NUREG-1628

Staff Responses to 

Frequently Asked 

Questions Concerning 

Decommissioning of 

Nuclear Power Plants 

Final Report 

U.S. Nuclear Regulatory Commission 

Office of Nuclear Reactor Regulation 

Washington, DC 20555-0001
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NUREG-1628

Staff Responses to 

Frequently Asked 

Questions Concerning 

Decommissioning of 

Nuclear Power Plants 

Final Report 

Manuscript Completed: May 2000 

Date Published: June 2000 

Prepared by 

I.L. Minns, M.T. Masnik, NRC 

R. Harty, E.E. Hickey, PNNL 

Pacific Northwest National Laboratory 

Richland, WA 99352 

Division of Licensing Project Management 

Office of Nuclear Reactor Regulation 

U.S. Nuclear Regulatory Commission 

Washington, DC 20555-0001
ABSTRACT

This report, through a question-and-answer format, provides U.S. Nuclear Regulatory'Commission 

(NRC) staff responses to frequently asked questions on the decommissioning process for commercial, 

nuclear power reactors. The questions were taken from a variety of sources over the past several years, 

including written inquiries to the NRC and questions asked at public meetings and during informal 

discussions with the NRC staff. In responding to the questions, the NRC staff attempted to provide the 

answers in a clear and non-technical form. 

With the increase in the number of power reactors beginning the decommissioning process and 

significant changes that occurred in the regulations since 1996, the staff realized that there was a general 

lack of understanding of the decommissioning process and the risks associated with decommissioning. 

This document was developed in response to the staff's concerns. The 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, low-level 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 also provides responses to questions related to public involvement in decommissioning as well 

as providing the public with sources for obtaining additional information on decommissioning.

iii
21
Contents 

AB STRACT ................................................................. 

ABBREVIATIONS ...................................................... xiii 

ACKNOWLEDGMENTS ........................................................... xv 

GENERAL ........ .................................................. 1 

1.1 How is decommissioning defined? ................. 1 

1.2 Why do nuclear power plants shut down permanently? ......................... 

1.3 Why are power reactors decommissioned? .................................... 1 

1.4 How does decommissioning proceed? ........................................ 1 

1.5 What are the benefits of decommissioning? .................................... 2 

1.6 What are the costs of decommissioning? ...................................... 2 

1.7 What are the options to decommissioning? .................................... 2 

2 DECOMMISSIONING PROCESS .......................................... 2 

2.1 What terms or definitions are important to the understanding of decommissioning? 2 

2.2 What is the difference between radioactive contamination and activation products, and where are contaminated materials and activated materials located? .............. 4 

2.3 How is a nuclear power plant decommissioned? .4................ ........ 4 

2.4 Who decides how a facility should be decommissioned? ....................... 4 

2.5 Is there a limit on the number of years that it would take to decommission a plant? ..... 5 

2.6 What alternatives are currently used for decommissioning? ....................... 5 

2.7 What are the benefits and costs of the DECON alternative? ....................... 6 

2.8 What are the benefits and costs of the SAFSTOR alternative? ..................... 6 

2.9 What are the benefits and costs of the ENTOMB alternative? .................... 7 

2.10 Is the choice of decommissioning alternatives a decision that is left entirely to the 

licensee, or does the NRC help make this decision? ............................. 7 

2.11 Must a licensee choose either DECON or SAFSTOR, or can it combine the two 

alternatives? .................................................... 7 

2.12 What main factors affect a licensee's choice of a decommissioning alternative? ....... 8 

2.13 How long does the dismantlement phase last?.................................. 8 

3 DECOMMISSIONED SITES .................................................... 8 

3.1 What NRC-regulated plants have been or are being decommissioned, and what 

decommissioning alternatives have been or are being used?....................... 8 

3.2 Have other non-NRC-regulated nuclear facilities been decommissioned? ............ 9 

3.3 What improvements have been made as a result of previous decommissioning 

experience? .................. ........................................... 9 

3.4 What research is being performed to find improved methods to be used during 

decommissioning? ....................... . ....................... 10 

3.5 What differences are there in decommissioning between different types of reactor 

designs? ...................................................... 10

v
4 NRC ACTIVITIES ............................................................ 11 

4.1 Licensing .............................................................. II 

4.1.1 Does the licensee have an NRC license even during the decommissioning 

process? ......................................................... 11 

4.1.2 Is the license the same kind that is used for an operating plant? .............. 11 

4.1.3 Could the NRC require a plant to cease operations and begin 

decomm issioning? ................................................. 11 

4.1.4 What would happen if a licensee refuses to decommission a plant that has 

ceased operations? .......................................... 11 

4.1.5 What would happen if the licensee's license expires before the decommissioning 

process is concluded? .............................................. 12 

4.1.6 Are there any restrictions as to what the licensee can do with the site after 

decommissioning is completed, and after the NRC has terminated the license? 

Would the NRC review these activities? ................................ .12 

4.1.7 How would license renewal of operating plants at multi-unit facilities affect the 

decommissioning? Would the licensee be allowed to delay the decommissioning 

of the plant even longer? ............................................ 12 

4.2 Regulations .............................................. .............. 12 

4.2.1 Who regulates the decommissioning process? ......................... 13 

4.2.2 Does the U.S. Environmental Protection Agency (EPA) have authority over the 

decommissioning process? ....... .............................. 13 

4.2.3 When were the current regulations for decommissioning written? ............ 13 

4.2.4 What publications contain the regulations for decommissioning? ............ 13 

4.2.5 What regulatory actions are required to decommission a nuclear facility? ...... 14 

4.2.6 Why isn't an environmental impact statement (EIS), or environmental assessment 

(EA) required at the start of the decommissioning process? ................. 16 

4.2.7 What activities can take place before submitting the PSDAR?.............. ; 17 

4.2.8 Can the licensee make changes to its plans or commitments after submitting the 

PSDAR? ........................................................ 17 

4.2.9 What would happen if the licensee wanted to switch from SAFSTOR to 

DECON? ................................................ 17 

4.2.10 Will the PSDAR or the License Termination Plan be reviewed by the 

U.S. Environmental Protection Agency (EPA)? ........................... 18 

4.2.11 Under what circumstances would the NRC refuse to allow a licensee to 

proceed with the decommissioqing? ................................... 18 

4.2.12 Would the NRC ever stop a licensee from decommissioning a plant on the 

basis of information in the Post-Shutdown Decommissioning Activities 

Report? ......................................................... 8 

4.2.13 Can the NRC require the licensee to follow the regulations, and how would the 

NRC know if the regulations were not being followed? .................... 18 

4.2.14 Does the NRC have to review and either approve or disapprove every decision 

the licensee makes related to decommissioning a facility'?................... 19 

4.2.15 How can the NRC make sure a licensee does not make a mistake in its 

evaluation of proposedzhanes, tests, or experiments? What happens if the 

licensee fails to identify an unreviewed safety question? ................... 19 

4.2.16 What would happen if the licensee decides to do something that will endanger 

the environment during the decommissioning process? .................... 20

vi
4.2.17 What were the previous regulations and why were they changed? ........... 20 

4.3 Inspection Program .................... ........................... 21 

4.3.1 What are the goals of the inspection program at nuclear power plants 

undergoing decommissioning? .................................. 21 

4.3.2 What types of inspections are conducted? ........................... 21 

4.3.3 Which areas receive the greatest emphasis during the inspections? ........... 22 

4.3.4 Are there some decommissioning activities that receive greater oversight 

than others. ................................... ............ 22 

4.3.5 Does the NRC independently monitor radioactive releases into the 

environm ent? ..................................................... 22 

4.3.6 Will the NRC have inspectors onsite during decommissioning? ............. 22 

4.3.7 Are inspection reports associated with the decommissioning activities available 

on the NRC's website? ....................................... 23 

4.3.8 Will plant security levels be de-emphasized during decommissioning? ........ 23 

5 SPENT FUEL ....................................................... 23 

5.1 W hat are "high-level wastes?" ....... ....................................... 23 

5.2 What is meant by the term "spent fuel?" ................................. 23 

5.3 Are there facilities or plans for facilities for the disposal of high-level radioactive 

waste? ................................. ...... 23 

5.4 Is the licensee allowed to store the spent fuel in the reactor vessel? ................. 24 

5.5 Since there are no facilities for permanent disposal, where will the spent nuclear fuel 

be kept during the decommissioning process? .................................. 24 

5.6 What are the long-range plans for disposition of spent fuel? .................... 24 

5.7 What happens if a disposal site for high-level waste is never licensed?............... 24 

5.8 Spent Fuel Pools ................................................ 24 

5.8.1 Why is spent fuel stored in a pool of water? ............................. 24 

5.8.2 Has the spent fuel pool been analyzed to determine the limits for heat removal 

due to spent fuel storage? ........................................... 25 

5.8.3 Do spent fuel pools leak, and if they do, how much radioactive material could be 

leaked, and where would it go? ................... ................... 25 

5.8.4 What would happen if there were a loss of heat-removal capability or water in a 

spent fuel pool when it was fully loaded? .............................. 25 

5.8.5 What would happen to the fuel in the spent fuel pool if there was an earthquake 

that ruptured the pool, or if an airjlane crashed into the pool? ............... 26 

5.8.6 What can be done to prevent the spent fuel pool from boiling dry? ........... 26 

5.8.7 How long can the licensee store the spent fuel in the spent fuel pool? ......... 26 

5.8.8 What will be done with the spent fuel pool after the fuel has been removed? ... 27 

5.8.9 Could the licensee transfer fuel to another licensee's facility? ............... 27 

5.8.10 Can the spent fuel be shipped to another facility's spent fuel pool for storage?.. 27 

5.8.11 Where can I find the regulations relating to the storage of spent fuel in a spent 

fuel pool? ........................................................ 27 

5.9 Independent Spent Fuel Storage Installation ................................... 27 

5.9.1 What is an independent spent fuel storage installation (ISFSI)? .............. 27 

5.9.2 Why would a licensee store spent fuel in an ISFSI rather than in the spent fuel 

pool? .......................... ........................ 27 

5.9.3 What is a dry-storage cask, and how does it keep the fuel from melting or from

vii
causing a nuclear reaction (criticality)? ................................. 28 

5.9.4 Who is responsible for reviewing proposed cask designs to ensure that they will 

safely confine the fuel, and what types of evaluations are required?.......... 28 

5.9.5 Are there any nuclear plants that already use dry-storage casks in an independent 

spent-fuel storage installation (ISFSI)? ................................. 28 

5.9.6 Can casks be used for both shipping and storage - or will they need to transfer 

the fuel from one cask to another at some later date? .................... 29 

5.9.7 How long may the licensee keep the spent fuel in an onsite ISFSI? ........... 29 

5.9.8 Can the spent fuel be shipped to another facility's ISFSI for storage? ......... 29 

5.9.9 Why is the licensing process on the independent spent-fuel storage installation 

(ISFSI) evaluated separately from the decommissioning process?............ 29 

5.9.10 Where are the regulations relating to use of ISFSIs? .................... 29 

5.9.11 Can the licensee use the site for other activities, such as a combustion turbine or 

natural gas plant, while fuel is stored onsite in an ISFSI? ................... 30 

6 RADIOACTIVE LOW-LEVEL WASTE ................. ............. ....... 30 

6.1 What is meant by low-level radioactive waste and how is it different from fuel? ....... 30 

6.2 How is the low-level radioactive waste disposed of? ............................. 31 

6.3 Is low-level waste disposal safe? . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 31 

6.4 Where can low-level radioactive waste be disposed of? .......................... 31 

6.5 What would happen if the waste site that was being used is closed? ................ 32 

6.6 Can the low-level radioactive waste be stored at the site in the event that the waste 

site is closed? What type of facility is required and how long can the waste be left 

at the site? ...... ............................................... 32 

6.7 Can radioactive waste be buried onsite? ...... ........ ....... ............ 32 

6.8 Can the reactor vessel be disposed of at a low-level waste (LLW) site? Can the reactor 

vessel be shipped intact, or does it have to be segmented? ...................... 32 

6.9 How are liquid wastes disposed of? .......................................... 33 

6.10 Where are the regulations relating to radioactive low-level waste disposal? ........... 33 

7 TRANSPORTATION ............................................... 33 

7.1 How is the spent fuel going to be shipped to a final repository? .................. 33 

7.2 How is low-level radioactive waste shipped to the disposal site? ................... 34 

7.3 Are there regulations on radiation levels during transportation?..................... 34 

7.4 Are there safety criteria for spent fuel shipping casks, and how are the criteria 

satisfied? ............................................................... 34 

7.5 Are there safety criteria for shipping containers for low-level waste, and how are 

the criteria satisfied before shipment? ........................................ 34 

7.6 What regulations apply to the transportation of radioactive material? ............... 35 

7.7 Will the public be informed beforehand of low-level waste shipments or shipments 

of spent fuel and the routes taken? . . . . . . . . . . . . . . . . . . . . . . . -- . . . . . . . . 35 

7.8 Are specific routes used for transporting radioactive material, and does the NRC 

approve the routes used for radioactive material shipments?.................... 35 

7.9 How safe are shipments of spent fuel casks? What would happen if the train or truck 

carrying a spent fuel cask was involved in an accident? .......................... 36 

7.10 What emergency response plans are in place for transportation accidents involving 

a spent fuel cask or low-level-waste packages? ................................. 36 

viii

V|
8 LICENSE TERMINATION AND THE ULTIMATE DISPOSITION OF THE FACILITY .... 37 

8.1 How does decommissioning end, and who decides that the decommissioning is 

com plete? .............................................................. 37 

8.2 Why is the license-termination plan filed so late in the process?.................. 38 

8.3 How will the licensee know where the radioactive material or contamination is located 

within the plant? ......................................................... 38 

8.4 What is included in the site characterization? .................................. 38 

8.5 What does "suitable for release" mean? Are there any restrictions on how the site 

can be used? .... I ....................................................... 38 

8.6 Why would the licensee be allowed to restrict use of the site? ................... 39 

8.7 What is residual radioactivity, and why is it important to the termination of the 

license? ................................................................ 39 

8.8 What are the criteria for residual radioactivity at the site at the end of decommissioning, 

assuming that the licensee is planning for unrestricted use of the'site?................ 39 

8.9 Why is there a difference between the residual radioactivity limits set by the NRC and 

those set by the EPA? Why do the licensees not have to use the limits set by EPA? .... 40 

8.10 What is a "total effective dose equivalent?" ................ 40 

8.11 Who would be considered an "average member of the critical group?" .............. 40 

8.12 Have any facilities been decontaminated to the point of being allowed unrestricted 

release? ..................................................... 41 

8.13 What are the criteria for residual radioactivity at the site at the end of decommissioning, 

assuming that the licensee is planning for restricted use of the site? ................. 41 

8.14 How does the dose based on the residual radioactivity levels relate to background dose 

levels? ................................................................ 43 

8.15 Why didn't the NRC set the final dose criteria for release of the site for unrestricted 

use to "pre-existing background" levels? ...................................... 43 

8.16 Is it possible that some isotopes are located in such a way that radiation-monitoring 

devices cannot accurately detect their levels of radioactivity? ..................... 43 

8.17 Will continued monitoring be required after the decommissioning process is complete 

to ensure that the radiation levels do not increase? ............................. 43 

8.18 What types of uses can be made of the plant site after decommissioning is 

completed? .......................... ............................... 44 

8.19 Could the licensee initiate an alternative use of the site before the decommissioning 

process is completed? ..................................................... 44 

8.20 What uses have been made of sites that were decommissioned in the past? . . . . . . . . . . . 44 

8.21 What regulations are related to license termination? ............................. 44 

9 HAZARDS ASSOCIATED WITH DECOMMISSIONING ............................ 45 

9.1 W orkers ............................................................... 45 

9.1.1 Where do the decommissioning workers come from? ..................... 45 

9.1.2 Is worker safety considered in the planning for and review of 

decommissioning? ............ ....................... 45 

9.1.3 How much occupational dose is received by workers during 

decommissioning? .................. . ....... ....................... 45 

9.1.4 Are there limits on the amount of occupational dose that may be received? .... 45 

9.1.5 Does the licensee have to estimate the occupational dose before the decommissioning

ix
process is initiated? ................................................ 46 

9.2 Public and Environment ................................................... 46 

9.2.1 Is the safety of the public considered in the planning for and review of 

decommissioning? .................................................. 46 

.9.2.2 How much dose will the public receive during the decommissioning process?.. 46 

9.2.3 Who estimates what the doses are, and how are these estimates made? ........ 46 

9.2.4 What types of effluent releases are expected, and where will they enter the 

environm ent? ..................................................... 47 

9.2.5 Can you measure the effluent release to know how much is really entering the 

environment? ......... .................................... 47 

9.2.6 Will there be continued environmental monitoring of the site and the offsite 

areas to measure releases of radioactive material during the decommissioning 

process? ......................................................... 48 

9.2.7 Who will perform the environmental monitoring? ........................ 48 

9.2.8 Are the monitoring reports available to the public? How would we find 

them ? ...................................................... ....... 48 

9.2.9 What types of chemicals will be released to the environment? ............... 48 

9.2.10 What hazards are presented to the public when the waste is shipped? ......... 48 

9.2.11 What types of accidents at the reactor site are considered and what would be 

the consequences to the public? ........................................ 49 

9.3 General ...................................................... 50 

9.3.1 In general, how safe is a decommissioning plant in contrast to an operating 

plant? ........................................................... 50 

9.3.2 Will there still be emergency preparedness plans and warning sirens in the 

vicinity of the plant? ......................................... 50 

9.3.3 What measures are taken to prevent vandalism and sabotage during 

decomm issioning? ................................................. 50 

10 FINANCES ........................................................ 50 

10.1 How much does it cost to decommission a nuclear power plant? .................. 50 

10.2 Who makes the estimates of the decommissioning costs? ...................... 51 

10.3 When are the estimates of the decommissioning costs made? ...................... 51 

10.4 If the first estimate of decommissioning costs is made at the time that the facility is 

licensed, are there methods for adjusting for inflation? ........................ 52 

10.5 How does the NRC ensure that the licensee will have the money when it is needed for 

decom m issioning? ........................................................ 52 

10.6 Does the NRC have a method to monitor the decommissioning trust fund moneys for 

each specific nuclear power facility? ........................................ 53 

10.7 Are the financial assurance reports available to the public? ..................... 53 

10.8 Do the financial assurance regulations apply for Federal Government licensees? ....... 53 

10.9 Is there any way to ensure that the licensee does not just spend all of the money in the 

first few years of decommissioning and have nothing left to complete the job? ........ 53 

10.10 What would happen if the cost of decommissioning exceeds the amount of money in 

the trust fund? .......................................................... 54 

10.11 What would happen if the plant has an accident, and there is not enough money in 

t he decommissioning trust fund to complete decommissioning and cleanup after the accident? 54 

x

Vi
S10.12 Will enough money still be in the decommissioning trust fund even if the facility ceases 

operation? ................................................................ 54 

10.13 Who pays for decommissioning? ............................................ 54 

10.14 What contingency plans are in place to ensure that decommissioning and long-term 

radioactive material storage will be properly performed in the event of financial 

default of the licensee? Who finances decommissioning if the licensee becomes 

bankrupt or insolvent? .................................................... 55 

10.15 What will happen if deregulation becomes a reality? How will deregulation affect 

anticipated revenue and the ability to decommission? ............................ 55 

11 SOCIOECONOMIC ISSUES .................................................... 56 

11.1 What impact would each of the alternatives have on the economy of the surrounding 

area, including work-force requirements? ............ .................... 56 

11.2 How many people are needed onsite during the decommissioning process? Is this 

more than during operations? ............................................... 56 

11.3 What is being done to assist communities that relied heavily on the operating plant 

for their tax base, now that the plant has permanently ceased operations? ............ 56 

12 PUBLIC INVOLVEMENT ...................................................... 57 

12.1 Public M eetings ......................................................... 57 

12.1.1 What meetings are planned to keep the public informed? .................. 57 

12.1.2 Where will the meetings be held? ..................................... 57 

12.1.3 How will the public be notified about the meetings? ...................... 57 

12.1.4 If I cannot attend a meeting, how do I find out what was said? .............. 58 

12.1.5 May I make comments at the meeting? ................................. 58 

12.2 Comment Period ................................................ 58 

12.2.1 When does the comment period for the decommissioning process start? ....... 58 

12.2.2 How do I make comments on the decommissioning process? ............... 58 

12.2.3 Where should I send my comments? ................................... 59 

12.3 Hearings ...................................................... 59 

12.3.1 Are hearings held on the decommissioning process? ..................... 59 

12.3.2 Why is there no opportunity for a public hearing in the early stages of 

decommissioning? ................................................. 59 

12.3.3 What is the process that is used for submitting contentions for a public 

hearing? ......................................................... 59 

12.3.4 Does the public really have any power to make changes if there are problems? . 60 

13 GETTING ADDITIONAL INFORMATION ........................................ 60 

13.1 Other than the public meeting, how can I get information about a nuclear power 

plant? .................................................................. 60 

13.2 What is the Public Electronic Reading Room, and how do I access it? ............... 61 

13.3 Does the NRC have a website? What kind of information can I obtain from it? ....... 61 

13.4 Does the NRC maintain a Public Document Room? ............................. 62 

13.5 What is the Federal Register and how can I get a copy of it? ...................... 62 

13.6 How can I get a copy of the Code of Federal Regulations?........................ 62 

13.7 How can I get answers to additional questions that were not addressed in this 

docum ent? ............................................................. 62

xi
14 BIBLIOGRAPHY .................................................. ,. . .... 66

xii

Vt
ABBREVIATIONS

ADAMS 

ALARA 

BWR 

CEDE 

CFR 

DOE 

DOT 

EA 

EIS 

EPA 

EST 

FRERP 

GElS 

GE-VBWR 

GPO 

HEPA 

HLW 

ICRP 

ISFSI 

LLNL 

LLW 

NCRP

xiii

Agency-Wide Documents Access and Management System 

As low as is reasonably achievable 

Boiling Water Reactor 

Committed Effective Dose Equivalent 

Code of Federal Regulations 

U.S. Department of Energy 

U.S. Department of Transportation 

Environmental Assessment 

Environmental Impact Statement 

U.S. Environmental Protection Agency 

Eastern Standard Time 

Federal Radiological Emergency Response Plan 

Generic Environmental Impact Statement 

General Electric - Vallicetos Boiling Water Reactor 

Government Printing Office 

High-efficiency particulate air filter 

High level waste 

International Commission on Radiation Protection 

Independent Spent Fuel Storage Installation 

Lawrence Livermore National Laboratory 

Low-level waste 

National Council on Radiation Protection and Measurements
NPDES National Pollutant Discharge Elimination System 

NRC U.S. Nuclear Regulatory Commission 

ODCM Offsite Dose Calculation Manual 

PSDAR Post-shutdown Decommissioning Activities Report 

PWR Pressurized Water Reactor 

SFP Spent Fuel Pool 

TEDE Total Effective Dose Equivalent 

TDD Telecommunications device for the deaf 

URL Universal Resource Locator 

U.S. United States 

xiv

Ii
ACKNOWLEDGMENTS 

This report, which represents the cumulative efforts of many, was facilitated by bringing together a number 

of subject-matter specialists from within NRC to make comments on the document. We greatly appreciate 

the guidance, assistance, technical review, and encouragement provided by Mr. Tim Johnson, Ms. Sherry 

Lewis, Ms. Cynthia Sochor, Mr. Chris Gratton, Mr. Robert Wood, Mr. Brian Richter, Mr. Stephen 

Klementowicz, Ms Eileen M. McKenna, Ms. Elizabeth A. Hayden, Dr. M. Nalluswami, Mr. Barry 

Zalcman, Ms. Diane Jackson, Mr. James H. Wilson, Mr. George Mencinsky, Mr. Spiros Droggitis, Mr. 

