the coast of Japan on March 
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No. Question Response

35 What other scenarios were considered for 

spent fuel pool accidents, and what were the results of the associated analyses? 

Specific to SONGS, the licensee provided analyses that alfollowing SFP scenarios: • Loss of all heat removal capability – Indicated that as o

2013, the time to boil is approximately 90 hours and th11 days available to restore cooling to the SFP before reaches 10 feet above the top of the fuel. 

• Total loss of cooling water inventory with air cooling – I

August 31, 2014, the fuel cladding will not exceed 565°

• Total loss of cooling water inventory without air cooling 

10 hour threshold would be reached after 31 months otime of permanent shutdown (or as of August 2013). 

• Shine from an empty SFP – Indicated that as of June 1

from the spent fuel assemblies is well below the accep100 mrem for a two hour period to a member of the pu

• Fuel handling accident – Indicated that the dose analys

handling accident in the Fuel Handling Building is well PAGs. 

Similar analyses were provided by the other decommission

36 Why does the NRC’s evaluation of EP 

exemption requests apparently disregard the risk associated with terrorist attacks and hostile actions? 

The licensee’s Security Plan must provide high assurance protection of the SFP. This is the same level of protection during reactor operations. In cases where exemptions are decommissioning reactors, the exemption eliminates the daction” and its related requirements, but there are elementevents that are maintained. Specifically, the classification events, notification of offsite authorities, and coordination wunder a CEMP approach are still required. The NRC’s letter to Congress regarding the National Acad“Study on the Security and Safety of Commercial Spent NuAccession No. ML050280428), stated in part, that “today, sprotected than ever. The results of security assessments clearly show that storage of spent fuel in both SFPs and in 
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No. Question Response

provides reasonable assurance that public health and safeand the common defense and security will be adequately pdesign basis threat provides reasonable hypothetical threasabotage, to which licensees must be able to respond. In order to assure that this threat statement remains a valid of physical protection systems, the staff routinely reviews aof intelligence information. Every 12 months the staff asseenvironment for that 12-month period and formally providethe Commission in a report. If significant information were into question the adequacy of the design basis threat statewould immediately notify the Commission. The NRC staff engages with the U.S. intelligence community so that it can respond to credible threats to licensees. Requirements for identifying anticipated threats, maintaininaddress those threats, protecting onsite workers, and recoprotect surrounding communities are integral to the regulatthreats that licensees must be capable of responding to alsphenomena and security events. In the event of a credible and secure operation of the facility, the licensee is required Operations Center, which is staffed continuously. The NRsituation and the licensee’s response to the threat. If condNRC will activate its Incident Response Center and begin other Federal and State agencies. 

37 Will the NRC continue to inspect 

emergency preparedness activities during decommissioning? 

After certification of permanent cessation of operation is refacility is no longer subject to Reactor Oversight Process (inspection activities will be continued under Inspection Ma“Decommissioning Power Reactor Inspection Program.” Tprogram is comprised of two major elements: (1) core Inspdiscretionary Inspection (i.e., reactive and initiative inspectinspection program will remain in effect until the license is In addition, Inspection Procedure (IP) 82501, “DecommissPreparedness Program Evaluation,” and IP 82401, “Decom
- 44 - 

No. Question Response

Emergency Preparedness Scenario Review and Exercise created to specifically address inspection activities in the EFinally, an NRC resident Inspector will typically remain ons12 months after the 50.82 certification is submitted to the Npresent for significant decommissioning activities including 
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Physical Security During Decommissioning

1. Why are decommissioning reactors no longer required to maintain the same physical 

security requirements they had when they were operating? 

Answer

If a licensee submits an exemption request from the physical security requirements for a 

power reactor, a detailed technical review of the request will be performed to ensure the 

exemption(s) are authorized by law, will not present an undue risk to the health and safety of 

the public, are consistent with the common defense and security; and special 

circumstances, as defined in the NRC’s regulations, are present. 

Decommissioning nuclear power plants that have ceased operations and no longer have 

fuel in the reactor core often request exemptions to certain physical security requirements 

that were intended for operating power plants. However, as long as there is fuel onsite, a 

decommissioning power plant must continue to maintain a physical protection program that 

provides high assurance that the activities involving special nuclear material are not inimical 

to the common defense and security and do not constitute an unreasonable risk to the 

public health and safety. The licensee must continue to demonstrate a security strategy to 

defend against spent fuel pool sabotage scenarios governed by the design basis threat.

2. How does the NRC provide oversight of the security program at decommissioning 

nuclear power plants? 

