ML20236S705
| ML20236S705 | |
| Person / Time | |
|---|---|
| Issue date: | 04/30/1998 |
| From: | Masnik M, Minns J NRC (Affiliation Not Assigned) |
| To: | |
| References | |
| NUREG-1628, NUREG-1628-DRF-FC, NUREG-1628-DRFT, NUDOCS 9807270179 | |
| Download: ML20236S705 (74) | |
Text
u~ --
uu <
e Fge. wnu Nuge6s -
@"" zD c
Staff Responses to Frequently Asked Questions Concerning Decommissioning of Nuclear Power Reactors Draft Report for Comment U.S. Nuclear Regulatory Commission Omce'of Nuclear Reactor Regulation )) .;
i ?! B C I 5 " 19 Jchn L. Minns and Dr. Michael T. Masnik fi
, % ,, ggo a ;
%, * * * * * /
l 72 980430 ;
l 1628 R PDR (
a
AVAILABILITY NOTICE Availability of Reference Materials Cited in NRC Publications Most documents cited in NRC publications will be available from one of the following sources:
- 1. The Nr.C Public Document Room, 2120 L Street, NW., Lower Level, Washington, DC 20555-06v1
- 2. The Superintendent of Documents, U.S. Govemment Printing Office, P. O. Box 37082 Washington, DC 20402-9328
- 3. The National Technical Information Service, Springfield, VA 22161-0002 Although the listing that follows represents the majority of documents cited in NRC publica-tions, it is not intended to be exhaustive.
Referenced documents available for inspection and copying for a fee from the NRC Public Document Room include NRC correspondence and internal NRC memoranda; NRC bulletins, circulars, information notices, inspection and investigation notices; licensee event reports; vendor reports and correspondence; Commission papers; and applicant and licensee docu-monts and correspondence.
The following documents in the NUREG series are available for purchase from the Government Printing Office: formal NRC staff and contractor reports, NRC-sponsored conference pro-ceedings, international agreement reports, grantee reports, and NRC booklets and bro-chures. Also available am regulatory guides, NRC regulations in the Code of Federal Reg %t-tions, and Nuclear Regulator; Commission Issuances.
Documents available from the Nationa: Technical Information Service include NUREG-series reports and technical reports prepared by oBer Federal agencies and reports prepared by the !
Atomic Energy Commission, forerunner agenc/ to the Nuclear Regulatory Commission.
Documents available from public and special technical libraries include all open literature items, such as books, journal articles, and transactions. Federal Register notices, Federal and State legislation, and congressional reports can usua!Iy be obtained from these libraries.
Documents such as theses, dissertations, foreign reports and translations, and non-NRC con-forence proceedings are available for purchase from the organization sponsoring the publica-tion cited.
Single copies of NRC draft reports are available free, to the extent of supply, upon written request to the Office of Administration, Distribution and Mail Services Section, U.S. Nuclear Regulatory Commission, Washington DC 20555-0001.
Copies of industry codes and standards used ir, a substant;ve manner in the NRC regulatory process are maintained at the NRC Library, Two White Filnt North,11545 Rockville Pike, Rock-ville, MD 20852-2738, for use by the public. Codes and standards are usually copi righted and may be purchased from the origina ing organization or, if they are American National Standards, from the American National 3tandards institute,1430 Broadway, New York, NY 10018-3308.
NUREG-1628 Staff Responses to Frequently Asked Questions Concerning Decommissioning of Nuclear Power Reactors Draft Report for Comment Manuscript Completed: March 1998 Date Published: April 1998 John L Minns and Dr. Michael T. Masnik Division of Reactor Program Management Omce of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 p ...
COMMENTS ON DRAFT REPORT Any interested party may subrnit comments on this report for consideration by the NRC staff.
Please specify the report number, draft NUREG-1628, in your comments, and send them by the due date published in the Federal Register notice to:
Chief, Rules Review and Directives Branch Office of Administration Mail Stop T6-D59 Washington, DC 20555-0001 Draft NUREG-1628 April 1998
ABSTRACT 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 publishet in the Federal Register on July 29,1996, and became effective 30 days from the date of publication (August 28,1996). The Dnal 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, r.nd (4) explicitly states the applicability of certain NRC requirements to permanently shutdown reactors.
Sections 1 and 2, define decommissioning and discuss altematives. Sections 3 and 4 focus on decommissioning experiences within the U.S. and on 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 and the ultimate disposition of the facility, hazards associated with decommissioning. Section 10 deals with how the public may become involved in the decommissioning process with an emphasis on the early phases of decommissioning.
Finances are discussed in Section 11. Section 12 provides the public with sources of additional information on decommissioning. The final section contains the references cited in the text and used to prepare this report.
The staff realized that there was a significant lack of public understanding of the decommissioning process and the risks associated with decommissioning. With the recent increase in the number of power reactors beginning the decommissioning process and the significant changes that occurred in the regulations in 1996, this report provides NRC staff responses to frequently asked questions on decommissioning nuclear power reactors.
This document, through a question-and-answer format, provides information to the public on decommissioning. 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 that an individual with no or little technical training could understand.
This document is being issued for public comment. As a result of public comment or peer review and discussions, the final document may be modified from this draft.
April 1998 iii Draft NUREG-1628
CONTENTS Page AB STRA CT . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . i i i AB B REVIATION S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . x iii AC KN O WLE DG M ENTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . x i v 1 G E NE RA L . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . !
1.1 How is decommissioning defined? . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . I 1.2 Why do nuclear power plants shut down? . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . I 1.3 Why are power reactors decommissioned? . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . I 1.4 How does decommissioning proceed? . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . I 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 DECOMMIS SIONING 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? . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 2.3 How is a nuclear power plant decommissioned? . . . . . . . . . . . . . . . . . . . . . . . . . 4 2.4 Is there a limit on the number of years that it would take to decommission a plant? . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 2.5 What alternatives are currently used for decommissioning? . . . . . . . . . . . . . . . . 4 2.6 What are the benefits and costs of the DECON alternative? . . . . . . . . . . . . . . . . 5 2.7 What are the benefits and costs of the SAFSTOR alternative? . . . . . . . . . . . . . . 5 2.8 What are the benefits and costs of the ENTOMB altemative? . . . . . . . . . . . . . . 6 2.9 Is the choice of decommissioning alternatives a decision that is left entirely to the licensee or does the NRC help make this decision? . . . . . . . . 6 2.10 Must a licensee choose either DECON or SAFSTOR, or can it combine the two alternatives? . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7 2.11 What main factors affect a licensee's choice of a decommissioning al ternati ve ? . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7 2.12 What impact would each of the alternatives have on the economy of the surrounding area, especially in terms of the work force requirements ? . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7 3 DECOMMISSIONED S ITES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 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? . . . . . 8
. 3.3 What improvements have been made as a result of previous decommissioning experience? ......................................8 3.4 What research is being performed to find improved methods to be used during decommissioning? . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 April 1998 v Draft NUREG-1628
Contents Page 4 N RC ACTIVITIES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9 4.1 LICEN S ING . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9 l 4.1.1 Does the licensee have an NRC license even during the decommissioning process? . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9 4.1.2 Could the NRC require a plant to cease operations and begin decommissioning? . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9 4.1.3 What would happen if a licensee refuses to decommission a plant that has ceased operations? . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9 4.1.4 What would happen if the license expires before the decommissioning process is concluded? . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10 4.1.5 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? . . . . . . . . . . . 10 4.2 REG ULATION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10 4.2.1 Who regulates the decommissioning process? . . . . . . . . . . . . . . . . . . . . . . . . . 10 4.2.2 When were the current regulations for decommissioning written? . . . . . . . . . . 10 4.2.3 What publications contain the regulations for decommissionir.g? . . . . . . . . . . . I1 4.2.4 What regulatory actions are required to decommission a nuclear facility? . . . . . I1 4.2.5 What activities can take place prior to submitting the PSDAR7 . . . . . . . . . . . . 14 4.2.6 Can the licensee make changes to its plans or commitments after submitting the PS DAR? . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14 4.2.7 What would happen if the licensee wanted to switch from S AFSTOR to DECON ? . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14 4.2.8 Will the PSDAR or the License Termination Plan be reviewed by the U.S. Environmental Protection Agency (EPA)? .................. 15 4.2.9 Under what circumstances would the NRC refuse to allow a licensee to proceed with the decommissioning? . . . . . . . . . . . . . . . . . . . . . . . 15 4.2.10 Would the NRC ever stop a licensee from decommissioning a plant on the basis of information in the PSDAR? . . . . . . . . . . . . . .. . . . . . . . . . 15 4.2.11 Can the NRC require the licensee to follow the regulations, and how would the NRC know if the regulations were not being followed ? . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16 4.2.12 Does the NRC have to review and either approve or disapprove every decision the licensee makes related to decommissioning a facility? . . . . 16 4.2.13 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? . . . . . . . . . . . . . . . . . 16 4.2.14 What would happen if the licensee decides to do something that will endanger the environment during the decommissioning process? . . . . . . . . . . . 17 4.2.15 What were the previous regulations and why were they changed? . . . . . . . . . . 17 Draft NUREG-1628 vi April 1998
Contents Page 4.3 IN S PECTION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 18 4.3.1 What are the goals of the inspection program at nuclear power plants undergoing decommissioning? . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 18 4.3.2 What types of inspections will be conducted? . . . . . . . . . . . . . . . . . . . . . . . . . . 19 4.3.3 WhicL areas receive the greatest emphasis during the inspections? . . . . . . . . . 19 4.3.4 Will the NRC have inspectors on site during decommissioning? . . . . . . . . . . . 19 4.3.5 Will plant security levels be de-emphasized during decommissioning? . . . . . . 20 5 S PENT FUEL . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 20 5.1 What are "high-level wastes"? . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 20 5.2 What is meant by the term " spent fuel"? . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 20 5.3 Are there facilities or plans for facilities for the disposal of high-level radioactive waste? . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 20 5.4 Is the licensee allowed to store the spent fuel in the reactor vessel? . . . . . . . . . 21 5.5 Since there are no facilities for permanent disposal, where will the spent nuclear fuel be kept during the decommissioning process? . . . . . . . . . . . 21 5.6 What are the long-range plans for disposition of spent fuel? . . . . . . . . . . . . . 21 5.7 What happens if a disposal site for high-level waste is never licensed? . . . . . . 21 5.8 S PENT FUEL POOLS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 21 5.8.1 Why is spent fuel stored in a pool of water? . . . . . . . . . . . . . . . . . . . . . . . . . . . 21 5.8.2 Has the spent fuel pool been analyzed to determine the limits Ier hee iemoval due to spent fuel storage? . . . . . . . . . . . . . . . . . . . . . . . . . . . . 22 5.8.3 Do spent fuel pools leak, and if they do, how much radioactive material could be leaked, and where would it go? . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 22 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? . . . . . . . . . . . . . . . . . . . 22 5.8.5 What can be done to prevent the spent fuel pool from boiling dry? . . . . . . . . . 23 5.8.6 How long can the licensee store the spent fuel in the spent fuel pool ? . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 3 5.8.7 What will be done with the spent fuel pool after the fuel has been remo ved ? . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 23 5.8.8 Could the licensee transfer fuel to another licensee's facility? . . . . . . . . . . . . . 23 5.8.9 Can the spent fuel be shipped to another facility's spent fuel pool for storage ? . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 23 5.8.10 Where can I find the regulations relating to the storage of spent fuel in a spent fuel pool ? . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 24 5.9 INDEPENDENT SPENT FUEL STORAGE INSTALLATION . . . . . . . . . . . . 24 S.9.1 What is an independent spent fuel storage installation (ISFSI)? . . . ....... 24 5.9.2 Why would a licensee store spent fuel in an ISFSI rather than in the spent fuel pool? . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 24 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)? . . . . . . . . . . . . . . . . . . . . . . . . 24 April 1998 vii Draft NUREG-1628
