Proceedings
of the
St. Michaels Workshop
on
Residual Radioactivity and
Recycling Criteria
September 27-28, 1989
St. Michaels, Maryland
Editors:
Anthony B. Wdlbarst
Hiromi Terada
Allan C.B. Richardson
Japan Atomic Energy Research Institute
and the
United States Environmental Protection Agency
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To order additional copies, contact:
Director, Criteria and Standards Division
Office of Radiation Programs (ANR-460)
U.S. Environmental Protection Agency
Washington, D.C. 20460
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PREFACE
Over the next several decades, many thousands of radioactively contaminated sites in the
United States, Japan, and other countries will become candidates for cleanup and
decommissioning.
Some sites are already in the process of cleanup - in the U.S, for example, under DOE
and Superfund programs.. For other facilities, such as the 500 operational commercial nuclear
power plants throughout the world, large-scale decommissioning and cleanup will become
increasingly common in the coming decade.
But suitable public health and environmental protection criteria and standards for such
decontamination and decommissioning programs are lacking or incomplete in virtually all
countries. In their absence, the responsible agencies and industries have had to resort to ad hoc
criteria. These often are inadequate and inconsistent, generate confusion, and lack public
confidence.
Health risks to some individuals could be significant. And since some radionuclides have
long half-lives and high environmental mobility, these risks could be long-lasting and wide-
spread. Because decontamination is anticipated to cost many tens of billions of dollars, this is
also an issue of substantial economic importance. The lack of established cleanup levels
exacerbates this aspect of the problem.
Thus, health protection criteria are needed so that sites can be cleaned up and made
available for other uses, either with or without restrictions based upon residual radioactivity.
On September 27 and 28, 1989, the Office of Radiation Programs of the Environmental
Protection Agency and the Japan Atomic Energy Research Institute together sponsored a
workshop on Residual Radioactivity and Recycling Criteria in St. Michaels, Maryland. Thirty-one
government and private sector radiological health experts from the United States participated,
and eleven from Japan.
The workshop provided a forum for an exchange of ideas and information among
individuals who deal directly with the issue of cleanup standards. Topics for discussion fell into
five general categories: Extent of the Clean-up Problem; Impacts of Clean-up Technologies and
Economics on Criteria; Health Effects; Desirable Characteristics of Criteria; and Recycling of
Materials and Equipment. There was a Panel discussion at the conclusion of the meeting.
Workshop participants made clear their feelings that the presented papers and the debate
were highly productive. We are therefore publishing this volume of Proceedings in the hope that
it will be of use as a resource document in the further development of criteria for the cleanup of
radioactively contaminated sites.
Masao Oshino, Director
Department of Health Physics
Japan Atomic Energy Research Institute
Richard J. Guimond, Director
Office of Radiation Programs
US Environmental Protection Agency
111
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PROCEEDINGS OF A WORKSHOP ON
RESIDUAL RADIOACTIVITY AND RECYCLING CRITERIA
CONTENTS
Page
I. Workshop Agenda 1
II. List of Participants ^ 4
III. Workshop Introduction and Overview, Richard J. Guimond, U.S. EPA 11
IV. Subrnitted papers
Session!: Extent of the Cleanup Problem 15
The Department of Energy's Five Year Plan, Roger P. Whitfield, U.S. DOE 17
Decommissioning Waste Characteristics, Timothy C. Johnson, U.S. NRC 28
Site Inventory of Residual Radioactivity in Japan, Shohei Kato, Fuyuhiko 44
Ishikawa, and Hideaki Yamamoto, JAERI
Session II: Impacts of Cleanup Technologies and Economics on Criteria 57
' Limitations of Cleanup Technologies, Thomas S. LaGuardia, TLG Engineering 59
Decontamination Technology for Decommissioning of Nuclear Facilities, 64
Hideo Yasunaka, Tamotsu Kozaki, and Takeo Gorai, JAERI
Low-Level Radioactivity Measurement Methods for Reuse or Recycling, 82
Iwao Manabe, Yukio Iwata, and Masao Oshino, JAERI
Disposal Capacity and Projected Waste Volumes Within the Low-Level Radioactive 90
Waste Compacts, Steven R. Adams, U.S. Ecology
Bench Scale Studies and Pilot Scale Design of a Modified Reduction-Chemical , 122
Extraction System for Radiation Contaminated Soils, Robert S. Dyer, U.S. EPA
Residual Radioactivity Cost Impact Evaluation, Richard P. Allen, Battelle Pacific 132
Northwest Laboratory ' *
i
Session III: Health Effects 147
Experience in Decontamination and Reuse of the Large-Scale Radiochemical 149
Laboratory and the Research Reactor at JAERI, Hideaki Yamamoto,
Kozo Matsushita, and Hozumu Yamamoto, JAERI
. Applied Exposure Modeling for Residual Radioactivity and Release Criteria, 158
Donald W. Lee, Oak Ridge National Lab
DOE Guidelines and Modeling in Determination of Soil Cleanup Standards, 166
Andrew Wallo, III, U.S. DOE
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Page
Session IV: Desirable Characteristics for Criteria 171
International Similarities and Differences to Regulating Nonradiation Hazards, 173
Rob Coppock, National Academy of Sciences , -v
What Are the Basic Requirements that Cleanup, Standards Should Satisfy?, , 179
Allan C.B. Richardson, U.S. EPA ^ ;.,-,„, ,,,.,: . ,
What Should Cleanup Standards Do?, Vern C. Rogers, Rogers & Associates *'*"' 186
Engineering Corporation - - - ; , , » . ,., ,. <•'.-. , -.
Current Status of Residual Radioactivity Criteria in Japan, ,Hideaki Yarnampto 193
and Masao'Qshino, JAERI " , , .- -
Establishment of Criteria for the Unconditional Release of the Shippingp'ortAtomic 197
Power Station Site, Lynn R. Wallis, G.E. Nuclear Energy «
NRC Residual Contamination Criteria, Timothy C. Johnson, U.S. NRC ' ' 212
Status and Implementation of'the NRC Policy on Exemptions 'from Regulatory Control, 216
Donald A. Cool, U.S. NRC
Surface Contamination Criteria for Free Release, Steven R. Adams, U.S. Ecology 220
EPA's Proposed Environmental Standards for Low-Level Radioactive Waste 237
Disposal and Criteria for Below Regulatory Concern. William F. Holcomb and
James M. Gruhlke, U.S. EPA ' ...
• • "• r' •• - * 'E .•••'••>.' ,•".,; .
EPRI Discussion Paper on BRC and De Minimi's Cpnce'pts, Jene Vange, ' 246
Vance & Associate •.•.,•.• v ^ , ,r,
Criteria for Release of Decommissioned Nuclear Facilities for Unrestricted Use, t
Joseph W. Ray, Battelle Memorial Institute ' '
Session V: Recycling of Materials and Equipment
A Research Program on the Recycling of Decommissioning Materials at JAERI, 257
Mjtsugu Tanaka and Hisashi Nakamura, JAERI ,,,,-
Effects of Residual Radioactivity in Recycled Materials on Scientific-and v 266
Industrial Equipment, Shohei Kato, Hideaki Yamarnoto, and Shigeru
Kumazawa, JAERI , ,.4
Development of International Exemption Principles for Recycle and Reuse, 281
William E. Kennedy, Jr., Battelle Pacific Northwest Laboratory »., ......
V. Summary and Panel Discussion •...., _,• , ;, . 290
Moderator: Anthony B. Wolbarst, U.S. EPA y.-..-.•. t- «
Panel: Donald A. Cool, U.S. NRC, Masao Oshino, JAERI, " "
Allan C.B. Richardson, U.S. EPA, and Andrew Wallo, III, U.S. DOE .,
VI
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RESIDUAL RADIOACTIVITY AND RECYCLING CRITERIA WORKSHOP
AGENDA
WEDNESDAY, SEPTEMBER 27, 1989
9:00 - 9:15 Welcome ~ J. William Gunter, U.S. EPA and Masao Oshino, JAERI :
9:15 - 9:40 Introduction and Overview - Richard J. Guimond, U.S. EPA
9:40-10:45 Extent of the Cleanup Problem - Chairman: J. William Gunter, U.S. EPA
• The Department of Energy's Five Year Plan, Roger P. Whitfield, U.S. DOE
• Decommissioning Waste Characteristics, Timothy C. Johnson, U.S. NRG
• Site Inventory of Residual Radioactivity in Japan. Presented by Shohei Kato, JAERI
Authored by Shohei Kato, Fuyuhiko Ishikawa, and Hideaki Yamamoto, JAERI
10:45-11:00 Break
11:00-12:30 Impacts of Clean-up Technologies and Economics on Criteria
Chairman: Andrevv Wallo, III, U.S. DOE
Limitations of Clean-up Technologies and Survey Instrumentation
• Limitations of Cleanup Technologies, Thomas S. LaGuardia, TLG Engineering
• Decontamination Technology for Decommissioning of Nuclear Facilities. Presented
by Hided Yasunaka, JAERI. Authored by Hideo Yasunaka, Tarnotsu Kozaki, and
Takeo Gorai, JAERI
• Low Level Radioactivity Measurement Methods for Reuse or Recycling. Presented
by Masao Oshino, JAERI. Authored by Iwao Manabe, Yukio Iwata, and Masao
Oshino, JAERI
12:30-1:30 Lunch"
1:30 - 2:30 Impacts of Clean-up Technologies and Economics on Criteria (continued)
Disposal Capacity, and Volume Reduction Techniques
• Processing of Decommissioning Waste, and Criteria for Release, H.W. Arrowsmith,
Westinghouse Electric Corporation
• Disposal Capacity and Projected Waste Volumes Within the Low-Level Radioactive
Waste Compacts, Steven R. Adams, U.S. Ecology
• Bench Scale Studies and Pilot Scale Design of a Modified Reduction-Chemical
Extraction System for Radiation Contaminated Soils, Robert S. Dyer, U.S. EPA
Economic Issues
•-•'. • Residual Radioactivity Cost Impact Evaluation, Richard P. Allan, Battelle Pacific
Northwest Laboratory
2:30 - 2:45
Break
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AGENDA (continued)
2:45-4:30 Health Effects - Chairman: Donald A. Cool, U.S. NRC
Risks to Individuals and in Populations . . „
• The Science and/or Art of Estimating Radiation Risks at Environmental Levels of
Exposure, William H. Ellett, National Academy of Sciences
• Experience in Decontamination and Reuse of the Large Scale Radiochemical
Laboratory and the Research Reactor at JAERI. Presented Hideaki Yamamoto,
JAERI. Authored by Hideaki Yamamoto, Kozo Matsushita, and Hozumu
Yamamoto, JAERI
Exposure Models. Results
• Applied Exposure Modeling for Residual Radioactivity and Release Criteria, Donald
W. Lee, Oakridge National Lab
• DOE Guidelines and Modeling in Determination of Soil Cleanup Standards, Andrew
Wallo, III, U.S. DOE
THURSDAY, SEPTEMBER 28, 1989
9:00 -10:30 Desirable Characteristics for Criteria - Chairman: Allan C.B. Richardson, U.S. EPA
Fundamental Principles/Possible Forms for Clean-up Criteria
• International Similarities and Differences in Approaches to Regulating Non-Radiation
Hazards, Rob Coppock, National Academy of Sciences
• What Are the Basic Requirements that Cleanup Standards Should Satisfy?, A|lan C.B.
Richardson, U.S. EPA
• What Should Cleanup Standards Do?, Vern C. Rogers, Rogers & Associates
Engineering Corporation , .
Criteria Currently in Use
• Current Status of Residual Radioactivity Criteria in Japan. Presented by Hideaki
Yamamoto, JAERI. Authored by Hideaki Yamamoto, and Masao Oshino, JAERI
10:30-10:45 Break
10:45-12:30 Desirable Characteristics for Criteria (continued) i
Criteria Currently in Use (continued)
• Establishment of Criteria for the Unconditional Release of the Shippingport Atomic
Power Station Site, Lynn R. Wallis, G.E. Nuclear Energy
• The Need for New Criteria for Cleanup of Land and Facilities Contaminated with
Residual Radioactivity, Stanley W. Neuder, Battelle Pacific Northwest Laboratory
• NRC Residual Contamination Criteria, Timothy C. Johnson, U.S. NRC
• Status and Implementation of the NRC's Policy on Exemptions from Regulatory
Control, Donald A. Cool, U.S. NRC
• Surface Contamination Criteria for Free Release, Steven R. Adams, U.S. Ecology
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AGENDA (continued)
12:30-1:30 Lunch :
1:30-2:00 Desirable Characteristics for Criteria (continued)
'•• Waste Management/Disposal Criteria/BRC ,
• EPA's Proposed Environmental Standards for Low-Level Radioactive Waste Disposal
and Criteria for Below Regulatory Concern. Presented by William F. Holcomb,
:- U.S. EPA. Authored by William F. Holcomb, and James M. Gruhlke, U.S. EPA
• EPRI Discussion Paper on BRC and De Minimis Concepts^ Jene Vance, Vance
. & Associates
• Residual Soil Contamination Criteria and BRC, Joseph W. Ray, Battelle
Memorial Institute
2:00-2:45 Recycling of Materials and Equipment - Chairman: Masao Oshino, JAERI
Inventory and Types of Materials and Contamination
• A Research Program on the Recycling of Decommissioning Materials at JAERI
, Presented by Hisashi Nakamura, JAERI. Authored by Mitsugu Tanaka and
~ Hisashi Nakamura, JAERI N
• Inventory and Types of Contamination in Recycled Materials, Mary E. Clark, Florida
Department of Health and Rehabilitative Services
2:45-3:00 Break
3IOO - 3:30 Recycling of Materials and Equipment (continued)
-•>••• Affects on Film. LSI. Nuclear Counting Industries !
• Effects of Residual Radioactivity in Recycled Materials on Scientific and Industrial
Equipment. Presented Shohei Kato, JAERI. Authored by Shohei Kato, Hideaki
Yamamoto, and Shigeru Kumazawa, JAERI
"•.••; • . --.. "• -,•'/•. :.--.• '• •. . ' v. • ': • • .'.',"•-• •'•'•• '- '-
f" Possible Forms for Recycling Criteria -
• Development of International Exemption Principles for Recycle and Reuse, William E.
Kennedy, Jr., Battelle Pacific Northwest Laboratory
•, Using NEPA to Decide Recycling Issues When There are no Explicit Standards,
Stanley Lichtman, U.S. DOE
3:30 - 4:30 Summary and Panel Discussion -- Donald A. Cool, U.S. NRC, Masao Oshino, JAERI,
. - .. Allan C.B. Richardson, U.S. EPA, and Andrew Wallo, III, U.S. DOE
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RESIDUAL RADIOACTIVITY AND RECYCLING
CRITERIA WORKSHOP
PARTICIPANTS •'.-•'•
Steven R. Adams
Chief, Radiological Control and Safety
Officer and Manager of Health Physics
U.S. Ecology
9200 Shelbyville Road, Suite 300
Louisville, KY 40207-7246
(502) 426-7160
Richard P\ Allen, Ph.D.
Battelle Pacific Northwest Laboratory
P.O. Box 999
Richland, WA 99352
(509) 376-9272
H.W. "Bud" Arrowsmith
President, Scientific Ecology Group
Westinghouse Electric Corporation
P.O. Box 2530
Oak Ridge, TN 37830
(615) 481-0222
R. Thomas Bell
Radiation Policy Division (RARP)
Defense Nuclear Agency
U.S. Department of Defense
Washington, D.C. 20305-1000
(202) 325-0459
Bruce Burnett
Office of Health Physics (HFZ-60)
Center for Devices and Radiological Health
5600 Fishers Lane
Rockville, MD 20857
(301) 443-2850
Mary Clark, Ph.D.
Office of Radiation Control
Department of Health and Rehabilitative Services
1317 Winewood Boulevard
Tallahassee, Florida 32399-0700
(904)487-1004
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Workshop Participants (con't)
Donald A. Cool, Ph.D.
Chief, Radiation Protection and Health Effects Branch
Office of Nuclear Regulatory Research
U.S. Nuclear Regulatory Commission (NL/S-007)
5650 Nicholson Lane
Rockville, MD 20852
(301) 492-3785
Rob Coppock, Ph.D.
National Academy of Sciences
2101 Constitution Avenue, N.W., NAS 356 G4
Washington, D.C. 20418
(202) 334-1637
Robert S. Dyer
Chief, Environmental Studies and Statistics Branch
Office of Radiation Programs (ANR-461)
U.S. Environmental Protection Agency
401 M Street, S.W.
Washington, D.C. 20460
(202) 475-9630
William H. Ellett, Ph.D.
Board on Radiation Effects Research CLS
National Academy of Sciences
2101 Constitution Avenue, N.W.
Washington, D.C. 20418
(202) 334-2743
Elliot C. Routes
Economics and Control Engineering Branch
Office of Radiation Programs (ANR-461)
U.S. EnvironmentarProtection Agency
401 M Street, S.W.
Washington, D.C. 20460
(202) 475-9644
Richard J. Guimond
Director
Office of Radiation Programs (ANR-458)
U.S. Environmental Protection Agency
401 M Street, S.W.
Washington, D.C. 20460
(202)475-9600 __
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Workshop Participants (con't)
J. William Gunter
Director
Criteria and Standards Division
Office of Radiation Programs (ANR-460)
U.S. Environmental Protection Agency
401 M Street, S.W.
Washington, D.C. 20460
(202) 475-9603
Seiichi Hitomi
Deputy General Manager
Department of Health Physics
Tokai Research Establishment
Japan Atomic Energy Research Institute
Tokai-mura, Naka-gun, Ibaraki-ken, Japan
William F. Holcomb
Acting Chief
Waste Management Standards Branch
Office of Radiation Programs (ANR-460)
U.S. Environmental Protection Agency
401 M Street, S.W.
Washington, D.C. 20460
(202) 475-9633
Osamu Hurukawa
Japan Radioisotope Association
Takizawa-mura, Iwate-ken, Japan
Mitsuo Ibuki
NGK-LOCK Inc.
3 Pickwick Plaza
Suite 401
Greenwich, CT 06830
Fuyuhiko Ishikawa
Chiyoda Co.
Tokai-mura, Naka-gun, Ibaraki-ken, Japan
Timothy C. Johnson
Division of Low Level Waste Management
U.S. Nuclear Regulatory Commission (5E4)
Washington, D.C. 20555
(301) 492-0558
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Workshop Participants (con't)
Shohei Kato
Senior Scientist
Department of Health Physics
Tokai Research Establishment
Japan Atomic Energy Research Institute
Tokai-mura, Naka-gun, Ibaraki-ken, Japan
Yutaka Kawakami
General Manager
Department of Health Physics
Tokai Research Establishment
Japan Atomic Energy Research Institute
Tokai-mura, Naka-gun, Ibaraki-ken, Japan
William E. Kennedy, Jr.
Battelle Pacific Northwest Laboratory
P.O. Box 999
Richland, WA 99352
(509)375-3849
Thomas S. LaGuardia
President
TLG Engineering, Inc.
148 New Milford Road East
Bridgewater, Connecticut 06752
(203)355-2300
Donald W. Lee, Ph.D.
Group Leader
Applied Physical Sciences Group
Oak Ridge National Lab
Building 2024, P.O. Box 2008
Oak Ridge, TN 37831-6045
(615) 574-5803
John Lehr
Director
Hazardous Waste and Remedial Actions Division
Office of Defense Waste and Transportation Management
Defense Programs (DP-124)
Department of Energy
Washington, D.C. 20545
(301) 353-3253
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Workshop Participants (con't)
Stanley Lichtman, Ph.D.
Group Leader, Waste Management Group
Office of NEPA Project Assistance
1000 Independence Avenue, S.W.
U.S. Department of Energy
Washington, D.C. 20585
(202)586-4610
Loring E. Mills
Vice President, Nuclear Activities
Edison Electric Institute
1111 19th Street, N.W.
Washington, D.C. 20036
(202) 778-6750
Hiroyuki Murakami
Chief, Department of Health Physics
Tokai Research Establishment
Japan Atomic Energy Research Institute
(Visitor Scientist in ORNL)
Hisashi Nakamura
Research Scientist, Department of JPDR
Tokai Research Establishment
Japan Atomic Energy Research Institute
Tokai-mura, Naka-gun, Ibaraki-ken, Japan
Stanley W. Neuder, Ph.D.
Battelle Pacific Northwest Laboratory
370 L'Enfant Promenade, S.W., Suite 900
Washington, D.C. 20024
(202) 646-5210 or 479-0500
Masao Oshino
Director
Department of Health Physics
Tokai Research Establishment
Japan Atomic Energy Research Institute
Tokai-mura, Naka-gun, Ibaraki-ken, Japan
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Workshop Participants (con't)
Joseph W. Ray, Ph.D.
Group Vice President and General Manager
Decontamination and Decommissioning Operations
Battelle Memorial Institute
505 King Avenue
Columbus, OH 43201-2693
(614) 424-5522
Allan C.B. Richardson
Chief, Guides and Criteria Branch (ANR-460)
Office of Radiation Programs
U.S. Environmental Protection Agency
401 M Street, S.W.
Washington, D.C. 20460
(202) 475-9620 .
Vern C. Rogers, Ph.D.
President
Rogers & Associates Engineering Corp.
Post Office Box 330
Salt Lake City, Utah 84110
(801)263-1600
John L. "Jack" Russell
Guides and Criteria Branch (ANR-460)
Office of Radiation Programs
U.S. Environmental Protection Agency
401 M Street, S.W.
Washington, D.C. 20460
(202)475-9620
Jene Vance
EPRI Contractor
Vance & Associates
P.O. Box 997
Ruidoso, NM 88345
(505) 336-4845
Lynn R. Wallis
Manager of Media and Environmental Information Programs
G.E. Nuclear Energy
175 Kurtner Avenue
San Jose, CA 95125
(408) 925-1149
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Workshop Participants (con't)
Andrew Wallo, III
U.S. Department of Energy
Office of Environmental Guidanpe and Compliance, EH231
1000 Independence Avenue
Washington, D.C. 20585 .
(202) 586-4996
Roger P. Whitfield ; .
Associate Director for Environmental Restoration
Office of Environmental Restoration and Waste Management, EM-40
U.S. Department of Energy ,
Washington, D.C. 20585
(202) 586-6331
Anthony B. Wolbarst, Ph.D.
Guides and Criteria Branch (ANR-460)
Office of Radiation Programs
U.S. Environmental Protection Agency
401 M Street, S.W.
Washington, D.C. 20460
(202) 475-9620
Hideaki Yamamoto
Research Scientist
Department of Health Physics
Tokai Research Establishment
Japan Atomic Energy Research Institute
Tokai-mura, Naka-gun, Ibaraki-ken, Japan
Hideo Yasunaka :
Head, Department of JPDR
Tokai Research Establishment
Japan Atomic Energy Research Institute
Tokai-mura, Naka-gun, Ibaraki-ken, Japan
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Introduction
Richard Guimqnd
Director of the Office of Radiation Programs
Environmental Protection Agency
INTRODUCTION
I am particularly pleased to be able to welcome you to this workshop on cleanup criteria
for radioactively contaminated sites, sponsored jointly by ORP and the Japan Atomic Energy
Research Institute (JAERI). We at ORP have undertaken several projects together with our
friends at JAERI, and we have found them to be colleagues of great value. I feel that this
workshop is an important part of our growing collaboration.
THE NEED FOR CLEANUP CRITERIA
In the United States, the issue of how to deal with radioactively contaminated facilities is
.one whose time has clearly come. The press is filled with reports of sites, some privately run and
others owned by the government, that are seriously contaminated and will have to be cleaned
up to protect the health and well-being of those who live nearby. The spector looms, moreover,
of one hundred commercial power plants, many of which will have reached the end of their useful
lives early in the 21st century, .and will have to be decommissioned. It is estimated that in the
U.S., there are perhaps a thousand significantly contaminated sites of various types that must
be cleaned up. And the Congress, the Federal agencies, and all aware citizens are concerned.
In the meantime, there has been little cleanup progress to date, One important reason
for this is that there are few applicable cleanup standards. There is nothing that says to the
owner or manager of a facility where radioactive materials are used, "If you clean your place up
to such-and-such a level, you can walk away from it and forget it."
In the absence of suitable public health and environmental protection criteria, cleanup
efforts have occurred ad hoc, with widely differing results. Standards that were written for the
decontamination of uranium and thorium mill tailing sites (40 CFR 192), for example, are being
used as cleanup criteria at Superfund and Department of Energy sites that are contaminated with
naturally occurring radioactive materials (where radium is usually the culprit). In some cases, this
has required stretching the standards, since they were originally designed for a much more
specific situation.
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As a consequence of the absence of criteria, cleanup has been postponed, in some
cases, in anticipation of standards that might be forthcoming. The cleanups that have been
undertaken have frequently been incomplete and inconsistent, have generated confusion, and
have lacked public trust.
There is thus a pressing need for clear and unequivocal criteria that will ensure uniform
and adequate protection of the public nation-wide. These standards should be consistent with
the requirements for the cleanup of other kinds of environmental hazards. And for economic and
ethical reasons, most if not all contaminated sites, large and small, will probably have to be made
clean enough for release for unrestricted public use. The public deserves nothing less, since we
cannot predict how long we can rely on the institutional controls that would be made necessary
by less than complete cleanup.
HOW CLEAN IS CLEAN ENOUGH?
At many sites, large amounts of radioactive contamination have become mixed into the
soil, sometimes to depths of several meters or more. It may have percolated down into the
aquifer, making water unsuitable for home use, for livestock, and for irrigation ofccrops. It may
have become imbedded in the walls and floors of buildings, from which it can not only produce
direct external irradiation, but also rub off onto hands and clothing.
After the sites are cleaned up and released for unrestricted use, people will then use the
land and groundwater to raise crops and animals, and will raise families and work within the •
buildings that remain. That raises the critically important central question: How clean is clean
enough for such unrestricted public use?
For a carcinogen like ionizing radiation, we have no threshold level below which we can
feel totally secure. We don't even have any legislative guidance on what an acceptable level of
residual radioactivity would be.
Should standards restrict individual lifetime risk of a fatal, radiation-induced health effect
to 10"4, or to 10"6, or to some other level? How important is the ALARA principle in establishing
cleanup criteria, and what levels of radiation should be considered Below Regulatory Concern?
There are no readily available answers to these fundamental questions.
The problem of cleanup criteria is complicated even further by the special difficulties of
disposing of radioactive materials, and by the particular distaste that people harbor for things
radioactive.
The disposal problems are unique. Unlike many other hazardous substances, radioactive
materials will not decompose or otherwise disappear by themselves. Nor can you simply
incinerate the soil and then put it back in place. Radionuclides have half-lives of thousands of
years or more, in some cases, and come in a wide variety of chemical forms, often mixed with
other hazardous materials. The sheer volume of low level waste projected to come from
decommissioning operations is staggering. It has been estimated that there would be one million
cubic meters of LLW left over from the decommissioning of our 100 commercial power plants.
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That's enough to build a wall one meter thick and three meters high between Washington, D.C.
and New York. Perhaps that's what we should do with it.
The sums of money involved are also quite remarkable. Frequently one hears figures like
one hundred or several hundred billions of dollars. By contrast, EPA's Superfund budget for the
1989 fiscal year has been $1.6 billion. And, of course, there will be many different organizations
and individuals that will be financially affected by the details of any standards and regulations that
are set, and ready and able to apply political pressure in support of their needs.
A few areas, such as the Nevada nuclear weapons testing range, are riot only highly
contaminated with wide-spread radioactivity, but also extremely remote and" inaccessible. The
costs of a complete cleanup would be very high, and the health benefits marginal. It might be
appropriate to perform a partial cleanup, and plan on institutional controls for an indefinite period
of time to prevent human intrusion. But is it reasonable to count on these services continuing
far into the future? :'
Another important complicating factor in the establishment of criteria is that the idea of
"radiation" tends to evoke emotions in our fellow citizens that are perhaps best described as
"intense". People frequently feel that sources of radiation are beyond their control, unnatural, and
certainly not freely chosen. Such a situation spells "outrage", and that outrage may well stand
in the way of setting good and equitable residual radioactivity standards. We must take into
account, and respect, the opinion of the public.
Meanwhile, decommissioning may result in the accumulation of large amounts of slightly
contaminated materials and equipment. There are thousands of tons of slightly radioactive
copper and nickel and steel which, if properly used, could be of significant economic and
societal value. But if released for general circulation, they could cause harm. In addition to the
potential health risks, they could be highly damaging to the photographic, microelectronics,
nuclear counting, and other sensitive industries. Restricted recycling, on the other hand, such
as employing contaminated steel for railroad tracks or in bridges, might fall ai the other end of
the benefit-risk scale. .
A VARIETY OF PERSPECTIVES IS ESSENTIAL " "'•
The selection of clean-up criteria for decommissioning and release of radioactively
contaminated sites and materials is an extremely complex task, one that requires hard work, the
exercise of intelligence, imagination, and good judgment. It is essential that the people involved
in the process represent a wide variety of perspectives. We need input from everyone who has
anything to do with the problem, from cradle to grave. Let me identify some of the people that
I have in mind:
• The people who prepare and employ the materials that cause the problem - They extract
the radioactive materials from ores and process it into yellowcake. they build'arid
operate the facilities that produce the radiopharmaceuticals arid the industrial
radionuclides, the electricity and the nuclear weapons. •'• " r ; '."""'^;>*;1'-;:'••>•"•''
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• The people who created the problem as it exists now - Some poured their chemical and
radioactive wastes into holes in the ground, and others stored it in leaking tanks and
forgot about it. Most disposed of it exactly as they were supposed to, in strict
accordance with the law. Unfortunately, the law itself was not adequate to deal with the
problem.
• The people who now have the problem, and must resolve it - the owners of nuclear power
plants that will soon be decommissioned; the Department of Energy, which has
awakened to a loud public clamoring about past mistakes at weapons production
facilities; and the managers of Superfund, saddled with sites that no longer have any
owners or responsible parties.
• The people who will establish cleanup levels for such sites, and figure out how actually
to clean them up - They are officials of Federal and State governments and their
contractors, and those in organizations in a position to offer expertise and advice, such
as the National Academy of Sciences, the National Council on Radiation Protection and
Measurement, and the Health Physics Society.
• And finally, the people who must actually carry out the cleanup work, and who will have
to ensure, after the fact, that a good enough cleanup job was done - These are the
utilities, and the Federal and State agencies, again, with the newspapers and
environmental groups looking over their shoulders.
It is important to understand and take into account the perspectives and experiences of
all these sorts and groups of people. And for this reason, I am extremely glad that you are here
today. I believe that you can play a constructive role in the early stages of developing residual
radioactivity criteria. And I believe that this gathering is an excellent collection of the kinds of
experts needed to address so pressing a problem.
s \
I thank you for coming, and wish you success in this important workshop.
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Session I
Extent of the Cleanup Problem
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Department of Energy
Environmental Restoration and Waste Management
Five-year Plan
Environmental Restoration Program
R. P. Whitfield
U. S. Department of Energy, Headquarters
ABSTRACT
On September 1, 1989, the Department of Energy (DOE) made available for public
comment the first Five-Year Plan for Environmental Restoration and Waste Management. This
plan establishes an agenda for compliance and cleanup against which progress will be
measured, and it establishes a 30-year goal for the completion of environmental cleanup. Specific
implementation plans are being developed by the DOE's field Operations Offices. The Five-Year
Plan and Operations Office Implementation plans are "living documents" that will be updated
annually. The Environmental Restoration (ER) program, as addressed by the plans, deals with
the assessment and cleanup of inactive potential release sites, the decontamination and
decommissioning of surplus nuclear facilities, and technology development needed for
remediation activities. Preparation of the Five-Year Plan began in March 1989, when a task force
was created, and a guidance was issued for field input that provided the basis for the plan.
Validated field input was integrated and manipulated electronically to generate the data needed
to establish the problem scope, priorities, funding requirements, and other elements of the plan.
The data show that the problems within the ER program include approximately 3,700 potential
release sites, more than 5,000 vicinity properties connected with the remediation of uranium mill
tailings, and approximately 500 contaminated facilities. The estimated funding requirement for all
ER activities for the period of 1991 through 1995 is $6.8 billion. In addition, several key needs
have been identified while preparing the plan. DOE has developed strategic objectives for ER
that include an aggressive applied research and development effort, and it is taking actions to
address the problems and needs associated with environmental restoration and waste
management. As part of the implementation process, these efforts include participation and
review by involved parties.
On September 1, 1989, the Department of Energy (DOE) made available for public
comment the first Five-Year Plan for Environmental Restoration and Waste Management. This
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plan establishes an agenda for compliance and cleanup against which progress will be
measured. This document will be the directive for the development of specific implementation
plans by DOE's eight field Operations Offices.
The Environmental Restoration Program (ER) is concerned with the assessment and
cleanup of facilities and sites that are no longer a part of active operations. Various amounts and
types of wastes have accumulated at these facilities and sites as a result of defense programs,
nuclear energy, and energy research program spanning nearly five decades. Included within the
scope of ER are Remedial Actions (RA) and Decontamination and Decommissioning (D&D). In
addition, technology development and demonstration necessary for the assessment and cleanup
of inactive sites and facilities are within the scope of the ER program.
The RA program is concerned with the assessment and cleanup of inactive, potential
release sites including burial grounds, spill sites, pits, cribs, lagoons, buried tanks, and uranium
mill tailings. Active disposal facilities do not fall within the scope of RA. The tasks associated
with ER encompass site discovery, preliminary assessment and inspection, site characterization,
analysis of cleanup options, selection of remedy, cleanup and site closure, and site monitoring.
The D&D program addresses the safe caretaking of surplus nuclear facilities until either
decontamination for reuse or their complete removal. This includes all tasks connected with
assessment and characterization, environmental review, engineering, D&D operations, and
closeout.
Preparation of the Five-Year Plan, including the portion dealing with ER began in the early
spring of 1989. A task force was created, and a guidance was issued to the eight Operations
Offices requesting input on activities proposed during the five-year planning window. The
guidance also defined planning areas, established criteria for assigning priority levels, and
designed the format and content of input on proposed activities.
Approximately 800 activity data sheets (ADSs) were prepared by the eight Operations
Offices for ER activities and were submitted as input to the overall Five-Year Plan. The ADSs
were submitted to the task force in the form of a data base diskette as well as hard copy. The
input from the field was reviewed by the task force to ensure accuracy, completeness, and
conformance with the guidance. The validated data were then compiled into a computer data
base by which the information was integrated and managed. Each ADS was assigned a Budget
and Reporting (B&R) code and subprogram category (e.g., ER), which allowed the input data to
be sorted electronically with respect to program (e.g., Defense Programs), subprogram, category
(CERCLA, RCRA, etc.), and phase (assessment or cleanup). The data base also included the
assigned priority level, descriptive keywords, and funding summaries. Other information provided
on the ADS hard copies, but not part of the data base included cost estimates, major milestones,
major items of cost, and a statement of the level of confidence in the information presented.
ER activities are organized into four interim priority categories. The priority assignments
will be reviewed by DOE on an annual basis and, to the extent that circumstances associated
with a specific activity change, its priority may change correspondingly. The priorities are listed
below.
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Priority-1 includes (1) protecting workers and the public from near-term (i.e., within 5
years) potential health risks, (2) containing near-term off-site spread of groundwater and soil
contamination, (3) preventing unnecessary disruption of ongoing assessment and cleanup work,
and (4) preclosure surveillance and D&D.
Priority 2 include activities, not otherwise assigned to Priority 1, that are required by in-
force agreements or agreements expected to be placed in force during 1991.
Priority 3 includes all activities, not assigned to Priorities 1 and 2, that will best (1) reduce
the potential for health and environmental risk, (2) promote regulatory compliance, (3) reduce
public concern, and (4) ensure no disruption in DOE's missions.
Priority 4 includes activities not covered under Priorities 1, 2, and 3. Priority 4 is
concerned with D&D activities that involve no present imperatives or significant benefits
associated with immediate cleanup.
Approximately 3,700 potential release sites have been identified for PA These site include
about 2,480,000 cubic meters of low level waste consisting of discarded materials such as tools,
paper, and rags, primarily in burial grounds, to be assessed and remediated as appropriate. In
addition, unknown portions of 192,000 cubic meters of pre-1970 buried transuranic (TRU) waste
are in inactive sites and as such are within the scope of ER. TRU waste refers to substances
contaminated with manmade radioactive elements, principally plutonium, having an atomic
number greater than that of uranium, a half-life greater than 20 years, and a concentration greater
than 100 nanocuries per gram. Examples of TRU waste include metal, glassware, process
equipment, soil, filters, and clothing. Remaining sites consist of hazardous and mixed hazardous
and radioactive waste releases. In addition, more than 5,000 vicinity properties are connected
with the Uranium Mill Tailings Remedial Action Program. The principal concerns connected with
these RAs pertain to groundwater and soil contamination.
Approximately 500 contaminated facilities are included under the D&D effort. Groundwater
and soil contamination are associated with only a relatively small number of facilities. The
• majority of activities involve assessment and cleanup of facilities from which there has been no
release of radioactive, hazardous, or mixed substances. The principal concerns pertain to the
collection, retention, and ultimate disposal of contaminating substances and debris.
The 30-year goal for ER is to ensure that risks to the environment and to human health
and safety posed by inactive and surplus facilities and sites are either eliminated or reduced to
prescribed, safe levels. A set of discrete strategic objectives connected with RA and D&D define
the overall approach to achieving this goal.
The objectives of RA are to (1) identify inactive, contaminated facilities and sites at DOE
nuclear installations, (2) assess these facilities and sites to determine the nature and extent of
contamination, (3) confine and contain existing contamination to the extent necessary for
minimizing its further spread, (4) provide for negotiated agreements with regulatory schedules
for the cleanup of these facilities and sites, (5) ensure that cleanup is carried out in strict
compliance with these agreements, and (6) provide long-term monitoring to ensure continuing
compliance.
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The strategic objectives associated with D&D are to (1) maintain facilities awaiting either
decontamination or decommissioning in a manner that limits worker, public, and environmental
exposure to potential hazards; (2) assess facilities to determine the nature and extent of
contamination; (3) decontaminate facilities designated for reuse to the extent necessary for
compliance with approved health and safety standards; and (4) decommission all other facilities
In accordance with the requirements set forth in an approved environmental compliance plan.
The role of applied R&D in the strategic approach for ER is to (1) provide an improved
technical and economic basis for dealing with environmental and health hazards through
development of improved and new assessment and cleanup technologies, (2) reduce the
potential for exposure of the public through development of automated remote handling
technologies, and (3) broaden the technical base by adapting technologies not previously
considered for application to this field. R&D activities that provide a benefit return in a short time
frame will be emphasized.
While preparing the plan, several key needs were identified, including (1) a centralized
management structure for Environmental Restoration and Waste Management Activities; (2) a
cultural transition from a production-oriented mentality to one stressing open communication,
clearly understood and demonstrated priorities for environmental stewardship, and accountable
management; (3) a national consensus on goals, objectives, and implementation strategy; (4) an
aggressive applied research, development, demonstration, testing, and evaluation program; (5)
adequate resources including sufficient personnel with proper qualifications to manage and
review the work; and (6) consistency in the implementation of environmental regulations.
DOE is taking actions to address the problems and needs associated with Environmental
Restoration and Waste Management. Specifically, DOE will (1) comply with environmental and
health related laws; (2) develop a national prioritization system with state, tribal, and other public
involvement; (3) contain known contamination and vigorously assess the uncertain nature and
extent of contamination to enable realistic planning, scheduling, and budgeting; (4) support the
establishment of interagency agreements and fulfill the requirements of existing compliance
agreements; (5) release health records of DOE employees for scientific analysis; (6) implement
waste minimization programs; (7) establish an Applied R&D program involving university research
capabilities, industry, national laboratories, and other federal agencies to determine and rank
R&D needs and pursue new and improved technologies for minimization and remediation; (8)
effect a cultural shift toward clear and open communications; (9) work diligently to achieve
congressional support; (10) take innovative steps to develop, motivate, and allocate the human
resources needed to implement compliance and cleanup activities; (11) recognize tribal
sovereignty and treaty rights related to tribal and ceded lands; and (12) continually examine
environmental regulations to ensure that DOE's compliance actions effectively reduce risk to
human health and the environment.
Issuance of the first Five-Year Plan initiates an ongoing process within DOE that sets the
path for achievement of the goal for completion of environmental cleanup within 30 years.
Operations Offices have been directed to develop five-year implementation plans that will include
participation and review by involved regional parties in the same manner as the DOE Five-Year
Plan. The implementation plans will be used in the management and implementation of actions
undertaken by each Operations Office. Future annual updates to the Five-Year Plan,
Implementation Plans, and accountability will follow the cycle illustrated in Fig. 1. The second
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and subsequent planning/implementation cycles will follow the federal budgeting calendar as
shown in Fig. 2. . ..,-•.
Fig. 3 illustrates funding requirements estimated for each year of the period from 1989
through 1995 by priority level and phase (assessment or cleanup). The amounts shown for 1989
are those currently appropriated. For 1990, the funds identified are estimated requirements for
all activities. The Funding levels shown for 1991 through 1995 are estimates of requirements for
funding all FiA, D&D, and R&D activities. They do not represent a projection of DOE budgets.
The total for this period is $6.8 billion.
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Figure 1
THE ENVIRONMENTAL RESTORATION PROGRAM
Included Within The Scope Of Environmental Restoration Are:
• Remedial Actions - Assessment And Cleanup Of Inactive Potential Release
Sites .'.'"
• Decontamination And Decommissioning - Safe Caretaking Of Surplus Nuclear
Facilities Until Either Decontamination For Reuse Or Their Complete Removal
• Technology Development And Demonstrations Necessary To Conduct
Cleanups
Includes Defense Programs, Nuclear Energy, And Energy Research Facilities
Figure 2
PROCESS USED TO ACCUMULATE AND INPUT DATA
ENVIRONMENTAL RESTORATION SECTION OF THE FIVE-YEAR PLAN
Need Established:
Guidance To The Field:
Field Input:
Validation Of ADSs:
An Agenda For Compliance And Cleanup Against Which
Progress Will Be Measured
Defined Planning Areas, Priority Levels, And Established Format
And Content Of Input On Proposed Activities
Approximately 800 Activity Data Sheets (ADSs) Were Prepared
By The Field For Proposed Environmental Restoration Activities
And Were Submitted As Input To The Five-Year Plan. The ADSs
Were Submitted In The Form Of A Data Base Diskette As Well As
Hard Copy
Input From The Field Was Reviewed To Assure Accuracy,
Completeness, And Conformance With Guidance
22
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Figure 3
PROCESS FOR INTEGRATING AND MANAGING DATA
ENVIRONMENTAL RESTORATION SECTION OF THE FIVE-YEAR PLAN
• A Computer Data Base Was Established Utilizing Validated Input From The Field
• The Activity Data Sheet B&R Codes Were Trie Key To The Accurate And Timely
Integration And Management Of Data.
• The Application Of B&R Coding Is Such That Activity Data Can Be Sorted
Electronically With Respect To Program/Subprogram (ER), Category (RCRA, CERCLA,
Etc), Phase (e.g., Assessment, Cleanup) .
• Additional Parameters On The Computer Data Base Include Assigned Priority Level,
Descriptive Key Words, And Funding Summaries
• Other ADs Information Utilized In The Preparation Of The Plan But Not Incorporated In
The Computer Data Base include Bases For Cost Estimates, Major Milestones, Major
Items Of Cost, And A Statement Of The Level Of Confidence In The Information
Presented
•- ." " : Figure 4
CRITERIA FOR PRIORITY RANKING OF
ENVIRONMENTAL RESTORATION ACTIVITIES FIVE-YEAR PLAN
Priority 1: Required For Near Term Protection Of Workers And General Public From
Potential Health Risk, Containing Near-Term Offsite Migration Of Soil/Ground-
water Contamination, Ongoing Assessment And Cleanup, Preclosure
Surveillance And Maintenance
Priority 2: Activities Not Assigned P1 Required By Agreements In Place Or Expected In
Place By FY 1991
Priority 3: Activities Not Assigned P1 Or P2 That Will Best Reduce The Potential For
Health And Environmental Risk, Promote Compliance, Reduce Public Concern,
Ensure No Disruption In DOE's Mission :;
Priority 4: Activities Not Included In The Above Consisting Primarily Of Decommissioning
And Decontamination Activities Where Immediate Cleanup Is Not Needed
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Figure 5
EXTENT AND NATURE OF THE PROBLEM
WITHIN THE SCOPE OF THE ENVIRONMENTAL RESTORATION PROGRAM
With Respect To Remedial Actions, Approximately 3,700 Release Sites
• Approximately 2,480,000 Cubic Meters Of Low Level Waste Consisting Of Discarded
Materials Such As Tools, Paper, And Rags, Primarily In Burial Grounds, To Be
Assessed And Remediated As Appropriate
• Approximately 192,000 Cubic Meters Of Pre-1970 Buried Transuranic Waste, A Portion
Of Which Is Within The Scope Of The ER Program
• Hazardous And Mixed Wastes (Radioactive Wastes Which Contain Hazardous
Substances)
• More Than 5,000 Vicinity Properties Connected With The Uranium Mill Tailings
Remedial Action Program
The Principal Concerns Connected With The Remedial Actions Pertain To Groundwater And
Soil Contamination
There Are Approximately 500 Contaminated Facilities Included Under D&D
• Approximately 400 Defense Programs Installations
• 30 Are Nuclear Energy Facilities
• The Remainder Are Connected With Energy Research
Soil And Ground-water Contamination Are Associated With Only A Small Number Of These
And The Majority Do Not Involve The Release Of Radioactive, Hazardous, Or Mixed
Substances.
The Principal Concerns Pertain To The Collection, Retention, And Ultimate Disposal Of
Contaminating Substances And Debris.
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Figure 5 (con't)
EXTENT AND NATURE OF THE PROBLEM
WITHIN THE SCOPE OF THE ENVIRONMENTAL RESTORATION PROGRAM
(Continued)
Several Key Needs Have Been Identified In The Process Of Preparing The Plan Including The
Following:
• A Centralized Management Structure For Environmental Restoration And Waste
Management Activities
• Cultural Transition From Production Oriented To One Of Open Communication, Clearly
Understood And Demonstrated Priorities For Environmental Stewardship, And
Accountable Management
• A National Consensus On Goals, Objectives, And Implementation Strategy
• An Aggressive Applied Research, Development, Demonstration, Testing, And Evaluation
Program
• Adequate Resources Including Sufficient Personnel With Proper Qualifications To
Manage And Review The Work
• Consistency In The Implementation Of Environmental Regulations
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Figure 6
HOW THE PROBLEMS AND NEEDS WILL BE DEALT WITH
IN TERMS OF THE FIRST FIVE-YEAR PLAN
The Specific Actions Being Taken By DOE To Address The Problems And Needs Associated
With Environmental Restoration And Waste Management Are:
• Comply With Environmental And Health Related Laws
• Develop A National Prioritization System With State, Tribal, And Other Public
Involvement
• Contain Known Contamination And Vigorously Assess The Uncertain Nature And Extent
Of Contamination To Enable Realistic Planning, Scheduling, And Budgeting
• Support The Establishment Of Interagency Agreements And Fulfill The Requirements Of
Existing Compliance Agreements
• Release Health Records Of DOE Employees For Scientific Analysis
• Implement Waste Minimization Programs
• Establish An Applied R&D Program Involving University Research Capabilities, Industry,
National Laboratories, And Other Federal Agencies To Determine And Rank R&D Needs
And Pursue New And Improved Technologies For Minimization And Remediation
• Cultural Change To One Of Clear And Open Communications
• Work Diligently To Achieve Congressional Support
• Take Innovative Steps To Develop, Motivate, And Allocate The Human Resources
Needed To Implement Compliance And Cleanup Activities
• Recognize Tribal Sovereignty And Treaty Rights Related To Tribal And Ceded Lands
• Continually Examine Environmental Regulations To Ensure That DDEs Compliance
Actions Effectively Reduce Risk To Human Health And The Environment
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Figure 7
STRATEGIC OBJECTIVES FOR ENVIRONMENTAL RESTORATION
30 Year Goal To Reduce Or Eliminate Risks To The Environment And Human Health
And Safety
Identify Inactive, Contaminated Sites And Facilities
Assess Nature And Extent Of Contamination
Confine, Contain Existing Contamination To Extent Needed To Minimize Further
Migration
Provide For Negotiated Agreements For Cleanups i
Ensure Cleanup Conducted In Strict Compliance With Agreements
Provide Long-Term Monitoring To Ensure Continuing Compliance
Figure 8
ROLEOFRDDT&E
IN THE STRATEGIC APPROACH FOR ENVIRONMENTAL RESTORATION
Provide Improved Technical And Economic Basis For Dealing With Problems Through
Development Of Improved And New Assessment And Cleanup Technologies
Reduce The Potential For Exposure Of Public . .;;
Broaden The Technical Base By Adapting Existing Technologies Not Previously Utilized
In This Fields
Emphasize Technologies That Provide A Benefit Return In A Short Timeframe
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Decommissioning Waste Characteristics
Timothy C. Johnson, G. W. Roles
U.S. Nuclear Regulatory Commission
ABSTRACT
This paper describes the expected general characteristics of wastes produced from
decommissioning nuclear facilities. For boiling water reactors and pressurized water reactors,
we summarize information extracted from studies performed by Pacific National laboratories
under contract to the U.S. Nuclear Regulatory Commission. These nuclear facilities will generate
the largest volumes and activities of decommissioning wastes. We also compare these studies
with current waste generation information and briefly address decommissioning waste projections
for other industries.
INTRODUCTION
Decommissioning wastes will be generated by a broad range of licensees. These
licensees include operators of nuclear fuel cycle facilities such as nuclear power reactors, reactor
fuel fabrication plants, and uranium hexaflouride conversion plants. Non-fuel cycle licensees
include hospitals, medical research institutions, colleges and universities, industrial research
laboratories, facilities involved with the production of radiopharmaceuticals, and other industrial
users of radioactive material. Wastes produced by these generators will be very diverse in terms
of volume, activity, and other physical, radiological, and chemical characteristics.
Decommissioning wastes will range from trash that is only suspected of being contaminated to
highly radioactive activated structural components from nuclear power reactors.
In this paper, we present some information on decommissioning wastes obtained from
studies performed by NRC contractors. We also compare this information with data on low-level
wastes (LLW) currently being disposed at LLW disposal facilities.
Our projections of decommissioning wastes are obtained from a series of studies
performed by Pacific Northwest Laboratories (PNL) for the Nuclear Regulatory Commission
(NRC). Two of these PNL studies summarize and classify projected wastes from
decommissioning boiling water reactors (BWR's) and pressurized water reactors (PWR'.s) [1,2].
Decommissioning wastes will be generated that will contain significantly more activity than that
in wastes normally generated during reactor operation. In contrast, decommissioning wastes
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from other fuel-cycle and non-fuel-cycle licensees should be similar to the wastes generated
during normal operations by these licensees. , -
Our data on current waste disposal was extracted from microfiche copies of shipment
manifests and computer-generated reports that NRG staff routinely obtain from the two existing
disposal facility operators [3 - 24]. The two disposal site operators are Chem-Nuclear Systems,
Inc., the operator at the Barnwell, SC disposal site, and U.S. Ecology, the operator of the Beaity,
NV and Hanford, WA disposal sites.
The LLW data is presented in terms of NRC's classification system for LLW. NRC's
regulations require that waste delivered to a low-level waste disposal site be classified into one
of three waste classes - Class A, B, or C - depending upon the concentrations of specific
radionuclides listed in Tables 1 and 2 of 10 CFR Section 61.55. Concentration limits are lowest
for Class A wastes and highest for Class C wastes. The waste classification system is used to
place the most restrictive waste disposal conditions on the most hazardous wastes. Wastes
exceeding Class C concentrations are considered not generally suitable for near-surface
disposal.
Below we present some of the key data relative to decommissioning wastes.
DECOMMISSIONING WASTES FROM NUCLEAR POWER PLANTS
Assuming a reference 1175-megawatt (electric) (MWe) PWR and a reference 1155-MWe
BWR, PNL developed estimates of waste volumes and activities generated from immediate
dismantlement of the power plants after a 40-year operating life. Each reactor was assumed to
have operated at a 75 percent capacity factor, resulting in 30 effective full-power-years of
operation.
Table 1 shows the projected waste stream volumes (in m3) by waste class from
dismantlement of the reference PWR. Virtually all the waste volume is Glass A. Class C and
greater-than-Class C (GTCC) wastes are activated metals from the reactor vessel and internals.
The most voluminous waste stream is Class A contaminated equipment and concrete.
Evaporator bottoms, resins, cartridge filters, and other dry active waste (DAW) are projected to
be both Class A and B wastes.
Table 2 shows the projected activities for the reference PWR. By far, the most active
wastes will be the core shroud and the other GTCC core internals. Although representing a small
volume, these activated metal wastes dominate the PWR decommissioning activity.
We believe that the PNL projections of the activity in evaporator bottoms, resins, cartridge
filters, and DAW are overly conservative. Current waste processing systems generated
substantially lower activities then the PNL report predicts.
Tables 3 and 4 present available 1987 disposal data by waste stream. Note that overall
concentrations for comparable waste streams are significantly lower than those projected by the
PNL report. For example, for PWR resins the PNL report suggests an average concentration of
about 740 curies per cubic meter (Ci/m3). This value exceeds the maximum recommended
activity loading (350 Ci/m3) for organic resins [28] by a factor of two. In actual decommissioning
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practice, radioactive waste concentrations are more likely to be similar, to process waste
concentrations from normal plant operations. ,
Table 5 presents the projected decommissioning waste volumes for the reference BWR.
Note again that virtually all the waste volume is Class A. Class C and GTCC wastes are activated
metals from reactor internals. The most voluminous waste stream is the Class A contaminated
equipment and concrete. Evaporator bottoms, solidified decontamination solutions, filter sludges,
resins, and other DAW are projected to be both Class A and B wastes; • . . • ;
Table 6 shows the projected activities for the reference BWR. By far, the most active
wastes will be the core shroud and the other core internals. Although representing a small
volume, these activated metal wastes dominate the BWR decommissioning activity. The PNL
projections of the activity in evaporator bottoms, filter sludges, resins, and DAW are considered
again to be very conservative. Current waste processing systems generate substantially lower
activities than the PNL report predicts. Tables 3 and 4 again suggest that overall concentrations
of comparable waste streams are significantly less than those projected by the PNL report. In
actual decommissioning practice, radioactive waste concentrations are more likely to be similar
to process waste concentrations from normal plant operations.
Table 7 shows a summary distribution of the PNL volume and activity data from PWR and
BWR decommissioning. .;..
Tables 8 and 9 present data showing the effects of decay on the Class A, B; C and GTCC
PWR and BWR decommissioning wastes. These data are based on the radionuclide distributions
assumed in the PNL reports. The tables clearly illustrate how the waste contamination is
dominated by short-lived radionuclides such as iron-55 (Fe-55: 2.6-year half-life) and cobalt-60
(Co-60: 5.26-year half-life). The activated metals that dominate the Class C and GTGC wastes
are also dominated by Fe-55 and Co-60.
DECOMMISSIONING WASTES FROM OTHER TYPES OF FACILITIES
PNL has estimated decommissioning waste volumes for many other fuel cycle and non-,
fuel cycle facilities [29 - 30]. Table 10 provides a summary of the waste volumes from-these
facilities. The volumes of decommissioning wastes from non-power reactor facilities will depend
on the particular operations of the licensee. The non-fuel cycle waste projections are for
equipment and individual components typically used in non-fuel cycle licensee operations.
Decommissioning wastes froms-non-utility licensees will contain the same radionuclides;
used during operations, and should also be physically and chemically similar. Because wastes
generated during normal operations are relatively low in activity, decommissioning wastes should
also be relatively low in activity. •. • • ? *.
Table 11 presents overall 1987 waste volume and activity data for disposals by industries
generating radioactive waste. These data show that, in terms of waste volume and activity, the
utilities are the principal generators of radioactive wastes. Except for the industrial sector, other
industries generate very little of the activity. Based on our reviews of shipment manifests, the
industrial sector activity is dominated by shipments from a few radioisotope production firms.
Therefore, with the exception of these firms, the radionuclide concentrations in these wastes are
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very small. Radionuclides found in non-utility wastes are quite varied reflecting the many uses
of nuclear materials in the commercial sector.
REFERENCES
[1 ] Murphy, E.S., Technology, Safety and Costs of Decommissioning a Reference Pressurized
Water Reactor Power Station, NUREG/CR-0130, Addendum 3, September 1984.
[2] Murphy, E.S., Technology, Safety and Costs of Decommissioning a Reference Soiling
Water Reactor Power Station, NUREG/CR-0672, Addendum 2, September 1984.
[3] U.S. Ecology, Radioactive Waste Report by Isotope, 1 /1 /87 through 12/31 /87, Beatty, NV
Facility, Report No. ECF883, May 23, 1988.
[4] U.S. Ecology, Radioactive Waste Report by Isotope, 1/1/87 through 12/31/87, Richland,
WA Facility, Report No. ECF883, May 23, 1988.
[5] U.S. Ecology, Radioactive Waste Report by State, Beatty, NV Facility, 1/1/87 through
12/31/87, Report No. ECF884, May 23, 1988.
[6] U.S. Ecology, Radioactive Waste Report by State, Richland, WA Facility, 1/1/87 through
12/31/87, Report No. ECF884, May 23, 1988.
[7] U.S. Ecology, Volume and Activity by Waste Type, Beatty, NV Facility, 1/1/87 through
12/31/87, Report No. ECF887, May 27, 1988.
[8] U.S. Ecology, Volume and Activity by Waste Type, Richland, WA Facility, 1/1/87 through
12/31/87, Report No. ECF887, May 27, 1988.
[9] Utility Data Institute, WASTENET: Radiation Waste Program, Generator Activity by Class,
Beatty, NV Facility, 1/1/87 through 12/31/87, Report No. ECF510, August 25, 1988.
[10] Utility Data Institute, WASTENET: Radiation Waste Program, Generator Activity by Class,
Richland, WA Facility, 1/1/87 through 12/31/87, Report No. ECF510, August 24, 1.988.
[11] Chem-Nuclear Systems, Inc., Total of Each Isotope ^Received per Waste Class for SC
DHEC, 1/1/87 to 12/31/87, Report No. 0060705TRR5493033, September 14,1988.
** ' • *J* ' , - ' ,
[12] Chew-Nuclear Systems, Inc., Volume Percentage by Class and Activity for Each State for
SCDHEC, 1/1/87 to 12/31/87, Report No. 0060705TRR5413018, August 30, 1988.
[13] Chem-Nuclear Systems, Inc., Summarized Volume and Activity by State per Class for
Generators for SCDHEC, 1/1/87 to 12/31/87, Report No. 0060705TRR5563003, September
9,1988.
31
-------
[14] Chem-Nuclear Systems, Inc., Summarized Volume and Activity by State per C/ass for
Generators for'SC DHEC, 1/1/87to 12/31/87, Report No. b060705fRR5853Q48, September
9,1988. .._.....•
[15] Chem-Nuclear Systems, Inc., Summary Breakdown of Isotopes Received for SC DHEC,
1/1/87 to 12/31/87, Report No. 0060705TRR5503007, August 31, 1988.
[16] Chem-Nuclear Systems, Inc., Volume by Waste Description for SC DHEC, 1/1/87 to
1/31/87, Report No. 0060705TRR5383010, September 9,1988.
[17] Chem-Nuclear Systems, Inc., Volume by Waste Description for SC DHEC, 2/1/87 to
2/28/87, Report No. 0060705TRR5383010, September 9,1988.
[18] Chem-Nuclear Systems, Inc., Volume by Waste Description for SC DHEC, 3/1/87 to
' 3/31/87, Report No. 0060705TRR5383010, September 12, 1988.
[19] Chem-Nuclear Systems, Inc., Volume by Waste Description for SC DHEC, 4/1/87 to
4/30/87, Report No. 0060705TRR5383010, September 12, 1988.
i. •
[20] Chem-Nuclear Systems, Inc., Volume by Waste Description for SC DHEC, 5/1/87 to
5/31/87, Report No. 0060705TRR5383010, September 12, 1988.
[21] Chem-Nuclear Systems, Inc., Volume by Waste Description for SC DHEC, 6/1/87 to
6/30/87, Report No. 0060705TRR5383010, September 12, 1988.
[22] Chem-Nuclear Systems, Inc., Volume by Waste Description for SC DHEC, 7/1/87 to
7/31/87, Report No. 0060705TRR5383010, September 12, 1988.
[23] Chem-Nuclear Systems, Inc., Volume by Waste Description for SC DHEC, 8/1/87 to
8/31/87, Report No. 0060705TRR5383010, September 12, 1988.