Robert Lewis, Mr. Vincent J Everett, Mr. Mark Deiligatti, Ms. Nancy Osgood, Mr. Patrick M. Madden, 

Mr. George Hubbard, and other NRC staff members. We also acknowledge the project managers of the 

Decommissioning Section: Mr. Thomas Frederichs, Mr. Duke Wheeler, Mr. Phillip Ray, Mr. Mike Webb, 

Mr. Dino Scaletti, Mr. John Hickman, Mr. Paul Harris, Mr. Dick Dudley, Mr. Lee Thonus, Mr. Mort 

Fairtile, Ms Ann P-Iodgdon, Mr. David J. Wrona, Mr. William Huffman and Mr. Stuart A. Richards, 

Director of Project Directorate IV & Decommissioning., A special thanks to Ms. Ray Sanders, Ms 

Barbara A. Calure, and Ms Elaine B Morris, Technical Editors.

xv
INTRODUCTION

On July 20, 1995, the Commission issued a "Notice of Proposed Rulemaking on Decommissioning of 

Nuclear Power Plants." On July 2, 1996, the Commission approved the final rule. The rule was published in the Federal Register on July 29, 1996, and became effective 30 days from the date of publication 

(August 28, 1996) (61 FR 39278). The final rule (1) redefines the decommissioning process, (2) defines 

terminology related to decommissioning, (3) requires licensees to provide the NRC with early notification 

of planned decommissioning activities at their facilities, and (4) explicitly states the applicability of certain 

NRC requirements that are specific for reactors that are permanently shut down. On July 21, 1997, the 

NRC published in the Federal Register a final rule entitled, "Radiological Criteria for License 

Termination" (62 FR 39058). This rule contains the regulations regarding the radiological criteria that the 

licensee must meet before the license can be terminated. These regulations superseded regulations written 

in 1998. 

The staff realized that there was a general lack of understanding of the decommissioning process and the 

risks associated with decommissioning. With the anticipated increase in the number of power reactors 

beginning the decommissioning process and the significant changes that occurred in the regulations since 

1996, this report, through a question and answer format, provides NRC staff responses to frequently asked 

questions on decommissioning of commercial nuclear power reactors. 

The questions were taken from a variety of sources over the past several years, including written inquiries 

to the NRC and questions asked at public meetings and during informal discussions with the NRC staff. In 

responding to the questions, the NRC staff attempted to provide the answers in a clear and non-technical 

form. 

Sections I and 2 define decommissioning and discuss alternatives. Section 3 focuses on decommissioning experiences within the United States. Section 4 describes how the NRC regulates the decommissioning 

process. Sections 5, 6, and 7 concern spent fuel, low-level waste during decommissioning, and 

transportation, respectively. Sections 8 and 9, respectively, consider questions and answers on license 

termination, the ultimate disposition of the facility, and hazards associated with decommissioning. Section 

10 addresses the financial aspects of funding decommissioning. Socioeconomic issues are discussed in 

Section 11. Section 12 discusses public involvement in the decommissioning process, with an emphasis 

on the early phases of decommissioning. Section 13 provides the public with sources of additional 

information on decommissioning. The final section contains a bibliography with relevant published 

materials.

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1 GENERAL

1.1 How is decommissioning defined? 

Title 10 of the Code of Federal Regulations, Section 50.2 (10 CFR 50.2) defines decommissioning as the 

safe removal of a facility from service and reduction of residual radioactivity to a level that permits 

termination of the NRC license. 

1.2 Why do nuclear power plants shut down permanently? 

Nuclear power plants cease operations for a variety of reasons. The NRC grants a license for a period of 

40 years. At the end of the license period, the licensee can seek to renew the operating license of the plant 

for another 20 years, or can cease operations and begin the decommissioning process. Some licensees 

choose to cease power operations before the 40-year licensing period has been completed. Reasons for 

this decision are usually financial. For example, the plant may require upgrades or repairs that are not 

economically justifiable, or the licensee may find other sources of power that are less expensive than 

nuclear generation. In addition to financial reasons for decommissioning, the NRC can order the licensee 

to cease operations for safety reasons. 

1.3 Why are power reactors'decommissioned? 

As one of the conditions for an operating license, the NRC requires the licensee to commit to decommis

sioning the nuclear plant after it ceases power operations. This requirement is based on the need to reduce 

the amount of radioactive material at the site in order to ensure public health and safety as well as the 

protection of the environment. 

1.4 How does decommissioning proceed? 

The regulations are written so that when a licensee announces its decision to permanently cease power 

operations at the nuclear power plant (or in extreme cases, when the NRC requires a licensee to cease 

operations), the decommissioning process is automatically initiated, and specific decisions regarding the 

decommissioning process must be made within 2 years. However, no major decommissioning activities 

can take place until the licensee has provided the NRC with specific information regarding the decom

missioning process as required by the decommissioning regulations discussed later. 

It is possible for the licensee to let the facility sit idle for a number of years before announcing its decision 

to permanently cease power operations (although the time could not extend beyond the duration of the 

operating license). However, it is not in the licensee's financial interest to delay this decision since the 

costs required to meet the regulations at an operating plant are much greater than the costs for a 

decommissioning plant.

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1.5 What are the benefits of decommissioning?

The major benefit of decommissioning for the licensee as well as the public is that the levels of radioactive 

material at the site are reduced to levels that permit termination of the license and use of the site for other 

activities, rather than leaving the radioactively contaminated material on the site so that it could adversely 

affect public health and safety and the environment in the future. 

1.6 What are the costs of decommissioning? 

The major costs of decommissioning are the large financial costs involved in funding the project. 

Substantial costs are incurred in the removal, shipment, and disposal of major components of the facility 

that are contaminated, such as pumps, valves, piping, the steam generators, the pressurizer and the reactor 

vessel. Decontamination of floors, walls and equipment also result in substantial costs. The 

occupational dose received by workers during decommissioning should also be considered a cost. 

1.7 What are the options to decommissioning? 

NRC's regulations do not allow the option of not decommissioning. Under NRC regulations, the original 

operating license for a nuclear power plant is issued for up to 40 years. The license may be renewed for 

up to an additional 20 years if NRC requirements are met. However, at the end of the licensing period, the 

regulations require that the facility be decommissioned. The alternative to decommissioning, at the end of 

the licensing period, is a "no action" alternative, implying that a licensee would simply abandon or leave a 

facility after ceasing operations. This is not considered to be a viable alternative to decommissioning. The 

objective of decommissioning is to restore a nuclear facility to such a condition that there is no 

unacceptable risk from the decommissioned facility to public health and safety or the environment. In 

order to ensure that at the end of its life, the risk from a facility is within acceptable bounds, some action is 

required. If nuclear poweri plants were not decommissioned, they could degrade and become radiological 

hazards. 

2 DECOMMISSIONING PROCESS 

2.1 What terms or definitions are important to the understanding of 

decommissioning? 

A number of terms are important to the understanding of decommissioning. These terms are listed and 

defined below. It is also important to gain an understanding of the units used for measuring radiation 

dose: rem and person-rem. 

Activation products are radioactive materials that were created when stable substances were bombarded by 

neutrons. For example, cobalt-60 is formed from the neutron bombardment of the stable isotope 

cobalt-59. In a reactor facility, neutrons are created inside the reactor vessel during the fission process. 

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These neutrons bombard (1) the metal around the reactor vessel, (2) the primary reactor coolant, and 

(3) the concrete near the reactor vessel and create activation products in these materials. 

Alpha radiation is a positively charged particle ejected spontaneously from the nuclei of some radioactive 

elements. It does not penetrate very far into material and it has a very short range even in air (a few 

centimeters). The most energetic alpha particle will generally fail to penetrate the dead layers of cells 

covering the skin and can be easily stopped by a sheet of paper. Alpha particles are hazardous when an 

alpha-emitting isotope is inside the body. 

Background radiation means the radiation that is in the natural environment, including cosmic rays and 

radiation from the naturally radioactive elements, both outside and inside the bodies of humans and 

animals. It also includes radon (from the ground) and global fallout (as it exists in the environment from 

the testing of nuclear explosive devices or from past nuclear accidents, such as Chernobyl, that contribute 

to background radiation and that are not under the control of the licensee). 

Contamination means undesired (for example, radioactive or hazardous) material that is (1) deposited on 

the surface of, or internally ingrained into, structures or equipment, or (2) mixed with another material. 

Dose or radiation dose is a generic term that means absorbed dose, dose equivalent, effective dose 

equivalent, committed dose equivalent, committed effective dose equivalent (CEDE), or total effective 

dose equivalent (TEDE). In the case of radiation dose, it is energy absorbed per unit mass. Dose is 

measured in rads. The metric form of the rad is a gray (Gy) (1 rad = 0.01 gray). Radiation dose received 

by a person is measured in units called "rem," which incorporates the biological harm of the radiation dose 

based on the type of ionizing radiation. A sievert (Sv) is the metric form of the rem (I rem = 

0.01 sieverts). 

Greater than Class C waste is radioactive waste that is not generally acceptable for near-surface disposal. 

It is waste for which form and disposal methods must be different, and in general more stringent, than 

those specified for Class C waste. Such waste must be disposed of in a geological repository. 

Half-life is the time required for half of any quantity of identical radioactive atoms to undergo radioactive 

decay, so that half of the atoms in the substance are no longer emitting radiation and are no longer 

considered to be radioactive. 

Person-rem is the sum of all the radiation dose equivalents (measured in rem) that were received by an 

individual or by all individuals in a population group. For example, if 1,000 people each received 1/10th 

of a rem (100 millirem), the corresponding population dose would be 100 person-rem. Doses to an 

individual are usually measured in millirem. A sievert is the metric form of the rem (1000 millirem 

I rem = 0.01 sievert). 

Radiation (ionizing radiation) means alpha particles, beta particles, gamma rays, x-rays, neutrons, 

high-speed electrons, high-speed protons, and other. particles capable of producing ions. Radiation, as 

used in this section, does not include non-ionizing radiation, such as radio or microwaves, or visible, 

infrared, or ultraviolet light. 

Radioactive decay is the spontaneous natural process by which an unstable radioactive nucleus releases 

energy or particles.

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Rem (see Dose).

Residual radioactivity means radioactive contamination or activation products that remain following 

decontamination and dismantlement of the facility. 

2.2 What is the difference between radioactive contamination and activation 

products, and where are contaminated materials and activated materials 

located? 

Radioactive contamination is radioactive material that is deposited on a nonradioactive surface. The 

material may be deposited from the air, or it may be dissolved in water and subsequently deposited into 

material such as concrete. Radioactive contamination is generally located on or near the surface of 

materials like metal or high-density concrete or painted walls. It would travel farther into unpainted 

surfaces or lower density concrete. Radioactive contamination can usually be removed from surface areas 

by washing, scrubbing, spraying, or, in extreme cases, by removing the outer surface of the material. 

Contaminated materials are transported through the facility by workers, equipment, and to some degree 

through the air. Although many precautions are taken to prevent the movement of contaminated material 

in a nuclear facility and to clean up any contaminated materials that may be found, it is most likely that 

contamination will occur in the reactor building, around the spent fuel pool, and around specific pieces of 

equipment in the auxiliary building. The areas known to contain contamination are marked by the 

licensee, who routinely checks for contamination. 

Activation products are radioactive materials that were created when stable substances were bombarded by 

neutrons. Typically these materials are the concrete and the steel that surround the fuel core. The 

radioactive decay of activation products is the main source of radiation exposure to plant personnel. 

2.3 How is a nuclear power plant decommissioned? 

To decommission a nuclear power plant, the radioactive material on the site must be reduced to levels that 

would permit termination of the NRC license. This involves removing the spent fuel (the fuel that has 

been in the reactor vessel), dismantling any systems or components containing activation products (such as 

the reactor vessel and primary loop), and cleaning up or dismantling contaminated materials. All activated 

materials generally have to be removed from the facility and shipped to a waste processing, storage or 

disposal facility. Contaminated materials may either be cleaned of contamination onsite, or the 

contaminated sections may be cut off and removed (leaving most of the component intact in the facility), 

or they may be removed and shipped to a waste processing, storage, or disposal facility. The licensee 

decides how to decontaminate material, and the decision is usually based on the amount of contamination, 

the ease with which it can be removed, and the cost to remove the contamination versus the cost to ship the 

entire structure or component to a waste-disposal site. 

2.4 Who decides how a facility should be decommissioned? 

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The licensee decides how to decommission the site. Frequently, licensees hire contractors that specialize 

in decommissioning sites to conduct part or most of the decommissioning. The process for decontami

nation and dismantlement may vary from site to site. Factors that are used to make these decisions include 

cost, worker exposure, availability of a waste site, and layout and structure of buildings. For example, at 

some sites, it may make more sense to segment the reactor vessel before removing it from the reactor 

building; in other cases, it would be appropriate to remove the reactor vessel intact through a hole cut in 

the side of the containment building and ship the reactor vessel intact. 

The NRC provides a process, which is discussed in the response to Question 4.2.14, that enables the 

licensee to make specific types of changes without prior NRC approval. If any major decommissioning 

activity does not meet the conditions specified by the NRC, the licensee is prohibited from undertaking the 

activity until; I): it submits a license-amendment request that describes the proposed activity and the 

potential impact associated with that activity, and 2) the NRC approves the request. 

2.5 Is there a limit on the number of years that it would take to decommission 

a plant? 

The NRC regulations state that decommissioning must be completed within 60 years of permanent 

cessation of operations. A duration of 60 years was chosen because it roughly corresponds to 10 half-lives 

for cobalt-60, one of the predominant isotopes remaining in the facility. By 60 years, the initial short-lived 

isotopes, including cobalt-60, will have decayed to background levels. In addition, the 60-year period 

appears to be reasonable from the standpoint of expecting institutional controls to be maintained. For 

periods beyond 60 years, institutional controls such as those discussed in the response to Question 8.13 

would be required. Completion of decommissioning beyond 60 years will be approved by the NRC only 

when necessary to protect public health and safety. 

2.6 What alternatives are currently used for decommissioning? 

The NRC has evaluated the environmental impacts of three general methods for decommissioning power 

facilities. 

DECON: The equipment, structures, and portions of the facility and site that contain radioactive 

contaminants are removed or decontaminated to a level that permits termination of the license shortly after 

cessation of operations. 

SAFSTOR: The facility is placed in a safe stable condition and maintained in that state until it is 

subsequently decontaminated and dismantled to levels that permit license termination. During SAFSTOR, 

a facility is left intact, but the fuel has been removed from the reactor vessel, and radioactive liquids have 

been drained from systems and components and then processed. Radioactive decay occurs during the 

SAFSTOR period, thus reducing the quantity of contaminated and radioactive material that must be 

disposed of during decontamination and dismantlement. 

ENTOMB: Radioactive structures, systems, and components are encased in a structurally long-lived 

substance, such as concrete. The entombed structure is appropriately maintained, and continued 

surveillance is carried out until the radioactivity decays to a level that permits termination of the license.

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2.7 What are the benefits and costs of the DECON alternative? 

The DECON option calls for prompt removal of radioactive material to permit restricted or unrestricted 

access. The advantages of DECON include the following: 

"* facility license is terminated quickly, and the facility and site become available for other purposes 

"• availability of the operating reactor work force that is highly knowledgeable about the facility 

"* elimination of the need for long-term security, maintenance, and surveillance of the facility, which 

would be required for the other decommissioning alternatives 

"* greater certainty about the availability of low-level waste facilities that would be willing to accept the 

low-level radioactive waste 

" lower estimated costs compared to the alternative of SAFSTOR, largely as a result of future price 

escalation because most activities that occur during DECON would also occur during the SAFSTOR 

period, only at a later date. (It is anticipated that the later the date for completion of the 

decommissioning, the greater the cost.) 

The disadvantages of DECON include the following: 

"* higher worker and public doses (because there is less benefit from radioactive decay such as would 

occur in the SAFSTOR option) 

"* a larger initial commitment of money 

"* a larger potential commitment of disposal-site space than for the SAFSTOR option 

"* the potential for complications if spent fuel must remain on the site until a Federal repository for spent 

fuel becomes available. 

2.8 What are the benefits and costs of the SAFSTOR alternative? 

The benefits of SAFSTOR include the following: 

"* a substantial reduction in radioactivity as a result of the radioactive decay that results during the 

storage period 

"* a reduction in worker dose (as compared to the DECON alternative) 

"a a reduction in public exposure because of fewer shipments of radioactive material to the low-level 

waste site (as compared to the DECON alternative) 

"* a potential reduction in the amount of waste disposal space required (as compared to the DECON 

alternative)

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"* lower cost during the years immediately following permanent cessation of operations 

"* a storage period compatible with the need to store spent fuel onsite. 

Disadvantages of SAFSTOR include the following: 

"* shortage of personnel familiar with the facility at the time of deferred dismantlement and 

decontamination 

"* site unavailable for alternate uses during the extended storage period 

"* uncertainties regarding the availability and costs of low-level radioactive waste sites in the future 

"• continuing need for maintenance, security, and surveillance 

"* higher total cost for the subsequent decontamination and dismantlement period (assuming typical price 

escalation during the time the facility is stored). 

2.9 What are the benefits and costs of the ENTOMB alternative? 

The benefits of the ENTOMB process are primarily related to the reduced amount of work in encasing the 

facility in a structurally long-lived substance, and thus, reducing the worker dose from decontaminating 

and dismantling the facility. In addition, public exposure from waste transported to the low-level waste 

site would be minimized. The ENTOMB option may have a relatively low cost. However, because most 

power reactors will have radionuclides in concentrations exceeding the limits for unrestricted use even 

after 100 years, this option may not be feasible under the current regulations. This option might be 

acceptable for reactor facilities that can demonstrate that radionuclide levels will decay to levels that will 

allow restricted use of the site. Three small demonstration reactors have been entombed. Currently, no 

licensees have proposed the ENTOMB option for any of the power reactors undergoing decommissioning. 

2.10 Is the choice of decommissioning alternatives a decision that is left 

entirely to the licensee, or does the NRC help make this decision? 

The choice of the decommissioning method is left entirely to the licensee. However, the NRC would 

require the licensee to re-evaluate its decision if the choice (i) could not be completed as described, 

(2) could not be completed within 60 years of the permanent cessation of plant operations, (3) included 

activities that would endanger the health and safety of the public by being outside of the NRC's health and 

safety regulations, or (4) would result in a significant impact to the environment. 

2.11 Must a licensee choose either DECON or SAFSTOR, or can it combine 

the two alternatives? 

A licensee need not restrict its choice of decommissioning options to either an immediate decontamination 

and dismantlement or to a storage period of 30 to 60 years, followed by decontamination and 

dismantlement. Generally licensees combine the first two options. For example, after power operations

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stop at a facility, a licensee could uie a short storage period for planning purposes, followed by removal of 

large components (such as the steam generators, pressurizer, and reactor vessel internals), place the facility 

in storage for 30 years, and eventually finish the decontamination and dismantlement process. 

2.12 What main factors affect a licensee's choice of a decommissioning 

alternative? 

The SAFSTOR alternative is often used at multi-unit sites when one or more of the units shuts down while 

others continue to operate. This is especially true for facilities that share some systems. In this case, the 

staff from the operating unit(s) assist in the maintenance and surveillance of the unit that is in storage. 

The choice of decommissioning options is also strongly influenced by potential uncertainties in low-level 

waste disposal costs and by concerns over the future availability of low-level waste sites. The licensee's 

rate regulator can also influence the choice of decommissioning alternatives. 

2.13 How long does the dismantlement phase last? 

The dismantlement phase typically takes between 3 to 5 years to complete, although it may take longer if 

there are constraints on access to low-level waste burial.sites, or ifthe licensee decides to proceed at a 

slower pace for programmatic reasons. 

3 DECOMMISSIONED SITES 

3.1 What NRC-regulated plants have been or are being decommissioned, and 

what decommissioning alternatives have been or are being used? 

As of May 2000, radiological decommissioning had been completed at three NRC-licensed power 

reactors: 

"* Pathfinder test reactor, Sioux Falls, South Dakota 

"* Fort St. Vrain 

* Shoreham Nuclear Power Station, Suffolk County, New York (The Shoreham plant had only operated 

2 effective full-power days.) 

Nine reactors are in various stages of dismantlement and decontamination: 

"• Saxton, Saxton, Pennsylvania 

"• Haddam Neck, Haddam Neck, Connecticut 

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"* Trojan, Rainier, Oregon 

"• Maine Yankee, Wiscassett, Maine 

"* Big Rock Point, Charlevoix, Michigan 

"* Yankee Rowe, Franklin, Massachusetts 

Rancho Seco, Sacromento, California 

San Onofre 1, San Clemente, California 

Humboldt Bay Unit 3, Eureka, California 

Ten nuclear power reactors are in,' or are planning, long-term storage: 

"* Indian Point 1, Buchanan, New York 

"* Dresden 1, Morris, Illinois 

"* Peach Bottom, Sacramento, California 

"• LaCrosse, LaCrosse, Wisconsin 

"• Millstone 1, Waterford, Connecticut 

"* Vallecitos Boiling-Water Reactor (GE-VBWR), Pleasanton, California 

"* Fermi 1, Monroe County, Michigan 

"* Zion Unit 1, Zion, Illinois 

Zion Unit 2, Zion, Illinois 

Three Mile Island, Unit 2, Middletown, Pennsylvania 

3.2 Have other non-NRC-regulated nuclear facilities been decommissioned? 

Two nuclear power plants owned by the U.S. Department of Energy (DOE)-Elk River in Minnesota and 

Shippingport in Pennsylvania-have been decommissioned using the DECON alternative. Three DOE 

facilities-Bonus in Puerto Rico, Hallam in Nebraska, and Piqua in Ohio-have been decommissioned 

using the ENTOMB alternative. Other DOE test reactors and production reactors are in the process of 

being decommissioned. 

3.3 What improvements have been made as a result of previous 

decommissioning experience?

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Some improvements in the process, such as the removal of large components, including the reactor vessel, 

and the use of a primary system chemical flush to reduce worker exposure, have resulted from the 

experience gained from previous plant decommissionings. Other improvements or technologies were 

developed as part of the cleanup process for the Three Mile Island, Unit 2, plant after the 1979 accident. 

These include strippable coatings of latex or plastic that are used for decontaminating surfaces and the 

increased use of robotics. In addition, most licensees have gained experience with decommissioning 

techniques during routine preventive maintenance programs or as part of repairs required during 

operations. 

3.4 What research is being performed to find improved methods to be used 

during decommissioning? 

The U.S. Department of Energy has taken the lead on decommissioning technology research. The 

following types of improvements are being investigated: 

" surface-removal techniques to remove the outer surface of a contaminated structure, such as lasers or 

microwaves combined with vacuums, electrohydraulic scabbling (water-pressure shock waves that are 

electrically controlled), and electrokinetic decontamination of concrete (gel electrolytes are used with 

electrodes to leach ionic contaminants from deep inside porous concrete) 

" cutting techniques, such as laser cutting or oxy-gasoline torches (which work twice as fast as an 

acetylene torch on 1-inch steel) to remove structures 

"* improved methods for worker protection, such as protective suits with liquid air-cooling apparatus and 

lightweight breathable suits with chemical absorption protective layers 

"* environmental-protection techniques, such as automated asbestos removal and in situ chemical 

conversion of asbestos to non-hazardous material 

"* survey/monitoring techniques, such as pipe-explorer internal survey/characterization systems and a 

remote 3-D characterization and archiving system (robotic sensor and mapping platforms analyze for 

hazardous organic and radioactive contaminants). 

Commercial firms are also developing promising avenues of research into usable technologies. 

3.5 What differences are there in decommissioning between different types of 

reactor designs? 