Contents Page 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 requi red ? . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 25 5.9.5 Are there any nuclear plants that already use dry-storage casks in an IS FS I ? . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 25 l
5.9.6 How long may the licensee keep the spent fuel in an onsite ISFSI? . . . . . . . . . 25 5.9.7 Can the spent fuel be shipped to another facility's ISFSI for storage? . . . . . . . 25 5.9.8 Why is the licensing process on the ISFSI evaluated separately from the decommissioning process? . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 25 5.9.9 Where are the regulations relating to use ofISFSIs? . . . . . . . . . . . . . . . . . . . . . 26 6 RADIOACTIVE LOW-LEVEL WASTE . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 26 6.1 What is meant by low-level radioactive waste and how is it different from fuel? . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 26 6.2 How is the low-level radioactive waste disposed of? . . . . . . . . . . . . . . . . . . . . 26 6.3 Is low-level waste disposal safe? . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 27 6.4 Where can low-level radioactive waste be disposed of? . . . . . . . . . . . . . . . . . . 27 6.5 What would happen if the waste site that was being used is closed? . . . . . . . . 28 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? . . . . . . . . . . . . . . . . . . 28 6.7 Can radior.ctive waste be buried on site? . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 28 6.8 Where are the regulations relating to radioactive LLW disposal? ......... 28 7 TRANS PORTATION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 28 7.1 How is the spent fuel going to be shipped to a final repository? . . . . . . . . . . . . 28 7.2 How is low-level radioactive waste shipped to the disposal site? . . . . . . . . . . . 29 7.3 Are there regulations on radiation levels during transponation? . . . . . . . . . . . . 29 7.4 Are there safety criteria for spent fuel shipping casks and how are the criteria satisfied? . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 29 7.5 Are there safety criteria for shipping containers for low-level waste and how are the criteria satisfied prior to shipment? . . . . . . . . . . . . . . . . 30 7.6 What regulations apply to the transportation of radioactive materi al ? . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 0 7.7 Will the public be informed beforehand of waste shipments and routes taken? . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 31 7.8 Are specific routes used for transporting radioactive material and does the NRC approve the routes used for radioactive material shipments ? . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 31 7.9 How safe are spent fuel cask shipments? What would happen if the train or truck carrying a spent fuel cask was involved in an accident ? . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 31 7.10 What emergency response plans are in place for transportation accidents involving a spent fuel cask or low-level-waste packages? . . . . . . . . . 31 Draft NUREG-1628 viii April 1998
Contents Page 8 LICENSE TERMINATION AND THE ULTIMATE DISPOSITION OF TH E FACILITY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 2 8.1 How does decommissioning end, and who decides that the decommissioning is complete? . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 32 8.2 What is included in the site characterization? . . . . . . . . . . . . . . . . . . . . . . . . . . 33 8.3 What does " suitable for release" mean? Are there any restrictions on how the site can be used? . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 33 8.4 Why would the licensee be allowed to restrict use of the site? . . . . . . . . . . . . . 34 8.5 What is residual radioactivity and why is it important to the termination of the license? . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 34 8.6 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? . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 34 8.7 What is a " total effective dose equivalent"? . . . . . . . . . . . . . . . . . . . . . . . . . . . 34 8.8 Who would be considered an " average member of the critical group"? . . . . . . 35 8.9 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? . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 35 8.10 How does the dose based on the residual radioactivity levels relate to background dose levels? . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 37 8.11 Why didn't the NRC set the final dose criteria for release of the site for unrestricted use to " pre-existing background" levels? . . . . . . . . . . . . . . . . . . . . 37 8.12 Is it possible that some isotopes are located in such a way that radiation-monitoring devices cannot accurately detect their levels of radioactivity? . . . . 37 8.13 Will continued monitoring be required after the decommissioning process is complete to ensure that the radiation levels do not increase? . . . . . . . . . . . . . . 37 8.14 What types of uses can be made of the plant site after decommissioning is completed? . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 38 8.15 What uses have been made of sites that were decommissioned in the past? . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 38 8.16 What regulations are related to license termination? . . . . . . . . . . . . . . . . . . . . . 38 9 HAZARDS ASSOCIATED WITH DECOMMISSIONING . . . . . . . . . . . . . . . . . . . . . 39 9.1 WO R KERS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 9 9.1.1 Where do the decommissioning workers come from? . . . . . . . . . . . . . . . . . . . . 39 9.1.2 Is worker safety considered in the planning for and review of decommissioning? . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 39 9.1.3 How much occupational dose is received by workers during decommissioning? . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 9 9.1.4 Are there limits on the amount of occupational dose that may be received? . . . 39 9.1.5 Does the licensee have to estimate the occupational dose before the decommissioning process is initiated? . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 40 April 1998 ix Draft NUREG-1628
Contents Page 9.2 P U B LIC . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 0 9.2.1 Is the safety of the public considered in the planning for and review of decommissioning? . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 40 9.2.2 How much dose will the public receive during the decommissioning process? . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 40 9.2.3 Who estimates what the doses are and how are these estimates made? . . . . . . 40 9.2.4 What types of effluent releases are expected, and where will they enter the environment? . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 40 9.2.5 Can you measure the effluent release to know how much is really entering the environment? . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 41 9.2.6 What hazards are presented to the public when the waste is shipped? . . . . . . . 41 9.2.7 What types of accidents at the reactor site are considered and what would be the consequences to the public? . . . . . . . . . . . . . . . . . . . . . . . . . . . . 42 9.3 G ENERAL . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ........... 42 9.3.1 In general, how safe is a decommissioning plant in contra to an operating plant? . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ...... 42 9.3.2 Will there still be emergency preparedness plans and warning sirens in the vicinity of the plant? . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 43
.10 FIN ANCES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 3 10.1 How much does it cost to decommission a nuclear power plant? . . . . . . . . . . 43 10.2 Who makes the estimates of the decommissioning costs? . . . . . . . . . . . . . . . . . 44 10.3 When are the estimates of the decommissioning costs made? . . . . . . . . . . . . . 44 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? . . . . . . . . . . . 45 10.5 How does the NRC ensure that the licensee will have the money when it is needed for decommissioning? . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 45 10.6 Do the financial assurance regulations apply for Federal Government licensees? . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 46 10.7 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? . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 46 10.8 What would happen if the cost of decommissioning exceeds the amount of money in the trust fund? . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 46 10.9 What would happen if the plant has an accident and there is not enough money in the decommissioning trust fund to complete decommissioning and cleanup after the accident? . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 46 10.10 Who pays for decommissioning? . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 46 10.11 What contingency plans are in place to assure 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? . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 47 Draft NUREG.1628 x April 1998
. Contents I
10.12 What will happen if deregulation becomes a reality? How will deregulation affect anticipated revenue and the ability to decommission? . . . . 47 11 PUB LIC INVOLVEM ENT . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 48 11.1 PUB LIC MEETING S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 48 11.1.1 What meetings are planned to keep the public informed? . . . . . . . . . . . . . . . . . 48 11.1.2 Where will the meetings be held? . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 48 11.1.3 How will we be notified about the meetings? . . . . . . . . . . . . . . . . . . . . . . . . . 48 11.1.4 If I cannot attend a meeting, how do I find out what was said? . . . . . . . . . . . . . 48 11.1.5 May I make comments at the meeting? . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 48 11.2 COMMENT PERIOD . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 49 11.2.1 When does the comment period for the decommissioning proce s s start ? . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 9 11.2.2 How do I make comments on the decommissioning process? . . . . . . . . . . . . . . 49 11.2.3 Where should I send my comments? . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 49 11.3 H EAR ING S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 9 11.3.1 Are hearings held on the decommissioning process? . . . . . . . . . . . . . . . . . . . . 49 11.4 GENERAL INFORM ATION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 50 11.4.1 Other than the public meeting, how can I get information about a nuclear power plant? . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 50 12 GETTING ADDITIONAL INFORMATION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 50 12.1 What is a local public document room and how can I find one? . . . . . . . . . . . 50 12.2 Does the NRC have a website? What kind ofinformation can l obtain from it ? . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 51 12.3 Does the NRC maintain an electronic bulletin board system? . . . . . . . . . . . . . 51 12.4 What is the Federal Register and how can I get a copy of it? . . . . . . . . . . . . . . 52 12.5 How can I get a copy of the Code of Federal Regulations? . . . . . . . . . . . . . . . 52 12.6 How can I get answers to additional questions that were not addressed in this document? . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 52 REFERENCES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 3 April 1998 xi Draft NUREG-1628
ABBREVIATIONS 1 ADS- automatic depressurization system CMT core makeup tank CVCS chemical and volume control system DBA design-basis accident DECIfr double-ended cold-leg (break)
ECCS- emergency core cooling system .,
EM Evaluation Model IRWST in<ontainment refueling water storage tank LOCA- loss-of-coolant accident MSLB main steam line break PCS passive cooling system PWR pressurized-water reactor SG steam generator SRP Standard Review Plan l
1 1
April 1998 xiii Draft NUREG-1628
ACKNOWLEDGMENTS -
l This report, which represents the cumulative efforts of many individuals who contributed to its l
publication over the last six months, was facilitated by bringing together a number of subject I 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 Wu, Ms. Stacey Rosenberg, Mr. Chris Gratton, Mr. R. Wood, Ms. Elizabeth A. Hayden, .