[24] Chem-Nuclear Systems, Inc., Volume by Waste Description for SC DHEC, 9/1/87 to
9/30/87, Report No. 0060705TRR5383010, September 12, 1988.
[25]. Chem-Nuclear Systems, Inc., Volume by Waste Description for SC DHEC, 10/1/87 to
0/31/87, Report No. 0060705TRR5383010, September 12, 1988.
[26] Chem-Nuclear Systems, Inc., Volume by Waste Description for SC DHEC, 11/1/87 to
11/30/87, Report No. 0060705TRR5383010, September 12, 1988.
[27] Chem-Nuclear Systems, Inc., Volume by Waste Description for SC DHEC, 12/1/87 to
12/31/87, Report No. 0060705TRR5383010, September 12, 1988.
[28] U.S. Nuclear Regulatory Commission, Low-Level Waste Licensing, Branch Technical
Position on Waste Form, May 1983.
[29] U.S. Nuclear Regulatory Commission, Final Generic Environmental Impact Statement on
Decommissioning of Nuclear Facilities, NUREG-0586, August 1988.
32
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[30] Murphy, E.S., Technology, Safety and Costs of Decommissioning Reference
Non-Fuel-Cycle Nuclear Facilities, Pacific Northwest Laboratory, NUREG/CR-1754,
February 1981.
33
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Table 1
Waste Stream Volumes (m3) from Dismantlement of a Reference PWR
Waste Streams
Neutron Activated
Core Shroud and
Adjacent Metal
Other Internals
Pressure Vessel
Concrete
Metal Cavity Liner
Contaminated Material
Equipment, Metal, and
Concrete Surfaces
Process Waste
Evaporation Bottoms
Resins
Cartridge Filters
Dry Activate
Waste
Total
Class A
Class C
GTCC
Total
39
222
707
14
982
16,078
266
195
461
17.521
60
60
57
9
154
214
133
17
133
133
116
222
707
14
1192
17
16,078
266
57
9
283
615
17.885
-------
' ": "' : "' '•'' ..-;..":' ' ' Table .2- ' -'"' - " '' ;
Waste Stream Activities (Ci) from Dismantlement of a Reference PWR
Waste Streams
Neutron Activated
Core Shroud and
Adjacent Metal
Other Internals
Pressure Vessel . 1
Concrete
Metal Cavity Liner
Contaminated Material
Equipment, Metal, and
Concrete Surfaces
Process Waste *
Evaporation Bottoms
Resins
Cartridge Filters
Dry Activate
Waste ;
Total
Class A
Class C . GTC(
Total
212
19,1186
2,121
21,529
997
13,812
5,501
5,501
234
14,046
36.572
41,998
5,040
528
47,566
531067
34,300
34,300
4,784,500 4,784,500
40,013
19,186
2,121
4/784,509 4,845,830
997
13,812
41,998
5,040
_ 762
61,612
4.784.500 4.908.439
35
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Table3 ,
Barnwell 1987 Waste Volumes and Activities
by Waste Type and Class
Waste Type
Volume (ft3)
Resin
Solid Combustibles
Solid Noncombustibles
Filter Media*
Cartridge/Mechanical
Filters**
Solidified Liquids
Equipment, Components
Biological
Incinerator Ash
Air Filtration Filters
Combustibles and Non-
combustibles (Mixed)
Total
Activity (Ci)
Resin
Solid Combustibles
Solid Noncombustibles
Filter Media*
Cartridge/Mechanical
Filters
Solidified Liquids***
Equipment, Components
Biological
Incinerator Ash
Air Filtration Filters
Combustibles and Non-
combustibles (Mixed)
Total
Class A
1.920E+5
1.733E+3
5.158E+4
2.193E+4
5.140E+3
5.836E+4
3.849E+2
5.077E+3
3.000E+1
3.045E+3
5.782E+5
9.175E+5
1.663E+4
8.740E+0
3.009E+2
3.114E+2
1.616E+2
8.332E+2
3.510E+1
1.357E+0
7.000E-5
5.685E+0
1.237E+3
1.953E+4
Class B
2.675E+4
1.190E+2
1.866E+2
1.160E+3
9.044E+2
7.836E+2
2.929E+2
1.016E+3
3.121E+4
1.717E+4
1.300E+2
5.977E+3
3.489E+2
4.456E+2
2.347E+2
1.177E+3
2.742E+3
2.822E+4
Class C
4.414E+3
2.730E+1
1.675E+3
9.700E+2
4.100E+0
7.090E+3
1.040E+4
7.216E+1
1.165E+3
1.516E+5
4.204E-2
1.633E+5
Total
2.231E+5
1.852E+3
5.179E+4
2.309E+4
7.719E+3
5.914E+4
1.648E+3
5.077E+3
3.000E+1
3.045E+3
5.793E+5
9.558E+5
4.420E+4
1.387E+2
6.350E+3
6.603E+2
1.772E+3
1.068E+3
1.528E+5
1.357E+0
7.000E-5
5.685E+0
3.979E+3
2.110E+5
* Used in liquids and other than resin or cartridges.
** Used in liquids.
*** Includes concentrates and sludges.
36
-------
Table 4
1987 Volumes and Activities of Richland and Beatty Wastes
Waste Types
Vials
Dry solid
Solidified liquid
Biological waste ' '
Filter media
Dewatered resin
Solidified resin
Absorbed aqueous liquid
Absorbed organic liquid
Aqueous liquid in vials
Animal carcasses in
absorbent
Gas •V-'T--'
Compacted dry active waste
Noncbmpacted dry active
waste
Other
Total
Richland, WA Site
Vol. (ft3) Act. (Ci)
1.500E+1
4.202E+5
2.729E+4
2.813E4-2
3.465fi+3'
3.851E+4
7.184E+3
2.967E4-4
4.857E+3
1.203E+4
1.187E-2
1.927E+4
2.284E+4
2.606E-2.
3.455E+2.
4.002E+3
7.481E+2
2.412E+2'
1.764E+0
9.921E+0
1.527E+1
5.700E+2 6.264E-1 •
2.075E+'2 6.478E+0
7.657E+3 5299E+0
5.566E+5 4.748E+4
Beatty, NV Site
Vol. (ft3) Act. (Ci)
1.158E+2
2.826E+5
3.578E+4
1.828E+2
1.823E+2
2.354E+3
7.376E+3
9.721E+2
2.352E+0
9.580E+3
6.029E+2
5.859E-3
6.649EOO
8.871E-1 "
9.026E+2
1.743EOO
2.250E+1 6.011E-3
1.602E+3;f'7.931E-l
2.250E+1,
5.431E+2
4.051E-3
2.013EOO
6.876E+2 1.239E+0
3.324E+5 1.110E+4
37
-------
Table 5
Waste Stream Volumes (m3) from Dismantlement of a Reference BWR
Waste Streams
Neutron Activated
Core Shroud
Other Internals
Reactor Vessel
Concrete
Contaminated Material
Equipment, Metal, and
Concrete Surfaces
Process Waste
Evaporation Bottoms
Solidified Decon
Solutions
Filter Sludge,
Resins
Dry Activate
Waste
283
Total
Class A
Class B Class C GTCC Total
15 •
8
90
113
17,229
492
120
54
468
1,134
18.476
15
15
148
210
358
373
53
53
47
47
47
83
8
90
228
53
17,229
640
%°
678
1,492
47 18.949
38
-------
Table 6
Waste Stream Activities (Ci) from Dismantlement of a Reference BWR
Waste Streams
Class A
Class B
Class C
GTCC
Contaminated Material
Equipment, Metal, and
Concrete Surfaces 8,490
Process Waste
Total
Neutron Activated
Core Shroud
Other Internals
Reactor Vessel
Concrete
750
2,160
180
3,090
10,300
10,300
239,000
239,000
6,301,700 6,301,700
. 250,050
2,160
180
6,301,70 6,554,090
8,490
Evaporation Bottoms
Solidified Decon
Solutions
Filter Sludges,
Resins
Dry Activate
Waste
283 ,
Total
1,440
105
227
562
2,334
13.914
31,200
1.260
32,460
42.760
32,640
105
227
1.822
, 34,794
239,000 6,301,700 6,597.374
39
-------
Table 7
Summary Distribution of Volume and Activity
Within Reactor Decommissioning Wastes
Class A
Volume (m3) (%)
Activity (Ci) (%)
PWR
17,521 (98)
36,600 (0.7)
BWR
18,476(97.5)
13,900 (0.2)
Volume (m3) (%)
Activity (Ci) (%)
Class C
Volume (m3) (%)
Activity (Ci) (%)
GTCC
Volume (m3) (%)
Activity (Ci) (%)
214 (1.2)
53,100 (1.1)
17 (0.1)
34,300 (0.7)
133 (0.7)
4,784,500 (97.5)
373 (2.0)
42,800 (0.6)
53 (0.3)
239,000 (3.6)
47 (0.3)
6,301,700 (95.5)
Volume (m3) (%)
Activity (Ci) (%)
17,885
4,908,400
18,949
6,597,400
40
-------
Table 8
Decay of Class A, B, and C Wastes From Reactor Decommissioning (Ci)
Time After
Disposal (yr)
Reference PWR
0
10
50
100
500
1,000
Reference BWR
0
10
50
100
500
1,000
Class A Class B
36,600
9,480
168
67
3.1
0.9
13,900
3,830
184
64.7
1.2
0.4
53,100
13,500
367
163
7.3
1.7
Class C
Total
Total
Reduction
Factor
34,300
9,890
1,090
726
45
10.2
124,000
32,900
1,630
956
55
12.8
1
4
76
130
2,300
9,700
42,800
11,900
684
. 286
10.2
2.8
239,000
65,100
5,310
3,500
231
64.9
296,000
80,800
6,180
3,850
242
68.1
1
4
48
77
1,200
4,300
Table 9
Decay of GTCC Wastes from Reactor Decommissioning
Reference PWR Reference BWR
Time Cvf) Activity (Ci) Red. Factor Activity (Ci) Red. Factor
0
5
10
30
50
100
300
500
1000
4,780,000
2,530,000
1,380,000
254,000
159,000
109,000
28,300
8,090
1,550
1
1.9
3.5
19
30
44
170
590
3,100
6,300,000
3,230,000
1,710,000
258,000
147,000
100,000
26,200
7,770
1,780
1
2
3.7
24
43
63
240
810
3,500
41
-------
Table, 10
Decommissioning Waste Volumes from Non-Power Reactor Facilities
Facility
Research and Test Reactors
Fuel Reprocessing
Non-TRU
TRU
Uranium Hexafluoride Conversion
Uranium Fuel Fabrication
Equipment, Concrete, etc,
Calcium Fluoride Waste
Spent Fuel Storage
Wet
Dry Well
silo
Vault
Cask
Non-Fuel-Cycle
Fume Hood
Glove Box
Small Hot Cell
Laboratory Workbench (4.6m x 0.9m x 0.75m)
Sink and Drain
Room (6m x 10m x 3m)
Waste Volume (m3)
160-4,930
3,100
4,600
1,259
1,100
29,600
2,720
6,700
920
500
42
* ,
4
4
2
7.2
0.4
7.5
42
-------
Table 11
1987 Low-Level Waste Volume and Activity Sorted by Generic Industry
Variable
Vol. (ft3)
Act. (Ci)
Hospitals
and
Health
Services
Colleges
and
Education
Services
Government Industry
Total
B*
R
Be
6.252E+5
2.422E+5
7.198E+4
9.394E+5
(50.9%)
9.740E+2
2.151E+4
2.248E+4
(1.2%)
1.091E+4
1.827E+4
9.900E+2
3.017E+4
(1.6%)
5.951E+4
9.916E+3
6.125E+4
1.307E+5
(7.1%)
2.592E+5
2.647E+5
1.982E+5
7.221E+5
(39.1%)
9.558E+5
5.566E+5
3.324E+5
1.845E+6
B
R
Be
2.009E+5
1.721E+4
1.632E+3
2.197E+5
(81,5%)
1.149E-1
2.420E+1
2.431E+1
(0.009%)
1.591E+1
5.201E+1
5.472E-1
6.847E+1
(0.025%)
6.086E+3
9.855E+2
5.293E+0
7.077E+3
(2.6%)
3.975E+3
2.922E+4
9.463E+3
4.266E+4
(15.8%)
2.110E+5
4.748E+4
1.110E+4
2.696E+5
B: Barnwell; R: Richland; Be: Beatty.
43
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Site Inventory of Residual Radioactivity in Japan
Shohei Kato, Fuyuhiko Ishikawa and Hideaki Yamamoto
Department of Health Physics
Japan Atomic Energy Research Institute
ABSTRACT .
The types and numbers of facilities in Japan that may require residual radioactivity criteria
for decommissioning have been investigated. The quantities of decommissioning wastes were
estimated. The characteristics of the residual radioactivity at the facilities are discussed, based
on our decommissioning experience.
INTRODUCTION
Applications of atomic energy have been developed in Japan under "the principle of
peaceful use" prescribed in the Atomic Energy Basic Law, which came into force in 1956 [1].
Nuclear power generation has grown to be one of the major sources of electricity in Japan.
Radioisotopes and particle accelerators have come into wide use. Most nuclear fuel facilities are
coming into a practical and commercial stage from an experimental stage.
No practical scale facility-has undergone decommissioning. However, some research
reactors and facilities using radioisptopes have been dismantled. Decommissioning of \arge scale
facilities is expected to begin in the near future.
In the present paper, an inventory of contaminated sites, characteristics of the residual
radipactivity, and some experiences with decommissioning are reported.
FACILITIES AND THE CHARACTERISTICS OF RESIDUAL RADIOACTIVITY
Reactor
There are four kinds of reactors: commercial power plants, research reactors, critical
experimental facilities and developmental reactors. The number of power plants has been
increasing since 1966. Thirty-six power plants (17 BWRs, 18 PWRs and one GCR) generate 27,000
MW, and 12 power plants are under construction, as shown in Figure 1 [2]. The electricity
generated by nuclear power plants was 25% of the total electricity demand in Japan in 1986.
44
-------
Assuming that a power plant operates for 40 years and is mothballed for 10 years after
shutdown, it is expected that decommissioning undertaken seriously will begin in 2020, and that
up to five power plants will be decommissioned every five years. This decommission will yield a
large amount of radioactive waste.
Ohta reported the estimated amount of radioactive waste produced in decommissioning
a PWR and a BWR (1100 MW), assumed to be closed for 10 years after shutdown, as shown in
Table 1 [3]. Radioactive concentration in the most radioactive waste is lower than 10"4 Ci/ton.
Radioactive metal waste and concrete waste are estimated to be 30,000 tons (5.9% of total
radioactive wastes) and 500,000 tons (93%), respectively for BWR, and 30,000 tons (6.1%) and
460,000 tons (91%), respectively for PWR.
Based on the data in NUREG/CR-0672 [4], the Japan Power Demonstration Reactor (JPDR)
Research Committee also estimated in detail the quantities of materials contained in a reference
BWR plant, as shown in Table 2 [5]. The table indicates that about 1,500 tons of nonferrous
metals will be generated from the decommissioning of a BWR power plant. The differences of the
estimated quantities are assumed to be caused from the differences in calculation conditions such
as history of operation, decontamination factors and extent of evaluation of quantities.
The data obtained from the JPDR decommissioning project are useful for estimating the
amount and characteristics of the radioactive waste from the decommissioning of a commercial
power reactor. JPDR is a light-water-cooled boiling water reactor which operated at a power of
45 MW for 10 years and at 90 MW for 1 year, and was shutdown in 1976, as shown in Table 3.
The reactor is under decommissioning at the present. Table 4 and Figure 2 show the estimated
amount of radioactive wastes from the JPDR decommissioning. In these statistics, amounts of
uneontaminated materials are not included. Metal waste is estimated to be 1,840 tons (1,640 tons
of contaminated metals and 200 tons of activated metals). Approximately 85% of the waste have
radioactivity concentrations lower than 10^Ci/ton. The amount of the metal waste from JPDR was
estimated to be one sixteenth of that from a typical 1,000 MW power plant, which is about twenty
times the size of JPDR. Major metals involved in JPDR are carbon steel and stainless steel. Also,
a small amount of copper waste comes from cables and heat exchangers. There are two kinds
of residual radioactivity, contamination and activation. Contaminated concrete was easily cleaned
up. By contrast, activation products are widely distributed in the concrete.
There are 11 research reactors, most of which were built in the 1950's. A research reactor,
JRR-3 was dismantled and is being converted into a new type of research reactor. A large amount
of concrete contaminated with tritium was generated from the dismantling of the reactor. Since
tritium is volatile and easily permeates concrete, it spread inside the whole facility and diffused into
the concrete. It was difficult to determine the residual activity level below which a cleanup is not
required.
There are 4 developmental reactors, JPDR, JOYO (a fast breeder experimental reactor, 100
MW in 1979), FUGEN (an advanced thermal reactor, 557 MW in 1977) and MONJU (a prototype
fast breeder-reactor, 714 MW which is under construction).
There are seven critical experimental facilities. In these facilities, the residual activity is
generally limited to several components and therefore the amount of radioactive waste from
dismantling is relatively small.
45 -
-------
Nuclear fuel cycle facilities. ;
A demand to develop nuclear fuel cycle facilities is increasing in Japan. The amount of
uranium deposits in Japan is an estimated 4,000 U3O8 tons. All uranium used in reactor fuel is
imported from foreign countries. At two sites, Ningyo-type and Tohko, uranium was mined and
milled experimentally. However, these sites are closed at present. A uranium mill facility at Tokai
had been operating from 1959 to 1970, and the facility was dismantled in 1978. A pilot plant for
milling and conversion started operation in 1982, producing UF6 at a rate of 300 tons/year. A pilot
uranium enrichment facility using the centrifuge method started operation at a rate of 50 ton
SWU/year in 1979. On the basis of this operation experience, a prototype plant will come into
operation in 1989. A commercial uranium enrichment plant is scheduled to be built in 1995. A
large amount of metal materials contaminated with alpha emitters is expected to be produced from
dismantling of the uranium enrichment plant!
There are two uranium conversion Plants. One facility started operation in 1962 and
processed at a rate of 475 tons of Uranium per year (tU/y). The other started in 1980 and
processes 550 tU/y., There are four uranium processing facilities with total processing capacity of
2,000 tU/y. :
A pilot plant for fuel reprocessing started treatment at a rate of 0.7 tU/d in 1970.
Construction plans for a commercial Preprocessing plant are in motion. There is no plan for
decommissioning a commercial scale fuel cycle facility in the near future.
However, some experimental facilities have been closed and dismantled. A distinctive
feature of nuclear fuel cycle facilities is that they may be contaminated with alpha emitters. Soil
produced from uranium mine contains naturally occurring radioactive materials such as U, Ra, Th.
The soil was disposed of according to the Mine Safety Law. In a fuel processing facility, floors
and walls, especially ventilation systems, were slightly contaminated with U. The dismantling
experience indicated that data of incidental contamination, as well as specifications of the facility,
were very useful for dismantling plans*
Facilities using radioisotopes
The trend in the number of facilities which have licenses to use radioisotopes or particle
accelerators is shown in Figure 3. The number of licenses has increased from 200 in 1959 to
4,257 in 1988. More than 80% of these facilities (4,257 facilities) are licensed to use only sealed
radionuclides. For the facilities in which unsealed radionuclides are handled, there is a possibility
of contamination with radionuclides. Table 5 shows the amount of radionuclides used in the
facilities. The number of facilities using unsealed radionuclides is 82 for medicine, 286 for
education and 391 for research. Major radionuclides are Ga-67, Tc-99m and Pm-149 for medicine,
H-3 and P-32 in education, and H-3 and C-14 for research. Important nuclides which should be
considered in decommissioning are relatively long-lived nuclides, with half-lives longer than 1 year,
such as H-3, C-14, Se-75, Pm-147.
JAERI has experience in decommissioning a large radio-chemical laboratory, to convert it
into a laboratory where no radionuclides are to be used. In the radio-chemical laboratory, a
distinguishing characteristic of residual radioactivity is contamination with H-3 and C-14 through
air contamination.
46
-------
Particle accelerators ; . ; ,,
The number of particle accelerators in use as of March 31, 1988 is shown in Table 6,
classified from the point of view of accelerator types and organizations [9]. The total ntirriber of
accelerators is 748. The number of Linear Accelerators or LINACS is 497, which is two thirds of
the total. The numbers of Cockcroft-Walton, Van de Graaff and cyclotron accelerators are 85, 46
and 34 respectively. Furthermore, Synchrotron Orbital Radiation (SOR) facilities1 (three in operation
and three others are under construction) and experimental fusion facilities may also be categorized
as particle accelerators. / . ' ~ '.•
The electron accelerators with energy lower than 1 MeV which are commonly used in
industrial processing do not come under "radiation generators" in the law concerning prevention
of radiation hazard due to radioisotopes, etc. [10]. Thus, these accelerators may be sold to scrap
markets free from regulatory restrictions.
It is expected that electron accelerators with energy lower than 10 MeV and positive-ion
accelerators with energy lower than 1 MeV also can be reused or recycled, because induced
activity produced by these accelerators is absent or negligible [11, 12, 13]. According to the
inquiry carried out by the High Energy Physics Research Institute (KEK) [14], about half of all
electron accelerators have energies greater than 10 MeV, and approximately 80% of the
positive-ion machines have energy greater than 1 MeV. Fusion experimental facilities also produce
neutrons and thus induce activity.
For these relatively high-energy accelerators, radioactivity is diluted and dispersed. The
radioactivity concentration is relatively low in most parts of the 'machine except in components
such as targets, target holders and positioning mechanisms, slits and collimators [11]. Thus most
components are expected to be potential candidates for reuse and recycle. It should be
reasonable to reuse or recycle a magnet (carbon steel), coil (copper or titanium), cyclotron vacuum
chamber (stainless steel, aluminum alloy or aluminum) and linac drift chamber (stainless steel or
aluminum).
For synchrocyclotrons of greatest energy (approx. 730 MeV), the total mass is 8,900 tons
[9]. Table 7 lists dominant radionuclides shown in the materials of particle accelerators [11, 14,
15, 16, 17]. By comparison with fuel cycle facilities, accelerators do not produce alpha-emitters,
but rather small amounts of short-lived radionuclides. The dominant radionuclide in aluminum is
Na-22. In concrete, Na-22, Co-60, Eu-152 and Eu-154 are important in the long term. In ferrous
materials, Mn-54, Fe-55 and Co-60 are dominant nuclides. In copper, Co-60 and Zn-65.are
important.
DISCUSSION '
An inventory of residual radioactivity sites in Japan and their characteristics were reviewed
in previous sections. There is no experience in decommissioning a commercial facility which is
supposed to generate a large amount of radioactive wastes. But decommissionings of small scale
experimental facilities are reported.
According to these experiences, the most practical way to reduce residual activity is to
prevent components and building structures from being contaminated, and to remove
47 "'
-------
contamination when present. It is then important to give consideration to contamination controls
and simplification of decommissioning in designing the facility.
t
It fs also essential to keep detailed documents concerning contamination incidents as well
as maintenance and/or remodeling records.
A large amount of residual radioactivity in a nuclear reactor is generated both by
contamination and by activation. Uranium mine and milling facilities produce soils containing
natural radioactive materials. Fuel processing and fabrication plants yield enriched uranium
contamination. The decommissioning of fuel concentration facilities is expected to generate a
large amount of metal materials contaminated with alpha emitters. A reprocessing plant is
contaminated with alpha emitters, fission products and activation products. Major residual
radioactivity in a facility using radioisotopes are to be distinguished from contamination with long
lived nuclides such as H-3, C-14. Some particle accelerator facilities may yield activation products.
Facilities which generate a large amount of useful metal materials from decommissioning
are nuclear reactors, fuel concentration plants, fuel reprocessing plants, and particle accelerators.
The residual radioactivity can be classified into three groups: natural radioactivity in the soil,
contamination and activation. The methods of cleanup will be selected depending on the
characteristics of the residual radioactivity. When natural radioactivity in soil is homogeneously
distributed, a cleanup is called recovery rather than decontamination. The components
contaminated with nonvolatile materials are easily cleaned up because such contamination is
limited to the surfaces of a facility or of equipment. However, the components contaminated with
volatile materials such as H-3 are not easily cleaned up, because the contamination spatially
distributes inside the facility and diffuses into materials. Components contaminated with activation
products are also difficult to reduce in activity. Proper clean-up methods should be selected
' depending on the material and the characteristics of the residual radioactivity.
Safety assessment for reuse of land and/or a facility should be conducted depending on
the characteristics of the residual radioactivity. It is also essential to choose a proper plan for
recycling materials, depending on the characteristics of the residual activity.
It is therefore necessary to give consideration both to the method of recycling or reuse and
to the characteristics of the residual radioactivity in establishing residual radioactivity criteria.
ACKNOWLEDGMENT ;
We would like to thank Mr. Hozumu Yamamoto, Department of Health Physics, JAERI for help and
support.
REFERENCES
[1 ] Atomic Energy Basic Law, 1956
[2] T. Asada et al.: Genshiryoku Handbook. Ohm Co. Tokyo, 1989
48
-------
[3] K. Ohta: Radioactive waste treatment and disposal from decommissioning of nuclear power
reactor, Genshiryoku-Kogyo 31, 29, 1985
[4] U. S. NRG: A technology, safety and costs of decommissioning a reference Boiling Water
Reactor station, NUREG/CR-0670, 1980
[5] JPDR Research Committee on Recycling .and .Reuse of Components from
Decommissioning of Nuclear Facility: Report on the recycle and reuse of components from
decommissioning of nuclear facilities, 1988 ,
[6] M. Ishikawa et al.: Present status of JPDR decommissioning program, Proc of 1987 Int.
Decommissioning Symp. Pittsburgh, 1987 „
[7] T. Hoshi et al.: Management of waste from JPDR decommissioning, ibid
[8] H. Yamamoto et al.: Experiences in decommission. and reuse of the large-scale
radiochemical laboratory and the research reactor in the Japan Atomic Energy Research
Institute, Proc. of The Residual Radioactivity and Recycling Criteria Workshop, St. Michaels,
MD. , Sept. 27-28, 1989
[9] Nuclear Safety Bureau in Science and Technology Agency: Statistics on the use of radiation
in Japan 1988, 1988
[10] The low concerning prevention from radiation hazards due to radioisotopes, etc., 1958
[11] J. H. Opelka et al.: Particle accelerator decommissioning ANL/ES-82, 1979
[12] IAEA: Radiological safety aspects of operation of electron linear accelerator, Technical
Report Series No. 188, IAEA, 1979 , , .
[13] IAEA, Radiological Safety aspects of operation of proton accelerator, Technical Report
Series No. 283. 1988
[14] Abstracts of the Workshop on the protection against the risks of ionizing radiations from
particle accelerators 1989, High Energy Physics Research Institute, 1989
[15] K. Masumoto, et al.: Residual radioactivities induced in stainless steel and some aluminum
alloy by an electron linear accelerators, Proc. of the 7th Meeting on linear accelerator, 1983
[16] H. Hirayama: Measurement of residual activities produced in various metal by KEK-PS, Proc.
of 2nd. symposium on accelerator science and technology, 1978
[17] M. G. Yalsintas: Shielding calculate at dismantled synchrocyclotron, CONF-870405-11.1987.
[18] A. B. Phillips et al: Residual radioactivity in a cyclotron and its surroundings, 1986.
49
-------
Table 1 Estimated quantities of wastes from decommissioning
of BWR and PWR
Radioactive BWR
Cocentration Metal Concrete
(Ci/t)
PWR
Metal Concrete
-3
> 10
io-4-icT3
2,700
3,240
540
1500 1,000
540 9,500
< 10
-4
30,780 502,200 30,500 457,500
Total(ton) 36,720 503,280 41,500 458,500
Source: K.Ohta,(1985)
50
-------
TABLE 2 ESTIMATED QUANTITIES OF MATERIALS CONTAINED IN A REFERENCE BWR PLANT (1155MWe)
"~ -~-^^ Mat e r i a l.s
Activity ^~^^
concentration level ^^-^^^
Radio-
active
210' Ci/t
10J ~10! Ci/t
10! ~10' Ci/t
10' ~ 10 Ci/t
10 ~10-'Ci/t
10-'~10-!Ci/t
Hr'-MO-'Ci/t
lO-'MO-'Ci/t
Activity-level
Unidentified
Sub-total
Activated
Contaminated
Activated
Contaminated
Activated
Contaminated
Activated
Contaminated
Activated
Contaminated
Activated
Contaminated
Activated
Contaminated
Activated
Contaminated
Activated
Contaminated
Activated
Contaminated
Non - radioactive
TOTAL
Equipments and Pipings
Carbon
Steel
0
0
0
0
0
69
0
.209
0
1,724
0
875
0
0
0
333
0
1.282
0
4.492
Low-
Alloy
Steel
0
0
0
0
148
0
. 0
568
0
29
0
0
0
0
0
1,330
0
220
148
2,147
Stainless
steel "
67
0 •
5
6 .
32
30
0-
509
0
519
0
0
0
0
0
0
0
1.431s1
104
2.495
Non-
• ferrous
metal "
0
0
0
0
0
0
0
0
0
1,415
0
58
0
0
0
0
0
20
0
1.491
TOTAL
67
0
5
6
180
99
0
1,286
0
3,687
0
931
0
0
0
1,663
0
2,953
252
10.625
67
11
279
3,687
931
0 .
Not available in NUREG/CR-0672 "
4.492
2,295
2.599
1,491
10,877
Structures and Build in g s
Concrete
0
0
0
0
0
0
0
: 0
0
891 .'
199
1,239
0
0
0
0
0
0
199
2,130
378, 572"
380,901
Rebar 31
(Carbon
Steel)
0
0
0
0
0
0
0
0
0
0
0
0
0
0
: 0
0 :
0
0
0
0
20,567
20, 567
Structural
Steel
0
0
0
0
0
0
66"
0
0
188"
7"
0
0
0
0
0
0
19
73
207
2.905
3.215
• TOTAL
0
0
0
0
0
0
66
0
0
1,079
206
1,239
0
0
0
0
0
•19
272
. 2,337
0
0
-. 0
66
1.079
1,445
0
0
19
2,609
402,074
404..683
TOTAL
(Equipments and Pipings +
Structures and Buildings)
67
0
5
6
180
99
66
1,286
•0
4, 766
206
2,170
0
0
0 -
1.663
0
2,972
524
.12.962
67
11
• 279
1,352 '
4,766
2.376
0
1.663
2.972
13,486
402,074
415,560
SOURCE : Report on the Recycle and Reuse of Components from Decommissioning of Nuclear Facilities 1988., JPDR Research Com!tee on Recycling.and Reuse of Components
from Decommissioning of Nuclear Facility ,
— This table is based on NUREG/CR-0672 and is supplemented with data of domestic power plants.
1) For recent BKR plants in. Japan, Inconel is used for some eqaipnents (e.g. decontamnation-.effhent concentrator : appro*. 13 tons) instead of stainless steel
2) Tnat is aluminium (for pressure-vessel-lid insulation material and pipings), titanium and copper alloy (for condensor pipings)
3) Quantities of reinforce-bar of cooling tower is not available in WEC/CR-0672. It is assumed to be lOOkg/m" (1,858 tons) using data of other buildings
4) Hade of low-alloy steel.
5) Includes charcoal .filters which consists of stainless steel and charcoal (neigh appro*. 33 tons),
6) Quantities of some equipments are not suimed up to the total.
7) Includes cooling tower which consists of concrete and asbestos (weigh appro*. 40,000 tons).
8) According to the data estimated by power generation company personnels, total mass is estimated to be 30,780 tons.
(unit : ton)
-------
Table 3' Major specifications and operation
history of JPDR.
Major Specifications : ,
Reactor _Ty_pe____BWR _
"Power JPDR-J L45_M_Wt__JPDR : H^ 90 MWt_
Reactor Pressure Vessel
Material
carbon steel internally
clad with SUS
Height 8m, Diameter 2m, _WaM_Thickness_7cm
Biological Shielding
Matericol
Thickness
Ion bars
reinforced concrete with
carbon steel liner
.1.5-3m
max. dia. 29mrn
iron bar ratio
Containment Vessel
Height 38m, Diameter 15m
'Operation History
Total Operation Time 17,000 hours
Total Output 21,500MWD(1.3FPY:
(Source : Ishikawa et al, 1987 )
Table 4 Kinds and activity of radioactive
waste generated from the JPDR
dicommissioning.
Kinds of Radioactive Waste
Activated
Components
Contaminated
Components
Core Internals
Control rods,
Core shroud, etc.
Pressure Vessel
Biological Shield
Concrete
Components
Concrete
Resin, etc.
Total
Activity (Ci)
4,500
46
12
• 4.7
0.2
0.5
4,600
Weight (Ton)
20 •
110
1,350
1,640
830
130
4,100
(Source : Hosh-i et al, 1987 )
52
-------
Table 5 Quantities of unsealded
radioisotopes used in fiscal year 1987.
(Source: Statistics on the use of radiation
in Japan, 1988) '
^"""--•^^ Category of
^""---^.Organ i za ti ons
Major Nuclldes(unit) T"*- — ^^
3H , (mCi)
"C (mCi)
"P (mCi)
"S (mCi)
«Ca (mCi)
51Cr (mCi)
"Fe (mCi)
"Ga (mCi)
"Se (mCi)
""Kr (mCi)
Generator
"Kr (mCi)
M"Tc (mCi)
Generator
""Tc (mCi)
Solution
- "'In (mCi)
" '"I (mCi)
'"I (mCi)
n'I (mCi)
'"Xe (mCi)
u'Pm (mCi)
2°'T1 (mCi)
Total
78 190
20 850
19 890
• 4 550
660
6 480
290
422 470
620
' 21 650
5 4SO
4 047 860
1 271 100
10 410
159 600
17 830
102 840
1 015 120
5 120 000
308 470
Hospitals &
Clinics
1 310
3 380
540
190
20
1 220
170
422 450
610
21 '650
—
4 044 570
1 270 240
10 320
159 580
8 560
101 490
1 014 540
—
308 410
Educational
Organizations
17 480
510
7 550
2 100
350
3 120
10
(2)
(5)
—
—
350
90
, 50
' (3)
3 390
460
—
—
(8)
Research
Institutions
42 920
14 760
11 460
2 080
250
: 1 980
80
(6)
(5)
—
60
720
220
10
- (6)
4 350
690
210
—
(2)
Industrial Firms
& Other
Organizations
16 480
2 200
340
. ' 180
40
160
30
20
(1)
—
5 420
2 220
550
30
20
1 530
200
370
5 120 000
50
Table 6 Number of radiation generators in
use (as of March 31, 1988)
^~^~---_^ Category of
— -^Organizations
Radiation Generators^"^---^^^
TOTAL
(Ratio JO
Cyclotrons
Synchrocycl Irons
Synchrotrons
!•"•
Linear Accelerators
Betatrons
Van de Graaff Accelerators
Cockcroft-Walton Accelerators
Transformer-type Accelerators
Microtrohs
TOTAL
(Ratio JO
7 4 8
(100.)
34 (4.5)
1 (0.1)
11 (1.5)
497 (66.4)
43 (5.9)
46 (6.3)
& 5 (11.6)
22 (3.0)
9 (1.2)
Hospitals and
Clinics
4 4 1
(59.0)
8
398
3 2
3 .. '
.Educational
Organizations
4 5
(6.1)
5
2
13
2 2
, 3
1
Research
Institutions
1 26
( 1 6, 8)
i i •
.• i
"•7
2 4
,2
2 7
3 9
1 4
1
Industrial
Firms
1 3 4
( 1 7 ."9)
15
4
7 0
6
' .. 8 '
2 4
5
4
• Other
Organizations
1
(0.1)
— :-•
1
SOURCE MATERIAL : STATISTICS ON THE USE OF RADIATION IN JAPAN 1988 Edited by Nuclear Safety Bureau in Science and Technology Agency
53 ' ' '
-------
Table 7 Major radionuclides commonly
identified in materials irradiated around
accelerators.
Irradiated Material
Aluminium
Concrete
Iron . Steel
Copper
Radionucli-des
High-energy electron accelerators(£35MeV)
"Na
!Na
22Na . 5iMn and 3iFe
"Na . s'Mn . S5Fe . s°Co and "Ni
Positive-ion accelerators
!Na
!Na . "Ca . !
-------
Figure 1 Trend of the installed electric
energy capacity on nuclear power plantsin•
Japan.(Source: Genshiryoku Pocketbook, 1989)
Metal
waste
Concrete
waste
-Activated (A)
-Contaminated (B)
, Activated (A)
•Contaminated (A)
Others jnfters, (A) (B)
[etc.
90 16 70'25
i i / /,
(200)
1,415 190 35
(1,640)
1,030
170 70
(1,270)
694
135 .1
(830)
80 20
(100)
0
500 ' 1000
Weight
1500 ton
Rodiooctivily
^ Level
(A):/iCi/g.
'(B)'./iCi/cm2
O
-------
4000
o>
in
3000
2000
_0
o
1000
0
to
c
o
" 11500
o
- o
*0
jhooo
"o
500
0)
JD
E
Total
I o o—O—0—0—°"
Educational
Organization
i i \ I i i i ...t._. I r 11 t t I i
'60
'65
'70 '75
•YEAR
'80
'85
Figure 3 Changes with the year in number of
users of radioisotopes in Japan. (Source:
Statistics on the use of radiation in Japan,
1988)
56
-------
Session II
Impacts of Cleanup Technologies and
Economics on Criteria
-------
-------
Limitations of Cleanup Technologies
Thomas S. LaGuardia, PE
President
TLG Engineering, Inc.
ABSTRACT
This discussion will include the limitations of cleanup technologies used in the typical
processes of decontamination and decommissioning of nuclear facilities. The key issue is to
determine the primary objective of decontamination campaigns^ namely, to achieve free release
of equipment or structures, or to reduce exposure to workers. A thorough radionuclide
characterization and contamination survey is a mandatory prerequisite to the development of a
properly planned program. The topics to be addressed are the problems associated with
performing effective decontamination without increasing LLW volume, radiation exposure or cost;
redeposition of contaminants, prevention of cross contamination, process limitations (corrosion,
temperature, pressure, application time, pH, flow rate, and waste volume), and actual effectiveness
versus vendor's reported effectiveness. Many decontamination programs have failed to meet their
objectives because of inadequate preplanning to characterize the contamination and identify the
process limitations.
INTRODUCTION
The beginnings of the nuclear era concentrated on processes. From research and
development in nuclear technology and weapons production facilities, the decontamination of
these facilities was considered a relatively minor concern. However, as the cost of radioactive
waste disposal increased rapidly and waste disposal space became scarce, the interest in
decontamination technologies grew and decontamination processes wer© developed to meet this
new demand.
Contamination sources included residual radioactivity in the mining and milling processing
of uranium and thorium, radium usage in industry for luminescent instruments, weapons production
plants, medical facilities, research laboratories and other processes. The development of nuclear
power facilities created perhaps one of the largest area for growth of decontamination technology
as the utility industry tried to clean up older plants to minimize the areas of each facility that are
considered radioactively controlled areas. The cleanup processes spawned new industries and
59
-------
products aimed at quickly and cost effectively reducing the levels of contamination to either
unrestricted use levels, or to levels permitting operating and maintenance without extensive
personnel radiological and respiratory protection. Decontamination process and equipment
vendors offered their services promising rapid application techniques and high decontamination
factors. In some cases these processes were only developed in the laboratory and never field
tested prior to their application at a contaminated facility. Not surprisingly, the results were
disappointing and in many cases the utility/vendor relationship ended in total dissatisfaction. The
primary failure in most of these cases was the lack of a planned decontamination program based
on an extensive characterization and identification of the program objectives. This paper will
address the limitations of decontamination technologies relative to the removal of residual
radioactivity and the possible recycling of materials.
FACILITY CHARACTERIZATION AND PROGRAM OBJECTIVES
The development of a decontamination program must begin with a thorough radiological
characterization of the facility to identify the radionuclides present, the concentrations in the var-
ious systems or structures, the nature of the contamination as to its inherent adherence to the
contaminated surface and the chemical composition of any corrosion product binding the con-
tamination within the corrosion layer. The characterization process is often overlooked, or per-
formed superficially without a clear understanding of the importance to the development of the
decontamination program objectives. It is important to realize that in many cases the failure of a
decontamination program is due to a chemical problem, not a radiological problem. The
management pressure and excitement of getting started and seeing progress is often weighed as
more important than carefully developing a realistic program objective that available or developed
decontamination processes can satisfactorily achieve.
The characterization pro usually requires several man-months of effort to plan and imple-
ment in the field. For internal system characterization, the system piping or components should
be accessed (opened) to obtain scraping samples and smears to completely identify the
radionuclides and concentrations with in the system. At the same location, external radiation
measurement should be made to correlate external exposure readings to internal contamination.
By this technique the extent of contamination can be readily estimated using external survey
techniques to bound the scope of the decontamination project. For external contamination on
structures, concrete cores or steel scrapings should be taken and analyzed for radionuclide
content and concentration. This data can be correlated to external survey measurements to
determine the extent of contamination on structure surfaces.
Once this characterization data has been analyzed radiologically, chemically and
mechanically (for adherence to the surface), a realistic program objective can be developed. The
management decision can then be made whether it is feasible to achieve complete
decontamination to release the material for unrestricted use and disposal in local landfills or
recycling, or whether to reduce radioactivity levels to minimize exposure to workers performing
operation or maintenance activities.
There have been far too many decontamination and decommissioning projects where
decontamination projects were started based on random and inadequate characterization data.
The project had to be stopped, re-characterized, planned and re-started at additional cost never
factored in to the original budget.
60
-------
DECONTAMINATION PROCESSES
In response to the need for effective decontamination processes, several techniques have
been developed for specific decontamination tasks. A brief discussion of these techniques is
included here as examples of typical methods currently being used. These same examples will
serve as focal points to discuss limitations in these and other processes from lessons learned in
their application.
High Pressure Water Lancing
This technique directs water through a lance with a jei or fan nozzle at pressures of about
2,000 to as high as 40,000 psi. The erosion action removes surface contamination from steel and
concrete surfaces. The process uses 3 to 5 gpm at about 2,000 psi, and about 5 to 10 gpm at
40,000 psi. The operator reaction forces above 10,000 psi are about 60 pounds or more, and
result in operator fatigue for prolonged use. At higher pressures the lance cannot be hand-held,
and remote/manual or automatic machine must be used. This technique is often used to reduce
overall contamination levels on contaminated surfaces rather than to achieve free release of the
material.
While this process is very effective in local surface decontamination, there are several draw-
backs that limit its usefulness. Operator fatigue was already mentioned. The copious quantities
of water used result in a significant radioactive waste management problem. The rapid generation
of wastes requires a large waste processing system and associated operators to dispose of the
wastes. These costs are often not considered in evaluating the effectiveness of this technique.
Portable water filtration units are available to process the water and removed contamination, and
to recycle the water through the lance. For lance pressures up to a few thousand pounds
pressure, the filtration unit consist of cartridge filters in a pressure housing, the filtration capacity
is sufficient to remove steel and concrete particulate, and to recycle the water through the pump
system successfully. However, at higher pressures and particularly at pressures as high as 40,000
psi, ultra-filters are needed to remove aJJ particulate so as to protect the pump cylinder surfaces
from scoring and damage. Ultra-filters cost about $20,000 to purchase, and operating costs are
also high. Pump repair is cost and time consuming adding to the overall cost of the
decontamination project.
Shot or Grit Blasting
This technique uses high pressure air mixed with steel shot or abrasive grit directed at the
work piece to abrade the surface and remove the contamination layer. An integral vacuum system
connected to the blaster head collects the shot or grit for separation by cyclonic action. The
heavier shot is recycled and reused, and the lighter particulate directed to a bag filter for disposal.
This technique is effective for the removal of paint and thin layers of contamination from
steel or concrete surfaces. If the total area to be decontaminated is relatively small (a few hundred
square feet) the volumetric amount of contaminated shot or grit wastes is high for the area being
treated. The process is relatively slow and labor intensive because of the small size of the blaster
head to needed to be able to effectively vacuum the spent shot back into the recovery hose. If
the work is being performed in a high radiation area, the worker exposure from adjacent
radioactive sources may be the controlling factor. If the contamination has penetrated more than
61
-------
an eighth of an inch or more into the surface this process would not be cost effective for achieving
free release of the surface.
Concrete Scarification
Concrete scarification (removal of one-quarter inch thick layers of the surface) is performed
using a pneumatically operated multiple piston / bit head moved over the contaminated surface.
The head removes up to one-quarter of an inch of surface per pass. The chips and dust gener-
ated in the process are captured within a vacuum shroud surrounding the head and directed to
a HEPA vacuum system for disposal. The technique is very fast and effective in surface removal,
and can remove up to about one inch of thickness by repeated passes. Beyond one inch
thickness, alternative techniques would have to be used. The scarification process (also called
scabbling) is ideally suited for concrete because most contamination rarely penetrates more than
about one-half inch in depth into concrete. Accordingly, complete removal of contamination is
possible permitting free release of the surface.
The process works well on floors because the chips remain within the shroud and can be
readily vacuumed up into a disposal drum. However, vertical surfaces is more difficult because
the heavier chips and dust tends to fall away from the shroud and further contaminating adjacent
areas. Very high vacuum systems and a tight shroud help to minimize this problem, but on large
projects the dust buildup in adjacent areas can become a problem.
Chemical Decontamination
The interior of process systems can be decontaminated using various pretreatment and
chemical solvents to remove the contaminant or the corrosion layer that may contain the con-
taminant. Numerous solvents have been developed for specific contamination films on carbon and
stainless steel surfaces with good overall success. The types of solvents are beyond the scope
of this paper, but are readily available in the literature. The process generally consists of flushing
a pretreatment chemical throughout the system to prepare the corrosion film for removal. A
subsequent flush of a mild or aggressive acid or caustic chemical removes all or most of the
corrosion layer. If the system or component is to be reused, a passivation flush may also be used
to chemically neutralize the surface and provide a protective film.
These processes have performed well when the corrosion film was adequately
characterized and the particular solvent tested both in the laboratory and in pilot testing on similar
surfaces. When these two critical steps are not taken the results are usually disappointing
because off-the-shelf processes or chemicals and concentrations are not always suitable under
all conditions. The processes generally require controlled temperatures, pH levels, concentrations
and flow rates to effectively remove the contamination film. Organic based solvents tend to break
down at high temperatures into their constituent elements and lose their effectiveness. As the
solvent action begins to dissolve the film, the solvent concentration changes and the dissolution
rate diminishes. In components of the system where, the solvent fluid velocity is reduced such as
within tanks, system traps and piping deadlegs, the scrubbing action of the fluid to remove the
dissolved corrosion film is great diminished. In these regions the dissolved contaminants may
precipitate out of solution and redeposit on the surface. Not only does the solvent fail to perform
its intended function, but may create a new "hot spot" of radioactive contamination which may be
worse than the original condition that the solvent flush was supposed to correct.
62
-------
The processing of the waste generated from chemical solvents is often overlooked as part
of the cost of decontamination. Organic solvents can be removed demineralization after filtration
of the particulate flushed along with the fluid stream. The demineralization deposits the dissolved
contaminants on the demineralization beds and the resulting "clean" effluent stream disposed to
an existing waste processing system for further treatment, or discharged as clean water if residual
concentrations are within specific disposal limits. However, the spent demineralizer beds are now
highly contaminated with the removed contamination from the process piping and must be
disposed of in accordance with normal plant procedures. The waste volume of the packaged bed
usually far exceeds the volume of radioactive waste removed from the process piping. This bed
disposal cost is often not recognized as an additional cost of processing. Nor is the additional
worker exposure incurred in disposing of these beds included in the estimated exposure involved
in the decontamination program.
Many chemical solvent processes are touted as being capable of decontaminating to
unrestricted use levels using single or multiple flushes. Such claims should be evaluated carefully
and the source of data checked for its relevance to the specific process system to be decon-
taminated. As noted earlier, all solvent processes should be laboratory tested and pilot tested in
the field before embarking on a major decontamination program. A few dollars invested on the
front end may save thousands or even millions of dollars on the tail end.
CONCLUSIONS
Cleanup technologies are available today that are effective in achieving high
decontamination factors for specific applications. The criteria to be met for unrestricted use of
decontaminated materials must be clearly understood so that a realistic program objective can be
adopted. Prior to embarking on any decontamination effort, a detailed radioactive contamination
characterization should be made to identify the type and concentration of contamination present
in the system or structure to be removed. An inadequate characterization will most surely lead to
disappointing results in the decontamination process applied. The program planner must take a
hard look at proposed decontamination techniques to determine if they are feasible for the task
at hand, and should verify vendor claims by contacting previous users of their success or failures
with the technique. The overall decontamination program costs, exposure and waste volume need
to be evaluated along with the projected decontamination factor when assessing the cost
effectiveness of proposed process. As expected, actual field experience is generally more reliable
than laboratory tests or vendor claims. The program planner heeds to ensure that the proposed
technique will perform the intended task. Nevertheless, as waste disposal costs continue to
increase at escalation rates far exceeding the national inflation rate, reliable decontamination
techniques are needed to minimize the volume of waste going to controlled burial grounds.
63
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Decontamination Technology for
Decommissioning of Nuclear Facilities
Hideo Yasunaka, Tamotsu Kozaki, Takeo Gorai
Department of Japan Power Demonstration Reactor
Japan Atomic Energy Research Institute
ABSTRACT
During the decommissioning of a nuclear reactor facility, appropriate decontamination
before and after dismantling of the facility is required for each stage of the process. As part of
the technology development for the Japan Power Demonstration Reactor (JPDR)
decommissioning program, several new decontamination methods have been developed at
JAERI.
This paper describes the decontamination methods for the decommissioning and
discusses their application for the reuse of dismantled components on the basis of
decontamination experience at JPDR.
INTRODUCTION , ' .
A nuclear reactor facility can be decommissioned in several ways: dismantling early after
shutdown, dismantling after mothballing, or entombment. In Japan, early dismantling is
necessary due to limited land resources for nuclear installations and, any site has to be
renovated in preparation for the next nuclear installation.
During decommissioning, appropriate decontamination is required for each stage of the
dismantling, not only to reduce occupational exposure but also to reduce decommissioning
waste. In Japan, no large scale decontamination has been performed, except in the JPDR, but
some small scale decontaminations have been executed: Therefore, data obtained from the
JPDR decontamination activities, such as decontamination of systems, dismantled components,
and concrete buildings, are valuable.
This paper describes the decontamination methods for decommissioning, including the
decontamination results for JPDR, and the decontamination techniques for reuse of dismantled
components.
64
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CLASSIFICATION OF DECONTAMINATION TECHNIQUE
The decontamination process may be divided into three major parts, as shown in Table 1:
decontamination before and after dismantling, and decontamination of the facility building. Each
part is described below.
Decontamination Before Dismantling
Decontamination Method:
Decontamination before dismantling is used to reduce the dose rate in the working area
to decrease occupational exposure. The process can be divided into two major parts:
decontamination of the primary coolant system, and of the large components, such as the
radwaste storage tank.
In general, decontamination before dismantling does not need a high decontamination
factor (DF, the ratio of activity before to that after decontamination), because its main purpose
is to reduce occupational exposure. (On the other hand, a high treatment efficiency of liquid
waste is required, in addition to a large throughput, such as can be provided by a cartridge filter
and ion exchange resin, because the object to be decontaminated has a large volume, and
produces a large amount of secondary waste.)
The decontamination before dismantling can be carried out by the means of the following
methods:
Chemical methods
- Reduction
- Oxidation
- Redox reaction
Mechanical methods
- Abrasives
- Hydro pressure
- Blasting
The method to be used in a particular case is selected according to the shape of the object and
the characteristics of its contamination. In general, chemical methods tend to be applied to the
system, while mechanical methods tend to be applied to the decontamination of the larger
components.
Characteristic of CRUD, and the Effect of Chemical Decontamination:
The decontamination efficiency is affected by the amount of cromium (Cr) in the CRUD
due to the CRUD's dissolution mechanism. Table 2 shows the representative system
decontamination methods and conditions. Fig. 1 shows the examination of the decontamination
methods used at JPDR, using sample specimens from the JPDR and two commercial power
plants (PWR and BWR).
The test results indicated that reducing decontamination methods (No. 1 to 5 in Table 2)
were not effective for high Cr-containing CRUD samples (more than 15% Crj, while a
decontamination factor (DF) of 10 was obtained in the case of low Cr-containing CRUD samples
(less than 12 % of Cr), as shown in Fig. I. Furthermore, the results also indicated that the
65
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addition of an NP or AP pre-oxidation step to the reducing decontamination step further improved
the decontamination efficiency for any Cr-containing CRUD (See No. 6 to 8 in Table 2 and Fig. 1).
Decontamination After Dismantling
Decontamination after dismantling intends not only to recycle the dismantled metal
components, but also to reduce the volume of radioactive waste. Most metal components are
pipes, valves, pumps and tanks consisting of stainless steel, carbon steel and aluminum. This
decontamination is carried out by removing the contamination with a base metal from the surface
during dissolution or abrasion processes, because the decontamination after dismantling requires
a high DF. However, there is no decontamination method which can apply to all kinds of
dismantled components, since the dismantled components have various shapes and materials,
as mentioned above. A few decontamination methods, therefore, are needed to treat them.
Representative decontamination methods of dismantled components are as follows.
Electropolishing Decontamination Method:
An electropolishing decontamination technique has been developed [1]-[3] for polishing
metal surfaces. In general, the object to be decontaminated is set as the anode in an electrolyte.
Electric current is then supplied to the anode and a cathode, thus promoting the anodic
dissolution of the metal surface material. Oxidation films and surface contamination, including
the deposited CRUD, are removed in this surface dissolution process. On the other hand, when
the object to be decontaminated is used as the cathode, the bubbling by hydrogen gas which
appears on the object's surface also promotes the removal of the deposited material.
Electropolishing decontamination is divided into a few methods, as shown in Table 3. Moreover,
Table 4 gives the characteristics of the electrolyte: phosphoric acid, sulfuric acid and sodium
sulfate (neutral salt).
This method is adequate for simple-shape components and is expected to attain a
considerably high DF. ;
Chemical Immersion Decontamination Method:
In general, the chemical immersion decontamination method utilizes a chemical bath, and
treats the object by two steps: pre-oxidation and reduction. This method uses many chemical
solutions, depending upon the material of the objects. However, a system decontamination
tends to become relatively complex, since this method uses two steps.
Also, there are several decontamination methods which need just one step of the redox
reaction [4]. .
The chemical immersion decontamination method is adequate for complex shape
components. Furthermore, simultaneous use of ultrasonic waves improves the decontamination
efficiency, especially for objects with small crevices and holes.
66
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Blasting Decontamination Method [5],[6]:
The blasting method decontaminates objects through the action of blasting abrasives,
which are jetted from a nozzle with a fluid (air, water, or other solvents). Sand is an example of
an abrasive, as is alumina, steel, boron carbide, silicon carbide and glass. This method uses a
large amount of abrasives and fluid, so that separation of the abrasives and the fluid from
contaminating slugs, and their reuse, are required to reduce the quantities of secondary waste.
A fajriy high DF is expected of this method if carried out under appropriate conditions,
which usually depends upon the object's material and the abrasives.
Decontamination of Buildings
A building is decontaminated not only to allow unrestricted use, but also to reduce the
volume of concrete waste. The removal of only the concrete surfaces is sufficient for concrete
decontamination, because contamination with Co-60 is limited to a few cm in depth.
Thermal stress methods such as a microwave irradiation, flame scarfing, and mechanical
methods, such as milling cutter, scabbier, drill & spoiler and blasting, are useful ways to
decontaminate. The microwave irradiation method is especially good for decontaminating
concrete buildings, because it does not make direct contact with the object being cleaned, and
thus prevents the contamination from spreading. Also, it is not necessary to apply much
mechanical force against the object (especially important for walls and ceilings).
DEMONSTRATION OF DECONTAMINATION METHOD IN JPDR
As part of the technology development for the JPDR decommissioning program at JAERI,
a number of decontamination methods have been developed to be applied to the system before
dismantling, to concrete surfaces, and to the dismantled JPDR components.
Several methods for system decontamination have already been demonstrated with the
primary coolant system of the JPDR. In addition, the electropolishing decontamination technique
for the dismantled components has started. And the decontamination systems for chemical
immersion and blasting are planned to be manufactured in 1989 and 1990, respectively.
The results of the system decontamination and the electropolishing decontamination
carried out in the JPDR are described below.
Decontamination Before Dismantling
Elemental Analysis of CRUD:
Results of elemental analysis of CRUD samples from the JPDR are shown in Table 5, in
addition to samples from the commercial nuclear power plants (BWR and PWR).
The CRUD in the JPDR contains 15 to 25% Cr, while most CRUD in the BWRs contains
less than 10% Cr.
67
-------
Development of New System Decontamination Method [7]:
The redox decontamination method using sulfuric acid-cerium(IV) (the SO method, No.
9 in Table 2) and the flowing abrasive decontamination method (No. 10 in Table 2) have been
developed at JAERI to overcome the Cr-rich CRUD problem as mentioned above.. The SC
method utilizing the redox reaction of Ce (IV) in a sulfuric acid solution is very effective in spite
of the fact that it is a single-step decontamination method. The flowing abrasive decontamination
method is independent of the chemical properties of CRUD, since this method utilizes only the
mechanical attributes of abrasives. When grains of abrasives are forced into suspension in the
water by circulation, the grains hit or rub the inner surfaces of pipes and components, thus
removing the CRUD mechanically. The DF obtained by this method can be adjusted by the flow
rate and the decontamination time. Figure 1 shows the test results after 12 hours of
decontamination at a flow rate of 4.5 m/sec.
Demonstration of System Decontamination:
Four system decontamination methods have been demonstrated at the four parts of the
JPDR primary coolant system, as shown in Fig. 2. The results by each decontamination method
are shown in Table 6, along with the decontamination conditions.
(A) CAN-DECON Method
The CAN-DECON method was performed under normal conditions at the reactor's water
clean-up line. This was done not only to get experience, but also to investigate unexpected
problems that might arise during system decontamination. The maximum DF of 90 was obtained
only at the regenerative heat exchanger, while the DF was 3 to 11 at the other parts of the line.
(B) Modified NP/NS-1
The modified NP/NS-1 decontamination method adds the NP pre-oxidation phase before
the reducing decontamination phase. The addition of the pre-oxidation phase can considerably
improve the decontamination efficiency even for high Cr-containing CRUD. In general, a 2-phase
decontamination method needs a large amount of solution for each phase, and the solution is
drained after each phase. For this reason, a liquid waste treatment system using a reverse
osmosis (RO) module was introduced. Reuse of the water treated by the system can
considerably reduce the total amount of water needed for this method. And, the liquid waste
resulting from this method was a concentrated solution, which is less than 10% in volume of the
total liquid used. The DF attained in this demonstration ranged from 90 to 740, and averaged
about 500.
(C) Redox Decontamination Method
The redox decontamination method [7] using sulfuric acid and Cerium (IV) involves two
processes: decontamination and liquid waste treatment. In the decontamination process [8],
electrolytic regeneration of Ce (IV) maintains the constant concentration of Ce (IV), which controls
the dissolution rate of the CRUD. The electrolytic regeneration keeps the ratio of Ce (IV) to Ce
(III) less than 1. In the liquid waste treatment process, an electrodialysis is performed to collect
metal ions and sulfuric acid from the liquid waste after their electrolytic reduction. The liquid
68
-------
waste, containing few ions and little sulfuric acid, is further cleaned by an ion exchange resin.
A DF of 300 to 1200 was attained, while the average DF was about 900 in the demonstration test
of the JPDR primary coolant system [9].
(D) Flowing Abrasive Decontamination Method
The flowing abrasive decontamination method can be used at room temperature.
Moreover, only an abrasive collector and a cartridge filter are required in the liquid waste
treatment process, since most of the removed CRUD exists in the form of suspended solution.
Therefore, a decontamination loop can be made very simple. The demonstration test of this
method showed a DF of 200 to 1660, while average DF was about 1,100 [7].
Electropolishinq Decontamination
Condition of Electropolishing and its Results:
From the results of the basic examination of eleciropolishing, optimum decontamination
conditions for each electrolyte were determined to be as follows.
(A) Decontamination Condition for each Electrolyte
ELECTROLYTE CONCENTRATION CURRENT DENSITY TEMPERATURE
phosphoric acid
sulfuric acid
Neutral salt
(Sodium sulfate)
80 wt%
10 wt%
20 wt%
0.2 A/cm'
0.2 A/cm2
0.4 A/cm2
60°C
60°C
60°C
(B) Optimum Cycle for Alternating Current Decontamination Method
Cathodic and an anodic electrolysis times are both 30 sec.