In 1988, the NRC issued the Final generic environmental impact statement (GELS), NUREG-0586 (NRC 

1988) a generic analysis of decommissioning at nuclear facilities, including nuclear power plants. The 

analysis looked at total estimated costs, occupational and public dose, and low-level waste volumes. The 

GEIS estimates for cost for the reference facility were generally higher by about 20% for the boiling water 

reactors (BWRs) than for pressurized water reactors (PWRs), depending on the decommissioning option 

selected. Occupational dose estimates in the GElS were slightly higher for the reference BWR, by 10 to 

50%, depending on the decommissioning option. Estimates of public dose were lower for the reference 

BWR by up to a factor of 2, depending upon the scenario. Burial-volume estimates for low-level waste for 

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reference BWRs and PWRs for SAFSTOR and DECON options were very close to the same. In early 

2000 the NRC staff has begun an effort to update the 1988 GElS for power reactors. The staff expects to 

issue a draft supplement in the year 2001. 

One other major difference between decommissioning BWRs versus PWRs is that BWRs are designed so 

that the spent fuel pool is located in the reactor building, rather than in a separate building that can be 

isolated from the rest of the facility. This eliminates the possibility of decontaminating and decommis

sioning the remainder of the facility while leaving the spent fuel pool building as a "nuclear island." 

4 NRC ACTIVITIES 

4.1 Licensing 

4.1.1 Does the licensee have an NRC license even during the decommissioning process? 

Yes. The NRC license (called a "Part 50 license" in reference to the location of the corresponding 

regulations for the license in the Code of Federal Regulations) is not terminated until the licensee can 

demonstrate that it meets the criteria for site release in the regulations. The NRC verifies the licensee's 

final radiation survey by reviewing it and/or conducting a separate survey. In addition, the licensee must 

demonstrate that the facility has been dismantled in accordance with the approved license-termination 

plan. 

4.1.2 Is the license the same kind that is used for an operating plant? 

Yes. A facility that has permanently ceased operations and is being decommissioned maintains the same 

license that it had during operations. This is called a "Part 50 license," since it is regulated under Part 50 

of Title 10, "Energy," of the Code of Federal Regulations. There are a number of differences between the 

specific regulations that apply to a decommissioning versus an operating plant. This results primarily 

because certain systems (for instance, the reactor coolant system) are not operating and do not need to be 

regulated once the fuel has been removed from the reactor vessel. 

4.1.3 Could the NRC require a plant to cease operations and begin decommissioning? 

The regulations allow the NRC to revoke, suspend, or modify a license in whole or in part for failure to 

operate a facility in accordance with the terms of the license or for violation of, or failure to observe, any 

of the terms and provisions of the Atomic Energy Act, regulations, license, or order of the Commission. 

The NRC may issue an order to a licensee to permanently cease operations. If such an order were issued, 

the licensee would have 60 years from the date it permanently ceased operation until the completion of 

decommissioning. 

4.1.4 What would happen if a licensee refuses to decommission a plant that has ceased 

operations?

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The Commission may levy a civil penalty against the licensee. The regulations allow the NRC to obtain a 

court order for the payment of a civil penalty for violations of any rule, regulation, or order, or for 

violation of any term, condition, or limitation of any license. In addition, the Atomic Energy Act provides 

for the Federal Government to assume responsibility for decommissioning if public health and safety are 

jeopardized because of inactivity on the part of the licensee. 

4.1.5 What would happen if the licensee's license expires before the decommissioning 

process is concluded? 

The decommissioning regulations state that the license for a facility that has permanently ceased 

operations will continue in effect beyond the expiration date to authorize possession of the facility until the 

Commission notifies the licensee in writing that the license is terminated. During such a period of 

continued effectiveness, the licensee shall take actions necessary to decommission and decontaminate the 

facility and shall continue to maintain the facility in a safe condition. 

4.1.6 Are there any restrictions as to what the licensee can do with the site after 

decommissioning is completed and after the NRC has terminated the license? Would 

the NRC review these activities? 

Frequently, after the radiological decommissioning process and the termination of the license, the licensee 

will remove nonradioactive facilities, or will remodel some of the remaining buildings for other industrial 

uses. The activities that take place after the licensee has demonstrated that the radiological hazard has 

been removed, and after the license has been terminated, are not within the jurisdiction of the NRC. The 

NRC has no further oversight of these activities once the license is terminated. 

4.1.7 How would license renewal of operating plants at multi-unit facilities affect the 

decommissioning? Would the licensee be allowed to delay the decommissioning of the 

plant even longer? 

The regulations specify that decommissioning will be completed within 60 years of permanent cessation of 

operations. Completion of decommissioning beyond the 60 years will be approved by the Commission 

only when necessary to protect public health and safety. Any request for such an extension would be 

evaluated by the Commission based on its merit. For example, a multi-unit facility with one unit that had 

permanently ceased operations may request and receive a license renewal for the other unit(s) that are 

continuing to operate. This would extend the license term for the operating unit(s) for an additional 20 

years. The licensee may request an extension to the 60-year decommissioning period in order to leave the 

shut-down facility in a SAFSTOR mode until the operating unit(s) permanently cease operation and are 

ready to be decommissioned. If this occurred, the Commission would likely evaluate the condition of the 

shut-down unit and the positive and negative attributes of leaving the unit in SAFSTOR for a period 

exceeding 60 years. 

4.2 Regulations 

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4.2.1 Who regulates the decommissioning process?

The NRC regulates and provides oversight of the radiological aspect of the decommissioning process until 

it agrees to terminate the license. The mission of the NRC is to ensure adequate protection of public health 

and safety, protection of the environment, and protection and safeguarding of nuclear materials and 

nuclear power plants in the interest of national security. The NRC functions include (1) licensing nuclear 

facilities, (2) developing regulations and regulatory guidance, (3) conducting inspections and enforcement 

activities to ensure compliance with the regulations, (4) reviewing changes to the license, changes to 

licensee programs, and unreviewed safety questions, and (5) providing licensees with information that 

would improve their performance. 

4.2.2 Does the U.S. Environmental Protection Agency (EPA) have authority over the 

decommissioning process? 

The NRC oversees and regulates the decommissioning process in the same way that it regulates the 

operation of the nuclear power plant. However, licensees are required to comply with EPA regulations 

related to liquid effluent discharges to bodies of water. Compliance with the Clean Water Act 

implementing regulations requires the licensee to obtain a National Pollutant Discharge Elimination 

System (NPDES) permit for routine discharges to any body of water. In many cases, the EPA may 

delegate the authority to write and monitor the use of NPDES permits to States; for most facilities, the` 

NPDES permit is issued by the State where they are located. In some States, the authority still remains 

with EPA. Likewise, the EPA has implemented regulations for discharges to the atmosphere (Clean Air 

Act), and permits are required for specific pollutants, although reactor decommissioning does not 

generally fall under these regulations. 

4.2.3 When were the current regulations for decommissioning written? 

The NRC continually reviews the decommissioning regulations that it is using and revises them as 

required to improve the regulatory process. On July 29, 1996, a final rule amending the regulations on 

decommissioning procedures was published in the Federal Register (61 FR 39278). On July 21, 1997, the 

NRC published (also in the Federal Register) a final rule entitled, "Radiological Criteria for License 

Termination" (62 FR 39058). This rule contains the regulations regarding the radiological criteria that the 

licensee must meet before the license can be terminated. These regulations superseded regulations written 

in 1988. Certain financial-assurance aspects of the regulations were amended by a final rule that was 

published in the Federal Register (63 FR 50465) on September 22, 1998. 

4.2.4 What publications contain the regulations for decommissioning? 

Regulations regarding the decommissioning of NRC-licensed plants appear in the Code of Federal 

Regulations. The Code of Federal Regulations is a codificaiion of the general and permanent rules 

published in the Federal Register by the executive departments and agencies of the Federal Government. 

The Code is divided into 50 titles, which represent broad areas subject to Federal regulation. Each title is 

divided into chapters; these usually bear the name of the issuing agency. Each chapter is further 

subdivided into parts covering specific regulatory areas.

13
The regulations related to the decommissioning of power reactors are included in Title 10, "Energy," 

Chapter I-Nuclear Regulatory Commission; for example, Part 20, "Standards for Protection Against 

Radiation"; Part 50--"Domestic Licensing of Production and Utilization Facilities"; and Part 

5 1--"Environmental Protection Regulations for Domestic Licensing and Related Regulatory Functions." 

The subparts related to decommissioning are 20.1402, "Radiological criteria for unrestricted use"; 

20.1403, "Criteria for license termination under restricted conditions"; 20.1404, "Alternate criteria for 

license termination"; 20.1405, "Public notification and public participation"; 20.1406, "Minimization of 

contamination"; 50.75, "Reporting and recordkeeping for decommissioning planning"; 50.82, 

"Termination of license"; 51.53, "Post-construction environmental reports";-and 51.95, "Post-construction 

environmental impact statements." These regulations state the technical and financial criteria for 

decommissioning licensed nuclear facilities. They address decommissioning, planning needs, timing, 

funding methods, and environmental-review requirements. 

See the response to Question 13.6 for instructions on how to obtain a copy of the regulation. 

4.2.5 What regulatory actions are required to decommission a nuclear facility? 

The regulations specify actions that both the NRC and the licensee must take to decommission a nuclear 

power plant. Once the decision is made to permanently cease operations, the licensee must notify the 

NRC, in writing, within 30 days. The notification must contain the date on which the power generation 

operations ceased or will cease. The licensee must remove the fuel from the reactor and submit a written 

certification to the NRC confirming its action. There is no time limit specified before the fuel must be 

removed or the corresponding certification received by the NRC. Once this certification has been 

submitted, the licensee is no longer permitted to operate the reactor or to put fuel back into the reactor 

vessel. This also reduces the licensee's annual license fee to the NRC and eliminates the obligation to 

adhere to certain requirements that are needed only during reactor operations. The licensee must submit a 

post-shutdown decommissioning activities report (PSDAR) to the NRC and the affected State(s) no later 

than 2 years after the date of permanent cessation of operations. The PSDAR must 

"* describe the planned decommissioning activities 

"* contain a schedule for the accomplishment of significant milestones 

"* provide an estimate of expected cost 

"* provide documentation that environmental impacts associated with site-specific decommissioning 

activities have been considered in previously approved environmental impact statements. 

If the environmental impacts that are identified have not been considered in existing environmental 

assessments, the licensee must address the impacts in a request for a license amendment regarding the 

activities. The licensee also must submit a supplement to its environmental report that relates to the 

additional impacts. The NRC will review this environmental assessment or supplement to the 

environmental statement in conjunction with its review of the license-amendment request. 

After receiving a PSDAR, the NRC publishes a notice of receipt, makes the PSDAR available for public 

review and comment, and holds a public meeting in the vicinity of the plant to discuss the licensee's plans. 

14

V
Although the NRC will determine if the information is consistent with the regulations, NRC approval of 

the PSDAR is not required. However, should the NRC determine that the informational requirements of 

the regulations are not met in the PSDAR, the NRC will inform the licensee in writing of the deficiencies 

and reqdire that they be addressed before the licensee initiates any major decommissioning activities. 

Upon completion of the required submittals, and allowing for a 90-day waiting period after submittal of 

the PSDAR, the licensee may commence major decommissioning activities. Major decommissioning 

activities include the following: 

" permanent removal of major radioactive components, such as the reactor vessel, steam generators, or 

other components that are comparably radioactive 

"* permanent changes to the containment structure 

"* dismantling components resulting in "greater than Class C" waste. 

Decommissioning activities conducted without specific prior NRC approval must not preclude release of 

the site for possible unrestricted use, must not result in there being no reasonable assurance that adequate 

funds will be available for decommissioning, and must not cause any significant environmental impact not 

previously reviewed. If any decommissioning activity does not meet these terms, the licensee is required 

to submit a license-amendment request before conducting the activity; this would provide an opportunity 

for a public hearing. 

Activities that are not considered to be "major decommissioning activities" may be performed in 

accordance with the license and technical specifications in effect for the facility even before the end of the 

90-day waiting period. Allowable activities would include such routine items as maintenance and 

low-level waste disposal of small radioactive components. 

Within 2 years following the date of permanent cessation of operations, the licensees must submit a 

site-specific cost estimate for the decommissioning project. The licensee is prohibited from using the full 

amount of money that was accumulated during operations for the decommissioning process until the 

site-specific cost estimate is submitted to the NRC. 

Within 2 years following permanent cessation of operation licenses are also required to submit for review 

and preliminary approval a description of how they intend to manage and provide funding for the 

management of irradiated fuel until title to the fuel is transferred to the Department of Energy (DOT). 

Unless the licensee receives permission to the contrary, the site must be decommissioned within 60 years. 

The licensee remains accountable to the NRC until decommissioning has been completed and the license 

is terminated. In order to conclude its obligations, the licensee must submit a license-termination plan. 

The license-termination plan must be submitted at least 2 years before the termination date. It must 

include the following: 

"* a site characterization 

"* identification of remaining dismantlement activities

15
0 plans for site remediation

"* detailed plans for the final survey of residual contamination on the site 

" a description of the end-use of the site (if restricted use is proposed, a description of institutional 

controls and maintenance and surveillance programs is needed) 

" an updated site-specific estimate of remaining decommissioning costs 

"* a supplement to the environmental report. 

After receiving the license-termination plan, the NRC will place a notice of receipt of the plan in the 

Federal Register and will make the plan available to the public for comment. The NRC will schedule a 

public meeting near the facility to discuss the plan's contents with the public. The NRC will also offer an 

opportunity for a public hearing on the license amendment associated with the licensee-termination plan. 

If the license-termination plan demonstrates that the remainder of decommissioning activities will be 

performed in accordance with the NRC's regulations, is not detrimental to the health and safety of the 

public, and does not have a significant effect on the quality of the environment, the Commission will 

approve the plan by a license amendment (subject to whatever conditions and limitations the NRC deems 

appropriate and necessary). Once the license amendment is granted, the licensee may achieve license 

termination by successfully demonstrating implementation of the license-termination plan as approved by 

the NRC. 

At the end of the license-termination-plan process, if the NRC determines that the remaining dismantle

ment has been performed in accordance with the approved license-termination plan, and if the final 

radiation survey and associated documentation demonstrate that the facility and site are suitable for 

release, then the Commission will terminate the license, and the decommissioning process is considered 

complete. 

4.2.6 Why isn't an environmental impact statement (EIS), or environmental assessment 

(EA) required at the start of the decommissioning process? 

In 1996, the NRC published the final rule on the current decommissioning regulations. The Commission 

stated in that rule that they had determined that under the National Environmental Policy Act of 1969, as 

amended, and under the Commission's own regulations, that major decommissioning (dismantlement) 

activities could proceed without an environmental assessment. The impacts of these major decommis

sioning activities were determined to be within the bounds of those that were assessed in the NRC's 

Generic Environmental Impact Statement (GEIS) on Decommissioning (NUREG-0586), which was issued 

in 1988. However, because the GEIS did not address site-specific situations, the final rule prohibits major 

decommissioning activities that could result in significant environmental impacts not previously reviewed. 

The review process also includes a requirement that the licensee provide the reasons for concluding that 

the planned activities are bounded by the GEIS and previous site-specific environmental impact 

statements. The license termination plan (LTP), submitted by the licensee near the end of the 

decommissioning process, is reviewed and approved by the NRC staff. The LTP is incorporated into the 

facility license by an amendment to the license. The amendment process includes an environmental 

assessment by the staff. 

16'

II
4.2.7 What activities can take place before submitting the Post-shutdown Decommissioning 

Activities Report? 

No major decommissioning activities may take place until 90 days after the PSDAR has been submitted. 

Major decommissioning activities are defined as "any activity that results in permanent removal of major 

radioactive components, permanently modifies the structure of the containment, or results in dismantling 

components for shipment containing greater than Class C waste." Major radioactive components are 

defined by the regulations as the reactor vessel and internals, steam generators, pressurizers, large-bore 

reactor-coolant-system piping, and other large components that are radioactive to a comparable degree. 

Examples of activities considered minor decommissioning activities are (1) normal maintenance and 

repair, (2) removal of certain, relatively small, radioactive components, such as control-rod drive 

mechanisms, control rods, pumps, piping, and valves, (3) removal of components (other than those defined 

above as major components) similar to those removed for maintenance and repair during plant operations, 

(4) removal of non-radioactive components and of radiation structures not required for safety, (5) 

shipment of reactor fuel offsite, and (6) site characterization and measurement of contamination levels. 

4.2.8 Can the licensee make changes to its plans or commitments after submitting the 

PSDAR? 

Yes. However, the regulations require the licensee to notify the NRC in writing before performing any 

decommissioning activity that is not consistent with, or could be considered to be a significant change 

from, the actions or schedules described in the Post-Shutdown Decommissioning Activities Report 

(PSDAR). Significant changes in cost are also to be reported to the NRC. Significant changes to the 

milestone schedule are used by the NRC staff to reschedule NRC inspections of the licensee's activities. 

Examples of changes in activities and schedule that would require prior NRC notification include, but are 

not limited to, changing from long-term storage to active dismantlement, changing the method used to 

remove the reactor vessel from cutting and segmenting to intact removal, or changing the schedule to 

affect major milestones. Examples of significant increases in cost associated with decommissioning the 

facility would include a new estimated cost more than 20 percent above the site-specific cost estimate or 

the PSDAR cost estimate, or a 25-percent increase in cost above a major milestone estimate. 

Written notifications to the NRC do not require a 90-day waiting period before initiation of activities. 

Typically, the staff would not require a public meeting to discuss the proposed changes unless the NRC 

staff determined that the change was significant enough to warrant a public meeting. 

4.2.9 What would happen if the licensee wanted to switch from SAFSTOR to DECON? 

If the licensee proposes changing the method of decommissioning, for example, from long-term storage 

followed by decontamination and dismantlement to prompt decontamination and dismantlement, a public. 

meeting would be held. Normally, a licensee's decision to conduct some limited decontamination and 

dismantlement during long-term storage (if such action had not been specified in the PSDAR) would 

require submittal of an update, but would not require a public meeting. If the expected environmental 

impact of any change in decommissioning activities is significantly greater than that predicted in a 

previous environmental impact statement or assessment (see NUREG-0586, "Final Generic Environmental 

Impact Statement on Decommissioning of Nuclear Facilities"), the licensee would be required to request a 

licensee amendment and provide a supplement to the environmental report for the facility that evaluates

17
the impact of the change. The NRC staff would review the licensee's submittal and would publish either 

an environmental assessment or a supplement to the facility's final environment statement in conjunction 

with review of the license-amendment request. 

4.2.10 Will the PSDAR or the License Termination Plan be reviewed by the 

U.S. Environmental Protection Agency (EPA)? 

The NRC will place a notice in the Federal Register of the availability of the PSDAR for comment. The 

EPA, as all Federal agencies, will be invited to comment on the PSDAR. 

4.2.11 Under what circumstances would the NRC refuse to allow a licensee to proceed with 

the decommissioning? 

It is highly unlikely that the NRC would refuse to allow a licensee to decommission a plant since the 

regulations require a licensee to complete the decommissioning process within 60 years of permanent 

cessation of operations. However, the NRC staff may determine that the information in the PSDAR is 

inconsistent with what is required by the regulations. In that case, the NRC will inform the licensee in 

writing of the deficiencies and will require the deficiencies to be addressed before starting major 

decommissioning activities. 

4.2.12 Would the NRC ever stop a licensee from decommissioning a plant on the basis of 

information in the Post-Shutdown Decommissioning Activities Report? 

A number of factors could cause the NRC to find the PSDAR deficient, and the NRC staff would prevent 

the licensee from proceeding with decommissioning as described in the PSDAR. The NRC could find the 

PSDAR deficient if the licensee's plan for decommissioning could not be completed as described (for 

example, if the plan called for an immediate decontamination and dismantlement of the facility and there 

were no waste-disposal facilities available for the licensee to use). The NRC could find the PSDAR 

deficient if the schedule contained a decommissioning process that could not be completed within 60 years 

of the permanent cessation of operations, unless it were shown that this action was necessary to protect 

public health and safety. Factors that would be considered for an extended decommissioning process 

include the unavailability of low-level-waste disposal capacity and other site-specific factors affecting the 

licensee's capability to carry out decommissioning in the given time period, including the presence of 

other operating nuclear facilities at the site. The NRC would find the PSDAR deficient if the licensee's 

decommissioning plans, as presented in the PSDAR, contained a decommissioning process that obviously 

could not be completed for the estimated cost (the NRC staff will base this decision on the generic 

guidelines and on previous facility decommissioning costs). The NRC would also find a PSDAR deficient 

if it contained activities that would endanger the health and safety of the public by being outside the 

NRC's health and safety regulations or that would result in a major detrimental'impact to the environment 

that was not bounded by the current environmental impact statements. 

4.2.13 Can the NRC require the licensee to follow the regulations, and how would the NRC 

know if the regulations were not being followed? 

The NRC reviews licensee procedures and works to ensure that licensees are properly adhering to the 

regulations. At the start of the decommissioning process, the NRC reviews the licensee's plans to enter

18
active decontamination and dismantlement. Generally, an NRC resident inspector will remain onsite for a 

time after cessation of operations. Additionally, all through the decommissioning process, NRC inspectors will periodically conduct special inspections of specific activities at the site. Site visits and inspections 

will be more frequent for plants that are undergoing decontamination and dismantlement, and less frequent for plants that are in a storage mode. If the licensee is not complying with regulations, the NRC will 

address the issue in accordance with NRC's enforcement policy (NUREG-1600). 

4.2.14 Does the NRC have to review and either approve or disapprove every decision the 

licensee makes related to decommissioning a facility? 

No. The NRC does not review every decision the licensee makes related to decommissioning the facility. 

The licensee may make any of the following types of changes without prior NRC approval: 

"* changes in the facility as described in the safety analysis report 

"* changes in the procedures as described in the safety analysis report 

"* conduct of tests or experiments not described in the safety analysis report. 

These changes, tests, and experiments can be made or conducted without prior Commission approval 

provided that the proposed change, test, or experiment does not require a change (1) in the technical 

specifications incorporated in the license or does not meet any of the following criteria: (1) that the 

likelihood of occurrence or the consequences of an accident or malfunction of structures, systems or 

equipments important to safety and previously evaluated in the safety analysis report may be increased 

more than a minimal amount, or (2) that the possibility of an accident or malfunction of a different type 

from any evaluated previously in the safety analysis report (SAR) may be created, or (3) create a 

possibility for a malfunction of a structure, system or component important to safety with a different result 

than any previously evaluated in the SAR, or (4) result in a design basis limit for a fission product barrier 

as described in the SAR. being exceeded or (5) result in a departure from a method of evaluation described 

in the SAR used in establishing the design basis or in the safety analysis. These revised requirements are 

contained in 10 CFR 50.59 and will be in effect January 2001. 

4.2.15 How can the NRC make sure a licensee does not make a mistake in its evaluation of 

proposed changes, tests, or experiments? What happens if the licensee fails to 

identify an unreviewed safety question? 

The licensee is required to periodically submit a report containing a brief description of any changes, tests, 

and experiments made at the facility, and to summarize the safety evaluation of each. This report must be 

filed with the NRC at least every 24 months for facilities that have submitted certifications for permanent 

removal of fuel. These reports are reviewed by the NRC. In addition, records of changes in the facility 

are required to be maintained until the date of termination of the license, and records of changes in 

procedures and records of tests and experiments are required to be maintained for a period of 5 years. 

These changes in the facility are normally inspected by the NRC annually. The Decommissioning Power 

Reactor Inspection Program Manual specifies the review of the documents that detail the licensee's basis 

for conducting operations in a safe manner. If the inspector disagrees with the licensee's evaluation or if

19
there are questions as to whether a proposed change is an unreviewed safety question, the NRC staff will 

evaluate this issue until it is appropriately resolved. 