Mr. Thomas Fredericks, Mr. Richard F. Dudley Jr., Ms. Ann Hodgdon, Mr. Anthony Markley, and other staff of the Decommissioning Section within NRR. The authors gratefully acknowledge the assistance provided by Rebekah Harty and other individuals of the Pacific Northwest National Laboratory (PNNL) in the preparation of an early draft of this report. A special thanks to Ms. Ray Sanders, Technical Editor.
Draft NUREG-1628 xiv April 1998
Staff Responses to Frequently Asked Questions on Decommissioning Nuclear Power Reactors l
1 GENERAL 1.1 How is decommissioning defined?
Title 10 of the Code of FederalRegulations, 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?
Nuclear power plants cease operations for a variety of reasons. The NRC grants a l!:ense 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 economicallyjustifiable 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 decommissioning 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 decommissioning 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.
April 1998 1 Draft NUREG-1628
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 radioactive contamination on the site so that it could adversely affect public heahh and safety and the environment in the future.
1.6 What are the costs of decommissioning?
The major costs of decommissioning are the la:ge financial costs involved in funding the project.
The occupational dose received by workers during decommissioning could also be considered a cost since there is no corresponding direct societal benefit as from power production (although there is a benefit from ensuring that the facility is decommissioned so that it does not become a liability in the future).
1.7 What are the options to decommissioning?
NRC's regulations do not allow the option of not decommissioning. The alternative to decommissioning 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 radioactive facility to such a condition that there is no unreasonable 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 power 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?
Several terms are important to the understanding of decommissioning: " radiation,"
" contamination," " activation products," " dose," " radioactive decay," and " half-life." It is also important to gain an understanding of the units used for measuring radiation dose: " rem" and
" person-rem."
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.
Contamination means undesired (e.g., radioactive or hazardous) material that is (1) deposited on the surface of, or intemally ingrained into, structures or equipment, or (2) mixed with another material.
Draft NUREG-1628 2 April 1998
Activation products are radioactive materials that were created when stable substances were i 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. 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.
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. Radiation dose receivel by a person is measured in units called " rem."
Radioactive decay is the spontaneous natural process by which an unstable radioactive nucleus releases energy or particles.
Halflife is the time required for half of any quantity ofidentical 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 (1000 millirem = 1 rem =
0.01 sievert).
Rem (see Dose).
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 mo,t likely that contamination will occur in the reactor building, around the spent fuel pool, and areand specific pieces of equipment in the auxiliary building. The areas known to contain contamim. tion 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. The radioactive decay of activation products is the main source of April 1998 3 Draft NUREG-1628
radiation exposure to plant personnel. Radiation activation products can be anywhere reactor coolant circulates, leaks, or is processed.
2.3 IIow is a nuclear power plant decommissioned?
To decommission a nuclear power plant, the radioactive material on the site must be reduced to j
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-storage facility. Contaminated materials may either be cleaned of contamination on site, 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 the waste-storage 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-storage site.
2.4 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. Completion of decommissioning beyond 60 years will be approved by the NRC only when necessary to protect public health and safety.
2.5 What alternatives are currently used for decommissioning?
The NRC has evaluated the environmentalimpacts 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 S AFSTOR 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.
Draft NUREG-1628 4 April 1998
2.6 What are the benefits and costs of the DECON alternative?
l The DECON option calls for prompt removal of radioactivity to permit restricted or unrestricted access. The advantages of DECON include the following:
! . facility license 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; a 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 oflow-level waste facilities that would be willing to accept the low-level radioactive waste; and 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);
e a larger initial commitment of money; e
a larger commitment of disposal site space than for the SAFSTOR option ; and the potential for complications if spent fuel must remain on the site until a Federal repository for spent fuel becomes available.
2.7 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 reduction in public exposure because of fewer shipments of radioactive material to the low-level waste site (as compared to the DECON alternative);
April 1998 5 Draft NUREG-1628
. a reduction in the amount of waste disposal space required (as compared to the DECON alternative);
= lower cost during the years immediately following permanent cessation of operations; and
. a storage period compatible with the need to store spent fuel on site.
l Disadvantages of S AFSTOR include the following:
= shortage of personnel familiar with the facility at the time of deferred dismantlement and decontamination; e site unavailable for alternate uses during the extended storage period; a uncertainties regarding the availability and costs oflow-level radioactive waste sites in the future;
= continuing need for maintenance, security, and surveillance; and
. higher total cost for the subsequent decontamination and dismantlement period (assuming typical price escalation during the time the facility is stored).
2.8 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. This option might be acceptable for reactor facilities that can demonstrate that radionuclides levels will decay to levels that will allow restricted use of the site. Three small demonstration reactors have been entombed. Currently, no power reactor licensees have proposed the ENTOMB option for any of the power reactors undergoing decommissioning.
2.9 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 reevaluate its decision if the choice (1) 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.
Draft NUREG-1628 6 April 1998
2.10 Must a licensee choose either DECON or SAFSTOR, or can it combine the two alternatives?
A licensee need not restrict its choice of decommiss'ioning 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 stop at a facility, a licensee could use a short storage period for planning purposes, followed by removal oflarge 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.
2.11 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 oflow-level waste sites. The licensee rate regulator can also influence the choice of decommissioning alternatives.
2.12 What impact would each of the alternatives have on the economy of the surrounding area, especially in terms of the work force requirements?
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 oflarge components like the steam generators and pressurizer. If the plant were placed in S AFSTOR, 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 fmther supplemented with contractor personnel for the final cleanup of the site.
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. Typically, additional public services are not required during decommissioning because the plant staff will be smaller than the operating staff. Contractual arrangements with State and local governments for emergency preparedness programs can be eliminated once activities at the facility can no longer exceed the U.S. Environmental Protection Agency (EPA) Protective Action Guidelines at the site boundary.
April 1998 7 Draft NUREG-1628
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 January 1998, radiological decommissioning had been completed at three NRC-licensed power reactors: the Pathfinder test reactor m Sioux Falls, South Dakota; Fort St. Vrain Generating Station in Plateville, Colorado; and the Shoreham Nuclear Power Station in Suffolk County, New York. (The Shoreham plant had only operated 2 effective full-power days.) Six nui;ar power reactors are now in various stages of dismantlement and decontamination: the Saxton reactor in Saxton, Pennsylvania: Haddam Neck plant in Haddam Neck, Connecticut; Maine Yankee plant in Wiscassett, Maine; Big Rock Point plant in Charlevoix, Michigan; Trojan plau 'n Rainier, Oregon; and Yankee Rowe plant in Franklin County, Massachusetts. Eleven nuclear power reactors are in, or are planning, long-term storage: Indian Point 1 in Buchanan, New York: Dresden 1 in Morris, Illinois; Humboldt Bay 3 in Eureka, California; Peach Bottom I in York County, Pennsylvania: San Onofre 1 in San Clemente, California; Rancho Seco in Sacramento, California; Lacrosse in Lacrosse, Wisconsin; Vallecitos Boiling-Water Reactor in Pleasanton, California; Fermi 1 in Monroe County, Michigan, Zion Units 1 and 2 in Zion, Illinois. Three-Mile Island, Unit 2 prematurely shut down as the result of an accident and holds a possession-only license.