Electropolishing decontamination was performed under the above conditions using
samples from JPDR and commercial power plants. Figure 3 shows the decontamination results
for JPDR's primary coolant system pipes (SUS 304) for each decontamination method. The
anodic decontamination method could rapidly decontaminate the samples in every electrolyte,
and a DF of more than 5000 was obtained. The results also showed that the decontamination
time required for the alternating current (a.c.) method was twice that of the anodic method, when
sodium sulfate solution was used as an electrolyte. However, sufficient results were attained by
every method after 20 minutes of decontamination. Figure 4 indicates the decontamination
results for the RTD sample (SUS 316) of the commercial power plant (PWR) by each electrolyte.
All samples contaminated with more than 105 Bq/cm2 were perfectly decontaminated within 20
minutes by each method, and a DF of about 10,000 was attained.
The Factors of Electropolishing Decontamination Method:
If electropolishing can be applied for a long time, it can bring about complete
decontamination. This method dissolves the contaminated metal surface. The decontamination
69
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time, however, is one of the economical factors which affects the decontamination cost. In the
case of decontaminating dismantled pipes or components, the time needed for electropolishing
depends on the following factors: material, shape, extent of contamination, and degree of
corrosion. These factors are described in detail below.
(A) Material
In electropolishing samples of stainless steel, a DF of more than 5000 was attained in a
short-time decontamination, as shown in Fig. 3 and Fig. 4. Some carbon steel samples, however,
could not be decontaminated, or else required very long decontamination times, even under the
same conditions as stainless steel, Fig. 5. In general, carbon steel has thick CRUD layers on
almost all surfaces, and corrodes non-uniformly. Though relatively good decontamination results
were attained when 10 wt% sulfuric acid was used as an electrolyte, a fairly long decontamination
time was required with phosphoric acid or sodium sulfate. However, the alternating current
decontamination method was effective even if the electrolyte was sulfuric acid or sodium sulfate.
(B) Shape of Object to be Decontaminated
Electropoiishing decontamination is very effective for simple components, such as pipes
and boards, but it is sometimes difficult with complex shaped components, such as valves and
pumps. The use of internal cathodes which fit to the shape of a component, however, can
improve the decontamination efficiency. Figure 6 shows the decontamination results with the
anodic method, using two internal cathodes for a valve casing. This valve casing was immersed
in the electrolyte after setting two internal cathodes on sides A and B, Fig. 6. Contamination on
side B was removed to below the lower limit of detection after 20 minutes of decontamination.
Contamination remained in side A even after 30 minutes of decontamination. Figure 6 suggests
that further decontamination was not effective for side A, and also that the anodic electrolysis
method was not adequate for narrow or complex parts such as side A. Therefore, the electrolytic
abrasive decontamination method was applied to remove the contamination that remained in side
A. This method uses a rotary cathode made of a sponge with abrasives, located at the top of
the pipe-shaft of a drill device. In addition, the rotary cathode is supplied with an electrolyte
through the pipe-shaft. Consequently, decontamination can be achieved through both the
mechanical rotation of the cathode and the electrochemical function of electropolishing. Figure
6 shows the results of the decontamination by this method, indicating that it could remove the
remaining contamination in under two minutes.
(C) Effect of a Weld
Figure 7 shows the decontamination results of electropolishing for pipes with welded
parts. Welding changes the characteristic of the base metal, thus leading to differences in
thickness, characteristics, and form of the CRUD. For this reason, decontamination by
electropolishing tends to become slightly difficult.
(D) Effect of a Small Hole or Crevice
Electropolishing decontamination of components with small holes or crevices is difficult,
because the electric current density becomes non-uniform. The internal cathodes or the
70
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electrolytic-abrasive decontamination method is recommended, following decontamination by the
immersion method.
(E) Effect of Edges Cut by the Thermal Method
An edge cut by the thermal method (such as plasma-arc), has contamination in the base
metal, which melted at the moment of cutting. In addition, the thermal cutting method spatters
melted metal, including contamination, near the cutting point. For this reason, the contamination
remains at the melting parts and the spatters, and a high DF can not be attained.
Decontamination of a Concrete Building
The microwave irradiation method had been developed in the JPDR to demolish concrete
surfaces to a depth of a few centimeters [10], [11]. An examination was performed using a
prototype microwave irradiation machine. In addition, existing mechanical concrete
decontamination methods such as milling cutter, scabbier, blasting, drill & spoiler, and
hand-breaker were also examined. The milling cutter [12] and the scabbier were adequate for
the removal of concrete surfaces to a depth of 3 to 5 mm. The microwave irradiation method is
adequate in removing concrete surfaces to a depth of a few cm. This is important for walls and
ceilings, because the method does not involve applying force against the object.
A demonstration machine has been manufactured, and a test of cold samples has been
performed. This machine is planned to be applied to the JPDR concrete building in the
controlled area from 1990 to 1991.
DISCUSSION
Demonstrations of decontamination methods at JPDR showed that the dismantled
simple-shape components contaminated at about 3.7 x 102 Bq/cm2 (1.0 x 102 |iCi/cm2) could
be decontaminated to below the lower limit of detection. In addition, examination of the system
and of the electropolishing decontamination of commercial power plant samples indicated that
the components of a commercial power plant could also be decontaminated to below the lower
limit of detection.
However, measuring decontaminated components becomes very hard when they are
decontaminated to a very low level. Because the measurement area has a relatively high dose
rate, in general, a shield is required to enclose the components. In addition, most components
to be decontaminated such as pipes, valves, and pump casings range from a few 10 kg to a few
100 kg in weight, but their plane surfaces are not large. A detector having a large detection-area,
therefore^can not be used with them.
At JPDR, a GM survey meter with a detection area of 20 cm2 is used mainly for measuring
decontaminated components. A very long time period is needed for these measurements. For
example, a GM survey meter used to measure a decontaminated pipe 400 mm in diameter and
700 mm in length requires 100 minutes under the following condition:
71
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Background : 80 cpm
Measurement time for each point: 10 sec
Detection area of the detector: 20 cm2
Time for change the position of the detector: 5 sec/point
Low limit of detection : 1.4 x 10~1 Bq/cm2
(aSBxIO
Number of workers required : 1 person
In this case, the electropolishing decontamination was finished within 30 minutes (during
which the electric current was supplied), and no contamination could be detected afterwards.
Most components produced by decommissioning can be completely decontaminated if
factors such as cost, occupational exposure, and secondary waste can be ignored. However,
these factors are not negligible when the decontamination for reuse is undertaken.
CONCLUSION
The results of the system and electropolishing decontamination studies indicate complete
decontamination for reuse is possible.
For system decontamination, a DF of about 50 provides sufficient reduction of the dose
rate (less than 30 mR/hr in the working area). However, the reuse or recycling of dismantled
components requires a higher DF of about 200. The results of the system decontamination
showed that an average DF of more than 500 could be attained if both the decontamination
method and conditions were appropriately selected according to the Cr content in the CRUD.
For the electropolishing decontamination, it was demonstrated that a high DF (more than
104) was obtained by the anodic electrolysis method for both stainless and carbon steal
simple-shaped components. Also, the complex-shaped components could be decontaminated
either by anodic electrolysis with an internal cathode or by electrolytic-abrasive decontamination
methods.
However, cost and occupational exposure are also important considerations in selecting
the decontamination method for reuse. The cost of contamination measurements of
decontaminated components, in particular, becomes important in the case of reuse.
REFERENCES
[1] E. L Childs, J. L. Long: Electrolyte Decontamination of Stainless Steel Using a Basic
Electrolyte, Nucl. Tech., 54 Aug. *1981
[2] H. W. Arrowsmith, et al.: Demonstration of Alternative Decontamination Technique at Three
Mile Island, PNL-SA-8143 (1979)
[3] R. P. Allen, et al.: Electropolishing as Decontamination Process, Progress and
Applications, PNL-SA-6858 (1978)
72
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[4] R. E. Lerch: Corrosion of Stainless-Steel in Solution of Cerium(IV)-Nitric Acid,
BNWL-cc-1646 (1968)
[5] J. F. Remark: Plant Decontamination Methods Review, EPRI-NP-1168 (1981)
[6] J. Sejvar and P. H. Dawson: Evaluation of Abrasive-Grit-High-Pressure Water
Decontamination, EPRI-NP-2691 (Oct. 1982)
[7] E. Tachikawa, H. Yasunaka, T. Suwa, T. Gorai, M. Kawasaki: Research and Development
on LWR System Decontamination. Mechanochemical-and RedoxDecontamination Method,
1988 JAIF Int. Conf. on Water Chem. in Nuclear Power Plants, Proceedings Vol. 2, pp.
443-448, Tokyo, 1988.
[8] T. Suwa, N. Kuribayahsi, E. Tachikawa: Development of Chemical Decontamination
Process with Sulfuric Acid-Cerium (IV) for Decommissioning. Single stop Process to
Dissolve chromium-rich Oxides, J. Nucl. Sci. Technol., 23, 622 (1986)
[9] T. Suwa, N. Kuribayahsi, E. Tachikawa: Development of Chemical Decontamination
Process with Sulfuric Acid-Cerium (IV) for Decommissioning. System Decontamination
Process with Electrolytic Regeneration of Ce (IV) from Ce (III), J. Nucl. Sci. Technol., 25,
574 (1988)
[10] H. Yasunaka, et al.: Microwave Irradiation Technology for Contaminated Concrete Surface
Removal. Demolition and Reuse of Concrete and Masonry, vol. I, pp. 280-289, Chapman
and Hall (London), 1988
[11] H. Yasunaka et al.: Microwave Decontaminator for Concrete Surface Decontamination in
JPDR, proceedings of 1987 International Decommissioning Symposium, Pittsburgh, Pa.
vol. 2, pp. IV-109-116, 1987
[12] N. Funakawa, T. Kinoshita and T. Tanaka; New Method for Decontamination of Concrete
with Milling Cutter, Proceedings of 1987 International Decommissioning Symposium,
Pittsburgh, Pa. vol. 2, pp. IV-217-229, 1987
73
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Table 1 Decontamination for Decommissioning
Decontamination before dismantling
Reduction of occupational exp osure
•Pipe line system decontamination—[^ical Method
'-Mechanical Method
'-Pool, Tank-
Hydro jet Method
Blast Method
Strippobfe coating Method, etc
o Decontamination after dismantling —
• Recycle of metal waste
•Reduction of radioactive waste
•Pipes, components
- Sectropolishing Method
-Chemical Immersion Method
Blast Method
-Ultrasonic Wave Method
LGel Method
Decontamination of Building
Concrete Surface
Unrestricted release of building
• Reduction of radioactive concrete waste
r Mechanical Method
• Scabbier
• Drill & Spoiler
• Steel Grit blast
L Thermal Stress Method
• Microwave irradiation
• Flame scarfing
Table 2 Decontamination Method tested in Basic Examination
No,
\
2
3
4
5
6
7
8
9
to
Name of Decon.
NS-Udil)
CAN- DECON
EBARA-DECON
mi- DECON
NS-1
NP/NS-1
(dil)
NP/CAN-DECON
AP/Citrox
SC Method
Rowing Abrasive
Decontamination condition
Decontamination Agents
NS- 1
LND- 101 A
ED -40 (Modified)
KD-203
NS-1 (cone.)
NP
Reduction
Decon.
NP
Reduction
Decon
AP
Ci
KMnO«
HNOj
Oxalic acid
NS- 1
KMnO*
HNOj
Oxalic acid
LND-101A
KMn04
NaOH
Oxalic acid
Ammonium Citrate., Dibasic
Iron Nitrate
Diethylthiourea
Sulfuric acid
Cerium (IV) Sulf ate
Ce (IV)
B4 C( Abrasives)
Cone.
0.7 wt%
Q1 wt%
2.0 wt%
0.1 wt%
7.0 wt%
1 g/t
5 g/t
1.42 g/t
0.7 wt%
1 g/t
5 g/t
1.42g/^
0.1 wt%
32 g/£
105 g/l
25g/t
5i^i
£Q/i
Ig//
14.5g//
Z06g//
(4.8 mmol )
20wt%
Temp.CC)
120
120
120
120
120
120
60
120
120
60
120
105
85
80
Room temp.
Time (hr)
24
24
24
24
100
6
0.75
24
6
0.75
24
4
48
48
12
Classification
Reducing
Decontamination
Pre- oxidation
+
Reducing
Decontamination
Redox
decontamination
Mechanical Method
Memo
Corrosive
Corrosive
Corrosive
Corrosive
,
74
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Table 3 Electropolishing Decontamination
o Immersion System
-Anodic Electrolysis Decontamination Method
Cathodic Electrolysis Decontamination Method
-Alternating Current Electrolysis Decontamination Method
0 Electrolytic-Abrasive
Decontamination System
In - Situ Technique
Pumped Stream Method
Contact Method
Table 4 Characteristics of Electrolyte
Electrolyte
Phosphoric acid
70-80%
Sulfuric acid
5- 20%
Neutral salt
20%
Surface finishing
bright surface
rough etched surface
rough etched surface
Solution life
50~70g/.2
(metal ion cone.)
20~30g/Jg
(metal ion cone.)
Long
Carry ing -out
Large
(Im/ /cm2)
Small
(0.5mf/cm2)
Small
«0.5m£/cm?)
Rinse
Difficult
Easy
Easy
Purpose
Reuse
Recycle
Recycle
75
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Table 5 Chemical Composition and Activity of CRUD
Type of Reactor
Commercial Power Plant
(PWR)
Commercial Power Plant
(BWR)
JPDR
J P D R
Part
RTD, 2-"*1
Reactor Water Clean-up Line
Reactor Water Clean-up Line
Recirculation Line x
Amount of CRUD
690 fig/cm2
990 *
714 r
400 «
Composition (%}
Fe
59.7
75.0
47.3
64.4
Cr
22.1
10.2
24.4
12.2
Ni
16.3
12.8
27.9
21.2
Co
0.29
0.32
0.01
0.34
Radioactivity
5.8 [iC\ /cm2
(2.1x105 Bq/cm2:
4.8 /iCi /cm2
(I.8xl05 Bq/cm2)
0.062 /»Ci/cm2
(a3xl03 Bq/cm2)
0.025 /iCi /cm2
(a3xl02 Bq/cm2)
Table 6 Condition & Results of System Decontamination Methods
applied to the JPDR Primary Coolant System
Decontamination method
Condition
Regents
Concentration
Temperature
Decontamination Time
Row rate
Regeneration of Decon.solution
Treotment of Waste Solution
Result
Decontamination Factor
(DF)
Form of Waste
CAN-DECON
Reducing Decon.
LND-tOtA
0.1 wt %
~120°C
24 hr
Cation resin •
Mixed bed resin
3 ~ 90
av.9
Cartridge Filter
Resin
NP / NS- 1 modified
Pre-oxidation
/HNOj V
NP VR.Mno]
06wto/70.5vrt%
U-bWI/0lo.1wtV.
~120°C
6 ;hr
Redua'ng Decon.
NS-1
0.7 wt%
~120eC
24 hr
non
Reverse Osmosis
90 - 740
av. 520
Condensed Waste Solution
(15% of system volume)
RO Module
Sulfuric acid-Ce(TV)
{ SC Method )
Redox Decon.
H2S04 + Ce4t
0.25 M 2mM
70~80°C
106 hr
Electrolytic reduction
Electrodialysis
Mixed bed resin
300 ~ 1200
av. 900
Condensed Waste Solution
Regenerated Sulfuric Acid
Ion Exchange Membranes
Cartridge Rlter
Resin
Rowing Abrasive
Mechanical Decon.
B4C Particles
20 wt %
Room Temp.
35 hr
4.8 ~ 6.7 m/sec.
non
Cartridge Rlter
200 - 1660
av. 1100
Abrasives
Sludge
Cartridge Filter
76
-------
15 20
Cr contents (%)
25
-»-LND101A -^- NP-LND101A
-o- NS-1 (0.7 wt%) -*r- NP-NS- 1
-v- KD-203 -x- AP-Citrox
-^- ED- 40 (Modified)
-a- NS-1(7wt%)
SC Method
Rowing Abrasive
Fig. 1 Relation between Cr-contents.and Decontamination Factor in case
of Basic Examination by Various Decontamination Methods
77
-------
steam
oo
emergency
condenser
Ce(IV)-H2S04 REOOX
DECONTAMINATION METHOD
; OF'.300-"1200
ov. 900
MODIFIED CONCENTRATED
DECONTAMINATION METHOD
(NP/NS-I )
OF! 90*740
av. 520
clean-up line
FLOWING ABRASIVE
DECONTAMINATION METHOD
DF : 200-1660
OV. 1100
DILUTE CHEMICAL /
DECONTAMINATION
METHOD (CAN- DECON)
DF: 3~90
ov.9
regenerative
heat exchanger
generator
demlneralizer
non-regenerative
heat exchanger •
demineralizer
Fig. 2 Decontamination Methods applied to the JPDR Pimary Coolant System
-------
(Bq/cm2) (/iCi/cm2)
- lO'1
Anodic Electrolysis
10wt% H2S04
80wt% H3P04
20 wt% Na2S04
Alternating Current Electrolysis
—*-- 10 wt% H2S04
—*-- 20 wt% Na,S04
10 20
Time (mm')
(Bq/crn2)
,, to1
1
Electrolyte
- IOwt%H2S04
- 80wt% H3P04
|- 20wt% Na2S04
o
o
a>
o
10* -
10
5 10 15
Time (min)
10 §
100
20
Sample
Decon. Method
Material
Temp, of Electrolyte
Current Density
RTD of Commercial power plant (PWR)
Anodic Electrolysis with Immersion System
sus-316 ;
02 A/cm2
Fig. 3 Results of Basic Examination of Electropolishing
using Stainless .Steel Samples of the JPDR
Fig. 4- Results of basic Examination of Electropolishing
with PWRs Samples
-------
(Bq/cmz)(/iCi/cm2)
T 102
10Z
10
o
o
o>
o
o
«*—
u_
=3
CO
id1
10'
id6
Anodic electrolysis
—o— 10 wt% H2S04
—o— 80wt% H3P04
—«— 20 wt% Na2S04
Alternating Current Electrolysis
—*~ 10 wt% H2S04
—*~ 20 wt% Na2S04
I
I
0 20
40 60 80
Time (min)
100 120
CM
I
O
10'
CO
10'
L.L.D
-Anodic Electrolysis Decontamination
Electrolytic- Abrasive
Decontamination
__i i i
10 20 3032
Decontamination time (min)
with Immersion System
Electrolyte
Material
Current density
Temp, of electrolyte
ntamination
SUS
0.2 A/cm2
80°C
j Electrolytic -Abrasive
i
i
}4J Electrolyte
i Material
i Current density
! Temp, of electrolyte
Decontamination
20 wt % Na2N03
SUS
5 A /cm2
Room Temp.
Fig. 5 Results of Basic Examination of Electropolishing
using Carbon Steel Samples of the JPDR
Fig. 6 Relation between Surface Activity of Complex
Sample and Decontamination Time
-------
DISMANTLED PIPE
Decontamination condition
(Edges cut by
i air- plasma ,-
, Welding
Electrolyte
Temp, of electrolyte
Current density
Decontamination time
Sample
L.L.D
80wt% phosphoric acid
60'C
0.2A/cm2
30min
SUS 304
(l.txlO"'Bq/cm2)
*
ID
Edges cut by air- plasma
Thermal cutting method
put contamination into
metal.
Complete decontamination
is not possible.
^\ \ (T\~\
rY|Iu
\} ) ®j~i
W— - •/ ^/ /
Ultra sonic decontamination
is not effective.
DF 50 ~ 400
1
.(F~l
Simple pipes
No contamination was detected
after decontamination.
. . • °
• :.- ,{
DF 600 ~ 1300
1
CD
Pipes including welding
Contamination remains on
welding parts.
It can be removed by mechanical
decontamination method.
n f
L» I
1
Ultra sonic decontamination
is not effective.
DF 700 ~ 900
fJ*Tl
0)
Pipes with seat
Contamination remains on
inner surface of hole and
around the seat.
DF 20 ~ 500
1
V \;
Pipes with spatters
Pipe cut by thermal method
has spatters near its edges.
Spatters remain after decon-
tamination. Contamination was
detected at the spatters.
-1 1— £4-
U ... u
v y y
Spatters can be removed by
mechanical decontamination
method.
No contamination detected
after the decontamination.
DF 600H30Q (After mechanical
decontamination. )
00
Fig. 7 Results of Electroplishing Decontamination of Various Parts of Pipes
-------
Low-Level Radioactivity Measurement Methods
for Reusing or Recycling
Iwao Manabe, Yukio Iwata and Masao Oshino
Department of Health Physics
Japan Atomic Energy Research Institute
ABSTRACT
When reusing facilities and land, or recycling materials, radioactivity surveys and
measurements are indispensable to make the program effective and to assure radiologic safety.
We discuss the procedures and techniques of radioactivity measurement needed to
confirm the contamination levels for reuse or recycling. These procedures can be divided into
three phases: planning, decontamination, and reuse or recycling. Radioactivity measurements
should be made in each phase: In the first, pre-investigating the contaminated spots or
materials; in the second, checking the contamination levels and classifying the materials; and in
the third, verifying the contamination levels and confirming the radiologic safety. Some methods
and instruments appropriate for each phase are proposed, and the measurements (including
minimum detectable limits) are discussed. New kinds of automated systems developed at JAERI
for verifying low level radioactive wastes are presented. One of these has a minimum detectable
limit of 10 mBq/g for Co-60 and Cs-137.
We shall illustrate the procedures and measurement techniques with three examples: the
reuse of research reactor facilities, the reuse of the radiochemical laboratory, and the dismantling
of a power reactor. Our experience so far has shown the effectiveness of the survey and
measurement techniques and of the automated measurement system for 200 liter drums.
INTRODUCTION
When reusing facilities and land, or when recycling materials, radioactivity surveys and
measurements are needed to make the program effective and to assure radiologic safety. Since
the objects of the surveys and measurements vary in amount, form and contamination level,
appropriate methods must be applied to each case. This may also require an apparatus or a
measurement system designed exclusively for a particular case.
We will discuss the procedures and techniques of radioactivity measurement needed for
reuse or recycling. This paper gives three application examples: reuse of research reactor
facilities, reuse of a radiochemical laboratory, and dismantlement of a power demonstration
reactor.
82
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PROCEDURES OF RADIOACTIVITY MEASUREMENT FOR REUSING OR RECYCLING
The major objects of reuse and recycling are buildings and land dedicated to processing
or treatment of radioactive materials, and structured materials used in the controlled area. The
major items requiring measurement are floors, walls and ceilings of a facility for reuse, and steel
plates and pipes removed from the facility for recycling. •* . '
The process of reuse or recycling of a facility consists of three phases: planning,
decontamination, and reuse or recycling. The radioactivity measurement procedure is divided
into three corresponding phases, as shown in Figure 1. The measurements in the first phase are
needed to make a reuse or recycling plan. The measurements in the second phase serve to
check the contamination levels of the buildings and the land, and to classify the structured
materials removed from the controlled area. The measurements in the third phase verify the
contamination level of the building, land, and the removed materials, and confirm their
radiological safety before reuse and recycling.
MEASUREMENT METHODS
Principal long-lived nuclides to be measured are commonly used nuclides such as Co-60,
Sr-90, Y-90 and Cs-137, and activation nuclides such as Cs-134, Eu-152 and Eu-154.
Various radioactivity measurement methods have been proposed for reuse and recycling.
It is important to choose a method or a combination of methods based upon their merits and the
contamination situation at hand. Measurement should be applied to large amounts of material.
There are four classes of methods:
A. Survey on the spot, using portable contamination meters.
A portable contamination meter offers easy and flexible operation. The instrument is
inexpensive, but requires substantial labor input. Typical measurement instruments for surface
contamination are shown in Table 1. A gas flow proportion counter with a large thin window has
a minimum detectable limit (MDL) less than 0.1 Bq/cm2 in a short counting time for
alpha/beta-ray.
This method is applicable for all three phases discussed in the preceding section.
B. Measurement of small specimens sampled at the selected location using sophisticated
instruments.
Typical instruments are gamma-ray spectrometers, low background alpha/beta counters,
and liquid scintillation counters. Such instruments provide detailed analysis with high precision,'
but they are expensive and require operators. Typical radioactivity measurement instruments and
their performances for sampled specimens are shown in Table 2, which also includes information
of ordinary survey meters used for measuring radioactive steel [1 ].
83
-------
It is generally possible to lower MDL by extending the counting time, or by increasing the
amounts or concentrations of specimens. However, the number and the kind of specimens and
total costs should be considered when making a decision.
This method is applicable for all phases, and for measuring contamination of. either
structured materials or removed materials. ,
C. Measurement of large amounts of material using an automatic measurement instrument.
Such a system measures mainly gamma-rays from removed materials in drum-sized
containers, and is better when custom designed for this purpose. The systems must be installed
near the working area. This is expensive, but it allows a large amount of material to be checked
effectively and quickly.
This method is applicable in the second and third phases.
Some systems for which this method has been developed are as follows:
• A measurement system for active neutron assays of TRU waste in 200 liter drums (MDL :
> 100 Bq/g for Pu-239) [2]
. An automatic measurement system using 4 Nai (Tl) gamma-ray spectrometers for waste
canned in 200 liter drums (MDL: 40 mBq/g for Cs-137)
• An automatic measurement system using 2 Ge gamma-ray spectrometers for waste canned
in 200 liter drums (MDL : 10 mBq/g for Cs-137, Co-60) [3], [4]. (described later)
D. Consecutive monitoring of materials conveyed continuously using an automated
instrument with a warning function.
These systems should also be custom designed. The method has the same features as the
third one, except for its inferiority in analyzing capability. Similar systems have been developed
for checking articles from controlled areas. Atypical one is the contamination inspection monitor,
which has MDL of 0.1 Bq/cm2 for Co-60.
In the planning stage, the first phase, it is important to make a careful investigation of the
facility's history and the characteristics of the contamination spot, e.g. major radionuclides,
contamination level and expanse, with the methods of A and B presented above.
As to the measurement technique, it is necessary to pay attention to the radionuclides
resulting from elsewhere, such as nuclear tests or Nature, which could potentially interfere with
the analysis.
MEASUREMENT EXPERIENCES FDR REUSE OF FACILITIES AT JAERI
This section describes three cases in which the policy discussed in the preceding two
sections was applied. Specific measurements were made for reuse programs of a research
84
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reactor facility, a radiochemical laboratory, and the land occupied by a power reactor
demonstration facility.
1) JRR-3 facilities (Japan Research Reactor No.3)
JRR-3 is a thermal-heavy water reactor fueled with natural uranium and operated for about
20 years. Investigation in the planning stage verified considerable H-3 contamination of the
whole reactor containment area, and surface spots of Cs-137 contamination scattered on the
floor.
H-3 was found to permeate deep into the concrete. H-3 activities of bored samples were
measured with a liquid scintillation counter. The counter has a MDL of 0.4 Bq/g. The floor
surface contamination, on the other hand, was surveyed with a gas flow p/oportion counter-type
contamination meter with a large thin window (GPCM). The instrument has a MDL of 10
mBq/cm . Samples of the removed materials were analyzed with a Ge gamma-ray spectrometry
system. The system has a MDL of 2 mBq/g.
These three methods were applied to investigate in the first phase, and to classify
materials in the second phase.
The reactor containment was released from control after appropriate cleanup, and a new
research reactor has been constructed within the containment. Contaminated materials were
classified into two categories: (1) materials contaminated with Cs-137 of with H-3 over 37 Bq/g,
and (2) materials contaminated with H-3 below 37 Bq/g. The former was disposed of as low-level
solid waste, and the latter has been under control in a storage facility.
2) Radiochemical laboratory
Research Laboratory Building No. 1 included many radiochemical facilities in which
unsealed radioactive materials have been treated for more than 20 years. It was remodeled for
reuse as an ordinary office building after cleanup.
A careful investigation in the planning stage, verified about 330 scattered spots of surface
contamination with C-14, Co-60, Ru-106, Cs-137, U and Th, with maximum concentration of 400
Bq/cm.
The surface contamination of the floor and equipment was surveyed with a GPCM.
Samples of scrapped materials, such as concrete fragments and pipe pieces, were analyzed with
a Ge gamma-ray spectrometry system.
The facilities were released from control after a series of elaborate cleanups and have
been reused as an ordinary office building. The scrapped waste materials were categorized into
two groups: (1) contaminated materials, and (2) those below the minimum detectable limit, and
sure to be uncontaminated. The former was disposed of as low-level solid wastes, and the latter
as ordinary industrial waste within the site.
3) JPDR facilities (Japan Power Demonstration Reactor)
85
-------
JPDR is a BWR-type demonstration reactor that has been operated for 17,000 hours. The
reactor facilities have been dismantled to allow study of the technology, and to reuse the site.
Heavily contaminated materials, such as spent fuel assemblies and interior parts of the reactor
itself, have been removed from the facilities.
A comprehensive investigation, in the planning stage, made clear the radiological impact
and led to a concrete dismantling program.
The contamination could be divided into three groups: (1) activated materials in the
biological shield around the core, (2) fission products in the piping, and (3) fission products on
the surface leaked from the piping.
Measurements of density and nuclide contamination have been made with a GPCM and
Ge gamma-ray spectrometry system, respectively.
An automated measurement system was developed exclusively to confirm the extremely
low-level waste contained in the 200 liter drums. The system has a computerized gamma-ray
spectrometer with two scanning Ge detectors, and it can measure the total radioactivity in a drum
in about 10 minutes. The system .has a"MDL of 10 mBq/g for Co-60 and Cs-137. The total
activity in a drum measured with the system is in agreement with that obtained by measurements
of sampled specimens.
CONCLUSION '
We have presented the radioactivity surveys and measurement techniques in connection
with reuse of facilities and land, or recycling of removed materials.
We have applied our low-level radioactivity measurement methods to the reuse of a
research reactor facility, a radiochemical laboratory, and the dismantling of the power
demonstration reactor in JAERI. The gas flow proportion counter-type contamination meter with
a large thin window (GPCM) for checking surface contamination and the Ge gamma-ray
spectrometer for measuring sampled specimens have proven to be excellent in sensitivity,
response, cost and measurements. And the automated measurement system for 200 liter drums
has been effective for large quantities of material.
REFERENCES
[1]
[2]
[3]
H. Kaminaga, K. Ohkubo, S. Suga and Y. Kajimoto; Estimation of Radioactive
Concentration in the Activated Steel by Survey Meters. Health Physics vol 13.103-111
(1978) _
JAERI-M 88-221 P.178-180 (1988)
Proceedings of the 1986 Fl Meeting of the Atomic Energy Society of Japan, G42 (1986)
[4] Proceedings of the 1988 Fl Meeting of the Atomic Energy Society of Japan, G9 (1988)
86
-------
Phase of
Reusing or
Recycling
PHASE 1
Planning
PHASE 2
Decontamination
PHASE 3
Reusing
Recycling
CX3
Procedure
.".:,'••-. of :;"•;• \
Measurement
Preinvestigation
survey
\
>
, ^
f
k
Planning
Contamination
survey :
Activity
measurement
for classification
Decontamination
Contamination
survey
Activity ;
measurement.
i Confirmation i
jy for safety [
Reusing of
facilities or land
Recycling of
removed materials
Figure 1 Procedure of Radioacitivity Measurement for Reusing or Recycling
-------
Table 1 Typical Measurement Instruments for Surface Contamination ;
Measured
item
Floor
Wall
Floor
Article,
Steel, etc.
Sample
shape
Flat
Flat
Plate
Box
Instrument
GM survey
meter
Gas flow
proportional
contami. meter
Plastic
scintillation
contami. meter
Floor contami.
monitor
Contamination
inspection
monitor
Detector
Type Size
(cm2)
GMtube 20
Gas flow 100
proport.
Plastic
scintil- 50
lation
Gas flow
1800
proport.
Plastic
scintil- 5250
lation
Radiation &
typical nu elides
p(r) 60Co,137Cs
a U
P 60CO,137CS
P 60Co,137Cs
p u
60Co
p U
60Co
Time
constant
or count
time*(s)
10
10
10
10
10*
600*
600*
Minimum
detectable
limit
(Bq/cm2)
: 0.4
0.01
0.05
0,B
0.04
0.1
0.04
0.1
00
00
-------
Table 2 Typical Radioactivity Measurement Instruments for Sampled .Specimen
Measured
item
Concrete
S,teel
Concrete
Steel Disk
SUS Pipe
Sample
amount
(g)
> 10
1
1
500
1
30cm (D)x
2,7c.m..(T).....
2.5cm(D)x
.50cm (L)
Instrument
GM counter
Low back
ground a/p
counter
SSD a-ray
spectrometer
Ge r-ray
spectrometer
Liquid scinti.
counter
Nal survey
meter
Nal survey
meter
Detector
Type Size
(cm2)
GM tube 20
Gas flow 20
proportional
Si surface 10
barrier •
HpGe 150
': -. (cm3)
LSC 20
(cm3)
Nal 2.5cmDx
Scinti.- 2.5cm,L
Nal 2.5cm Dx
Scinti. 2.5cm L
Radiation &
typical nuclides
p(r) 60co,137cs
a U,Th
P 50Co,137CS
?... U'Th
r 60co,137cs
13fe,'52Eu
() 3H,UC
r. 59Fe,54Mn
r 51Cr
Time
constant
or count
time*(s)
600*
600*
600-*
10000*
500*
"j&OO*"'
10
10
Minimum
detectable
limit
(Bq/g)
0.4
0.005
0,01
0.04
0.002
0.003
.0.4
0.1
0.4
oo
-------
Disposal Capacity and Projected
Waste Volumes Within the Low-Level
Radioactive Waste Compacts
Steven R. Adams
US Ecology, Inc.
ABSTRACT
Waste volume and activity projection analysis have been performed for the Southwestern
and Central States projects. A detailed discussion of the data and projections from fuel cycle
generated waste, medical wastes, academic, government and non-medical industry waste is
presented. The effects of nuclear power plant decontamination, volume reduction technology,
U.S. NRC/U.S. EPA BRC decision, and mixed waste production are discussed. The projected
waste volumes and activity projections are compared with present disposal capacity.
In the late 1970s, the governors of Washington, South Carolina and Nevada declared that
they would close their facilities to out-of-state LLW if swift progress were not made in developing
new disposal sites in other regions of the country. In response to the concerns of these states,
Congress in 1980 passed the Low-Level Radioactive Waste Policy Act. This Act required each
state to be responsible for providing disposal capacity for LLW generated within its own borders.
To accomplish the objective of the 1980 Policy Act, each state had the option of either
establishing its own disposal facility or entering into a compact with other states to develop a
regional facility. The Act encouraged the formation of compacts and regional solutions to LLW
management issues. By January 1, 1986, the deadline for fulfilling the objective of the Act,
thirty-six states had formed seven compacts, which were ratified by Congress. These states
accounted for approximately 68% of the LLW shipped for disposal in 1986.
No new disposal facilities had been established, however. As it became apparent that
no additional states would meet the January 1,1986, compliance date, the states and compacts
began negotiating amendments to the 1980 Act. The result was compromise legislation, the
Low-Level Radioactive Waste Policy Amendments Act of 1985, which allows access to the three
existing disposal facilities on a limited basis until January 1, 1993. States and regions without
their own facilities, are now required to meet a series of milestones. These are deadlines by
90
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which specific actions or tasks must be completed to ensure that new LLW disposal facilities will
be developed in states or regions by January 1,1993. A penalty system of escalating surcharges
and eventual loss of access to waste disposal facilities has been devised to foster compliance
with the milestones. States and regions that meet the milestones are eligible for a 25% rebate
on surcharges. The design of low-level radioactive waste (LLRW) disposal facilities is dependent,
among other criteria, on the determination of types, kinds and quantities of waste generated
within each compact. This paper discusses the projected types, kinds and quantities of waste
for the Southwest and Central States LLRW compacts.
SOUTHWESTERN INTERSTATE COMPACT
The low-level radioactive waste being considered for the Southwestern Compact is based
on historical data taken from US Ecology radioactive waste shipment records. AH shipments
originating from California, Arizona, North Dakota, and South Dakota from 1985 through 1987
have been considered in the analysis. After reviewing shipment records from North and South
Dakota, the conclusion was reached that their waste volumes and activities were insignificant and
would not be considered further in the analysis.
The various categories of waste (known as waste streams) present in the Southwestern
Compact can be divided into two major groups. Table 1 lists those wastes associated with fuel
cycle facilities (nuclear power stations) and Table 2 lists those wastes associated with non-fuel
cycle facilities. The tables identify those major individual waste streams that, based on historical
data, account for more than five percent of the total volume or activity generated for that waste
category (fuel cycle or non-fuel cycle), or, based on future projections, will account for more than
five percent of the disposed waste volume or activity.
The low-level radioactive wastes (LLRW) expected to be buried at the Southwestern
Compact proposed Ward Valley site near Needles, California are produced by several categories
of generators including:
1) Nuclear power stations
2) Government
3) Medical
4) Academic Institutions
5) Industry (non-medical)
NUCLEAR POWER STATIONS
The volumes of wastes that can be expected to be produced by the generators in each
category are described in the following sections.
Dry solid waste, which consists of contaminated tools, clothing, and equipment, can be
grouped into compatible or noncompatible waste. Current operating practices are to attempt to
compact as much waste as possible to reduce burial volumes. Thus, several nuclear power
stations go through sorting processes to determine which wastes can be compacted to a smaller
volume. This type of waste material is usually packaged in 55-gaIlon drums or some type of
large metal box.
91
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Nuclear power stations generate large amounts of radioactive liquids. Typical liquid
wastes are from laundry operations, laboratory work, and flushing of various filter media
systems. All liquid wastes must be either dewatered, absorbed, or solidified before disposal at
the Ward Valley site. Dewatering, absorption, solidification, and packaging of liquid waste, filter
media, and resins must be performed by the waste generator in compliance with 10 CFR 61.56,
the U.S. NRG Technical Position on Waste Form, and the applicable California regulations,
standards, and guides prior to acceptance for disposal at the Ward Valley LLRW disposal facility.
The final type of waste generated at nuclear power stations includes absorbed aqueous
and nonaqueous liquids, such as oils, paints, and various other miscellaneous liquids. Any
absorbed liquids accepted for disposal at the proposed Ward Valley site will have been absorbed
using absorbent approved by the state of California.
GOVERNMENT
Government generators consists of several types of facilities, including naval operations
and research facilities. Naval operations generally produce the same types of wastes as nuclear
power stations. The research facilities produce various miscellaneous wastes including dry
solids, liquids, and others.
MEDICAL
Medical facilities include hospitals, medical research facilities, and any industrial facility
that produces products for medical use (such as radiopharrnaceutical plants and some biological
research labs).
Medical facilities use several radionuclides for diagnostic and therapeutic treatments.
These radionuclides are produced artificially in radionuclide production facilities. Normally, LLRW
is generated in the form of glass, plastic, trash, and small quantities of metal and liquid.
However, it is important to note that these small quantities of liquids may possess very high
concentrations of various radionuclides. The proposed Ward Valley site is projected to receive
large amounts of tritium (H-3) wastes from several radiopharrnaceutical plants. This tritium
wastes is usually sealed in structural concrete within stainless steel 2R containers which are
placed within 55-gallon drums or within liners.
Other wastes produced by medical facilities include liquids scintillation vials and fluids.
The vials, which are made of glass and occasionally polyethylene, are usually filled to less than
50 mL with fluid. Only scintillation fluid that are rendered non-hazardous and which have been
packaged in sufficient absorbent material to absorb twice the volume of the liquid or solidified
in compliance with 10 CFR 61.56(a)(2) and (8) will be accepted for disposal.
ACADEMIC INSTITUTIONS
Colleges and universities produce various wastes as a result of research projects which
can cover a wide variety of fields including medicine, nuclear fuel cycle, and biology.
Probably the largest amount of waste generated by academic institutions is from
biological research programs. This waste consists of animal carcases, animal bedding, excreta,
92
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and vegetation arid culture media. These types of wastes are usually disposed of after being
packed in lime and absorbent material.
Several universities operate neutron activation analysis (NAA) facilities that generate small
amounts of waste material in the form of activated metal foils, miscellaneous trash used in lab
procedures, and other small amounts of dry solid and liquid wastes. The liquid wastes are
packed in absorbents or solidified before disposal.
Another type of waste generated at academic institutions is produced from accelerator
target bombardment. Accelerator targets are used to produce radionuclides by direct
bombardment with charged practical beams or by indirect reactions of the target fragments with
other materials. Spent targets are commonly made of titanium foils containing absorbed tritium.
INDUSTRY (NONMEDICAL)
The last category of waste generators is the industrial group. These facilities manufacture
items such a density gauges, well logging devices, radiography sources, X-ray fluorescence
tubes, and static eliminators.
The waste generated from these operations include dry solids, liquids, and industrial and
institutional sealed sources that have been spent.
WASTE PROJECTIONS
Thirty-year projections have been made of the waste expected to be disposed of at the
proposed Ward Valley facility. The basic assumption is that the character and technology of the
industry will not change thus the current disposal trends will continue to hold constant for the
30-year operational life of the facility.
Actual radioactive waste shipment record wers used as baseline data for the 30-year
projection analyses made for the proposed facility. The radioactive waste shipment records from
1985 through 1987 were sorted by month into 36 monthly summaries for four categories:
1. Fuel cycle generator volume
2. Fuel cycle activity
3. Nonfuel cycle generator volume
4. Nonfuel cycle generator activity.
These four categories were plotted to determine if any further modifications were needed
on the data before performing the actual 30-year projections. From these plots, it was
determined that some of the data needed to be smoothed in order to obtain acceptable results
from the projection analyses. The fuel cycle data, for example, showed large variations in
monthly volumes shipped to the disposal facilities. Further investigation revealed that the peak
months for waste volume shipped immediately preceded waste burial regulatory changes or price
increases at the disposal facilities. Therefore, the waste generators were apparently attempting
to deplete their waste inventory while the burial and price conditions were more favorable to
them. This had the opposite effect on the months immediately following the regulatory or price
changes. These months showed large reductions in the amounts of. waste shipped for burial.
93
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Standard mathematical models normally do not handle large fluctuations like these satisfactorily.
Thus, the data was smoothed in order to provide better baseline data for the 30-year projections.
The smoothing technique was a simple five term averaging model. The resultant data
maintained the general trends of the original data but without the large fluctuations caused by
regulatory changes or price increases. Projections were performed using standard S-curve
models. The S-curve model is based on the function Z = exp (a + b/l) and fits an S-shaped
curve through the data.
The coefficients (a and b) were obtained by least squares after taking the natural
logarithm of Z. This trend analysis resulted in 30-year summaries for the four waste categories
discussed above. These summaries are shown in Tables 3 and 4, along with the information on
decontamination wastes and tritium wastes.
Tritium wastes are a major source of activity projected to be buried at the proposed Ward
Valley facility. Large quantities of tritium wastes are generated at several pharmaceutical
companies in California. The tritium waste accounts for over 80 percent of the total activity (but
less than 6 percent of the volume) projected to be buried at the facility. The projected activities
and volumes of the tritium wastes were determined from waste surveys sent to the two major
generators of this type of waste. /
In addition to analyzing the waste projections based on he different waste generator
categories, the projections were also broken down into the different types of waste (or waste
streams) expected to be produced. ;
Tables 5 and 6 depict annual volumes of the various waste streams for fuel cycle and
nonfuel cycle generators, projected to be buried at the proposed Ward Valley facility. These
volumes are based on total projected volumes (from Table 4) as modified by individual waste
stream average volume percentages (from Tables 1 and 2). Again, the assumption has been
made that these volume percentages will not change over the operational lifetime of the
proposed facility. ;
Tables 7 and 8 show the annual activities of the various waste streams projected to be
buried at the proposed Ward Valley facility. These concentrations are based on overall activities
(Table 3) as modified by individual waste stream average activity percentages given in Tables 1
and 2.
The radionuclides used in this analysis are listed in Table 9. The activity shown is in the
activity projected to be present in the disposal trenches at the proposed Ward Valley facility at
the end of its 30-year operational lifetime (this activity includes radioactive decay and buildup
through the 30-year period). The percent column indicates the percentage of the activity that the
individual radionuclide accounts for in relation to the total activity. The tritium (H-3) waste
accounts for a large percentage of the total activity. Because of this high total activity, the H-3
wastes are treated as a separate waste stream in the analysis.
AH radionuclides shipped for burial during the reference period were totaled according to
the various fuel cycle and nonfuel cycle waste streams. An assumption was made that these
radionuclides would continue to be produced during the 30-year operational life of the proposed
94
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Ward Valley disposal facility. This allowed a projection to be made on the concentrations of
radionuclides in each waste stream. These concentrations are shown in Tables 10 and 11.
Note that the values in Tables 10 and 11 include only the radionuclides that were
indicated on the actual radioactive waste shipment manifests from the US Ecology database.
Thus, no decay of short-lived radionuclides or ingrowth of daughter products is shown unless
they were shown on the original manifests.
Class A waste is generally composed of radionuclides with short half-lives. This type of
waste poses the smallest threat to the environment since the majority of its activity decays
relatively quickly after burial.
Class B waste is composed of a mixture of radionuclides, some with short half-lives and
some with longer half-lives. The short-lived radionuclides are usually of very high activities and
are placed into this classification as a safety precaution. The longer-lived radionuclides are
composed of some transuranics and other radionuclides that are relatively mobile in the
environment.
Class C waste is composed of radionuclides that could potentially pose the greatest
threat to the environment. This type of waste is primarily composed of transuranic radionuclides
and high concentrations of relatively mobile radionuclides. '
At the proposed Ward Valley facility, Class B and C wastes are disposed of in a separate
trench from the majority of the Class A wastes. In accordance with guidance received from the
California Department of Health Services (DHS), stabilized Class A wastes with a contact dose
rate greater than 30 R/hr are buried with the B and C wastes in the dedicated "BC30" trench.
The waste streams previously identified can be further broken down to provide more
detailed information concerning the wastes to be buried at the proposed Ward Valley facility.
Several assumptions must be made to provide a basis for the detailed breakdown. First,
historical data is used to provide percentage breakdowns on waste classes, waste streams, and
radionuclides. These initial breakdowns are assumed to hold constant for the life of the facility.
This is believed to be a valid assumption, as major generator waste production rates are not
expected to change significantly during the operational life of the proposed facility.
Thus, the 3-year (from 1985 through 1987) average breakdown of wastes by classification
is a good measure of the projected breakdown. These 3-year averages and the resulting
projections are shown in Table 12.
FUTURE GENERATION RATE ALTERATIONS
Available information from waste generator survey questionnaires completed by major
generators on waste processing techniques indicates that there will be little change over the first
5 years of operation at the Ward Valley facility. This implies that the current volume reduction
(VR) and volume increase (VI) factors achieved by various waste processing techniques will
remain fairly constant. Volume reduction is achieved at fuel cycle facilities through the use of
sorting of dry active wastes (DAW), compacting wastes as appropriate, and evaporation of water
from liquid wastes. A volume increase of the waste results from most solidification techniques
'.-.., - 95 . / -'
-------
(with the exception of bitumen) and packaging systems. Unless otherwise noted, all data
presented in this section is based on "as-shipped for burial" (processed) volumes of wastes.
NUCLEAR POWER STATION DECONTAMINATION WASTES
Nuclear power stations may periodically generate an additional type of waste called
"decontamination waste." Predominately resins containing chelating agents, their kind of waste
is discussed below.
During the operation of a nuclear power station, a very thin metal oxide layer forms in
several systems of the plant, including the reactor, steam, and water piping systems. This layer,
often referred to as "crud," is predominantly an iron-nickel oxide which contains radioactivity
mostly in the form of Cobait-58 and Cobalt-60. These radionuclides are significant gamma
radiation emitters and can severely limit the normal maintenance and plant inspection procedures
necessary for the safe operation of the facility.
Several chemical cleaning agents have been developed that greatly increase the ability
to remove this layer of crud. These chemical agents form complex ions containing the
radioactive metal ions from the crud. However, the resulting complex ions are highly soluble and
must be handled in a different manner during subsequent burial. Being soluble, the chelated
waste may exhibit enhanced transport and mobility properties in the soil, thus potentially
migrating faster than other nonchelated wastes buried in the same area. Furthermore, these
complex ions have greater bonding potential than the standard waste forms that may be buried
at the proposed facility. Since the bonding properties of the chelating agent are stronger than
the surrounding backfill material, any radionuclides that may have leaked from other nearby
wastes may become attached to the chelate's complex ion structure. These radionuclides would
then migrate at the same increased rate as the original chelated radionuclides. The Ward Valley
facility's operating procedures shall require the separation of wastes containing chelating agents
from other waste in the disposal trenches; the distance between chelate and nonchelate waste
is expected to be a minimum of 10 feet. This practice reduces the potential for nonchelated
radionuclides becoming attached to the chelated radionuclides.
Little information is available concerning the amount of decontamination wastes containing
chelating agents to be disposed of at the facility. Therefore, generic data from NUREG-0782
(NRC81) has been used to project the chelated waste volumes. NUREG-0782 assumes that a
decontamination of the primary coolant system at a nuclear power station occurs every 5 to 10
years. Conservatively assuming that a full decontamination is performed every 7 years (an
average of the 5 and 10 year estimates) results in the generation of approximately 68,700 cubic
feet (processed volume of chelated wastes) to be disposed of over the lifetime of the Ward Valley
facility.
This volume is obtained in the following manner. NUREG-0782 estimates that 1677 cubic
feet of resin waste are generated as a result of decontaminating the primary coolant system of
a typical nuclear reactor. There are nine nuclear reactors in the Southwestern Compact.
Conservatively, an assumption is made that all nine reactors perform their first decontamination
in 1991, thus resulting in a total of 41 decontaminations over the lifetime of the proposed Ward
Valley facility. This number of decontaminations takes into account the projected shutdowns of
Rancho Seco Nuclear Generating Station and San Onofre One Nuclear Station. It also takes into
96
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account the expected decontaminations that these two reactors will perform during the initial
phases of shutdown. This waste is listed as DECON in Tables 4 and 5. Current trends in the
nuclear industry are to use a decontamination process developed in Britain known as low
oxidation metal ions (LOMI). This process is based on picolinic acid and vanaddium, a dissolving
agent classified as a chelate.
NUCLEAR POWER STATION DECOMMISSIONING WASTES
Another significant source of waste is generated from the decommissioning of nuclear
power stations. Six nuclear power stations are projected for shutdown in the SOr-year lifetime of
the proposed Ward Valley facility. These plants are the 63 megawatt electric (MWe) Humboldt
Bay Reactor (currently shut down and in a safe storage mode), the 916 MWe Rancho Seco
Nuclear Station, the 436 MWe San Onofre One Nuclear Station, the 1084 MWe Diablo Canyon
Unit One Nuclear Station, the 1106 MWe Diablo Canyon Unit Two Nuclear Station, and the 58.5
MWe Pathfinder reactor located in South Dakota (currently in a safe storage mode). The analysis
performed assumes that these nuclear stations go into a safe storage mode following shut down
and primary coolant system decontamination. Thus, no additional waste volumes or activities
are included for nuclear station decommissioning activities.
No inventory limitations are currently proposed for carbon-14, tritium, technetium-99,
iodine-129, or the transuranic species. The waste projections result in total site inventories that
do not adversely affect the environment; i.e., dose rates resulting from buried wastes are
expected to be significantly less than the performance objectives of 10 GFR 61.41. Therefore,
there is no basis for placing limitations on the amounts of any radionuclides.
WASTE DURING CLOSURE PERIOD
The proposed Ward Valley facility is expected to have a nominal 5-year closure period
immediately following the operational period. During the first 12 to 18 months of this closure
period, the site cleanup and building and structure removal is performed. The generation of any
radioactive wastes (due to ground or building surface contamination) is expected to occur during
the first six months of this period. However, because of the procedures utilized to unload and
dispose of radioactive wastes during facility operation, there is no significant waste projected to
be generated during the closure of the facility. Thus, a minimal volume of radioactive waste, less
than 75 cubic feet, is projected to be generated during closure activities at the site.
DISPOSAL CAPACITY - ' -
Based upon regulatory guidance issued by the California Department of Health Services
(DHS) there are dedicated trenches for Class A waste. Class B and C wastes, which have longer
periods of concern, and must meet additional waste form and stability requirements are disposed
of separately from the Class A waste that does not meet these stability requirements?
Additionally, the DHS has determined that all wastes with a contact dose rate of 30 R/hour or
more, irrespective of class, must meet the regulatory stability requirements and be disposed of
in the trench with the Class B and C wastes. Four Class A trenches and one trench for the Class
B and C waste, the BC30 trench, are planned.
97
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To ensure that all waste be buried beneath the deepest projected scouring action caused
by a Probable Maximum Flood (PMF) all wastes are buried at least 20 feet below the original
ground surface at the time of completion. Trench dimensions at the original ground surface
depend on the depth and side slope. Soil stability may allow side slopes of one horizontal to
one vertical. This is the case presumed for the optimistic calculation of the potential disposal
capacity. The trench design may require a 1.5 horizontal to 1.0 vertical side wall design. This
is the case presumed for the pessimistic calculation of the potential disposal capacity. Another
parameter that effects the potential disposal volume is the efficiency in the use of trench volume.
The efficiency is dependent on the method of disposal, waste package design, and the amount
of backfill required to shield waste packages. At the Ward Valley facility the waste packages will
be stacked using front-end loaders. Experience at the US Ecology LLRW disposal facility in
Beatty, Nevada has shown that efficiences range from about 25 to about 33 percent. The four
Class A trenches are approximately 1,546 feet long by 290 feet wide at grade by 60 feet deep.
An optimistic calculation results in a potential volume of 2.4 E+7 cubic feet for the four Class A
trenches. A pessimistic calculation results in a potential disposal volume of 6.8 E+6 cubic feet
for the four Class A trenches. The single BC30 trench is 1,546 long by 226 feet wide at the
surface and 42 feet deep. A shallower excavation is used to ensure that any moisture
accumulation in a Class A trench does not migrate into the Class BC30 trench. The optimistic
calculation of the BC30 trench disposal capacity results in 2.5 E+6 cubic feet while pessimistic
calculation results in a capacity of 8.5 E+5 cubic feet.
CENTRAL MIDWEST INTERSTATE LLRW COMPACT
The majority of the low-level radioactive waste (LLRW) generated within the Central
Midwest Compact region (the states of Illinois and Kentucky) is generated by the nuclear power
reactors within the State of Illinois. Wastes from these reactors constitute approximately 80
percent of the volume, and more than 99 percent of the radioactivity in the LLRW from the Central
Midwest region. It is projected that approximately 185,000 cubic feet per year of LLRW from
power reactors will be sent to the Compact's disposal facility during the early years of its
operation, beginning about 1993. During the 21st century the 14 reactors in the state of Illinois
should begin to be shut down. As this process continues during the subsequent 30 years, the
volumes of waste sent to the Compact's disposal facility could increase significantly. The
increase will be dependent on the methods used for decontamination and dismantlement.
The LLRW generated in the Central Midwest region is currently treated by the waste
generators to achieve a significant reduction in volume. It is estimated that if current treatment
practices continue at the time the regional disposal facility begins operation, approximately
480,000 cubic feet of waste that will be generated annually will be reduced to under 200,000
cubic feet, a reduction by a factor of 2.4 (RAE87). The only way to significantly reduce the
activity in the LLRW is to delay the decontamination and dismantlement of the power reactors
for a period of decades. Figure 1 illustrates the projected history of volumes of waste entering
the regional disposal facility for two scenarios for reactor dismantlement. The first scenario
shows the volumes if the reactors are dismantled immediately after the shutdown and the second
scenario where the reactors are dismantled after 50 years of storage.
98
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CHARACTERISTIC OF LLRW IN THE CENTRAL MIDWEST COMPACT
A data base was assembled containing the projected characteristics of LLRW going to
the disposal facility in 1993. The two primary sources of information in the data base were the
1986 annual surveys of waste generators conducted by the Illinois Department of Nuclear
Science (IDNS) and the Kentucky Division of Radiation and Product Safety. The IDNS survey
requested projections of volumes and activities of waste that will be produced through 1993.
This information was used to forecast waste from Illinois generators. The 1986 volumes and
activities of waste generated in the state of Kentucky were used as the projection for the early
years of the disposal facility operation (RAE87).
The Illinois and Kentucky surveys were supplement and expanded by a number of
telephone calls to the larger waste generators or to any waste generators who had a significant
inconsistency in the information reported on the survey forms. Cognizant officials from each
state were also contacted by telephone and asked to identify potential sources of radioactive
waste that may have been omitted from the survey forms. Projections of reactor
decommissioning wastes, which could be a significant contributor to the wastes in the first half
of the twenty-first century, were made using information developed by the U.S. Nuclear
Regulatory Commission.
A total of 60 waste generators that account for more than 99 percent by volume of 1986
waste shipments from the Central Midwest region were identified, along with their projected rates
of generation and shipment of LLRW for disposal in the early 1990s. A source of LLRW not
considered in the studies by IDNS and Kentucky in estimating the volumes and activities for
disposal are future generators hot presently disposing of LLRW. A significant source of LLRW
in the future could be clean-up and decontamination projects of the U.S. DOE and the U.S. EPA.
No formal study has been promulgated to study the future generation of LLRW due to actions
initiated by these federal agencies. Informal discussions with these agencies has indicated that
projections of LLRW generated in the Central Midwest Compact would be sporadic and would
most likely be less than 10,000 cubic feet per year or a small fraction of the total waste volumes.
The activity generated from these projects is estimated to be very low, a maximum of 50 curies.
POTENTIAL IMPACTS OF VOLUME REDUCTION
The potential for reducing volumes of waste entering the regions disposal facility by
supercompaction and incineration was investigated (CMI87). Two scenarios were investigated
for both supercompaction and incineration to determine their effect on waste volumes. In the first
scenario only supercompactable or incineratable waste which are not currently being treated by
compaction or incineration by the waste generator would be sent to a regional treatment facility.
In the second scenario, all supercompactable or incineratable waste would be sent to the
regional treatment facility even though in some cases the waste is currently being compacted or
incinerated by the waste generator. Tables 13 and 14 show the potential impacts on annual
waste volumed of supercompaction and incineration at a regional treatment facility. Figure 2
summarizes the analyses of the use of regional treatment facilities for supercompaction and
incineration of reactor, nonreactor, and total wastes. It shows the reduction in waste volumes
that are achieved for reactor wastes and nonreactor wastes when regional incineration and
supercompaction facilities are used. It can be seen that the largest reduction in waste volumes
99
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from the operation of a regional treatment facility come from the treatment of reactor wastes, if
those wastes are sent to the treatment facility (RAE87).
DISPOSAL CAPACITY ,
The options for disposal of LLRW at the Central Midwest Compact and choice of specific
design parameters will depend on the characteristics of the disposal site. The disposal site has
yet to be chosen by the Compact. The disposal site, the allowable waste forms, and disposal
unit design have yet to be decided by the Illinois Department of Nuclear Safety and the Compact
Commission. These subjects are undergoing extensive scrutiny at this time. Prototype designs
that have been reviewed include above and below ground vaults, modular concrete canisters,
earth mounded concrete bunkers, augered holes, and mined cavities. Below-ground vaults,
modular concrete canisters, earth mounded concrete bunkers, and augered holes are considered
to be the most likely to be licensed.
CENTRAL INTERSTATE LLRW COMPACT
The Low-Level Radioactive Waste Amendment Act of 1985 (Public Law 99-240) revised
the original act and granted Congressional consent to the states of Arkansas, Kansas, Louisiana,
Nebraska and Oklahoma to join together and form the Central .Interstate Low-Level Radioactive
Waste Compact. In 1987 the Compact Commission selected US Ecology to site, design, license,
construct, operate, and close a facility for managing the nonfederal, low-level radioactive and
mixed waste generated in the Compact region. US Ecology chose Bechtel National, Inc. (BNI)
to be its prime subcontractor. BNI performed a survey of the low-level radioactive waste
generators to determine an estimate of the annual LLRW volumes generated in the Compact.
The results of the survey are very preliminary and will be augmented by further research on this
subject. The results of the survey are listed in Table 15. The survey results were used in
preparing initial facility design. This initial design consisted of four above-grade vaults for Class
B and C waste with a capacity of 100,000 cubic feet each, one vault for Class B and C waste with
a capacity in the range of 65,000 to 75,000 cubic feet, and one small mixed-waste vault. Review
of these plans by the waste generators resulted in the requests for the mixed-waste disposal
capacity to be increased to 50,000 cubic feet during the 30 year operating period of the disposal
facility. The capacity of the aboveground disposal vaults and the waste generation rate data on
which the capacity requirements are based is still in a dynamic state and may be modified during
the licensing period.
REFERENCES
[1] NRC81 United States Nuclear Regulatory Commission, Draft Environmental Impact
Statement on 10 CFR Part 61, Licensing Requirements for Land Disposal of Radioactive
Waste, NUREG-0782, National Tech. Information Service, Springfield, Virginia, 22161.