4.2.16 What would happen if the licensee decides to do something that will endanger the 

environment during the decommissioning process? 

NRC regulations state that the licensee must not perform any decommissioning activity that, (1) forecloses 

release of the site for possible unrestricted use, (2) causes any significant environmental impact not 

previously reviewed, or (3) results in there no longer being reasonable assurance that adequate funds will 

be available for decommissioning. If any major decommissioning activity could not meet these 

conditions, the licensee is prohibited from undertaking the activity until it submits a license-amendment 

request that describes the proposed activity and the potential impact associated with that activity. The 

license-amendment request would provide an opportunity for a public hearing. The NRC staff will 

evaluate the licensee's procedures for ensuring that these three restrictions are part of the screening criteria 

that are used for proposed changes to the facility. Additionally, periodic inspections of the licensee's 

activities will consider the environmental protection issues. 

4.2.17 What were the previous regulations and why were they changed? I 

When the NRC issued decommissioning regulations in 1988, it was assumed that decommissioning would 

take place after the facility's operating license expired. The licensee was obligated to submit a preliminary 

decommissioning plan 5 years before the license expired. The preliminary decommissioning plan 

contained a cost estimate for decommissioning and an up-to-date technical assessment of the factors that 

could affect planning for decommissioning. This included (1) the choice of alternative, (2) the major 

technical actions necessary to carry out decommissioning safely, (3) the current situation with regard to 

disposal of high-level and low-level radioactive waste, (4) the residual radioactivity criteria, and (5) other 

site-specific factors that could affect decommissioning planning and cost. 

The previous rule also required that no later than 1 year before expiration of the license (or within 2 years 

of permanent cessation of operations for plants closing before their license expires), a licensee had to 

submit an application for authority to decommission the facility. The application was to be accompanied 

by or preceded by a proposed decommissioning plan. The proposed decommissioning plan was to include 

(1) the choice of the alternative for decommissioning with a description of the activities involved, (2) a 

description of controls and limits on procedures and equipment to protect occupational aind public health 

and safety, (3) a description of the planned final radiation survey, (4) an updated cost estimate for the 

chosen alternative and a plan for ensuring the availability of adequate funding, and (5) a description of the 

technical specifications, quality assurance provisions, and physical security plan provisions in place during 

decommissioning. A supplemental environmental report that described any substantive environmental 

impacts that were anticipated but not already covered in other environmental impact documents was also 

required. 

The NRC reviewed the decommissioning plan. The Commission would approve the plan if the plan 

demonstrated that the decommissioning would be performed in accordance with regulations and there 

were no security, health, or safety issues. The Commission would also require that notice be given to 

interested persons. However, the NRC could add other conditions and limits to the plan that it deemed 

appropriate. The license would then be terminated if the Commission determined that the 

decommissioning had been performed in accordance with the approved decommissioning plan and the 

20

31
order authorizing decommissioning, and if a final radiation survey and associated documentation 

demonstrated that the facility and site were suitable for release for unrestricted use. 

The regulations were revised for several reasons. First, the experience gained in the early decommis

sioning activities associated with several facilities did not reveal any activities that required NRC review 

and approval of a decommissioning plan. Second, environmental impacts associated with decommis

sioning those early facilities resulted in impacts consistent with those evaluated in the "Generic 

Environmental Impact Statement on Decommissioning of Nuclear Facilities," NUREG-0586. And finally, 

experience gained from reviewing numerous decommissioning oversight activities at a number of these 

facilities also indicated that the decommissioning activities were in general no more complicated than 

activities normally undertaken at operating reactors without prior and specific NRC approval. In August 1996, the revised rule that redefined the decommissioning process and required licensees to provide the 

NRC with early notification of planned decommissioning activities at their facilities went into effect. The 

rule made the decommissioning process more efficient and uniform. It provided for greater public 

participation in the decommissioning process and gave plant personnel a clearer understanding of the 

process for changing from an operating organization to a decommissioning organization. The current 

regulatory process for decommissioning a facility is described in the response to Question 4.2.5. 

4.3 Inspection Program 

4.3.1 What are the goals of the inspection program at nuclear power plants undergoing 

decommissioning? 

The goals of the inspection program at nuclear power plants undergoing decommissioning are to 

" obtain sufficient information through direct observation and verification to determine if 

decommissioning is being conducted safely, if the spent nuclear fuel is being stored safely, and if 

activities at the site are being conducted in accordance with all applicable regulations and 

commitments 

" determine if the administrative controls that the licensee has in place are adequate and in accordance 

with regulatory requirements (the controls include self-assessment, audits and corrective actions, 

design control, safety review, maintenance and surveillance, radiation protection, and effluent 

controls) 

" identify any significant declining performance trends and verify that the licensee has taken actions to 

reverse any trend. 

4.3.2 What types of inspections are conducted? 

Three types of inspections will be conducted: core, temporary instruction compliance, and discretionary. 

Core inspections involve the inspection of a number of functional areas at specified frequencies. The 

functional areas are facility management and control, decommissioning -support activities, spent fuel 

safety, and radiological safety. Each of these functional areas is further broken down into a number of

21
subelements. Each subelement has a specific frequency based on whether the plant is actively removing 

components, is in a storage mode, or is undergoing site-termination activities. 

Temporary instruction compliance inspections are inspections designed to investigate generic problems 

that have been identified. Examples are spent fuel pool siphoning and spent fuel pool seismic protection. 

Discretionary inspections are detailed reviews of a particular functional area. Examples are high-level 

waste transportation, internal and external dose assessment, and confirmatory surveys. 

4.3.3 Which areas receive the greatest emphasis during the inspections? 

The inspection effort during decommissioning places the greatest emphasis on radiological controls, 

management, procedure compliance, spent fuel pool operations, and the safety review program. 

4.3.4 Are there some decommissioning activities that receive greater oversight than others? 

Many of the activities that occur during decommissioning are very routine and occur frequently in 

operating plants. These activities include decontamination of surfaces and components, and waste 

packaging and disposal, among others. The NRC inspectors have previously reviewed the process the 

licensee uses for many of these activities and will continue to review the process as any changes are made. 

The types of activities that will receive the most attention from the NRC inspectors include any activities 

that relate to the safe storage of irradiated fuel and any activities that could result in offsite releases in 

excess of the regulatory limits (for example, removal of a large component like'a steam generator). 

4.3.5 Does the NRC independently monitor radioactive releases into the environment? 

The NRC does not independently monitor radioactive releases on a continuous basis. However, the NRC 

does occasionally have an independent measurement of the releases to verify that the licensee's program is 

adequate to continuously analyze samples and to measure the amount of radioactive material that leaves 

the plant. In addition, the NRC reviews the licensee's documented procedures for monitoring radioactive 

releases into the environment and reviews the results periodically. All licensee's submit effluent release 

results to the NRC on an annual basis. 

4.3.6 Will the NRC have inspectors onsite during decommissioning? 

The inspection effort at a plant undergoing decommissioning is significantly less than at an operating 

reactor site. Operating reactor facilities have between one and two onsite inspectors, who maintain a 

continual presence at the plant or in the vicinity, for each unit on the site, However, because of the 

reduced hazard during the decommissioning process, NRC inspectors may not maintain a continual 

presence. The NRC will generally remove the onsite inspector from a single-unit site within 1 year from a 

reactor's permanent cessation of operations. The resident inspector's stay at a site can be extended 

because of special circumstances, however. Rather than stationing a resident inspector at the site during 

the entire decommissioning process, the NRC will provide subject-matter experts to cover specific 

activities occurring at the site. For example, if the licensee is planning to remove a large component, the 

NRC might send at appropriate times an expert in radiation protection, an expert in heavy lifting and polar 

cranes, and an expert in packaging radioactive waste. Inspections are performed by the NRC headquarters 

staff and NRC regional personnel. This requires attention to scheduling so that NRC personnel are 

22

31
available to review the licensee's procedures and to inspect before and during specific activities. The 

extent of onsite presence at the facility will depend on the activities that are taking place. During active 

decommissioning, NRC personnel may be at the facility 2 or 3 weeks of the month. During storage 

operations, they would be present several times a year. 

4.3.7 Are inspection reports associated with the decommissioning activities available on the 

NRC's website? 

The inspection manual chapters are available from the NRC website 

(http://www.nre.gov/NRC/IM/index.html). The inspection reports are available from the Public Document 

Room and from the Electronic Reading Room. Section 13 describes the process for obtaining information 

from the Public Document Room and the Electronic Reading Room. 

4.3.8 Will plant security levels be de-emphasized during decommissioning? 

Licensees have requested and been granted exemptions to specific sections of 10 CFR 73.55 that would 

reduce the security requirements below those of an operating reactor. This is done only after a thorough 

review by NRC security specialist. 

5 SPENT FUEL 

5.1 What are "high-level wastes?" 

High-level radioactive waste (HLW), as it pertains to commercial nuclear power reactors, is mainly 

irradiated (spent) reactor fuel. 

5.2 What is meant by the term "spent fuel?" 

Spent nuclear fuel is uranium-bearing fuel elements that have been used at commercial nuclear power 

reactors. Although spent (used) fuel can no longer produce enough heat to produce electricity, it contains 

highly radioactive material resulting from the fission process that takes place within the reactor. As a 

result, it still continues to generate radiation and heat. This heat and radiation are caused by "radioactive 

decay" of the products of the fission process. The heat and radioactivity in spent fuel necessitate that any 

shipment be made in containers or casks that provide the necessary degree of protection. In practice, this 

means that a cask must shield and contain the radioactivity and dissipate the generated heat. 

5.3 Are there facilities or plans for facilities for the disposal of high-level 

radioactive waste? 

At this time, there are no facilities for permanent disposal of high-level radioactive wastes. On January 7, 

1983, the President signed into law the Nuclear Waste Policy Act, which defined the goals and structure of 

a program for permanent, deep geologic repositories for the disposal of high-level radioactive waste and

23
unreprocessed spent fuel. Under this Act, the U.S. Department of Energy (DOE) is responsible for 

developing permanent disposal capacity for the spent fuel and other high-level nuclear wastes. At the 

present time, DOE, as directed by Congress, is investigating a site in Yucca Mountain, Nevada, which 

would be built and operated by DOE and licensed by the NRC, for a possible disposal facility. 

5.4 Is the licensee allowed to store the spent fuel in the reactor vessel? 

No. The licensee must submit a certification indicating that it has permanently removed the fuel from the 

reactor vessel before it can start the decommissioning process. When the NRC receives this certification, 

the licensee is prohibited from loading fuel back into the reactor vessel. The licensee has incentives to 

permanently remove the fuel from the reactor vessel because certification (along with the certification of 

permanent cessation of operations) reduces the licensee's annual license fee to the NRC and eliminates the 

obligation to adhere to certain requirements needed only during reactor operations. 

5.5 Since there are no facilities for permanent disposal, where will the spent 

nuclear fuel be kept during the decommissioning process? 

Until the repository is approved and constructed, spent nuclear fuel is being stored primarily in specially 

designed, water-filled basins (spent fuel pools) at individual reactor sites around the country. Another 

option for storage is in an independent spent fuel storage installation (ISFSI), which is located either at the 

reactor site or elsewhere. The spent fuel may be stored in air-cooled dry casks at an ISFSI (see Section 

5.9). These options will continue to be available during the decommissioning process. 

5.6 What are the long-range plans for disposition of spent fuel? 

The fuel would be stored in an independent spent-fuel storage installation (ISFSI) or in the spent fuel pool 

until a Federal repository became available. Then the spent fuel would be shipped to the Federal 

repository for final disposition. 

5.7 What happens if a disposal site for high-level waste is never licensed? 

The NRC has stated in its regulations that "The Commission has made a generic determination that, if 

necessary, spent fuel generated in any reactor can be stored safely and without significant environmental 

impact for at least 30 years beyond the licensed life for operation (which may include the term of renewed 

license) of that reactor at its spent fuel storage basin or at either onsite or offsite independent fuel-storage 

installations." Further, the Commission believes there is reasonable assurance that at least one mined 

geological repository will be available in the first quarter of the 21st century, and sufficient repository 

capacity will be available within 30 years beyond the licensed life for operation of any reactor to dispose 

of the commercial high-level waste and spent fuel originating in such reactor and generated up to that 

time. 

5.8 Spent Fuel Pools 

5.8.1 Why is spent fuel stored in a pool of water? 

24

V!
Even after the nuclear reactor is shut down, the fuel continues to generate decay heat. Decay heat results 

from the radioactive decay of fission products. The rate at which the decay heat is generated decreases the 

longer the reactor has been shut down. So the longer the spent fuel has been out of the reactor, the less 

heat that it gives off. Storing the spent fuel in a pool of water is a way to provide an adequate heat 

sink for the removal of heat from the irradiated fuel. In addition, the fuel is located far enough under 

water that the radiation emanating from the fuel is shielded by the water to adequately protect the workers 

from the radiation. 

5.8.2 Has the spent fuel pool been analyzed to determine the limits for heat removal due to 

spent fuel storage? 

Yes. The regulations give criteria that must be met for fuel storage and handling. This includes designing 

fuel-storage systems to ensure adequate safety under normal and postulated accident conditions. The 

system is to be designed with suitable shielding for radiation protection, with appropriate containment, 

confinement, and filtering systems, and with a heat-removal capability that is reliable and that can be 

tested to ensure that it meets the requirements for removing the heat produced by the spent fuel. 

5.8.3 Do spent fuel pools leak, and if they do, how much radioactive material could be 

leaked, and where would it go? 

All nuclear power plants have a reinforced-concrete spent fuel pool (SFP) structure designed to retain its 

function, even following the design-basis seismic event (that is, seismic Category I or Class I 

[earthquake]) that is anticipated for the area. The SFP also has a welded, corrosion-resistant liner. All 

plants except for one have leak-detection channels positioned behind liner plate welds to collect any 

leakage and to direct the leakage to a point at which it can easily be monitored. Nearly all nuclear power 

reactors have passive features preventing draining or siphoning of the SFP to a coolant level below the top 

of stored, irradiated fuel. Excluding paths used for irradiated fuel transfer, passive features at nearly all 

nuclear reactors prevent draining or siphoning the coolant to a level that provides inadequate shieldipg for 

fuel seated in the storage racks. 

In the event that SFP coolant inventory decreases significantly, several indicators are available to alert 

operators to that condition. The primary indication is a low-level alarm. A secondary indication of a loss 

of coolant is provided by area radiation alarms. These primary and secondary alarms indicate a loss of 

shielding that occurs when SFP coolant inventory is lost. Except for the SFP located inside the contain

ment building, the area radiation alarms are set to alarm at a level low enough to detect a loss of coolant 

inventory early enough to allow for recovery before radiation levels could make such a recovery difficult. 

The level of radioactivity in the water of the spent fuel pool is low. In addition, nuclear plants have 

cleaning systems to maintain the purity of the water in the spent fuel pools. All nuclear plants have a 

groundwater monitoring system around the facility so that if a system leaks, there is a method for alerting 

the licensee to the problem as well as for providing information regarding the location of the 

contamination. 

5.8.4 What would happen if there were a loss of heat-removal capability of water in a spent 

fuel pool when it was fully loaded?

25
The consequences of losing the heat-removal capability or water (coolant) in a spent fuel pool depends on 

the amount of time since the fuel was last used for power operation inside the reactor. If fuel was recently 

used for power operation, there may be enough decay heat to cause the spent fuel pool coolant to heat up 

to the boiling point if forced cooling were lost to the spent fuel pool. If plant operators took no action, 

boiling would cause the level in the spent fuel pool to decrease over time. However, operators have 

redundant sources of water to add to the pool to maintain coolant level should a loss of forced cooling 

occur. Operators are alerted to a loss of level condition by a series of alarms at the cooling system control 

station and in the main control room. Given the unlikely event that no operator action is taken, the pool 

level would decrease at a very slow rate (about one foot every several hours to weeks, depending on the 

age of the stored fuel). The longer the time interval since the last batch of fuel was used to generate power 

in the reactor, the longer it would take for the spent fuel to boil off the spent fuel pool coolant. Boiling the 

spent fuel pool coolant is, however, an acceptable method for cooling spent fuel and has a minimal effect 

on public health and safety. In the unlikely event that a large loss of coolant uncovers the spent fuel, if 

sufficient time has not elapsed since the fuel was used to generate power in the reactor, the spent fuel may 

have enough decay heat to overheat the cladding in air and cause it to ignite. The resulting fire could carry 

radioactive particles offsite and the consequences could be significant. However, the NRC staff considers 

this a very low probability accident because of design features required at all spent fuel storage pools that 

minimize the possibility of losing all of the spent fuel pool coolant. 

5.8.5 What would happen to the fuel in the spent fuel pool if an earthquake ruptured the 

pool, or if an airplane crashed into the pool? 

Spent fuel pools are designed to withstand earthquakes greater than any earthquake that actually occurred 

or is expected to occur in the area of the plant. Therefore, the probability of the spent fuel pool rupturing 

due to an earthquake is very low. However, in the unlikely event that a very large earthquake does occur, 

one that is larger than the pool was designed to withstand, the pool structure could fail and allow the 

coolant to drain out. The consequences of an accident like this are discussed in the response to 

Question 5.8.4, above. In the unlikely event that an aircraft crashed into the spent fuel pool, the pool 

structure could be severely damaged and not capable of maintaining coolant level. In this event, 

consequences such as those discussed in Question 5.8.4 could result. However, the staff has evaluated the 

possibility of an aircraft impacting the spent fuel pool and consider it a very low probability event. 

5.8.6 What can be done to prevent the spent fuel pool from boiling dry? 

A cooling system removes decay heat from the spent fuel pool. The coolant in the spent fuel pool is 

maintained below a specific temperature and the level of the water is maintained at a specific height over 

the spent fuel. Temperature indicators are installed and are either equipped with an alarm or require visual 

surveillance on a daily basis. High/low-water-level monitors are installed in spent fuel pools. The monitor 

alarms at the spent fuel pool and in the control room when the spent fuel pool's water level is not within 

the specified limit. 

5.8.7 How long can the licensee store the spent fuel in the spent fuel pool? 

The Commission has made a generic determination in 10 CFR 51.23 that, if necessary, spent fuel 

generated in any reactor can be stored safely and without significant impacts for at least 30 years beyond 

the licensed life for operation.

26
5.8.8 What will be done with the spent fuel pool after the fuel has been removed? 

The spent fuel pool will be decontaminated and most likely dismantled. 

5.8.9 Could the licensee transfer fuel to another licensee's facility? 

If the fuel that came out of the reactor vessel were not fully expended (had not been thoroughly used or 

"burned up"), it could be shipped to another licensee's facility for use there. This would offset some of 

the licensee's costs. Spent fuel that has no further value in producing power could also be shipped to other 

facilities. Amendment to one or both of the facility licenses would be required before fuel transfer. The 

fuel would have to be transported in a licensed cask that met NRC requirements for shipping. 

5.8.10 Can the spent fuel be shipped to another facility's spent fuel pool for storage? 

Yes. NRC regulations do not prohibit spent fuel from one facility being stored in the spent fuel pool at 

another facility. As noted in the response to Question 5.8.9, amendment to one or both of the facility licenses would be required before shipment. The spent fuel would have to be transported in a licensed 

cask that met NRC requirements for shipping. 

5.8.11 Where can I find the regulations relating to the storage of spent fuel in a spent fuel 

pool? 

Regulations regarding the storage of fuel in a spent fuel pool appear in the Code of Federal Regulations. The Code of Federal Regulations is a codification of the general and permanentrules published in the 

Federal Register by the executive departments and agencies of the Federal Government. The Code is 

divided into 50 titles, which represent broad areas subject to Federal regulation. Each title is divided into 

chapters; these usually bear the name of the issuing agency. Each chapter is further subdivided into parts 

covering specific regulatory areas. 

The regulations related to spent fuel pools are in Title 10, "Energy," Chapter I - Nuclear Regulatory Commission, Part 50, "Domestic Licensing of Production and Utilization Facilities," along with the 

regulations and standards for construction permits and operating licenses for nuclear power plants. 

See the response to Question 13.6 for instructions on how to'obtain a copy of the regulations. 

5.9 Independent Spent Fuel Storage Installation 

5.9.1 What is an independent spent fuel storage installation (ISFSI)? 

An independent spent fuel storage installation or ISFSI is a complex designed and constructed for the 

interim storage of spent nuclear fuel and other radioactive materials associated with spent fuel storage. ISFSIs may be located at the site of a nuclear power plant or at another location. The most common 

design for an ISFSI at this time is a concrete pad with dry casks containing spent fuel bundles. 

5.9.2 Why would a licensee store spent fuel in an ISFSI rather than in the spent fuel pool?

27
ISFSIs are used by operating plants that require increased spent fuel storage capability because their spent 

fuel pools have reached their capacity for holding spent fuel. Decommissioning facilities also use ISFSIs. 

Licensees that remove the spent fuel from their pools and place it in an ISFSI can then continue and/or 

complete the decommissioning process on the power generation facilities and subsequently terminate the 

facility license. In some instances, the license for the nuclear power reactor can be terminated when the 

ISFSI, which has a separate license, is located on the facility site. 

In general, it is both less expensive and easier to maintain spent fuel in dry storage in an ISFSI than in a 

spent fuel pool. The cost of an ISFSI as compared to spent fuel pool storage depends on many factors 

unique to each facility. Dry cask storage may not be less expensive than pool storage for every licensee.. 

5.9.3 What is a dry-storage cask, and how does it keep the fuel from melting or from 

causing a nuclear reaction (criticality)? 

Dry storage involves sealing used or spent fuel above ground in airtight steel (or in steel and concrete) 

containers or casks that provide both structural strength and shielding. The spent fuel is surrounded by 

inert gas inside the cask. All casks are passive designs in that they involve no mechanical devices for 

cooling or ventilation. Casks must be safety tested to withstand a variety of disasters, such as floods, 

projectiles originating from a tornado, temperature extremes, and lightning strikes. The cask is designed 

to preclude the possibility of an inadvertent criticality under all credible accidents or conditions. It must 

also provide adequate confinement, shielding, and heat removal during normal and accident conditions, 

including cask tipovers and drop accidents. Casks are placed vertically on a concrete pad or inside a 

concrete storage building or are inserted horizontally into a steel-reinforced concrete vault, depending on 

cask design. They will receive spent fuel that has been cooling in the reactor's spent fuel pool for at least 

5 years. Depending on the design, casks hold from 7 to 68 spent fuel assemblies. The maximum amount 

of heat that is generated by the radioactive decay of fission products within the spent fuel cask will be less 

than that given off by 240 100-watt light bulbs, and this amount of heat will gradually decrease with time. 

5.9.4 Who is responsible for reviewing proposed cask designs to ensure that they will safely 

confine the fuel, and what types of evaluations are required? 

The NRC is responsible for reviewing proposed cask designs to ensure that they will safely confine the 

fuel and prevent fuel cladding (which surrounds the fuel) from degrading. The NRC regulations cover the 

testing, manufacture, and maintenance of casks used in dry storage. This includes an evaluation of natural 

events (earthquakes, high winds and tornadoes, wind-driven projectiles, flooding), an evaluation of 

accidents (explosions, fires, drops, tipovers, airplane accidents), and an evaluation of sabotage. 

5.9.5 Are there any nuclear plants that already use dry-storage casks in an independent 

spent-fuel storage installation (ISFSI)? 

The first dry-storage installation was licensed by the NRC in 1986. As of May, 2000, there were 12 

nuclear power facilities using dry storage: , Surry, Oconee, H.B. Robinson, Calvert Cliffs, Fort St. Vrain, 

Palisades, Point Beach, Prairie Island, Davis-Besse, Susquehanna, Arkansas Nuclear One and North Anna. 