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 alternativ6. Three DOE facilities-Bonus in Puerto Rico, Hallam in Nebraska, and Piqua in Ohio-have been decommissioned using the ENTOMB alternative, 3.3 What improvements have been made as a result of previous decommissioning experience?
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 decommissioning. 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 iclude strippable coatings oflatex 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 Depanment of Energy has taken the lead on decommissioning technology research. The following types ofimprovements are being investigated:
Draft NUREG-1628 8 April 1998
. 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 electredes to leach ionic contaminants from deep inside porous concrete);
. cutting techniques, to remove structures, such as laser cutting, or oxy-gasoline torches (which work twice as fast as an acetylene torch on 1-inch steel);
e 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; and a survey / monitoring techniques, such as pipe-explorer internal survey / characterization systems and remote 3-D characterization and archiving system (robotic sensor and mapping platforms analyze for hazardous organic and radioactive contaminants).
Promising avenues of research are developed into usable technologies by commercial firms.
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 is not terminated until the licensee can demonstrate that it meets the criteria for site release in the regulations. This is demonstrated by a final radiation survey that is reviewed and verified by the NRC staff. In addition, the licensee must demonstrate that the facility has been dismantled in accordance with the approved license termination plan.
4.1.2 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.
4.1.3 What would happen if a licensee refuses to decommission a plant that has ceased operations?
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 April 1998 9 Draft NUREG-1628
order, or for violation of any term, condition, or limitation of any license. In addition, the Atomic Energy Act provides for the Federal Govemment to assume responsibility for decommissioning if public health and safety are jeopardized because of inactivity on the part of the licensee.
4.1.4 What would happen if the licensee's license ap!res before the decommissioning process is concluded?
l 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.
4.1.5 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 thejurisdiction of the NRC. The NRC has no further oversight of these activities once the license is terminated.
4.2 REGULATIONS 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 enferccment 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 When were the current regulations for decommissioning written?
The NRC continually reviews the decommissioning regulations that it is using and revises them as required in order to improve the regulatory process. On July 29,1996, a final mle amending the regulations on decommissioning procedures was published in the Federal Register. On July 21,1997, the NRC published (also in the Federal Register) a final rule entitled, " Radiological Draft NUREG-1628 10 April 1998
l l - Criteria for License Termination." 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.
4.2.3 .What publications contain the regulations for decommissioning?
Regulations regarding decommissioning of NRC-licensed plants appear in the Code ofFederal Regulations. The Code ofFederalRegulations 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 funher subdivided into parts covering specific regulatory areas.
The regulations related to decommissioning of power reactors are included in Title 10
(" Energy"), Chapter I-Nuclear Regulatory Commission, e.g., Part 20, " Standards for Protection Against Radiation"; Part 50 " Domestic Licensing of Production and Utilization Facilities"; and Part 51 " 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, "Altemate 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 te.chnical and financial criteria for decommissioning licensed nuclear facilities. They address decommissioning, planning needs, timing, funding methods, and environmental review requirements.
See the answer to question 12.5 for instmetions on how to obtain a copy of the regulation.
4.2.4 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 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 annuallicense 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, e contain a schedule for the accomplishment of significant milestones, April 1998 11 Draft NUREG-1628
. provide an estimate of expected cost, and a provide documentation that environmentalimpacts 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 l
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.
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 require 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:
a permanent removal of major radioactive components, such as the reactor vessel, steam generators, or other components that are comparably radioactive; e permanent changes to the containment stmeture; and
. 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.
Draft NUREG-1628 12 April 1998 L
Within 2 years following the date of permanent cessation of operations, the licensee 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.
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,
. plans for site remediation, a 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, and 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 is authorized to implement the license termination plan.
At the end of the license termination plan process,if the NRC determines that the remaining dismantlement 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.
April 1998 13 Draft NUREG-1628
4.2.5 What activities can take place prior to submitting the PSDAR?
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 i permanent removal of major radioactive components, permanently modifies the structure of the l containment, or results in dismantling components for shipment containing greater than Class C waste." Major radioactive components are defm' ed by the regulations as the reactor vessel and intemals, steam generators, pressurizer, 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) remos a 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 off site; and (6) site characterization and measurement of contamination levels.
4.2.6 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 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 milestorie 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.
4.2.7 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 Draft NUREG-1628 14 April 1998
4 L
activities is significantly greater than that predicted in a previous environmental impact
( , statement, or assessment (see NUREG-0586, " Final Generic Environmental Impact Statement on l-l 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 l - the impact of the change. The NRC staff would review the licensee's submittal and would l - publish either an environmental assessment or a supplement to the facility's final environment statement in conjunction with review of the license amendment request.
4.2.8 L Will the PSDAR or the License Termination Plan be reviewed by the U.S.
l Environmental Protection Agency (EPA)?
l 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.
L
' 4.2.9 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 preceding the initiation of major decommissioning activities.
.'4.2.10 Would the NRC ever stop a licensee from decommissioning a plant on the basis of
- information in the PSDAR?
.I
- 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 cbuld not be completed within 60 years of the permanent cessation of operations, unless it were sho'wn 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 ifit 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 f- in a major detrimental impact to the environment that was not bounded by the current I. environmentalimpact statements.
' April 1998 15 Draft NUREG-1628
- TQ r .
w a
s 4.2.11 Can the NRC require the licensee to follow the regulations, and how would the NRC know if the regulations were n'ot being followed?
The NRC reviews licensee procedures and work 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 active decontamination and dismantlement. Generally, an NRC resident inspector will remain on site 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).
4.2.12 Does the NRC have to review and either approve or disapprove every decision the licensee makes related to decommissioning a facility?
Nc. 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, unless the proposed changes involve the modi 6 cation of the technical specifications that are incorporated into the license or if an unreviewed safety question is identified:
a changes in the facility as described in the safety analysis report, a changes in the procedures as described in the safety analysis report, and a 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 unless the proposed change, test, or experiment involves a change (1) in the technical specifications incorporated in the license or (2) an unreviewed safety questions. An unreviewed safety question is involved (1)if the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety and previously evaluated in the safety analysis repon may be increased,(2)if the possibility of an accident or malfunction of a different type from any evaluated previously in the safety analysis report may be created, or (3)if the margin of safety as defined in the " basis" for any technical specification is reduced.
4.2.13 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 falls to identify an unreviewed safety question?
The licensee is required to periodically submit a report containing a brien description of any changes, tests, and experiments made at the facility, and to summarize the safety evaluation of each. This repon 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 Draft NUREG-1628 16 April 1998
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 J or if there are questions as to whether a proposed change is an unreviewed safety question, the l i
NRC staff will evaluate this issue until it is appropriately resolved.
4.2.14 What would happen if the licensee decides to do something that will endanger the environment during the decommissioning process?
The 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 availab!:: for decommissioning. If any major decommissioning activity would 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 1. 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 changes that are to be made to the facility. Additionally, periodic inspections of the licensee's activities will focus on the environment.
4.2.15 What were the previous regulations and why were they changed?
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 ceatained 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) oiher 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 cf controls and limits on procedures and equipment to protect occupational and public health and safety, (3) a description of the planned final radiation survey, (4) an updated cost estimate for the chosen l alternative and a plan for assuring 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 j
substantive environmental impacts that were anticipated but not already covered in other l environmental impact documents was also required.
April 1998 17 Draft NUREG-1628
The NRC reviewed the decommissioning plan. The Commission would approve the plan if the j plan demonstrated that the decommissioning would be performed in accordance with regulations !
and there were no secway, health, or safety issues, and after notice was 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 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 decommissioning 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 decommissioning 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 uore 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 4.2.4.
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 e 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);
e identify any significant declining performance trends and verify that the licensee has taken actions to reverse any trend; Draft NUREG-1628 18 April 1998 I
5 r
. focus on the important elements of decommissioning to reduce the regulatory burden on
. the licensees; ~and -
e adequately manage the inspection resources to ensure adeq'uate coverage at all' facilities undergoing decommissioning.
l43.2 What types of haspections 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 subelements. Each subelement has a specific frequency based on whether the i 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 detail'ed reviews of a particular functional nea. Examples are high-level waste transportation, internal and external dose assessment, and confirmatory surveys.
. 433 L Which areas receive the greatest emphasis during the inspections?-
The inspection effort during decommissioning places the greatest emphasis on radiological controls, management, procedure compliance, effluent controls, and the safety review program.
f43.4 Will the NRC have inspectors on ' site during decommissioning?
.The inspection effort at a plant undergoing decommissioning is significantly less than at an operating reactor site. Operating reactor facilities have onsite inspectors who maintain a
['
' continual presence at the plant or in the vicinity. However, because of the reduced hazard during
'the decommissi,oning 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 L permanent cessation of operations. The resident inspector's stay at a site can be extended L . because of special circumstances, Rather than stationing a resident inspector at the site, the NRC
~will provide subject matter experts to cover specific activities occurring at the site. For example, L 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 available to i I
review the licen'see's procedures and to inspect before and during specific operations. The extent
- of onsite presence at the facility will be dependent on the activities that are taking place. During 1 April 1998 19 Draft NUREG-1628 l
_ _ _ _ - _ _ - - - - _ _ _ __ ___ _ _ - - _ _ _ - - _ _ - - __ _ _ J
l o
i 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.