[2] RAE87 Central Midwest Interstate Low-Level Radioactive Waste Commission, Potential
Impacts of Source and Volume Reduction Techniques on the Central Midwest Compact's
Waste Management System, prepared by Rogers and Associates Engineering Corp.,
October, 1987.
100
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TABLE 1
FUEL CYCLE WASTE STREAMS
i r
1
| WASTE DESCRIPTION
— : — -
1
(dewatered resin
..
(solidified
(liquids
(filter media
| (sludge/cartridges)
(dry active solid
(wastes (DAW)
! —
1
(solidified resin
(decontaai nation
wastes
_ i
MAJOR
WASTE STREAM
YES
YES
NO
YES
YES
NO
PERCENT AC
VOLUME
7.0
17.9
1
1.8
64.9
8.4
see
note 1
E OF TOTAL
ACTMTY
44.2
1.7
3.8
I 1
25.4
24.9
see
note 1
NOTE 1: The primsry reactor coolant decontamination waste stream has been intentionally emitted fro. this
table since th« available data concerning this waste stream is limited. Actual projections
concerning this waste stream can be seen in Tables 3 and 4.
101
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TABLE 2
NON-FUEL CYCLE WASTE STREAMS
WASTE DESCRIPTION
solidified liquids
biological' wastes
dry active sol id
wastes
tritium wastes
"
ion exchange resin
absorbed aqueous
liquids
Absorbed noo*
queous liquids
non-aqueous liquids
in vials
MAJOR
WASTE STREAM
'-
YES
NO
YES
YES
YES
YES
NO
NO
PERCENT*)
VOLUME
17.5
3.5
64.7
see
note 1
1.0
8.2
1.5
/
3.6
X OF TOTAL
ACTIVITY
'
2.3
<0.1
88.5
see
note 1
6.6
2.5
-
<0.1
••
<0.1
I
MAJOR GENERATORS
radtopharMceut i ca I
coepanies, labs
research centers
and lab*
all non-fuel cycle
generator*
radiophanaaceutical
companies
government and
academic institutes
research centers
and labs
,
all non-fuel cycle
generators
research centers
and labs
MOTE 1: The tritium waste activity his been intentionally omitted from this table since it would
drastically skew the data. The «nunt of tritisttd wast* projected to be buried at the facility
accounts for sere than eighty percent of the total projected burial activity.
102
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TABLE 3
YEARLY ACTIVITY PROJECTIONS
FOR VARIOUS WASTE CATEGORIES/STREAMS
activities «r«.giv«B in curi.s**
YEAR FUEL CYCLE
1991
1992
1993
1994
1995
1996
1997
1998
1999
2000
2001
2002
2003
2004
2005
2006
2007
2008
2009
2010
2011
2012
2013
2014
2015
2016
2017
2018
2019
2020
557.6
555.4
553.8
552.5
551 .4
550.6
549.8
549.2
548.7
548.2
547.8
547.5
547.1
546.8
546.6
.546.4
546.1
493.5
469.7
469.6
469.4
469.3
469.1
469.0
468.9
468.9
468.8
468.7
468.6
468.5
NON-FUEL
CYCLE
935.8
936.9
937.7
938.3
938.9
939.3
939.6
940.0
940.2
940.4
940.7
940.8
941.0
941.2
941.4
941.5
941.6
941.7
941.8
941.9
941.9
942.0
942.1
942.1
942.2
942.2
942.3
942.3
942.4
942.4
DECON
12870.0
12870.0
12870.0
12870.0
12870.0
12870.0
12870.0
12870.0
12870.0
12870.0
12870.0
12870.0
12870.0
12870.0
12870.0
12870.0
12870.0
6545.0
6545.0
6545.0
6545.0
6545.0
6545.0
6545.0
6545.0
6545.0
6545.0
6545.0
6545.0
6545.0
TRITIUM
30400.0
33840.0
37544.0
41538.4
45852.2
50517.4
55569.1
61046.0
66990.6
73449.7
80474.7
88122.2
96454.4
105539.8
115453.8
126279.2
138107.1
151037.8
165181.6
180659.8
197605.8
216166.4
236503.0
258793.3
283232.6
310035.9
339439.5
371703.5
407113.9
445985.3
TOTALS
44763.4
48202.3
51905.5
55899.2
60212.5
64877.3
69928.5
75405.2
81349.5
87808.3
94833.2
102480.5
110812.5
119897.8
129811.8
140637.1
152464.8
159018.0
173138.1
188616.3
205562.1
224122.7
244459.2
266749.4
291188.7
317992.0
347395.6
379659.5
415069.9
453941.2
TOTALS
15467.5
28222.6
303875.0
4810637.0
103
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TABLE 4
YEARLY VOLUME PROJECTIONS
FOR VARIOUS WASTE CATEGORIES/STREAMS
**A11 volumes ar« giv«n in cubic f«et**
NON-FUEL
YEAR FUEL CYCLE CYCLE
OECON
TRITIUM
TOTALS
1991
1992
1993
1994
1995
1996
1997
1998
1999
2000
2001
2002
2003
2004
2005
2006
2007
2008
2009
2010
2011
2012
2013
2014
2015
2016
2017
2018
2019
2020
50693.4
50659.3
50633.3
50612.6
50595.9
50582.2
50570.7
50560.8
50552.4
50545.0
50538.4
50532.7
50527.6
50522.9
50518.7
50514.9
50511.7
45663.3
43462.0
43459.7
43457.7
43455.8
43453.8
43452.4
43450.4
43449.1
43447.6
43446.4
43445.1
43444.0
82511.1
82439.9
82385.4
82342.4
82307.6
82278.9
82254.8
82234.2
82216.5
82201.0
82187.5
82175.4
82164.7
82155.1
82146.4
82138.5
82131.4
82124.8
82118.7
82113.2
82108.1
82103.3
82098.9
82094.8
82090.9
82087.3
82083.9
82080.7
82077.7
82074.8
2436.9
2436.9
2436.9
2436.9
2436.9
2436.9
2436.9
2436.9
2436.9
2436.9
2436.9
2436.9
2436.9
2436.9
2436.9
2436.9
2436.9
2102.3
2102.3
2102.3
2102.3
2102.3
2102.3
2102.3
2102.3
2102.3
2102.3
2102.3
2102.3
2102.3
1437.5
1585.0
1746.5
1923.4
2117.2
2329.7
2562.7
2818.2
3098.5
3406.1
3743.7
4114.3
4521.2
4968.1
5458.9
5998.0
6590.3
7241.1
7956.2
8742.1
9605.8
10555.1
11598.6
12745.7
14006.8
15393.2
16917.5
18593.5
20436.4
22462.8
137078.9
137121.1
137202.1
137315.3
137457.6
137627.7
137825.1
138050.1
138304.3
138589.0
138906.5
139259.3
139650.4
140083.0
140560.9
141088.3
141670.3
137131.5
135639.2
136417.3
137273.9
138216.5
139253.6
140395.2
141650.4
143031.9
144551.3
146222.9
148061.5
150083.9
TOTALS 1426759.8 2465527.9
68757.2 234674.1
4195719.0
104
-------
TABLE 5
FUEL CYCLE GENERATOR
ANNUAL WASTE STREAM VOLUMES
all volumes are in cubic feet **
YEAR
1991
1992
1993
1994
1995
1996
1997
1998
1999
2000
2001
2002
2003
2004
2005
2006
2007
2008
2009
2010
2011
2012
2013
2014
2015
2016
2017
2018
2019
2020
DRY ACTIVE
WASTES (DAW)
32896.2
32874.1
32857.2
32843.8
32833.0
32824.0
32816.6
32810.2
32804.7
32799.9
32795.7
32792.0
32788.6
32785.6
32782.9
32780.4
32778.3
29632.1
28203.6
28202.1
28200.8
28199.5
28198.3
28197.3
28196.1
28195.2
28194.2
28193.5
28192.6
28191.9
SOLIDIFIED
LIQUIDS
9080.2
9074.1
9069.4
9065.7
9062.7
9060.3
9058.2
9056.4
9054.9
9053.6
9052.4
9051.4
9050.5
9049.7
9048.9
9048.2
9047.6
8179.2
7784.9
7784.5
7784.1
7783.8
7783.4
7783.2
7782.8
7782.6
7782.3
7782.1
7781.9
7781.7
FILTER
MEDIA
934.0
933.3
932.9
932.5
932.2
931.9
931.7
931.5
931.4
931.2
931.1
931.0
930.9
930.8
930.7
930.7
930.6
841.3
800.7
800.7
800.7
800.6
800.6
800.6
800.5
800.5
800.5
800.4
800.4
800.4
DEUATERED
RESINS
3528.3
3525.9
3524.1
3522.7
3521.5
3520.6
3519.8
3519.1
3518.5
3518.0
3517.5
3517.1
3516.8
3516.4
3516.1
3515.9
3515.7
3178.2
3025.0
3024. £
3024.7
3024.6
3024.4
3024.3
3024.2
3024.1
3024.0
3023.9
3023.8
3023.7
SOLIDIFIED
RES I MS
4254.7
4251 .9
, 4249.7
4247.9
4246.5
4245.4
4244.4
4243.6
4242.9
4242.3
4241 .7
4241.2
4240.8
4240.4
4240.1
4239.7
4239.5
3832.5
3647.8
3647.6
3647.4
3647.3
3647.1
3647.0
3646.8
3646.7
3646.6
3646.5
3646.4
3646.3
TOTALS
50693.4
50659.3
50633.3
50612.6
50595.9
50582.2
50570.7
50560.8
50552.4
50545.0
50538.4
50532.7
50527.6
50522.9
50518.7
50514.9
50511.7
45663.3
43462.0
43459.7
43457.7
43455.8
43453.8
43452.4
43450.4
43449.1
43447.6
43446.4
43445.1
43444.0
TOTAL 925860.4
255560.7
26286.3
99303.7
119748.7
1426759.8
105
-------
TABLES
NON-FUEL CYCLE GENERATOR
ANNUAL WASTE STREAM VOLUMES
all volumes are in cubic feet **
YEAR
1991
1992
1993
1994
1995
1996
1997
1998
1999
2000
2001
2002
2003
2004
2005
2006
2007
2008
2009
2010
2011
2012
2013
2014
2015
2016
2017
2018
2019
2020
DRY ACTIVE
WASTES (DAU)
53384.7
53338.6
53303.4
53275.6
53253.1
53234.4
53218.9
53205.5
53194.1
53184.1
53175.3
53167.5
53160.6
53154.3
53148.8
53143.7
53139.0
53134.7
53130.9
53127.2
53123.9
53120.8
53118.0
53115.3
53112.8
53110.5
53108.3
53106.2
53104.3
53102.5
SOLIDIFIED
LIQUIDS
14439.4
14426.9
14417.4
14409.9
14403.8
14398.8
14394.5
14390.9
14387.8
14385.1
14382.8
14380.7
14378.8
14377.1
14375.6
14374.2
14372.9
14371.8
14370.7
14369.8
14368.9
14368.0
14367.3
14366.5
14365.9
14365.2
14364.6
14364.1
14363.6
14363.1
SOLIDIFIED
RESINS
825.1
824.3
823.8
823.4
823.0
822.7
822.5
822.3
822.1
822.0
821.8
821.7
821.6
821.5
821.4
821.3
821.3
821.2
821.1
821.1
821.0
821.0
820.9
820.9
820.9
820.8
820.8
820.8
820.7
820.7
ABSORBED
AQUEOUS
LIQUIDS
6765.9
6760.0
6755.6
6752.0
6749.2
6746.8
6744.8
6743.2
6741.7
6740.4
6739.3
6738.3
6737.5
6736.7
6736.0
6735.3
6734.7
6734.2
6733.7
6733.2
6732.8
6732.4
6732.1
6731.7
6731.4
6731.1
6730.8
6730.6
6730.3
6730.1
ABSORBED
NON-AQUEOUS
LIQUIDS
1237.6
1236.5
1235.7
1235.1
1234.6
1234.1
1233.8
1233.5
1233.2
1233.0
1232.8
1232.6
1Z32.4
1232.3
1232.1
1232.0
1231.9
1231.8
1231.7
1231.6
1231.6
1231.5
1231.4
1231.4
1231.3
1231.3
1231.2
1231.2
1231.1
1231.1
NON-AQUEOUS
LIQUIDS IN
VIALS
2970.4
2967.8
2965.8
2964.3
2963.0
2962.0
2961.1
2960.4
2959.7
2959.2
2958.7
2958.3
2957.9
2957.5
2957.2
2956.9
2956.7
2956.4
2956.2
2956.0
2955.8
2955.7
2955.5
2955.4
2955.2
2955.1
2955.0
2954.9
2954.7
2954.6
ANIMAL
CARCASSES
IN LIKE
2887.8
2885.3
2883.4
2881.9
2880.7
2879.7
2878.9
2878.1
2877.5
2877.0
2876.5
2876.1
2875.7
2875.4
2875.1
2874.8
2874.5
2874.3
2874.1
2873.9
2873.7
2873.6
2873.4
2873.3
2873.1
2873.0
2872.9
2872.8
2872.7
2872.6
TOTALS
82511.1
82439.9
82385.4
82342.4
82307.6
82278.9
82254.8
82234.2
82216.5
82201.0
82187.5
82175.4
82164.7
82155.1
82146.4
82138.5
82131.4
82124.8
82118.7
82113.2
82108.1
82103.3
82098.9
82094.8
82090.9
82087.3
82083.9
82080.7
82077.7
82074.8
TOTAL 1595197.0
431466.1
24653.7 202171.8
36981.4
88757.4
86291.8 2465527.9
106
-------
TABLE 7
FUEL CYCLE GENERATOR
ANNUAL WASTE STREAM ACTIVITIES
** all volumes ,are in cubic rnillicuries **
YEAR
1991
1992
1993
1994
1995
1996
1997
1998
1999
2000
2001
2002
2003
2004
2005
2006
2007
2008
2009
2010
2011
2012
2013
20U
2015
2016
2017
2018
2019
2020
DRY ACTIVE
HASTES (DAW)
141852.4
141296.0
140873.2
140540.9
140272.9
140052.2
139867.3
139710.0
139574.7
139457.0
139353.8
139262.4
139180.9
139108.0
139042.1
138982.5
138928.1
125548.5
119476.3
119440.1
119406.7
119375,8
119347.0
119320.2
119295.1
119271.7
119248.7
119228.0
119208.6
119190.2
SOLIDIFIED
LIQUIDS
9497.8
9460.5
9432.2
9410.0
9392.0
9377.3
9364.9
9354.4
9345.3
9337.4
9330.5
9324.4
9318.9
9314.0
9309.6
9305.6
9302.0
8406.2
7999.3
7996.8
7994.6
7992.5
~~ 7990.6,
7988.8
7987.1
7985.6
7984.1
7982.7
7981.4
7980.2
FILTER
MEDIA
20951.0
20868.8
20806.4
20757.3
20717.7
20685.1
20657.8
20634.6
20614.6
20597.2
20582.0
20568.5
20556.5
20545.7
20536.0
20527.1
20519.1
18543.0
17645.4
17640.1
17635.1
17630.6
17626.3
17622.4
17618.7
17615.2
17612.0
17608.9
17606.0
17603.3
DEUATERED
RESINS
246216.5
245250.7
244516.8
243940.1
243474.9
243091.8
242770.8
242497.9
242263.1
242058.8
241879.5
241720.8
241579.5
241452.8
241338.6
241235.0
241140.7
217917.4
207369.0
207306.3
207248.3
207194.6
207144.6
207098.1
207054.6
207013.9
206975.7
206939.8
206906.0
206874.1
SOLIDIFIED
RESINS
139114.8
138569.2
138154.5
137828.6
137565.8
137349.4
137168.0
137013,8
136881.1
136765.7
136664.4
136574.7
136494.9
136423.3
136358.8
136300.3
136247.0
123125.6
117165.6
117130.2
117097.4
117067.1
117038.8
117012.5
116987.9
116965.0
116943.4
116923.1
116904.0
116886.0
TOTALS
557632.5
555445.2
553783.1
552476.9
551423.3
550555.8
549828.8
549210.7
548678.8
548216.1
547810.2
547450.8
547130.7
546843.8
546585.1
546350.5
546136.9
493540.7
469655.6
469513.5
469382.1
469260.6
469147.3
469042.0
468943.4
468851.4
468763.9
468682.5
468606.0
468533.8
TOTAL 3934711.3
263446.7
581132.4
6829470.7
3858720.9
15467482.0
107
-------
TABLE 8
«
• NON-FUEL CYCLE GENERATOR
ANNUAL WASTE STREAM ACTIVITIES
•** all volumes are in cubic millicuries **
YEAR
1991
1992
1993
1994
1995
1996
1997
1998
1999
2000
2001
2002
2003
2004
2005
2006
2007
2008
2009
2010
2011
2012
2013
2014
2015
2016
2017
2018
2019
2020
DRY ACTIVE
UASTES (DAW)
828066.8
828999.5
829713.6
830278.0
830735.3
831113.5
831431.3
831702.2
831936.0
832139.7
832318.7
832477.3
832618.9
832745.9
832860.7
832964.7
833059.6
833146.4
833226.0
833299.5
833367.5
833430.5
833489.1
833543.7
833594.8
833642.6
833687.5
833729.7
833769.5
833807.1
SOLIDIFIED
LIQUIDS
21421.7
21445.9
21464.3
21478.9
21490.8
21500.5
21508.8
21515.8
21521.8
21527.1
21531.7
21535.8
21539.5
21542.8
21545.7
21548.4
21550.9
21553.1
21555.2
21557.1
21558.9
21560.5
21562.0
21563.4
21564.7
21566.0
21567.1
21568.2
21569.3
21570.2
SOLIDIFIED
RESINS
61491.2
61560.4
61613.5
61655.4
61689.3
61717.4
61741.0
61761.1
61778.5
61793.6
61806.9
61818.7
61829.2
61838.6
61847.2
61854.9
61861.9
61868.4
61874.3
61879.8
61884.8
61889.5
61893.8
61897.9
61901.7
61905.2
61908.6
61911.7
61914.7
61917.4
ABSORBED
AQUEOUS 1
LIQUIDS
23714.7
23741.4
23761.8
23778.0
23791.1
23801.9
^23811.0
"23818.8
23825.5
; 23831.3
23836.4
23841.0
23845.0
23848.7
23852.0
23854.9
23857.7
23860.1
23862.4
23864.5
23866.5
23868.3
23870.0
23871.5
23873.0
23874.4
23875.6
23876.8
23878.0
23879.1
ABSORBED I
(OH- AQUEOUS
LIQUIDS
5416.0
5422.1
5426.8
5430.5
5433.5
5436.0
5438.0
5439.8
5441.3
5442.7
5443.8
5444.9
5445.8
5446.6
5447.4
5448.1
5448.7
5449.3
5449.8
5450.3
5450.7
5451.1
5451.5
5451.9
5452.2
5452.5
5452.8
5453.1
5453.3
5453.6
ION-AQUEOUS
LIQUIDS IN
VIALS
189.1
189.3
189.4
189.6
189.7
189.7
189.8
189.9
189.9
190.0
190.0
190.1
190.1
190.1
190.1
190.2
190.2
190.2
190.2
190.2
190.3
190.3
190.3
190.3
190.3
190.3
190.3
190.3
190.4
190.4
ANIMAL
CARCASSES
IN LINE
470.8
471.3
471.7
472.0
472.3
477.5
472.7
472.8
473.0
473.1
473.2
473.3
473.3
473.4
473.5
473.5
473.6
473.6
473.7
473.7
473.8
473.8
473.8
473.9
473.9
473.9
474.0
474.0
474.0
474.0
TOTALS
935896.1
936950.2
937757.3
938395.2
938912.1
939339.5
939698.7
940004.9
940269.1
940499.3
940701.6
940880.9
941040.9
941184.5
941314.2
941431.8
941539.0
941637.1
941727.1
941810.2
941887.0
941958.2
942024.4
942086.2
942143.9
942197.9
942248.7
942296.4
942341.4
942383.8
TOTAL 24970895.6 645986.1 1854306.6
715131.4 163324.1
5701.0 14196.1 28222557.6
108
-------
TABLE 9
MAJOR NUCLIDES CONSIDERED FOR ANALYSIS
MUCLIDE
ACTIVITY
PZRCEXTT
AM-241
C-14
CM-243
CM-244
CO-60
CS-137
FE-55
H -3
I -129
NB-94
NI-59
NI-63
NP-237
PU-238
PU-239
PU-240
FU-241
RA-226
RN-222
SR-90
TC-99
TH-231
TH-234
U -235
U -238
TOTALS
1.660E+0
2.746E+2
2.480E-2
1.433E-1
1.831E-I-4
9.702E+3
3.224E+3
4.811E+6
9.888E+0
1.443E-1
1.793E+1
1.463E+3
1.900E-3
8.317E-1
4.467E-1
4.207E-1
5.098E+1
1.610E+1
1.610E+1
2.245E-H3
2.768E+0
2.757E+0
1.726E+2
2.757E+0
1.726E+2
4.844E+6
0.4
0.2
<0.1
99.3
99.9
All activities are given in curies.
109
-------
TABLE 10
FUEL CYCLE GENERATOR
RADIONUCLIDE CONCENTRATION BREADKOWN BY WASTE STREAM
** all concentrations are in microcuries per cubic centimeter **
RADIO-
MUCLIDE
AC- 11 OH
AH-241
IA-140
BE-7
•1-207
C-14
CO-109
CE-141
CE-144
CM-242
CM-243
CM-244
CO-57
CO-58
CO-60
CX-51
CS-134
CS-136
CS-137
EU-154
EU-155
FE-55
FE-59
H-3
1-129
1-131
1-133
KK-85
LA- 140
UN -54
m-99
ww
HA-24
nn fc^
M-94
v^
Nt-95
«-S9
+ r
HI-63
NI-65
MP-237
ttr *•*•
M-210
MY ACTIVE
WASTES (DAW)
1.049E-02
O.OOOE+OO
O.OOOE+OO
1.484E-06
o.oooe+oo
4.526E-04
O.OOOE+OO
8.904E-06
7.123E-05
1.484E-06
o.oooe+oo
O.OOOE+OO
8.904E-06
3.300E-02
1.468E-02
7.806E-03
4.541E-04
O.OOOE+OO
1.732E-03
O.OOOE+OO
O.OOOE+OO
2.667E-02
2.003E-04
H.046E-03
4.452E-06
1.677E-04
O.OOOE+OO
O.OOOE+OO
O.OOOE+OO
1.658E-03
O.OOOE+OO
U.OOOE+00
O.OOOE+OO
3.491E-03
O.OOOE+OO
2.553E-02
O.OOOE+OO
O.OOOE+OO
O.OOOE+OO
SOLIDIFIED
LIQUIDS
3.828E-OS
4.283E-09
5.112E-09
4.725E-08
O.OOOE+OO
1.313E-05
O.OOOE+OO
8.290E-10
3.454E-09
3.689E-08 ,
4.283E-09
4.283E-09
1.255E-05
1.359E-03
9.860E-04
O.OOOE+OO
1.514E-03
O.OOOE+OO
4.184E-03
O.OOOE+OO
O.OOOE+OO
1.327E-03
1.133E-07
1.833E-02
6.356E-08
4.6S4E-05
1.658E-09
O.OOOE+OO
5.748E-08
1.412E-04
1.6S8E-09
9.3S4E-08
O.OOOE+OO
4.463E-08
O.OOOE+OO
9.950E-04
O.OOOE+OO
O.OOOE+OO
O.OOOE+OO
FILTER
MEDIA
1.673E-05
O.OOOE+OO
1.127E-05
5.649E-06
O.OOOE+OO
2.034E-03
2.552E-08
1.782E-04
6.918E-04
5.427E-07
O.OOOE+OO
O.OOOE+OO
,5.243E-04
2.120E-01
9.395E-02
1.820E-01
1.847E-03
6.023E-06
7.273E-03
O.OOOE+OO
O.OOOE+OO
3.841E-02
1.459E-02
1.410E-03
2.077E-05
7.137E-04
O.OOOE+OO
O.OOOE+OO
O.OOOE+OO
2.714E-02
O.OOOE+OO
O.OOOE+OO
2.552E-08
3.250E-02
4.811E-05
3.307E-02
O.OOOE+OO
O.OOOE+OO
O.OOOE+OO
OEWATERED
tESINS
O.OOOE+OO
1.316E-08
O.OOOE+OO
1.316E-08
O.OOOE+OO
1.897E-04
O.OOOE+OO
1.841E-05
1.02SE-03
3.307E-08
1.316E-08
1.316E-08
5.949E-04
3.721E-01
2.085E-01
9.172E-06
3.422E-01
8.303E-07
7.966E-01
O.OOOE+OO
O.OOOE+OO
1.071E-01
7.019E-03
4.949E-03
«.754E-06
2.675E-05
2.133E-09
O.OOOE+OO
1.43SE-06
8.546E-02
O.OOOE+OO
7.688E-06
O.OOOE+OO
2.750E-04
6.3S4E-04
6.492E-02
O.OOOE+OO
2.133E-09
O.OOOE+OO
SOLIDIFIED
RESINS
5.680E-04
7.767E-07
O.OOOE+OO
2.131E-03
O.OOOE+OO
1 .754E-03
O.OOOE+OO
O.OOOE+OO
1.721E-05
1.121E-06
1.769E-09
6.537E-07
2.149E-04
6.784E-02
9.206E-02
9.489E-04
1.100E-01
O.OOOE+OO
4.028E-01
O.OOOE+OO
O.OOOE+OO
1.474E-01
9.94SE-03
1.514E-02
1.683E-06
1.219E-03
O.OOOE+OO
O.OOOE+OO
1.142E-04
3.411E-02
O.OOOE+OO
O.OOOE+OO
O.OOOE+OO
7.147E-04
6.478E-05
9.501E-02
O.OOOE+OO
O.OOOE+OO
O.OOOE+OO
TOTALS
1.111E-02
7.941E-07
1.128E-05
2.13&E-03
O.OOOE+OO
4.443E-03
2.S52E-OB
2.056E-04
1.805E-03
3.217E-06
1.921E-08
6.712E-07
1.355E-03
6.863E-01
4.101E-01
1.908E-01
4.560E-01
6.8S4E-06
1.213E+00
O.OOOE+OO
O.OOOE+OO
3.210E-01
3.175E-02
4.088E-02
3.372E-05
2.173E-03
3.791E-09
O.OOOE+OO
1.157E-04
1.48SE-01
1.658E-09
7.781E-06
2.552E-08
3.696E-02
7.483E-04
2.195E-01
O.OOOE+OO
2.133E-09
O.OOOE+OO
110
-------
TABLE 10(con't)
RADIO-
MUCUDE
WJ-238
PO-239
W-240
WJ-241
HO- 103
RU-106
S8-122
SB- 124
S8-12S
SN-113
SR-89
S8-90
TA-182
TC-99
TC-99M
TE-125M
XE-131H
XE-133
ZN-6S
ZR-95
DRY ACTIVE
WASTES
1.484E-06
O.OOOE+00
0.0006*00
1.172E-04
O.OOOE+00
3.265E-05
." O.OOOE+00
1.619E-03
2.S20E-05
, O.OOOE+00
O.OOOE+00
5.936E-06
O.OOOE+00
.. 8.904E-06
1 .039E-05
1.4ME-06
O.OOOE+00
O.OOOE+00
1.197E-05
4.912E-W
SOLIDIFIED
LIQUIDS
3.261E-08
2.0S9E-08
4.283E-09
2.780E-06
O.OOOE+00
O.OOOE+00
1.658E-09
2.297E-03
1.726E-04
O.OOOE+00
O.OOOE+00
1.037E-05
O.OOOE+00
1.064E-07
8.290E-10
1.072E-05
.5.112E-09
"1.658E-09
8.290E-10
2l625E-09
FILTER
MEDIA
2.5S2E-08
5.776C-08
4.970E-08
2.080E-04
1. 5786-04
1.2686-05
O.OOOE+00
4.800E-03
7.208E-04
1.938E-05
2.312E-06
2.554E-05
O.OOOE+00
4.561E-05
O.OOOE+00'
O.OOOE+00
1.143E-05 ?
2.934E-05
2.367E-03
1.918E:02
DEUATERED
RES IMS
1.3166-08
1 .3166-08
1.316E-08
T.794E-04
O.OOOE+00
O.OOOE+00
3.092E-06
1.062E-01
2.131E-04
O.OOOE+00
1.104E-03
1.576E-03
O.OOOE+00
8.424E-06
2.133E-09
0;600E+00 *
O.OOOE+00
O.OOOE+00
1.119E-04
1.339E-04
'>', : i- .
SOLIDIFIED
RESINS
2.2786-06
1.9676-06
1.769E-09
2.945E-04
O.OOOE+00
O.OOOE+00
O.OOOE+00
1.5586-04
8.055E-OS
O.OOOE+00
2.8266-06
1.031E-03
O.OOOE+00
2.209E-06
O.OOOE+00
O.OOOE+00
O.OOOE+00
O.OOOE+00
1.363E-04
3.349E-04
TOTALS
3.834E-06
2.059E-06
6.891E-08
8.020E-04
1.578E-04
4.533E-05
3.094E-06
1.151E-01
1.2156-03
1.9386-05
•1.109E-03
2.649E-03
O.OOOE+00
6.5266-05
1.039E-05
1.220E-05
1.144E-05
2.934E-05
2.627E-03
2.014E-02
111
-------
TABLE 11
NON-FUEL CYCLE GENERATOR
RADIONUCLIDE CONCENTRATION BREADKOWN BY WASTE STREAM
**
all concentrations are in microcuries per cubic centimeter **
AtSORIED ABSORBED NON-AQUEOUS ANIMAL
RADIO- DXY ACTIVE SOLIDIFIED SOLIDIFIED AQUEOUS NON-AQUEOUS LIQUIDS IN CARCASSES
NUCLIDE WASTES (DAW) LIQUIDS RESINS LIQUIDS LIQUIDS VIALS IN LIME TOTALS
AC-227
AC-22B
AG-108M
AG-110M
AL-26
AK-241
AU-195
•A- 133
BA-140
BE-7
•I -205
11-207
•8-82
C-14
CA-4S
CD-109
CE-139
CE-141
CE-144
CF-252
CL-36
CM-242
CM-243
04-244
CO-57
CO-58
CO-60
01-51
CS-134
CS-136
CS-137
CS-141
CU-64
CO-67
EU-152
EU-154
EU-155
FE-55
2.101E-07
4.51SE-09
2.600E-06
1.319E-06
0. 0006+00
2.298E-05
1.197E-05
9.329E-07
O.OOOE+00
1.694E-07
8.358E-07
2.666E-07
O.OOOE+00
2.138E-03
3.352E-05
3.636E-05
4.518E-OV
1.276E-06
4.554E-06
4.51SE-09
1.324E-06
2.711E-08
5.602E-07
2.950E-06
5.963E-04
1.582E-05
3.726E-01
3.0S3E-04
3.683E-05
o.oooe+oo
9.416E-02
8.35SE-07
9.036E-09
2.101E-07
1.145E-04
1.520E-05
4.491E-05
2.556E-02
O.OOOE+00
O.OOOE+00
O.OOOE+00
O.OOOE+00
O.OOOE+00
3.936E-06
4.092E-08
8.184E-09
O.OOOE+00
8.184E-09
O.OOOE+00
O.OOOE+00
O.OOOE+00
6.177E-03
2.782E-06
4.062E-05
O.OOOE+00
3.273E-08
8.559E-05
O.OOOE+00
4.665E-07
3.273E-08
O.OOOE+00
1.473E-07
3.110E-07
3.368E-OS
9.566E-03
2.204E-05
2.561E-04
O.OOOE+00
1.381E-02
O.OOOE+OO
O.OOOE+00
O.OOOE+00
1.447E-03
6.211E-05
8.090E-04
1.773E-02
O.OOOE+00
O.OOOE+00
O.OOOE+00
2.635E-05
O.OOOE+00
8.157E-04
O.OOOE+00
O.OOOE+00
1.289E-06
1.432E-07
O.OOOE+00
O.OOOE+00
O.OOOE+00
8.327E-04
O.OOOE+00
O.OOOE+00
O.OOOE+00
O.OOOE+00
6.771E-03
O.OOOE+00
O.OOOE+00
1.432E-07
O.OOOE+00
O.OOOE+00
O.OOOE+00
4.085E-03
1.478E-01
8.031E-04
3.339E-02
2.B64E-07
1. 5716+00
O.OOOE+00
O.OOOE+00
O.OOOE+00
6.144E-02
2.246E-03
3.406E-02
1. 3406-01
O.OOOE+00
O.OOOE+00
O.OOOE+00
4.838E-06
2.712E-05
6.966E-08
O.OOOE+00
O.OOOE+00
O.OOOE+00
1.747E-08
O.OOOE+00
O.OOOE+00
O.OOOE+00
1.277E-02
8.462E-05
3.423E-06
O.OOOE+00
8.034E-07
5.764E-07
O.OOOE+00
1.167E-05
O.OOOE+00
O.OOOE+00
O.OOOE+00
1.039E-03
2.253E-06
2.876E-03
1.429E-03
1.104E-05
O.OOOE+00
8.696E-04
O.OOOE+00
O.OOOE+00
1.799E-06
1.163E-05
9.0B2E-07
5.921E-06
3.663E-04
O.OOOE+00
O.OOOE+00
O.OOOE+00
9.548E-08
O.OOOE+00
O.OOOE+00
O.OOOE+00
O.OOOE+00
O.OOOE+00
9.S48E-08
O.OOOE+00
O.OOOE+00
O.OOOE+00
4.207E-03
O.OOOE+00
O.OOOE+00
O.OOOE+00
O.OOOE+00
O.OOOE+00
O.OOOE+00
4.392E-06
O.OOOE+00
O.OOOE+00
O.OOOE+00
8.593E-07
1.685E-04
1.110E-03
1.038E-04
1.719E-05
O.OOOE+00
4.923E-04
O.OOOE+00
O.OOOE+00
O.OOOE+00
1.012E-05
9.548E-08
2.482E-06
9.684E-04
O.OOOE+00
O.OOOE+00
O.OOOE+00
O.OOOE+00
O.OOOE+00
O.OOOE+00
O.OOOE+00
O.OOOE+00
O.OOOE+00
O.OOOE+00
O.OOOE+00
O.OOOE+00
O.OOOE+00
4.884E-04
2.029E-06
2.387E-07
O.OOOE+00
4.336E-06
O.OOOE+00
O.OOOE+00
3.819E-06
O.OOOE+00
O.OOOE+00
O.OOOE+00
8.828E-OS
O.OOOE+00
3.501E-06
6.970E-05
O.OOOE+00
O.OOOE+00
1.472E-06
O.OOOE+00
O.OOOE+00
O.OOOE+00
O.OOOE+00
O.OOOE+00
O.OOOE+00
3.978E-07
O.OOOE+00
O.OOOE+00
O.OOOE+00
O.OOOE+00
O.OOOE+00
O.OOOE+00
O.OOOE+00
O.OOOE+00
O.OOOE+00
O.OOOE+00
O.OOOE+00
O.OOOE+00
O.OOOE+00
8.405E-04
2.881E-04
1.113E-05
O.OOOE+00
3.364E-05
O.OOOE+00
O.OOOE+00
6.424E-06
O.OOOE+00
O.OOOE+00
O.OOOE+00
2.476E-05
O.OOOE+00
O.OOOE+00
2.251E-04
O.OOOE+00
O.OOOE+00
O.OOOE+00
4.S01E-07
O.OOOE+00
5.189E-05
O.OOOE+00
O.OOOE+00
O.OOOE+00
1.SS7E-07
2.101E-07
4.518E-09
2.600E-06
3.261E-05
2.712E-05
8.426E-04
1.201E-05
9.411E-07
1.289E-06
4.338E-07
8.358E-07
2.666E-07
O.OOOE+00
2.746E-02
4.110E-04
9.178E-05
4.518E-09
4.008E-05
6.862E-03
4.518E-09
2.809E-05
2.031E-07
5.602E-07
3.097E-06
1.7SOE-03
4.306E-03
5.339E-01
2.961E-03
3.371E-02
2.864E-07
1.681E+00
1.286E-06
9.036E-09
5.389E-05
6.302E-02
2.324E-03
3.492E-02
1.786E-01
112
-------
TABLE 11 (con't)
AiSORBED ABSORBED NON-AQUEOUS AHINAL
IADIO- DRY ACTIVE SOLIDIFIED SOLIDIFIED AQUEOUS NON-AQUEOUS LIQUIDS IN CARCASSES
MUCLIDE IMSTES (DAW) LIQUIDS RESINS LIQUIDS LIQUIDS VIALS IN LIME
TOTALS
FE-S9
SA-67
6D-153
GE-68
HF-175
HF-181
HG-203
MO-166H
-123
-124
-125
-129
-131
-135
IN-111
IN-113M
IN-114
IN-114M
IR-192
r-40
K-42
KR-85
LA- 140
MN-54
MN-56
NO-99
NA-22
NA-24
NB-94
N8-95
NI-59
NI-63
HI -65
KP-237
NP-239
P-32
r-33
M-233
M-210
M-212
HI- 147
PO-210
PT-195
PU-238
1.022E-05
1.543E-05
3.727E-07
2.553E-07
O.OOOE+00
O.OOOE+00
3.434E-07
5.568E-06
7.073E-06
8.584E-08
6.153E-03
7.981E-06
1.105E-04
6.619E-07
2.808E-05
O.OOOE+00
O.OOOE+00
3.501E-07
1.204E-06
9.962E-07
O.OOOE+00
3.693E-04
O.OOOE+00
5.944E-05
6.777E-08
4.706E-05
9.074E-06
1.807E-08
3.115E-06
9.036E-07
3.482E-04
9.S32E-04
O.OOOE+00
4.292E-08
O.OOOE+00
5.334E-03
4.021E-07
O.OOOE+00
7.632E-OS
4.518E-09
5.571E-04
3.638E-OS
3.011E-06
1.177E-05
O.OOOE+00
S.273E-08
O.OOOE+00
O.OOOE+00
O.OOOE+00
5.729E-08
8.184E-09
O.OOOE+00
1.515E-05
O.OOOE+00
2.63C€-04
6.203E-06
3.026E-05
O.OOOE+00
O.OOOE+00
O.OOOE+00 ,
O.OOOE+00
O.OOOE+00
O.OOOE+00
O.OOOE+00
O.OOOE+00
3.026E-04
8.184E-09
3.7S6E-05
9.002E-08
O.OOOE+00
1.274E-05
1.555E-06
O.OOOE+00
O.OOOE+00
4.239E-06
9.142E-04
O.OOOE+00
O.OOOE+00
O.OOOE+00
2.08SE-04
O.OOOE+00
O.OOOE+00
O.OOOE+00
O.OOOE+00
O.OOOE+00
O.OOOE+00
O.OOOE+00
2.030E-06
3.390E-04
O.OOOE+00
O.OOOE+00
O.OOOE+00
6.359E-05
3.655E-04
O.OOOE+00
O.OOOE+00
O.OOOE+00
O.OOOE+00
o.oooe+oo
1.331E-02
6.87SE-06
O.OOOE+00
O.OOOE+00
O.OOOE+00
O.OOOE+00
O.OOOE+00
O.OOOE+00
O.OOOE+00
O.OOOE+00
o.oooe+oo
1.432E-07
2.772E-03
O.OOOE+00
O.OOOE+00
O.OOOE+00
O.OOOE+00
O.OOOE+00
1.905E-05
1.4096-04
2.170E-02
O.OOOE+00
0.0006+00
O.OOOE+00
O.OOOE+00
O.OOOE+00
O.OOOE+00
O.OOOE+00
O.OOOE+00
O.OOOE+00
O.OOOE+00
O.OOOE+00
3.506E-04
3.221E-05
3.423E-06
1.799E-06
9.082E-07
O.OOOE+00
O.OOOE+00
2. 6206-07
O.OOOE+00
0.0006+00
O.OOOE+00
1.328E-02
3.493E-07
1.419E-03
O.OOOE+00
2.152E-05
O.OOOE+00
O.OOOE+00
8.733E-08
1.939E-06
3.877E-06
6.462E-06
o.oooe+oo
1.747E-08
3. 9306-06
O.OOOE+00
6.462E-06
4.1106-05
O.OOOE+00
1.1006-06
8.733E-08
6.637E-07.
3.706E-05
O.OOOE+00
O.OOOE+00
O.OOOE+00
8.966E-02
1.1406-05
O.OOOE+00
O.OOOE+00
O.OOOE+00
O.OOOE+00
O.OOOE+00
O.OOOE+00
2.270E-07
1.442E-05
0.0006+00
O.OOOE+00
O.OOOE+00
O.OOOE+00
O.OOOE+00
O.OOOE+00
5.729E-07
O.OOOE+00
7.066E-06
1.774E-04
9.548E-08
1.719E-05
O.OOOE+00
, 1.766E-05
O.OOOE+00
O.OOOE+00
O.OOOE+00
O.OOOE+00
o.oooe+oo
O.OOOE+00
7.066E-03
O.OOOE+00
2.941E-05
O.OOOE+00
O.OOOE+OO
9.548E-07
O.OOOE+00
O.OOOE+00
2.492E-05
O.OOOE+00
1.186E-04
O.OOOE+00
O.OOOE+00
O.OOOE+00
6.0586-04
O.OOOE+00
0.0006+00
O.OOOE+00
O.OOOE+00
O.OOOE+00
O.OOOE+00
O.OOOE+00
O.OOOE+00
7.957E-08
4.177E-06
O.OOOE+00
O.OOOE+00
O.OOOE+00
O.OOOE+00
O.OOOE+00
• O.OOOE+00
O.OOOE+00
5.888E-06
5.581E-04
O.OOOE+00
2.665E-06
O.OOOE+00
1.671E-06
O.OOOE+00
O.OOOE+00
O.OOOE+00
O.OOOE+00
O.OOOE+00
3.9786-08
O.OOOE+00
O.OOOE+00
O.OOOE+00
O.OOOE+00
1 .472E-05
5.172E-07
O.OOOE+00
O.OOOE+00
O.OOOE+00
O.OOOE+00
O.OOOE+00
O.OOOE+00
O.OOOE+00
o.oooe+oo
9.5306-04
5.172E-07
O.OOOE+00
3.5806-07
O.OOOE+00
O.OOOE+00
O.OOOE+00
O.OOOE+00
O.OOOE+00
6.297E-05
4.321E-05
3.237E-05
O.OOOE+00
O.OOOE+00
O.OOOE+00
O.OOOE+00
O.OOOE+00
4.189E-04
1.228E-07
2.294E-03
O.OOOE+00
1.032E-04
O.OOOE+00
3.256E-04
3.339E-05
9.002E-07
1.473E-05
O.OOOE+00
O.OOOE+00
O.OOOE+00
O.OOOE+00
O.OOOE+00
4.9106-07
O.OOOE+00
O.OOOE+00
1.526E-05
O.OOOE+00
O.OOOE+00
3.564E-05
O.OOOE+00
O.OOOE+00
O.OOOE+00
o.oooe+oo
O.OOOE+OO
8.008E-05
O.OOOE+00
O.OOOE+00
O.OOOE+00
O.OOOE+00
O.OOOE+00
1.228i-07
O.OOOE+00
O.OOOE+00
4.5896-04
6.628E-05
3.454E-05
1.163E-06
6.359E-05
3.656E-04
6.135E-07
6.141E-06
4.411E-04
1.316E-05
2.273E-02
1.332E-02
1.689E-03
6.619E-07
3.946E-04
3.339E-05
9.002E-07
1.517E-05
3.143E-06
4.873E-06
6.502E-06
7.738E-03
1.689E-07
2.903E-03
1.578E-07
6.825E-05
7.965E-05
1.573E-06
4.215E-06
8.060E-05
4.940E-04
2.372E-02
O.OOOE+00
4.292E-08
O.OOOE+00
9.684E-02
1.232E-05
O.OOOE+00
7.668E-05
4.518E-09
5.571E-04
3.6506-05
S.011E-06
3.646E-04
113
-------
TABLE 11 (con't)
RAD 10-
NUCLIDE
K-239
K-240 •
PO-241
RA-226
RA-228
Ri-83
M-86
RE- 184
RU-103
RU-10S
RU-106
sa-122
$•-124
tt-125
SC-46
SC-47
SE-75
SI -32
SM-113
S«-119m
SN-126
SR-85
SR-89
5R-90
SR-91
TA-182
TI-160
TC-99
TC-99M
TE-123H
TH-228
TH-230
TH-232
TL-201
TL-204
TL-208
U-233
U-234
U-23S
U-238
XE-131N
XE-133
Y-88
Y-90
DRY ACTIVE
UASTES (DAW)
5.390E-06
5.453E-06
5.S33E-04
3.636E-04
8.109E-07
1.588E-06
9.241E-06
2.259E-09
1.100E-06
O.OOOE+00
6.555E-06
O.OOOE+00
3.161E-03
1.332E-05
1.1B6E-06
O.OOOE+00
4.235E-06
O.OOOE+00
4.427E-06
9.307E-07
2.736E-0&
1.319E-06
3.840E-08
4.5S8E-02
4.604E-06
7.387E-07
O.OOOE+00
4.978E-05
1.086E-05
2.047E-06
8.652E-07
2.48SE-08
3.689E-06
9.562E-06
5.376E-07
6.777E-09
O.OOOE+00
1.751E-06
2.580E-OS
3.8S8E-03
O.OOOE+00
5.674E-06
9.442E-07
1.118E-04
SOLIDIFIED
LIQUIDS
8.184E-09
O.OOOE+00
4.494E-04
O.OOOE+00
O.OOOE+00
O.OOOE+00
1.097E-06
O.OOOE+00
O.OOOE+00
O.OOOE+00
O.OOOE+00
O.OOOE+00
6.244E-06
O.OOOE+00
O.OOOE+00
O.OOOE+00
2.954E-06
O.OOOE+00
O.OOOE+00
O.OOOE+00
O.OOOE+00
4.910E-08
2.660E-06
1.794E-04
O.OOOE+00
O.OOOE+00
1.637E-08
7.365E-07
O.OOOE+00
O.OOOE+00
O.OOOE+00
O.OOOE+00
1.620E-06
O.OOOE+00
O.OOOE+00
O.OOOE+00
O.OOOE+00
O.OOOE+00
1.141E-04
1.221E-04
O.OOOE+00
O.OOOE+00
O.OOOE+00
1.372E-04
SOLIDIFIED
RESINS
2.569E-04
2.569E-04
2.209E-02
O.OOOE+00
3.180E-05
O.OOOE+00
O.OOOE+00
O.OOOE+00
O.OOOE+00
O.OOOE+00
O.OOOE+00
O.OOOE+00
5.299E-05
2.721E-06
O.OOOE+00
O.OOOE+00
O.OOOE+00
O.OOOE+00
5.299E-06
O.OOOE+00
O.OOOE+00
O.OOOE+00
O.OOOE+00
2.982E-01
O.OOOE+00
2.650E-05
O.OOOE+00
6.717E-05
O.OOOE+00
O.OOOE+00
O.OOOE+00
O.OOOE+00
O.OOOE+00
O.OOOE+00
O.OOOE+00
O.OOOE+00
O.OOOE+00
O.OOOE+00
1.432E-07
1.289E-06
O.OOOE+00
O.OOOE+00
O.OOOE+00
2.982E-01
ABSORBED
AQUEOUS
LIQUIDS
1.747E-08
O.OOOE+00
8.925E-06
6.986E-08
O.OOOE+00
O.OOOE+00
7.021E-06
O.OOOE+00
O.OOOE+00
O.OOOE+00
1.572E-07
O.OOOE+00
O.OOOE+00
5.764E-07
O.OOOE+00
O.OOOE+00
6.991E-05
O.OOOE+00
8.209E-07
O.OOOE+00
O.OOOE+00
1.589E-06
3.249E-06
5.68SE-04
O.OOOE+00
O.OOOE+00
O.OOOE+00
1.327E-05
O.OOOE+00
O.OOOE+00
8.733E-08
5.240E-08
1.048E-07
O.OOOE+00
O.OOOE+00
O.OOOE+00
O.OOOE+00
1.921E-07
1.27SE-06
3.727E-05
O.OOOE+00
O.OOOE+00
O.OOOE+00
S.292E-06
ABSORBED
NON-AQUEOUS
LIQUIDS
O.OOOE+00
O.OOOE+00
1.929E-05
O.OOOE+00
O.OOOE+00
O.OOOE+00
2.473E-05
O.OOOE+00
O.OOOE+00
O.OOOE+00
O.OOOE+00
O.OOOE+00
O.OOOE+00
9.S48E-08
O.OOOE+00
O.OOOE+00
3.S33E-06
O.OOOE+00
O.OOOE+00
O.OOOE+00
O.OOOE+00
O.OOOE+00
O.OOOE+00
6.493E-05
O.OOOE+00
O.OOOE+00
O.OOOE+00
2.397E-05
O.OOOE+00
O.OOOE+00
O.OOOE+00
O.OOOE+00
9.S48E-08
O.OOOE+00
O.OOOE+00
O.OOOE+00
O.OOOE+00
O.OOOE+00
2.026E-04
7.3335-05
O.OOOE+00
O.OOOE+00
O.OOOE+00
2.005E-06
NON-AQUEOUS
LIQUIDS IN
VIALS
O.OOOE+00
O.OOOE+00
O.OOOE+00
O.OOOE+00
O.OOOE+00
O.OOOE+00
2.773E-05
O.OOOE+00
O.OOOE+00
O.OOOE+00
O.OOOE+00
O.OOOE+00
O.OOOE+00
O.OOOE+00
O.OOOE+00
O.OOOE+00
O.OOOE+00
O.OOOE+00
4.336E-06
O.OOOE+00
O.OOOE+00
4.336C-06
O.OOOE+00
2.387E-07
O.OOOE+00
O.OOOE+00
O.OOOE+00
5.291E-06
1.030E-05
O.OOOE+00
O.OOOE+00
O.OOOE+00
O.OOOE+00
3.342E-06
O.OOOE+00
O.OOOE+00
O.OOOE+00
O.OOOE+00
O.OOOE+00
O.OOOE+00
O.OOOE+00
9.866E-06
O.OOOE+00
O.OOOE+00
ANIMAL
CARCASSES
IN LIME
O.OOOE+00
O.OOOE+00
2.537E-06
O.OOOE+00
O.OOOE+00
O.OOOE+00
1.228E-07
O.OOOE+00
3.683E-05
7.775E-07
O.OOOE+00
O.OOOE+00
O.OOOE+00
O.OOOE+00
4.071E-05
O.OOOE+00
6.670E-06
O.OOOE+00
4.292E-05
1.211E-05
O.OOOE+00
1.91SE-05
1.363E-05
O.OOOE+00
O.OOOE+00
O.OOOE+00
O.OOOE+00
1.723E-05
3.305E-04
O.OOOE+00
O.OOOE+00
O.OOOE+00
O.OOOE+00
1.308E-04
O.OOOE+00
O.OOOE+00
O.OOOE+00
O.OOOE+00
O.OOOE+00
8.184E-08
O.OOOE+00
2.113E-04
O.OOOE+00
8.184E-06
TOTALS
2.624E-04
2.624E-04
2.312E-02
3.637E-04
3.261E-05
1.588E-06
6.994E-05
2.2S9E-09
3.793E-05
7.775E-07
6.713E-06
O.OOOE+00
3.220E-03
1.671E-05
4.190E-OS
O.OOOE+00
8.731E-05
O.OOOE+00
5.781E-05
1.304E-05
2.736E-06
2.644E-OS
1.957E-05
3.449E-01
4.604E-06
2.724E-05
1.637E-08
1.774E-04
3.517E-04
2.047E-06
9.525E-07
7.724E-08
S.509E-06
1.437E-04
5.376E-07
6.777E-09
O.OOOE+00
1.943E-06
3.440E-04
4.092E-03
O.OOOE+00
2.268E-04
9.O2E-07
2.985E-01
114
-------
TABLE 11 (con't)
AiSORBED
RADIO- DRY ACTIVE SOLIDIFIED SOLIDIFIED AQUEOUS
NUCLIDE WASTES (DAW) LIQUIDS RESINS LIQUIDS
ABSORBED NON-AQUEOUS ANIMAL
NON-AQUEOUS LIQUIDS IN CARCASSES
LIQUIDS VIALS IN LINE
TOTALS
YB-169 9.420E-07
ZN-65 8.4S3E-05
ZR-95 8.132E-08
O.OOOE+00 O.OOCE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 9.420E-07
O.OOOE+00 0.0006+00 1.231E-W O.OOOE+00 1.S91E-07 4.583E-06 2.12£E-04
O.OOOe+00 3.523E-OS 1.747E-08 1.404E-05 O.OCOE+00 O.OOOE+00 4.937E-05
115
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TABLE 12
PERCENTAGE BREAKDOWN OF WASTE
STREAM VOLUMES BY WASTE CLASSIFICATION
HASTE STREAM
WASTE CLASSIFICATION
A UNSTABLE A STABLE
B
FUEL CYCLE
DAW
Solidified
Liquids
Filter Media
Dewatered
Resins
Solidified
Resins
NON-FUI
DAW
Solidified
liquids
Solidified
Resins
Absorbed Aqueous
Liquids
Absorbed Non-
Aqueous Liquids
Non-Aqueous
Liquids in
Vials
Animals Carcasses
in Lime
94.8
61.4
64.9
67.4
63.2
;L CYCLE
96.5
84.9
100.0
100.0
100.0
100.0
100.0
1.9
38.6
20.6
4.8
12.7
2.5
. 3.6
0.0
0.0
0.0
0.0
0.0
1.9
0.0
14.5
25.5
17.3
0.4
9.8
0.0
0.0
0.0
0.0
0.0
1.4
0.0
0.0
2.3
6.8
0.6
1.7
0.0
0.0
0.0
0.0
0.0
116
-------
Scenario
TABLE 13
POTENTIAL IMPACTS ON ANNUAL WASTE VOLUME
OF INCINERATION AT REGIONAL TREATMENT
FACILITIES (Source RAE87)1
Volume of Waste
Treated in The
Disposal Facility
Current Treatment Practices
Continue for Waste That is
Currently Being Treated
All Appropriate Wastes are
Incinerated
27,000 ftd
249,000 ft3
Total Volume of
Waste for Disposal
154,000ft3
(83 percent)2
136,000ft3
(74 percent)
1.
2.
3.
The estimated annual volume of LLW in the Central Midwest Compact in the 1990's is
185,000 ft3 if projected treatment practices are followed. Projected treatment processes
are essentially the processes now in use.
Numbers shown in parenthesis state the percent of base case annual waste volume
(current treatment practices continue).
As currently treated, these wastes have a volume of about 74,000 ft3 when shipped for
disposal.
117
-------
TABLE 14
POTENTIAL IMPACTS ON ANNUAL WASTE VOLUME OF
INCINERATION AT REGIONAL TREATMENT FACILITIES (Source RAE87)1
Scenario
Current Treatment Practices
Continue for Waste That is
Currently Being Treated
All Appropriate Wastes are
Incinerated
Volume of Waste
Treated in The
Disposal Facility
27,000 ft3
247,0003 ft3
Total Volume of
Waste for Disposal
162,000 fT f
(88 percent)2
159,000 ft3-
(86 percent)
1. The estimated annual volume of LLW in the Central Midwest Compact in the 1990's is
185,000 ft3 if projected treatment practices are followed. Projected treatment processes
are essentially the processes now in use.
2. Numbers shown "in parenthesis state the percent of base case annual waste volume
(current treatment practices continue).
3. As currently treated, these wastes have a volume of about 67,000 ft3 when shipped for
disposal.
118
-------
TABLE 15
ESTIMATE OF THE ANNUAL LLRW VOLUMES GENERATED
IN THE CENTRAL INTERSTATE COMPACT
Nuclear Power Plants
Medical Facilies
Industrial Facilities
LLRW Generator Class A(ft3/Yr) Class B (ft3/Yr)
Cooper
Arkansas 1 and 2
Wolf Creek
River Bend
Waterford 3
Fort Calhoun
Generic Act. Metals
1,521 ft3/yr
3,045
Class C (ft3/Yr)Mixed (ft3/Yr)
11,343ft3/yr
7,410ft3/yr "
4,070 ft3)yr
11,738ft3/yr
7,789 ft3/yr
3,968 ft3/yr
"'.•'''.•'
•; 0' -.;•. ' '. -; •
•,'•0-' ••••:' >' .;• . •.
lesft^yr
1,090ft3/yr
240 ft3/yr
, 0 ft3/yr
seotnyr
142ft3/yr
" , •; '
0
0
0 ft3/yr 30 ft3/yr
0 tt3/yrUnknown
0 ft3/yrUnknown
0 ft3/yrUnknown
7 ft3/yrUnknown
0-ft3/yr
>n3/yr
113tt3/yr
0
TOTALS
50,875 ft3/yr
1,997ft3/yr
7ft3/yr
143tt3/yr
-------
1000
800
O
o
til
2
600
200
IMMEDIATE DISMANTLEMENT
NOTE: DASHED LINES
ARE FOR DELAYED
DISMANTLEMENT
1090 2010 2Q3O 2O5O
YEAR
2070
2090
FIGURE 1 IMPACTS OF IMMEDIATE AND DELAYED
DISMANTLEMENT OF UTILITY REACTORS
ON VOLUMES OF LLW FOR DISPOSAL IN
THE CENTRAL INTERSTATE LLRW
COMPACT
120
-------
ALL
WASTES
REACTOR
WASTES
NON-
REACTOR
WASTES
CURRENT
TREATMENT
CONTINUES
MAXIMUM
TREATMENT
CURRENT
TREATMENT
CONTINUES
MAXIMUM
TREATMENT
CURRENT
TREATMENT
CONTINUES
MAXIMUM
TREATMENT
. ' n \
% i i
* • -f<
U 1
. :.- . ,; i- r i • .. •
' i ••'' .
Mj I 'fa « WCINERATION
I 1; -..'•' i
50.000 100.000 150.000
ANNUAL WASTE VOLUME (ft3/yr)
200.000
FIGURE 2. POTENTIAL IMPACTS OF REGIONAL TREATMENT FACILITIES.
121
-------
Bench Scale Studies and Pilot Scale Design
of a Modified Volume Reduction-Chemical Extraction
System for Radiation Contaminated Soils
Robert S. Dyer
US Environmental Agency
Office of Radiation Programs
Volume reduction is a subject of interest to everyone associated with the management
of radioactive waste. This interest is focused in three general areas. Economics is one area - -
- regulatory , transportation, and disposal site factors have during recent years significantly
increased the cost to dispose of a cubic yard of low level radioactive waste. From a societal
standpoint, waste generators' concerns center on the limited current availability of low level waste
burial sites, the uncertainty associated with availability of future State compact sites, and the
national policy preferring treatment prior to, or instead of, burial. A second general area of
interest is concerned with the broad question of "how clean is clean?". In other words, what
exactly are the applicable or relevant and appropriate regulations (referred to as ARARs), should
there be one national cleanup standard or should each site have its own standard, and should
the standard be related to the specific activity of radionuclides or a dose limit? Thirdly, volume
reduction is a generic issue. Every manager of low level radioactive waste, whether in a nuclear
power plant, medical institution, Federal facility, or Superfund site, can benefit from the use of
volume reduction technology.
The obvious objective of volume reduction is to concentrate the radioactive matrix or
separate the non-radioactive from the radioactive components. A wide variety of technologies
are in current use, including dryers, compactors, physical separation systems, incinerators,
concentrators, evaporators, and chemical systems. Choosing the appropriate volume reduction
technology of course depends on the waste form and the strategy for ultimate disposal.
The US Environmental Protection Agency (EPA) Superfund program is directed toward
remediation of sites on the National Priority List (NPL). The waste form at many of the NPL sites,
that are contaminated with radioactivity, is soil. The only disposal strategy available for
radioactive soil is burial. Since the contaminated soil volumes are very large, burial costs will be
very high. To investigate technologies that could reduce the volume of soil requiring burial, the
US EPA Office of Radiation Programs is conducting a project with the Superfund program that
is focused on the remediation of soil contaminated with radium. Factors that must be considered
in this soil remediation technology include long-term effectiveness and permanence; reduction
122
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of toxicity, mobility, or volume through treatment; short-term effectiveness; implementability; and
cost.
The VORGE (Volume Reduction /Chemical Extraction) project was established to
investigate technologies for possible application at the Montclair and Glen Ridge, New Jersey
Superfund sites. These two sites were contaminated with ores and ore processing debris from
the radium industry that operated in the area in the 1920s. The project has two broad objectives:
(1) reduce the activity level of a major portion of the soil to levels protective of public health and
return reconstituted soil to the sites, and (2) reduce the volume of contaminated soil by use of
physical or chemical techniques that can potentially be applied in residential areas. The. project
organization consists of two components: Laboratory Studies and Field Treatability.