28

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5.9.6 Can casks be used for both shipping and storage - or will they need to transfer the 

fuel from one cask to another at some later date? 

Some casks are used solely for shipping spent nuclear fuel, some for storage purposes only, and some are 

dual purpose and can be used for both storage and shipment of the spent fuel. These casks are called dual 

purpose casks, and are licensed for both activities. If the spent fuel is stored in a cask that is not licensed 

for shipment, then it will need to be transferred to a cask licensed for shipment before it is removed from 

the site. Depending on the cask design, the transfer process can be done at the ISFSI site, or it may require 

a pool of water, such as the spent fuel pool. 

5.9.7 How long may the licensee keep the spent fuel in an onsite ISFSI? 

The NRC has determined that spent fuel can be stored onsite for at least 30 years beyond the licensed 

operating life of nuclear power plants-safely and with minimal environmental impact. This includes 

storage in the spent fuel pool or at either onsite or offsite independent spent fuel storage installations. 

5.9.8 Can the spent fuel be shipped to another facility's ISFSI for storage? 

Yes. Regulations allow for spent fuel from one facility to be stored in an ISFSI located at another facility 

with appropriate license conditions (this only applies if the second facility has a specific type of license 

called a "site specific Part 72" license in reference to the location in the regulations that describe the 

license requirements). The spent fuel would have to be transported in a cask that met NRC requirements 

for shipping casks. 

5.9.9 Why is the licensing process on the independent spent-fuel storage installation (ISFSI) 

evaluated separately from the decommissioning process? 

Both operating plants and plants that have permanently ceased operations and are decommissioning use 

ISFSIs. ISFSIs are not unique to decommissioning plants. The initial development of the 

decommissioning regulations occurred in the early 1980's. At that time the NRC and the industry assumed 

that by the time facilities begun decommissioning the U.S. Department of Energy's (DOE) high level and 

waste repository would be accepting spent fuel for ultimate disposal. Therefore, spent fuel onsite during 

decommissioning was not expected to be an issue. Consequently development of regulations related to 

ISFSI's occurred separately from the development of decommissioning regulations. Since the ISFSI may 

in some cases remain at the site longer than a nuclear facility that is undergoing immediate 

decommissioning, it is appropriate that ISFSI be capable of being licensed separately. The 

decommissioning of the ISFSI is also handled separately from the decommissioning of the nuclear power 

plant. Site specific ISFSI licenses requires the evaluation of the ISFSI separately from the remainder of the 

facility, although other site activities adjacent to the ISFSI are considered to evaluate their impact on the 

storage of the spent fuel. An ISFSI can be constructed and operated either under the same license that is 

used for the operating or decommissioning facility (called a "Part 50 license" in reference to the location 

in the Code of Federal Regulations that describes the license requirements), or under a site-separate 

license (called a "Part 72 license" in reference to the location in the Code of Federal Regulations that 

describes the licensing requirements for the ISFSI). 

5.9.10 Where are the regulations relating to use of ISFSIs?

29
Regulations regarding the ISFSIs appear in the Code of Federal Regulations. The Code of Federal 

Regulations is a codification of the general and permanent rules published in the Federal Register by the 

executive departments and agencies of the Federal Government. The Code is divided into 50 titles, which 

represent broad areas subject to Federal regulation. Each title is divided into chapters; these usually bear 

the name of the issuing agency. Each chapter is further subdivided into parts covering specific regulatory 

areas. The regulations related to ISFSIs are in Title 10, "Energy," Chapter I - Nuclear Regulatory 

Commission, Part 72, "Licensing Requirements for the Independent Storage of Spent Nuclear Fuel and 

High-Level Radioactive Waste." 

See the response to Question 13.6 for instructions on how to obtain a copy of the regulation. 

5.9.11 Can the licensee use the site for other activities, such as a combustion turbine or 

natural gas plant, while fuel is stored onsite in an ISFSI? 

The NRC will conduct a site-specific analysis if the licensee requests that the site be used for other 

activities, such as a combustion turbine or natural gas plant, while the fuel is stored onsite in an ISFSI. 

The decision as to whether to allow the alternate activities will depend on the results of the site-specific 

analysis. 

6 RADIOACTIVE LOW-LEVEL WASTE 

6.1 What is meant by low-level radioactive waste, and how is it different from 

fuel? 

Low-level waste (LLW) is any radioactive waste that is not classified as high-level waste, spent nuclear 

fuel, transuranic waste (containing manmade elements heavier than uranium that emit alpha 

radiation--transuranic waste is produced during reactor fuel assembly, weapons fabrication, and chemical 

processing operations), or uranium or thorium mill tailings. LLW often contains small amounts of 

radioactivity dispersed in large amounts of material, but may also have activity levels requiring shielding 

and remote handling. It is generated by reactor fuel production, reactor operations, isotope production, 

medical procedures, and research and development activities. LLW usually comprises the following 

material contaminated with radionuclides: rags, papers, filters, solidified liquids, ion-exchange resins, 

tools, equipment, discarded protective clothing, dirt, construction rubble, concrete, or piping. 

NRC regulations classify LLW on the basis of potential hazards, such as the concentrations of short-lived 

and long-lived radionuclides, in accordance with 10 CFR 61.55. Thus, LLW usually, but not necessarily, 

includes waste with relatively low concentrations of radionuclides. Although the classification of waste 

can be complex, Class A waste generally contains lower concentrations of longer half-life radioactive 

material than Class B and C wastes. Greater than Class C waste are not considered low-level radioactive 

waste and must be handled and disposed of differently from Class A, B and C wastes. See 10 CFR 61.55, 

"Waste Classification," for a detailed description of waste classification. 

30

1!
6.2 How is the low-level radioactive waste disposed of.

Low-level waste is commonly disposed of by burial in near-surface shallow trenches. After they are filled 

with containers, the trenches are usually covered with a low-permeability cover (such as clay). They are 

then often covered with a gravel drainage layer and a layer of topsoil. Vegetation is planted on top for 

erosion control. There is no intent to recover the wastes once they are disposed of. The volume of waste 

that is being disposed of each year is decreasing as the result of industry efforts to compact or incinerate 

part of the waste. 

6.3 Is low-level waste (LLW) disposal safe? 

LLW facilities are sited in areas that are away from surface water and where the groundwater is located at 

depths sufficiently beneath the trenches to minimize nuclide migration. Sites and the surrounding areas 

are monitored using a system of wells to determine if there is any leakage of radioactivity into the 

groundwater. 

A combination of natural site characteristics and engineered safety features is used to ensure the safe 

disposal of LLW. In addition, restrictions of types and amounts of waste disposed of at a site, as well as 

the analysis performed as part of the licensing to demonstrate compliance with performance objectives in 

NRC regulations, increase the safety of LLW disposal. 

The natural characteristics of an LLW disposal site are relied on in the long term, and they should promote 

disposal-site stability and attenuate the transport of radionuclides away from the disposal site into the 

general environment. Sites generally must possess the following characteristics: (1) relatively simple geology, (2) well-drained soils free from frequent ponding or flooding, (3) lack of susceptibility to surface 

geological processes, such as erosion, slumping, and landslides, (4) a water table of sufficient depth so that 

groundwater will not periodically intrude into the waste or discharge onsite, (5) lack of susceptibility to 

tectonic processes, (6) no known potentially exploitable natural resources, (7) limited future population growth or development, and (8) capability of not being adversely impacted by nearby facilities and 

activities. 

Engineered barriers are manmade structures designed to improve the site's natural capability to isolate and 

contain waste. They consist of various engineered system components, including the following: (1) a 

layered earthen cover, (2) a disposal vault, (3) a drainage system, (4) waste forms and containers, 

(5) backfill material, and (6) an interior moisture barrier and low-permeability membrane. 

Regulations specify the allowable radiation dose from the LLW facilities to the workers and to the public. 

6.4 Where can low-level radioactive waste be disposed of? 

There are currently three active, licensed disposal facilities. All three sites are located in Agreement States 

and are regulated by the States.0') They are located in Barnwell (South Carolina), Hanford (Washington), 

(a) The concept of Agreement states was set up by the Atomic Energy Act, and it permits NRC to 

relinquish to the States (on a state-by-state basis), cdrtain of its authoAityo oregulate specific areas, 

31
and Clive (Utah). The site in Utah is restricted to specific types of low-level waste (LLW). The site in 

Washington State is restricted to waste from the Northwest and Rocky Mountain regional compacts. The 

site in South Carolina accepts LLW from all States except North Carolina. There are several additional 

sites currently under consideration for LLW disposal. 

The Low-Level Radioactive Waste Policy Amendments Act of 1985 authorized the formation of regional 

compacts. A compact is a group of States (not necessarily with contiguous borders) that have decided to 

share resources to develop and maintain an low-level waste disposal site. Nine compacts are currently 

active, although most of them do not have a low-level waste site that has been developed or licensed. The 

Act contains a system of milestones, incentives, and penalties to ensure that States and compacts will be. 

responsible for their own waste. Compacts can restrict access to the disposal site from States outside the 

compact. 

6.5 What would happen if the waste site that was being used is closed? 

If the low-level waste site is closed, the licensee would need to find another waste site that is willing to 

take the waste. Options such as compaction or incineration would be investigated as a means of reducing 

the magnitude of this waste stream. Until a new waste site is located, the licensee would need to 

temporarily store the waste onsite. 

6.6 Can the low-level radioactive waste be stored at the site in the event that 

the waste site is closed? What type of facility is required and how long can 

the waste be left at the site? 

The NRC has historically discouraged the use of onsite storage as a substitute for permanent disposal, but 

has not limited the amount of time that the waste can be stored. However, low-level waste (LLW) is 

normally stored onsite on an interim basis before being shipped offsite for permanent disposal. Onsite 

storage facilities are designed to minimize personnel exposure. High-dose-rate LLW is isolated in a 

shielded storage area and is easily retrievable. The lower dose-rate LLW is stacked or stored to maximize 

packing efficiencies. The NRC has guidelines regarding the storage facility, including the following: (1) 

shielding used should be controlled by dose rate criteria for both the site boundary and any adjacent offsite 

areas and (2) a liquid drainage collection and monitoring system should be present. The drain should be 

routed to a radwaste processing system. 

6.7 Can radioactive waste be buried onsite? 

Onsite burial is not acceptable unless (1) the licensee applies for permission and is specifically authorized 

under 10 CFR 20.2002 and (2) such burial is not prohibited by compact (that is, approved by the State). 

6.8 Can the reactor vessel be disposed of at a low-level waste (LLW) site? 

Can the reactor vessel be shipped intact, or does it have to be segmented? 

such as the disposal of low-level radioactive waste.

32
A reactor vessel can not be disposed of at a LLW site unless it meets waste classification requirements specified in the regulations and any site-specific requirements specified in the disposal facility's license. 

In most cases where disposal of the reactor vessel has occurred, the reactor vessel internals have been 

.removed before any parts of the reactor vessel were shipped to an LLW disposal site. In two cases, the 

Trojan Nuclear Plant and the Saxton Facility, the reactor vessel was removed from the building and 

shipped intact to a LLW disposal site. Licensees are required to demonstrate that the shipment meets the 

regulations for package integrity and that the package meets the acceptance criteria for the LLW disposal 

site, such as criteria for radionuclide concentration and waste form. 

6.9 How are liquid wastes disposed of? 

Liquid wastes are processed onsite. The liquid portion is separated from the solid portion. The solid 

portion is disposed of in the low-level waste site, as long as it meets the criteria for low-level waste. The 

concentration of radioactive material in the liquid portion is measured, and if the concentration is below 

the Federal limits for release of effluents, the liquid portion may be released offsite (for instance, to sewers 

or a nearby body of water). Otherwise, the liquid portion is solidified (by mixing with concrete or similar 

solidifying or absorbing material) and disposed of as solid low-level waste. 

6.10 Where are the regulations relating to radioactive low-level waste (LLW) 

disposal? 

Regulations related to radioactive low-level waste disposal appears in the Code of Federal Regulations. The Code of Federal Regulations is a codification of the general and permanent rules published in the 

Federal Register by the executive departments and agencies of the Federal Government. The Code is 

divided into 50 titles, which represent broad areas subject to Federal regulation. Each title is divided into 

chapters; these usually bear the name of the issuing agency. Each chapter is further subdivided into parts 

covering specific regulatory areas. 

The regulations related to LLW disposal are in 10 CFR Part 61 and 10 CFR Part 20 Subpart K. See the 

response to Question 13.6 for instructions on how to obtain a copy of the regulations. 

7 TRANSPORTATION 

7.1 How is the spent fuel going to be shipped to a final repository? 

The spent fuel will be shipped in specially designed shipping casks. The design, construction, use, and 

maintenance of these commercial shipping containers are regulated by the NRC. In some cases, a 

dual-purpose cask, which is designed to work as a storage cask as well as a shipping cask, will be used. 

The shipping casks can be loaded onto trucks, trains, or barges for shipment.

33
7.2 How is low-level radioactive waste shipped to the disposal site? 

Most low-level radioactive waste is shipped in packages authorized by the U.S. Department of 

Transportation, although some packages for larger quantities of low-level waste require NRC certification. 

Low-level radioactive waste packages can be loaded onto trucks or trains for shipment to the low-level 

waste site. 

7.3 Are there regulations on radiation levels during transportation? 

Yes. There are regulations governing the radiation level that stipulate limits (1) on the outer surface of the 

vehicle that is carrying the radioactive material, (2) at 6.6 feet (2 meters) from the surface of the vehicle, 

and (3) at the position occupied by the driver of the vehicle. Measurements of the applicable radiation 

levels are required before a vehicle is allowed to leave with packages. 

7.4 Are there safety criteria for spent fuel shipping casks, and how are the 

criteria satisfied? 

Yes. Safety standards for spent fuel shipping casks are detailed in NRC regulations. These regulations are 

compatible with safety standards issued by the International Atomic Energy Agency. Casks must be 

designed to withstand a series of tests that simulate accident as well as normal conditions of transportation. 

The normal conditions of transportation that must be considered include temperature, pressure, vibration, 

water spray, impact, penetration, and compression tests. The following tests are used to provide 

reasonable assurance that the casks will withstand serious transportation accidents: 

"* a 30-foot drop onto a flat, unyielding surface 

"* a 40-inch drop onto a vertical steel rod (puncture test) 

"* a 30-minute exposure to a fire of 14757F 

"* submersion in 50 feet of water. 

A cask design must be reviewed by the NRC staff to verify its resistance to accidents. Applicants must 

demonstrate to NRC that their design satisfies all applicable requirements. That demonstration may 

involve comparative evaluations with approved designs, analyses, and partial-scale tests. An approval 

certificate must be issued by the NRC before a particular cask design can be used to transport spent fuel. 

7.5 Are there safety criteria for shipping containers for low-level waste (LLW), 

and how are the criteria satisfied before shipment? 

Yes. LLW is shipped in containers that are designed to NRC or U.S. Department of Transportation (DOT) 

standards. Packaging requirements for shipments of specific radioactive materials are based on a number 

of factors, including the material activity, quantity, form (normal or special), specific activity, fissile 

properties, and other characteristics (physical, chemical, and nuclear properties). For smaller quantities of 

LLW, one of three types of containers is used, depending upon the material activity, or in some cases, the 

34

a
specific activity. These container types are "strong tight containers," industrial packages, and Type A 

containers. Safety criteria for these types of containers, which are found in the DOT regulations, increase 

along with the package categories from "strong tight" to Type A containers. Type A containers must be 

able to withstand the normal conditions of transport as described in the answer to Question 7.4. 

Wastes that contain higher levels of radioactivity are transported in Type B containers. Type B containers 

must be able to withstand the normal and accident conditions and are certified by the NRC as described in 

the response to Question 7.4. 

For all'types of containers, additional regulations apply that (1) specify preliminary determinations that 

must be ascertained before the first use of any container (for example, no cracks, no defects) and 

(2) specify routine determinations that must be satisfied before each shipment (for example, proper 

package for specific contents, properly installed closure devices). 

7.6 What regulations apply to the transportation of radioactive material? 

The transportation of radioactive materials is regulated jointly at the Federal level by the U.S. Department 

of Transportation (DOT) and the NRC. The responsibilities of the two agencies are delineated in a 

memorandum of understanding. (See Federal Register of July 2, 1979.) In general, the areas regulated by 

the agencies are as follows: 

DOT- Regulates shippers and carriers of radioactive material and the conditions of transport (including 

routing, tiedowns, radiological controls, vehicle requirements, hazard communication, handling, storage, 

emergency response information, and employee training). DOT regulations are located in the Code of 

Federal Regulations, Title 49, "Transportation." 

NRC - Regulates users of radioactive material and the design, construction, use, and maintenance of 

shipping containers used for larger quantities of radioactive material and fissile material (such as 

uranium). NRC regulations are located in the Code of Federal Regulations, Title 10, "Energy," Part 71, 

"Packaging and Transportation of Radioactive Material." 

See the response to Question 13.6 for instructions on how to obtain a copy of the regulations. 

7.7 Will the public be informed beforehand of low-level waste shipments or 

shipments of spent fuel and the routes taken? 

No. The NRC does not monitor, or inform the public, or require the licensees to inform the public about 

the timing of shipments of low-level radioactive material wastes or die routes that the shipments will take. 

NRC does require licensees to provide advance notification to State governments regarding shipments of 

spent fuel and other specific types of waste shipments. 

7.8 Are specific routes used for transporting radioactive material, and does the 

NRC approve the routes used for radioactive material shipments? 

Specific routes are generally not required for transporting low-level radioactive materials. DOT's Federal 

Highway Administration has established routing requirements for spent fuel shipments. Essentially, these

35
requirements limit these shipments to Interstate System Highways and city bypasses that minimize time in 

transit. Under NRC's regulations for physical protection of spent fuel, licensees are required to request 

and obtain advance approval from the NRC of the routes used for road and rail shipments of spent fuel and 

of any U.S. ports in which vessels carrying spent fuel shipments are scheduled to stop. 

7.9 How safe are shipments of spent fuel casks? What would happen if the 

train or truck carrying a spent fuel cask was involved in an accident? 

According to the analyses performed and the statistical record to date, shipments of spent fuel casks are 

very safe. There have been no recorded injuries or fatalities attributed to spent fuel in over 

3,000 shipments in the U.S. over the last 30 years. In 1986, the NRC contracted with Lawrence Livermore 

National Laboratory to perform a study to determine the level of safety provided during shipments of spent 

fuel from U.S. commercial nuclear power plants. The study estimated that about one accident in every 80

million shipment miles could cause cask damage that would be significant enough to cause a radiological 

hazard that could equal or slightly exceed existing compliance values. Further information on this study 

can be found in NUREG/BR-01 11, "Transporting Spent Fuel: Protection Provided Against Severe 

Highway and Railroad Accidents." 

7.10 What emergency response plans are in place for transportation accidents 

involving a spent fuel cask or low-level-waste packages? 

DOT requires carriers to receive emergency-response training that includes written procedures. In 

addition, DOT requires emergency-response procedures to accompany each shipment. 

If an accident occurs, State and local governments are primarily responsible for overseeing the response of 

the carrier, shipper, and others, and for taking any actions deemed necessary to protect public health and 

safety. 

The authorities that are likely to respond to transportation incidents (for example, police, fire fighters) 

have been provided with Emergency Response Guidebooks by DOT; these identify the potential hazards 

of materials and discuss corresponding mitigative actions. Each State has emergency plans in place for 

responding to incidents involving transportation of radioactive material. To assist the States in the 

development of these plans, a guide exists entitled "Guide and Example Plan for Development of State 

Emergency Response Plans and Systems for Transportation-Related Radiation Incidents"; it was prepared 

by the Western Interstate Nuclear Board and Regional Training Committee, Region VIII, Denver, 

Colorado, in 1975. An NRC- sponsored survey was conducted to ascertain the States' readiness to 

respond to radiological transportation incidents. In the responses to the survey, the States indicated that 

they have the necessary plans and procedures in place. This survey is documented in NUREG/CR-5399, 

"Survey of State and Tribal Emergency Response Capabilities for Radiological Transportation Incidents," 

which was prepared by Indiana University in Bloomington, Indiana, early in 1990. 

To assist State and local governments, the Federal Government has established an interagency plan called 

the Federal Radiological Emergency Response Plan under the coordination of the U.S. Department of 

Energy (DOE), which charges eight regional coordinating offices with the responsibility and authority for 

convening radiological assistance teams. Upon request from a State, DOE will respond immediately to an 

emergency, including providing assistance at the scene. 

36

31
8 LICENSE TERMINATION AND 

THE ULTIMATE DISPOSITION OF THE FACILITY 

8.1 How does decommissioning end, and who decides that the 

decommissioning is complete? 

Licensees must submit an application for license termination at least 2 years before the requested 

termination date of the license. The license-termination plan must include 

* a site characterization 

• identification of remaining dismantlement activities 

* plans for site remediation 

detailed plans for the final survey of residual contamination on the site 

* a method for demonstrating compliance with the radiological criteria for license termination. For 

restricted release, the license-termination plan should include a description of the site's end use and 

documentation on public consultation, institutional controls, and financial assurance needed to comply 

with the requirements for license termination for restricted release or alternative criteria. 

a an updated site-specific estimate of remaining decommissioning costs 

* a supplement to the environmental report that describes any new information or significant 

environmental changes associated with the licensee's proposed termination activities. 

After receiving the license-termination plan, the NRC places a notice of the receipt of the plan in the 

Federal Register and makes the plan available to the public for comment. The NRC also schedules a 

public meeting near the facility to discuss the plan's contents with the public. Because this is an action 

that involves a license amendment, there is also an opportunity for members of the public to request a 

hearing. If the license-termination plan demonstrates that the remainder of decommissioning activities 

will be performed in accordance with the NRC's regulations, is not detrimental to the health and safety of 

the public, and does not have a significant effect on the quality of the environment, then the Commission 

approves the plan by a license amendment, subject to whatever conditions and limitations the NRC deems 

appropriate and necessary. At this point, the licensee may achieve license termination by successfully 

demonstrating implementation of the license-termination plan, as approved by NRC. 

The NRC will determine if the remaining dismantlement is performed in accordance with the approved 

license-termination plan. If this is the case, and if the final radiation survey and associated documentation 

demonstrate that the facility and site are suitable for release, then the Commission terminates the license, 

and the decommissioning process is considered to have been completed.

37
8.2 Why is the license-termination plan filed so late in the process?

The initial decommissioning activities (decontamination and dismantlement) are not significantly different 

from routine operational activities such as replacement or refurbishment. Because of the framework of 

regulatory provisions that are in place in the licensing basis for each facility, these activities do not present 

significant safety issues for which a detailed plan such as the license-termination plan is warranted. 

Therefore, it is appropriate that the licensee be permitted to conduct these activities without the need for a 

license amendment. At the license-termination stage (towards the end of the decommissioning process), 

the Commission must consider (1) the licensee's plan for assuring that adequate funds will be available for 

final site release, (2) the radiation-release criteria for license termination, and (3) the adequacy of the plans 

for the final survey that is required to verify that the release criteria have been met. 

8.3 How will the licensee know where the radioactive material or 

contamination is located within the plant? 

During operation, the plant is required to keep records of radiological surveillances that document where 

contaminations occur and locations of activation products and other sources of radiological materials. 

This is the basis for a plant's initial assessment of the location of radioactive material and contamination. 

Radiological surveys continue during all phases of decommissioning, and the records of these surveys are 

also kept. This information is used as part of the basis for the site-characterization plan; however, 

additional measurements are made at the site-characterization stage. The characterization plan must be 

designed to demonstrate compliance with appropriate dose or risk-based regulations. 