Answer

The NRC provides oversight of licensee security programs at decommissioning nuclear 

power plants through a security inspection program that verifies compliance with applicable 

regulatory requirements. The security inspection program examines licensee activities in 

order to assess performance and to assure that the licensee’s overall security program is 

meeting the objective of providing high assurance of protection against the design basis 

threat. The following attributes of a licensee’s security program are inspected for 

decommissioning nuclear power plants: (1) access authorization; (2) access control; 

(3) equipment performance, testing, and maintenance; (4) protective strategy evaluation; 

(5) protection of safeguards information; (6) security training; and (7) target sets. 

3. What does the term “high assurance” mean for the security of a decommissioning 

nuclear power plant as compared to “reasonable assurance”? 

 

Answer

The term “high assurance” in regard to the security of a decommissioning nuclear power 

plant means that the objective of the physical protection program is to prevent a spent fuel 

sabotage event. Specifically, the program must ensure that the capabilities to detect, 

assess, interdict, and neutralize threats up to and including the design basis threat of 

radiological sabotage as stated in 10 CFR 73.1, are maintained at all times. The term 

“reasonable assurance” means that a defense-in-depth strategy should be used and would 

imply that various barriers to prevent or mitigate an event could fail (e.g., plant design, 

operator response/training, etc.). 
- 46 - 

4. Does the NRC participate in or observe any on-site security drills performed at 

decommissioning reactors to verify their security capabilities? 

Answer

Yes, the NRC staff participates in or observes onsite security drills performed by the 

licensee at decommissioning reactors as part of the NRC’s ongoing inspection activities to 

verify the site’s security capabilities. Specifically, the NRC’s inspection program verifies that 

the licensee continues to meet the applicable security regulations.

The NRC inspection program typically includes NRC observation of licensee-conducted 

force-on-force exercises. Licensee force-on-force exercises assess their ability to defend 

against the design basis threat for radiological sabotage and provide valuable insights that 

enable the NRC to evaluate the effectiveness of licensee security programs. They are an 

essential part of the oversight of the security of these facilities. 

NRC inspectors from Headquarters and the Regional offices monitor licensees’ security 

activities throughout the year. The inspectors provide firsthand, independent assessments 

of plant conditions and licensee performance, document their findings in writing, and 

conduct follow-up inspections to ensure that the licensee has made any necessary 

corrections.

5. Should there be a potential security threat at a decommissioning reactor, what NRC 

processes are in place to determine the nature of the potential threat, ensure the 

capability of each site to address any threats, and to protect the workers and the 

surrounding communities? 

Answer

Through its rulemaking, licensing, and inspection activities, the NRC evaluates the 

licensee’s plans and procedures for identifying and responding to anticipated threats to the 

safe operation of the facility. The NRC has established regulations and policies that contain 

design standards and programmatic requirements to address normal operations, anticipated 

external events, and design basis accidents. The NRC reviews the licensee’s plans for 

compliance with the regulations and policies through its licensing process. The NRC 

provides oversight of the licensee’s implementation of its license requirements through its 

inspection activities. 

The design basis threat provides reasonable hypothetical threats for radiological sabotage, 

to which licensees must be able to respond and repel. In order to assure that this threat 

statement remains a valid basis for the design of physical protection systems, the staff 

routinely reviews and analyzes a range of intelligence information. Every 12 months the 

staff assesses the threat environment for that period and formally provides its conclusions to 

the Commission in a report. If significant information were received that called into question 

the adequacy of the design basis threat statements, the staff would immediately notify the 

Commission. The NRC staff also continuously engages with the U.S. intelligence 

community so that credible threats can be immediately communicated to licensees. 

Requirements for identifying anticipated threats, maintaining preparedness to address those 

threats, protecting onsite workers, and recommending actions to protect surrounding 
- 47 - 

communities are integral to the regulations. The anticipated threats that licensees must be 

capable of responding to include natural phenomena and security events. 

In the event of a credible threat to safety and security of the facility, the licensee is required 

to notify the NRC Operations Center, which is staffed continuously. The NRC staff monitors 

the situation and the licensee’s response to the threat. If conditions warrant, the NRC will 

activate its Incident Response Center and begin coordinating with other Federal and State 

agencies. In addition, the NRC will notify licensees of potential threats to their facilities 

received from other Federal, State, or tribal sources so that licensees can prepare as 

appropriate. 

6. How will the spent fuel dry casks be secured once all of the fuel has been removed 

from the pool and placed on the ISFSI pad at decommissioning facilities? 