Laboratory studies are being conducted at the EPA National Air and Radiation
Environmental Laboratory in Montgomery, Alabama. Studies include soil characterization
protocol development and analysis, evaluation of physical separation techniques (e.g., screening,
sieving, magnetic separation), and evaluation of chemical extraction techniques. The radium
contaminated soils from the two NPL sites mentioned above have been characterized as to
particle size distribution, radioactivity distribution, and mineral and materials composition. Some
of the information derived from the characterization studies is shown in Figure 1 for a Montclair
soil sample. The laboratory analyses demonstrated that soils from different sites, as expected,
exhibit different compositions and radioactivity distributions. For example, the Glen Ridge
samples were much higher in activity than Montclair samples, and also had more slag,
unprocessed ores, and radium process precipitates. This demonstrates the importance of
representatively sampling a given site, and emphasizes the fact that the soil composition and
radioactivity distribution dictate the design of the treatment technology . The analyses also
showed that the majority of the total activity is in the sand-silt particle size range and that the
highest concentrations of radium are in the silt-clay particle size range. The radium in these soils
is present as ore materials (e.g., carnotite and uraninite), acid precipitates (radiobarite and
amorphous silica), and adsorbed bnto materials (slag, quartz, rock, clay minerals).
The initial evaluation of physical separation techniques emphasized screening and sieving.
The tests showed that a significant percentage of the radioactivity could be separated into the
smaller particle size fractions and that the percentage was highest when a combination wet
sieve/vigorous wash technique was used. As shown in Figures 1 and 2 for Montclair soil, the
smaller particles are higher in activity (i.e., activity increases as particle size decreases), and there
is a dramatic decrease in the activity levels of the sand-gravel fractions for the 'Wet
Sieve/Vigorous Wash" as compared with the "Dry Screen", corroborating the inverse relationship
between particle size and radioactivity. Figures 1 and 2 also show that the small-to-middle sand
size (0.15mm to 0.30mm) is the most appropriate particle size range for separation. Laboratory
tests demonstrated that 25 - 40% (by weight) of the contaminated soil can be cleaned and
returned to the site by use of the wet sieve/vigorous wash technique. The remaining 60 - 75%
(by weight) of the soil, consisting of silt, clay, and small sand size particles, would be
disposed of.
Some of the components of the contaminated soil have magnetic properties, such as
ferruginous particles. The tests to date on magnetic separation demonstrate that the technique
has potential application to the soils used in this project. There is evidence that radium
contaminated debris was burned in furnaces and the resulting slag and ash materials that were
123
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disposed of in community landfills contained iron compounds. Some of the slag material in the
test soils can be removed by a magnetic field, and the laboratory results indicate that the activity
of the "wet sieve/vigorous wash" fraction can be further reduced by about five to ten percent.
Further tests and evaluations are planned. The expectation is that, provided the laboratory data
indicate the technique will be cost-beneficial, a field treatability magnetic separation component
will be evaluated.
Chemical extraction techniques are widely used in uranium mining and milling processes.
Tests and evaluations are underway at the National Air and Radiation Environmental Laboratory
to determine applicability of chemical extraction to the Superfund soils. The objective is to
identify acids, salts, or complexing agents, at varying temperatures and concentrations, that
potentially could be applied in later stages of the volume reduction system. Initial efforts have
been aimed at chemical extraction of radium from the residue from the "wet sieve/vigorous wash"
technique, but later tests will investigate application to the unprocessed soil. In addition,
techniques for regenerating any acid solutions, salts, or complexing agents used will also be
evaluated. Below is an example of the preliminary results from acid extraction, after application
of 3M (3 molar) nitric acid at 85 degrees centigrade and a 2.5 to 1.0 liquid to soil ratio.
Radium concentration (picoCuries per gram)
in a Glen Ridge soil sample before extraction
60
180
810
Radium concentration (picoCuries per gram)
after extraction
7
13
60
Conclusions to date from the laboratory studies can be summarized as follows:
• The "wet sieve/vigorous wash" technique reduced 25 - 40% of the Montclair soil from
180 picoCuries radium per gram of soil to 12 -15 picoCuries radium per gram of soil.
• The wash water can be recycled (radium is not sufficiently solubilized in the wash
water to require removal).
• Reconstituting the soil treated only by the Vet sieve/vigorous wash" technique (12-15
picoCuries radium per gram) with clean fill will produce 5-7 picoCuries radium per gram
soil for return to the site, in general conformance with the 40 CFR 192 standards
promulgated under authority of the Uranium Mill Tailings Radiation Control Act (UMTRCA)
of 1978.
• Preliminary study indicates additional activity reduction may be obtained by magnetic
separation and by chemical extraction.
In the Field Treatability/Pilot Scale demonstration phase of the VORCE project, a field
treatability unit will be designed based on the results of the laboratory studies, and the unit will
be constructed, tested and evaluated. The main objective is to translate the laboratory scale
success to the field scale by using off-the-shelf mineral beneficiation equiprnent, modified as
necessary. Equipment components envisioned for use on the VORCE project have been widely
124
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used by the beneficiation industry for many years. These processes and equipment components
have however never been used for cleaning soil contaminated with radioactivity. The evaluation
phase of the field treatability effort will include thorough field testing of all the equipment and
process control parameters.
Figure 3 is a simple mass flow diagram for a VORCE pilot scale field treatability unit. The
diagram shows the vigorous wash technique and the potential planned inclusion of the magnetic
separation and chemical extraction techniques. Figure 4 is the conceptual design for the
vigorous wash technique. This design is briefly described below.
The soil is added onto the "grizzly screen" which removes large sized material (rocks, tree
roots, trash). The remaining soil material falls into the "trommel/screen"which provides vigorous
mixing, scouring of particle surfaces, and separation and removal of washed gravel size material.
The tumbling/cascading action occurring in the pressurized water spray environment scours
radioactive material from the gravel- and sand-sized soil particles and breaks down soft soil
agglomerates. Scouring removes the majority of radioactive particles adhering to gravel surfaces
and a portion of the radioactive particles adhering to sand surfaces. The cleaned gravel size
materials are separated from the soil stream by the screen section of the "trommel/screen".
The dirty water containing sand, silt, and disintegrated clay balls passes through the
screen openings and into the sloped-bottom water pool of the first stage "classifier".
Classification is produced by differential suspension whereby settled material is carried along the
inclined bottom and discharged at the open end of the trough. Vigorous washing of this settled
material is produced by the tumbling/rolling action of the rotating spiral ribbon which grinds sand
grain against sand grain and thereby removes deleterious material. In differential suspension,
rounded, denser, larger soil particles tend to sink faster than angular, less dense, smaller soil
particles. The latter particles tend to remain suspended near the surface of the water pool and
thereby overflow the weir. The "first stage classifier" therefore produces two product streams:
a. Material which has settled to the bottom of the water pool, traveled the length of the
rotating screw to the upper end of the trough, and has been discharged into the input of
the "attrition mill".
b. Material which has remained in suspension in the water pool and follows the water
overflowing the weir into the input of the "sloped plate clarifier". This material will be
greatly enhanced in radioactivity.
The material discharged at the upper end of the "first stage classifier" receives prolonged
agitation washing in the "attrition mill". The "attrition mill" is composed of two cells, each
containing a high speed turbine shaft with two propellers set at opposing pitches. The propellers
force the material grains to abrade each other while carried in water suspension within the two
cells, thereby scouring superficial adhesions and coatings and suspending them. These newly
suspended, small particle size adhesions and coatings will be enhanced in radioactivity and the
coarser particles from which they were removed will be cleaned of radioactivity. The output from
the "attrition mill" is discharged to the water pool of the "second stage classifier".
The "second stage classifier" provides the final washing and particle size separation. The
"second stage classifier" produces two product streams:
125
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a. Material which has settled to the bottom of the water pool, traveled the length of the
rotating screw to the upper end of the trough, and discharged onto a stockpile. This
material will have been cleaned by the washing, scrubbing, scouring action of the
"trommel/screen", "first stage classifier", "attrition mill", and second stage classifier", and
.may be returned to the original excavation site.
b. Material which has remained in suspension in the water pool and follows the water
overflowing the weir into the input of the "sloped plate clarifier". This material will be
greatly enhanced in radioactivity.
The "clarifier" separates solids from water and produces a sludge, containing the
radioactive materials that have been removed from the original contaminated soil. The "plate and
frame filter press" removes the majority of the water from the clarifier sludge, thereby allowing the
water to be recycled. The solid filter cake produced by the "plate and frame filter press" contains
the radioactivity removed from the original contaminated soil and should be disposed of as
radioactive waste.
Figure 5 shows a general layout of the VORCE field treatability pilot scale unit. The
treated soil stream from the trommel and from the second stage classifier will be sampled and
analyzed to determine the radium concentration. Soil that meets the cleanup requirement will
be moved to the treated soil pile for mixing with fill soil. The resulting reconstituted soil will be
returned to the site. If, because of unusually high activity input soil, the streams are not
sufficiently clean, they will be moved to the reject soil pile for recirculation through the system
up to three additional cycles. The filter cake will be analyzed and packaged for disposal.
To summarize, laboratory scale tests and evaluations of techniques for reducing the
volume of radium contaminated soil for disposal have demonstrated that a significant portion of
the contaminated soil can be cleaned by a vigorous water wash technique and returned to the
excavation site. The laboratory work has also demonstrated the importance of soil
characterization, and has provided preliminary results that indicate magnetic separation and
chemical extraction techniques can provide additional volume reductions. A conceptual design
for a field treatability pilot scale unit to duplicate the laboratory results has been completed. The
design relies on readily available components. Construction, test, and evaluation of the unit is
planned.
126
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Figure 1: Soil Characterization Data (Example)
(180 picoCuries Ra226/gram Montclair soil)
ID
s
o
a
in
<*•
o
0 +
Gravel
26
44
8
Size Class
53
98
35
Clay
decreasing particle size
20 . 1
396 688
53 5
EH Slag
E9 Coal
H Glass/Trash
E3 Rock Particles
C] Quartz/Feldspar
M Ores/Acid Precipitates
@ Clay Minerals
H! Heavy Minerals
(volume or weight) of Total Soil
Average pCi Ra226/gram
% of Total Soil Ra226
127
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Figure 2: Wet Sieve/Vigorous Wash Data (Example)
(180 picoCuries Ra226/gram Montclalr soil)
350-
ra
CE
O
O
300-
250-
200-
150-
100-
50-
/ \
note: Activity values for particle
sizes smaller than 0.038mm are
greater than the vertical axis
maximum value of 350 pCi/g
tf
o- Dry Screen
I
—&— WetSieve/Vigorous Wash
\
\
.01
10
silt
-X-
Particle Size (mm)
sand
gravel
128
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Figure 3: VORCE Flow Diagram
1.0 Ton
Contaminated
from Site
Soil
VORCE,
Vigorous
Wash
System
0.6 Ton Fill
(1 pCi/g)
0.4 Ton'
VORCE Product
(12 - 15 pCi/g) for
Reconstitution and
Return to Site
JL
Ufiffi;
VORCE
Magnetic
Component
I
1.0 Ton
VORCE Reconstituted Soil
(5 - 7 pCi/g) for Return to Site
••#"VORCE Reconstituted Soil
for Return to Site
Rll
Contaminated Particles
for Disposal
0.6 Ton
Contaminated
for Disposal
Soil
"^"VORCE Reconstituted Soil
for Return to Site
Fill
Contaminated Material
for Disposal
129
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Soil from Sits
Figure 4: Conceptual Design
Return to Site
Make Up
Water
Return to Site
Sand
Filler
Radium
Removal
Unit
Holding
Tank
; -1/4" to +0.15mm
t
Return to Site
Plate & Frame
Filter Press
.*»- Disposal
130
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Figure 5: Plant Layout
Operations Office
(Trailer) ._
Filter Cake
for Disposal "S^.
Treated Soil.
a
Treated Soil,for
Sampling and •
Radioanalyaia
.a
Radiation Measurement
Lab (Trailer)
Fill Soil
O
VORGE System
Treated Soil
Reconstituted Soil
Input
Site Soil
Output
Reconstituted Soil
Stockpiles of Sit* Soil
.13-
Reject Soil for
Rocirculation
131
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Residual Radioactivity Cost Impact Evaluation
Richard P. Allen
Pacific Northwest Laboratory
ABSTRACT
Studies are in progress at Pacific Northwest Laboratory to evaluate the cost differentials
(impacts) for compliance with different residual radioactivity limits governing the release of land
and structures for unrestricted use following the decommissioning of licensed nuclear facilities.
The initial work has focused on the decommissioning of a reference uranium fuel fabrication
facility to achieve alternative effective dose equivalent levels ranging from 80 to 1 mrem/yr for the
maximally exposed individual. Concrete and soil removal and disposal costs were calculated for
the reference uranium fuel fabrication facility as a function of these target levels. The disposal
cost evaluation considered both low-level waste (LLW) disposal and the effect of establishment
of a below regulatory concern (BRC) disposal option based on compliance with a 1, 4 or 10
mrem/yr individual dose level. The results indicate that removal+disposal costs for concrete
structures are relatively insensitive to decreasing target levels for the LLW disposal option, but
costs decrease at the lower target levels for the BRC waste disposal option. The availability of
the BRC waste disposal option favors demolition over the preservation of decommissioned
structures. For contaminated soil, the removal+disposal costs increase with decreasing target
levels for the LLW disposal option, but similarly can decrease at the lower target levels for the
BRC waste disposal option.
INTRODUCTION
The preliminary studies described in this paper used a reference uranium fuel
fabrication facility decommissioning study (Elder and Blahnik 1980) to develop a cost analysis
methodology and illustrate its application to the evaluation of decommissioning cost impacts.
This work was supported by the U.S. Nuclear Regulatory Commission under Contract
DE-AC06-76RLO 1830, NRC FIN L11949. Pacific Northwest Laboratory is operated for the
U.S. Department of Energy by Battelle Memorial Institute. The results and conclusions
presented in this paper are those of Pacific Northwest Laboratory and do not represent official
NRC policy.
132
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The analysis was limited to the land and structures remaining at the completion of equipment
dismantling and building decontamination operations, and assumed a starting dose of 100
mrem/yr. Based on these guidelines, concrete and soil removal and disposal costs were
calculated for representative decontamination and decommissioning (D&D) operations capable
of achieving alternative (target) residual radioactive contamination levels of 80 to 1 mrem/yr for
the maximally exposed individual. Low-level waste (LLW) disposal costs were calculated based
on shipment of the LLW to a regional waste compact site. The effect of the establishment of a
below regulatory concern (BRC) waste disposal option based on compliance with a 1, 4 or 10
mrem/yr individual dose level also was evaluated.
The reference facility plant area selected to provide a conservative case for the impact
evaluation was an 8 x 8 x 5 m ground floor room with reinforced concrete floor, concrete block
walls and reinforced concrete ceiling. The site area was representative soil containing
radionuclides deposited from normal plant operations. The physical condition of the reference
facility at the conclusion of the decommissioning study (starting point of this evaluation) was
defined by reviewing the proposed decommissioning operations as specified in Elder and Blahnik
(1980), supplemented by discussions with participants in the original study. Decommissioning
of the plant and site areas would be effected by:
1. Disassembling, decontaminating, packaging and disposing of all equipment.
2. Removing all remaining piping, electrical conduit, and local filter systems and ducting.
The main ventilation ducts were left in place since they were protected by HEPA filters
on the equipment and room ducting.
3. Decontaminating the floors, walls and ceilings with detergent cleanser, powered
brushes, and hand wipes. The floor tiles and strippable seal coatings would be
removed.
4. Removing hot spots in the floors and walls by chipping and vacuuming.
5. Removing all inside and outside piping from the pipe trenches, along with any hot
spots in the trenches.
6. Removing contaminated material and liners from the waste treatment lagoons, along
with any contaminated soil beneath the liners.
The reference decommissioning operation thus would leave the buildings intact, with
all contaminated equipment and fixtures removed and the interior surfaces decontaminated.
There would be no general site decontamination, just selective removal of hot spots associated
with pads, piping and the lagoons. ,
COST EVALUATION
Unlike other facilities, the fractional activities for a uranium fuel fabrication plant are not
location or process specific; i.e., the residual radioactive mixture is the same for all structure and
site surfaces. However, the initial contamination levels and the concentration profiles can vary
133
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significantly as a function of the handling+processing step, the form of the contamination
(gaseous, liquid, powder), and the type and condition of the surface (metal, concrete, porous,
sealed, etc.).
The radionuclide inventory estimates in Elder and Blahnik (1980) addressed the
equipment removed during the decommissioning operation, but not the structure and site
surfaces except for the lagoons. A survey of recent decontamination, decommissioning and
radiological survey literature was conducted to provide some guidance on post-decommissioning
contamination levels and profiles and also to obtain current D&D cost, worker dose and waste
volume data. The only residual contamination information relevant to the uranium fuel fabrication
plant reference case was some limited soil contamination data from the post-remedial-action
radiological survey of a decommissioned mixed oxide fuel development and fabrication facility
(Flynn et al. 1984). All uranium and thorium readings were within the range of normally expected
background concentrations except for one area with isotopic ratios indicative of a release or
spillage of enriched uranium. The contaminated zone was subsurface (5-10 cm depth), with an
average radionuclide concentration of 6.4 pCi/g.
This lack of specific data or estimates defining the initial residual concentration values
for the reference facility is not critical since the cost and worker dose impacts are evaluated using
differences rather than absolute values. The initial concentration values were therefore calculated
based on an assumed starting dose of 100 mrem/yr. It was further assumed that the plant walls
and ceilings contain 50% and 10%, respectively, of the corresponding floor contamination
inventory.
A concentration versus depth functional relationship must be established for each
structure and site area in order to estimate the effectiveness of the applied decontamination
processes and the associated costs and worker exposure as a function of the amount of material
removed. The literature survey cited earlier identified several different types of observed
contamination distribution profiles reflecting the origin and subsequent history of the various
possible surface-contaminant interactions, i.e., contamination incorporated in surface coatings,
diffusion or mixing into the base material, previous decontamination operations, subsequent
coverage with paint or clean soil, contaminated versus activated material, etc. No concentration
profile information directly relevant to the uranium fuel fabrication plant reference case was
identified other than the previously noted example of subsurface soil contamination.
In the absence of applicable data, diffusion theory was used to develop a generic
concentration versus depth relationship. The concentration profile for a substance deposited on
a surface and allowed to diffuse into the base material for a time t is given by the exponential
expression: C = « exp (-x^Bt), where C is the concentration at the depth x. Solving a diffusion
problem involving a series of line sources or a single replenished source yields an expression
of the form: C = « erfc (x/6t), where erfc is the error-function complement. This expression was
selected as the best representat'' -* *j generic concentration profile, since the contamination
diffused into a surface could repiw. ^ cumulative effect of multiple sources of varying
strength applied at different times and for variable time intervals.
The required relationship between removal depth and the percent of the total
contamination inventory remaining after a decontamination operation was derived by integrating
the error function. An adjustable scaling factor was used to define the depth corresponding to
134
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a specified percent removed or remaining. Figure 1, for example, illustrates the calculated depth
profile curves for a 50% penetration depth of 2 mm, 1 cm, and 3 cm from tHe original deposition
surface. The same basic contamination inventory versus depth relationship thus can be used
to represent contamination that is predominantly near the surface or that has penetrated a
significant distance into concrete or soil.
Once the concentration profile and corresponding removal depths have been defined,
the specific decontamination and dismantlement processes applicable to each of the major
surface types (metal, concrete, soil, asphalt, etc.) can be selected. The major selection
considerations for decontaminating low activity surfaces are 1) required removal depth, 2) access
constraints, 3) contamination control and recontamination potential, 4) waste type and volume,
and 5) cost. The following are the candidate decontamination and dismantlement processes for
concrete and soil for the reference uranium fuel fabrication facility:
Concrete Decon. (0-1 cm):
Concrete Decon. (1-5 cm):
Concrete Demolition:
Soil Removal:
Scabbier
Spaller
Backhoe-Mounted Ram
Modified Backhoe
The values in parenthesis are the optimum removal depth ranges for the scabbier and
the spaller. Demolition of walls, floors, etc., is required if the amount of material removed impairs
structural integrity, For this evaluation, the demolition criteria are 1) removal of more than
one-half of the 10-cm floor thickness, 2) removal of more than one-fourth of the 20-cm wall or
10-cm ceiling thickness, since these would be decontaminated on both surfaces, and 3)
automatic demolition of the ceiling along with the walls. ~~
The decontamination costs and associated waste volumes for the reference fuel
fabrication facility structure and site surfaces were calculated using the following steps and
as'sociated assumptions:
Concrete Decontamination
1. Calculate the concentration profiles for the defined surfaces and an assumed 50%
penetration depth using the generic concentration profile derived previously and initial
inventory values of 1700 yCi/m2 for the floor, 850 jiCi/m2 for the walls, and 170 jiCi/m2
for the ceiling. The penetration depths, reflecting the volume containing 50% of the total
inventory as measured from the original deposition surface, are assumed to range from
0.2 cm to 5cm.
2. Calculate the total amount of material that must be removed for each surface to reduce
the dose to the specified target level. These calculations are based on the inventory
factors (total |iCi/m2 of surface) and the specified surface activity factors (dpm/100 cm
of surface). It is assumed that the higher concentration in the floor is removed first, and
that the walls and ceiling are decontaminated only when their dose contribution
becomes significant. Since there may be substantial additional costs or technical
difficulties associated with the precise establishment of the required removal depth, the
calculated depth is increased by 50% to reflect the uncertainties inherent in an actual
decommissioning operation.
135
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3. Calculate the LLW/BRC waste boundary for each surface to establish the required
removal depths for each waste type. The radionuclide concentration defining BRC
concrete waste from the reference uranium fuel fabrication facility is assumed to be 20
times the allowable residual concentration for the same target dose levels. For example,
BRC waste at the 4 mrem/yr BRC level has the same average radionuclide concentration
as the residual concrete for an 80 mrem/yr target dose level. The LLW/BRC waste
boundary is established by starting at the maximum residual contamination removal
depth and then calculating the average radionuclide concentration for the removed
material until the BRC limit is reached. The calculated LLW removal depth is increased
by 50% to reflect uncertainties in establishing this boundary in an actual
decommissioning operation.
4. Specify, for each removal depth, the appropriate decontamination or dismantlement
technique and calculate the number of applications required to remove material to at
least 80% of the specified depth. Use of the 80% value rather than the full depth
minimizes excessive waste generation just to remove a small remaining fraction of the
required inventory, and is consistent with the field uncertainties in removal depths and
the amount of material actually removed with each technique.
5. Calculate the decontamination cost to reach each target residual radioactivity level
starting from the initial conditions.
Soil Decontamination
1. Calculate the concentration profile for soil using the generic concentration profile
derived previously and an initial inventory value of 210 nCi/m2 for soil homogeneously
contaminated to a depth of 15 cm, and to twice this value for contaminated soil covered
with 15 cm of clean soil.
2. Calculate the total amount of soil that must be removed to reduce the dose to the
specified target level, assuming no subsequent coverage with clean soil. As with the
concrete surfaces, the calculated depth is increased by 50% to reflect the uncertainties
inherent in an actual decommissioning operation.
3. Calculate the LLW/BRC waste boundary, the decontamination costs, and the disposal
volumes using the procedures outlined for concrete. The radionuclide concentration
defining BRC waste from the reference uranium fuel fabrication facility site is assumed
to be four times the allowable residual concentration for the same target dose levels.
Waste Handling and Disposal
It is assumed 1) that all removed concrete and soil that is classified as LLW is sent to
a generic regional waste compact site for final disposal, 2) the disposal volume for the scabbled,
spalled or demolished concrete is ~ 1.4 times the removal volume and 3) the soil disposal volume
is equal to its removal volume.
A representative distance of 390 miles from the reference facility to the disposal site was
established by averaging the one-way distance from nuclear power stations to the existing
136
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disposal sites. A truck transportation cost of ~$100/m3 of removed material for concrete with
a density of 2300 kg/m3 was calculated by assuming a representative one-way charge of
/mile and a weight limit of 20,000 kg. The corresponding transportation cost for the 1500
kg/m3 soil is $7D/m3.
Current LLW disposal charges for the three existing disposal sites were averaged to give
a disposal cost of $1500/m3 for the removed concrete and $1100/m3 for the soil. Adding
packaging costs of $200/m3 for the concrete and $130/m3 for the soil gives a total waste
handling+disposal cost of $1800/m3 of removed concrete and $1300/m3 of excavated soil.
It is further assumed that any removed concrete and soil that can be classified as below
regulatory concern waste is sent to a local sanitary landfill for final disposal. Other assumptions
are 1) the waste is transported 30 miles at a cost of $17/1000 kg, or $40/m3 of removed material
for the concrete and $25/m3 for the soil, and 2) the disposal charge is $26/1000 kg, or $60/m3
of removed material for the concrete and $40/m3 for the soil.
This gives a total BRC waste handling and disposal cost of $100/m3 of removed
concrete and $65/m3 of excavated soil, or almost a factor of 20 less than the corresponding
direct cost for LLW handling and disposal. However, these savings are at least partially offset
by increased costs for monitoring, verification and certification of the BRC waste.
RESULTS AND DISCUSSION
The following are the key cost impact evaluation results for the reference concrete
room and site area. The associated figures show the calculated concrete or soil removal and
waste disposal costs required to go from the starting condition to a surface contamination level
less than the prescribed limit. The removal and disposal costs are displayed as separate bands
that sum to the total cost to illustrate the variation of the individual cost elements and the total
cost with the target levels. The cost differential or "impact" of decreasing the target levels is
reflected by the slope of the upper boundary, or total cost line. It must be emphasized that these
cost figures do not include front end and indirect but substantial costs such as management,
planning, engineering, training, radiological support, and safety because they would be incurred
for any target level requiring significant additional decontamination.
Concrete Decontamination
With all removed material classified as LLW, the concrete room removal+disposal costs
increase substantially with increasing contamination penetration depth (Figures 2 and 3) at the
lower target levels. However, the total cost is either constant or changes very little with residual
contamination level for target levels below 30 mrem/yr at any particular assumed penetration
depth. For contamination concentrated near the surface (Figure 2), the total cost is quite
insensitive to decreases in the target level. This reflects the discrete rather than continuous
nature of the concrete removal processes and the ability to remove all of the contamination with
one application. In the case of the reference room and the 0.2-cm penetration depth, for example,
the amount of material removed by one application of the spaller (2.5 cm) is more1 than sufficient
to reach any specified target level including 1 mrem/yr.
137
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For the same contamination inventory distributed over a depth of several centimeters
(Figure 3), the removal+disposal costs are comparatively low at the highest target levels, and
then increase sharply to a maximum LLW disposal cost reflecting demolition of the reference
room. The cost is low initially because of the reduced concentration levels associated with the
deeper penetration of the contamination. However, once the target level decreases sufficiently
to require significant decontamination, the amount of material that must be removed requires
demolition of the structure.
The preceding analyses were based on the classification and disposal of all removed
concrete as low-level waste at $1800/m3. An alternative option is to classify removed material
complying with a 1, 4 or 10 mrem/yr individual dose level as BRC waste that can go to a local
sanitary landfill at a disposal cost of $100/m3. The effect of this BRC option on removal+disposal
costs as a function of target level is quite dramatic as illustrated in Figure 4 for an assumed BRC
level of 10 mrem/yr. The removal+disposal costs at the highest target levels are the same. The
material nearest the surface has the highest contamination concentration, and is classified as
LLW regardless of the available disposal options.
As the target levels decrease, however, lower concentration material is removed that
meets the BRC criterion. The total cost decreases rather than increases at the lower target levels
since demolition is a comparatively low-cost operation and the resulting large waste volumes do
not have a corresponding large-disposal cost. In addition, if the definition of BRC waste is based
on an average contamination concentration," the volume of LLW and the attendant
removal+disposal costs will decrease rather than increase at the lower target levels. This results
from an effective dilution of the original near-surface LLW by lower concentration material from
deeper in the structure; i.e., the volume of BRC waste increases at the expense of the LLW.
This LLW cost decrease can be converted to a constant LLW removal+disposal cost
by defining BRC waste based on a maximum rather than an average contamination concentration
as illustrated in Figure 5. However, the total cost is still lowest at the lower target levels where
demolition is required. Since most of the BRC cost for intermediate target levels is for removal
operations to preserve the structure, a further cost reduction is possible by choosing to demolish
the structure regardless of the target level. The availability of a BRC waste disposal option thus
could favor demolition over preservation for decommissioned structures since the removal and
disposal costs would be the lowest of any option and almost independent of the residual level.
Soil Decontamination
The soil LLW removal+disposal costs for contamination concentrated within the first
few centimeters of the surface are independent of the target level. Unlike the concrete case, the
soil LLW removal+disposal costs increase significantly with decreasing target levels for deeper
assumed contamination penetration values as shown in Figure 6. However, the availability of a
BRC waste disposal option for soil results in a moderation, or even a decrease, in total
removal+disposal costs at the lower residual contamination levels (Figure 7), but this beneficial
effect occurs at lower target levels for soil than for concrete.
These results for the uranium fuel fabrication facility reference room and site area
provide valuable insights regarding key factors that can affect the cost impacts for compliance
with different residual radioactive contamination levels. However, this evaluation is quite limited,
138
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and the results and conclusions must be regarded as preliminary. Additional work is needed to
address the number of rooms and site areas that are contaminated, the effect of the differing
initial contamination levels for these areas, and the impact of the cost variations for these specific
decontamination operations on the total decommissioning cost.
REFERENCES
[1] Elder, H. K. and D. E. Blahhik. 1980. Technology, Safety and Costs of Decommissioning
a Reference Uranium Fuel Fabrication Plant. NUREG/CR-1266, Vols. 1 and 2, Prepared for
the U.S. Nuclear Regulatory Commission by Pacific Northwest Laboratory, Richland,
Washington.
[2] Flynn, K. F., A. L Justus, C. M. Sholeen, W. H. Smith, and R. A. Wynveen. 1984. Post-
Remedial-Action Radiological Survey of the Westinghouse Advanced Reactors Division
Plutonium Fuel Laboratories, Cheswick, Pennsylvania, October 1-8,1981. DOE/EV-0005/36,
ANL-OHS/NHP-84-100, Argonne National Laboratory, Argonne, Illinois.
139
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140
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Session III
Health Effects
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Experience in Decontamination and Reuse
of the Large-scale Radiochemical
Laboratory and the Research Reactor
at the Japan Atomic Energy Research Institute
i
Hideaki Yamamoto, Kouzou Matsushita and Hozumu Yamamoto
Department of Health Physics
Japan Atomic Energy Research Institute
ABSTRACT
This paper reviews, from a health physics viewpoint, some of our experience in
decontamination and reuse of a heavy-water cooled research reactor and a large-scale
radiochemical laboratory at the Japan Atomic Energy Research Institute. The research reactor
was decommissioned with plans for its reactor containment to be reused for a new reactor.
Residual radioactivity particular to this case was tritium remaining in concrete matrices of the
reactor containment. The primary health risk possibly encountered in reusing this type of facility
would be caused by inhalation of gaseous tritium released from concrete. The radiochemical
laboratory was decontaminated and reused as office rooms. The health risk from reusing this
facility would result from both internal and external exposure to various forms of radioactivity.
Generalizing from this experience, the information needed for estimating the risks in these
types of facilities is examined. Possible forms of residual radioactive criteria for each of these
cases are discussed.
INTRODUCTION
More than 30 years have passed since the establishment of the Tokai Research
Establishment of the Japan Atomic Energy Research Institute (JAERI). Therefore, in the
Establishment, there are some old facilities which have lost their usefulness or flexibility to users'
expanding requirements.
Typical examples of these timeworn facilities at JAERI were the Japan Research Reactor
No. 3 (JRR-3), and a large-scale radiochemical laboratory called the Research Laboratory
Building No. 1. These two facilities needed to be repaired, to fill new needs.
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In this paper, decontamination experiences with these two facilities and their reuse are
discussed from a health-physics viewpoint. In addition, some characteristics of health risks, and
data needed for estimation of the risks due to residual radioactivity in these types of facilities, are
pointed out.
REUSE OF BUILDINGS
Objects for decontamination or reuse can be categorized into "sites", "buildings" and
"materials". The reuse of "buildings" is discussed in this paper. The two "buildings" discussed
here can be classified as a heavy-water cooled research reactor in the "reactor category" (test
reactors, research reactors, commercial power plants, reactors in ships) and a chemical
laboratory in the "laboratory category" (physics laboratories, including accelerators, chemical
laboratories), respectively. The Japan Research Reactor is being prepared for "restricted reuse",
and the Research Laboratory Building No. 1 for "unrestricted reuse".
JRR-3
JRR-3 (10 MW), the first reactor designed, manufactured and installed with only Japanese
domestic techniques, was operated from 1962 to 1983 in the Tokai Research Establishment.
Experiments in design, construction, and operation were performed with it for developing and
testing reactors. It also produced radioisotopes used in various fields of industry and science.
The reactor fuels were natural and 1.5% enriched uranium oxides, and the coolant was
heavy-water.
During its 20 year operation, users of the reactor carried out increasingly complicated
experiments, and required a higher quality of performance of the reactor. Eventually, it could no
longer satisfy the users' needs sufficiently.
The dismantlement program planned to remove the reactor block without taking apart the
reactor containment, and to construct a new reactor with higher power and efficiency inside the
same containment. The reactor containment consists of reinforced concrete and is 33 m in
diameter, 27 m in height, and cylindrical in shape. It should be emphasized that reuse is
intended to be restricted, and that the inside of the reactor containment will be treated as a
"radiation controlled area" as before.
The one-piece removal of the 2,250 ton reactor block attracted much public attention.
The duration of the dismantlement continued from August 1984 to March 1987.
Types and Levels of Residual Radioactivities
When reactor JRR-3 ceased its operation, a considerable amount of radionuclides
remained in its installation and equipment. Most of the residual contamination was removed with
the reactor block. However, tritium remained in the concrete walls, floors, and ceilings of the
containment.
The tritium contamination of the reactor containment was due to the heavy-water coolant.
The heavy-water was activated and tritium was generated and stored in the coolant system. The
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tritium concentration in the coolant was about 40 MBq per cubic centimeter in 1984, after about
20 years of operation.
The gaseous heavy-water containing the tritium diffused into the reactor containment,
deposited onto its surfaces, and penetrated into the concrete structures. The heavy-water
released from the coolant system onto floor surfaces during maintenance work directly
penetrated concrete structures, and evaporated and then contaminated other floors, walls and
ceilings.
Figure 1 shows an example of tritium concentration distribution in the concrete of the
reactor containment. Widespread tritium contamination in the reactor containment should be
noted. The tritium concentration levels in the concrete samples of the basement floor, i.e.,
around the coolant system, were higher and more scattered than on the first floor. The basement
floor was probably contaminated directly with tritiated heavy-water released from the coolant
system, while the first floor was contaminated by tritium vapor. Typical profiles of tritium
concentration in the concrete of floors and a wall are shown in Figure 2. The tritium penetrated
about 60 cm into floors and walls of JRR-3.
The tritium in the concrete, released to the atmosphere of the reactor containment, could
generate airborne contamination. The tritium concentration in the atmosphere of the reactor
containment was measured before the removal pf the primary coolant system. The maximum
concentration was approximately 0.7 Bq per cubic centimeter of air in 1979. This value
decreased to about 70 micro-Bq per cubic centimeter of air after removal of the reactor block.
The latter level of the tritium concentration is considered to be caused by the release of residual
tritium in the concrete structures inside the reactor containment.
Decontamination
The decontamination process, which aimed to reuse the reactor containment, involved
surveys of contamination level, the decontamination itself, and measurements of the level of
residual radioactivity afterwards. Although various types of decontamination were performed on
the reactor, only the decontamination of tritium in the reactor containment is discussed in this
paper-
To estimate the levels of contamination "powder" samples of the walls, floors and ceilings
were taken using a special drill. The "powder" samples were immersed in distilled water for 24
hours, and then the tritium in the water was measured with a liquid scintillation counter to
estimate the tritium concentration in the concrete. The minimum detectable limit of this method
was 0.4 Bq per gram of concrete.
Contaminated concrete was removed with drills. Some concrete matrices were removed at
a depth of several tens of centimeters.- •
Residual Radioactivity Criteria
Contaminatedconcrete was Removed down to a level of residual radioactivity derived from
the "limit for surface contamination" prescribed in an ordinance of the responsible Minister. The
limit is 0.4 Bq per square centimeter for alpha-emitters and 4 Bq per square centimeter for other
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radionuclides. The limit is one of the conditions for designating an area as a "radiation controlled
area", where some restrictions for radiation protection purposes are to be implemented. This
residual radioactivity criterion was chosen due to plans to release the reactor containment as a
"radiation controlled area" temporarily, before it was reused as a containment for the new reactor.
The residual radioactivity criteria were derived from the existing regulatory limit, without estimating
any possible risk associated with this level.
It was confirmed that the residual radioactivity levels in the decontaminated concrete did
not exceed the criteria. This was done by measuring tritium concentration in the samples of
concrete powder using a method similar to that of the pre-decontamination survey.
, The dose for an individual working for 2,000 hours per year in the reused reactor
containment was estimated. With the level of residual tritium concentration measured after the
removal of the reactor block, such exposure was estimated to result in an effective dose
equivalent of about 3 micro-Sv per year due to inhalation of the contaminated air. The tritium
concentration in the air of the reactor containment was assumed to be 70 micro-Bq per cubic
centimeter.
Features of Health Risk
In this section, we identify some features of health risk which are encountered in reusing
heavy-water cooled reactors. In the JRR-3 case, where a new reactor is to be installed and
operated, the health risks of the reuse will result from both the operation of the new reactor and
the residual radioactivity from the old. Although the risk due to the former will dominate, for the
purpose of identifying general features of the health risk in reusing heavy-water cooled reactors,
it is of value to consider the risk due to the residual radioactivity separately. The consideration
will be informative for a case in which a contaminated reactor containment is reused for other
than containing a reactor.
The residual radioactivity health risk peculiar to JRR-3 is from tritium. During a reactor
operation, tritium penetrates concrete matrices in the floors, walls and ceilings. And during a
period of reuse, the residual tritium escapes from the concrete to the air in the reactor
containment. The dominant health risk of tritium is due to internal exposure from inhalation of
contaminated air. External exposure of skin by submersion in the air will also be encountered.
Skin can be directly exposed when it touches a contaminated surface. Moreover, if a ventilation
system is installed in the reactor containment, environmental contamination outside of the
containment caused by gaseous effluents containing tritium can also be a source of health risk.
One critical group, with respect to the health risk caused by a reuse of a heavy-water
cooled reactor containment, will be the workers within the containment area. But under certain
circumstances, the general public living around the reused facility can also be designated as a
critical group.
Although the residual radioactivity criteria derived from the current regulation were
implemented for JRR-3, alternative residual radioactivity criteria based on a risk assessment could
be established. To establish residual radioactivity criteria in this way, a model would be needed
for estimating the extent of risk. In the case of a heavy-water cooled reactor, the model should
refer to the relation between residual tritium concentration in concrete matrices and airborne
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tritium concentration in a room. In order to relate the concentration in concrete matrices to that
in air, it is necessary to take a suitable average of the observed concentrations of tritium in the
concrete matrices, because of "hot spots" of high contamination. Analyses of tritium behavior
in a reactor containment, (i.e., its dynamics of diffusion and transport, effects of room
temperature, humidity and ventilation) is indispensable in making a risk model. (One must
analyse tritium behavior to calculate airborne tritium concentrations.)
The risk depends directly on the airborne tritium concentration in the reactor containment.
A practical residual radioactivity criterion, therefore, would be a limit for airborne concentration
or for the concentration in the concrete.
RESEARCH LABORATORY BUILDING NO. 1
Research Laboratory Building No. 1 was an assembly of chemical laboratory rooms where
various radioisotopes were handled. It had a concrete framework and consisted of six stories,
including two underground ones. In the first period of the decommissioning plan, the west half
of the Building, which covered an area of 4,000 square meters with 70 rooms, was designated
for decontamination and reuse. Unsealed radioisotopes were handled in the whole area of the
third story and a part of the second story. Sealed radioisotopes were handled or X-ray
generators were installed in the other rooms. The Building was used from 1959 to 1982 and
therefore, most of the equipment and its function had lost its effectiveness. The Building needed
to be repaired for maintaining safety and security. But it was judged to be impossible to reuse
the Building as a radioisotope handling facility because most of the equipment and installations
were too timeworn to handle radioisotopes. Therefore, the reuse plan for the Building was to
remove all interior equipment and installations, to decontaminate its framework, and to reuse it
as offices, meeting rooms or laboratories where radiation is not to be used. This was an example
of releasing a building for unrestricted reuse. The decontamination lasted from December 1982
to March 1984.
Types and Levels of Residual Radioactivity
At the end of Research Laboratory Building No. 1 's use, there remained various kinds of
contaminated equipment: hot cells, fume hoods, desks, chemicals stackers, etc. Contaminated
spots, generated by radioisotopes scattered during chemical handling, were distributed over the
surfaces of walls, floors and ceilings. In the drainage system, contamination was distributed
inside piping, and on floors and walls around leaky pipes.
Typical contaminating radionuclides were Ru-106, Cs-137, and isotopes of uranium and
thorium. C-14's vapors contaminated wall and ceiling surfaces and the insides of some
ventilation ducts. The maximum contamination level was measured at a spot orvthe floor. It was
a Cs-137 contamination of 400 Bq per square centimeter. A room was entirely contaminated with
C-14, where contamination was found on the floor, walls, ceilings and frames of the windows,
with a maximum contamination level of 80 Bq per square centimeter. Only C-14 contaminated
a whole room surface; contamination by other radionuclides were found only in spots or on
fractions of room surfaces.
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Decontamination
The decontamination process required for reuse in the Building included a survey of the
levels of contamination, the decontamination itself, and measurement of the levels of residual
radioactivity. Before the decontamination, in-situ surveys of levels and locations of contamination
were conducted with gas-flow radiation counters with large detection windows. Alpha-emitters,
i.e., isotopes of uranium and thorium, were also measured with these counters for their beta
emissions. Radionuclides from the contaminated areas were identified with
gamma-spectroscopy. For rooms where only one kind of radionuclide was handled, the nuclide
could be identified from the history of the room, without consulting results of the spectroscopic
analyses.
During the decontamination, all the interior equipment, piping, and ducts were removed.
Loose or unfixed contamination on floor or wall surfaces was washed off. The contamination
remaining after this washing treatment was removed with drills or scabbling equipment. Residual
contamination was found to a depth of only a few millimeters, so it was easily removed.
Residual Radioactivity Criteria
Because of the planned unrestricted reuse of the Building, practically no residual
radioactivity was allowed to remain. This requirement was attained and affirmed by an in-situ
measurement of surface density of radioactivity throughout the entire decontaminated Building.
The numerical criterion for the surface density of radioactivity was set equal to the minimum
detectable limit of the radiation measurement instrument, i.e., 0.4 Bq per square centimeter for
fixed beta contaminants. Finally, the Building was released from radiation control.
External and internal doses due to total surface contamination of 0.4 Bq per square
centimeter of Cs-137 were estimated to be about 20 micro-Sv/y and 0.1 micro-Sv/y, respectively.
Features of Health Risk
In this section, we identify some features of the health risks resulting from residual
radioactivity which would generally be encountered in the reuse of a large-scale radiochemical
laboratory.
Observed patterns of residual radioactivity distribution in the Building fall into four categories:
1) extended contamination of inner surfaces of rooms (e.g., contamination with volatile radioactive
materials, such as C-14), 2) hot spot contamination of surfaces, 3) localized surface
contamination on and around piping or ducts for exhaust air or liquid effluents and 4)
contamination on-the equipment surfaces. Unfixed contamination will possibly spread over the
building as well as to the outside environment through exhaust air or liquid effluents.
Health risks due to exposure to radiation from contamination can be encountered through
the following pathways: 1) external exposure from the surface contamination, 2) internal exposure
caused by inhalation of resuspended or volatilized radioactive materials from the surface
contamination, and 3) external exposure through submersion in air containing resuspended or
volatilized radioactive materials. Unfixed contamination can be easily decontaminated. Moreover,
most of the fixed contamination remains at shallow depths in the building material and can easily
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be removed. Thus, in the above-mentioned exposure pathways, those with volatile radionuclides
will be most critical. Although the risks can expand to the environment outside of the building
through radioactive liquid or gaseous effluents, the critical group will consist primarily of users
of the Building following clean-up.
To establish a residual radioactivity criterion based on a risk assessment, a model which
could express the relationship between surface contamination levels and exposure doses would
be needed. This model should be applicable for various kinds of radionuclides. In addition, the
model should provide a method of averaging various levels of spot or localized contamination,
as well as contamination dispersed through ventilation or drainage systems. Modeling the
behavior of volatile radioactive materials, such as C-14, would also be needed.
The risks depend mainly on the surface density of radionuclides on the floors, walls and
ceilings. A practical form of a residual radioactivity criterion can therefore be expressed in terms
of surface density.
CONCLUSION
Experiences in the decontamination and reuse of a heavy-water cooled reactor and a
large-scale radiochemical laboratory were reviewed. The two reuse facilities discussed in this
paper were quite different from one another
Following cleanup, JRR-3 was released for restricted use, but use of the Research
Building No. 1 was unrestricted. It seems natural to set different levels of residual radioactivity
criterion for the two types of reuse.
The main residual radionuclide in JRR-3, a heavy-water cooled reactor/was tritium. The
Research Laboratory Building No. 1, on the other hand, a large-scale radiochemical laboratory,
was contaminated with various kinds of radionuclides. Therefore, in order to set a residual
radioactivity criterion nuclide-specific, radionuclides should be selected appropriately.
The two facilities also differed with regard to the health risk due to residual radioactivity.
The health risk in a heavy-water cooled reactor was caused mainly by internal exposure resulting
from the inhalation of tritium. In a large-scale radiochemical laboratory, the health risk was
caused by both external and internal exposure to various radionuclides. To model these health
risks, it is necessary to explain the relation between specific activity of a radionuclide and
individual exposure dose for a heavy-water cooled reactor, and between surface density of
radionuclides and individual exposure dose for a large-scale radiochemical laboratory.
If "operational" residual radioactivity criteria are based on some dose criteria, criteria for
these two facilities should be different from each other in form.
It seemed appropriate to us that in the case of reusing buildings, an operational residual
radioactivity criterion should be derived from a basic general criterion established by the
regulatory authorities, in a case-by-case basis for each facility, according to its type of reuse, its
main residual radionuclides, and its features of the health risk.
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ACKNOWLEDGEMENTS
*
The authors would like to thank related staff of the Radiation Control Division II,
Department of Health Physics, JAERI, who provided us with the valuable observation data
referred to in this paper.
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Applied Exposure Modeling for
Residual Radioactivity and Release Criteria'
D. W. Lee
Energy Division**
Oak Ridge National Laboratory
ABSTRACT
The protection of public health and the environment from the release of materials with
residual radioactivity for recycle or disposal as wastes without radioactive contents of concern
presents a formidable challenge. Existing regulatory criteria are based on technical judgment
concerning detectability and simple modeling. Recently, exposure modeling methodologies have
been developed to provide a more consistent level of health protection. Release criteria derived
from the application of exposure modeling methodologies share the same basic elements of
analysis but are developed to serve a variety of purposes. Models for the support of regulations
for all applications rely on conservative interpretations of generalized conditions while models
developed to show compliance incorporate specific conditions not likely to be duplicated at other
sites. Research models represent yet another type of modeling which strives to simulate the
actual behavior of released material. In spite of these differing purposes, exposure modeling
permits the application of sound and reasoned principles of radiation protection to the release
of materials with residual levels of radioactivity. Examples of the similarities and differences of
these models are presented and an application to the disposal of materials with residual levels
of uranium contamination is discussed.
The submitted manuscript has been authored by a contractor of the U.S. Government
under contract No. DE- AC05-84OR21400. Accordingly, the U.S. Government retains a
nonexclusive, royalty-free license to publish or reproduce the published form of this
contribution, or allow others to do so, for U.S. Government purposes.
Operated by Martin Marietta Energy Systems, Inc., under contract
DE-AC05-84OR21400 with the U.S. Department of Energy.
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INTRODUCTION
Modeling studies of the fate of radioactive materials released to the environment have
been performed over the past forty years. Within the past decade, they have been applied to the
establishment of regulatory criteria and standards. Models are a useful means for examining the
consequences of the release of radioactive materials because they provide a consistent
presentation of the significant transport mechanisms and exposure pathways affecting human
health and the environment. ICRP [1,2] has provided the basis for evaluating doses on a
consistent basis with the application of constant risk from exposure to radiation. This has
allowed models to examine scenarios for the exposure of workers and members of the public
and compare the relative doses providing a technically defensible basis for establishing
regulatory standards. Maximum individual doses and collective doses to larger groups of people
can be addressed by models on an equivalent basis to provide consistent levels of protection
appropriate to the waste management practice being considered. Exposure models include
doses received from external exposure, inhalation, and ingestion pathways and are typically
tailored to the specific management practice and radioactive materials being evaluated. The
application of exposure modeling to residual radioactivity and recycle provides a means for
addressing many of the unresolved issues in current regulations, such as the criteria for the
release of materials with bulk activity compared to those with surface activity. With a reasonable
representation of the source terms, bulk and surface activity can be addressed separately and
the resulting consequences evaluated on the consistent basis of potential doses to a member
of the public.
Exposure models differ in the detail and depth of analysis, but all exposure models share
the same basic conceptual elements. Many of these elements require the use of approximations
or assumptions that limit the validity of the model for some applications. For the application of
exposure models to residual radioactivity and release criteria, the identification of a dose limit is
an important consideration. Without the identification of a dose limit, model results are difficult
to evaluate and compare, which leads to results which are likely to be suggestive rather than
conclusive. Another common element in exposure modeling is the definition of the source of
radiation. .Often this aspect of modeling is the most important and difficult component of the
analysis. Models typically resort to some level of approximate analysis with respect to the source
term that is subsequently propagated throughout the analysis. Assumptions invoked in defining
the source term are typically the best indicator of the range of applications that an exposure
model can be used and remain technically defensible. Having an established source term,
models require the identification of exposure scenarios likely to occur as a result of the waste
management practice under consideration. These scenarios are identified with the intent of
encompassing potential exposures to the public and the environment in the future and are
formulated with a conservative interpretation of possible events. With the establishment of the
exposure scenario, the source is transported through environmental media to the point of human
exposure by a modeled representation of the facility and site performance. At the point of human
exposure, the concentrations of contamination are related to doses using the exposure scenarios
identified. Model results alone rarely provide a definitive answer of the potential dose as
compared to the prescribed dose limit because of the assumptions and approximations
incorporated into the model. Consequently, interpretation of the results is necessary to support
any decision resulting from the application of exposure modeling.
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While exposure models all share the same basic conceptual elements, the variety of
applications and different concerns addressed have resulted in a large number of models to
consider in model selection. In spite of the extensive list of available models for performing
exposure analysis, the existing models are of three basic types. These types can be regarded
as generalized or generic models, simulation or research models, and compliance or site-specific
models. Each model type has an important role in understanding the consequences of
management decisions regarding radioactive materials. Generic models provide a general
interpretation of potential exposures from a waste management practice and are useful in the
support of rulemaking, such as the modeling used to support 10 CFR 61 [3]. Research models
provide scientific simulation of the fate and behavior of radioactivity using best available scientific
analysis of important components of exposure models and are useful for advancing the
state-of-the-art in modeling potential exposures. Site-specific models provide conservative
estimates of facility and site performance for demonstrating compliance with regulatory
requirements for specific applications. This type of modeling is useful for demonstrating
protection of public health and the environment when waste, facility and site characteristics are
important considerations in determining the acceptability of the waste management practice.
These site-specific factors frequently are critical to making decisions that are both appropriate
and defensible.
Modeling potential exposure from a waste management practice provides forecasts of
future consequences of the release of radioactive materials. In the evaluation of the
consequences of the release of materials with residual levels of radioactivity, the potential uses
of these materials following release is limitless. Once released, control is difficult to impossible
to restore. Modeling can not and should not be expected to predict the future use of materials
released for recycle. Since materials that are recycled will ultimately be disposed of as ordinary
waste and any waste may be salvaged for reuse, a conservative approach to exposure modeling
that is defensible is to consider all materials with residual levels of radioactivity as wastes that
are disposed of and subject to salvage in the future. By considering future salvage, exposure
modeling needs to address the public adjacent to the disposal facility as well as an inadvertent
intruder who comes in contact with the waste.
APPLICATION OF EXPOSURE MODELING TO BRC WASTE DISPOSAL
An approach to exposure modeling using generic models for BRC waste disposal has
been prepared by EPA in support of the proposed 40 CFR 193 rulemaking process. The
exposure model used to support the proposed rule evaluates doses in the context of risk to
provide support to the selected dose limit for BRC waste disposal. A site-specific exposure
model for a site with a defined waste stream is in contrast to a generalized exposure model. A
site-specific model applied to the BRC disposal of uranium contaminated wastes from the Y-12
Plant in Oak Ridge illustrates the differences between generic and site-specific exposure
models [4].
As a first step in defining a site-specific model, an annual dose limit of 4 mrem effective
dose equivalent from all pathways was selected, which is consistent with the proposed BRC limit
being considered by EPA. Since the wastes to be evaluated include only depleted uranium
(0.02% U-235), modeling needs to focus on the likely pathways of exposure associated with long
half-life materials where the predominant doses result from ingestion and inhalation of
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contaminated material. The disposal concept is to have a dedicated facility within the DOE
Reservation in Oak Ridge which limits inadvertent intrusion into the waste until governmental
control of the Oak Ridge Reservation is lost. Exposure modeling requires the evaluation of doses
by considering concentrations of contamination; however, uranium is managed within the Y-12
Plant on the basis of activity. The management of uranium on an activity basis is important for
uranium accountability requirements but necessitates the conversion of uranium activity to
concentration by the use of the waste density. This conversion is not easily made since waste
materials are likely to have a broad range of densities depending on the generating process.
These site-specific issues and the unusual site characteristics of the Oak Ridge Reservation are
not typicallylncluded in generic models. The use of generic exposure models for this application
would necessitate extensive modifications to existing generic models. A more comprehensive
technique for ensuring that important site-specific conditions are presented as part of the
exposure model is to utilize site and facility data to define the scope and detail of the exposure
model. By focusing on the known data and associated issues, site-specific analysis can become
an extremely useful tool in waste management for examining alternatives and developing
management practices that are acceptable to regulatory requirements and protective of public
health and the environment.
Site-specific exposure modeling is typically founded on the data that are available to
examine potential exposures; however, numerous assumptions are typically required to provide
a complete picture of the waste management practice. Assumptions are chosen to be
conservative but not all assumptions are well supported and can only be regarded as best
estimates. For the application to Y-12, these assumptions are described for offsite and intruder
exposure scenarios, site performance, waste form and the disposal facility. The details of the
assumptions are shown in Table 1. While many assumptions are invoked the use of data is
maximized. Important site-specific data are incorporated into the transport calculations that are
determined from site characterization investigations of the proposed disposal site. The
combination of the assumptions and data are used to define the limiting concentration of uranium
in a dumpster unit that contains uranium contaminated materials and meets the prescribed dose
limit. The resulting dumpster loadings and activity limits are presented in Table 2.
DISCUSSION
The preceding exposure modeling methodology illustrates the differences in detail and
approach between site-specific and generic modeling. The most striking difference is the
focused orientation of the site-specific analysis in contrast to the exhaustive analyses common
to generic models. Emphasis is placed in site-specific modeling on those factors that .are
significant to the protection of public health and the environment to guide acceptable
management decisions. In humid environments, where the dominant pathway for the transport
of contamination is by surface water and groundwater, examination of the potential for
atmospheric exposure is tangential to selecting acceptable management practices. Similarly, the
consideration of decay and direct exposure from the wastes is not a major consideration for
uranium contaminated wastes. Uranium contaminated wastes provide an excellent example of
the usefulness of site-specific analyses because uranium is a ubiquitous element that needs to
be addressed as a contaminant in the context of the increment over natural background.
Additionally, the activity of uranium contaminated materials can vary widely depending on the
isotopic composition of the uranium source material. Failing to consider the specific differences
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between the various forms of uranium can lead to widely varying results. The unique features
of uranium need to be considered separately along with the unique institutional and site
characteristics associated with uranium contaminated materials that could be considered for
management as BRC wastes.
Exposure modeling for site-specific applications can be anticipated to yield results that
are representative of the materials to be considered for release with residual levels of
radioactivity. Results from site-specific analyses can be expected to yield results that differ from
generalized models whenever material properties or site characteristics are unusual. Since one
or the other of these conditions is commonly encountered, generic modeling results can be
misleading unless the results are subjected to careful interpretation. Likewise, results from
site-specific modeling need to be carefully considered against results from generic models to
ensure site-specific models incorporate all of the significant pathways of exposure. Research
models have an important role in determining the appropriateness and correctness of the
simplified model components typically invoked in generic or site-specific models. Consequently,
each type of exposure model has an important role in providing protection to public health and
safety, and no one type of model should be ignored or excluded when considering an approach
to modeling a specific disposal system.
CONCLUSIONS
Exposure modeling provides a basis for the application of sound and reasoned principles
in radiation protection to the release of materials with residual levels of radioactivity. Exposure
modeling supports criteria for the release of materials that are technically defensible and
consistent while having the flexibility of examining specific or unique applications. As a result,
exposure modeling supports rulemaking, the determination of compliance with existing rules and
the improved understanding of the consequences of the release of materials with residual levels
of radioactivity. By the use of generic exposure models, criteria can be established that define
overall performance objectives and specific exemption levels. Exceptions from specific exemption
levels for unique materials or sites can be accommodated by the application of site-specific
exposure modeling.
REFERENCES
[1 ] International Commission on Radiological Protection, Limits for Intakes of Radionuclides
by Workers, ICRP Publication 30, Part 1, Ann. ICRP 2, No. 3/4 (1979).
[2] International Commission on Radiological Protection, Recommendations of the
International Commision on Radiological Protection, ICRP Publication 26, Ann. ICRP 1, No.
3 (1977).
[3] O. I. Oztunali and G. W. Roles, Update of Part 61 Impacts Analysis Methodology -
Methodology Report, NUREG/CR-4370, U.S. Nuclear Regulatory Commission (1986).
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[4] D. W. Lee, Pathways Analysis Model for Below Regulatory Concern (BRC) Uranium
Wastes, ORNL Report, Oak Ridge National Laboratory, Oak Ridge, Tennessee, in
preparation.
[5] G. T. Yeh and D. S. Ward, FEMWASTE: A Finite-Element Model of WASTE Transport
Through Saturated-Unsaturated Porous Media, ORNL-5601, Oak Ridge National Laboratory,
Oak Ridge, Tennessee, 1981.
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Table 1
Assumptions Applied to BRC Disposal of Depleted Uranium
Offsite Exposure Scenario
Contaminated groundwater consumption
Point of consumption at facility boundary
Exposure occurs at highest concentration
Intruder Exposure Scenario
Intrusion occurs 50 years after facility closure
Foodstuffs consumed that are grown in contaminated soil
Contaminated groundwater consumption
Contaminated milk and meat are consumed
External exposure and inhalation
Site Performance
Dilution of leachate occurs in unsaturated zone by infiltrating
precipitation
Consumption of contaminated water occurs at the top of the saturated
zone without aquifer dilution
50% of site surface area is dedicated to disposal
Dilution of leachate determined by application of FEMWASTE code
Waste Form and Disposal Facility
50% of waste is soluble
Release of soluble portion occurs uniformly over a 10-year period
50% of the leachate generated is collected by the leachate collection
system
Waste is diluted by a factor of 4 within the disposal unit by
uncontaminated materials
Waste resembles soil at the time of intrusion
Waste and uncontaminated soil are uniformly mixed across the disposal
site area upon intrusion
Average concentrations in disposal units are the concentrations in
each and every dumpster disposed of at the disposal facility
FEMWASTE assumptions that include:
Leachate concentrations controlled by solubility limits
Kd determined by lowest experimental equilibrium value
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Table 2
BRC Limits for Depleted Uranium Calculated by Exposure Modeling
Dumpster Size
(Cubic Yards)
6
10
12
Activity
(uCi)
Mass
(g)
200
300
400
400
700
900
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DOE Guidelines and Modeling in
Determination of Soil Cleanup Guidelines
Andrew Wallo 111
U.S. Department of Energy
Environmental Guidance Division
ABSTRACT
This presentation will summarize the current guidelines used by the Department of Energy
for their remedial actions under the Office of Nuclear Energy's Formerly Utilized Sites Remedial
Action Program and Surplus Facilities Management Program. It will discuss soil criteria and
surface contamination limits used for buildings and equipment. A brief description of the
computer code, RESRAD, which is used to develop soil criteria and estimate doses from residual
radioactivity will also be provided.