8.4 What is included in the site characterization? 

The purpose of the site characterization is to ensure that the final radiation surveys are conducted to cover 

all areas where contamination existed, remains, or has the potential to exist or remain as well as to provide 

data for planning further decommissioning activities. The site characterization contains a description of 

(1) the radiological contamination on the site before any cleanup activities associated with 

decommissioning took place, (2) a historical description of site operations, spills, and accidents, (3) a map 

of remaining contamination levels and contamination locations, and (4) a description of the survey 

instruments and supporting quality assurance practices used in the site-characterization program. 

8.5 What does "suitable for release" mean? Are there any restrictions on how 

the site can be used? 

There are two broad categories of uses for the facility after the license termination. The first is 

"6unrestricted use," and the second is "restricted use." These will be discussed separately. 

Unrestricted use means that there are no restrictions on how the site may be used. The licensee is free to 

continue to dismantle any remaining buildings or structures and to use the land or sell the land for any type 

of application. 

Restricted use means that the licensee has demonstrated that further reductions in residual radioactivity 

would result in net public or environmental harm, or residual levels are as low as is reasonably achievable.

38
The licensee must have made provisions for legally enforceable institutional controls (for example,. 

restrictions placed in the deed for the property describing what the land can and cannot be used for), which 

provide reasonable assurance that the radiological criteria set by the NRC will not be exceeded. In 

addition, the licensee must have provided sufficient financial assurance to an amenable independent third 

party to assume and carry out responsibilities for any necessary control and maintenance of the site. There 

are also regulations relating to the documentation of how the advice of individuals and institutions in the 

community who may be affected by the decommissioning has been sought and incorporated in the license

termination plan related to decommissioning by restricted use. 

Although power reactor licensees can choose either a restricted or unrestricted option for release, the 

restricted option is primarily for materials licensees and would not normally be selected by reactor 

licensees because of the low levels of site contamination. 

8.6 Why would the licensee be allowed to restrict use of the site? 

There can be situations in which restricting site use can provide protection of public health and safety by 

reducing the total effective dose equivalent in a more reasonable and cost-effective manner than 

unrestricted site use. This protection is afforded by limiting access to the site, limiting the amount of time 

that an individual spends onsite, or by restricting agricultural or drinking water use. For many facilities,. 

the time period requiring this type of restriction can be fairly short, and need only be long enough to allow 

radioactive decay to reduce radioactivity to levels that permit the Site to be released for unrestricted use. 

8.7 What is residual radioactivity, and why is it important to the termination 

of the license? 

The term "residual radioactivity" means radioactivity in structures, materials, soils, groundwater, and other 

media at a site resulting from activities under the licensee's control. This includes radioactivity from all 

licensed and unlicensed sources used by the licensee, but excludes background (natural) radiation. It also 

includes radioactive materials remaining at the site as a result of routine or accidental releases of 

radioactive material at the site and previous disposals at the site. Criteria for the termination of the license 

are based on the residual radioactivity levels remaining at the site at the end of decommissioning. 

8.8 What are the criteria for residual radioactivity at the site at the end of 

decommissioning, assuming that the licensee is planning for unrestricted 

use of the site? 

The Commission has established a dose of 25 millirem (0.25 millisievert) per year total effective dose 

equivalent to an average member of the critical group as an acceptable criterion for release of any site for 

unrestricted use. The dose limit includes the dose from drinking groundwater. The licensee will be 

required to show that the site can meet this criterion before the license will be terminated for unrestricted 

use. In addition, the licensee will need to show that the amounts of residual radioactivity have been 

reduced to levels that are as low as reasonably achievable. This concept, known as ALARA, means that 

all doses are to be reduced below required levels to the lowest possible level considering economic and 

societal factors. Determination of levels that are ALARA must take into account consideration of any

39
detriments, such as deaths from transportation accidents that are expected to potentially result from 

decontamination and waste disposal. 

8.9 Why is there a difference between the residual radioactivity limits set by 

the NRC and those set by the EPA? Why do the licensees not have to use 

the limits set by EPA? 

NRC continues to rely on the findings from two international organizations, the International Commission 

on Radiation Protection (ICRP) and the National Council on Radiation Protection and Measurements 

(NCRP). Both organizations have acknowledged the difficulty in setting acceptable levels of risk for the 

public; however, both ICRP and NCRP have established a dose of 100 mrem/yr to an individual member 

of the public as the level that is acceptable for exposure to radiation from sources other than medical 

procedures. The ICRP and the NCRP further established the need to reduce this annual dose rate by using 

the principle of "optimization," considering the cost effectiveness of additional dose reduction. Following 

these recommendations, the NRC adopted a level of 25 mrem/year as the value for residual radioactivity at 

a site under consideration for license termination. 

EPA's radiation dose limit of 15 mrem/year results from a different technical analysis for establishing an 

acceptable risk to the public and a value for residual radioactivity other than that of NRC where radiation 

is the only.contaminant considered. In addition, the NRC also has a "cleanup" requirement of"As Low 

As Reasonably Achievable" (ALARA) as discussed in the response to Question 8.8. The use of the 

ALARA requirement usually results in cleanups that are below the EPA's requirements also. 

Nuclear reactors are licensed by the NRC, and the NRC is responsible for making the safety and 

environmental determination for termination of the license. Therefore, licensees are required to meet the 

NRC's requirements for residual radioactivity. However, since the NRC value of 25mrem/year is a limit 

a licensee can choose to further reduce the value of residual radioactivity at a site to achieve annual dose 

values less than 25 mremlyear. 

8.10 What is a "total effective dose equivalent?" 

The total effective dose equivalent is a term that is used to express how the radiation dose is calculated to 

an individual. It means that the dose from radioactive material outside of the individual (external 

radiation) and the dose from any radioactive material that the individual may have inhaled or ingested 

(internal radiation) have been considered. For the latter case, the internal radiation dose is considered for a 

period of 50 years following the intake of the radioactive material. In addition, weighting factors are used 

that are specific to the body organs or tissues that are irradiated. These weighting factors are used to 

account for the variation in sensitivity of different organs or tissues to radiation. 

8.11 Who would be considered an "average member of the critical group?" 

The "critical group" means the group of individuals reasonably expected to receive the highest exposure to 

residual radioactivity within the assumptions of a particular scenario. The average dose to a member of 

the critical group is represented by the average of the doses for all members of the critical group, which in 

turn is assumed to represent the most likely exposure situation. For example, when considering whether it 

is appropriate to "release" a building (allow people to work in the building without restrictions) that has 

40

11
been decontaminated, the critical group would be the group of regular employees that would work in the 

building. If radiation in the soil is the concern, then the scenario used to represent the maximally exposed 

individual is that of a resident farmer. The assumptions used for this scenario are "prudently 

conservative" and tend to overestimate the potential doses. The added sensitivity of certain members of 

the population, such as pregnant women, infants, and children, are accounted for in the analysis. 

However, the most sensitive member may not always be the member of the population that receives the 

highest dose. This is especially true if the most sensitive member (for example, an infant) does not 

participate in specific activities that may provide the greatest dose or if they do not eat specific foods that 

cause the greatest dose. 

8.12 Have any facilities been decontaminated to the point of being allowed 

unrestricted release? 

Yes, two commercial reactors, Shoreham and Fort St. Vrain, have had their licenses terminated and the 

sites released for unrestricted use. The Pathfinder reactor has had its license terminated, and most of the 

site has been released for unrestricted use. In addition, many non-reactor facilities, such as hospitals and 

industrial and pharmaceutical manufacturers have, had their licenses terminated and the site returned to 

unrestricted use. 

8.13 What are the criteria for residual radioactivity at the site at the end of 

decommissioning, assuming that the licensee is planning for restricted 

use of the site? 

The Commission has also established criteria for restricted use of the site. These are more complex than 

for unrestricted use and employ a tiered approach. Residual radioactivity at the site must have been 

reduced so that there would be reasonable assurance that the total effective dose equivalent from the 

residual radioactivity to the average member of the critical group would not exceed 25 millirem 

(0.25 millisievert) per year with institutional controls in place and either 100 millirem (I millisievert) per 

year or 500 millirem (5 millisievert) peryear with no institutional controls. Institutional controls include 

such engineered controls as fences and such restrictions on the site's deed that activities like a park or 

farming would not be allowed. Institutional control could also include ownership by the Federal or State 

government, thus providing for a legal mechanism to restrict public access. 

For the first case, 100 millirem (I millisievert) per year, the licensee must demonstrate the following: 

"Further reductions in residual radioactivity would result in net public or environmental harm, or they 

were not made because the residual levels are as low as reasonably achievable, taking into account the 

consideration of any detriments, such as traffic accidents, that may be expected to potentially result 

from decontamination and waste disposal. 

"* Provisions have been made for legally enforceable institutional controls to provide assurance that the 

25 millirem (0.25 millisievert) per year average dose to the average member of the critical group will 

not be exceeded.

41
" Funds have been placed into an account segregated from other assets and outside of the licensee's 

administrative controls that will be used to pay for any necessary control and maintenance of the site, 

or that a surety method, insurance, or other guarantee method has been established. 

" The licensee has sought advice from affected parties and in seeking that advice provided for 

(1) participation by representatives of a broad cross section of community interests, (2) an opportunity 

for a comprehensive collective discussion on the issues, and (3) publicly available summary of the 

results of all such discussions. 

For the second case of 500 millirem (5 millisievert) per year, the licensee must demonstrate the following: 

" Further reductions in residual radioactivity necessary to comply with the 100-millirem 

(l-millisievert)-per-year value are not technically achievable, are prohibitively expensive, or would 

result in net public or environmental harm. 

" Provisions have been made for legally enforceable and durable institutional controls (which may also 

include Federal, State, or local government control of sites), as well as provisions for a verification of 

the continued effectiveness of the institutional controls at the site every 5 years after license 

termination to ensure that the institutional controls are in place, and the restrictions are working. 

" Sufficient financial assurance has been provided to enable a responsible government entity or 

independent third party to carry out periodic rechecks of the site no less frequently than every 5 years 

to ensure that the institutional controls remain in place as necessary to meet the 25-millirem 

(0.25-millisievert)-per-year criterion. Sufficient financial assurance must also be provided to assume 

and carry out responsibilities for any necessary control and maintenance of those controls. 

" The licensee has sought advice from affected parties and in seeking that advice provided for 

(1) participation by representatives of a broad cross section of community interests, (2) an opportunity 

for a comprehensive collective discussion on the issues, and (3) a publicly available summary of the 

results of all such discussions. 

In addition, alternate criteria exist for the case in which the 25-millirem-per-year limit is found to be 

inappropriate. In this situation, it must be unlikely that the dose from all manmade sources combined, 

other than medical, would be more than 100 millirem (1 millisievert) per year. These alternate criteria are 

expected to be used only in rare cases. The licensee must 

"* submit an analysis of possible sources of exposure to provide assurance that public health and safety 

would continue to be protected 

"* demonstrate that it has employed, to the extent practical, restrictions on the site use 

"• reduce doses as low as reasonably achievable, taking into consideration any detriments, such as traffic 

accidents, that are expected to potentially result from decontamination and waste disposal 

" submit a license-termination plan, which specifies that it plans to decommission by using alternate 

criteria and documents how the licensee has sought and addressed advice from affected parties and in 

seeking that advice provided for (1) participation by representatives of a broad cross section of

42
community interests, (2) an opportunity for a comprehensive collective discussion on the issues, and 

(3) a publicly available summary of the results of all such discussions. 

The use of alternate criteria to terminate a license requires the approval of the Commission after a 

consideration of the NRC staff's recommendations that address any comments provided by the 

U.S. Environmental Protection Agency or the public. 

8.14 How does the dose based on the residual radioactivity levels relate to 

background dose levels? 

This dose can be compared with the background dose of approximately 300 millirem (3 sievert) per year 

that is anticipated to the average person living in the United States. Background radiation means radiation 

from cosmic sources, naturally occurring radioactive material, including radon, and global fallout as it 

exists in the environment from the testing of nuclear explosive devices or from earlier nuclear accidents, 

such as Chornobyl, that contributes to background radiation and is not under the control of the licensee. 

"Distinguishable from background" means that the detectable concentration of a radionuclide is 

statistically different from the background concentration of that radionuclide in the vicinity of the site. 

8.15 Why didn't the NRC set the final dose criteria for release of the site for 

unrestricted use to "pre-existing background" levels? 

For those facilities in which soil or building contamination exists, it would be extremely difficult to 

demonstrate that an objective of "return to background" had been achieved. In addition, the removal of 

soil or concrete to "pre-existing background" levels is generally not desirable from the perspective of risk 

to public health and safety and protection of the environment. For example, at some point, the removal of 

increasingly larger volumes of concrete and soil would also result in a greater net risk from transportation 

accidents. 

8.16 Is it possible that some isotopes are located in such a way that 

radiation-monitoring devices cannot accurately detect their levels of 

radioactivity? 

It is unlikely that radioactive material located inside a piece of equipment or a structure is not detected 

during the final radiation survey. The structures, systems, and components that have radioactive 

contamination exceeding NRC's limits will be decontaminated or dismantled and shipped to a 

low-level-waste disposal site. The licensee must keep records of information during the operating phase 

of the facility that could be used to identify where spills or other occurrences involving the spread of 

contamination in and around the facility, equipment, or site have been located. 

8.17 Will continued monitoring be required after the decommissioning 

process is complete to ensure that the radiation levels do not increase? 

No. For'sites that have been determined to be acceptable for unrestricted use, there are no requirements 

for further measurement of radiation levels. It is not expected that these radiation levels would

43
change--other than to be reduced over time-because the radioactive material will have been removed 

from the site, and there would be no mechanism for further contamination or radiological releases. 

For sites that have been determined to be acceptable for license termination under restricted conditions, 

additional measurements of radiation are only required for sites that have residual radioactivity in excess 

of 100 millirem (1 millisievert) per year, but less than 500 millirem (5 millisievert) per year. These 

measurements are to be made by a responsible government entity or independent third party, including a 

governmental custodian of a site. The measurements are to be carried out no less frequently than every 

5 years to ensure that the institutional controls remain in place as necessary to meet the criterion of 

25 millirem (0.25 millisievert) per year to an average member of the critical group. 

8.18 What types of uses can be made of the plant site after decommissioning is 

completed? 

Once the license has been terminated and the site released for unrestricted use, there are no restrictions on 

the type of use. Possible uses could range from restoring the natural habitat, to farming, to continued use 

as an. industrial site (possibly leaving buildings and installing a gas-, coal-, or oil-powered generating 

plant). If a site is being decommissioned under restricted-release or alternate-release criteria, then NRC 

approval must be obtained, on a case-by-case basis, for the future uses of the plant site before the license is 

terminated. 

8.19 Could the licensee initiate an alternative use of the site or partial site 

release before the decommissioning process is completed? 

Requests by licensees to initiate alternative uses of the site or site partial release before the 

decommissioning process is completed would be reviewed by the Commission on a case-by-case basis. 

The NRC staff has proposed a rulemaking plan that would procedurize the process by which a licensee 

would release part of its reactor facility or site for unrestricted use before receiving review and approval 

of its license termination plan. 

8.20 What uses have been made of sites that were decommissioned in the 

past? 

The licensee that held the license for the Fort St. Vramn nuclear plant in Colorado has chosen to build a 

natural-gas-powered boiler for use with the existing turbine generator. The Pathfinder site has a 

natural-gas electric-generating plant. The Shoreham site is currently not being used. 

8.21 What regulations are related to license termination? 

Regulations regarding license termination appear in the Code of Federal Regulations. The Code of 

Federal Regulations is a codification of the general and permanent rules published in the Federal Register 

by the executive departments and agencies of the Federal Government. The Code is divided into 50 titles, 

which represent broad areas subject to Federal regulation. Each title is divided into chapters; these usually 

bear the name of the issuing agency. Each chapter is further subdivided into parts covering specific 

regulatory areas. 

44

VI
The regulations related to license termination are in Title 10, "Energy," Part 20, "Standards for Protection 

Against Radiation," Subpart E, "Radiological Criteria for License Termination." See the response to 

Question 13.6 for instructions on how to obtain a copy of the regulations. 

9 HAZARDS ASSOCIATED WITH DECOMMISSIONING 

9.1 Workers 

9.1.1 Where do the decommissioning workers come from? 

The majority of workers for an immediate decontamination and dismantlement program will likely be 

people who worked on the operating plant. These workers are most familiar with the facility and its 

history. Some jobs, however, may be contracted out to companies that have gained experience at other 

plants in specialized areas of decommissioning or dismantlement. There will be very few employees 

during the storage phase in facilities that are placed in a storage mode. A new group of workers will likely 

need to be hired who are most likely unfamiliar with the plant, but who will probably have had some 

decommissioning experience at other facilities. 

9.1.2 Is worker safety considered in the planning for and review of decommissioning? 

Yes. Worker safety is considered both in terms of the radiological hazard (their exposure to radiation) and 

in terms of industrial safety. 

9.1.3 How much occupational dose is received by workers during decommissioning? 

The amount of occupational dose received during the decommissioning process will depend on the design 

and size of the facility as well as on the plans for decommissioning. A greater amount of occupational 

dose is anticipated to be incurred for an immediate decontamination and dismantlement than for a storage 

period followed by dismantlement. Estimates were given in NUREG-0586, "Final Generic Environmental 

Impact Statement on Decommissioning of Nuclear Facilities," (NRC 1988) that ranged from 333 

person-rem for a 30-year storage period to 1874 person-rem for an immediate decontamination and 

dismantlement. This can be compared with the 1996 annual average for an operating plant: 126 

person-rem for pressurized-water reactors and 235 person-rem for boiling-water reactors. The person-rem 

numbers are the doses that are received by all the workers. The dose to any one worker is expected to be 

below the 5-rem-a-year regulatory limit and is usually well below this limit. 

Since that study was performed, estimates for occupational dose from decommissioning range from 591 

person-rem for the Trojan nuclear plant to 1215 person-rem for Maine Yankee (includes the dose from 

transportation of the low-level waste [LLW] and 996 person-rem for Haddam Neck (includes the 

occupational dose from the transportation of the LLW). All three of these plants are using an immediate 

decontamination and dismantlement type of decommissioning. 

9.1.4 Are there limits on the amount of occupational dose that may be received?

45
Yes. The regulations state that the licensee shall control the occupational dose to individual adults to an 

annual limit of 5 rem (total effective dose equivalent to the entire body) or to an organ dose equal to 

50 rem. There are also dose limits to the eyes, the skin, and the extremities. 

9.1.5 Does the licensee have to estimate the occupational dose before the decommissioning 

process is initiated? 

No. However, at the time that the license-termination plan is updated, the licensee is required to update its 

environmental report as appropriate to reflect any new information or significant environmental change 

associated with the applicant's proposed decommissioning activities. The environmental report contains 

an estimate of occupational dose, so the licensee needs to estimate the occupational dose for 

decommissioning to determine if the estimates are within the range given in the environmental report for 

routine operations. 

9.2 Public and Environment 

9.2.1 Is the safety of the public considered in the planning for and review of 

decommissioning? 

Yes. The safety of the public is a major concern, even though the potential for hazards to the public from 

the decommissioning process and potential accidents is much less than it is when the facility is operating. 

9.2.2 How much dose will the public receive during the decommissioning process? 

The major source of exposure to the public is from the shipment of low-level waste from the reactor site to 

the low-level waste disposal site. Estimates made in a generic study of nuclear power reactor 

decommissioning (the GELS, NUREG-0586) range from 3 person-rem for a 30-year storage period to 

21 person-rem for immediate decontamination and decommissioning. The estimated public dose from the 

Trojan nuclear plant decommissioning is 4.8 person-rem. The estimated dose to the public from 

decommissioning the Haddam Neck plant is 11 person-rem. The radiation dose is received by people who 

travel along the same route as the trucks that are transporting low-level waste. However, because of the 

variability in the timing of each shipment, the short period of time that any person would be near any of 

the trucks, and the small dose that is allowed 6 feet from the side of a truck (10 millirem per hour), the 

dose that is received by any one person traveling down the highway or stopped at a rest stop is a very 

small fraction of the annual dose that the person would receive from background radiation. 

Minor sources of exposure to the public include radioactive effluent releases during the decommissioning 

process as discussed in the response to Question 9.2.4. 

9.2.3 Who estimates what the doses are, and how are these estimates made? 

The licensee estimates the doses. The doses are estimated using assumptions about the amount of 

radioactive material that will be released or the proximity of the public to the source of radiation. The 

doses are calculated using NRC-approved scenarios, assumptions, parameter values, and conceptual 

46

21
models. The NRC reviews the licensee's estimates of the doses and often recalculates the doses using its 

own assumptions for activities with the potential for significant worker exposure. 

9.2.4 What types of effluent releases are expected, and where will they enter the 

environment? 

Three important radiation-exposure pathways need to be considered in the evaluation of the radiation 

safety of normal reactor decommissioning operations: -inhalation, ingestion, and external exposure to 

radioactive materials. During decommissioning, inhalation is considered to be the dominant pathway of 

public radiation exposure, since exposure to radioactive surfaces and ingestion can be minimized or 

eliminated as radiation pathways to the public. During the transport of radioactive wastes, inhalation and 

ingestion can be minimized or eliminated as radiation pathways to the public by containing the waste in'a 

form or a container that does not allow for release to the air or water. Therefore, for transportation of 

radioactive waste, external exposure to radioactive materials is considered to be the only pathway of 

concern. 

During decontamination and dismantlement activities (DECON), radioactivity in air effluents is expected 

to be limited primarily to airborne radioactive particulates. The particulates will be filtered through , , 

high-efficiency particulate air (HEPA) filters in the ventilation systems of containment buildings, auxiliary 

buildings, and fuel-handling buildings. These are the buildings that contain the major sources of 

radioactive materials at a nuclear power facility. Air effluents from these buildings are monitored because 

the NRC has set limits on the amount of radioactive material that may be released. During storage 

(SAFSTOR), there should be little or no effluent releases since there are no decommissioning activities 

being performed. However, monitoring still continues to verify that any releases are minor. 

Liquid radioactive wastes are collected, stored, and processed, depending on the amount of particulates 

and the source of the liquid wastes. The liquid waste is processed before any releases are made to the 

environment. There are limits on the amount of radioactive material or other types of wastes that may be 

released to the environment. The regulatory limits for radioactive discharges can be found in 10 CFR Part 

20. Additionally, there are limits on the discharge of non-productive wastes. These limits are described 

on the National Pollutant Discharge Elimination System waste discharge permit that is issued to the 

specific facility. These permits are regulated by the States or the U.S. Environmental Protection Agency. 

Direct exposure can occur during both SAFSTOR and DECON. Public exposure would only result if 

members of the public were close enough to the source of radiation to receive a dose. The farther away a 

person is from radioactive material, the smaller the dose. The site boundary determines how far from the 

source people should be to avoid direct exposure; at that distance, exposure would be negligible. 

However, the dose from a shipment of nuclear waste could be 10 millirem per hour at a distance of 6 feet 

from the side of the truck. This means that a person standing there for 1 hour would receive a maximum 

dose of 10 millirem, which is a very small fraction of the average annual exposure to background 

radiation. 

9.2.5 Can you measure the effluent release to know how much is really entering the 

environment? 

Yes. Air effluent releases and liquid releases from all licensed light-water power reactor sites are 

monitored in accordance with the licensee's Offsite Dose Calculation Manual (ODCM). Release limits

47
given in the ODCM are based on NRC regulations. Special filters are used to collect particulate materials 

being released from the plant. Special air samplers and monitors are used to determine the amount of 

material that is released to the environment. Similar monitors are used to determine the amount of 

material released in liquid effluents. 

9.2.6 Will there be continued environmental monitoring of the site and the offsite areas to 

measure releases of radioactive material during the decommissioning process? 