- 4.3.5 W' ill 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 on a
. case-by-case basis.-
' 5 SPENT FUEL 5.1 .What are "high-level wastes"? -
High-level radioactive wastes (HLWs) are (1)' irradiated (spent ) reactor fuel; (2) liquid waste resulting from the operation of the first-cycle solvent-extraction system, and the concentrated wastes from subsequent extraction cycles in a facility for reprocessing irradiated fuel; and (3) solids into which such liquid wastes have been converted. HLW is primarily in the form of spent fuel discharged from commercial nuclear power reactors.- A byproduct of nuclear reactions, they are in the form of either spent fuel or materials remaining after reprocessing of spent fuel. i 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. This spent (used) fuel contains radioactive material resulting from the fission process that takes place within the reactor. The radioactive material is formed within ceramic fuel pellets about the diameter of an aspirin tablet but twice as long as this diameter. - These
~
pellets are contained in about 13-foot-long sealed metal tubes or rods. The rods are arrayed to form a spent fuel assembly. When spent fuel is removed from the reactor, the self-sustaining .
' fission process has stopped; however, spent fuel assemblies still generate significant radiation and heat.' This heat and radiation are caused by " radioactive decay" of the pr4ucts 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 Am 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 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, for a possible disposal facility, which would be built and operated by DOE and licensed by the NRC.
l.
Draft NUREG-1628 - 20 April 1998
h . 5.4 : Is the licensee allowed to store the spent fuel in the reactor vessel?
L
- The licensee must submit a cenification indicating that it has 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. 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 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?
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 April 1998 -21 Draft NUREG-1628
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 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 reactors have a reinforced-concrete spent fuel pool (SFP) structure designed to retain its function, even following the design-basis seismic event (i.e., seismic Category I or Class 1 (earthquake)) that is anticipated for the area. The SFP also has a welded, corrosion-resistant liner. 'All reactors 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 prev _enting 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 shielding 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 containment 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 water in spent fuel pools is constantly cleaned and requirements exist regarding the purity of water and the type of water chemistry that is used. 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 or water in a
. spent fuel pool when it was fully loaded?
Depending on the amount of time since the reactor last operated, it would take several days for
. the temperature of the water to rise to boiling from the constant temperature that it is normally Draft NUREG-1628 22 April 1998 -
1 l
maintained. After sufficient time, the heat generated by the fuel would be insufficient to even cause boiling in the pool. The longer the time interval since the last batch of fuel was moved from the reactor to the spent fuel pool, the longer it would take for the water to n:ach the boiling point. After approximately a week or more from the onset of boiling without mitigating action, the water level in the spent fuel pool would drop to a level at which it would no longer provide adequate radiation protection for workers. If the water level were allowed to fall lower than this level, dose rate levels would continue to increase, and at some point the fuel pool would boil dry.
This would cause very high radiation levels in the vicinity of the spent fuel pool. The NRC has found that, after sufficient time has elapsed since the fuel was last used to generate power, even under the worst case, of a catastrophic immediate loss of spent fuel pool coolant, no offsite doses to the public it excess of Federal limits would occur. The NRC staffis preparing a draft guide (DG-1069), which discusses fire protection in the spent fuel area.
5.8.5 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 required 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.
5.8.6 How long can the licensee store the spent fuel in the spent fuel pool?
The Commission he made a generic determination 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.
5.8.7 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.
5.8.8 Could the licensee transfer fuel to another licensee's facility?
If the fuel that came out of the reactor vessel was 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.
'5.8.9 'Can the spent fuel be shipped to another facility's spent fuel pool for storage?
o Yes. NRC regulations do not prohibit spent fuel from one facility being stored in the spent fuel j pool at another facility. As noted above, 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.
April 1998 23 Draft NUREG-1628
L 5.8.10 Where can I find the regulations relating to the storage of spent fuelin a spent fuel pool?
L 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 permanent l
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 spent fuel pools are in Title 10 (" Energy"), Chapter I-Nuclear Regulatory Commission, Part 50," Domestic Licensing of Production and Utilizat'on Facilities,"
along with the regulations and standards for construction permits and operating licenses for nuclear power plants.
i See the answer to question 12.5 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 ad 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?
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. Thus, the license for the nuclear power reactor can be terminated while the ISFSI, which has a separate license, is still located on the facility site.
In addition, it is both cheaper and easier to maintain spent fuel in dry storage in an ISFSI than in
, a spent fuel pool.
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
. Draft NUREG-1628 24 April 1998 o _ _ _ _ _ _ _ _ _ _ _ _ . . . . ._ :. . ._ .. ..
j mechanical devices for cooling or ventilation. Casks must be safety-tested to withstand a variety i
of disasters, such as floods, projectiles originating from a tomado, 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 either vertically on a concrete pad inside a concrete storage building or are inserted horizontally into a stet.1-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 56 spent fuel assemblies. The maximum amount of heat that is generated by the cask will be less than that given off by 240100-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 confm' e 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 (metal rods around 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 ISFSI?
The first dry-storage installation was licensed by the NRC in 1970. Currently,10 nuclear power plants are using dry storage: Surry, Oconee, H.B. Robinson, Calvert Cliffs, Fort St. Vrain, Palisades, Point Beach, Prairie Island, Davis-Besse, and Arkansas Nuclear One.
5.9.6 Ilow long may the licensee keep the spent fuel in an onsite ISFSI?
The NRC has determined that spent fuel can be stored on site 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 fuel storage installations.
5.9.7 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 ISFSIlocated at another facility with appropriate license conditions. The spent fuel would have to be transported in a cask that met NRC requirements for shipping casks.
5.9.8 Why is the licensing process on the 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; thus the April 1998 25 Draft NUREG-1628
licensing process for an ISFSI is not contained in the decommissioning regulations. In addition, 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 the licenses be separate for the two facilities. The decommissioning of the ISFSIis also handled separately from the decommissioning of the nuclear power plant.
5.9.9 Where are the regulations relating to use of ISFSIs?
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 depanments 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 W aste."
See the answer to question 12.5 for instructions on how to obtain a copy of the regulation.
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 uranium-enrichment processes, 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 oflonger half-life radioactive material than Class B and C wastes.
6.2 IIow is the low-level radioactive waste disposed of?
LLW 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).
-Draft NUREG-1628 26 April 1998
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 ofindustry efforts to compact or incinerate part of the waste.
6.3 Is low-level waste disposal safe?
LLW facilities are sited in areas that are away from surface water and where the groundwater is located 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 assure 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 nass wasting, erosion, slumping, and landslides; (4) a water table of sufficient depth so that ground-water will not periodically intrude into the waste or discharge on site; (5) lack of susceptibility to tectonic processes; (6) no known poteritially 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 stmetures designed to improve the site's natural ability 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. They are located in Barnwell (South Carolina),
Hanford (Washington), and Clive (Utah). The site in Utah is restricted to specific types of low-activity waste. 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.
April 1998 27 Draft NUREG-1628
L
?
i
- 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 I
~
have decided to share resources to develop and maintain an LLW disposal site. Nine compacts I
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 thut 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 should 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 on site.
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, LLW is normally stored on site on an interim basis before being shipped off site 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 he buried on site?
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 (i.e.,
approved by the State).
i.
16.8 ' Where are the regulations relating to radioactive LLW disposal?
The regulations related to LLW disposal are in 10 CFR Part 61 and 10 CFR Part 20 Subpart K.
. See the answer to question 12.5 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 wili be shipped in specially designed shipping casks. The design, construction,
- Draft NUREG-1628 28 April 1998
i h.
i l
2 1
.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 i shipment.
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. These 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 is measured at the outside surface L of a package, and in some cases, of the vehicle itself. This depends upon whether the shipment is l being made as exclusive or non-exclusive use shipments. For non-exclusive-use shipments, the radiation level must not exceed applicable limits (1) at the external surface of the package and (2) at 3.3 feet (1 meter) from the external surface of the package. For exclusive-use shipments, the radiation level at 1 meter from the external surface of the package is allowed to be higher than it is for non-exclusive-use shipments, but additional regulations are stipulated that govern radiation levels (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 O 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 l_ 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 I compression tests. The tests used to provide reasonable assurance that the casks will withstand serious transportation accidents include the following:
i j
. a 30-foot drop onto a flat, unyielding surface
. a 40-inch drop.onto a vertical steel rod
- a 30-minute exposure to a fire of 1475 *F 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 l partial-scale tests. An approval certificate must be issued by the NRC before a particular cask design can be used to transport spent fuel.
29
7.5 Are there safety criteria for shipping containers for low-level waste and how are the criteria satisfied prior to shipment?
Yes. LLW is shipped in containers that are designed to NRC and U.S. Department of
' Transportation (DOT) standards. Packaging requirements for specific radioactive materials shipments 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 specific activity.
These container types are strong tight containers (STCs), industrial packages (IPs), 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 STCs 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.-
Two additional types of containers are used for LLWs that contain higher levels of radioactivity.
The safety criteria for these container types, Type B containers and NRC Type A for low specific activity (LSA) containers, are stipulated in the NRC regulations. Type B containers must be able to withstand the normal and accident conditions that are described in the answer to question 7.4.
For all types of containers, additional regulations apply that (1) specify preliminary J i
determinations that must be ascertained before the first use of any container (e.g., no cracks, no
. defects) and (2) specify routine determinations that must be satisfied before each shipment (e.g., .,
1 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 regulatedjointly at the Federal level by the U.S.
l 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 generhl 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, hand _ ling, storage, emergency response information, and employee training). DOT regulations are located in the Code ofFederal Regulations, Title 49 (" Transportation").