INTRODUCTION
The Department of Energy currently implements remedial actions and decontamination
projects at its Formerly Utilized Sites Remedial Action Program (FUSRAP) sites and its Surplus
Facilities Management Program (SFMP) sites using the "Department of Energy Guidelines for
Residual Radioactive Material at FUSRAP and SFMP Sites", Revision 2, March 1987. The
Guidelines are supported by a computer code, RESRAD, which is used to estimate potential
doses from residual radioactive material and support the development of soil cleanup criteria.
This presentation will provide a brief history of the development of the guidelines, some general
discussion of the guidelines and a general overview of the RESRAD code.
BACKGROUND
In the late 1970's and early 1980's the Department of Energy began the survey and
ultimate decontamination of several facilities that had been used by the Departments predecessor
agencies in the development of nuclear energy. Criteria and standards for these sites were
proposed on a site-by-site basis by the field offices responsible for the specific remedial action.
The criteria and approaches varied from site to site. In reviewing the criteria DOE headquarters
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realized that general guidelines were needed and at their request the national laboratories made
several attempts to generate generic criteria. In 1983, DOE consolidated these criteria and
procedures in order to identify generic soil criteria for all remedial actions.
The DOE headquarters established a working group comprised of representatives from
concerned program offices; the Office of Environment, Safety, and Health; DOE Operations
Offices; and the involved national laboratories. The working group activities were also
coordinated with the U.S. Environmental Protection Agency and the U.S. Nuclear Regulatory
Commission. These representatives attended several meetings of the working group in an
advisory capacity.
The initial charter of the working group was to develop an acceptable set of generic soil
criteria. However, the recommendations from the first working group meeting resulted in an
action which expanded the charter to cover all aspects of decontamination. Because potential
doses from residual radioactivity are dependent on many site specific factors, the working group
also recommended that a generic procedure for developing soil limits be developed rather than
generic concentration limits. This recommendation resulted in the development of the RESRAD
computer code which supports the development of soil criteria. The following is a summary of
the working group's major recommendations:
The guidelines should be consistent with other available standards where they are
appropriate including the EPA standards for Uranium Mill Tailings Remedial Action
(UMTRA) (40 CFR 192); NRC surface contamination limits (Guidelines for
Decontamination of Facilities and Equipment Prior to Release for Unrestricted Use
or Termination of Licenses for Byproduct, Source or Special Nuclear Material, July
1982); and DOE Orders.
It was recommended that DOE adopt the standards, guidance and models in
publications 26 and 30 of the International Commission on Radiological Protection
(ICRP). This includes the use of the concept of effective dose equivalent rather
than critical organ dose (ICRP publication 2) for defining dose limits and that a
100 mrem/year effective dose equivalent be adopted for longterm exposures
instead of the 500 mrem/year whole body limit then in effect.
Soil criteria for radionuclides other than those covered under 40 CFR 192 (radium
and thorium) should be derived for each site to ensure compliance with the dose
limit. These criteria should be based on a conservatively assumed plausible-use
(realistic) scenarios.
The as low as reasonably achievable (ALARA) process should be incorporated
into the guidelines for all phases of a remedial action.
No attempt should be made by DOE to define "de minimis" or "below regulatory
concern" levels.
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GUIDELINES ;
The guidelines were developed in accordance with these recommendations and were first
issued to the field for implementation in February'of 1985. The current guidelines are revision 2,
dated March 1987 (they are included as DOE-wide guidelines in DOE Order 5400.5, "Radiation
Protection of the Public and Environment"). The revisions were generally procedural in nature
and were directed toward resolving implementation problems experience with the earlier versions.
The guidelines include a primary dose limit of 100 mrem/yearfor remedial actions and a
requirement to bring potential doses as far below that limit as is reasonably achievable (the
ALARA process). It also adopted as secondary limits, the 40 CFR 192 concentration limits for
radium and thorium in soil as well as the radon standards for buildings. The Nuclear Regulatory
Commission (NRG) surface contamination limits ("Guidelines for Decontamination of Facilities and
Equipment Prior to Release for Unrestricted Use or Termination of Licenses for Byproduct,
Source or Special Nuclear Material", NRC, July 1982) were included as guidelines to be used for
the release of usable buildings and equipment.
Soil criteria for radionuclides other than radium and thorium are derived and are applied
over each 100 square meter area of a site. Hot spot criteria are also included to ensure no
adverse impacts will result due to averaging the residual radioactive material over 100 square
meters. Guidance is provided for interim and longterm management of remedial action residues
and wastes.
Procedures for supplemental limits or exceptions to generic or derived limits are also
provided in the guidelines to address those situations where special site-specific circumstances
indicate they are required. In general, the use of exceptions and supplemental limits is
discouraged. Exceptions are used in those cases where remediation cannot effectively reduce
residual material to levels that will allow use without radiological restrictions. An exception may
be granted when the above condition is demonstrated and when it is shown that partial remedial
measures coupled with controls on use of the site will adequately protect the public.
Supplemental limits are applied to a portion of a site or facility for which the derived or generic
limits are not appropriate and because of the nature of the area either more stringent or less
stringent limits are necessary. However, supplemental limits do not consider use restrictions in
their development. The steps and requirements for justification of either exceptions or
supplemental limits are described in detail in the guidelines.
MANUAL AND RESRAD COMPUTER CODE
As noted above the guidelines are supported by a separate but integral implementation
manual (DOE/CH/8901, "A Manual for Implementing Residual Radioactive Material Guidelines")
and computer code (RESRAD). The manual describes the pathway methodology to be used in
deriving soil criteria and the associated dose and transport factors necessary for their
development. It provides guidance on the application of the ALARA process and on the use of
hot spot criteria. The manual also describes the computer code RESRAD and its use.
RESRAD can be used to calculate site-specific soil guidelines on the basis of any selected
dose limit. It may also be used to calculate doses to a hypothetical on-site resident or worker
(maximally exposed individual member of a critical population group).
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Dose calculations are done for the external radiation exposure pathway and the internal
exposure pathways including ingestion of drinking water (surface or groundwater), vegetation,
meat, milk, and fish, and inhalation. The program contains vegetable/soil transfer factors, food
transfer factors, radidriuclide distribution coefficients, dose conversion factors and decay and
ingrowth data. Input parameters include:
Physical parameters such as size, thickness, density, and porosity
of contamination, the cover and subsurface layers.
Hydrological parameters including conductivity, gradient, and water
table depth.
Geochemical parameters such as distribution coefficients.
Meteorological parameters including precipitation,
evapotranspiration, erosion, runoff and mass loading.
Usage and consumption parameters for the scenarios.
All of the input parameters, can be site-specific and entered by the user. Under
circumstances where information on the site is limited, default values can be used. However, the
dose conversion factors, and the uptake and transfer factors are standard and are not readily
changed by the user.
The current version of RESRAD and the associated methodology does not address the
indoor radon pathway. At present the DOE Guidelines have adopted the 40 CFR 192 radium
limits and as a result, this pathway has not yet been included. The code does not compute
collective dose to offsite individuals. If the site under review is large enough that such
evaluations are warranted other models are used for offsite dispersion.
The code runs on IBM Compatible PC and PS/2 computers with a hard disk drive and
400K of memory. A math coprocessor and mouse are recommended but not required. The
program is menu driven and very user friendly. Internal interactive help files are available.
SUMMARY
Use of the DOE guidelines (100 mrem/year plus ALARA) with RESRAD and the associated
methodology described in the manual to derive soil limits for DOE remedial actions has been
successfully applied in the field. Generally, during planning, an authorized concentration limit for
cleanup is selected at a level that would result in doses, under current use or realistic use
scenarios, that are well below the 100 mrem/year dose limit. The ALARA process is also
extended to field operations which ensures that the user of the site or area will not receive doses
anywhere near the 100 mrem/year limit.
While this code, the associated methodology and the supporting manual were developed
to support the DOE Guidelines, the methodology and code can be a useful tool for determining
potential doses to individuals from soil contamination and in evaluating various remedial action
options. The code and methodology offers a consistent approach with considerable site-specific
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flexibility. Once the site specific data has been entered into the data fields, the program
computes potential doses and soil guidelines (based on a specified dose) quickly and, hence,
can be used as a tool to optimize cleanup options or even to support certain ALARA process
determinations.
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Session IV
Desirable Characteristics for Criteria
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International Similarities and Differences
in Regulating Nonradiation Hazards
Rob Coppock
Staff Director
Panel on Policy Implications of Global Warming
Committee on Science, Engineering, and Public Policy
National Academy of Sciences / National Academy of Engineering
Institute of Medicine
ABSTRACT
Assessment of the risk of cancer and other health effects, and of injury or other
deleterious consequences of technological hazards, involves intricate interactions
between science and policy. These interactions are best understood in terms of (1) a
gradual overlay or replacement of traditional administrative approaches to regulation with
"scientific" principles and methods, and (2) the emergence of different styles or
approaches to regulation in different countries.
All regulatory styles involve some mix of consultation, reliance on expertise, self-
regulation, and consideration of political context. But variation in their respective
contribution in different countries leads to characterization of "typical" regulatory styles:
evidential (United States); consensual (Sweden, United Kingdom); authoritative (France,
Japan); corporatist (The Netherlands, West Germany). The persistence of these political
and administrative traditions suggests that they will hold for other areas of regulatory
action.
INTRODUCTION
Many people, when travelling in foreign countries, observe first-hand that things
work differently overseas. This was brought home to me rather forcefully on my first visit
to Britain. I had been out of college only a couple of years, and was to visit the London
office of the contract research firm for which I worked. When I arrived at my hotel, a
message from the head of the local office was waiting. It said, "Please ring before you
call." I was utterly baffled. Fortunately, I had been a student who plagued his professors
with questions. I asked the concierge at the hotel what the message might mean. Of
course he was able to provide a translation: I should telephone the office before visiting.
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The Japanese visitors at this meeting may have had no trouble with such a
message. But for Americans, both "ring" and "call" are used most commonly in reference
to the telephone, which obviously is not the case in Britain. Such differences in American
and British English are common. It has been said that the United States and the United
Kingdom are two countries separated by a common language.
Confusion based on different use of the same terms is not limited to everyday
experience. For years I have studied the use of scientific evidence and judgment in
regulatory decisions in different countries. I have observed on several occasions the
frustration of European researchers as they attempted to gain wider application of some
of the concepts of risk assessment and cost-benefit analysis. I believe they often used
the same concepts to describe quite different procedures because they were thinking in
terms of the processes that occurred at home in their own countries. Since those
processes differed, they had slightly different, sometimes quite different, meanings for the
concepts.
Today I will address similarities and differences in the regulation of nonradiation
hazards in different countries. The meeting organizers originally asked me to talk about
similarities and differences in the clean-up of residues of radioactive materials at storage
and disposal sites. However, despite considerable effort, I have been unable to locate
sufficient published materials on that subject. So I must rely instead on topics I have
studied and can discuss with some degree of authority. I hope that the participants will
be able to extend my observations to the topic of the workshop.
DIFFERENT RELIANCE ON SCIENTIFIC EVIDENCE AND ADVICE
On August 8,1987, a resident doctor of the hospital at Osaka University received
a liver transplant at the Cromwell Hospital in London.[1] Since no definite policy on
organ transplants existed in Japan, the transplant had been arranged overseas. There
was no study comparing the risks and benefits of liver transplants and other more
conventional therapies in Japan. This is in sharp contrast to the situation in the United
States, where statistics on success rates and other indices of medical performance are
commonly available for separate hospitals and sometimes individual physicians.
This example highlights an important difference between the typical approaches
to pubic health policy in the two countries. Japan seems to rely on guaranteed safety
based on an established, formal set of relationships among organizations and individuals
rather than on explicit information about the specific situation as is typical in the United
States. Furthermore, the general preferences in the two countries apply in a variety of
settings. We can anticipate that they will influence the development of the radiation clean-
up criteria under discussion at this workshop as well.
To set the stage for a more detailed look at the relevant influences on regulatory
approaches, let me describe what has been called the "emerging scientific state."
THE EMERGING SCIENTIFIC STATE
Since the 1960s, it has been common to consider interactions between science
and public policy in terms of both "science for policy" and "policy for science."[2] The
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concept of science for policy refers to the fact that modern governments depend on
science to perform, or at least provide input for, many of their traditional functions. Policy
for science derives from the emergence of science as a national resource. It is the first
of the two, science for policy, that concerns us here.
In describing what they call the "scientific state," Schmandt and Katz [3] identify
three principal characteristics of the interactions between science and policy: (i) science
affects the public policy process in distinctive ways that can be identified as stemming
uniquely from the realm of science; (2) the resulting changes gradually replace, modify,
or supersede, as the case may be, older agenda, organizations, and procedures of policy
making and administration; and (3) science induces change in all industrial countries, but
the resulting social environments differ, depending on national circumstance, traditions,
and goals. Schmandt and Katz point out that science enters the policy process in three
distinctive ways: as product, as evidence, and as method.
The search for socially acceptable policy responses to complex and uncertain
hazards often pits new options against more familiar legal and political options. Science
is often seen as providing these new policy options. In other words, science produces
policy products for consideration.
But in another sense, or perhaps in other instances, science provides information
about all the options under consideration. The interpretation of scientific evidence
concerning the likely consequences of alternative options is a central facet of
contemporary industrial states. The political and administrative systems vary from country
to country, however, and it should be no surprise that the way these interpretations affect
policy decisions also vary.
Finally, many activities involved in policy making are based on explicit and logical
analysis. From science comes formal methods for analysis of quantitative data, gathered
and interpreted according to accepted method. From the policy environment comes the
requirement for timely action and the acceptance of uncertainty, incomplete information,
and compromise.
Each way that science enters the policy process-^as product, as evidence, and as
method-interacts with the traditions found in various countries.
COMMON CHARACTERISTICS OF REGULATION
Although there are differences in the political and administrative traditions, the
basic issue-the choice among complex and uncertain options-is similar everywhere.
Four major characteristics describing regulatory approaches used in all countries have
been identified: consultation, dependence on expertise, self-regulation and self-policing,
and political considerations.^]
Some form of dialogue or consultation among the affected parties is a feature of
all regulatory approaches, although there are notable differences in what is discussed,
when, and with whom. Consultation is often an unavoidable part of the political and
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administrative system, but even when not legally required it is often used to help reach
agreement about the proposals that are most likely to succeed.
Experts with special understanding and knowledge provide vital input to all
regulatory processes. The nature of interactions between experts and public officials vary
among countries, however, and attitudes about expertise, consultation, and participation
are changing the nature of the expert's role.
Some degree of self-regulation and self-policing is a part of every regulatory
regime. But some countries rely oh self-regulation to a much greater degree than others,
making it a virtual cornerstone of their whole approach.
In the past, regulatory decisions were commonly thought of as primarily technical
issues. Now, in contrast, political and social factors often appear to outweigh technical
concerns. On occasion, options are chosen more for their political palatability than their
technical feasibility.
Let me now turn to the way approaches to regulation vary.
TYPICAL STYLES OF REGULATION
Four main approaches to regulation have been identified: evidential, consensual,
authoritative, and corporatist.[5] A particular style is usually dominant in a given country,
although different styles may be adopted by some agencies or organizations within the
same country. There does appear to be a tendency toward convergence among
countries, with less marked differences than was the case a few decades ago. This
would be consistent with the notion that the "scientific state" is gradually replacing more
traditional administrative procedures throughout the industrialized world.
THE EVIDENTIAL APPROACH
.*
In the evidential model, for which the United States is the prime example,
regulatory approaches result in the involved parties becoming adversaries, often arguing
the facts and interpretations of them in legal battles. Because of these conflicts,
regulations are increasingly formulated in terms of precise targets, standards, and
procedures. There is also separation of powers, and dispersion of authority among units
and levels of government. There tends to be complex structures of interest groups inside
and outside government, and the configuration of those coalitions shift with issues rather
than party politics.
* This category is commonly labeled "adversarial" in the literature. I believe,
however, that "evidential" more accurately describes the key factors that underlie its
differences with other approaches.
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• THE CONSENSUAL APPROACH
A regulatory approach aimed at achieving consensus may confine its consultation
to an elite group of civil servants, experts, and influential politicians and industrialists, as
in the United Kingdom. Or it may extend involvement more broadly, as in Sweden.
However, the elite consensual approach, common in several European countries, differs
from the evidential approach in the United States in several ways. Explicit specifications
and formal rules are avoided, allowing government considerable discretion when
addressing specific circumstances. Procedures tend to be more intrusive~in order to deal
with each case individually, it is necessary to have detailed information about it. Dealing
with each case in detail requires considerable confidentiality, especially with regard to
proprietary commercial data. Both intrusiveness and confidentiality encourage self-
regulation with little legal prosecution but considerable, and generally effective, moral
suasion. The consensual approach often results in development of an "elite club" of
those involved, which treats those not involved with disdain or even contempt. This
makes it doubly difficult for outsiders to influence the development of consensus politics.
THE AUTHORITATIVE APPROACH
Regulation by centralized authority is most often found in countries like France or
Japan which have a strong central government and a weak legislature, and where
regional or local government is largely limited to executing decisions issued by national
bodies. In this approach, regulators have considerable leeway in setting standards and
enforcing compliance. Consultation with relevant parties tend to be infrequent, although
those that take place tend to be crucial to the outcome. Regulations are usually imposed
and enforced with limited scope for legal redress unless officials can be shown to have
acted arbitrarily or contrary to statutory procedures. The implementation of regulations
generally leave room for negotiations recognizing the unique attributes of the specific
situation.
THE CORPORATIST APPROACH
The corporatist regulatory style, most often associated with the Netherlands and
West Germany, combines aspects of the analysis found in the evidential approach with
the consultation found in consensual models. The general framework of regulations, and
sometimes the specific content, are established in groups where individuals "represent"
the relevant economic or social interest groups. There are many consultative bodies and
advisory mechanisms. These groups tend to be construed in ways that correspond to
the configuration of relevant interests. Evidence is used more to determine the
boundaries of choice rather than determine the preferred outcome. Debate may be open
to anyone, but decisions are generally in closed groups so as to maintain the established
balance of interests. Corporatist approaches tend to restrict the influence of smaller or
newer groups.
PRACTICAL IMPLICATIONS OF DIFFERENT APPROACHES
Let me now speculate about the implications of these differences in approach for
the development of criteria for radioactive materials in Japan and the United States. If
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control of air pollution can serve as an example, we may expect Japan to formulate
general goals for clean-up and to specify levels of contamination as "guidelines."
Although this process would involve considerable analysis and consultation with experts,
the relationship of the two might not be extensively documented. The determination of
the exact level of permissible residual contamination would be determined on a case-by-
case basis, subject to extensive negotiation. The United States, in contrast, is more likely
to develop explicit levels of permissible contamination for general application in a lengthy
rule-making procedure subsequently defended in court. The rule-making decision would
likely refer to extensive evidence and support analysis. Extensive records would likely
document the handling of each site.
CONCLUSIONS
In this short talk, I have described several aspects of the similarities and
differences among the approaches to regulating nonradiation hazards in different
countries. I also briefly suggested how these observations might apply to the
development of criteria for clean-up of radioactive materials. My comments have probably
raised more questions than provided answers. I would be happy to try to answer any
questions you might have.
REFERENCES
[1] Morioda, Hohru, Dr. Eng., Risk Assessment and Risk Mangement in Japan -
Administrative Guiding without Explicit Guidelines and Probabilistic Concepts,
Second US-Japan Workshop on Risk Management, Ocotber 26 - 30,1987, Suita,
Japan.
[2] Brooks, Harvey, The Government of Science, Cambridge, MA: The MIT Press,
1968; and Mesthene, Emmanuel G., and Schmandt, Jurgen, Science and the
Politics of Government, Paris; Organization for Economic Cooperation and
Development, 1963.
[3] Schmandt, Jurgen and Katz, James Everett, The Scientific State: A Theory with
Hypotheses, Science, Technology, & Human Values, 11 (1968):40-52.
[4,5] O'Riordan, Timothy, Approaches to Regulation in: Otway, Harry and Peltu,
Malcom Regulating Industrial Risks: Science, Hazards and Public Protection
(London: Butterworth's, 1985)20-29.
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What are the Basic Requirements that Cleanup
Standards Should Satisfy?
Allan C.B. Richardson
Chief, Guides and Criteria Branch
Office of Radiation Programs
U.S. Environmental Protection Agency
It is essential that we, as a society, be clear about what we expect to achieve through
standards for cleanup. Fortunately, in radiation protection, as opposed to protection against
other environmental threats, we have been thinking about the basics for a long time. As a result,
our objectives are well developed.
Most of these objectives were laid out almost forty years ago, in the late 1950's, and have
changed only slightly since then. There are three; they were first set forth as guidance for
regulation of radiation exposure by Federal agencies, by the Federal Radiation Council, in 1960
(Figure 1):
1. "There should not be any man-made radiation exposure
without the expectation of benefit resulting from such
exposure."
2. "Every effort should be made to [keep] radiation doses
as far below [the recommended limits] as practicable."
3. "[Recommended limits] for the general population are:
0.5 rem per year (whole body)
5 rem in 30 years (gonads)"
Figure 1. U.S. Federal Radiation Council (1960)
These three objectives have been characterized, by the International Commission on
Radiological Protection (ICRP), as the principles of justification, optimization, and limitation; we
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will use that convenient terminology here. Justification is a judgment that is applied when a
practice involving the use of radiation is introduced. In the 1960 Federal guidance, it was stated
as a requirement that there should not be any radiation exposure without the expectation of a
net benefit. This requires a broad judgment. Often, it is a political one. It is generally not made
by radiation protection professionals, although they usually have the opportunity for input.
The second objective, optimization of protection, is more commonly known in the United
States as the ALARA ("as low as reasonably achievable") principle. It has been with us since the
early 1950's, and flows from the assumption that there is a linear relationship between radiation
dose and the risk of deleterious effects on health. This assumption leads to the conclusion that,
since any dose carries some risk, dose should be maintained at the lowest level that is
economically reasonable.
The third objective is limitation of risk to the individual. In the 1960 Federal radiation
protection guidance it is expressed in terms of dual limits on dose to members of the general
population, one for exposure of the whole body and one for genetic exposure. It must be
satisfied without regard to the economic cost. This third requirement is needed to supplement
the first two because they assure only that the effects of radiation on the population as a whole
is reduced to reasonable levels. They do not assure that the distribution of these effects is such
that no individual receives an unreasonably large share of the risk from exposure to a particular
source.
These three principles were set forth in essentially the same form by the ICRP in their
Publication No.2 in 1959, the year before the 1960 Federal guidance was issued. We have noted
that there were dual limits for the individual. These disappeared in 1976, when the ICRP revised
its recommendations (ICRP Publication No.26), in the next major statement of basic principles.
This is shown in Figure 2.
1. No practice shall be adopted unless its introduction
leads to a positive net benefit.
2. All exposure shall be kept as low as reasonably
achievable, economic and social factors being taken into
account.
3. The dose equivalent to individuals shall not exceed the
limits [0.5 rem in any year, and, in the case of chronic
exposure, 0.1 rem per year.]
Figure 2. International Commission on Radiation Protection (1977)
(as interpreted by the "Paris Statement" of 1985)
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The three principles remain. However, one important distinction was introduced in the
application of limitation. The ICRP now makes a distinction between limits for short-term and for
long-term exposure. There is one limit for chronic exposure, say, from a planned release from
a fixed facility, and a second limit for non-recurring short-term exposure, e.g. from a contaminated
site prior to cleanup. This distinction evolved gradually from the 1976 ICRP recommendations,
and it was not stated explicitly until 1985.
Notice that the phrase "as low as practicable" in the former FRC Guidance has now
become "as low as reasonably achievable, economic and social factors being taken into
account," a much more realistic statement of this requirement.
As some of you know, the Environmental Protection Agency (EPA) inherited the functions
of the former Federal Radiation Council when EPA was formed in 1970, and in 1987 we revised
the Federal Radiation Protection Guidance for workers. These new recommendations replaced
that portion of the 1960 Federal Guidance that applied to workers. It was developed
cooperatively by all of the affected Federal agencies.
We have also recently convened a new Federal interagency group to develop guidance
on residual radioactivity. As a first task this group is developing recommendations to update the
remaining portion of the 1960 Federal guidance, which applies to the general public. That work
is now near completion. Figure 3 shows excerpts from the current draft. Although I can
guarantee that the words will change before it is published for public co'mment, the basic ideas
should remain the same.
"For those sources of exposure covered:
1. "There should be no exposure of members of the
general public. . . unless it is justified by the
expectation of an overall benefit...
2. "A sustained effort should be made to ensure that doses
to individuals and in populations are maintained as low
as reasonably achievable."*
3. "The combined radiation doses incurred in any year from
all sources of exposure ... should not normally exceed
a limiting value of 0.1 tern ...
"[This] includes consideration of economic and social factors, and applies to radiation exposure
that may occur now and in the foreseeable future."
Figure 3.
Draft Revision of Federal Radiation Protection Guidance (1989)
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As you can see, the basic objectives have not changed dramatically in the thirty years
since 1960. There is, however, one significant change that is not embodied in previous
statements of basic principles. That is, a distinction is now made between limits on dose from
individual sources of exposure and limits that apply to the sum of doses to an individual from all
sources. This distinction has been evolving in the ICRP, as well as in regulatory practice in the
United States. The ICRP now places primary importance on limits that apply to individual
sources, and much less on limits for total dose to individuals. Limits for individuals apply to all
of the various sources of exposure that an individual may be subjected to, and, of course, is
achieved - in a regulatory fashion - by applying limits to individual sources. A consequence
of this, of course, is that no individual source can take up all of the limit on allowable dose to an
individual.
This was expressed well in a recent publication of the International Atomic Energy Agency
(IAEA) (Figure 4). What they said is perhaps obvious - the problem is to account for the
presence of other sources, the continued operation of these other sources in the future, and the
introduction of new sources. For this reason, they introduced the idea of a source-related limit
that is lower than the dose limit - a source upper bound.
"Since [the primary dose] limit is individual-related ..., account
must be taken of the presence of other sources, the continued
operation of these sources in the future, and the eventual
introduction of new sources. For this reason a source-related limit,
lower than the dose limit and called a source upper bound, must be
set... as the boundary condition for optimizing." (IAEA Safety Series
77, 1986)
"Authorized limits for sources will normally be a fraction of the
limiting value for individual dose from all sources combined." (Draft
revision of Federal guidance, 1989)
Figure 4. Source-related limits
Notice that this is not the limit for sources; it is just the upper bound for source-specific
limits and actual limits will generally be lower, because they also must take account of
optimization of control. In the draft revision of Federal guidance for the general public (Figure
4) we have addressed these problems through simple recognition that authorized limits for
sources will normally be a fraction of the limits for dose from all sources combined.
The distinction between individual and source-related limits has been recognized in the
U.S. for some time. It is reflected by a series of source-related standards established during the
last decade and a half - usually in response to public opinion expressed in the form of statutory
mandates from Congress. The standards established under the Clean Air Act (40 CFR Part 61)
for a host of types of sources of radionuclide emissions is the most recent example. Previous
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examples include the EPA's standards for the nuclear fuel cycle (40 CFR Part 190), high level
waste (40 CFR Part 191), and uranium mill tailings (40 CFR Part 192).
We have not had to give serious attention to the idea of a source upper bound in the U.S.
because most of the standards we have set have been based, at least in part, on the application
of optimization, and have resulted in values that were low compared to individual-related limits.
That is, the individual limit in this country, which is still five hundred millirems per year, has never
been in danger of being exceeded. This would still be the case with an individual limit of 100
millirems per year, as currently recommended by the ICRP.
For residual radioactivity this may no longer be true, because it is highly unlikely that
optimization considerations are going to be controlling. That is, individual risk is going to be the
most important consideration for residual radioactivity, not optimization, simply because the
collective doses involved are likely to be small in most cases, and the costs high.
• ~ , - ' " f
We have seen that three principles govern decisions for radiation protection, in general.
These are justification, optimization, and limitation. Just how useful are they going to be in the
case of residual radioactivity? The first, justification, is not particularly helpful for decisions on
criteria for residual radioactivity. Presumably, justification decisions were in place long before the
need for cleanup of a site for a new use came along. Some take the point of view that no
exposure is justified from residual radioactivity because there is no benefit for a practice which
has been discontinued. But this is not a useful approach/since obviously it is not practicable
to impose a requirement of zero dose. Others (including the ICRP) argue that unless there is a
net benefit from cleanup, it should not proceed. That is a judgment not easily achieved or, in
most cases, defended. Clearly, in some cases the effort will be great, but necessary in order to
achieve acceptable levels of individual risk. And in others careless treatment of natural resources
cannot be rewarded by a decision that cleanup is not "justified" because the effort required from
the owner is too great compared to the benefit. We have to conclude that justification is not a
very useful consideration for establishing cleanup criteria.
Optimization, on the other hand, certainly can be applied. In practice, however, unless
the contamination is very long-lived or inexpensive to remove it will not usually lead to individual
doses that are acceptably small. Of the three radiation protection objectives, I believe that it will
usually be found that limitation is the most important, because we will usually be faced with the
need to decide what fraction of the overall limit on dose to an individual it is appropriate to allow
residual radioactivity to take up.
Next, we must consider whether there are other applicable or useful concepts or criteria,
aside from the general principles for radiation protection. There are several; these are shown in
Figure 5. The first is the concept of "acceptable risk" as applied to regulation of chemicals in this
country. There is a growing consensus that risks on the order of one in ten thousand to one in
a million over a lifetime are appropriate regulatory objectives. Risks larger than these are only
acceptable in special situations where extenuating circumstances exist, such as during
emergencies.
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o Acceptable risk: 1 (T4 to 10'6 lifetime
o Groundwater policy:
- Groundwater classification
- Drinking water standards
o Below regulatory concern:
- International guidance
- NRC Proposals for exempt practices
- EPA Proposals for low-level wastes
Figure 5. Other Constraints on Cleanup Criteria
The radiation doses that correspond to this range of lifetime risk are disconcertingly low.
The range is on the order of 0.04 to four millirems per year, for lifetime exposure. Expressed
another way, it is essentially a range of from background to background plus four millirems per
year.
A second set of useful criteria may be found in national groundwater policy. These are
established in legislation (e.g. the SARA amendments to CERCLA) and contain some very tight
and rather specific requirements for cleanup, with respect to groundwater. For example, under
current interpretation and practice, if groundwater meets the classification criteria for current or
potential use as drinking water (and that means most groundwater) then the applicable standards
are those for drinking water. That standard is four millirems per year for man-made radioactivity.
So, with respect to groundwater contaminated as a result of residual radioactivity in the
soil, we are back to the same constraints for water as those noted above for the upper end of
the range of "acceptable" risk.
A final concept, which may be useful in thinking about criteria for cleanup of residual
radioactivity, is that of "below regulatory concern" (BRC). The numbers under consideration for
BRC levels of radiation dose are in the same range as those already mentioned above. For
example, the IAEA has issued recent guidance on acceptable criteria for exemption. For any
specific practice, they recommend individual doses that are on the order of one millirem per year,
coupled with a finding that exemption is the optimal choice, i.e., ALARA. They offer the further
advice that if the practice involves less than 1 man-Sievert (100 man-rem), then you can assume
that it is already the optimal choice, i.e., a further demonstration of ALARA is not required. The
NRC is also considering BRC criteria. The levels they are proposing (10 millirem/year) would not,
however, satisfy current EPA concepts of acceptable risk, or existing drinking water standards.
We can now summarize the requirements for residual radioactivity cleanup criteria. First,
the criteria have to be a fraction of the overall limit for individuals - that is, a fraction 100
millirems per year. Second, they must optimize protection, considering the totality of current and
future health risks in relation to the costs. However, in looking at these health effects, usually
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through projection of collective dose from residual radioactivity (even if the collective dose
involves the world's population and exposures far into the future), we almost invariably find that
this criterion will not be as limiting as the first. (The primary case where collective dose is limiting
that I am aware of is Carbon-14, as demonstrated in a study for the IAEA of exemption levels
for low level waste by Kennedy in 1987.)
Finally, we will have to satisfy other policy constraints, such as those of acceptable risk
and groundwater policy. Based on current practice, these may end up involving the same level
of cleanup as is required to satisfy general criteria for BRC, at least at the international level.
I want to make one final point that is often overlooked. We tend to concentrate on dose
as the unit of measure by which we express radiation limits, because it is related directly to risk.
It is probably a mistake to do this in the case of cleanup criteria for residual radioactivity. Dose
limits imply the need for a site-specific assessment of doses to hypothetical receptors for each
cleanup. That is expensive and it often will be difficult to achieve without controversy. Further,
we do simply not need great precision, at each site, in relating contamination levels to the low
levels of risk involved after cleanup. It should be enough to do generic site assessments and
then to set standards expressed in units that are directly related to contamination levels.
What we need, then, are simple numerical cleanup criteria that apply to all sites. These
must satisfy three different kinds of objectives that may each end up being limiting - individual
risk, population risk, or a national policy objective, as in situations when groundwater may be
contaminated. Finally, these numerical criteria should be expressed directly in terms of
contamination levels, or total amount of contaminant.
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What Should Cleanup Standards Do?
V.C. Rogers
Rogers and Associates Engineering Corporation
ABSTRACT
Standards for residual radioactivity on soils and other materials should provide adequate
protection for: individuals using the materials or occupying the site, other critical population
groups that may be exposed due to their proximity to the site or materials, and the general
population at risk from atmospheric transport of contaminants or from ingestion of contaminated
water or food. Past analyses have shown that the individual site reclaimer or recycled material
use is the critical receptor, but that risks to other com'ponents of the public should also be
considered. Doses received by site reclaimers or recycled material users are generally
proportional to the nuclide concentration (pCi/g) for bulk contamination and nuclide surface
concentration (dpm/100 cm2) for surface contamination. Therefore, standards should limit the
concentration of individual nuclides or groups of nuclides. Limiting concentration values should
be based on appropriate exposure scenarios and multipathway risk assessments. For example,
the ease of nuclide removal or leachability from the material should affect the concentration limits.
INTRODUCTION
, This paper addresses the basic requirements for cleanup standards for residual
radioactivity. Specifically, the following questions will be addressed in this paper:
1. What segment of the public is the most critical in terms of radiation
protection?
2. Should only a dose or risk standard be established?
3. Should standards inclu.de limits on concentrations of nuclides?
4. Are multipathway risk assessments necessary for establishing cleanup
standards?
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Standards for residual radioactivity on soils and other materials should provide adequate
protection to the public by ensuring that risks from radiation exposures are at acceptable levels.
Separate consideration should be given to individuals using the materials or occupying the site,
other critical population groups that may be exposed due to their proximity to the site or
materials, and the general population at risk from atmospheric transport of the contaminants or
from ingestion of contaminated water or food. Cleanup standards should also contain provisions
for reducing potential doses to the public to "as low as reasonably achievable" (ALARA).
In general, for nearly all decommissioning and decontamination problems doses to
individuals using the site or using recycled contaminated materials is the critical receptor.
However, risks to the other components of the public should also be considered. In particular,
these risks are a factor in evaluating whether the cleanup level is ALARA.
DEVELOPING CLEANUP STANDARDS
The major steps in developing cleanup standards, shown in Figure 1, begin by collecting
information on contaminated sites: the number of sites, the type and extent of contamination,
and the nature of the material that is contaminated. .In addition costs to clean up the
contamination should be estimated. The increase in costs for cleaning to lower levels of
contamination should also be determined.
The next major step leads to the determination of risk or dose performance objectives.
These risk/dose limits may be established by consistency with risk/dose limits from related
regulations. They also may be developed from a cost/benefit analysis that involves comparing
the decrease in risk from the cleanup to the added costs of the cleanup effort for decreasing
levels of residual contamination.
After the risk/dose performance standard is determined, then the risk assessments provide
the nuclide concentration limits consistent with the performance standard.
It is important to establish performance objectives for cleanup standards. The
performance objectives should pertain to a health or risk limitation to the above components of
the public and should provide a consistent basis for allowing the effective implementation of the
standards. When evaluating doses and risks to populations or individuals in the population,
potential exposures in the long term should be considered in addition to potential exposures over
the near term.
FACTORS INFLUENCING POTENTIAL DOSES FROM RESIDUAL CONTAMINATION
Many factors affect the extent of risk to individuals from residual contamination levels.
The factors, shown in Figure 2, can be grouped into four general categories. The first is the
physical and chemical characteristics of the contamination: Is it easily dispersible? Is it easily
leached by water? If dispersible, is the airborne component respirable? Questions such as these
must be determined before evaluating the dose from a particular contamination level.
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The pathway between the waste and humans is another barrier affecting the potential
dose. Dust loading in the air, transport, decay and dilution in the groundwater are several
pathway-related factors that must be considered. .
The environmental barrier is a third factor and is closely related to the pathway
component. For example, if the contaminant is ingested via groundwater, does direct
consumption of the contaminated water constitute the individual's total intake, or is the
contaminated water also used to irrigate crops and water animals that are eventually consumed
by humans.
The time of exposure is the next factor that influences doses from a given contaminated
site. If the dominant exposure pathway is direct gamma radiation, is the individual in proximity
to the contamination for 75 percent of his total time, 25 percent of his total time, or half of his
working time?
The final factor that affects doses and risks to humans are the nuclide-specific dose
factors, i.e., the dose or dose commitment per pCi inhaled or ingested. Some nuclides with very
large dose factors, may constitute an extremely small fraction of the total activity at the site but
may dominate the dose to an exposed individual because of the large dose factor.
For the component of the public that involves individuals occupying the site or using the
recycled materials, many of the natural barriers identified in Figure 2 are bypassed. In general,
the only remaining significant barriers for this situation are the physical and chemical form of the
contamination, the exposure time as defined by the exposure scenario, and the dose factors.
Because some of the barriers are circumvented, this segment of the public is the critical receptor
and generally forms the basis for limits in cleanup standards. Exposures to this segment of the
public are also highly dependent of the definition of the exposure scenario. Consequently, doses
to inadvertent intruders can vary widely for a given degree of contamination on a site, and it is
important to construct meaningful, reasonably conservative exposure scenarios for this
component of the public.
For the critical population group or the general population, the doses are directly related
to the nuclide inventory, as contrasted with onsite exposures that are directly related to the
nuclide concentrations. If the onsite reclaimer is considered in establishing cleanup standards,
then the performance objective based on an annual dose or risk can be directly related to
individual nuclide concentrations. Since individual nuclide concentrations are also a function of
the scenario characteristics, concentration limits serve to standardize the exposure scenario
considered for the residual radioactivity. Therefore, specific concentration limits remove a
considerable amount of the ambiguity and variability that occurs when standards are based only
upon a dose or health risk performance objective. A dose limit is difficult to implement for
specific sites. An equivalent nuclide concentration cleanup level is needed by the organization
doing the cleanup. If a dose performance objective is the only specific criteria in the standard,
ambiguities can arise in defining the appropriate exposure scenario, in defining the characteristics
and model for the pathway calculations, and in selecting appropriate food chain and dose
conversion factors. An example of the variation that can occur in radionuclide concentration limit
that can occur from a specific dose limit is shown in Figure 3. For a dose limit of 25 mrem/yr the
limiting Cs-137 concentration can have multiple values depending on the amount of clean cover
soil over the contaminated area. As shown in the figure, this concentration limit can vary by four
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orders of magnitude. In converting a dose standard to an implementable concentration limit
extensive dialogue of necessity takes place between the regulatory body and the site owner.
Therefore, limiting the standard to a dose performance objective adds significantly to the site
owner's burden, as well as to the regulatory burden. Experience has shown that, in general, site-
specific concentration limits are driven to their lowest relevant values if the standard contains only
a dose performance objective.
For most radionuclides, multipathway performance assessments of residual contamination
reveals that for alpha-emitters the highest doses occur from inhalation of airborne contaminant.
For mobile beta-emitters, highest doses occur from water transport and ultimate ingestion. For
immobile beta-emitters, highest doses generally occur from biointrusion and onsite agricultural
scenarios, and for a few strong gamma-emitters, highest doses occur from direct gamma
exposure. Because many of the nuclides can be grouped and their characteristics generalized,
concentration limits can be developed for groups of nuclides. In addition, for some industry
sectors, the nuclide mix is sufficiently constant that the standard may be implemented by
specifying a total activity concentration limit.
SUMMARY
Doses to individuals occupying a contaminated site or using contaminated materials
receive the highest doses for a fixed degree of contamination than for offsite individuals or
populations. A dose performance standard, therefore, to a critical individual is an appropriate
basis for cleanup standards. However, the cleanup standards should go beyond this point and
should include nuclide concentration limits for bulk contamination as well as for surface
contamination. The nuclide concentration limits should also consider physical and chemical
characteristic information. Higher standards should apply to nuclides that are not easily removed
and are not mobile unless the gamma pathway dominates the dose. For many circumstances
it may be appropriate to apply group nuclide concentrations.
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DATA ON
CONTAMINATED SITES
. COSTS TO CLEAN-UP SITES
EXPOSURE SCENARIOS
AND PATHWAY
CHARACTERISTICS
MULTIPATHWAY
RISK ASSESSMENT
RISK LEVELS FROM
RELATED REGULATIONS
ACCEPTABLE RISK LEVELS CONSIDERING CLEAN-UP COSTS
AND RELATED REGULATIONS
NUCLIDE CONCENTRATION LIMITS
FROM RISK ASSESSMENT
CONCENTRATION LIMITS FOR NUCLIDE GROUPS
AND AVAILABILITY FACTORS
RAE-102984
FIGURE 1. STEPS IN DEVELOPING STANDARDS FOR RESIDUAL LEVELS OF RADIOACTIVITY
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i
EXPOSURE TIME
ENVIRONMENTAL
PATHWAY
CHEMICAL AND
PHYSICAL FORM
CONTAMINATION
RAE-102985
FIGURE 2. BARRIERS BETWEEN MAN AND CONTAMINATION.
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10'
E
0
-m
CO 10
O
Q
10
x
10
-1
500 rnrjam/yr
100 mrem/yr
10~1 1 10 102 103 104 105
AVERAGE Cs-137 CONCENTRATION (pCi/g)
RAE-102306
FIGURE 3. RELATIONSHIP BETWEEN DOSE LIMIT AND Cs-137 CONCENTRATION LIMIT
FOR SUPERFUND SITE.
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Current Status of Residual Radioactivity
Criteria in Japan
Hideaki Yamamoto and Masao Oshino
Department of Health Physics
Japan Atomic Energy Research Institute
ABSTRACT
In this paper, current Japanese regulations concerning residual radioactivity criteria, and
efforts to establish such criteria, are reviewed. In addition, the authors comment on desirable
forms of residual radioactivity criteria. .
Japanese laws and regulations relating to the utilization of atomic energy or radiation do
not explicitly mention any residual radioactivity criteria. Although some governmental
organizations state policies on reusing nuclear facilities, lands or materials, and efforts to
establish criteria have been started, no conclusions have been reached. Considering these
circumstances, the authors describe their ideas on desirable forms of residual radioactivity criteria
for reusing facilities, lands or materials.
INTRODUCTION
No Japanese regulations concerning the utilization of atomic energy or radioisotopes
explicitly refer to criteria for judging the safety or propriety of reuse of nuclear sites, facilities,
contaminated materials or radioactive wastes. However, unrestricted reuse of some nuclear
facilities has been authorized as a result of judgments based on the current regulations. In this
paper, current Japanese regulations implicitly relating to residual radioactivity criteria, and the
bases for establishment of the criteria, are reviewed. In addition, the authors comment on the
desirable form of residual radioactivity criteria.
REGULATORY SYSTEM OF UTILIZATION OF ATOMIC ENERGY AND RADIATION
The Japanese laws which regulate the utilization of atomic energy and radiation are 'The
law for the regulations of nuclear material, nuclear fuel material and reactors (Reactor Law)" and
'The law concerning prevention of radiation hazards due to radioisotopes, etc. (Rl Law)".
Facilities that are candidates for reuse comprise two groups, which are regulated separately by
these two laws. Facilities in the nuclear fuel cycle, such as reactors or nuclear fuel reprocessing
plants, are regulated by the "Reactor Law". Radioisotope handling facilities and accelerators are
regulated by the "Rl Law".
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RULES REGULATING DISMANTLING OF REACTORS AND DISCONTINUATION OF
RADIOISOTOPE HANDLING
Reuse of a facility begins after most of the utilization or handling of atomic energy or
radiation has ceased. Although a permissible residual radioactivity level at this point is not
provided in the above-mentioned laws, some rules for decommissioning are prescribed. The
"Reactor Law" provides rules on dismantlement of a reactor, ordering the owner to submit a
report to the responsible Minister. The report should cover methods to dispose of nuclear
materials and to decontaminate the reactor. If the Minister judges the cleanup methods to be
inappropriate, he can order the owner to improve his methods. For example, appropriateness of
the residual radioactivity level set by the owner for his dismantled reactor is judged by the
Minister. Each case is judged individually (case by case basis).
Similar rules are prescribed for radiation utilization in the "Rl Law". Each submitted plan to
decontaminate and to set a residual radioactivity level is judged by the responsible Minister.
EXAMINATION CONCERNING ESTABLISHMENT OF RESIDUAL RADIOACTIVITY CRITERIA
In the following sections, the basis for establishing residual radioactivity criteria by
governmental organizations is reviewed.
Reuse of decommissioned reactors
The decommissioning of the Japan Power Demonstration Reactor, which was used for
research and development work related to power generation, is now in progress. At the start of
the decommissioning project in 1985, the Nuclear Safety Commission, one of the advisory organs
of the Japanese government, issued a report on its regulatory policy for the decommissioning
research reactors.
In the report, the Commission stated two conditions that should be confirmed when the
decommissioning process is completed: 1) all of the nuclear materials in the reactor should be
removed, and 2) the radioactive wastes generated in the operation should be properly disposed
of. The report suggested that the radioactive waste disposal methods should include reuse or
recycling of extremely low-level radioactive wastes. However, this report did not refer to any
concrete criteria for a residual radioactivity level.
Reuse of low-level radioactive waste repositories
The Nuclear Safety Commission issued a report in 1985, concerning regulatory policy on
land disposal of low-level radioactive solid wastes (LLW). In this report, the Commission made
suggestions concerning reuse of LLW repositories, and on residual radioactivity criteria for them.
The Commission pointed out that after radioactive materials in the wastes have decayed
to a sufficiently low level, restricted reuse may be possible, under certain institutional controls to
prohibit or restrict any actions that possibly could pose a hazard to the public.
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In addition, the report referred to a stage at which a repository could be exempted from
any regulatory control. A repository at that stage could be reused without restoration. According
to the report, the primary considerations for exemption of a repository from regulatory control are
the levels of radioactivity concentration in the disposed wastes and in the soil of the repository.
The report stated that, using an appropriate dose estimation model, residual radioactivity criteria
(below%hich no radiation control should be needed) could be derived from a certain level of the
radiation dose to the public.
Reuse or recycling of radioactive wastes
The Nuclear Safety Commission also indicated in the "Regulatory Policy Report" of LLW,
a possibility of restricted reuse of extremely low-level radioactive wastes (ELLW). Some examples
of ELLW reuse or recycling are the reuse of contaminated concrete chips or blocks as fillers for
land development, and of metal piping as construction supplies or raw materials.
The reuse or recycling of radioactive wastes should be regulated on a case-by-case basis,
subject to estimation of dose to the reusers from residual radioactivity. Therefore, the primary
residual radioactivity criterion for restricted reuse or recycling of ELLW should be a certain dose
level to the reusers.
Dose to be exempted from regulatory control
In its report submitted in 1987, the Radiation Council suggested a criterion for unrestricted
recycling of radioactive wastes, such as metals/ coming from decommissioning the reactors.
With regard to shallow land disposal of LLW, the Council concluded that the criterion for
exemption of LLW site should be set so that the estimated dose is sufficiently low compared with
the dose limit for the public, i.e., an effective dose equivalent of 1 mSv per year (100 mrem). The
criterion should ensure that the public will not be exposed to radiation beyond the dose limit (1
mSv/y), even taking into account the possibility of present and future exposure from other
sources and practices.
The Council adopted an individual dose of 10 micro-Sv (1 mrem) per year as the criterion
for LLW disposal, in concordance with both the International Commission on Radiological
Protection and the International Atomic Energy Agency. The report suggested a similar approach
in establishing the criterion for unrestricted reuse or recycling of
radioactive wastes.
Investigations for establishment of residual radioactivity criteria
An investigation for establishing residual radioactivity criteria for reuse or recycling of
radioactive wastes has been started by the Science and Technology Agency of Japan. Some
scientific investigations for this problem are also being carried out by the Japan Atomic Energy
Research Institute and some other research organizations.
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CONCLUSION AND PRIVATE COMMENTS
As mentioned above, although current Japanese regulations implicitly refer to a residual
radioactivity criterion for reuse, no generic criteria have been established yet. No general
conclusions have been reached regarding desirable criteria characteristics.
The criteria for reuse of buildings or sites, are considered to be certain dose levels to the
reusers. A building or site owner should follow regulatory procedures pertaining to reuse (e.g.,
the report on decommissioning to the responsible Minister), thereby establishing the operational
criteria for residual radioactivity, on a case-by-case basis.
On the other hand, the criteria for reuse or recycling contaminated material or radioactive
wastes are the radioactivity concentration levels. Types of reused materials or radioactive
wastes, and the ways to reuse them, can be easily identified in streams of reuse. Therefore, it
should be practical to set the criteria for each reuse stream, derived from a certain basic criterion,
such as a dose limit.
REFERENCES : >
[1] The law for the regulations of nuclear material, nuclear fuel material and reactors,
(1957).
[2] The law concerning prevention from radiation hazards due to radioisotopes, etc,
(1958). ,
[3] Nuclear Safety Commission: Regulatory policy for decommissioning of research
reactors, (1985).
[4] Nuclear Safety Commission: Regulatory policy for land disposal of low-level radioactive
solid wastes, (1985).
[5] Radiation Council: Dose to be exempted from regulations concerned with shallow land
disposal of radioactive wastes, (1987).
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Establishment of Criteria
for the Unconditional Release
of the Shippingport Atomic Power Station Site
Lynn R. Wallis and Kenneth J. Eger
General Electric Company
ABSTRACT
This paper describes the method used to develop site release criteria for the Shippingport
station. It also provides information on initial site characterization, initial radiological conditions,
and the application of the DOE document "A Manual for Implementing Residual Radioactivity
Guidelines."
Although the direct use of release limits is discussed as an alternate to'basing the limits
on a cost benefit basis, the paper does show that both the direct use and cost benefit analysis
were important in defining the final release criteria.
The development of a procedure which (T) considered the cost of postulated remedial
actions and (2) employed an appropriate number of scenarios to demonstrate compliance with
applicable limits and the ALARA philosophy is also discussed.
INTRODUCTION
The Shippingport decommissioning project was the first to involve a large-scale,
commercial nuclear power plant. Shippingport was also the first nuclear power plant to be
decommissioned which had a long period of power operation.
The Shippingport Atomic Power Station was located 25 miles northwest of Pittsburgh,
Pennsylvania on the south bank of the Ohio River. The reactor was a four-loop, 72 MWe
pressurized water reactor. It was operated by Duquesne Light Company from late 1957 until
October 1982 and produced 7.4 billion kwh of electricity.
The reactor and containment were housed in four underground concrete vaults. Two
boiler chambers each housed two of the coolant loops. Because of the low radiation levels that
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existed in the reactor containment chambers, gross reactor systems decontamination was not
required. One of the unique features of the project was the one-piece removal of the reactor
pressure vessel. (Figure 1)
The Shippingport project involved the dismantling of reactor equipment from the plant and
disassembly of its buildings, as well as removal, transport and disposal of radioactive materials.
(Table 1) Irradiated materials were shipped by barge to DOE's disposal site at its Hanford facility
in Richland, Washington. (Figure 2) The Shippingport decommissioning project has restored the
site to a condition suitable for unrestricted use.
RELEASE CRITERIA: BASIS ,
DOE's decision to decommission Shippingport was made in August 1982. The site was
turned over to GE (the Decommissioning Operations Contractor) in September 1984. At that time
the radiological criteria which had to be met before the site could be returned to its owner,
Duquesne Light Company, had not been defined. The first step toward definition of the release
criteria was made in October 1984 when DOE notified GE that Regulatory Guide 1.86 was to be
used to define allowable surface contamination limits. At the same time, GE was asked to
perform a sensitivity analysis to estimate the costs to reduce future site doses to the public. We
were asked to evaluate 3 levels: 100, 25 and 10
mrem/year.
The second step occurred on Sept. 30, 1985 when DOE issued their working draft of "A
Manual for Implementing Residual Radioactivity Guidelines."
INITIAL SITE SURVEYS
An initial surface survey of the site was performed in June 1985. Measurements were
made using a scintillation detector held 5 cm off of the ground, and surface soil samples were
taken for gamma isotopic analysis. Samples were typically taken where dose rate readings were
highest. Only two of the three isotopes (besides tritium) which had been identified in plant
systems (Cobalt-60, Cesium-137 and Antimony-125) were found in the soil. Table 2 provides a
summary of the concentration of radionuclides found in the soil. Dose rate readings were not
taken where radiation from neighboring structures and equipment raised the general background,
nor were samples taken in paved areas.
Although the initial site survey data was limited, it was sufficient to perform a preliminary
dose assessment the latter part of 1985. A post-decommissioning rough estimate of 22.4
mrem/year to a hypothetical site resident was calculated by summarizing the exposures from the
six applicable pathways defined in DOE's draft "A Manual for Implementing Residual Radioactivity
Guidelines." This analysis not only provided an initial assessment of the expected dose but also
identified "direct exposure" as the predominant exposure pathway. The discovery that 98% of
the calculated annual dose would come from direct exposure suggested that cost effective
remedial action should be directed at the layers of near surface soil.
The second phase of the initial site survey conducted in late 1985 also determined the
degree of on-site concrete contamination. Samples were taken from areas that were known to
be contaminated as well as those areas that were thought to be clean.
198 ' ... .
-------
A hand drill was used to break up concrete over an area of about 100 cm2 to a depth of
not more than 0.6 cm. A total of about 10 grams of concrete was collected for each sample, and
a gamma isotopic analysis was made using an intrinsic germanium detector. Table 3 shows the
results of the concrete sampling program. As expected, concrete from the Fuel Handling
Building canal and from sump areas had the highest concentrations, Cobalt-60 made up more
than 99% of the activity detected.
The final part of the initial survey was a subsurface soil sampling program conducted in
September 1986. Fifty 10 cm holes were drilled, ranging in depth from 0.5 to 4 meters. Soil was
collected in 0.5 meter increments (about 800 grams) and a gamma isotopic analysis performed
on site. Table 4 summarizes the concentrations of residual radioactive materials found in the
subsurface soil.
SCENARIOS
In early 1986 a program was started to define final site release criteria. At the same time
a program was also launched to prepare a cost benefit analysis. In the course of this work, four
scenarios were evaluated which covered potential future uses of the site.
(1) Residential:
(2) Occupancy:
(3) Souvenir:
(4) Exposed Slab:
A home is built on the site (including excavation of a basement down to
elevation "-3 m"). The family that resides there has a garden, and
livestock, and uses water from an on-site well. Particulars about this
scenario are defined in detail in DOE's "A Manual for Implementing
Residual Radioactivity Guidelines."
A concrete substructure which is open at the top somewhere within the
top 3 meters of the site surface has been excavated and turned into an
office area. The use is based on an occupancy of 40 hours/week and 50
weeks per year. The minimum size considered to be occupiable is a room
having a floor area of at least 10 m2 and a minimum dimension of 3 m.
A piece of concrete weighing less than 50 kg and located within three
meters of the surface is excavated and taken as a souvenir. The exposed
person was assumed to live 2 m from the souvenir all year.
A monolithic block of concrete not amenable to occupancy, but existing
within 3 m of the surface is exposed by excavation. The minimum area
considered would be 2 m2 and minimum dimension 1 m. An individual sits
on it for 168 hours while it is exposed.
For completeness, the subsurface concrete not covered by the latter three scenarios is
included with the adjacent soil. Thus any contribution to the annual dose from this material is
considered via the residential scenario.
199
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COST BENEFIT ANALYSIS
An annual dose was computed for a hypothetical family (residential scenario) establishing
a farm on the most contaminated location of the Shippingport site. Figure 3 is a diagram of the
farm site showing a dose from residual radioactivity of 3.5 mrem/year.
The cost benefit analysis showed that remedial action that could not be justified on a strict
cost benefit ratio could still be undertaken as a matter of prudence because the overall cost was
low.
A dose assessment was also made for a worst case "occupancy scenario." It was
assumed that the Fuel Handling Building Canal was excavated and converted into an office
space. In this case an annual dose as high as 3.82 rem could be received by the office workers.
This analysis showed reduction of the canal dose derived in the worst case scenario
(occupancy by office worker) to negligible levels would cost 13.2 man-rem and $898,000. The
analysis showed that such high costs should not be expended because the expected benefit
would be small.
SITE RELEASE CRITERIA
GE's assessment of the 4 worst case scenarios and cost benefit analysis were then
reviewed by DOE. Based on this data, DOE issued its Shippingport site release criteria on
January 6,1987. It was defined as "...100 mrem/year total committed effective dose equivalent
to the maximum exposed individual of the general public under the worst case scenario..." In
addition GE was directed to apply the philosophy of ALARA to the release of the site.
Adherence to the acceptable surface contamination levels defined in Table 1 of Regulatory
Guide 1.86 was established as a goal for the project. Provision was made for DOE concurrence
for release at higher levels (but not in excess of the limits), if achievement of the goal should be
too costly. Figure 4 gives a time line showing all of the major events leading up to establishment
of these criteria.
LIMITING CONDITIONS FOR RELEASE
Radiological conditions which could exist on the site without causing the release criteria
to be exceeded were derived based on the four referenced scenarios (Figure 5).
• Top 3 meters of soil: The average concentration must be less than 6 pCi/gram for Co-60.
Vertical averaging is limited to 0.15 meters, and horizontal averaging to 100 m2
• Other soil: The average concentration must be less than 100 pCi/gram Co-60. Horizontal
averaging is limited to 100 m2, vertical averaging is limited to 3 m.
• Occupiable concrete substructure: The average exposure three feet from any wall in the
"room" must be less than .05 mrem/hour.
200
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• Souvenir: The average dose rate from a souvenir must not exceed 0.01 mrem/hours at
2m.
• Exposed slab: The average dose rate at contact (5 cm) must not exceed 0.6 mrem/hour.
Two definitions supplemented these limiting radiological conditions. One placed a restriction on
the magnitude and extent of hot spots allowed. The other specified the modifications required
when confronted with a mixture of isotopes.
• Hot Spot Criteria: Allowable hot spots can have concentrations up to ten times the
average - provided that the area bounding the hot spot does not exceed the value
determined using the following relationship:
Maximum Area of Hot Spot = 100 m2 (Average Concentration] 2
(Hot Spot Concentration) •
• Mixtures of isotopes: Limiting concentrations (residential scenario) must be reduced by
4% for each pCi/gram of cesium-137 in the top 3 m of soil and by 2% for each pCi/gram
of antimony-125 when cobalt is not the only isotope present. For deeper soil, the
reduction would be 1% per pCi/gram of cesium-137 and 0.4 per pCi/gram of
antimony-125.
APPLICATION OF ALARA
Reduction of radiation doses to values "As Low As Reasonably Achievable (ALARA)" at
Shippingport was required as part of the release criteria. The usual cost ratio of $2000 per man-
rem avoided was gsed to determine whether dose reduction was justified. Actions with cost
ratios in excess of this rate were not taken, while actions with lower cost ratios were pursued.
Table 5 summarizes the application of ALARA in the residential scenario. The normal
decommissioning effort independent of efforts to apply ALARA would result in a dose of 8.8
mrem/year to a future resident. Installation of a drainage cap already part of the
decommissioning program reduces the dose to 3.5 mrem/year.
One other decommissioning activity helped reduce the dose from 3.5 to 2 mrem/year.
The top 3 feet of soil in areas formally occupied by buildings was removed.
ALARA in this, scenario was met with activities already planned as part of the
decommissioning program.
The probability that the Occupancy, Souvenir or Exposed Slab scenarios could become
real exposure pathways is low for two reasons. The chance of the required sequence of events
occurring is remote, and the real occupancies would be much less than the postulated ones.
Therefore, reducing the dose to values less than the limiting ones was not cost beneficial. This
is consistent with direction received from DOE which states that action to reduce the expected
dose rate from the canal to values less than 100 mrem/year "would result in increased costs and
worker exposure... disproportionate to the benefit of further decontamination." GE believes for
these cases that the reduction in dose to a value of 100 mrem/year meets ALARA requirements.
201
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APPLICATION OF TABLE 1 TO REGULATORY GUIDE 1.86
Table 1 of Regulatory Guide "Acceptable Surface Contamination Levels" was established
as a goal for reducing residual contamination, in this case, on surfaces of concrete left below
grade.