Answer

Once all of the fuel has been removed from the spent fuel pool and placed in storage casks 

in an Independent Spent Fuel Storage Installation (ISFSI) it will be protected by the 

requirements within the security orders issued following the terrorist attacks of September 

11, 2001, and the requirements in 10 CFR Part 72, “Licensing Requirements for the 

Independent Storage of Spent Nuclear Fuel, High-Level Radioactive Waste, And ReactorRelated Greater Than Class C Waste“ and 10 CFR 73.55, “Requirements for Physical 

Protection of Licensed Activities in Nuclear Power Reactors Against Radiological Sabotage”

as modified by § 72.212 Conditions of general license issued under § 72.210. 

The Code of Federal Regulations provides basic security requirements for ISFSIs, which 

include well-trained and armed security officers, physical barriers, intrusion detection and 

surveillance systems, vehicle barriers, and coordination with local response organizations. 

The NRC also issued security Orders requiring additional security measures for ISFSIs. In 

order to protect the information from individuals with a malevolent intent, the actual 

requirements in these security orders are designated at Safeguards Information (SGI) and 

are not publicly available. 

7. In late May 2014, the Commission released its votes that unanimously accepted the 

staff’s recommendations not to require decommissioning reactors to be subjected to 

NRC’s force-on-force exercises, which are designed to validate the facilities’ security 

capability. What is the basis for not requiring force-on-force exercises? 

Answer

The NRC determined that the force-on-force inspections for decommissioning power 

reactors are not warranted because the current security inspection program, which includes 

the observation of licensee-conducted force-on-force exercises, provides adequate 

oversight and verification of the security posture given a reduction in both risk and the 

number of target sets at decommissioning power reactors. 

8. According to the NRC inspection manual, decommissioning reactors with spent fuel 

still in the spent fuel pool receive 12-24 hours’ worth of security inspections each 

year. But operating reactors receive well more than 100 hours’ worth of security 

inspections. 
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Additionally, NUREG-1738 concluded that “SFP [spent fuel pool] fires could have 

health effects comparable to those of a severe reactor accident....Large seismic 

events that fail the SFP are the dominant contributor [to causing an SFP fire].” 

If the NRC believes that the consequences of an accident at a spent fuel pool could 

be equivalent to those of an accident at an operating reactor, why did the 

Commission assert in its votes that the validation of security measures in place at the 

spent fuel pool could be met via security inspections when the time spent on those 

inspections for decommissioning reactors is nowhere near equivalent to the time 

spent on security inspections for operating reactors? 

Answer

NRC’s current decommissioning power reactor inspection program was tailored to reflect an 

inspection protocol for decommissioning reactors with fuel that had aged for many years. 

After the events of September 11, 2001, the NRC recognized the need to revise the security 

requirements and inspection programs for new reactors, operating power reactors, and 

decommissioning reactors. The security inspection activities conducted by the NRC at each 

of these facilities differ based on the structures, systems, and components that require 

protection due to the facility’s operating status and risk profile. Thus, there are differences 

within the security inspection programs for new reactors, operating reactors, and 

decommissioning reactors. 

The NRC just completed revising the security inspection program for decommissioning 

reactors. The revision addressed issues associated with recent and future 

decommissioning of nuclear power reactors. The revised security inspection program for 

decommissioning nuclear power reactors will involve a total of 95 hours of security 

inspections per site, each year. These inspections are designed to ensure that licensees 

are effectively implementing their physical protection programs consistent with applicable 

requirements. Within the scope of this overall physical protection program evaluation, the 

staff will observe and assess licensee-conducted force-on-force exercises. 

9. How do the exemptions from security requirements relate to waste confidence? Does 

that document assume that all security requirements are waived? 

Answer

The environmental impacts of security exemptions are neither explicitly evaluated nor 

resolved generically in the Waste Confidence Generic Environmental Impact Statement 

(GEIS). The GEIS assumes the baseline condition that all licensees will comply with current 

NRC requirements for security at decommissioning nuclear power plants. 

The Waste Confidence analysis does not assume the granting of any security exemptions. 

The exemptions from NRC security requirements are evaluated on a case-by-case basis 

and are subject to separate, site-specific safety, security and environmental reviews. The 

analysis in the Waste Confidence GEIS has no bearing on the granting or denial of a 

security exemption. The granting of security exemptions does not reduce the security level 

at a site such that the probability of a successful attack is increased. Consequently, the 

granting of security exemptions prior to or during the continued storage period does not 

change the impact determinations in the Waste Confidence GEIS.