Yes. The radiological environmental monitoring program that was in place at the nuclear plant will 

continue even after the plant is shut down. The program will be modified to appropriately monitor the 

types of releases that may occur during decommissioning and to monitor results at appropriate intervals of 

time. Not all measurements will be made on a continuous basis. The licensee uses the results of the 

environmental monitoring program to calculate the dose to the public. They follow a procedure that they 

have developed that is in a document called the Offsite Dose Calculation Manual. 

9.2.7 Who will perform the environmental monitoring? 

The Radiological Environmental Monitoring Program is conducted by the licensee. The procedures and 

results of the Radiological Environmental Monitoring Program are inspected and reviewed by NRC staff 

to ensure that all requirements are being met. The NRC does not independently monitor radioactive 

releases on a continuous basis. However, the NRC does occasionally have an independent measurement 

of the releases to verify that the licensee's program is adequate to analyze samples and to measure the 

amount of radioactive material that leaves the plant. 

9.2.8 Are the monitoring reports available to the public? How would we find them? 

Yes, each site prepares an annual environmental monitoring report that describes the Radiological 

Environmental Monitoring Program and the results of the program for that calendar year. The reports are 

available to the public through the Public Document Room and the Public Electronic Reading Room. See 

the responses to Questions 13.1 and 13.2 for instructions to access these sources of information. 

9.2.9 What types of chemicals will be released to the environment? 

Chemicals that are used at the operating facility and chemicals that are used specifically during the 

decontamination process must be disposed of in accordance with appropriate Federal and State 

regulations. Liquid releases from the plant to rivers, streams, or lakes must meet the requirements given in 

the facility's National Pollutant Discharge Elimination System permit, which is issued by the State or by 

the U.S. Environmental Protection Agency. Release of chemicals to waste sites must meet the 

requirements of that waste site. Mixed wastes (hazardous chemicals and radioactive material) are 

generally shipped to a licensed mixed waste processor. 

9.2.10 What hazards are presented to the public when the waste is shipped? 

Very minor amounts of radiation will be received by people in other vehicles driving alongside, or parked 

beside, trucks carrying low-level radioactive waste from the facility to the low-level waste site. The 

hazard from transportation accidents is discussed in responses to Questions 7.5 and 7.9. However, there is

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also a potential hazard resulting from the release of radioactive material due to a major accident involving 

a vehicle carrying radioactive waste. The low-level waste is shipped in a solid form, so any danger 

resulting from a typical vehicular accident would likely be confined to a small radius around the accident. 

9.2.11 What types of accidents at the reactor site are considered and what would be the 

consequences to the public? 

Once the reactor permanently shuts down, the risk to the public is greatly reduced; however, there are still 

several accidents that may occur with consequences offsite. The accidents that have the potential for the 

greatest offsite doses are those that involve the spent fuel that has recently been moved from the reactor to 

the spent fuel pool. Over time, the hazard from the spent fuel diminishes as the radioactive material in the 

fuel decays. 

Licensees are required to examine their sites and decommissioning plans to identify postulated accidents 

that could occur during decommissioning. An analysis of these accidents is required in their Final Safety 

Analysis Report, which is part of the licensing basis for the plant. Except for the fuel-related accidents in 

the first year(s) after the facility ceases operation, the offsite consequences of these accidents are very 

small and do not require offsite emergency response. Examples of the types of accidents that are 

considered by the licensees include 

"* Cask or heavy load-handling accident with a subsequent drop into spent fuel pool 

"* Loss of cooling for the spent fuel pool or loss of water from the spent fuel pool 

* Materials handling event (non-fuel) 

"* Radioactive liquid waste releases 

"* Accidents from handling spent resin 

"* Fire 

"• Explosions 

"* External events 

"* Transportation accidents. 

If a licensee requests an exemption to a regulation because they believe it no longer applies due to the 

decommissioning state of the plant, they must show that the regulation is not needed including 

consideration of the risk to the public. Additional information regarding the consequences of spent fuel 

pool accidents is given in the response to Question 5.8.4.

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9.3 General

9.3.1 In general, how safe is a decommissioning plant in contrast to an operating plant? 

At the time that the plant permanently ceases operations and the fuel is removed from the reactor, the risk 

to the public from an accident drops significantly. 

9.3.2 Will there still be emergency preparedness plans and warning sirens in the vicinity of 

the plant? 

For some period of time after the licensee ceases reactor operations, the offsite emergency preparedness 

planning will be maintained. This period of time depends on when the reactor was last critical as well as 

onsite-specific considerations. Offsite emergency planning may be eliminated when the fuel has been 

removed from the reactor and placed in the spent fuel pool, and sufficient time has elapsed, and there are 

no longer any postulated accidents that would result in offsite dose consequences that are large enough to 

require offsite emergency planning. There would be no requirement to maintain offsite systems to warn 

the public. Onsite emergency plans will be required for both the spent fuel pool and the Independent 

Spent Fuel Storage Installations, but offsite plans will not be required. 

If, however, an operating plant is located at the same site as the decommissioning plant, the emergency 

preparedness plans will still be in effect for the operating plant. 

9.3.3 What measures are taken to prevent vandalism and sabotage during 

decommissioning? 

The facility is required to have a security plan when a plant is being decommissioned; however, as the 

hazards are removed from the nuclear reactor site, security requirements are modified. Many plants 

reduce the area that they keep very secured to the area around where the spent fuel is stored. This is 

known as the "nuclear island." Reducing the size of the area that has strict security measures allows for 

better control of the material that must be safeguarded. Security measures for the nuclear island are 

designed to prevent sabotage or removal of the nuclear material. 

10 FINANCES 

10.1 How much does it cost to decommission a nuclear power plant? 

The total cost of decommissioning depends on many factors, including the sequence and timing of the 

various stages of the program, location of the facility, current radioactive waste burial costs, and plans for 

spent fuel storage. The minimum amounts that are required for reasonable assurance of funds for 

decommissioning are $290 million for pressurized-water reactors and $370 million for boiling-water 

reactors. These costs are in 1999 dollars and are adjusted annually, as further specified in the regulations. 

These are minimum amounts to show reasonable assurance, rather than estimates, of what it would cost to 

decommission a specific nuclear reactor. 

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11I
Actual site-specific costs incurred and estimated costs of decommissioning give a better indication of what 

the process costs. The Fort St. Vrain nuclear plant, which was a 330-megawatt-electric high-temperature 

gas-cooled reactor, ceased power operations in 1989 and underwent immediate decontamination and 

dismantlement. The decommissioning effort was completed in late 1996, and the license was terminated. 

The total cost of decommissioning was $189 million. 

The cost for decommissioning the Trojan nuclear plant (an I 130-megawatt-electric pressurized-water 

reactor) is estimated to be on the order of $210 million in 1993 dollars, which does not include $42 million 

for non-radioactive site remediation or $110 million for the independent spent fuel storage installation and 

related fuel management. The Trojan nuclear plant is also planning an immediate decontamination and 

decommissioning from shutdown in 1993 to license termination in 2002. 

The estimated cost for decommissioning the Haddam Neck nuclear plant, a 619-megawatt- electric 

pressurized-water reactor, is $344.4 million in 1996 dollars, not including $82.3 million in spent fuel 

storage costs (for a total of $426.7 million). 

The estimated cost for decommissioning Maine Yankee, an 830-megawatt-electric pressurized- water 

reactor, is $274.9 million in 1997 dollars. This does not include costs for spent fuel management 

($53.4 million) or for site restoration ($49.2 million), for a total of $377.6 million. 

The estimated cost for decommissioning Big Rock Point, a 67-megawatt-electric boiling-water reactor, is 

$290 million in 1997 dollars. 

The estimated cost for decommissioning Rancho Seco, a 913-megawatt-electric pressurized-water reactor 

is $441 million in 1995 dollars. 

The estimated cost for decommissioning Yankee Rowe, a 175-megawatt-electric pressurized- water 

reactor is $306.4 million in 1995 dollars. 

Decommissioning costs vary, based on plant size and design, local labor and radiological waste burial 

costs, and the specific process that is being u/sed for decommissioning. 

10.2 Who makes the estimates of the decommissioning costs? 

The licensee makes the site specific estimates of the decommissioning costs or hires a contractor who has 

extensive experience in making these estimates. The site specific estimates are reviewed by the NRC. 

10.3 When are the estimates of the decommissioning costs made? 

The NRC has regulations regarding the methods used to reasonably ensure that funds will be available to 

decommission the facility. The NRC has specified a table of minimum amounts required to demonstrate 

reasonable assurance of funds for decommissioning by reactor type and power level ($105 million for 

pressurized-water reactors and $135 million for boiling-water reactors in 1986 dollars). The licensees 

must adjust this rate annually. Licensees may also perform site-specific estimates that could result in cost 

estimates that are higher than the generic-formula amounts specified in 10 CFR 50.75 (c).

51
An estimate is made at or about 5 years preceding the projected end of operations. At this time, power 

reactor licensees shall submit a preliminary decommissioning cost estimate, which includes an up-to-date 

assessment of the major factors that could affect the cost to decommission. If the amount of money 

available is inadequate, the licensee has approximately 5 years to adjust the money in the decommis

sioning trust fund to ensure that appropriate funds are available for decommissioning. 

An estimate is submitted at the time that the post-shutdown decommissioning activities report (PSDAR) is 

submitted (no later than 2 years following permanent cessation of operations). This estimate may be (1) a 

site-specific cost estimate that is based on the activities and schedule that are also discussed in the PSDAR, 

(2) an estimate based on actual costs at similar facilities that have undergone similar decommissioning 

activities, or (3) a generic cost estimate. The NRC recommends that licensees planning an immediate 

decontamination and dismantlement submit a site-specific cost estimate in the PSDAR; however, a more 

generic one would be acceptable for facilities that are submitting their PSDAR in advance of the 2-year 

requirement. If a storage period is planned during decommissioning, the licensee should provide a method 

of adjusting the cost estimate and funding throughout the duration of the storage. 

The regulations also require a site-specific cost estimate within 2 years following permanent cessation of 

operations if one has not already been submitted. 

Finally, at the time that the license-termination plan is submitted (at least 2 years before the date when the 

license terminates), an updated site-specific estimate of any remaining decommissioning costs is required. 

10.4 If the first estimate of decommissioning costs is made at the time that the 

facility is licensed, are there methods for adjusting for inflation? 

NRC regulations provide an adjustment factor for cost escalation that takes into account escalation factors 

for labor, energy, and waste burial. The labor and energy escalation factors are obtained from regional 

data issued by the U.S. Department of Labor's Bureau of Labor Statistics. The waste-burial cost

escalation factor is taken from an NRC report, "Report on Waste Burial Charges" ( NUREG-1307). 

10.5 How does the NRC ensure that the licensee will have the money when it 

is needed for decommissioning? 

Financial assurance is provided by the following methods: 

Prepayment. In this case, at the start of operations, the licensee deposits into an account enough funds to 

pay the decommissioning costs. The account is segregated from the licensee's other assets and remains 

outside the licensee's administrative control of cash or liquid assets. Prepayment may be in the form of a 

trust, escrow account, government fund, certificate of deposit, or deposit of government securities. 

External sinking fund. An external sinking fund is a fund established and maintained by setting funds 

aside periodically into an account segregated from licensee assets and outside the licensee's administrative 

control. The total amount of these funds would be sufficient to pay decommissioning costs at the time that 

it is anticipated that the licensee will cease operations. An external sinking fund may be in the form of a 

trust, escrow account, government fund, certificate of deposit, or deposit of government securities. 

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Surety method, insurance, or other guarantee method. A surety method may be in the form of a surety 

bond, letter of credit, or line of credit. Any surety method or insurance used to provide financial assurance 

must be open-ended, or if written for a specific term, such as 5 years, must be renewed automatically. An 

exception is allowed when the issuer notifies the Commission, the beneficiary, and the licensee of its intent 

to not renew within 90 days or more preceding the renewal date. The surety or insurance must also 

provide that the full face amount be paid to the beneficiary automatically preceding the expiration date 

without proof of forfeiture if the licensee fails to provide a replacement acceptable to the Commission 

within 30 days after receipt of notification of cancellation. In addition, the surety or insurance must be 

payable to a trust established for decommissioning costs, and the trustee and trust must be acceptable to 

the Commission. The surety method or insurance must remain in effect until the Commission has 

terminated the license. 

10.6 Does the NRC have a method to monitor the decommissioning trust fund 

moneys for each specific nuclear power facility? 

The NRC monitors the decommissioning funds. The regulations regarding how the funds are to be 

accumulated are very specific. Accounts for decommissioning are segregated from the licensee's other 

assets and remain outside of the licensee's administrative controls as discussed in the response to Question 

10.5 (although there is a provision for combining the costs for long-term storage of the spent fuel onsite 

with the costs for decontamination and dismantlement of the facility as long as the funds for each are 

clearly identified). In addition, the regulations require that the licensee report at least once every 2 years 

on the status of its decommissioning funding for each reactor or part of a reactor that it owns. This same 

report is required annually for facilities that have already permanently ceased operations, or facilities that 

are within 5 years of the projected end of their operation, or if it appears that they might be within 5 years 

of shutting down. 

10.7 Are the financial assurance reports available to the public? 

The financial assurance reports that are discussed in the previous section and that are submitted to the 

NRC on an annual or biannual basis are part of the public record and can be obtained from the Public 

Document Room or the Public Electronic Reading Room by the process described in the response to 

Question 13.2 or 13.4. 

10.8 Do the financial assurance regulations apply for Federal government 

licensees? 

Certain Federal government licensees may provide a statement of intent containing a cost estimate for 

decommissioning and indicating that funds for decommissioning will be obtained when necessary as long 

as the obligations of such licensees are granted by and supported by the full faith and credit of the 

U.S. government. 

10.9 Is there any way to ensure that the licensee does not just spend all of the 

money in the first few years of decommissioning and have nothing left to 

complete the job?

53
The NRC has placed regulations regarding the amount of money that can be used from the decommis

sioning fund at various stages of the decommissioning process. The licensee is allowed to use 3 percent of 

the generic amount of funds that are specified in the regulations for power plants based on their size and 

type for decommissioning planning that may occur, even while the facility is still operating. Appropriate 

activities include engineering design, work package preparation, and licensing activities. 

After submitting the certification of permanent cessation of operations and the certification that the fuel 

has been removed from the reactor vessel, the licensee may use an additional 20 percent of the funds for 

any legitimate decommissioning activities. The licensee is prohibited from using the remaining 77 percent 

of the generic decommissioning funds until a site-specific cost estimate is submitted to the NRC. This cost 

estimate must be submitted within 2 years following permanent cessation of operations. Also, the NRC's 

regulations require licensees to use funds from decommissioning trusts only for those legitimate 

decommissioning activities consistent with the NRC's definition of decommissioning (see the response to 

Question 1.1). 

10.10 What would happen if the cost of decommissioning exceeds the amount 

of money in the trust fund? 

The various cost estimates (at the time of licensing, 5 years before anticipated shutdown, with the Post

Shutdown Decommissioning Activities Report submittal, 2 years following shutdown, and 2 years 

preceding the anticipated termination of the license) are a method of re-evaluating the decommissioning 

costs at various times and stages in the facility's life to ensure that there will be adequate funds available 

to complete the decommissioning process. If there is insufficient money in the trust fund, licensees would 

be required to obtain the additional funds needed to complete the decommissioning of the facility. 

10.11 What would happen if the plant had an accident, and there was not 

enough money in the decommissioning trust fund to complete 

decommissioning and cleanup after the accident? 

Licensees are required to carry insurance, which is separate from the decommissioning funding 

requirements, in an amount that would allow cleanup of the site to such a level that decommissioning 

could be completed with the full amount of the decommissioning trust fund. Currently, $1.06 billion per 

operating unit is required for such insurance coverage. 

10.12 Will enough money still be in the decommissioning trust fund even if the 

facility ceases operation? 

In the event that a facility shuts down prematurely before the decommissioning trust is fully funded, the 

licensee is still required to fully fund the decommissioning. Most utilities are diversified and are able to 

recapture these costs as part of their continued business. To date, none of the utilities of the prematurely 

shut-down facilities have defaulted on their decommissioning funding obligation. Contingency plans are 

in place to ensure that decommissioning will be properly performed in the event of financial default of the 

licensee as discussed in the response to Question 10.14. 

10.13 Who pays for decommissioning? 

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21
The particular licensee that holds the license for the facility pays for decommissioning. Subject to the 

public utilities commission that regulated the utility, the money for decommissioning is collected as part of 

the price of electricity; thus, the funds for decommissioning are ultimately paid by the ratepayer in the 

electric bill. As the electric utility industry deregulates, many states are choosing to require payment of 

decommissioning costs through the imposition of non-by-passable charge as part of a customer's electric 

bill. 

10.14 What contingency plans are in place to ensure that decommissioning and 

long-term radioactive material storage will be properly performed in the 

event of financial default of the licensee? Who finances decommissioning 

if the licensee becomes bankrupt or insolvent? 

The Atomic Energy Act contains provision for the Federal government to assume responsibility for 

decommissioning if public health and safety are jeopardized because of inability on the part of the 

licensee. 

Bankruptcy does not necessarily mean that a power reactor licensee will liquidate. To date, the NRC's 

experience with bankrupt power reactor licensees has been that they file under Chapter 11 of the 

Bankruptcy Code for reorganization, not liquidation (for example, Public Service Company of New 

Hampshire, El Paso Electric Company, and Cajun Electric Cooperative). In these cases, bankrupt 

licensees have continued to provide adequate funds for safe operation and decommissioning, even as 

bondholders and stockholders suffered losses that were often severe. Because electric utilities typically 

provide an essential service in an exclusive franchise area, the NRC staff believes that, even in the 

unlikely case of a power reactor licensee liquidating, its service territory and obligations, including those 

for decommissioning, would revert to another entity without direct NRC intervention. 

10.15 What will happen if deregulation becomes a reality? How will 

deregulation affect anticipated revenue and the capability to 

decommission? 

The NRC has issued a final policy statement on its expectations and intended approach to nuclear power 

plant licensees as the electric utility industry moves from an environment of rate regulation toward greater 

competition. This policy statement was issued on August 19, 1997, and published in the Federal Register 

(62 FR 44071). The policy statement addresses NRC concerns about the adequacy of decommissioning 

funds. The statement indicates that the NRC believes that its current regulatory framework is generally 

sufficient to address the expected changes, but in order to remove any ambiguities in its regulations and 

address situations that may not be adequately covered, the Commission is considering revising its financial 

and decommissioning funding-assurance requirements. 

Deregulation may force some licensees to separate.their systems into functional areas, with their 

NRC-licensed nuclear plants potentially no longer being rate regulated. This would cause some licensees 

to cease being an "electric utility," as defined in NRC regulations. If this occurs, the NRC will require the 

licensees to meet the more stringent decommissioning funding-assurance requirements that apply to 

non-electric utilities. Electric utilities are permitted to accumulate funds for decommissioning over the 

remaining terms of their operating licenses. NRC regulations require most other licensees to provide

55
funding assurance for the full estimated cost of decommissioning, either through full up-front funding or 

by some allowable guarantee or surety mechanism. 

In addition, the policy statement emphasizes that the NRC retains the right to assess the timing of 

decommissioning trust fund deposits and withdrawals and the liquidity of decommissioning funds for 

licensees that no longer are subject to rate regulatory oversight. 

The NRC issued a final rule on September 22, 1998, that modified its decommissioning funding-assurance 

requirements in anticipation of rate deregulation. 

11 SOCIOECONOMIC ISSUES 

11.1 What impact would each of the alternatives have on the economy of the 

surrounding area, including work-force requirements? 

The biggest socioeconomic impact occurs before decommissioning starts, at the time the plant ceases 

operations, and the tax income created by the plant is substantially reduced. The surrounding communities 

may find their property tax base reduced in half or more, depending on the presence or absence of other 

industries in the area. Typically, additional public services are not required during decommissioning 

because the plant staff will be smaller than the operating staff. 

11.2 How many people are needed onsite during the decommissioning 

process? Is this more than during operations? 

After cessation of operations, the number of workers in the plant will be reduced. Plants that are currently being decommissioned using the DECON alternative have work forces in the range of 100 to 200 persons (for a single-unit plant). This is approximately one-third to one-tenth the number of persons who were 

employed at the plant during its operation. 

These personnel are periodically supplemented with contract personnel during major decommissioning activities, such as the removal of large components like the steam generators and pressurizer. The onsite 

workforce could exceed 600 personnel during these peak periods of activity. If the plant were placed in 

SAFSTOR, the number of workers would be further reduced. Decommissioning plants that are located at 

the same site as operating facilities generally have a staff of 20 or fewer during SAFSTOR. Single-unit plants (not located next to operating units) require a larger staff and may have 20 to 70 employees during SAFSTOR. After the SAFSTOR period, the number of workers would increase to the range of 100 to 200 

and would be further supplemented with contractor personnel for the final cleanup of the site. 

11.3 What is being done to assist communities that relied heavily on the 

operating plant for their tax base, now that the plant has permanently 

ceased operations? 

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The NRC has no mechanism for providing economic assistance to communities that relied heavily on the 

operating plant for their tax base. This matter is generally left to local and State governments to address. 

12 PUBLIC INVOLVEMENT 

12.1 Public Meetings 

12.1.1 What meetings are planned to keep the public informed? 

Two public meetings are required during the decommissioning process. The first occurs before major 

decommissioning activities begin, when the post-shutdown decommissioning activities report is submitted. 

The second takes place when the licensee submit a license-termination plan, which describes how the site 

will be returned to a condition that makes radiological controls no longer necessary, must be submitted by 

the licensee. In both cases, the NRC will publish notifications of the public meetings in the Federal 

Register and in local media. The meetings will be held in the vicinity of the power plant to encourage 

local participation. 

Although not required by the regulations, the NRC holds an initial public informational meeting shortly 

after the licensee submits the certification of permanent cessation of operations. At the meeting, the NRC 

presents its process for regulating decommissioning, the licensee presents its current plans for shutting 

down the facility and for decommissioning it (if any such plans have been made), and questions and 

comments from members of the public are addressed. 

12.1.2 Where will the meetings be held? 

Meetings are held in the vicinity of the facility. Often they are held at a local hotel, county courthouse, 

school, or library. 

12.1.3 How will the public be notified about the meetings? 

NRC issues a news release and publishes a notice of the date, time, and location of the public meeting in 

the Federal Register and in the local papers. The notice is also posted on the meetings section of the 

NRC's website (www.nrc.gov). See the response to Question 13.3 for additional information on accessing 

the NRC's website.

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12.1.4 If I cannot attend a meeting, how do I find out what was said?

A written transcript of the meeting is prepared. A signup sheet is made available for members of the 

public to leave their addresses if they wish to receive a copy of the transcript. An individual who is unable 

to attend the meeting may contact the NRC project manager whose name, address, and telephone number 

are listed in the Federal Register notice and other published notices. Individuals may also call the NRC 

Office of Public Affairs, 1-301-415-8200, or the NRC project manager for the facility being 

decommissioned. 

12.1.5 May I make comments at the meeting? 

Yes. A portion of the meeting will be devoted to questions from the public. Additionally a sign-up sheet 

will be available at the start of the meeting for individuals requesting time for making comments or for 

reading prepared statements. Questions and short comments will also be accepted from the floor. 

12.2 Comment Period 

12.2.1 When does the comment period for the decommissioning process start? 

There are two separate comment periods. The first occurs following the licensee's submittal of the 

post-shutdown decommissioning activities report (PSDAR) discussed earlier. A notice of the receipt of 

the PSDAR and the scheduling of a public meeting to be held in the vicinity of the licensee's facility is 

printed the Federal Register, posted in local places, and printed in a local newspaper. The PSDAR will be 

available for public comment at that time. 