In the residential scenario all the subsurface material is considered to be like soil so
surfaces have no meaning, and the Table is not applicable.
In contrast, the Table 2 regulatory guide values can be considered to apply to surfaces
of underground concrete when it is exposed per the occupancy scenario. There is no direct
correlation, however, since the contamination exists not on the surface of the concrete, but mixed
in the near surface concrete matrix. This anomaly can be resolved by using a reference
measurement. A near surface (0.6 cm) concentration of 100 pCi/gram Co-60 gives a contact
reading of approximately 100 ncpm. This corresponds to the reading expected from a smooth
surface at 5000 dpm/100 cm2 (the Table 1 limit). The 100 pCi/g concrete also generates an area
dose rate of approximately 0.05 mrem/hour (.01 mrem/hour for each of the walls and the floor).
Since this second value is the limiting dose rate for the occupancy scenario, achievement of the
100 mrem/year value also demonstrates equivalent achievement of the Table 1 goal.
Limitation on averaging in Table 1 to areas of 100 cm2 and 1 m2 was not applied since
the exposure received according to the occupancy scenario is from the general habitation of the
room. Compliance was demonstrated instead by requiring the average dose rate in the
occupiable portions of the room to be less than 0.05 mrem/hour. This reduced the effort required
to prove that the criteria was met without reducing the assurance that the 100 mrem/year dose
rate limit would not be exceeded.
The souvenir scenario and the exposed slab scenario were treated differently. Each
represents a case where the probability is small that the actual use will correspond to the
identified one. In addition, it is unlikely that the occupancy assumed in the two scenarios would
be as extensive as that modeled. Therefore, the expected dose rate will be much less than the
limiting one (100 mrem/year) and it is "unlikely to result in an unreasonable risk to the health and
safely of the public."
Therefore cleaning miscellaneous pieces of rubble and nonoccupiable trenches beyond
the specified limits (100 mrem/year by the applicable scenario) would not be cost beneficial.
SUMMARY
Site release criteria was generated for the Shippingport Station Decommissioning Project.
An implementation plan translated the criteria into limiting conditions for the site. This
plan demonstrated that ALARA criteria were met, and that the Table 1 limits in Regulatory Guide
1.86 were reached.
202
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The annual dose to a member of the public via the most restrictive probable scenario (the
residential scenario) after decommissioning would be less than 5 mrem. In addition, extra costs
above and beyond the applied decommissioning procedures to achieve this low dose were less
than $30,000. The Shippingport experience should be considered by other nuclear facilities as
an incentive to keep operations under strict control. When unnecessary spread of contamination
can be avoided, the costs of decommissioning (both financial and radiation exposure) can be
minimized.
203
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TABLE 1
SHIPPINGPORT STATION DECOMMISSIONING PROJECT
KEY STATISTICS
Reactor Vessel Package - 900 Tons
Radwaste Volume - 3000 Cu Yds
Radioactive Contents -14,500 Curies
Vessels/Tanks -130
Chamber Steel - 22,400 Tons
Contaminated Concrete - 50 Cu. Yds
Non-Contaminated Rubble -15,000 Cu. Yds
Contaminated Pipe - 56,000 LF
Non-Contaminated Pipe - 55,000 LF
Asbestos Waste - 500 Cu. Yds
Isotope
Co-60
Cs-137
Sb-125
TABLE 2
SUMMARY OF SURFACE SURVEY RESULTS
Frequency
Of Detection
(% of Samples)
64.3
85.4
0
Distribution of Detectable
Activity (pCi/q) *
Maximum
68**
5.3
3.5
Median Mean
0.69 1.04
0.39 0.52
* Only listed for those samples having statistically detectable activity
** One sample (68 pCi/g) contained a resin bead so the second highest concentration is also
included. The initial value was not included in the average.
204
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TABLE 3
SUMMARY OF CONCRETE SURFACE SURVEY RESULTS
Area
B/D Boiler Chamber
A/C Boiler Chamber
Fuel Handling Building
Canal
# of Samples
15
15
39
Surface Activity
(pCi/g)
Range Median
1.3 to 5.9
8.0 to 591
2.2
4.1
88 to 35,500 6,820
TABLE 4
SUMMARY OF SUB-SURFACE SURVEY RESULTS
Frequency
isotope Of Detection
Distribution of Detectable
Activity (oCi/a)*
(%of Samples)
All
Samples**
All
Holes
47
44
1
Maximum
94
100
4
Median
4.00
1.58
3.06
Mean
0.220.32
0.210.25
1.961.85
Co-60
Cs-137
Sb-125
* Only listed for those samples having statistically detectable activity.
** Usually there were 16 samples per hole.
205
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TABLES
APPLICATION OF ALARA USING THE RESIDENTIAL SCENARIO
Category
Limit
Site (As Is)
Site (Per Plan)
Site With
Soil Removal
Annual Dose*
100mrem/year
8.8 mrem/year
3.5 mrem/year
2.0 mrem/year
* 1990mrem
** LTRA - Lower Than Reasonably Achievable
ALARA
N/A
N/A
Yes
LTRA
**
206
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FIGURE 1
Lifting Beam
and Skirt
\
to
o
-J
NST
V J
33.2'
h-io.5'*H
PWR
Shlpplngport
72MWe
PWR
Shipplngport
with
Neutron Shield Tank,
Lifting Beam and Skirt
.3'
PWR
San Onofre 2/3
1100 MWe
-------
FIGURE 2
SHIPPINGPORT STATION DECOMMISSIONING PROJECT
RPVI/NST PACKAGE AND OTHER COMPONENTS ON BARGE
O
00
, STEAM GENERATORS (2)
. RCS PUMPS (2)
BLOWOFFTANK
.AREA AVAILABLE FOR
300 TONS LSA WASTE
PRESSURIZER
VALVES & PIPING
21 BOXES
RPVI/NST PACKAGE W/SHIPPING CRADLE
AREA AVAILABLE FOR 300
TONS LSA WASTE
4000 TON NOMINAL CAPACITY
OCEAN BARGE _
DEMINERALIZER,
VALVES & PIPING
16 BOXES
HEAT EXCHANGERS
2 BOXES
FLASH TANK
RCS PUMPS (2)
STEAM
GENERATORS (2)'
-------
FIGURE 3
FARM-SITE SHOWING DECOMMISSIONINED
BUT UNREMEDIATED CONFIGURATION AND THE
RESULTANT D'OSE VIA THE RESIDENTIAL SCENARIO
Soil Concentration
(pCi/g)
Co-6D Cs-137
100* or 100*
Dose
mrem/year
0.11
.085
1.48
— — —
1.0.
.29
.094
.060
.063
.96
.16
. ' .081
.047
.055
1.11 s
.410
.002
.0003
.0004
Resident's Basement
Excavated Soil
Drainage
Soil Section
Removed To
Show Concrete
*Limiting Concentration Assumed .
Total Dose 3.5 mrem/year
Surface Soil
(5cm)
Subsurface
Layers
.463
209
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FIGURE 4
ACTIVITIES LEADING TO THE
IMPLEMENTATION OF SITE RELEASE CRITERIA FOR THE
SHIPPINGPORT ATOMIC POWER STATION
DOC Agreement
To Prepare
Cost-Benefit
Analysis
DOE Request
For Sensitivity
Analysis —>
Issue of
DOE Manual
(Working Draft)
(Review Draft)*—)
First
Assessment Of
Dose To Future
Resident
Release
Criteria
Recommended
By DOC
Decision To
Accelerate
The Project-)
Cost Benefit
Analysis
Issued by DOC
Release Criteria
Issued by DOE-}
6
7
8
9
t
10
11
r
12
1984
~f
7/7/fr
Z^^L^.
Surface Survey^
(Intermittent)
T f
yy//
^-^A
Concrete
(Widely In
X
A
X
\\
Y
X"
v
, \
X
X
\'
t
\l
\V
ft
\
\S
V
Survey •— ' V V
termittent) \
Development
12345678
1985
Compu
9 10 11
12
ter
1
of
Subsurface Soil)
) Survey
Program
2
3
/)
5
6
7
8
9'
10
11
12
1986
*^
\
^
.
•—-.^
.Implementation
Plan Submitted
To DOE
For
Concurrence — -*^
1 2 3
4 ,!
i 6
1987
-------
FIGURE 5
LIMITING CONDITIONS
PROPOSED FOR THE RELEASE OF THE
SHIPPINGPORT STATION DECOMMISSIONING PROJECT SITE
<100 mrem/Year
Top 3 Meters
Below
3 Meters
Monolithic
Concrete
[6]
[100]
60Co Concentration (pCi/g) = [ 3
211
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NRC Residual Contamination Criteria
Timothy C. Johnson
Division of Low-Level Waste Management and Decommissioning
U.S. Nuclear Regulatory Commission
ABSTRACT
The Nuclear Regulatory Commission (NRC) is currently addressing the issue of residual
contamination criteria. A Commission Policy Statement on exemptions from regulatory control is
expected in the very near future. Based on this policy NRC staff will develop interim guidance on
acceptable levels of residual contamination early in 1990.
CURRENT CRITERIA FOR REACTORS
The current guidance for terminating nuclear reactor licensees is found in Regulatory Guide
1.86, Termination of Operating Licenses for Nuclear Reactors" [1]. This regulatory guide provides
methods and procedures considered acceptable by the NRC staff for reactor license termination.
It also contains decontamination guidance for release for unrestricted use. This regulatory guide
applies to research, test, and power reactor license terminations.
Prior to terminating a license and releasing the site for unrestricted use
Regulatory Guide 1.86 recommends that the licensee should --
a. Make a reasonable effort to eliminate residual contamination.
b. Not apply coverings to radioactive surfaces of equipment and structures until the
contaminated levels are below those in Table I of Regulatory Guide 1.86.
c. Determine the radioactivity in the interior surfaces of pipes, drain lines, or ductwork by
making measurements at all traps and other appropriate access points if the contamination
can be shown to be representative of the actual contamination. Inaccessible points on
structures, equipment, or scrap should be considered contaminated to levels in excess of
, the limits for unrestricted release.
d. Make a comprehensive radiation survey.
212
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The NRC staff, and previously the Atomic Energy Commission staff, have used the surface
contamination limits given in Table I of Regulatory Guide 1.86 for over twenty years. Table I of
Regulatory Guide 1.86 is presented below. These limits were developed based on the variations
in natural background and the lower limits of detection of survey instruments. These criteria are
not based on a dose objective that relates nuclide concentration levels to exposures to the public.
This situation has led to the need for criteria based primarily on individual dose that is directly
related to public health and safety.
In addition to the structure and equipment surface contamination levels in Regulatory Guide
1.86, the NRC staff has applied a limit for gamma-emitting nuclides of 5 uR/hr above background
as measured at 1 meter from the surface applicable to radioactive material other than surface
contamination [3]. Assuming a reasonably conservative occupancy time of 2,000 hours/year, the
maximum exposure to an individual would be 10 mRem/yr. '
TABLE I
ACCEPTABLE SURFACE CONTAMINATION LEVELS
NUCLIDE3
U-nat, 0-235, U-238 and
associated decay products
Transuranics, Ra-226
Ra-228, Th-230, Ih-228,
Pa-231, Ac-227, 1-125, 1-129
AVERAGE130
5000 dpm a/100 cm2
100 dpm/100 cm2
Th-nat, Th-232, Sr-90, Ra-223, 1000 dpm/100 cm2
Ra-224, U-232, 1-126, 1-131. 1-133
Beta-gamma emitters
(nuclides with decay modes
other than alpha emission or
spontaneous fission) except
Sr-90 and others as noted
5000 dpm Br/100 cm2
MAXIMUM"
15,000 dpm or/100 cm2
300 dpm/ 100 cm2
3000 dpm/100 cm2
15,000, dpm Br/100 cm2
REMOVABLE138
100Q dpm a/100 cm2
20 dpm/100 cm2
200 dpm/100 cm2
1000 dpm Br/100 cm2
• Where surface contamination by both alpha- and beta-gamma emitting nuclides exists, the limits established
for alpha- and beta-gamma emitting nuclides should apply independently.
b As used in this table dpm (disintegrations per minute) means the rate of emission by radioactive material as
determined by correcting the counts per minute observed by an appropriate detector for background, efficiency
and geometric factors associated with, the instrumentation. .,
0 Measurements of average contaminant should not be averaged over more than 1 square meter. For objects of less
surface area, the average should be derived for each such object.
d The maximum contamination level applies to an area of not more than 100 cm2.
e The amount of removable radioactive material per 100 cm2 of surface area should be determined by wiping that
area with dry filter or soft absorbent paper, applying moderate pressure and assessing the amount of
radioactive material on the wipe with an appropriate instrument of known efficiency. When removable
contamination on objects of less surface area is determined, the pertinent 'levels should be reduced
proportionally and the entire surface should be wiped.
Regulatory Guide 1.86 is currently undergoing revision to make it compatible with the
license termination procedures in the new decommissioning rule published in the Federal Register
on June 27,1988 [2]. The revised regulatory guide will reference updated residual contamination
criteria. It is expected to be issued for comment after the interim residual contamination limits are
issued early in 1990.
213
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NEW CRITERIA DEVELOPMENT •»
The NfiC is currently a member of an interagency task force to develop standards for residual
criteria. However, this group is not expected to complete its work until the mid-1990's. In the
interim the NRC staff recognizes the need1 for guidance in this area.. The Commission has
committed to providing this guidance by December 1989. This guidance is directly related to'NRC
activities in the areas of Below Regulatory'Concern (BRC) waste disposal practices and other
regulatory decisions involving the exemptions of radioactive material from regulatory control. The
following discussion presents important background information that will lead to the development
of residual contamination criteria.
Section 10 of the Low-Level Radioactive Waste Policy Amendments Act of 1985 [4] required
that, within six months, the NRC establish standards and procedures, and the technical capability
to act in an expedited manner on petitions to exempt specific waste streams from regulation. NRC
responded with three actions.
First, On August 29, 1986 the NRG published in the Federal Register [5] a Commission
Policy Statement and Staff Implementation Plan. These two documents provide guidance to
potential rulemaking petitioners outlining the decision criteria the Commission intends to use to
expeditiously process BRC waste stream petitions.
Second, the IMPACTS-BRC computer code for calculating radiological impacts from
unregulated disposal was adapted for personal computer use and a draft user guide was
published in July 1986 (Volume 2 of NUREG/CR-3585) [6]. Subsequently, the NRC staff contracted
with Sandia National Laboratory for technical assistance to critique, validate, and verify the
computer code.
Third, on December 2,1986 the NRC published in the Federal Register an advanced notice
of proposed rulemaking (ANPR) [7] requesting comments on the development of a generic BRC
level for wastes. Over 90 comments were received in response to the ANPR reflecting diverse
views on how the NRC should proceed. Many commenters opposed the concept of any level of
radioactivity being BRC and others urged NRC to proceed promptly on the generic rulemaking.
In March 1988 the Commission delayed the rulemaking and directed the staff to first prepare for
Commission consideration options for a broad policy statement that establishes a generic limit for
exposures that are below regulatory concern.
The policy statement addresses exemption decisions as a whole, not only those BRC
issues for waste management, but also licensing applications for consumer products, existing
exempt quantity limits, decommissioning, and effluent releases. This policy statement would
provide for more efficient and consistent regulatory actions in connection with exemptions from
specific NRC requirements. A draft policy statement was prepared and discussed at the
International Workshop on Rules for Exemption from Regulatory Control sponsored by the NRC
and the Nuclear Energy Agency in October 1988. An advance notice of a policy statement was
issued for public comment in the Federal Register on December 12,1988 [8]. The Advance Notice
recommended a 10 mRem/yr individual dose criterion as one basis for establishing a floor for
curtailment of ALARA activities. This value considered optimum use of Commission resources to
address matters of radiological protection, the variations in background, risk perceptions, BRC
versus de minimis distinctions, the linear non-threshold hypothesis, and practical implementation.
, ,, 214
-------
The policy and the comments received are currently being considered by the Commission. We
expect the Commission to take action in the very near future.
Based on the Commission action and dose objectives that are set, the NRC staff will
prepare interim criteria on residual contamination early in 1990. Included will be guidance on
residual contamination levels on a nuclide-by-nuclide basis. These nuclide-by-nuclide data will be
developed from pathway analyses performed under a contract with Pacific National Laboratories.
These analyses are based on direct exposure, ingestion, inhalation, and groundwater pathways.
REFERENCES
[1] U.S. Nuclear Regulatory Commission Regulatory Guide 1.86, Termination of Operating
Licenses for Nuclear Reactors, June 1974.
[2] General Requirements for Decommissioning Nuclear Facilities, Final Rule, Federal Register,
Vol. 53, No. 123, pp. 24018-24056, June 27, 1988.
[3] Letter to Dr. Roland A Finston, Stanford University, from John A. Stolz, NRC, March 17,
1981.
[4] Low-Level Radioactive Waste Policy Amendments Act of 1985, Public Law 99-240, January
15,1986.
[5] Radioactive Waste Below Regulatory Concern; Policy Statement, Federal Register, Vol. 51,
No. 168, pp. 30839-30847, August 29,1986.
[6] ; Forstom, d.M., Goode, D.J, De Minimis Waste Impacts Analysis Methodology,
NUREG/CR-3595, Volume 2,-July 1986.
[7] Radioactive Waste Below Regulatory Concern; Generic Rulemaking, Federal Register, Vol.
51, No. 231, pp. 43367 - 43369, December 2, 1986.
[8] Policy Statement of Exemptions from Regulatory Control, Federal Register, Vol. 53, No. 238,
pp. 49886 - 49891, December 12,1988.
215
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Status and Implementation of
the NRC Policy on
Exemptions from Regulatory Control
Donald A. Cool, Ph.D., Chief
Radiation Protection and Health Effects Branch
Office of Nuclear Regulatory Research
U.S. Nuclear Regulatory Commission
ABSTRACT
The U.S. Nuclear Regulatory Commission (NRC) is currently considering a broad policy
on exemptions from regulatory control. An advance notice of policy development was published
for public comment in December 1988, and an international workshop and public meeting were
held in October 1988 and January 1989 respectively. The policy will establish the framework
within which the NRC will consider specific regulations and licensing actions that exempt certain
practices from all or part of the normal system of regulatory control. Included within the policy
will be the specification of numerical individual dose criteria that define levels below which the
Commission believes thatfurther compliance with the ALARA principle is unwarranted. The policy
will have broad applicability in areas such as waste disposal, decommissioning, and consumer
products.
The NRC also is developing, as part of the initial implementation of the policy, interim
guidance on residual contamination criteria for soils and structures that correspond to the
individual dose criterion. This guidance will probably take two different forms. First, values of
dose per unit concentration for exposure pathways such as inhalation, secondary ingestion, and
direct radiation will be provided for various radionuclides to facilitate site-specific analyses.
Second, a generic release for unrestricted use scenario will be evaluated and the concentration
of each radionuclide corresponding to the individual dose criterion tabulated.
Beginning in early 1988, the U.S. Nuclear Regulatory Commission (NRC) began
development of a broad policy on Exemptions From Regulatory Control. This effort was in
addition to the actions already underway to implement the Low-Level Radioactive Waste Policy
Amendment Act of 1985 which directed the NRC to develop procedures for determining the types
and quantities of radioactive wastes that could be considered to be below regulatory concern.
The thrust of the broad policy, however, is to establish a framework within which the NRC can
216
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consider all types of exemption, including consumer products, decontamination and
decommissioning for unrestricted use, and potentially the recycle of radioactive materials.
The purpose of a broad policy statement is to establish the basis upon which the
Commission may initiate the development of appropriate regulations or make licensing decisions
to exempt certain practices from some or all regulatory controls. While we envision this policy
to be directed principally toward rulemaking activities, it may also be applied to license
amendments or license applications involving the release of licensed radioactive material either
to the environment or to persons who would be exempt from the NRC's regulations.
In December 1988, an Advance Notice of Proposed Policy Development was published
in the Federal Register which contained a proposed policy statement and solicited comment on
the approach being considered for exemptions. The NRG received over 200 comment letters
from members of the public, public interest papers, the industry, and various state, federal, and
local governmental organizations. The NRC also solicited comments during a public meeting in
January 1989 and the international perspective on exemptions during an international workshop
on exemptions held in October 1988. As a result of these comments and the information
gathered during the meetings, a revised policy statement was prepared by the NRC staff and
submitted for Commission consideration in June 1989. The Commission is presently considering
the staff proposal.
The proposed policy would establish numerical criteria for individual and collective dose
that would define a region in which exemption of a practice from regulatory controls should be
a rather straightforward undertaking. However, a practice resulting in individual or collective
doses in excess of the numerical criteria would not automatically be excluded from consideration
for exemption. Instead, further analysis would be required to determine if an exemption was the
appropriate regulatory approach in that particular situation.
The structure of the proposed policy statement was very similar to the recommendations
of the International Atomic Energy Agency (IAEA) in their Safety Series No. 89. However,
although fundamental principles of radiation protection were considered in recommending the
individual and collection dose criteria, the specific values selected represented a policy judgment
based on risk and resource allocation considerations. As a result, the criteria do not exactly
agree with the bases for, or magnitudes of, similar criteria selected or under consideration
nationally by the U.S. Environmental Protection Agency (EPA), by other countries, or by
international agencies such as the IAEA.
It is important to note that, in this policy, the NRC does not assert an absence or
threshold of risk at low radiation dose levels but rather establishes a baseline where further
government regulation to reduce risks is unwarranted. The presence of natural background
radiation and variations in the level of this background are used to provide a perspective on
which to judge.the relative significance of the-radiological risks involved in the exemption
decisionmaking process.
A major consideration in exempting any justified practice from some or all regulatory
controls hinges on the general question of whether or not application or continuation of
regulatory controls is necessary and cost-effective in reducing dose. To determine if exemption
is appropriate, the Commission must determine if one of the following conditions is met:
217
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1. The application or continuation of regulatory controls on the practices does not
result in any significant reduction in dose received by individuals within critical
groups and by the exposed population; or
2. The costs of the regulatory controls that could be imposed for dose reduction are
not balanced by the commensurate reduction in risk that could be realized.
The basic numerical criteria contained in the proposed policy statement are illustrated by
Figure 1. The individual dose criterion was established at 10 mrem (0.1 mSv) and a collective
dose criterion set at 500 person-rem (5 person-Sv). Each of these criteria applies to a practice
which is defined in such away so as to avoid fractionation or deliberate dilution of materials that
should otherwise be controlled. If the individual and collective doses from a practice were to fall
within the criteria, then, under the proposed policy the practice would be considered as meeting
the basic conditions for exemption.
One facet of the proposed policy which has not been finalized is the way in which the
potential for exposures to multiple exempted practices is to be handled. The Commission is
considering several mechanisms to deal with this issue, including a case-by-case evaluation of
other existing exemptions to assure that exposure to several exempted practices will not result
in doses which are a significant fraction of the dose limits for members of the general public. A
second possibility is to reduce the individual dose criterion if the practice under consideration
has the potential for being widespread within the population, such as for consumer products.
The exemption policy does not constitute regulatory permission to transfer radioactive
materials to an uncontrolled status. The policy only describes the criteria and framework under
which the NRC will take specific rulemaking or licensing actions. The NRG staff anticipates that
the typical approach will be for an outside group, such as the nuclear industry, to petition the
NRC for an exemption action. In fact, the Nuclear Utility Management and Advisory Council
(NUMARC) is planning to submit such a petition in the near future to consider certain waste
streams from commercial nuclear power plants as below regulatory concern. The staff will
consider such petitions, and, if warranted by the technical evidence, publish a proposed rule for
public comment. NRC action on any final rule would then be based upon the proposed rule and
the public comments received.
One area in which the NRC recognizes the need for additional information to translate the
proposed exemption policy into actual regulatory decisions is the area of residual contamination
criteria. The NRC staff, with contractual support from PNL, is currently developing interim criteria
for soils and structures. The NRC staff envisions that the criteria will be presented in two formats
to facilitate use by NRC, licensees, and other groups. First, the various pathways that could
result in exposure, including direct radiation, inhalation, secondary ingestion, food pathways, and
groundwater will be examined and the dose per unit concentration for soils or structures
tabulated for a variety of radionuclides. We believe that this type of information is essential to
a site-specific analysis of potential doses resulting from a decontamination or decommissioning
action.
The NRC staff also envisions the need for generic values for each radionuclide that would
correspond to a dose equal to the individual dose criterion of the exemption policy. These
generic values could be replacements for the guidance currently available in Regulatory Guide
218
-------
1.86 and the NMSS Branch Technical Position on Disposal Onsite Storage of Uranium and
Thorium in Soils. Therefore, the staff expects to develop a combination of exposure pathways
into a generic "release for unrestricted use" scenario and to develop values for each radionuclide.
A licensee could then simply demonstrate that the residual radioactivity levels on a site are less
than the corresponding guidance values and thus show that the dose would be less than the
exemption dose criterion. The NRC actions could then focus upon verification of residual
radioactivity levels, rather than upon the appropriateness of the pathways analysis.
In summary, the NRC is currently developing both a broad policy on exemptions from
regulatory control and interim guidance on the appropriate levels of residual radioactivity for
decontamination. Although the NRC is not specifically considering the recycle of materials at this
time, decisions on recycle most likely would also be covered within the criteria of the exemption
policy.
PROPOSED EXEMPTION POLICY
FOR A JUSTIFIED PRACTICE
IUUU
-100
E
£
E
«_~, *r\
*~ 10
0)
(0
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Q
— 1
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= i i i i
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:
- "
~
—
i
n
-
_
5
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_-
_^
1 III
1 1 1 1 1 | Illlllll t Illlllll 1 Illlllll 1 Itlll
Not Exemptable
Public Dose Limit- —
*"™^
I1
. .
i Possibly Exemptable
i
Exemptable j
i
i
1 1 1 1 1 i IIIIIIM i 1 1 1 1 1 1 1 * i iiiiiiii i 1 1 1 1
'-
-
—=
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— =
-
^^ m
z
.
II
10 100 1,000 10,000
Collective Dose (Person-rem)
100,000
219
-------
Surface Contamination Criteria
for Free Release
Steven R. Adams
US Ecology, Inc.
ABSTRACT
Recycling of materials, equipment, and facilities during decontamination,
decommissioning, or remedial action projects requires criteria for surface radioactivity guides.
A consensus standard developed by the Health Physics Society on permissible limits of residual
surface radioactivity for unrestricted release has gone through an evolutionary process during
the last 18 years.
Brief history of surface contamination limits from 1974 through 1989 is presented.
Comments on the practical application, limits, measurement methods, and hazardous analysis
relating to the limits are reviewed.
The process to develop a consensus standard in the United States on permissible limits
of residual surface radioactivity on materials, equipment, and facilities to be released for
uncontrolled use began in September 1971 with the formation of a subcommittee of the Health
Physics Society Standards Committee (HPSSC), The standard has thus gone through an
evolutionary process of over 18 years, including the usual subcommittee deliberations, meetings
of the working group with HPSSC, with regulatory agencies, and with industry representatives,
reviews of the literature, and consideration and responses to comments and criticisms in several
ballots (Shapiro, 1980). In 1974 the subcommittee completed a proposed draft as ANSI N328
(1974), and it was formally sent to American Nation Standards Institute (ANSI) Committee N13.
ANSI published the standard as a draft for comment ANSI N13.12. However, N328 was adopted
by the NRC in the form of Regulatory Guide 1.86. Subsequently the Health Physics Society
Standards Committee working group drafted revisions of ANSI N13.12. Proposed ANSI N13.12
has never been published as an ASNI Standard. In 1983 DOE published ORO31 as an
adaptation of the proposed N328. DOE later dropped the N328 version in favor of the Regulatory
Guide 1.86 version and issued a final adoption of 1.86 jn July 1985.
220
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PROPOSED ANSI N328 (1974),
The limits selected by the N328 working group were indirectly tied to MPC values but it
was made clear that these Ijmits were by a consensus based on what appeared to be safe and
practical with the existing technology. The limit for Sr-90 was arbitrarily taken as 1000 dis/min
per 100 cm2 because it was approximately the value of background Sr-90 contamination
produced by fallout from past above ground nuclear weapon tests. In addition, the Committee
decided to use 5000 dis/min per 100 cm2 as an upper limit for group 2 radionuclides, since
higher values could lead to unnecessary high direct radiation exposure. Nuclides with maximum
beta energies less than 150 keV were not considered. An abbreviation of the table of limits from
N328 is shown in Table 1.
REGULATORY GUIDE 1.86 (1974) AND NRG (1982)
An abbreviated version of the table of limit values from Regulatory Guide 1.86 (1974) and
NRC (1982) is shown in Table 2. The only significant difference between the 1.86 (1974) table
and the NRC (1982) table is that the NRC (1982) table has a footnote which states: 'The average
and maximum radiation levels associated with surface contamination resulting from beta-gamma
emitters should not exceed 0.2 mrad/h at 1 cm and 1.0 mrad/h at 1 cm respectively, measured
through 7 milligrams per square centimeter of total absorber."
DOE GUIDELINE (JULY 1985)
The DOE (1985) table of limits is identical to the Regulatory Guide 1.86 (1974) table. It
includes the NRC (1982) footnote on average and maximum dose rates of 0.2 and 1.0 mrad/h
at 1 cm. The ORO 831 (1983) guidelines were essentially identical to N328 (1974) except that
a footnote was added which stated: "In the event of the occurrence of mixtures of radionuclides,
the fraction contributed by each constituent to its own limit shall be determined, and the sum of
the fraction shall be less than 1."
DRAFT ANSI N13.12
An abbreviation of the table of limits in ANSI N13.12 (1978) is shown in Table 3. ANSI
N13.12 (1978) is different from N328 (1974) in that "nondetectable" was used for Group 1, and
"nondetectable By" and 2000 were used for Group 2. An abbreviation of the table of limits in
proposed ANSI N13.12 (1981) is shown in Table 4. The limits in proposed ANSI N13.12 (1981)
essentially returned to those of N28 (1974) except that the total value for Group 2 was changed
from 1000 dpm/100 cm2 to 5000 dpm/100 cm2 and the iodines were moved to Group 2.' The
1983 revision of proposed N13.12 was identical to proposed N13.12 (1981) except that it
excluded depleted uranium, U-228 and Th-232 from Group 1. Table 5 is an abbreviation of the
table of limits in proposed ANSI N13.12 (1983).
The most recent version of the proposed standard ANSI N13.12 (1988) is shown in Table
6. The most significant modification from the 1983 version is addition of a Group 4 that includes
natural, depleted, and enriched uranium to less than 10% U-235, Th-232, and their decay
products. The removable limit is 200 dpm/100 cm2 of gross alpha disintegration and the fixed
plus removable limit is 1000 dpm/100 cm2 of gross alpha disintegration.
221
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PRACTICAL APPLICATION
The standard presents requirements and suggestions for measuring and sampling to
obtain a complete assessment of the surface being examined. Both the direct and indirect (wipe)
monitoring methods are required to be used. Acceptable practices for the purposes of the
standard are described for direct alpha and beta activity monitoring and indirect surface
monitoring.
The direct alpha activity monitoring requires the distance from the detector window to the
surface being monitored to not exceed 0.5 cm, the total absorber of the detector window and the
interviewing medium shall not exceed 2 mg/cm2 and the scanning speed must be slow enough
to ensure a detection frequency of 50% or more at the guide level. Table 7, taken from Table
B-1 of ANS11988, lists the frequency or probability of observing "n" or more counts from a point
source of 300 d/minute for several efficiencies and counting times. For direct monitoring of betas
the detector window and intervening medium must transmit at least 50% of the incident beta
particles, the detector window must be no more than 1.5 cm from the surface, and the scanning
speed must be slow enough to ensure a source detection probability of 50% or more.
For indirect (wipe) monitoring method the instruments must have sufficient counting
efficiency so that the average limiting alpha or beta disintegration rates could be determined with
an accuracy of better than ± 50% at the 90% confidence level in a counting time of less than 5
minutes. Table 8 provides the frequency or probability for observation of "n" or more counts from
a wipe source of 20 dis/minute.
The standard presents examples of instruments and monitoring methods that will ensure
compliance with surface contamination guidelines.
LIMITS
The standard was limited in that the following subjects were beyond its scope:
• Control of radioactivity that is dispersed in material, or both on the surface and
dispersed in material;
• Surface radioactivity from radionuclides detectable only by beta particles with Emax
s:150keV;
• Radioactivity in soils; and,
• Radioactivity on clothing and persons.
The reviews by the ANSI N13 Committee and HPSSC resulted in the following objections:
1. The limit of 300 dis/min-100 cm2 for most alpha emitters was too low to measure
practically.
2. There was an objection to placing low limits on natural Uranium and Thorium than
in previous versions, 200 removable and 1000 fixed, rather than 1000 removable
222
-------
3.
and 5000 fixed. It was also stated that there was no convincing evidence that
uranium should not be included in the least restrictive category. 'This standard,
if adopted, will severely impact operations at many uranium handling facilities
inferring, without evidence to substantiate such, that past operations have been
health threatening."
The Appendix should provide a more thorough discussion to support the
technical basis for the values in the draft standard.
4. There is no explicit distinction between fixed and loose radioactive contamination.
These objections were representative of three major classes of concerns expressed to the
Working Group over the history of the standard - the ability to measure the limits, the assignment
of radionuclides to specific groups, and the magnitude of the actual values of the limits, including
the rationale for choosing the values. These points will be discussed in succession.
i. Some individuals believe that the limit of 300 dis/min-100 cm2 for total alpha surface
radioactivity was too low to measure. This conclusion was based on an extensive evaluation of
survey instruments for alpha radiation by the Los Alamos National Laboratory and Pacific
Northwest Laboratory (Olsher 1986). All models tested were zinc sulfide scintillation detectors.
One of the evaluation procedures was a survey of a 4 by 4-foot masonite surface divided
into a grid of thirty-six 8 by 8 inch squares. Nine spots of alpha activity were painted on the
surface with a Th-232-based paint. Alpha activity levels ranged from 64 dis/min to 672 dis/min.
There were several different groups of participating survey monitors, including one group with
no previous experience and one group consisting of experienced survey technicians.
Participants used monitors with an audio output (one click for each count). They first estimated
the background, and then very slowly scanned the surface of the grid listening for an increased
"click" rate. Monitors were instructed to locate all of the hot spots and measure their emission
rate in counts/min. They were given 10 minutes to monitor a surface. Based on the data a
statistical analysis was used to determine the alpha source activity for a 50% detection frequency.
The best detection record was achieved by participants with no previous survey
experience. They took a long time to monitor each surface, much longer than experienced
personnel, and much longer than the 10 minutes allotted, The alpha source activity for a 50
percent detection frequency depended on the detector used. The values were as follows:
223
-------
Activity Standard Deviation
dpm dpm
Ludlum 3
Eberline ESP-1
Bicron Analyst
AN/PDR 60
AN/PDR 56F
305
376
478
301
1149
23
43
50
33
294
The probes used with the Ludlum 3, the Eberline ESP-1, and the Eberline PAC 4G have
the following characteristics:
METER
Active area (cm )
Window (mg/cm )
Probe dimensions (cm)
EBERLINE ESP-1
AC-3-7 SCINT
50
0.5
14.4X5.08
EBERLINE PAC 4G
AC-21 GAS PROP
50
0.85
18.5X3.3
LUDLUM 3
43-5 SCINT
50
1
19.7X5.7
The Ludlum AC-21 gas proportional probe used with the PAC 4G alpha monitor is
included for reference. All probe backgrounds were about 1 c/min.
The ANSI N13.12 (1988) presents calculations of the expected response of a survey
instrument to surface radioactivity over which it passes. The response is preferably in terms of
aural clicks, although individual needle deflections can sometimes be detected. For a point
source, the analysis is in terms of the actual time taken to pass over the source and the
efficiency of detection. The time for a detector of length 14 cm to pass over a spot at the
recommended speed of 10 cm/sec is 1.4 sec. A surface activity of 300/min and 30 percent
detection efficiency would give an expected count of something less than 2.1 counts during
traversal by the detector along its axis. The Poisson probability of observing no counts during
the traverse would be e"2>1 X 2.1 °/0! = 0.122 and the probability of observing 1 or more counts
is 1-0.122 = 0.88. Thus the 300 dis/min spot of contamination seems is detectable, if borderline,
and this is borne out by the survey results quoted above (Shapiro 1987).
2. Objections To The Uranium And Thorium Limits
The U.S. Atomic Energy Commission established a limit on the intake by inhalation of
airborne uranium ore dust in 1960. In the following 20 years, it was learned that thorium-230, one
of the long lived decay products that accompanies natural uranium, had a biological half life in
the lungs of 1 year, in contrast to the 120 day biological half life for uranium previously assumed
to also hold for thorium. To use this finding would have required a reduction in the permitted
airborne concentration of uranium ore dust. However, according to McGuire (1983) there is a
mitigating factor associated with uranium ore dust particle size, "...the uranium ore dust in
uranium mills was found to occur with very large particle sizes (10-micron activity median
aerodynamic diameter, AMAD)...The two effects are of about the same magnitude but in
opposing direction. Thus the present uranium ore dust intake limit in NRC regulations should
provide a level of protection consistent with that provided for other airborne radioactive
materials."
224
-------
Does the large median aerodynamic diameter in mills for uranium dust extend to uranium
residual surface radioactivity in other locations? We don't know. However, the low specific
activity of uranium has also been invoked by the IAEA (1963) as reducing the perceived hazard
of surface radioactivity - 'The specific activity of a radionuclide affects the probability that a given
quantity of the radioactivity may enter the body and affects its subsequent behavior in the
body...As the specific activity is an important inherent property of a radionuclide, it cannot be
ignored in making a classification." IAEA ranked uranium 236 in toxicity among 263
radionuclides, in the same grouping as H-3 and 1-129. Pu-239 was ranked 4, 1-131 was 41.
Wrixon (1979) divided radionuclides into two categories, the most hazardous and all the rest.
He used a resuspension factor of 5 X10"5 rrf1 for all radionuclides except those with low specific
activity. 'There is some evidence that resuspension factors for low specific activity materials,
such as natural uranium, are at least ten times lower than those found for plutonium...there is
considerable merit in using a different resuspension factor for low specific activity materials, since
such materials are frequently handled in the absence of other radionuclides...where resuspension
is the main source of airborne contamination, the mass of material that would have to be
resuspended is relevant."
Duggan (1972) noted that less restrictive toxicity classifications for Th-nat and U-nat given
by IAEA were based on acute exposures with a maximum inhaled mass of 10 mg and that
industrial exposures (presumably a more diffuse distribution of activity) will reach much higher
levels since they occur continually over the year - "It is therefore concluded that in any toxicity
classification of the radionuclides, which is to be applied in situations where continual exposure
in industrial conditions is envisaged, both Th-nat and U-nat should be in one of the more toxic
categories and not (as in the IAEA classification) in the group of lowest toxicity.
In considering all these factors, and the low limits for airborne concentrations of uranium,
the working group felt that it could not retain the traditional lowest toxicity classification for natural
uranium, but it also felt that it was being too conservative to classify it with the most toxic
materials. For this reason, a separate category was established for the very low specific activity
radionuclides, intermediate between the lowest and highest toxicity groups.
3. Analysis Of The Hazard From Surface Radioactivity
Early in its deliberations, the working group decided that a workable standard had to be
a performance standard - that it was not possible to present a rigorous technical analysis to
develop limits in terms of risk. In the control of nonradioactive materials, monitoring for surface
contamination is often bypassed completely in favor of limiting control to air monitoring to
determine compliance with prescribed limits for airborne levels.
Thus, the trial use of the standard over 15 years, including repeated critical reviews that
accompanied successive balloting by the HPSSC, has served to validate its status as a
performance standard. The limits and measurement methods appear to be practicable, and most
critical comments require only minor changes. Methods of calculating the hazard from surface
radioactivity are given in the literature, however, and it is of interest to apply these methods, to
hazard evaluations at the limits given here or to compare recommended limits. This should
provide an indication of the magnitude of control achieved.
225
-------
The standard presents an equation for the concentration of activity in the air as a function
of the surface radioactivity. An equilibrium situation is assumed where the activity is removed by
ventilation at the same rate at which it becomes airborne. The equation is:
Surface activity X Area X fractional removal rate
Concentration =
Room volume X rate of air changes
dis/min-m3 =
dis/min-m2 X m2 X fraction/h
m X air changes/h
The standard gives an example using Area/Volume = 0.44 m"1; 5 air changes/hr;
fractional removal rate of
At the guide level for plutonium of 30,000/min-m, the
equilibrium activity works out to 1.2 X 10
environment given in 10 CFR 20.
r14
uci/ml, which is 20 percent of the limit for Pu in the
Another approach that is often used is in terms of the Resuspension factor. This is the
ratio of the air concentration to the surface concentration of activity. It may be derived from the
equation above as:
dis/min-m3 m2 X fraction/h
Resuspension = dis/min-m^ = m3 X air changes/h
The equivalent resuspension factor for the example above is:
0.44/m X lO^/h - 5 = 0.9 X 10"6
It is obvious that the evaluation of airborne activity, whether through the use of a rate of
resuspension or as a resuspension factor, is strongly dependent on values which are difficult to
assign in a general case. Healy (1971) has proposed resuspension rates which are "reasonably,
but not overly, conservative" varying from 5 X lO^/h for vigorous activity, including cleaning or
children at active play to 10^/h for "quiet, no movement." Tabulated values of resuspension
factors have varied from lO^/m to 10 /m.
In one example in the standard, it is assumed that a person might ingest all the
contamination from 10 cm2 of surface each day, an assumption first used by Dunster (1962) and
quoted widely since. The consequences of this assumption are derived by referring the
calculated intake to the Annual Limit on Intake for the radionuclide considered.
The ALI given in ICRP 26 for Radium-226 is 200,000 Bq and this results in a whole-body
dose commitment of 5000 mrem. For a daily intake at the alpha limit of 300 dpm/100crrr
(amounting to 0.5 Bq/day), the committed dose from a year's intake is 4.6 mrem. Because of
the long half-lives, the committed doses are not accumulated until after many years, so the actual
annual dose is much less.
Various other modeling exercises have been done. All come up with doses below limits
set by professional and regulatory agencies at the surface radioactivity levels given in the
standard.
226
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4. On The Distinction Between Fixed And Loose Radioactive Contamination
Healy (1971) did not distinguish between fixed and loose contamination - "In view of the
possibility of abrasive or similar actions on items during repair or modification, it is believed that
these limits should apply to the total contamination and not just to the "loose" contamination
unless other information rules out such actions on the item."
Wrixon et al. (1979) made note of the problem of distinguishing between fixed and loose
radioactivity - "By definition, resuspension and subsequent inhalation of surface contamination
can occur only with loose activity. A clear distinction between loose and fixed contamination
cannot be readily made however. What appears at one time to be firmly fixed may, through
movement and other physical processes, become loose. Furthermore, the resuspension factors
used in the calculations here are derived from a consideration of the total surface activity. The
activity removed by wiping the surfaces involved in the experiments was of the order of 10% of
the total activity; in these cases also, the Derived Levels (DLs) are normally appropriate for the
direct measurement of contamination. If, however, the contamination clearly is firmly fixed, higher
levels than the DLs may be permitted."
"It is recommended therefore, that monitoring be carried out by direct measurement.
Where this is not feasible, as for example, in the presence of high levels of gamma radiation, the
normal methods of wiping should be used and the assumption made that only 10 percent of the
surface contamination has been removed."
After considering the arguments for relying solely on direct measuring in monitoring for
surface radioactivity, the Working Group felt it prudent to retain the use of both direct and indirect
measurements. Depending on the radionuclide and the monitoring circumstances, the wipe test
could be the method of choice for finding surface radioactivity. In any event, it served as a check
on the direct measurement to provide an additional control for compliance with the standard
before release.
CONCLUSION
The guidelines in the proposed ANSI N13.12 are not based up a risk or safety analysis.
The relationship between dose and surface radioactivity is very tenuous. Some calculations and
analysis are presented in an appendix to the standard and thus are not part of the standard itself.
The proposed ANSI N13.12 standard was developed as a consensus performance
standard and has been a de facto standard for over a decade. The lowest limit that by
consensus are achievable has evolved by a continuous evaluation of the practicability of
decontaminating equipment to prescribed limits and on detecting those limits. The guide
concentrations in the proposed ANSI N13.12 were developed as a performance standard, tested
through use and consensus.
227
-------
REFERENCES
[1] Duggan, M. J. 1972. The Toxicity Classification of Th-nat and U-nat. Health Phys. 22:
102-103.
[2] Dunster, H. J. 1962. Surface Contamination Measurements a an Index of Control of
Radioactive Materials. Health Phys. 8:353
\ •
[3] Healy, J. W. 1971. Surface Contamination: Decision Levels. Los Alamos Scientific
Laboratory Report LA-4558-MS (UC-41).
[4] IAEA 1963. A Basic Toxicity Classification of Radionuclides. Vienna: IAEA.
[5] McGuire, S. A. 1983. The NRC's Limit on Intake of Uranium Ore Dust. United States
Nuclear Regulatory Commission Report NUREG-0941. Office of Nuclear Regulatory
Research.
[6] Olsher, R. H., J. S. Haynie and E. Schultz. 1986. Alpha RADIAC Evaluation Project. Los
Alamos National Laboratory Report LA-10729.
[7] Schultz, N. B. and A. F. Becher. 1963. Correlation of Uranium Alpha Surface
Contamination, Air-Borne Concentrations, and Urinary Excretion Rates. Health Phys.
9:901-909.
[8] Shapiro, J. 1980. The Development of the American National Standard, Control of
Radioactive Surface Contamination on Materials, Equipment and Facilities to be Released
for Uncontrolled Use. Proceedings of the 5th International Congress of the International
Radiation Protection Association. New York: Pergamon Press.
[9] Shapiro, J. 1987. HPS/ANSI Health Physics Guide. Proceedings of the 1987 International
Decommissioning Symposium, October 4-8, Pittsburgh, PA.
[10] Wrixon, A. D., G. S. Linsley, K. C. Binns and D. F. White. 1979. Derived Limits for Surface
Contamination. National Radiological Protection Board report NRPB-DL2. Harwell,
England: NRPB
228
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TABLE 1
Surface Contamination Limits:
Abbreviated Table From Proposed ANSI N328 (1974)
Nuclide
Group 1: Nuclides for which the MFC for
air is 2 x 10"13 Ci/m3 or less or for which
the MPC for water is 2 x 10"7 Ci/m3 or less;
includes Pu-239, Ra-226, Pb-210, 1-125,
1-129, etc.
Group 2: Those nuclides not in Group 1 for which
the MPC for air is 1 x 10'12 Ci/m3 or less or
for which the MPC for water is 1 x 10"6 or less;
includes Po-210; Sr-90; Th-232; U-232, etc.
Group 3: Those nuclides not in Group 1 or Group 2.
dpm/100cm
Total Removable
100
1,000
5,000
20
200
1,000
TABLE 2
Surface Contamination Limits:
Abbreviated Table From Regulatory Guide 1.86 (1974)
Nuclide
Transuranics, Ra-226, Ra-228,
Th-230, Th-228, Pa-231, Ac-227,
1-125, 1-129
Th-Natural, Th-232, Sr-90, Ra-223,
Ra-224, U-232, 1-126, 1-131, 1-133
U-Natural, U-235, U-238, and
associated decay products
Beta-gamma emitters except Sr-90
and others noted
Average Total
100
dpm/100cm
Maximum Total
300
Removable
20
1,000
5,000
5,000 BY
3,000
15,000
15,000 BY
200
1,000
1,000-Y
229
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TABLE 3
Surface Contamination Limits:
Abbreviated Table From Draft ANSI N13.12 (1978)
Nuclide
'Group 1: nuclides for which the MFC for air
ts 2 x 10~13 Ci/m3 or.less or for which the
nonoccupational MFC for water is
2 x 10~7 Ci/m3 or less; includes Pu-239, Ra-226,
Pb-210, 1-125,1-129, etc.
Group 2: those nuclides not in Group 1 for
which the MFC for air is 1 x 10'12 Ci/m3
or less or for which the MFC for water is
1 x 10"6 Ci/m3 or less; includes Po-210;
Sr-90; Th-232; U-232, etc.
Group 3: those nuclides not in Group 1
or Group 2
dpm/100cm
Total Removable
Nondetectable 20
2,000 200
Nondetectable By
5,000
1,000
TABLE 4
Surface Contamination Limits:
Abbreviated Table From Proposed ANSI N13.12 (1981)
Group Description
dpm/100cm
Total Removable
1 Alpha emitters except
Nat-U and Nat-Th
2 Sr-90, 1-125,1-129,
1-131, Ra-228
3 Beta emitters not in Group 2
with Emax > 150 keV; Nat-U and
Nat-Th
300
5,000
5,000
20
200
1,000
230
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TABLE 5
Surface Contamination Limits:
Abbreviation of Table From Proposed ANSI N13.12 (1983)
Group Description
1
Alpha emitters except
Nat-U and Nat-Th, depleted
uranium, U-238 and Th-232
2 Sr-90,1-125,1-129,1-131, Ra-228
3 All radionuclides not in groups 1
or 2 except beta emitters with
dpm/100cm
Total
300
5,000
5,000
Removable
20
200
1,000
231
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TABLE 6
Surface Contamination Limits:
Table From Proposed ANSI N13.12 (1988)a
Group Description
1 All alpha emitters except those
with extremely low specific .
activity and their associated
decay products as listed in
Group 4; Pb-210, Rac-228
2 Sr-90, 1-125, 1-129
1-131d
ACTIVITY GUIDEb
(dpm/100 crn2)
Removable (Fixed plus removable)
20
300
200
5000
All beta and gamma emitters not
specified in Groups 1, 2, and 4
except pure beta emitters with
1000
5000
•max
;>150KeVe
Uranium (natural, depleted,
enriched), Th-natf
200
1000
a. A rationale for these surface activity guides is presented in Appendix B. Where both alpha and
beta-gamma emitting radionuclides exist, the limits tablished for alpha and beta-gamma
emitting nuclicles shall apply independently.
b. The levels may be averaged over one square meter provided the maximum surface activity
in any area of 1 00 cm2 is less than three times the guide values. For purposed of averaging,
any square meter of surface shall be considered to be above the activity guide G if: (1) from
measurements of a representative number n of sections it is determined that 1/n 2n S, > G,
where S, is the dis/min-100 cm2 determined from measurement of section i; or (2) it is
determined that the sum of the activity of all isolated spots or particles in any 1 00 cm area
exceeds 3G.
For purposes of this standard, the disintegration rate refers to those disintegrations which
result in the emission of alpha particles, beta particles or electrons with Emax k 150 KeV, or
photons of energy
the dose.
20 KeV, provided these particles are responsible for more than 80% of
c. Pb-21 0 is included due to an alpha emitter, Po-21 0, in its decay chain and Ra-228 is included
due to an alpha emitter, Th-228, in its decay chain.
d. These are the radionuclides undergoing beta or electron capture decay that present the
greatest hazards as surface radioactivity.
232
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f.
TABLE 6 (con't)
The pure beta emitters with maximum energy less than 150 KeV are excluded because
detection by direct methods is not practical and they must be treated on a case-by-case
basis. However, radionuclides that are detectable by direct mesurement with appropriate
instrumentation through emission of low-energy X and gamma rays (as in electron capture)
or through the presence of short-lived decay products are included in this category.
U-nat and Th-nat include gross alpha disintegration rates of natural uranium, depleted
uranium, uranium enriched to less than 10% U-235, Th-232, and their decay products.
233
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TABLE 7
FREQUENCY FOR PROBABILITY FOR OBSERVATION WITH A SURVEY
INSTRUMENT OF "n" OR MORE COUNTS8 FROM A POINT SOURCE
OF 300 DIS/MIN, ALPHA
Ec
(counts per
disintegration)
0.5
0.4
(seconds
counting
time)
1.67
1.0
0.67
0.5
0.33
1.67
1.0
0.67
0.5
0.33
n = 1
99
92
81
72
57
97
87
74
64
49
n = 2
92
72
50
36
86
60
39
n = 3
79
46
65
33
0.3
0.2
1.67
1.0
0.67
0.5
0.33
1.67
1.0
0.67
0.5
0.33
92
78
64
53
40
82
67
49
72
45
46
51
27
24
a. Background is 1 (cpm).
b. The survey speed of the detector is v = d/t(cm/s) where t is the counting time and d is the
width of the detector.
c. The mean counts/disintegration with the source under the window of the stationary detector.
234
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TABLES
FREQUENCY FOR PROBABILITY FOR OBSERVATION OF
MORE COUNTS3 FROM A WIPE SOURCE OF 20 DIS/MIN, ALPHA
"n" OR
(counts per
disintegration)
0.1
0.2
0.3
0.4
0.5
t°
(seconds
counting
time)
0.5
1.0
1.5
2.0
0.5
1.0
1.5
2.0
0.5
1.0
1.5
2.0
0.5
1.0
1.5
2.0
0.5
1.0
1.5
2.0
n = 1
78
95
99
100
92
99
100
100
97
100
100
100
99
100
100
100
100
100
100
100
n =2
80
94
98
71
96
100
100
86
99
100
100
94
100
100
100
97
100
100
100
n = 3
58
83
94
__
96
98
100
68
97
100
100
83
99
100
100
91
100
100
100
n = 4
66
85
74
94
99
92
99
100
66
98
100
100
80
100
100
100
a. Background used is 1 cpm.
235
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HPS STANDARDS COMMITTEE
PROPOSED ANSI N328-1974(?)
HPS STANDARDS COMMITTEE
ANSI
DRAFT ANSI N13.12
(AUGUST 1978)
PROSPOSED ANSI N13.12
(AUGUST 1981)
PROPOSED ANSI N13.12
(JUNE 1983)
PROPOSED ANSI N13.12
(JUNE 1988)
PROPOSED GUIDELINE ORO 830
(MARCH 1983)
DOE
I
AEC / NRC
AEC REG. GUIDE 1 .86
(JUNE 1974)
NRC GUIDELINE
(NOV. 1976 and JULY 1982)
DOE
GUIDELINES FOR
FUSRAP and SFMP SITES
(JULY 1985)
Figure 1. Interrelation of Surface Contamination Guidelines,
236
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EPA's Proposed Environmental Standards for
Low-Level Radioactive Waste Disposal and
Criteria for Below Regulatory Concern
William F. Holcomb and James M. Gruhlke
Environmental Protection Agency
Office of Radiation Programs
ABSTRACT
The Environmental Protection Agency has developed generally applicable environmental
standards for land disposal of low-level radioactive waste. The elements of the proposed
standards will include: (a) exposure limits for pre-disposal management and storage operations;
(b) criteria for other agencies to follow in specifying wastes that are Below Regulatory Concern
(BRC); (c) post-disposal exposure limits; (d) ground-water protection requirements; and
(e) qualitative implementation requirements.
To support the concept of BRC, the Agency has developed technical information, cost
data and a risk assessment methodology for analyzing promising candidate waste streams. The
BRC criteria are based on general population health risks, maximum annual exposures to critical
population groups, and the costs now associated with the regulation of these wastes.
The regulatory package for these standards is presently under interagency review.
Promulgation of the proposed standards is expected in 1990,
INTRODUCTION
In August 1983, the Environmental Protection Agency (EPA) published an Advanced
Notice of Proposed Rulemaking (ANPRM) [1 ], stating the Agency's intention to develop generally
applicable environmental standards for the land disposal of low-level radioactive waste (LLW).
The intent is that these standards (40 CFR 193) must be met by facilities that dispose of LLW,
whether the facilities are licensed and regulated by the Nuclear Regulatory Commission (NRC)
or their Agreement States, or are owned and operated by the Department of Energy (DOE).
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The EPA LLW Standard is intended to cover disposal of all AEA materials not covered by
other EPA Standards, i.e., all radioactive waste that is not high-level and transuranic radioactive
waste or spent nuclear fuel, as defined in 40 CFR Part 191, or, uranium or thorium mill tailings
subject to 40 CFR Part 192. The Standard will have the following elements:
(a) Low-level waste pre-disposal management and storage. This
would include limits on radiation exposure to individuals during
processing, management, and storage of LLW.
(b) Definition of radiation exposures related to low-level radioactive
waste disposal that are sufficiently small that they do not need to
be regulated regarding their radiation hazard (i.e., a level "below
regulatory concern").
(c) Limits on radiation exposure to individuals after the disposal site is
closed.
(d) Ground-water protection requirements for both pre- and post-
disposal phases.
(e) Qualitative implementation requirements.
STANDARDS RATIONALE
Individual Radiation Exposure Limits During Management and Storage (Pre-Disposal)
This element would limit annual effective whole body exposure from all environmental
pathways to any member of the public from facilities which process, manage, or store LLW. This
would include the operation phase of regulated LLW disposal facilities, i.e., while they are
receiving and emplacing waste; and "away from generator" LLW management, processing and
storage facilities.
The Office of Radiation Program's analyses indicates a standard around 25 mrem/yrfrom
all pathways would be consistent with the technology and other similar standards.
"Below Regulatory Concern" Criteria
Criteria are being proposed for identifying LLW with sufficiently low levels of radioactivity
to qualify as "Below Regulatory Concern" (BRC). Any waste meeting these criteria could be
disposed of as a non-radioactive waste. However, if it had Resource Conservation and Recovery
Act (RCRA) hazardous characteristics, it would have to be disposed of in compliance with RCRA
regulations. The EPA would not be involved in identifying or selecting specific LLW types which
qualify as BRC wastes, the NRC, States and DOE would implement the use of our criteria for
determining which wastes would qualify for disposal by less restrictive means.
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In arriving at a proposed BRC level, EPA carefully weighed and considered many factors.
The Office's analyses has indicated that a standard around 4 mrem/yr would provide protection
of the public and the environment.
Individual Radiation Exposure Limits for Post-Disposal
Our standard will establish limits on exposure through all pathways to members of the
public from the land disposal of LLW. EPA's post-disposal limit would apply to any DOE or
NRC/State-licensed LLW land disposal method or facility constructed after the effective date of
the rule and apply to existing disposal facilities within a certain time frame.
A persuasive reason would be needed to significantly depart from a 25 mrem/yr level.
EPA's technical analysis has not revealed any such reason so far.
Ground-Water Protection
The protection of the Nation's groundwaters is of major importance in EPA and such a
consideration is particularly appropriate in land-based waste disposal standards.
Two sets of ground-water protection requirements will be proposed and public comments
solicited. In both proposals Class I groundwaters require the highest levels of protection and
represents those that are highly vulnerable to contamination and serve as irreplaceable sources
of drinking water for large populations. It is appropriate to give these groundwaters the highest
level of protection, i.e., non-degradation. The two proposals differ only with respect to the
protection levels for Class II groundwaters which represent all non-Class I present or potential
sources of drinking water. The first proposal would protect Class II groundwaters from high yield
aquifers (which are or could be a community water supply) to an annual effective dose equivalent
of 4 mrem, while Class II groundwaters from low yield aquifers (which generally could not provide
a community water supply) would be protected as a part of the 25 mrem/yr all pathways pre- and
post-disposal performance standards. The second proposal would protect all Class II
groundwaters, which is by far the largest category of groundwaters, to an annual effective dose
equivalent of 4 mrem. This level is comparable to the 4 mrem/yr Maximum Contaminant Level
(MCL) for manmade beta particle and photon radioactivity established for public water supplies
by EPA's drinking water standards under the Safe Drinking Water Act [2].
Finally, both proposals recommend the same levels of protection for Class III
groundwaters. Class III A groundwaters are protected to the level applicable to the highest class
of groundwater to which it is interconnected. Class III B groundwaters have a low degree of
interconnection with other classes of groundwater and would be protected as a part of the 25
mrem/yr all pathways pre- and post-disposal performance standards.
Qualitative Requirements
Qualitative requirements are being proposed which would make clear the context and
assumptions within which we expect the Standard to be implemented.
These requirements would address areas not appropriate for quantitative requirements
and compensate for the uncertainties that necessarily accompany plans to isolate radioactive
239
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wastes from the environment for a long time. They would include: (a) limiting the dependence
on active institutional controls (such as guarding, maintenance or cleanup of releases) after
disposal to no more than 100 years, (b) providing passive institutional measures (such as
permanent markers, records or archives or government ownership) which should reduce the
chance of inadvertent human intrusion beyond the active institutional control period; (c) requiring
monitoring during disposal and post-disposal phases which should be done with techniques that
would not jeopardize the isolation of the wastes; and (d) suggesting site location away from
areas containing materials not widely available from other sources (such as minerals, fuels and
groundwaters).
BRC RATIONALE
Philosophical Approach
We believe that when some LLW streams contain sufficiently small concentrations of
radioactivity, there is no reason, from a public health point of view, not to dispose of these
wastes as we would any "non-radioactive" trash. However, this is not to say that there is no
remaining risk or that any form of waste disposal is risk free. EPA recognizes that this remaining
risk should be estimated, and adequate assurance must be given that it is not unreasonable in
light of the benefits of deregulating it.