'NRC-Regulates users of radioactive material and the design, constmetion, use, and maintenance (of shipping containers used for larger quantities of radioactive material. NRC regulations are located in the Code of Federal Regulations, Title 10 (" Energy"), Part 71, (" Packaging and Transportation of Radioactive Material").
- See the answer to question 12.5 for instructions on how to obtain a copy of the regulations,
- j. ;
=
n ,
1 Draft NUREG-1628 30 April 1998 4
l 7.7 Will the public be informed beforehand of waste shipments and 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 radioactive material wastes or the routes that the shipments will take.
7.8 Are specific routes used for transporting radioactive material and does the NRC approve the routes used for radioactive material shipn.ents?
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 requirements limit these rhipments 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 spent fuel cask shipments? 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, spent fuel cask shipments are very safe. There have been no recorded injuries or fatalities attributed to spent fuel in over 1,000 shipments in the U.S. over the last 30 years. In 1986, the NRC contracted with Lawrence Livermore National Laboratory (LLNL) 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-0111," 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.
l The authorities that are likely to respond to transportation incidents (e.g., police, fire fighters) have been provided with Emergency Response Guidebooks by DOT; these identify the potential i hazards of materials and discuss corresponding mitigative actions. .
I i
j Each State has emergency plans in place for responding to incidents involving transportation of l April 1998 31 Draft NUREG-1628
s radioactive material. To assist the States in the development of these plans, a guide exists titled
" Guide and Example Plan for Development of State Emergency Response Plans and Systems for Transponation-Related Radiation Incidents"; it was prepared by 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 l transportation incidents. In the responses to the survey, the States indicated that they have the 1 necessary plans and procedures in place. This survey is documented in NUREG/CR-5399, l
" 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 (FRERP) 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.
8 LICENSETERMINATION 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 two years before the requested termination date of the license. The license termination plan must include a a site characterization e identification of remaining dismantlement activities
= plans for site remediation e 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 repon After receiving the license termination plan, the NRC places a notice of the receipt of the plan in the FederalRegister, and makes th:: 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. 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
' Draft NUREG-1628 32 April 1998
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 is authorized to implement the license termination plan.
At the end of the license termination plan process, if the NRC determines that the remaining dismantlement 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 terminates the license, and the decommissioning process is considered to have been completed.
8.2 What is included in the site character 2zation?
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; and (3) a map of remaining contamination levels and contamination locations. The purpose of the site characterization is to assist in the planning for remediation, the selection of remediation techniques, and the assessment of radiological impacts and cost estimates.
8.3 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
" unrestricted 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, and the licensee made provisions for legally enforceable institutional controls (e.g., 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 ofindividuals 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.
April 1998 33 Draft NUREG-1628
8.4 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 on site, or by restricting agricultural or l drinking water use. For many facilities, the time period requiring this type of restriction can be l 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.
8.5 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.
8.6 What a-e 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 etfective 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.
Determination oflevels that are as low as reasonably achievable must take into account consideration of any detriments, such as deaths from transportation accidents that are expected to potentially result from decontamination and waste disposal.
8.7 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.
Draft NUREG-1628 34 April 1998
t 8.8 Who would be considered an " average member of the critical group"?
The critical group is an individual or relatively homogeneous group ofindividuals expected to received the highest exposure within the assumptions of the particular scenario. The average 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, the critical group for a scenario in which people work inside a building would be the group of regular employees working in a building that has been decontaminated. If the site were convened to residential use, the critical group could be people whose occupations involve resident farming at the site, not an average of all the residents on the site.
8.9 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 millisiever:) per year with institutional controls in place and either 100 millirem (1 millisievert) per year or 500 millirem (5 millisievert) per year 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, or ownership by the Federal or State government, thus providing for a legal mechanism to restrict public access.
For the first case,100 millirem (1 millisievert) per year, the licensee must demonstrate the following:
=
Further reductions in residual radioactivity would result in net public or environmental harm or were not made because the residual levels are as low as reasonably achievable taking into account 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.
Funds have been placed into an account segregated from other assets and outside of the licensee's administrative contro!s 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, April 1998 35 Draft NURF.G-1628
(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 (1-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 assure 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 opprtunity 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 millisieurt) 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, e 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, and a submit a license termination plan, which specifies that it plans to decommission by use of altemate 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 community interests, (2) an opportunity for a Draft NUREG-1628 36 April 1998
comprehensive collective discussion on the issues, and (3) a publicly available summary of the results of all such discussions.
The use of altemate 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 (EPA) or the public.
8.10 How dces 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 radoa, and global fallout as it exists in the environment from the testing of nuclear explosive devices or from earlier nuclear accidents, such as Chomobyl, that contributes to background radiation and is not under the control of the licensee. " Distinguishable from background" means that the detectable concentration of a radionuclides is statistically different from the background concentration of that radionuclides in the vicinity of the site.
8.11 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 ofincreasingly larger volumes of concrete and soil would also result in a greater net risk due to deaths from transportation accidents.
8.12 Is it possible that some isotopes are located in such a way that radiation monitoring devices cannot accurately detect their levels of radioactivity?
'n is urilikely that radioactive material located inside a piece of equipment or a structure is not detected during the fm' al 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.
8.13 Will continued monitoring he required after the demmmissioning 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 change-other than to be reduced over time-because the radioactive material will April 1998 37 Draft NUREG-1628
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 l
l 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.
8.14 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, tMre ve no restrictions on the type of use. Possible uses could range from restoring the natural habicat, to farming, to continued use as an industrial site (possibly leaving buildings and installing n gas ,
coal , or oil-powered generating plant).
8.15 What uses have been made of sites that were decommissioned in the past?
The licensee that held the license for the Fort St. Vrain 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.
8.16 What regulations are related to license termination?
Regulations regarding license termination appear in the Code of Federal Regulations. The Code ofFederal 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 license termination are in Title 10 (" Energy"), Part 20, " Standards for Protection Against Radiation," Subpart E," Radiological Criteria for License Termination."
See the answer to question 12.5 for instructions on how to obtain a copy of the regulations.
Draft NUREG-1628 38 Aprii 1998
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. Somejobs, however, may be contracted out to companies that have gained experience at other plants in specialized areas of decommissioning or dismantlement.
Facilities that are placed in a storage mode will initially have very few employees during storage and will likely have to hire a new group of workers 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 ofindustrial safety.
9.1.3 Ilow 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 a generic study of decommissioning (published in 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?
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.
April 1998 39 Draft NUREG-1628
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 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 only exposure anticipated to the public is from the shipment of low-level waste from the site to the low-level waste disposal site. Estimates made in a generic study of nuclear power reactor decommissioning 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 for an hour at a rest stop is a very small fraction of the annual dose that the person would receive from background radiation.
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 assumptions, models, and codes. The NRC may review 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 Draft NUREG-1628 40 April 1998
i 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, radioactivity in air effluents is expected to be limited primarily to airborne radioactive particulate. The particulate 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.
Liquid radioactive wastes are collected, stored, and processed in either the clean radioactive waste system or the dirty radioactive waste system, depending on the amount of particulate and the source of the liquid wastes. The liquid waste is processed before it is released 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 (NPDES) waste discharge permit. These permits are regulated by the States or the U.S. EPA.
Direct exposure would only result from the public being 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 I 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 (OCDM). Release limits given in the OCDM are based on NRC regulations.
9.2.6 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 earlier. However, there is a April 1998 41 Draft NUREG-1628
potential hazard resulting from the release of radioactive material due to a major accident involving a vehicle canying radioactive waste.
9.2.7 What types of accidents et the reactor site are considered and what would be the consequences to.the public?
A generic study of decommissioning analyzed the results of the following accidents:
a explosion in an LLW storage area of a liquid propane gas cylinder leaked from a front-end loader
. explosion of an oxyacetylene torch during segmenting of the reactor vessel shell e explosion and/or fire in the ion exchange resin a large leak during primary system decontamination flush
.. segmentation of reactor coolant system piping with unremoved contamination contained in the pipe a loss of contamination control envelope during oxyacetylene cutting of the reactor vessel shell .
. vacuum bag rupture of a contained vacuum cleanup device e accidental cutting of contaminated piping
. accidental spraying of concentrated contamination with the high-pressure spray l'
- filter damage from blasting surges (for boiling-water reactors)
The analysis of the preceding accidents showed that radiation doses to the maximally exposed member of the public at the site boundary from accidental airborne radioactivity releases during decommissioning operations were calculated to be quite low, with total body doses for a 50-year dose commitment ofless than 0.044 millirem (0.00044 millisievert) to the total body of the maximally exposed individual.
I 9.3 GENERAL l
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.
Draft NUREG-1628 42 April 1998
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 is dependent on when the reactor was last critical as well as on site-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 system; to warn the public. Onsite emergency plans will bc required for both the spent fuel pool and the ISFSIs, 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.
10 FINANCES 10.1- How much does it cost to decommission a nuclear power plant?
The total cost of decommissioning is dependent on the sequence and timing of the various stages of the program. The minimum amounts that are required for reasonable assurance of funds for decommissioning are $105 million for pressurized-water reactors and $135 million for boiling-water reactors. These costs are in 1986 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.
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 1130-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 (ISFSI) 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).
April 1998 43 Draft NUREG-1628
1 The estimated cost for decommissioning Maine Yankee, an 830-megawatt-elecmc pressurized- I 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 iccommissioning 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-3 water react is $306.4 million in 1995 dollars.
10.2 Who makes the estimates of the decommissioning costs?
The licensee makes the estimates of the decommissioning costs, or hires a contractor who has extensive experience in making these estimates. The 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 assure 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 3105 million for pressurized-water reactors and $13.5 million for boiling-water reactors in 1986 dollars). 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).