There is a second opportunity for public comment after the licensee submits a license termination plan 

which describes the remaining activities necessary to terminate the NRC license. Again, the NRC 

publishes a notice of the receipt of the license-termination plan in the Federal Register, and schedules a 

public meeting to be held in the vicinity of the licensee's facility. This notice is also published in the local 

vicinity of the site and in a local newspaper. Comments may be made in writing or orally at the public 

meeting. 

12.2.2 How do I make comments on the decommissioning process? 

Comments and questions may be submitted in writing to the NRC project manager for the facility. 

Comments and questions can also be addressed at the public meeting following receipt of the Post

Shutdown Decommissioning Activities Report (PSDAR) or the License Termination Plan. A written 

transcript containing these comments is prepared. All comments and questions received at the meeting 

will be responded to in a written memorandum that will be made available to the public. Additionally, a 

memorandum that documents whether or not the information provided in the PSDAR satisfies NRC 

requirements will also be prepared and made available to the public. A signup sheet will also be available 

at the public meeting for individuals to request copies of the memorandum, and a copy will be mailed to 

those who request one from the NRC project manager. An address and telephone number for the specific

58
NRC project manager will be published in the Federal Register along with the notice of the receipt of the 

PSDAR and the schedule of the public meeting. 

12.2.3 Where should I send my comments? 

NRC publications and notices identify the correct address and person to write to when submitting 

comments. 

12.3 Hearings 

12.3.1 Are hearings held on the decommissioning process? 

Whenever there is an amendment to the facility's license, there is an opportunity for a hearing. For 

example, if the NRC finds the license-termination plan acceptable, it approves the plan by license 

amendment, which allows the opportunity for members of the public to request a hearing. The Post

shutdown Decommissioning Activities Report (PSDAR) is not a request for a license amendment, 

therefore, there is no opportunity at this stage in the process for a hearing on the contents of the PSDAR. 

The public meetings that follow the submittal of the PSDAR provide an additional opportunity for public 

participation. 

12.3.2 Why is there no opportunity for a public hearing in the early stages of 

decommissioning? 

The early stages of decommissioning, including the submittal of the Post-shutdown Decommissioning 

Activities Report, is not considered a major licensing action and does not, therefore, require a license 

amendment. If a license amendment is requested, there will be an opportunity for a public hearing. 

12.3.3 What is the process that is used for submitting contentions for a public hearing? 

The process for requesting or participating in a hearing on a license amendment is set forth in the NRC's 

regulations in Part 2, "Rules of Practice of Domestic Licensing Proceedings and Issuance of Orders" in the 

Code of Federal Regulations, Title 10, "Energy." Information on obtaining a copy of these regulations is 

given in the response to Question 13.6. 

The regulations state that after publication of the proposed action (license amendment) in the Federal 

Register, any person whose interest may be affected by the proceeding may file a request for a hearing or a 

petition for leave to intervene if a hearing has already been requested. The Federal Register notice 

describes the time period in which this request or petition must be made. The person or group submitting 

the request or petition must explain in the petition (among other things) their interest in the proceeding and 

how that interest may be affected by the results of the proceeding. At a later time, they must also submit a 

list of contentions, which are specific statements of the issue of law or fact that are being raised. A brief 

explanation of the contention is required, which includes facts or expert opinion that supports the 

contention and sufficient information to show that a genuine dispute exists. Specific requirements for the 

information required in these documents and for the timing of the information submittals is specified in the 

NRC's regulations.

59
12.3.4 Does the public really have any power to make changes if there are problems? 

Yes. There are several methods by which the public can make changes. If the licensee has requested an 

action requiring a license amendment, then the process for intervening in this action is discussed in the 

response to Question 12.3.3. If the action of concern does not involve a license amendment, any member 

of the public may raise potential health and safety issues in a petition to the NRC to take specific 

enforcement action against a licensed facility. This provision is contained in the NRC's regulations and is 

often referred to as a "2.206 petition" in reference to its location in the regulations (Chapter 2, Section 206 

of Title 10, "Energy." In general, the petition is submitted in writing to the Executive Director for 

Operations and includes specific facts supporting the request for the NRC to take enforcement action. 

Unsupported assertions of safety problems or general opposition to nuclear power are not considered 

sufficient grounds for action. Further information on this approach is found on the NRC's website under 

Public Outreach (http://www.nrc.gov/OPA). 

Workers at the nuclear power plant, or members of the public that have specific concerns of a safety

related nature, can bring safety concerns (allegations) directly to the NRC (although workers are 

encouraged to take their concerns directly to their employer initially). The NRC's toll-free safety hotline 

is 1-800-695-7403. 

Members of the public may also voice concerns at the public meetings that are held in conjunction with the 

licensee's submittal of the Post-Shutdown Decommissioning Activities Report or License Termination 

Plan, or they may request information or identify concerns in writing. General correspondence should be 

addressed to the Office of Secretary, U.S. Nuclear Regulatory Commission, Washington, D.C. 20555

0001. 

13 GETTING ADDITIONAL INFORMATION 

13.1 Other than the public meeting, how can I get information about a 

nuclear power plant? 

The NRC has many ways of keeping the public informed. These include printed materials, electronic 

access. and public meetings. NRC's website contains information of interest to the public 

(htt2://www.nrc.gov). This site also has a link to the Public Electronic Reading Room as described in the 

response to Question 13.2. 

A comprehensive listing of information sources can be found in an NRC publication, NUREG/BR-0010, 

Rev. 2. Citi-en's Guide to L-S. Nuclear Regulatory Commission Information. 

The NRC maintains a Public Document Reading room in Washington, D.C. as discussed in the response to 

.Question 13.4 

Copies of documents can also be ordered from the Government Printing Office at the following address: 

60

V'
Superintendent of Documents 

U.S. Government Printing Office 

P.O. Box 37082 

Washington, DC 20402-9328 

Phone: 202-512-1800 

Copies can also be read online or downloaded electronically from the NRC's website. The URL (address 

or Universal Resource Locator) for this is http://www.nrc.gov/NRC/NUREGS/BR0010/index.html. 

13.2 What is the Public Electronic Reading Room, and how do I access it? 

The Public Electronic Reading Room provides access to the NRC's new records-management system of. 

publicly available information, the Agency wide Documents Access and Management System (ADAMS). 

Within this system you can access two libraries: the Publicly Available Records System, and that Public 

Legacy Library. 

This system, which was implemented on October 12, 1999, marks a change in the previous practice where 

records were available only in paper or microfiche copies at either the main NRC Public Document Room 

in Washington, D.C. or at 86 local public document rooms at libraries near nuclear power plants and other 

regulated facilities throughout the United States. Access to the NRC Public Electronic Reading Room will 

now be possible from personal computers, including those located in most public libraries. 

ADAMS is an electronic information system that allows access to NRC's publicly available documents via 

the Internet. It permits full text searching and the capability to view document images, download files, and 

print locally. It also provides a more timely release of information by the NRC and faster access to 

documents by the public than before. 

ADAMS can be accessed via the Internet at the NRC's website using the following URL: 

http://www.nrc.gov/NRC/ADAMS/index.htmi. This site contains instructions for installing and running 

ADAMS as well as information on obtaining assistance during installation or use. 

13.3 Does the NRC have a website? What kind of information can I obtain 

from it? 

The NRC maintains a website at http://www.nrc.Lov. The websitc includes 

"* general information about the NRC 

"* information on nuclear reactors (including the nuclear plant performance, plant information books, and 

daily reports, including plant status reports and daily events reports) 

" information on radioactive waste disposal 

"* information on current rulemaking processes 

"* news releases and fact sheets on a variety of nuclear issues

61
"* information on public involvement and public meetings, including sources for additional information 

"* copies of the post-shutdown decommissioning activities reports (PSDARs) for plants that have 

permanently ceased operations. 

13.4 Does the NRC maintain a Public Document Room? 

Yes. The Public Document Room contains a comprehensive collection of unrestricted documents related 

to NRC licensing proceedings and other significant decisions and actions. Among the wide variety of 

agency documents available to the public at the Public Document Room are NRC NUREG reports and 

manuals, transcripts and summaries of Commission meetings and NRC Staff and licensee meetings; 

existing and proposed regulations and rulemakings; licenses and amendments; and correspondence on 

technical, legal, and regulatory matters. Technical Reference Librarians are available to assist on-site 

users and those who call or write with information requests. The Public Document Room is located at 

2120 L Street, N.W., Washington, D.C. It is open Monday through Friday from 7:45 a.m. to 4: 15 p.m. 

eastern time, except Federal holidays. Information and service to offsite users is available by calling (202) 

634-3273 or 1-800-397-4209. The fax number is (202) 634-3343 and the email address is Wdranrc.gov. 

The mailing address is U.S. Nuclear Regulatory Commission, Public Document Room, Washington, D.C. 

20555. 

13.5 What is the Federal Register and how can I get a copy of it? 

The Federal Register is a daily publication announcing rules, policies, and other important actions of the 

Federal Government. Copies are available at many local libraries. Copies are also available at the NRC's 

Public Document Room in Washington, D.C. You may also search the Federal Register database on the 

lnternet at http://www.access.gpo.gov/su docs/aces/aces 140.htmi. 

13.6 How can I get a copy of the Code of Federal Regulations? 

Title 10 of the Code of Federal Regulations (the section containing regulations enforced by the NRC) is 

available on the NRC's website under the reference library (http://www.nrc.gov/NRC/CFR/index.html). 

Most local libraries have personal computers available for free access to the NRC's website via the 

Internet. The entire collection of codes of Federal Regulations is often available at local libraries in the 

reference section. They are also available for purchase from the Government Printing Office (GPO) by 

credit card at 202-512-1800, Monday through Friday, 8 a.m. to 4 p.m. EST (fax 202-512-2233, 24 hours a 

day) or by check by writing to the Superintendent of Documents, Attn: New Orders, P.O. Box 371954, 

Pittsburgh, PA 15250-7954. For GPO Customer Service, call 202-512-1803. 

13.7 How can I get answers to additional questions that were not addressed in 

this document? 

The NRC has established a toll-free number, 1-800-368-5642, for general inquiries from the public 

concerning information on NRC activities. The NRC accepts calls from telephones equipped with a 

telecommunication device for the deaf (TDD) at a special main switchboard number, 301-415-5575. TDD 

numbers are also available in the NRC Library (301-415-5609). 

62

j!
The phone number for the Public Affairs Office located at NRC headquarters is 301-415-8200. 

Regional public-affairs offices are also available to answer questions on specific nuclear facilities at 

"* Region I (Philadelphia) 610-337-5000 

"* Region I (Atlanta) 404-562-4400 

"* Region III (Chicago) 630-829-9500 

* Region IV (Dallas) 817-860-8100 

Questions may be addressed in writing to 

Office of Public Affairs 

U.S. Nuclear Regulatory Commission 

Washington, DC 20555-0001 

Electronic inquiries can be sent to opa@nrc.gov.

63
14 BIBLIOGRAPHY

Federal Register Notices 

"Clarification of decommissioning funding requirements," Federal Register, Vol. 60, p. 38235, July 26, 

1995 (10 CFR Parts 30, 40, 50, 51, 70, and 72). 

"Decommissioning of nuclear power reactors," Federal Register, Vol. 61, p. 39278, July 29, 1996 

(10 CFR Parts 2, 50, and 51). 

"Definitions - Decommissioning," Federal Register, Vol. 62, p. 39089, July 21, 1997 (10 CFR 30.4). 

Electric utility industry; restructuring and economic deregulation; public health and safety effects; policy 

statement. Federal Register Vol. 62, p. 44071, August 19, 1997. 

"Financial assurance requirements for decommissioning nuclear power reactors," Federal Register, 

Vol. 63, p. 50465, September 22, 1998 (10 CFR Parts 30 and 50). 

"General requirements for decommissioning nuclear facilities," Federal Register, Vol. 53, p. 24018, 

June 27, 1988 (10 CFR Parts 30, 40, 50, 51, 70, and 72). 

"Public notification and public participation," Federal Register, Vol. 62, p. 39089, July 21, 1997 (10 CFR 

20.1405). 

"Radiological criteria for license termination," Federal Register, Vol. 62, p, 39058, July 21, 1997 (10 CFR 

Parts 20, 30, 40, 50, 51, 70, and 72). 

"Timeliness in decommissioning of materials facilities," Federal Register, Vol. 59, p. 36026, July 15, 

1994 (10 CFR Parts 2, 30, 40, 70, and 72). 

Generic Letters 

GL 79-19, "Packaging of Low-Level Radioactive Waste for Transport and Burial." 

GL 79-30, "Packaging, Transport and Burial of Low-Level Radioactive Waste." 

Information Notices 

IN 80-24, "Low-Level Radioactive Waste Burial Criteria." 

IN 83-5, "Obtaining Approval for Disposing of Very-Low-Level Radioactive Waste - 10 CFR 

Section 20.302." 

IN 85-92, "Surveys of Wastes Before Disposal From Nuclear Reactor Facilities." 

IN 86-90, "Requests to Dispose of Very Low-Level Radioactive Waste Pursuant to 10 CFR 20.302." 

64

Pi
IN 87-3, "Segregation of Hazardous and Low-Level Radioactive Wastes." 

IN 88-8, "Chemical Reactions With Radioactive Waste Solidification Agents." 

IN 88-16, "Identifying Waste Generators in Shipments of Low-Level Waste to Land Disposal Facilities." 

IN 89-13, "Alternate Waste Management Procedures in Case of Denial of Access to Low-Level Waste 

Disposal Sites." 

IN 89-27, "Limitations on the Use of Waste Forms and High Integrity Containers for the Disposed of 

Low-Level Radioactive Waste." 

IN 90-9, "Extended Interim Storage of Low-Level Radioactive Waste by Fuel Cycle and Materials 

Licensees." 

IN 90-31, "Update on Waste Form and High Integrity Container. Review Status, Identification of 

Problems With Cement Solidification." 

IN 90-75, "Denial of Access to Current Low-Level Radioactive Waste Disposal Facilities." 

IN 91-3, "Management of Wastes Contaminated With Radioactive Materials ('Red Bag' Waste and 

Ordinary Trash)." 

IN 91-65, "Emergency Access to Low-Level Radioactive Waste Disposal Facilities." 

IN 94-7, "Solubility Criteria for Liquid Effluent Releases to Sanitary Sewage Under the Revised 

10 CFR 20." 

IN 94-23, "Guidance to Hazardous, Radioactive and Mixed Waste Generators on the Elements of a Waste 

Minimization Program." 

IN 96-47, "Recordkeeping, Decommissioning Notifications for Disposals of Radioactive Waste by Land 

Burial Authorized Under Former 10 CFR 20.304, 20.302 and Current 20.2002." 

NUREG-Series Reports 

NUREG-0586, "Final Generic Environmental Impact Statement on Decommissioning of Nuclear 

Facilities," August 1988. 

NUREG-0945 (Draft), "Draft Environmental Impact Statemenit for 10 CFR Part 61," September 198 1. 

NUREG-0945, "Final Environmental Impact Statement for 10 CFR Part 61," November 1982. 

NUREG-1 101, "Onsite Disposal of Radioactive Waste," 1986. 

NUREG-1307, "Report on Waste Burial Charges," Rev. 4, June 1994.

65
NUREG-1337, "Standard Review Plan for the Review of Financial Assurance Mechanisms for 

Decommissioning Under 10 CFR 30,40, 70, and 72," Rev. 1, August 1989. 

NUREG-1444, Supplement 1, "Site Decommissioning Management Plan," November 1995. 

NUREG-1496 (Draft), "Generic Environmental Impact Statement in Support of Rulemaking on 

Radiological Criteria for Decommissioning of NRC-License Nuclear Facilities," Vols. I and 2, August 

1994. 

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

Criteria for License Termination of NRC-Licensed Nuclear Facilities," Vol. 1, July 1997. 

NUREG-1500, "Working Draft Regulatory Guide on Release Criteria for Decommissioning: NRC Staff's 

Draft for Comment," August 1994. 

NM G- 1501, "Background as a Residual Radioactivity Criterion for Decommissioning," August 1994. 

NUREG-1505, "A Nonparametric Statistical Methodology for the Design and Analysis of 

Final Status Decommissioning Surveys," August 1995. 

NUREG-1506, "Measurement Methods for Radiological Surveys in Support of New Decommissioning 

Criteria," August 1995. 

NUREG-1507, "Minimum Detectable Concentrations With Typical Radiation Survey Instruments for 

Various Contaminants and Field Conditions," August 1995. 

NUREG- 1520, "Standard Review Plan for the Review of a License Application for a Fuel Cycle Facility," 

1995. 

NUREG-1573, "Branch Technical Position on a Performance Assessment Methodology for Low-Level 

Radioactive Waste Disposal Facilities," April 1997. 

NUREG-1577, Rev 1, "Standard Review Plan on Power Reactor Licensee Financial Qualifications and 

Decommissioning Funding Assurance (March 1999). 

NUREG/BR-01 11, "Transporting Spent Fuel: Protection Provided Against Severe Highway and Railroad 

Accidents," March 1987. 

NUREG/BR-0241, "NMSS Handbook for Decommissioning Fuel Cycle and Materials Licensees," March 

1997. 

NUREG/CR-0130, "Technology, Safety, and Costs of Decommissioning a Reference Pressurized Water 

Reactor Power Station," June 1978 (Addendum 1, July 1979; Addendum 2, July 1983; Addendum 3, 

September 1984; Addendum 4, July 1988). 

NUREG/CR-0672, "Technology, Safety and Costs of Decommissioning a Reference Boiling Water 

Reactor Power Station," June 1980 (Addendum 1, July 1983; Addendum 2, September 1984; 

66

1I
Addendum 3, July 1988; Addendum 4, December 1990).

NUREG/CR-5512, "Residual Radioactive Contamination From Decommissioning, Technical Basis for 

Translating Contamination Levels to Annual Total Effective Dose Equivalent," Vols. 1 and 2, October 

1992. 

NUREG/CR-5849 (Draft), "Manual for Conducting Radiological Surveys in Support of License 

Termination," June 1992. 

NUREG/CR-5884, "Revised Analysis of Decommissioning for the Reference Pressurized Water Reactor 

Power Station," November 1995. 

NUREG/CR-6174, "Revised Analyses of Decommissioning for the Reference Boiling Water Reactor 

Power Station," (Pacific Northwest National Laboratory), July 1996. 

NUREG/CR-6232, "Assessing the Environmental Availability of Uranium in Soils and Sediments" 

(Pacific Northwest Laboratory), June 1994. 

Regulatory Guides 

DG 1067, "Decommissioning of Nuclear Power Reactors" (draft guide), June 1997. 

DG 1071, "Standard Format and Content for Post-Shutdown Decommissioning Activities Report," 

December 1997. 

RG 1.86, "Termination of Operating Licenses for Nuclear Reactors," June 1974. 

Miscellaneous 

Branch Technical Position, "Disposal or Onsite Storage of Thorium or Uranium Wastes From Past 

Operations," 46 FR 52601, October 1981. 

"Branch Technical Position onsite Characterization for Decommissioning," 1994. 

Commission Paper From EDO, Mr. J. Taylor, to Chairman Jackson and Commissioners Rogers and Dicus, 

"Resolution of Spent Fuel Storage Pool Action Issues," February 1, 1996. 

FC 83-23, Policy and Guidance Directive, "Guidelines for Decontamination of Facilities and Equipment 

Prior to Release for Unrestricted Use or Termination of Byproduct, Source and Special Nuclear Material 

Licensees," November 1983. 

"Guidelines for Decontamination of Facilities and Equipment Prior to Release for Unrestricted Use or 

Termination of Byproduct, Source and Special Nuclear Material Licensees," August 1987. 

PG-8-08, Policy and Guidance Directive "Scenarios for Assessing Potential Doses Associated With 

Residual Radioactivity," May 1994.

67
SECY-94-145, "Increase of Tritium and Iron-55 Unrestricted Use Limits for Surface Contamination at 

Shoreham and Fort St. Vrain." May 1994. 

68

Ii
NRC FORM NO U.S. NUCLEAR REGULATORY COMMISSION 1. REPORT NUMBER 

(2-W .aeftnrd by NRC, AM V.. &,ai., Rev. 

MM. 3 BIBUOGRAPHIC DATA SHEET Addu us.. 

Sk anNUREG-1628 

2. TITLE AND SUBTITLE Final 

Staff Responses to Frequently Asked Questions Concerning Decommissioning of Nuclear Power Plants _ m. ATEREPORT___-r_ 

Final Report June 2000 

4. FIN OR GRANT NUMvBER 

6. DJTmOR(S) 6. TYPE OF REPORT 

J.L Manns, M.T. Masnik Technical 

7. PERIOD COVERED Elkive oD.ms) 

S. PERFORMING ORGANIZATION - NAME AND ADDRESS 9MM.pm vde oW Oft* orJegkno U.& kcM RMogan cn 6- Widnm addiue-, • wIwc, 

Division of Licensing Project Management 

Office of Nuclear Reactor Regulation 

U.S. Nuclear Regulatory Commission 

Washington, DC 20555-0001

S. b?'UrNUt( AN IZATION*NA1A ADDRESS #rCl. 4Wpe 'Si as ebov. fronhcz paidC*w C drhg9iw U.S AbwRogulmy Commisakn. and maft-gadr)

Same as above.

10. SUPPLEMENTIARY NTESI

11. ABSTRACT I200 wards Ur) 

This report through a question and answer format, provides U.S. Nuclear Regulatory Commission (NRC) staff responses to frequently asked questions on the decommissioning process for commercial nuclear power reactors for the 21st century. The questions were taken from a variety of sources over the past several years, Including written inquiries to the NRC and question. asked at public meetings and during Informal discussions with the NRC staff. In responding to the questions, the NRC staff 

attempted to provide the answers In a dear and non-technical form. 

With the Increase In the number of power reactors beginning the decommissioning process and significant changes that occurred In the regulations since 1996, the staff realized that there was a general lack of understanding of the decommissoning process and 

the risks associated with decommissioning. This document was developed in response to the staffs' concerns. The report contains a definon of decommissioning and a discussion of alternatives. It also provides a focus on decommission experiences In the United States and how the NRC regulates the decommissioning process. Questions related to spent fuel, low-level waste, and transportation related to decommissioning are answered. Questions related to license termination, the ultimate deposition of the facility, partial site release, socloeconomics and finances for completing decommissioning and hazards associated with 

decommissioning are also addressed. This document also provides reposes to questions related to public Involvement in 

decommissioning as wel as providing the public with sources for obtaining additional Information on decommissioning.

12. KE=Y WORDS/DESCRIPTORS ¢Lkt mud. @-Fnaa met wa ,ktmwchva hs~cg me I3& AVAJLA STATEOEUNr 

Decommissioning- nuclear power reactors unlirMted Decommissioning- process, sites, regulations, licensing, Inspection programs " , Spent Fuel pools- independent spent fuel storage Installation (ISFSI) trn Pep,) Radioactive low level waste unclassified 

Transportation M 

Licenses termination/Final Disposition of Facility Hazards assodated with Decommissioning N u5. ass FiE F ina n ces1 .N U B R F PA S 

Sodoeconomics 

Public Involvement is. PRIC 

Partial site release

ý rumm OaQ W-W)
Federal Recycling Program 

II
NUREG-167Z --TALFF-IUJPONSES TO YREQUEUNT76Y-AKYD-QUESTIONS CONCERNING-DECOMAMISIONING . JUNE 2000-

OF NUCLEAR POWER PLANTS

UNITED STATES 

NUCLEAR REGULATORY COMMISSION 

WASHINGTON, D.C. 20555-0001

SPECIAL STANDARD'MAIL 

POSTAGE AND FEES PAID, 

USNRC 

PERMIT NO. G-67