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 decommissioning 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 en actual costs at similar facilities that have ur brgone 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 c,f adjusting the cost estimate and funding throughout the duration of the storage.
Draft NUREG-1628 44 April 1998
. A-
- k' -
L The regulations also require a site-specific cost estimate, within 2 years following permanent
~ccssation of operations, if one has not already been submitted.
- Finally, at the time that the license termination plan is submitted (at least 2 yec.rs before the date when the license terminates), an updated site-specific estimate of any remaining decommissioning costs is required.~
10.4 - lIf the first estimate of decommissioning costs is made at the time that the facility is licensed, are there methods for adjusting for inflation?
c 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 wsste burial cost escalation factor is taken from an NRC report, " Report on Waste Burial Charges."
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 meth,ds:
Prepayment.' In this case, at the start of operations, the licensee deposits into an account enough funds to pay the dec' commissioning 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.
Externalsinkingfund.1 An extemal sinking fund 'is a fund ,stablished 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 y l 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. j Surety method, irisurance, or other guarantee method. A surety method may be in the form oia-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 unless,90 days or more preceding the renewal date, the issuer notifies the
- Commission, the beneficiary, and the licensee ofits intent to not renew. The surety or nisurance 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 :
L costs, and the trustee and trust must be accepiable to the Conimission. The surety method or
-insurance must remain in effect until the Commission has terminated the license.
April 1998 c .45 Draft NUREG-1628
~
10.6 Do the financial assurance regulations apply for Federal Government licensees?
Federal Government licensees, such as the Tennessee Villey Authority, are only required to have a statement of intent containing a cost estimate for decommissioning, and indicating that funds for decommissioning will be obtained when necessary.
10.7 Is there any way to ensure that the licensee does not just spend all of the money in the first few years of decommissbning and have nothing left to complete thejob?
The NRC nas placed regulations regarding the amount of money that can be used from the decommissioning 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 regula.iuns fer 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 ior 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.
10.8 What would happen if the cost of decommissioning exceeds the annount of money in the trust fund?
The various cost estimates (at the time oflicensing,5 years before anticipated shutdown, with the PSDAR submittal,2 years following shutdown, and 2 years preceding the anticipated termination of the license) are a method of reevaluating 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 I decommissioning process. If there is insufficient money in the trust fund, typically a licensee will secure a line of credit and borrow the funds to complete the decommissioning of the facility.
1 10.9 What would happen if the plant has an accident and there is not enough money in i 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. j Currently, $1.06 billion per operating unit is required for such insurance coverage.
10.10 Who pays for decommissioning?
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. !
l Draft NUREG-1628 46 April 1998 l
t . . . . _ _ _ _ _ - . _ - . .
- - - - - o
l 10.11 What contingency plans are in place to assure that decommissioning and long-term radioactive material storege 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 arejeopardized because ofinability 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 (e.g., 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.
10.12 "Ihat will happen if deregulation becomes a reality? How will deregulation affect 2nticipated revenue and the ability 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 regu'.ation toward greater competition. This policy statement was issued on August 19,1997, and published in the Federal Register. The policy statement addresses NRC concerns about the adequacy of decommissioning funds. The statement indicates that the NRC believes that its current regusatory 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 fur.:tional 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 funding assurance for the full estimated cost of decommissioning, either through full up-front funding or by some allowable guarantee or surety mechanism.
l I
I 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, i
l April 1998 47 Draft NUREG-1628
_ - _ _ - _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ - - _ _ _ _ _ _ . a
x -
J l
11 PUBLICINVOLVEMENT .
- 11.le :PUBLIC MEETINGS:
11.1.1 LWhat meetings are planned to keep the public informed?
< Two p0blic meetings are required during the decommissioning process. The first occurs before J
major decommissioning activities begin, when the post-shutdown decommissioning activities T, report (PSDAR) is submitted.' The second takes place when the licensee believes the project is
- close to completion.' At that time, a license termination plan, which describes how the site will be returned to a condition that makes radiological controls ~ no longer necessary, must be ,
f submitted by the licensee. In both cues; 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. !
1 i
Although not required by the regulations, the NRC will likely hoki an initial public informational -
~
meeting shortly after the licensee submits the certification of permanent cessation of operations. q At the meeting, NRC presents its process for regulating decommissioning, the licensee presents 1
, 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.'
11.1.2 Where will the meetings be held? R
' Meetings'are held in the vicmity of the facility. Often they are held at a local h'otel, county ~ _i courthouse, school, or library. - l 11.1.3 How will we be notified about the meetings?
~ NRC issues a news release and publishes a notice of the date, time, and location of the pubhc-
~
~o ; meeting in the Federal Register and in the local papers."The notice is also sent to the local public
, document room.
~ 11.1.4 If I cannot attend a meeting, how do I fmd out what was said? ,
A written transcript .of the meeting is prepared.fA signup sheet is made available for members 'of 1 o ;the public to leave their addresses if they wish to receive a_ copy of the transcript. An individual 1 ,
who is unable to attend the meeting may contact the NRC project manager whose name, address, and telephone number are listed in the FederalRegister notice and other published notices.
, In'dividuals may also call the NRC Office of Public Affairs,1 301-415-8200, or the NRC project manager for the facility being decommissioned.
2 11.1.5 May I make comments at the meeting? -
]
1Yes.- A~~ portion of the meeting will be devoted to questions from the public. Additionally. a sign- 'l up sheetLwill be available at the stan of the meeting for individuals requesting time for making ' !
Dhaft NUREG-1628 48 - April 1998 - :I
L comments or for reading prepared statements. Questions and short comments will also be -
!L accepted from the floor.
b l f 11.21 COMMENT PERIOD l
11.2.1 When does the comment period for the decommissioning process start? -
There are two separate comment periods. The first occurs following the licensee's subnuttal 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 when the licensee believes the project is close to completion. At that time, a license termination plan, which describes the remaining activities q necessary to terminate the NRC license, must be submitted by the licensee. Again, the NRC publishes a notice of the receipt of the license termination p!an 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.
11.2.2 - How do I make comments on the decommissioning process?
1 l
Comments and questions may be submitted in writing to the NRC project manager for the i facility. Comments and questions can also be addressed at the public meeting following receipt of the PSDAR. 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 NRC project L manager will be published in the Federal Register along with the notice of the receipt of the PSDAR and the schedule of tne public meeting.
11.2.3 Where should I send my comments?
NRC publ; cations and notices identify the correct address and person to write to when submitting comments. .
1 11.3 HEARINGS 11.3.1 Are hearings held on the decommissioning process?
There is no opportunity for a hearing on the contents of the PSDAR. If the NRC finds the license termination plan acceptable, it approves the plan by license amendment, which allows the April 1998 49 Draft NUREG-1628
- _ _ =
opportunity to request a hearing. This provides an additional opportunity for public participation.
11.4 GENERAL INFORMATION 11.4.1 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 that the public is interested in (http://NRC. gov). A comprehensive listing ofinformation sources can be found in an NRC publication. NUREG/BR-0010, Rev. 2, Citizen's Guide to U.S. Nuclear Regulatory Commission Information. Copies can be ordered from the Government Printing Office at the following address:
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 on line 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 The NRC also maintains a collection of materials about individual power plants in a local public docoment room located in the vicinity of each plae site, usually in a library. This facility, usually a local library, is available for public access.
12 GETTING ADDITIONALINFORMATION 12.1 What is a local public document room and how can I rmd one?
The NRC, in cooperation with several types of libraries, established a local public document room (LPDR) in the vicinity of each civilian nuclear power reactor site. These LPDRs maintain microfiche collections of all NRC publicly available documents issued since January 1,1981.
Documents issued before 1981 are not on microfiche but are maintained as paper copies and pertain only to the local facility. Documents in these collections include hearing transcripts, safety evaluation reports, environmental impact statements, emergency plans, inspection reports, licensee event reports, and generic communications.
The public is inviWd to read documents in the LPDR and to utilize available copying equipment for a fee. The library reference staffs and the LPDR staff in Rockville, Maryland are available to assist with the location of documents in the collections.
Draft NUREG-1628 50 April 1998
A list of LPDR libraries will be sent upon request by calling 1-800-638-8081 (numbers for headng-impaired persons (TDD numbers) are 202-634-3333 or 1-800-635-4512) or by writing to FOI/LPDR Branch U.S. Nuclear Regulatory Commission Mail Stop: T-6D8 Washington, DC 20555-0001 12.2 Does the NRC have a website? What kind of information can I obtain from it?
The NRC maintains a website at http://www.NRC. gov The website includes a generalinformation about the NRC
. information on nuclear reactors (including the nuclear plant watchlist, the Systematic Assessment of Licensee Performance (SALP) reports, plant information books, and daily reports, including plant status reports, daily events reports and headquarter's daily reports) a information on radioactive waste disposal a information on current rulemaking processes a news releases
. information on public involvement and public meetings, including sources for additional information .
. copies of the post-shutdown decommissioning activities reports (PSDARs) for plants that i have permanently ceased operations -
1 12.3 Does the NRC maintain an electron!c bulletin board system? !
The NRC has entered into an interagency agreement with the National Technical Information Service (NTIS), a central repository for scientific, technical, and engineering infoarction.
NTIS provides access to more than 100 computer bulletin board systems through a facility l- known as FedWorld. FedWorld can be accessed electronically by using a computer and modem
! to dial 703-321-3339 (up to 14,400 bits per second, eight data bits, no parity, one stop bit, full duplex, American National Standards Institute terminal emulation). Internet users can T:Inet to l . FedWorld at fedworld. gov or file transfer protocol (FTP) to ftp.fedworld. gov. Users who wish to
. only access NRC information can dial toll-free to 1-800-303-9672 with the same communication parameters at speeds up to 28,800 bits per second. The following information is available:
April 1998 51~ Draft NUREG-1628
. daily press releases ,
l
- . public petitions
- enforcement program information a generic communications a public document room system information
- NRC rulemakings 12.4' 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 ate also available at the NRC's local public document rooms around the country and at the NRC's Public Document Room in Washington, D.C. You may also search the Federal Register database for 1995 and 1996 on the internet.
12.5 How can I get a copy of the Code of Federal Regulations?
Copies of.the Code ofFederal Regulations are found in the local public document reading rooms and often at local libraries in the reference section. They are available for purchase from the Govemment Printing Office by credit card at 202-512-1800, Monday through Friday,8 a.m. to 4 p.m. EST (fax 202-512-2233,24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 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.
12.6 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 NPC Library (301-415-5609).
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 NRC policies, programs and activities:
- Region I(Philadelphia) 610-337-5000
- Region II(Atlanta) 404-562-4400
- Region III(Chicago) 630-829-9500
- Region IV (Dallas) 817-860-8100
- Walnut Creek Field Office (San Francisco) 510-975-0200 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, Draft NUREG 1628.. 52 April 1998
c..
J; b .
n s' , . I C REFERENCES
- e i ,
(U.S. Nuclear Regulatory Commission, Washington, D.C.
FederalResister 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)
~
, ' Defm* itions -Decommissioning," Federal Register, Vol. 62, p. 39089, July 21,1997 (10 CFR 4
-30.4) 1" General requirements for decommissioning nuclear facilities," Federal Register, Vol. 53, p.
-24018, June 27,'1988 (10 CFR Pans 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, l'997 '
-(10 CFR Parts 20, 30,40,~ 50,51,70, and 72)
' " Timeliness in decommissioning of m' aterials facilities," Federal Register, Vol. 59, p. 36026, L. July 15,1994 (10 CFR Pans 2,30,40,70, and 72) .
! Generic letters
, iGL 79-19, 'APackaging of Low-Level Radioactive Waste for Transport and Burial"
' GL 79-30,L" Packaging, Transport and Burial of Low-level Radioactive Waste" 1
Information Notices -
IN 80-24,'." Low-12 vel 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" LApril 1998. .
~53 Draft NUREG-1628 4 1 i l
- .. L _ -_---_w f = f: -- ----__-- --__________----__--____-___________--___--_-__A
. References : j
- IN 87-3, " Segregation'of Hazardous and Low-Level Radioactive Wastes" l
IN'88 78,L" Chemical Reactions With Radioactive Waste Solidification Agents" IN 88-16, " Identifying Waste Generators in Shipments of Low-Ixvel Waste to Land Disposal Facilities"
'IN 89-13, t' Alternate Waste Management Procedures in Case of Denial of Access to Low-Level; Waste Disposal Sites" I
!IN 89-27,'" Limitations on the Use of Waste Forms and High Integrity Containers for the q
- Disposal of Low-level Radioactive Waste" l
' IN 90-9, " Extended Interim Storage ofliw'-Level Radioactive Waste by Fuel Cycle and '
Materials Licensees"
. IN 90-31, " Update on Waste Fo'rm and High Integrity Container. Review Status, Identification of Problems With Cement Solidification" Li . - 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 OrdinaryTrash)" i 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, "Recoidkeeping, Decommissioning Notifications for Disposals of Radioactive Waste l by Land Burial Authorized Under Former 10 CFR 20.304,20.302 and Current 20.2002" D
- NUREG-Series Reports h6 -l NUREG-0586', " Final Generic Environmental Impact Statement on Decommissioning of Nuclear -
Facilities," August 1988 L t' NUREG-0945 (Draft), " Draft Environmental hepact Statement for 10 CFR Part 61," September b"
~
1981- l l NUREG-0945, " Final Environmental Impact Statement for 10 CFR Part 61," November 1982.
NUREG-1101,"Onsite Disposal of Radioactive Wasto" 1986
}
L
[ , x Draft NUREG-1628 54~ . April 1998
.h _ __ __ . '
- l References 1
- NUREG-1307," Report on Waste Burial Charges," Rev. 4, June 1994 l NUREG-1337, " Standard Review Plan for the Review of Financial Assurance Mechanisms for l L ' 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 L 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, 'l Working Draft Regulatory Guide on Release Criteria for Decommissioning:
- NRC Staffs Draft for Comment August 1994 NUREG-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-1.evel Radioactive Waste Disposal Facilities," April 1997 NUREG/BR-0111, " 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 1 Water Reactor Power Station," June 1978 (Addendum 1, July 1979; ~ Addendum 2, July 1983; Addendum 3, September 1984; Addendum 4, July 1988)
April 1998 : 55 Draft NUREG-1628 =
References -
NUREG/CR-%72, " Technology, Safety and Costs of Decommissioning a Reference Boiling
. Water Reactor Power Station," June 1980(Addendum 1, July 1983; dAd endum 2S, eptemb er 1984; Addendum 3, July.1988; Addendum 4, December 1990) i 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 - l 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 i
~ 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 ,
i 4 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 on Site 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 F;ior to Release for Unrestricted-Use or Termination of Byproduct, Sogarce and Special Nuclear Material Licensees," August 1987 Draft NUREG-1628 - 56 April 1998
- (
l
-1 References
!: l, PG-8-08, Policy and Guidance Directive " Scenarios for Assessing Potential Doses Associated 1With Residual Radioactivity," May 1994
- e 1 SECY-94-145 " Increase of Tritium and Iron-55 Unrestricted Use Limits for Surface
- Contarhination at Shoreham and Fort St. Vrain." May 1994 i
- sr.
l~
i i,
V : r, .
g u.
l::
- (
i ii !,I !'
i
.~ . . i April 1998'- -
- 57- Draft NUREG-1628 . 1
- m ,
C_ _Jb__.... _ .i .i'-- . .4 .e -
NRC FORM 33s u.s. NUCLEAR REGULATORY coMMisslON 1. REPORT NUMBER 9 49) (Assigned by NRC, Add Yol., supp., Rev.,
NRCM 11o2. and Addendum Numbers,it anyj 32ot aro2 BIBLIOGRAPHIC DATA SHEET (See tastruchons on the reverse)
- 2. TITLE AND SUBTITLE NUREG-1628 Stiff Responses to Frequently Asked Ouestions Concerning Decommissioning of Nuclear Power Plants 3. DATE REPORT PUBLISHED Dr'ft Report for Comment MONTH
] YEAR April 1998
- 5. AUTHOR (S) 6. TYPE OF REPORT J.L Minns, M.T. Masnik Draft
- 7. PERIOD COVERED (Inciustve Dates)
- 8. PERFORMING ORGANIZATION - GAME AND ADDRESS (itNRC, prowde Dwston, ot#ce or Rega, US Nucipar Regulatory Comtrussot, and madng acMress; # contractor, prowde name end madne odoress }
Division of Reactor Program Management Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission W:shington., D.C. 20555-0001 0, SPONSORING ORGANIZATION NAME AND ADDRESS (ttNRc, type "Same es above", # contractor, prowde NRc Demon, onice or Region, U.S NucharRegulatory comirnsmor, and enene address }
Same as above
- 10. SUPPLEMENTARY NOTES J.L. Minns, Project Manager
- 11. ABSTRACT poo words or bes; Through a question-and-answer format, this document presents information to the public on decommissionitig. The questions were taken from a variety of sources over the past severa! years, including written inquiries to the NRC and questions asked at public meetings and dunng informal discussions with the NRC staff. In responding to the questions, the NRC staff attempted tKnswer in a clear and non-technical form, one that an individual with no or little technical training could understand.
Questions are posed on the following categories: the decommissioning process and decommissioned sites; licensing; regulations; the inspection program; spent fuel, spent fuel pools, and spent fuel storage; radioactive low-levri waste; tr;nsportation, license termination and the ultimate disposition of the facility; hszards; finances; and public drivolvement. ,
. 1 LQ docen4snt is being issued for public comment. As a result of this comment, peer review, and discussions, the final I document may be modified from this draft.
I i
- 12. KEY WORDS/DEScRIPTORs (List mords or puisses mat ws essest researchers m ecsong me report ; 3 AvAitAsiuty 6iATEMENT
' Decommissioning nuclear power reactors; Decommissioning process; Decommissioning sites; unlimited
- i. Decommissioning regulations; Decommissioning licensing; Decommissioning inspection programs; 14 secon.TvctAssiFicATION
- 8 pent fuel; Spent fuel pools; independent spent fuel storage installation (ISFSI); Transportation; rrtus eagej Finances; Public involvement; License termir.ation; Disposition of site unclassified 5s Report) unclassified II.EUMBER Of PAGES to. price
- i% f oMM N6 4249) T,ve tor , nao electroi.ically produced by Elae Federal Forms. inc.
4-L
! Prinhal ..
on recy M
- \
i Federal Recycling Program l
i _
0 gg I
3ll p4 g
P l
2 it
?!-
!s:
3 15.
- g .o 5
e
~
B
- g .
fgga ll! )
- ess; j l 111- e '
4!i5- 1
-4 A
'