It should be noted that EPA's concept of BRC is not one of a de minimis level. BRC is
based on a careful analysis of specific sources of exposures, e.g., low-level radioactive waste
disposal, and the methods of exposure control. A BRC decision is a conclusion, that relative to
a specific practice, certain radiological impacts will be small and not worth the effort that would
be necessary to further reduce them. The concept of a de minimis level does not consider
specific practices, costs or facility locations, but refers to a level of negligible risk from all sources
of radiation exposure or any other potential insult. If a generic de minimis risk level were
determined, it would be at or below any BRC level.
Concepts for Establishing BRC Levels
We believe there are two essential elements when deciding not to regulate a beneficial
practice which can cause radiation exposure. The first is that the deleterious impact of the
practice on health in the exposed population, taken as a whole, is small enough that the effort
and expense of regulation is not warranted. The second is that the risk to any person is small
compared with other risks in society.
In developing our BRC proposal for low-level waste an initial premise was that since this
would be a criterion for not regulating certain waste streams, and that wastes that met the criteria
would not receive any monitoring or follow-up, there would be no long-term confirmation of the
results. This implies that the BRC level should be well below the limit for regulated disposal.
Additionally, as we see it, the concept of "As Low As Reasonably Achievable" (ALARA) is indeed
related to BRC. Such a regulatory cutoff may be appropriately viewed as a floor to ALARA. The
proper perspective, we believe, is that there is a standard or upper limit (the level for regulated
disposal) below which one practices ALARA, and then at some lower level (BRC) there could be
240
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a regulatory cut-off where we say that for this practice it isn't worth the effort to go any lower.
This should allow regulatory resources to be expended more effectively.
Basic Criteria
There are several necessary considerations in determining a regulatory cutoff. One of the
most important of these is economics, i.e., the costs of implementing control. There are two
important consequences of this: First, there may be no single number applicable to all practices
as the costs of control may vary with practice. Therefore, there may be a separate and distinct
number associated with each activity or "practice." Second, as we previously indicated, such a
cutoff by the nature of its determination, serves as a floor to "As Low As Reasonably Achievable"
(ALARA), but only for the specific practice.
We believe that a BRC decision consideration should at least:
1. Exempt a relatively small collective dose which does not significantly change with
individual dose in the area of the BRC decision. We have no magic number for this collective
dose, but its size should be considered relative to the total collective dose if the whole of the
practice were unregulated.
2. Represent an individual dose that is well below the overall regulatory limit for the
practice. This is so it will represent a small individual risk and can be readily differentiated from
the regulatory limit, which is presumed to represent an acceptable risk.
3. Be formulated so that it does not increase other environmental impacts.
4. Be compatible with legal authorities and other control actions dealing with
nonradioactive pollutants.
5. Be able to be practically implemented using available management systems, analytical
techniques, and instruments.
6. Result in the possibility of some resource savings for the regulator and the practice.
However, if it holds little hope for ultimately saving resources, it would negate one of our prime
reasons for considering a BRC in the first place (i.e., allocation of resources to the more
significant risks).
7. Be supported by an analysis that provides a reasonable and sufficient basis for the
decision-makers to arrive at a judgment.
Risk Assessment Methodology and Health Impacts
To set a BRC level that would provide adequate public health and environmental
protection, a methodology was developed for assessing health impacts, i.e., the cumulative
population health effects and critical population group (CPG) dose exposure.
To estimate the possible doses and economic impacts resulting from potential BRC
deregulation, EPA modeled several scenarios of possible waste streams.disposal methods, and
241
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various demographic and hydrogeologic/climatic settings. Surrogate types of LLW were chosen
to represent a wide variety of waste generators, such as power reactors, uranium fuel fabrication
and processing facilities, industrial facilities, medical facilities [3]. The BRC scenarios included
a variety of disposal methods, i.e., municipal sanitary landfills, dumps, on-site landfills, and
incineration methods situated in rural, suburban, and urban demographic settings.
In determining radiation doses for the CPG who might collect the wastes, on-site disposal
facility workers doing routine incineration and disposal operations, reclaimers, inadvertent
intruders, and nearby residents exposed to water, food, and inhalation pathways were included.
Health effects to the general population over 10,000 years were calculated. Individual radiation
doses were calculated as a committed annual effective radiation dose equivalent for 10,000 years
to an individual in the CPG. Estimated lifetime risks were also calculated for the CPG [4].
Cost Benefit Analysis
EPA also performed a national economic assessment of potential BRC waste streams.
A cost-benefit evaluation was done for a range of alternative exposure levels from 0.1 to 15
mrem/yr. AH of the alternative non-zero BRC exposure levels-'could reduce the volume of
regulated LLW. Volumes of regulated waste could be reduced by up to 43 percent, with
attendant cost-savings ranging up to over $700 million over a 20-year period, except at the very
low levels of exposure where it would require considerable expense to regulate those materials
presently not being regulated [4,5],
EPA's Approach to the Proposed Criteria .
Foremost in our approach was protection of the public and the environment.
approach was to develop an exposure level with assurance of no undue risk.
Our
EPA believes that at the BRC level, the health risk should be very small when compared
to other risks we encounter in our daily lives. The Agency considered: (a) the risk from natural
background radiation, which is on the order of 3x10^ to 10'2 lifetime; and (b) the lifetime risk of
developing fatal cancer, which is about 2x10"1.
Another consideration was that if a person were exposed to several deregulated waste
streams, the total risk should still be small. Our health impact analysis, even usjng multiple waste
streams, indicated lifetime risks for many of the disposal scenarios to be less than IxlO"4.
Based on the technical and economic evaluation, EPA considered several specific
regulatory options for BRC all of which represented levels of protection more stringent than for
proposed limits on regulated LLW disposal. These options ranged upward from the position that
there would be no BRC criteria allowed, in which case the regulated disposal of all wastes would
be required as long as any radioactivity remained.
The other options considered included criteria for the less restrictive disposal of BRC with
limits of 0.1 mrem/yr, 1 mrem/yr, 4 mrem/yr and 15 mrem/yr to the individual which results in a
lifetime risk of 2.8x10"6, 2.8x10'5,1.1 x10"4 and 4.2x10"4, respectively.
242
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Another regulatory option was to accept the current practice, wherein EPA would either
stop its BRC effort or endorse the existing practice of case-by-case deregulations.
In considering a BRC level, we also took note of other regulated risk levels used by other
government programs. Of special note was the 4 mrem/yr dose level (1.1x10"4 lifetime risk) set
for man-made radionuciides in the National Interim Drinking Water Standards (2).
Another consideration was the cumulative population health effect. In our BRC range (0.1
to 15 mrem/yr) the additional health effects would be of the order of less than 0.001 to 0.05 per
year over the 10,000 years.
All of these considerations and evaluation together with our risk analysis indicate that
many wastes could be disposed of without consideration of their very low levels of radioactivity.
As a result, we are proposing a BRC value of 4 mrem/yr (a lifetime risk of 1.1x10"4). We believe
this provides reasonable assurance of protecting the public and the environment with minimal
incremental risks.
Implementation Guidelines
* •
There are several basic factors EPA considers necessary for an agency making a specific
waste stream deregulatory decision.
An evaluation of the collective dose, and thus the risk to the population, is a necessary
part of any cost-benefit evaluations. It must be estimated to assure there is no significant impact
on public health. It will also serve as an indication of, and thus a measure to deter the practice
of, waste dilution to reach the BRC individual exposure limits. Regardless of whether collective
dose is a specific parameter in any BRC limits, the total potential health impact of an exemption
needs to be considered in some way as a part of any decision for deregulation.
It is also important to be able to characterize with reasonable certainly the waste streams'
physical, chemical, and radiological characteristics.
The waste should have negligible potential recycle value so it will not be attractive to
scavengers who might otherwise take the waste and use it.
It is important that sufficient recordkeeping and reporting requirements be included to
provide the regulatory agency with adequate information to understand where and how much
BRC waste is being disposed of. Records containing this information will be needed to
document how the BRC concept is working in actual practice and will serve as a "final
accounting" for the BRC wastes before regulatory control is irreversibly lost.
Consistency With International Policy
The Agency also reviewed the guidance for similar exemptions being developed by other
countries and international organizations. The Canadian Atomic Energy Control Board [6] and
the United Kingdom's National Radiological Protection Board [7] have issued documents
proposing a 5 mrem/yr dose criterion to members of the public to be used for case-by-case
analysis of applications for license exemptions for radioactive waste disposal.
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The International Atomic Energy Agency (IAEA) has considered a de minimis dose of 1
mrem/yr to the average individual in the CPG for ocean dumping and for quantities of solid
radioactive waste for uncontrolled disposal by incineration and landfill [8,9]. The International
Commission on Radiological Protection has also adopted a radiation principle of an annual
individual dose exemption criterion of 1 mrem [10]. These latter de minimis recommendations
differ from BRC, in that they are value judgments of negligible risk and make no attempt to
consider the cost of regulation.
In 1988, the IAEA issued its recommendations on "Principles for the Exemption of
Radiation Source and Practices from Regulatory Control." They concluded that individual doses
of about 1 mrem/yr from each exempted source, leading to a total dose from all such sources
of a few mrem/yr, were reasonable, but only if the societal impact was sufficiently low. In this
regard, they suggested that 100 man-rem from the entire practice would qualify it for exemption
without further analysis.
CONCLUSION
We are now in the final stage of proposing the LLW waste standards. We have prepared
regulatory support documents which will be available when the proposed standard is published
in the Federal Register.
The EPA Low-Level Radioactive Waste Management program staff believes the Standards
covering the above described areas would provide adequate protection to members of the
general public with a reasonable balance of risks and costs.
REFERENCES
[1 ] U.S. Environmental Protection Agency, 40 CFR Part 193, Environmental Radiation
Protection Standards for Low-Level Radioactive Waste Disposal, Advanced Notice
of Proposed Rulemaking, Federal Register, 48(170):39563, August 31,1983,
[2] U.S. Environmental Protection Agency, 40 CFR Part 141, Interim Primary Drinking
Water Regulations - Promulgation of Regulations on Radionuclides, Federal
Register, 41 (133) :28402-28405, July 9, 1976.
[3] W.F. Holcomb and J.M. Gruhlke, The EPA's Criteria Development for Radioactive
Waste that is Below Regulatory Concern, Proceedings of the 8th Annual DOE Low-
Level Waste Management Forum, Volume VII, pp 6-18, CONF-870859, EG&G
Idaho, Inc. Idaho Falls, ID 1987.
[4] W.F. Holcomb and J.M. Gruhlke, EPA LLW Standards Program: Below Regulatory
Concern Criteria Development, Proceedings of the Symposium on Waste
Management 87", Volume I, pp 327-331, University of Arizona, Tucson, AZ, 1987.
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[5] C.E. Foutes, Economics of EPA's BRC Criterion: Impacts and Implications,"
Proceedings of the 9th Annual DOE Low-Level Radioactive Waste Management
Conference, CONF0870859, EG&G Idaho, Inc., Idaho Falls, ID, 1988.
[6] Atomic Energy Control Board, The Basis for Exempting the Disposal of Certain
Radioactive Materials from Licensing, Proposed Regulatory Policy Statement,
AECB Consultative Document C-85, Ottawa, Ontario, Canada, May 6, 1985.
[7] National Radiological Protection Board, Small Radiation Doses to Members of the
Public, NRPB, ASP7, London (HMSO), January 1985.
[8] International Atomic Energy Agency, Considerations Concerning De Minimis
Quantities of Radioactive Waste Suitable for Dumping at Sea Under a General
Permit, IAEA-TECHDOC-244, Vienna, Austria, February 1981.
[9] International Atomic Energy Agency, De Minimis Concepts in Radioactive Waste
Disposal - Concentrations in Defining De Minimis Quantities of Solid Radioactive
Waste for Uncontrolled Disposal by Incineration and Landfill, IAEA-TECDOC-282,
Vienna, Austria, February 1983.
[10] The International Commission on Radiological Protection, Radiation Protection
Principles for the Disposal of Solid Radioactive Waste, ICRP Publication 46,
Pergamon Press, New York 1985.
[11] International Atomic Energy Agency, Principles for the Exemption of Radiation
Sources and Practices from Regulatory Control, IAEA Safety Series No. 89, Vienna,
Austria, September 1988.
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EPRI Discussion Paper on BRC
and De Minimis Concepts
Jene N. Vance, Vance & Associates
Patricia J. Robinson, Electric Power Research Institute
ABSTRACT
The purpose of this discussion paper is to examine the definitions, relationships and
characteristics of the terms: de minimis, below regulatory concern (BRC) and generic BRC. In
the past few years these terms have been used in various contexts such as: disposal of very
low-level wastes, dose levels to members of the public, and the unrestricted release or use of
slightly radioactive materials. While all of the terms have been taken to imply radiation risk levels
of little significance, the use of the terms has not been wholly consistent and there appear to be
multiple and sometimes conflicting interpretations of the terms. The use of these terms is
explored herein. In addition to this discussion paper, the results of the Electric Power Research
Institute (EPRI) $2M research program to establish the technical basis for petitioning the NRC to
exempt very low-level nuclear reactor wastes from LLW disposal facilities are also summarized
in this paper.
INTRODUCTION
The generally accepted uses of these terms, as discussed in detail below, are:
de minimis: a low exposure level based on the corresponding risk which has been
determined to be neglible or trivial based solely on a comparison with other generally
accepted risks.
\
activity-specific BRC: an exposure level set for a specific regulated activity based on an
cost-risk reduction evaluation showing that further controls to reduce exposures beyond
are not justified.
generic BRC: an exposure level set for a class of comparable activities based on a
cost-risk reduction evaluation of a typical activity in that class showing that further
controls to reduce exposures beyond this level are not justified.
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DE MINIMIS
The term "de minimis" is an abbreviation of the phrase "de minimis con curat lex" which
implies: the law is not concerned with trivialities. The initial use of the term in radiation protection
matters was to identify the need for establishing or codifying a radiation exposure level (or risk)
to a member of the public which would be considered trivial by the members of the public. The
approaches suggested for the establishment of such a dose were generally on a comparative
risk basis against other society risks readily accepted or against naturally occurring background
radiation levels. From a regulatory point of view, any activity or practice which had projected
doses at or below the de minimis dose would not require further regulation, including actions or
controls to reduce the doses below the de minimis level. Thus, implicit in the de minimis
definition is that a COST-BENEFIT evaluation would not be required at this dose level to
determine if further dose reductions can be justified. Likewise, the definition indicates that the
de minimis dose would be universally applicable to any activity or practice which had the
potential to expose members of the public. Thus the important features of a de minimis exposure
limit, identified from the definitions and uses, are that the dose level would apply to all activities
and would be based on a comparative risk basis, rather than a COST-BENEFIT basis.
BRC
Subsequent to these initial discussions, the de minimis concept was supplanted with the
concept of below regulatory concern (BRC). This new concept was intended to differentiate
between de minimis, which implies the establishment of a trivial risk without the need for
additional regulatory considerations, and BRC which would embody other regulatory
considerations such as cost/benefit or evaluations.
Although this seems to be the basis for the introduction of the BRC concept, considerable
confusion has since resulted in the use and application of the term. Much of the confusion may
stem from the unfortunate choice of the words "below regulatory concern" which connote a
degree of regulatory triviality rather than a dose level which represents a regulatory
cost-effectiveness limit or objective. For example, BRC is often described using words such as:
"minimal risk," "trivial risks from a regulatory standpoint," "insignificant risk level," all of which tend
to emphasize the smallness of the risk as the basis for the dose level rather than the fact that
further regulatory efforts to reduce the dose below the BRC dose level are not considered
cost-effective on the basis of a cost-risk reduction evaluation. Examples of cost-effective limits
are demonstrated by the 10CFR50, Appendix I dose objectives and the 40CFR190 dose limit both
of which are cost-benefit based limits, but which contain no reference to BRC. Obviously, these
cost effective limits represent a dividing line between doses which are above a cost-effective level
and are therefore of regulatory concern and doses which are below regulatory concern for these
activities. Doses at the limits are neither above regulatory concern nor are they below regulatory
concern. Perhaps a better choice of words for a cost-effective objective or limit which would fall
between the adequate protection limit and a de minimis level would be a Regulatory
Cost-Effective Limit rather than BRC.
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GENERIC BRC
Although there have been recent discussions regarding the establishment of a "generic
BRC" dose or risk, the basis for establishing a generic BRC dose level is not clear. If generic
BRC is intended to apply all regulated activities then it isn't clear that a generic BRC level can
be established by a cost-risk reduction evaluation without doing an injustice to BRC levels which
would be established on a source-specific basis. By their nature, cost risk reduction evaluations
are performed on a source-specific and regulatory control-specific basis in order for the
cost-effectiveness evaluation to be valid. If a generic cost-effectiveness evaluation is performed
for all sources and all control measures then It is likely that a more restrictive dose level would
be justified on the basis of the most cost-effective control measures applied to the easiest source
to control. This would, in effect, invalidate the cost-effectiveness evaluations for some of the
other sources.
A clear example of how a cost-benefit evaluation of too broad a class of unrelated
activities could lead to an inappropriate generic BRC level is provided by EPA's evaluation of low
level waste disposal. EPA lumped together both reactor generated and medical wastes even
though only medical wastes contain significant quantities of carbon-14. As a result, the
carbon-14 in medical wastes led to a significantly lower generic BRC level than would have
resulted from an evaluation of reactor wastes alone. This shows that to be meaningful, a generic
BRC level must be calculated only for a class of related activities which can be represented
adequately as a typical member of that class.
Notwithstanding the potential for overly restrictive doses being applied to some activities
or practices, a conservatively low generic BRC dose level may still be beneficial for licensees who
would not choose to perform a source-specific cost-benefit evaluation which potentially could
justify a higher dose level. The obvious benefit would be the elimination of the need for holding
an additional rulemaking on a proposed activity-specific BRC dose level. However, there may
be other ways to eliminate the need for source specific rulemaking, as discussed below under
implementation. Even if a generic BRC dose level was established, there should be a higher
dose level allowed as indicated by the results of the cost-benefit evaluation for those licensees
that would choose to perform a cost-benefit evaluation for a proposed practice.
MULTIPLE SOURCES
To account for the potential for multiple sources or multiple exposures to a maximum
exposed individual from more than one source it is important to identify the dose level for which
the accounting is important. For example, the basic regulatory dose limit (proposed 100
mrem/yr) is fundamentally a risk-based limit and as such multiple sources must be taken into
account to ensure that the limit is not exceeded for the maximum exposed individuals. Likewise,
the de minimis dose level is also risk-based and therefore should take into account multiple
exposures. It would be preferable to account for the potential of multiple source exposures at
the de minimis level in the licensee's compliance implementation rather than by an arbitrary, and
likely overly conservative, reduction in the de minimis dose level. Arbitrary factors of 5 to 10 have
been suggested to account for multiple sources or exposures. However, in reality for most of
the regulated activities the likelihood of overlap in the exposure of the maximum exposed
individual from several activities is exceedingly small.
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Because the BRC dose levels are cost-benefit and not risk-based, there is no need to
account for multiple sources by reducing the BRC dose levels by some arbitrary factor unless
there is a potential to exceed the basis regulatory dose limit. In fact further reductions in any
cost-benefit BRC dose level invalidates the cost-benefit evaluation on which the dose level is
based. And, as noted above for a de minimis dose, the potential for any significant overlap of
exposures to the maximum exposed Individual for more than one of the BRC activities is
extremely remote. However, in the unusual case where there is a significant potential for multiple
exposures, or where a BRC level is a significant fraction of the basic radiation limit, a more
detailed evaluation of the potential for multiple exposures could be required.
IMPLEMENTATION
DeMinimis
It would be necessary to conduct a rulemaking to establish a de minimis dose or risk
level. The rulemaking would focus on the basis for establishing that a given risk would be
considered trivial by a member of the public. In addition to establishing a regulatory cut-off, the
rulemaking could also establish the need for the NRC to evaluate and accept the dose
computation models used to determine the expected doses from a proposed practice or activity.
If NRC pre-approval of the dose compliance models is required, it would allow the NRG to
account for the potential for overlap of exposures from multiple sources and also to account for
recycle, re-use or use of licensed materials in indeterminate exposure pathways. Both overlap
and indeterminate pathways could be accounted for by approving a dose limit for specific
applications, which would be below the codified de minimis dose. As an alternative to this
approach, the de minimis dose limit established in the regulations could be reduced by some
arbitrary factor to account for multiple sources. It would still be necessary to account for
indeterminate exposure pathways in the dose computation models.
BRC
To establish a BRC dose level for a source-specific practice, the NRC could follow the
procedure defined by the NRC policy statement on BRC waste disposal wherein waste
generators on a national scale would petition the NRC for a rulemaking to allow waste disposal
of wastes by means other than in an NRC-licensed disposal facility as long as the doses were
within the BRC level. However, in following this process, the phrase 'a few millirems' will not be
taken as a pre-determined number suitable for a risk based limit. Rather, the phrase would be
interpreted as implying that a cost-benefit evaluation is the appropriate process for setting the
BRC limits, and the regulations would prescribe the need for a cost-benefit evaluation in
accordance with the definition for a BRC dose level. Thus, consistent with the definition of BRC,
the regulations should not specify any exposure limits but would require applicants to establish
the appropriate exposure limits by a cost-benefit evaluation. It would also be beneficial for the
NRC to publish general guidance on an acceptable methodology for conducting a cost-benefit
evaluation including the cost/risk averted ratio used.
For the generic BRC dose level it will also be necessary for the NRC to hold a rulemaking
hearing to establish the dose level. The rulemaking would focus on the cost-benefit evaluations
for all or a significant portion of the regulated practices and activities. In the cost-benefit
evaluations performed by the NRC, both the expected and the indeterminate exposure pathways
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(as appropriate for recycle or reuse scenarios) would be accounted for in the evaluations. The
dose computation models used by the licensees in seeking approval of a proposed practice in
compliance with the codified BRC dose level should be essentially identical to the NRC
cost-benefit models and would therefore be pre-approved. The basis for applying the generic
BRC dose level to a practice or activity which was not included in the rulemaking cost-benefit
evaluations is not clear, unless the cost-benefit evaluations were conducted with an excessive
degree of conservatism to account for unidentified practices or sources.
SUMMARY
Overall, the above definitions for de minimis, BRC and generic BRC are internally
consistent and consistent with the fundamental principles of radiation protection and dose
limitation. With the exception of the term BRC, the above definitions are reasonably consistent
with the use and applications of the terms by various agencies in various contexts. For the term
BRC, the concept that the dose level should be cost-benefit based has been lost in many of the
uses of term, which in some degree can be attributed to the unfortunate choice of the term
"below regulatory concern." The inclusion of the cost benefit consideration in determining a BRC
dose raises a question regarding the validity of establishing a broad-based generic BRC because
of the nature of cost-benefit evaluations. However, if the generic BRC dose level is not
established too conservatively, it could still be useful for licensees not electing to petition the
NRC under the existing BRC policy statement. The ultimate benefit of a generic BRC dose level
is that it could, in practice, become a de minimis dose level. It would appear that items such as
multiple sources and indeterminate exposure pathways can be accounted for in the
implementation portion of the regulation. For the nuclear power industry for waste disposal and
unrestricted release of materials, it would appear that the greatest benefit would accrue from
rulemakings which (1) established a de minimis dose level, and (2) implemented the BRC policy
statement by the NRC's adoption of a source-specific cost-benefit-based BRC limit for all power
plant wastes considered together as one composite. A generic BRC dose level is not likely to
provide a significant benefit for activities such as waste disposal and unrestricted release of
materials. Other nuclear power licensee's activities could have a greater benefit from the
establishment of a generic BRC dose level.
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Criteria for Release of
Decommissioned Nuclear Facilities
for Unrestricted Use
Dr. Joseph W. Ray
Battelle Memorial Institute
INTRODUCTION
A recent report by the U.S. General Accounting Office cited the need for residual
radiation criteria for nuclear facilities decommissioning. Government is emphasizing public
health and environmental protection in the operation and remediation of facilities. At the
same time, an increasing number of private sector facilities will be ready for decommissioning.
Effective and workable criteria on which to base the decision to release nuclear facilities, as
well as associated equipment and materials, for unrestricted use are clearly important. This
workshop itself lends further testimony to the need for such criteria.
Release criteria must provide suitable protection for public health and safety and the
environment. They must also be workable in the sense that they can be implemented and
monitored effectively. Today I would like to bring you a perspective on unrestricted use
release criteria from the viewpoint of an owner, operator and decommissioner of a nuclear
facility who is subject to the regulatory or policy oversight of three federal and two state
agencies, as well as county and city government units.
First I should be specific about what I mean by unrestricted use. I am using the term
literally - free from any restrictions on future use. By this I mean that a building is suitable for
use as a cafeteria or laboratory without any restrictions: that a piece of furniture is suitable for
use in an office or for donation to a day care center without any restrictions: that items of
equipment are suitable for donation to a school or for recyclable scrap without any
restrictions: that a piece of land is suitable for a playground or for residential development
without any restrictions. This absence of restrictions means no further control or monitoring,
regulatory or otherwise, over what is done with the released property.
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DECOMMISSIONING AT BATTELLE
At Battelle, we are currently dealing with the decommissioning of all or part of 15
buildings that have been used for nuclear materials research at various times over the past 46
years. The facilities are operated under a U.S. Nuclear Regulatory Commission (NRC)
license, and are also subject to the relevant U.S. Department of Energy Orders as well as
applicable regulations of the U.S. and Ohio Environmental Protection Agencies. Most of the
affected buildings are currently used for non-nuclear research, and all are expected to be
used for R&D offices and laboratories following decommissioning.
The facilities include a hot cell laboratory which was used until recently for post-
irradiation examination of power reactor fuel elements, and a former research reactor which
was partially decommissioned in 1978. The remaining 13 buildings were used for R&D on
properties and fabrication methods for uranium and thorium, as well as for research with
radiotracers.
The residual radioactivity in these buildings is primarily low levels of natural uranium
and thorium. The residual radioactivity is contained within the buildings, and no hazardous or
radioactive materials were buried or otherwise disposed of at the site. In short, this
decommissioning is far less dramatic than some of the decontamination activities being
undertaken elsewhere. However, it is probably representative of a large fraction of future
decommissioning projects.
RELEASE OF BUILDINGS AND EQUIPMENT
Annex C in fuel cycle facility licenses is essentially the non-reactor equivalent to
Regulatory Guide 1.86, and both are similar to proposed ANSI Standard N13.12. Annex C (or
Reg Guide 1.86) is extensively used as the basis for releasing buildings ~ as well as
equipment and material - for unrestricted use. A close reading of Annex C suggests that it is
intended primarily for application to a scenario involving continued non-residential use of the
building. It does not appear to be applicable to all plausible future uses of a building.
For example, consider a laboratory building with interior surfaces contaminated with
natural uranium. If this building is released for unrestricted use according to the acceptable
surface contamination levels of Annex C, then the interior surfaces of the building could
average 5000 dpm/100 cm2. A little arithmetic shows that, to a first approximation, dpm/100
crn2 is approximately equal to pCi/g for uranium and thorium in a thin surface layer. This
implies surface contamination of 5000 pCi/g, which is equivalent to about 0.7 weight percent
uranium in the surface layer. Suppose that several years after release of the building a new
owner, in converting it for office use, decides to clean up the walls and floors by removing a
thin surface layer. In so doing, he creates a particulate waste material containing a few tenths
of a percent of uranium by weight. This concentration of uranium is licensable under 10 CFR
40.13. Of course, the new owner knows nothing about uranium waste - the building was,
years before, released for unrestricted use. As a result, a few tons of licensable source
material may end up in a local landfill. Suitably treated and packaged, this material is unlikely
to pose a threat to public health or the environment. However, if handled or disposed of in a
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condition under which the waste material could become airborne, it is likely that air
concentrations of uranium in excess of the limits of 10 CFR 20, Appendix B, could occur.
Annex C also does not appear to have been generated for application to unrestricted
release of equipment and materials from the decommissioned facility. Here the principal
concern is the likelihood that the released item will find its way into use by the general public.
For example, such items may be reused by the current owner, sold or donated to other
organizations, or disposed of as scrap with subsequent recycling. The lack of control after
release together with the high probability of public contact, suggest very conservative criteria
should be used for release. The logical alternatives are either no detectible radiation above
background (generic criterion), or levels low enough that pathways analysis shows the
exposure to the general public to be below the basic dose limit (situation-specific criteria).
Clearly generic criteria are much simpler to implement and monitor. However, they run
the risk of being pegged to "worse case" and hence being too restrictive (and costly) for
many situations, or of being pegged to some "typical case" and hence not restrictive enough
for some situations. The situation-specific approach has appeal because release criteria can
be pegged to a risk-based Basic Dose Limit. The important question is "What Basic Dose
Limit do I use?" Limits ranging from 4 to 100 mrem/year above background are being
proposed by various groups. Adopting a uniform and defensible Basic Dose Limit is crucial
to any credible criteria, be they generic or situation-specific.
SOIL CRITERIA :
Generic soil release criteria for radium and thorium are available in 40 CFR 192.
Situation-specific criteria must be used for other nuclides. Typically this takes the form of a
site-specific pathways analysis which assesses the annual dose to a maximally exposed
member of the general public under an appropriately conservative scenario. Hydrological and
geological parameters are highly variable and highly site specific, and the most appropriate
future-use scenario is also dependent on the site location, ownership, and other factors. For
these reasons, the situation-specific approach is appropriate for soil release criteria.
As I noted above, a uniform and defensible Basic Dose Limit is essential to credible
criteria. For soil release criteria, a generally accepted pathways analysis model is also
important. Such models are available and others are under development. From the
viewpoint of an owner, operator, and decommissioner of a nuclear facility, the availability of a
personal computer based model which is acceptable to all cognizant regulatory bodies is
especially important. In combination with a uniform Basic Dose Limit, this would greatly
enhance the implementation of site-specific release criteria.
AIR AND WATER RELEASE
Criteria are provided in 10 CFR 20.106 for concentrations of nuclides in air or water
released to unrestricted areas. In addition, 10 CFR 20.303 provides criteria for release of
water into sanitary sewer systems. The criteria are nuclide-specific and criteria for
combinations of nuclides are provided.
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CONCLUSION
In summary, appropriate generic criteria for release of buildings and associated
equipment and materials for truly unrestricted use are not available. The prudent approach is
to use situation-specific analysis, taking into consideration specific nuclides, concentration
levels, and plausible future use of facilities, equipment and materials.
Situation-specific pathways analysis is also appropriate for soil release criteria.
Implementation of the situation-specific approach would be enhanced by the adoption of a
uniform and defensible Basic Dose Limit, and by the availability of a generally accepted
pathways analysis model.
Albert Einstein once said "Everything should be as simple as possible, but no simpler."
We should keep this in mind as we further develop criteria for the release of decommissioned
nuclear facilities for unrestricted use.
REFERENCES
[1] Nuclear Regulation: NRC's Decommissioning Procedures and Criteria Need to be
Strengthened, U.S. General Accounting Office, GAO/RCED-89-119, May 1989.
[2] Annex C, Guidelines for Decontamination of Facilities and Equipment Prior to Release
for Unrestricted Use or Termination of Licenses for Byproduct, Source, or Special
Nuclear Material, U.S. Nuclear Regulatory Commission, Division of Fuel Cycle and
Material Safety, July 1982.
[3] Regulatory Guide 1.86, Termination of Operating Licenses for Nuclear Reactors, U.S.
Atomic Energy Commission, June 1974.
[4] Control of Radioactive Surface Contamination on Materials, Equipment, and Facilities to
be Released for Uncontrolled Use, ANSI N13.12, Draft Standard, 1988.
[5] Title 10, Code of Federal Regulations, Part 40 - Domestic Licensing of Source Material.
[6] Title 10, Code of Federal Regulations, Part 20 - Standards for Protection Against
Radiation.
[7] Title 40, Code of Federal Regulations, Part 192 - Health and Environmental Protection
Standards for Uranium and Thorium Mill Tailings.
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Session V:
Recycling of Materials and Equipment
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A Research Program on the Recycling of
Decommissioning Materials at JAERI
Mitsugu Tanaka and Hisashi Nakamura
Department of Japan Power Demonstration Reactor
Japan Atomic Energy Research institute
ABSTRACT
Research on reuse of wastes from reactor decommissioning has been done at JAERI
since 1987. This paper describes a fundamental research program on reuse of radioactive metal
wastes, and discusses future plans for the development of reuse techniques as a realistic method
of waste disposal. With respect to the first of these, the melting testing program for radioactive
metal coming from the decommissioning of the Japan Power Demonstration Reactor (JPDR) is
going to reveal the material balance and the radioactivity balance during melting. The second
is at the stage of designing research programs intended to demonstrate real reuse of
decommissioning materials.
INTRODUCTION
The generation of electricity by nuclear power in Japan has increased remarkably in
recent years, with 36 commercial nuclear power plants (about 28,000 MW) in operation as of the
end of November, 1988. But, the life of a nuclear power reactor is only about 30 to 40 years.
It is predicted that the decommissioning of commercial nuclear power plants in our country will
begin in the second half of the 1990's. The report "Long-term Program on Nuclear Energy
Development and Utilization," issued by the Atomic Energy Commission in June, 1982 provided
concrete guidelines for decommissioning nuclear power plants. Because of our country's small
size and insufficient space to construct hew nuclear power plants, the report recommended, as
the basic decommissioning policy, quick dismantling (5 to 10 years) after shutdown, for the reuse
of the site for follow-on reactors. The decommissioning program of the Japan Power
Demonstration Reactor (JPDR), under progress at the Japan Atomic Energy Research Institute
(JAERI), is based on this basic policy, and aims both at the development of reactor dismantling
technologies and at the improvement of dismantling safety.
It is expected that decommissioning a nuclear power plant will generate a large volume
of dismantling waste. For example, it is estimated that dismantling a 1,100 MW nuclear power
plant will result in 500 to 550 thousand tons of waste (including 40 to 50 thousand tons of metal)
within a concentrated period of time. To carry out the decommissioning smoothly, the
establishment of a rational system for treating and disposing of the waste is essential. If we can
use the dismantling waste as a resource, this can decrease the amount of waste in a very
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meaningful way. In addition, this direction is in harmony with national policy; the reuse of
extremely low-level radioactive solid wastes is suggested in the report "Philosophy of Safety
Regulation for Land Disposal of Low-Level Radioactive Solid Wastes" submitted by the Nuclear
Safety Commission in October, 1985. But dismantling wastes from nuclear installations differ
from the usual industrial wastes, and safe and rational techniques for recycling them have not
been established yet. From this viewpoint, JAERI started its reuse research with the JPDR
decommissioning materials in parallel with investigations of the technical problems for reusing
the dismantling waste safely.as resources. The objectives of this reuse research are to establish
the techniques for reusing the dismantling waste from decommissioning nuclear power plants,
and to contribute to the discussion on technical guidelines for reuse.
PRELIMINARY SURVEY ON DISMANTLING WASTE REUSE
In order to establish a system for safe and economical reuse of dismantling wastes,
development of the uses according to characteristics (material, form, radioactivity level, etc.) of
the waste is indispensable. The results of a preliminary survey on the uses of metal and
concrete waste are demonstrated in the following sections.
Reuse of metal waste
The system (treatment technique, distribution route of treatment and recovery, etc.) for
the reuse of usual industrial metal waste is already established, because metal waste has
inherent economical value. Therefore, it is expected that the existing reuse system will function
well for the waste under an exemption level. The exemption level has not yet been determined
in Japan. Although the exemption level to be decided upon in the future may exert influence on
radioactive metal waste reuse, the development of the possible uses of radioactive metal will play
an important role. From this viewpoint, with the ease in promoting reuse taken into account, the
uses of metal waste for applications in controlled areas or non-controlled areas within the nuclear
installations have been discussed, and for use off-site, on the condition that remanufactured
articles are confined to a certain region. The results to date are summarized as follows:
Promising uses in controlled areas
- Reinforcing bars and other structural materials for building
- Tanks
- Casks, canisters and waste containers, etc.
Promising uses in non-controlled areas
- Reinforcing bars
- Pipings, etc.
Promising uses off-site
- Base materials (caissons, piles, etc.)
- Reinforcing bars
- Pipelines, etc.
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Reuse of concrete waste
As base information for discussing the uses of concrete waste coming from the
dismantling of nuclear installations, the present situation of concrete waste coming from the
demolition of usual structures was investigated. A survey was conducted on the annual
consumption and disposal of concrete in our country, the treatment and disposal processes of
concrete, and their costs. Discussions on the uses for concrete and the solutions to remaining
problems were carried out on the prospects of reusing concrete wastes. According to the
survey, the amount of concrete waste disposed of in 1986 is equivalent to the concrete wastes
from dismantling 30 to 40 1,100 MW nuclear power plants. Seventy percent of the concrete
waste was disposed of, and only the remaining thirty percent was reused. Even if concrete waste
is reused, its applications are restricted to roadbed or backfilling materials. However, it is
anticipated that in the near future, disposal means like backfilling will be difficult to use, owing
to land shortage and protection of the environment.
Although most of the uses are not characterized well, some typical examples are as
follows:
- Structures of nuclear installations (structural wall, shielding wall, sectioning wall)
- Waste disposal pit and container
- Roadbed material
- Cementation material
It is essential that the safety, economics, and quality of these applications are thoroughly
examined, and that a rational reuse system is established through the development of the
necessary technology.
MELTING TESTING PROGRAM FOR RADIOACTIVE METAL
There are two ways to reuse metal dismantling wastes. One is to reuse contaminated
equipment (pump, pipings, etc.) as it is after decontamination in other nuclear installations. The
other is to produce remanufactured articles (waste container, construction materials, etc.) by
melting the metal, and utilizing them within nuclear installations or in another place. With respect
to the latter case, the behavior of radionuclides, during melting and solidification of the
radioactive metals, must be revealed for the safety of workers and the public. But, at present in
Japan, there is no relevant data. Therefore, JAERI has started research titled "Reuse Technology
Development of Low-Level Radioactive Wastes", performed under contract with the Science and
Technology Agency since 1987. In this research, tests of the melting of radioactive metal waste
from the dismantling of the JPDR, and of other materials, are planned. The objectives of these
melting tests are to investigate and assess the movement of radionuclides during solidifying
processes and melting, and its influence on the environment. The schedule of the program is
shown in Fig. 1.
Experimental equipment
The production of the experimental equipment to be used in the melting tests of
radioactive metals will be completed by 1990. The schematics of the melting equipment and the
flow of work procedures are shown in Fig. 2.
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The melting equipment consists of a melting furnace, casting equipment, sampling
apparatus, filtering equipment, radiation monitors, etc. A high frequency induction furnace (500
kg capacity) was considered most appropriate for the kind of work to be undertaken. Some of
its positive characteristics are:
a) Adequate homogeneity of radioactivity in melted metal
b) Feasible melting of carbon steel and stainless steel
c) Little secondary wastes
d) High reliability, etc.
The melting furnace will be surrounded by a containment chamber to minimize the spread
of contamination by radionuclides during operation. The filtering system to collect dust is
composed of a cyclone, bag filters and a High Efficiency Particulate Air (HEPA) filter.
Contents of testing
Fundamental melting tests with radioactive metals will be performed as a function of
materials, condition of radioactivity (activated, contaminated), etc. Accordingly, three types of
tests are planned: cold tests, tracer tests with radioactive isotopes (Rl), and actual waste tests
with JPDR decommissioning materials. Data on the materials balance, radioactivity balance, and
dose rate in the working environment during the melting and solidifying processes will be
collected, as well as the operation characteristics of the experimental equipment. The tests are
summarized below.
(1) Cold test
A cold test serves as a trial run of the melting equipment and provides an understanding
of the materials balance. Operational data on melting equipment (relation between melting time
and temperature, off-gas temperature, etc.) and on the movement of materials (adhesion to the
equipment, production of slag and dust, etc.) will be collected.
(2) Tracer test with Rl
The evidence suggests that some metal wastes are contaminated with radionuclides
which are produced both by the activation of the base metal and impurities in structural materials
and by fission products arising from the breakage of fuel rods. The aim of this test is to
investigate the behavior of these radionuclides in the melting and solidifying processes.
This is difficult to examine using JPDR decommissioning materials because of its
decreased radioactivity. For this reason, the amount of radionuclide movement from melting
metal to solidified metal, furnace wall, slag, pipings and off-gas will be measured in commercially
available metals coated with radioisotope tracers that imitate the nuclides found in contaminated
material of the light water reactor. The radionuclides to be used are Mn-54, Co-60, Zn-65, Sr-85,
Cs-137, etc. In addition, the radiation dose rates on the equipment surfaces and in the working
environment will be measured, to obtain radiation control data.
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Testing parameters and assessment items are as follows.
1) Testing parameter
a) Materials (carbon steel, stainless steel)
b) Radioactivity
c) Condition of radioactivity (activated, contaminated)
d) Nature of flux
e) Melting temperature
2) Assessment items
a) Material movement from charged material to ingot, slag, off-gas, etc.
b) Movement of radionuclides to ingot, slag, off-gas, etc .
c) Radioactivity distribution within an ingot
e) Radiation dose rate in working places
f) In-air radioactivity of working places
(3) Actual waste test with JPDR decommissioning materials
Contaminated and activated metals such as pipings, valves, and a piece of the pressure
vessel from decommissioning the JPDR will be melted. By these melting tests, the safety of
melting real radioactive metal will be demonstrated.
FUTURE PLAN
The key to success in decommissioning a reactor is to treat and dispose of large volumes
of dismantling waste. In the future, establishing better ways of treating and disposing of waste
will become increasingly important. In particular, the economics of treatment and disposal of
waste, and environmental protection from waste, will be regarded as major problems. Therefore,
we must find a way to solve these problems, from a long-term standpoint, irrespective of the
types of waste. The reuse plans for the future are described below and illustrated in Fig. 3.
Production and application of remanufactured articles
Before reusing dismantling waste, whether metal or concrete, it is very important to
examine the safety of both the manufactured items and the manufacturing processes. The
necessary data, however, does not exist. Therefore, integrated research is planned for both
metal and concrete wastes. In the research, feasibility for restricted or unrestricted use of
remanufactured articles will be investigated through the production of such articles. For metal
waste, a large scale melting facility will be constructed, taking into account the results obtained
from the fundamental melting test described earlier. All the processes (melting to forming) will
be tested to produce high value-added remanufactured articles (waste containers, etc.) using the
metal wastes of the JPDR. As to concrete wastes, remanufactured articles like construction
materials made from radioactive and non-radioactive concrete waste or composites with metal
will be produced. Also, confirmation testing, intended to examine the performance of these
remanufactured articles, will be carried out. Furthermore, buildings such as waste storage
facilities will be constructed using some of waste to confirm its actual integrity.
261
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Effective utilization of radioactive waste
Activated metal waste generally must be stored, owing to its high radiation dose rate. This
makes treatment and disposal difficult. However, if the radiation energy emitted from the waste
is used effectively, the radiation can become a benefit. For example, radiation energy is used
widely for chemical reactions, disinfection, food preservation, etc. In addition, utilization of
radiation is being tried as a means to purify the environment, such as purification of exhaust
gases, decomposition and removal of contamination in sewage water; and disinfection and
composing of sewage sludge. Activated metal waste could be utilized effectively in the fields of:
- Disinfection of sewage water, medical apparatus and agricultural products
- Purification of exhaust gases, etc.
Accordingly, the feasibility of reusing highly activated metal waste as a radiation source will be
investigated, including the related forming techniques and processes, equipment, and economy.
CONCLUDING REMARKS
JAERI has started research on reuse of decommissioning material. In this paper, the
fundamental reuse research program of metal waste is described, together with the future plans
including production and application of remanufactured articles. Reuse technology may be an
alternative means of disposing waste. The radioactive metal melting test program in progress
is expected to bring a great deal of useful fundamental data on recycling metal waste. Similarly,
plans described here utilize the vast volumes of dismantling wastes from a decommissioned
reactor. "
REFERENCES
[1] Ishikawa, M., Kawasaki, M., Yokota, M., Ezure, H., Hoshi, T., and Tanaka, M., Present
status of JPDR Decommissioning Program, Proceedings of the 1987 International
Decommissioning Symposium, Westinghouse Hanford Company, Vol.1 ,1987, p. 111-18.
[2] Tanaka, M., Yanagihara, S., Ishikawa, M., and Kawasaki, M., The Japan Power
Demonstration Reactor Decommissioning Programme, Proceedings of the International
Conference on Decommissioning of Major Radioactive Facilities, Institution of Mechanical
Engineers, 1988, pp. 25-31.
262
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Fiscal Year
'87
'88
'89
'90
'91
'92
'93
1.
Melting of
radioactive
metal
2.
Assessment for
decommissioning
material reuse
Design
of
melting
furnace
Cold test
Construction
of melting
furnace
Assessment
Hot test
Safety, Economy, Technical Problems, etc.
Fig.l Schedule of radioactive metal melting testing
-------
Containment chamber
-p-
1
n - '
i Q j
. Slag container
^>
Drum
^ C(
Cyclone Bag filter
Radioactivity
B)
1
1
» fi r
1_1 U
©
\ !
Cutting machine
Decommissioning
Slag container
waste ,
L Container I
[ High frequency
,induction furnace
Air curtain
High frequency
power source
Casting
Casting mold
Cooling water
Blower
To existing
ventilation
system
HEPA filter
Off-gas treatment
\) . Constituents
Measurement and Analysis
Fig.2 Schematic diagram of melting equipment and flow of work
-------
High efficiency filter
Off-gas treatment system
Wave breaker, etc.
Steel box filled with concrete
Fig.3 Flow diagram of decommissioning material reuse
Range of fundamental
research
-------
Effects of Residual Radioactivity in Recycled
Materials on Scientific
and Industrial Equipments
Shohei Kato, Hideaki Yamamoto, Shigeru Kumazawa and Takao Numakunai
Department of Health Physics
Japan Atomic Energy Research Institute
ABSTRACT
Health effects, as well as scientific and industrial effects, should be assessed in
establishing residual radioactivity criteria for recycled materials. In the present study, a
comprehensive literature survey has been carried out on the effects of residual radioactivity on
scientific and industrial equipment. To develop a method for assessing the scientific and
industrial impacts of recycling materials, an investigation also was conducted on measures to
reduce these effects, on the trend in rates of production of the equipment, and on the progress
of the technologies.
The kinds of equipment most likely to be affected by residual radioactivity are large scale
integrated circuits (LSI), photographic films, and low-background radiation counters. Procedures
for assessment of the effects of residual radioactivity on these have been suggested. Recycling
of contaminated materials is expected to bring both gains and losses. An optimum residual
radioactivity level was proposed as an index for the assessment of the effects. The optimum
residual radioactivity level is defined as the level at which the difference between economic
benefits and costs from recycling of contaminated materials is maximum. We discussed the
relationship between the optimum residual radioactivity level and a residual radioactivity level that
will be derived to produce no significant radiological hazard.
INTRODUCTION
The decommissioning or remodeling of nuclear reactor, nuclear fuel facilities, particle
accelerators and radioisotope handling facilities generates large amounts of slightly radioactive
material and equipment. The establishment of residual radioactivity criteria (RRC) which enable
us to reuse or recycle such material and equipment is absolutely essential for the exploitation of
extremely low-level radioactive wastes as potential resources, as well as for the reduction of their
volumes.
266
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The RRC will ensure that the health effects will be negligibly small from recycling
contaminated materials. When is it sufficient to consider only health effects?
Modern science and technology have been and will be making remarkable progress to
change our society. There are various types of high technology industry that can use only
extremely high quality materials. But the probability of contaminated materials effecting such
technologies is expected to increase.
For example, the trend in large scale integrated circuits (LSI) is towards higher levels of
integration. However, a reduction in memory cell size, with a corresponding reduction in signal
charge, leads to increases in the sensitivity to noise. May and Woods [1 ] recognized that alpha
particles from radioactive impurities in the packages could cause a soft error in dynamic RAM
and CCDs. Therefore, the LSI industry has been seeking extremely high quality material as a
constituent material.
This phenomena indicates that scientific and industrial effects as well as health effects
should be considered in developing RRC. Assessments on health effect from residual
radioactivity in recycling material have been conducted by the Commission on European
Communities (CEC) [2] and the U.S. Nuclear Regulatory Commission (NRC) [3]. However, an
assessment of scientific and industrial effects has not been carried out.
In assessing health effects from radiation quantitatively, there are many procedures to be
investigated; analysis of exposure pathways, identification of critical pathway and development
of assessment models. These same procedures should also be investigated when assessing
the industrial effects from residual radioactivity.
This is a pilot study on the effects of residual radioactivity on scientific and industrial
equipment. Its purpose is to obtain background information to establish the residual radioactivity
criteria. The study was started from 1988 as a three-year program.
1988: Investigation of the effects of residual radioactivity on scientific and industrial
equipment.
1989: Development of a calculation code for assessment of the scientific and industrial
effects of recycling contaminated materials.
1990: Assessment of the effect on possible scientific and industrial products by using the
calculation code.
In the present paper we describe the effects of residual radioactivity on scientific and
industrial equipment and the development of an assessment method.
THE EFFECTS ON SCIENTIFIC AND INDUSTRIAL INSTRUMENTS
With unrestricted recycling, contamination levels in the material may be very low, so it is
important to consider the effects from slight contamination.
267
-------
The rapid progress of science and technology will increase the possibility of residual
radioactivity affecting the scientific and industrial fields, in particular, new technologies such as
LSI, and high-sensitivity photographic film.
The investigation of the effects was conducted to cover all kinds of instruments. The
investigation was focused on the following items: phenomena and mechanism of the effect,
occurrence rate of the effect, occurrence rate from other causes, measures to reduce the effects,
and the trend of production.
Large Seals Integrated Circuit (LSI)
(1) Effects of radiation on LSI
Radiation effects on LSI consist of the single event, and the total dose effect.
Furthermore, the single event consists of soft error and lutchup.
Total dose effect:
Ionizing radiation produces stored charge and oxide interface states on the chip surface.
MOS technologies can seriously be affected, since they involve surface effect devices. The major
effects are a shift in threshold voltage of the transistor, and an inversion of p-type regions under
field oxides. This effect is a semipermanent phenomena and can be evaluated in terms of the
cumulative exposure which causes the failure of LSI function. The dose level depends on the
type of device and integration density. The cumulative dose which leads to the effect is from 103
to 104 rad for nMOS-DRAM, nMOS-SRAM, CMOS-SRAM [4, 5, 6].
Lutchup:
Lutchup is a parasitic phenomenon occurring in CMOS structures due to feedback in the
p-n-p-n structure. Lutchup occurs when currents caused by radiation in the substrate, or well,
forward bias a diffusion-background junction. Once turned on, the current flow will continue until
power is removed or until the device is damaged. Lutchup is induced by irradiation of high
energy cosmic rays. The error rate of lutchup is one hundredth to one thousandth of soft error
rate [6, 9].
Soft error:
Soft error is the upset of stored data by the passage of alpha particles through the
memory array area. There are two modes of upset in LSI, memory cell mode and bit-line mode,
as shown in Figure 1. In the cell mode, an upset is caused by noise electrons flowing into a
storage capacitor. In the bit-line mode, an upset is caused by noise electrons flowing into the
n+ diffused layer in the floating bit-line in a read cycle.
The relation between soft error rate (SER) and alpha flux indicates that the SER is
proportional to alpha flux, as shown in Figure 2 [1,8]. This figure also indicates that the soft error
rate increases as the integration density of LSI increases. Figure 3 shows that devices with
smaller critical charges exhibit higher SER [1]. It is known that soft error is scarcely induced by
beta-rays or gamma-rays [1].
268
-------
From the above review, it becomes clear that the soft error is the most important effect
induced by radiation emitted from contaminated recycling material.
(2) Factors affecting soft error and reduction techniques
SER depends on critical charge, cycle time, power supply voltage, integration density of
LSI and flux, energy, injection point and angle of alpha particle [7,11,12,13].
The information densities of new generations of DRAM have increased four-fold every 3
years since the introduction of a 1k DRAM in 1974 [14]. 64 M DRAM devices are expected to
be produced by the year 2000. As mentioned above, the higher integration density tends to have
small critical charge, which easily causes soft errors in a DRAM. These trends indicate that soft
errors will be very important problem in future.
Many techniques to reduce the soft error rate have been developed. Information on the
reduction techniques is essential to estimate soft error rate and the cost of the reduction. Many
measures such as reduction of alpha-ray producing impurity in the material, overcoat, error
detecting, error correcting codes, and improvement in the design of the circuit or structure of LSI
are being developed [7,15].
(3) Materials used in LSI
Figure 4 shows the cross sections of three types pf JC packages: the stacked ceramic
type package, the pressed hermetic ceramic type, and the plastic molding type [16]. The LSI
parts likely to be contaminated with recycling materials are the ceramic (AI2O3), cap (Fe-Ni), steel
(Au-Sn) and lead (Au-gild) for the static type, and ceramic and lead for the pressed hermetic type,
and lead for the plastics type.
The major metallic materials used in LSI are Al, Au and Cu.
(4) Trend of production
Information on production and application of LSI in society is important to assess the
economic impact of recycling contaminated material. Figure 5 shows a,trend in the yield of
DRAM in terms of cumulative bit number of devices [16]. The yield of 64k DRAM, reached a
maximum in 1984. Thereafter, the production of 265k DRAMs increased. Today, the 1M DRAM
is a major device. The number of each device changes year by year, but the logarithm of the
total number of bits of a DRAM increases in an approximately linear fashion, by a factor of 1,000
times every ten .years. The logarithm of the number of DRAMs also increases as a linear function
of the year, at a rate of 10 times every 10 years.
Figure 6 shows industrial fields using ICs, and the cost fraction of an IC [13]. In Japan,
an application fraction of ICs for industrial products is large. The demand for ICs from office
automation systems, like personal computers, is expected to increase in the future. The fraction
of IC cost to total cost of the instrument for all instruments is expected to increase. For example,
the fraction for an automobile will increase from 2.4% in 1985 to 8.1% in 1990.
269
-------
The ORRL is an index in establishing the RRC. For example, no net benefits are obtained
from Case 3. In Case 1, recycling material contaminated with radioactivity at the level derived
from health effects may be advisable. In Case 2, the ORRL may be desirable for recycling.
The ORRL can be estimated using system engineering methods such as non-linear
planning. A calculation code is under development to estimate the ORRL and to assess
quantitatively the profits and costs from recycling of contaminated materials generated from
decommissioning.
CONCLUSIONS
In establishing RRC, not only the health effects but also the industrial effects from residual
radioactivity should be considered. Many assessment methods of health effects have been
already developed by the CEC, NRG, and others. However, the assessment method for industrial
effects from residual radioactivity has not been established. In the present study, a literature
survey on the industrial and scientific effects from residual radioactivity has been carried out to
identify the effects and to develop an assessment method.
The potential fields in which scientific and industrial effects occur from residual
radioactivity in recycling material are LSI, photographic films, and low-background radiation
counters. Furthermore, countermeasures to reduce the effects, trends of production rate, and
technological advances were surveyed.
Based on this information, an optimum residual radioactivity level (ORRL) is proposed as
an index for the assessment. An assessment code to estimate the ORRL and to evaluate the
scientific and industrial effects quantitatively is under development. Using this code, the industrial
effects from residual radioactivity can be assessed in establishing RRC for both unrestricted and
restricted recycling.
ACKNOWLEDGMENT
This work was primarily suggested by Mr. Allan Richardson, Office of Radiation Programs,
U.S. Environmental Protection Agency. We would like to acknowledge Dr. Kentaro Umeda,
Nuclear Application and System Analysis Co. Ltd., for valuable contributions to the literature
survey. We would also like to thank Mr. Masao Aono, Fuji Film Co., and Dr. Noboru Shiono,
Nippon Telegraph and Telephone Public Co., for fruitful discussions.
REFERENCES
[1 ] T.C. May and M.H. Wood, A new physical mechanism for soft errors in dynamic memory,
in Proc. of the 16th Int. Reliability Physics Symp. 1978, p. 73; also IEEE Trans. April 1979,
ED-26 (1) pp. 2-9
[2] Commission of European Communities, Radiological protection criteria for the recycling
of material dismantling of nuclear installation, Radiation Protection No. 43, CEC, 1988
272
-------
[3] I.R. O'Donell, et al, Potential radiation dose to man from recycle of metals reclaimed from
a decommissioned nuclear power plant, NUREG/CR-0131, 1975
[4] G.W. Hughes, et al, Radiation hardened MOS technology, Solid State Technology 22,7 pp.
70-76, 1979
[5] N. Shiono, etal, Radiation effects on LSI, OYOBUTSURI 55, pp. 243,1986
[6] D. M. Long, State-of-the-art review: Hardness of CMOS and bipolar integrated circuit. IEEE
Trans. On Nucl. Sci. NS-27, pp. 1974, 1980
[7] P.M. Carter, et al, Alpha particle induced soft error in NMOS RAMs: a review, IEEE
Proceedings 134, pp. 32-44,1987
[8] S. Hirai, Determination of Uranium, Thorium and alpha emitters in semiconductor,
BUNSEKI9, pp.639, 1988
[9] W.A. Kelasinski, et al, Simulation of cosmic ray induced soft error and Lutchup in
integrated circuit computer memories. IEEE Trans. Nucl. Sci., NS-26 (6) pp. 5087-5091
1979
[10] K.K.M.M. Haque, et al, Soft error rate in 64k and 156k DRAM., Electron Lett., 22, pp.
1188-1189,1986
[11 ] D.S. Yaney, et al, Alpha-particle tracks in silicon and their effect on dynamic MOS RAM.,
IEEE Tran. on Electron Device, ED-29, pp. 10-16,1979
[12] T. Toyabe, et al., A soft error rate model for Mos dynamic RAM. IEEE Trans on Electron
Devices ED-29 pp. 7302, 1982
[13] The latest trend of semiconductor market, KAGAKUKOGYO 51, pp. 343-347, 1987
[14] K. Shimohigashi, et al, Alpha-particle-induced soft error in semiconductor memories
[15] K. Kudo, et al., Uranium, Thorium and alpha emitters in LSI constituent materials.,
RADIOISOTOPES 31, pp.490-499, 1982
[16] Data and Chart of LSI., NIKKEI ELECTRONICS 1987 11,16 pp. 239-240, 1987
[17] M. Aono, Private communication, 1988
[18] S. Takada, High sensitivity photographic films, KOTAIBUTSURI, 21, pp. 835-842, 1986
[19] P. J. Donald, et al., The effects of background radiation on the keeping quality of X-ray
films., Photographic Science and Engineering 6, pp. 212-215, 1964
[20] K. Kurakado et al., Quasiparticle excitation in a superconducting tunnel junction by
alpha-particle., Physical Review B22, 1 pp. 168-173,1980
273
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[21] J.O. Willis, et al, Radiation damage in YBa2CU3O7 by fast neutrons., LA-UR-87-38033,
DE88.003157 ~ " "
[22] H. Hanafusa, et al., Drawing condition dependence of radiation-induced loss in optical
fiber., Electron Lett. 22, 2 pp. 245-258,1987
[23] H. Hori, et al., Electron irradiation induced amorphyzation at dislocations in NiTL, Jpn. J.
Appl. Phys. 22, 2 pp. L94-L96, 1983
[24] J.L. Brimhall, et al., Th amorphous phase transition in irradiated NiTi alloy., Radiation Effect
90, pp. 241-258, 1987
[25] H. Yamaoka: Recent studies radiation effect of organic insulation., Genshiryoku-shi 26,9,
pp.739-747, 1984
[26] T. Kinoshlta, Radiation damage of ceramics, defect kinetics and phase stability under
electron irradiation., Genshiryoku-shi 28,11, pp. 1009-1014,1986
[27] A. Tominaga, et al., Measurement of Travelling Time of Blast Furnace Burden with Co-60,
Tetsu to Kou 45, 689,1959.
274
-------
• POTENTIAL WELL FILLED
WITH ELECTRONS
. P-TYPE SILICON IN
"INVERSION"
.-ONE MILLION ELECTRONS
'0'
- 5 MeV« PARTICLE
• POTENTIAL WELL EMPTY
• P-TYPE SILICON IN "DEEP"
DEPLETION
-1.4 MILLION ELECTRON HOLE
PAIRS GENERATED TO A DEPTH
OF-25/1
•POTENTIAL WELL REMAINS
FILLED
•NO APPRECIABLE
COLLECTION
•POTENTIAL WELL NOW
FILLED
Soft error due to cell contribution
Source: T.C. May and M.H.Wrod (1978)
1-0'
'O1
a-ray
Memory
Cell
Sense
Amplifier
V
e~
Dummy
Cell
'0'—'1
Soft error due to bit line contribution
Figure 1 Mechanism of soft error in dynamic
memory.
-------
10
10'
t_
10"
Alpha-Flux (a/cmh)
10~2 1 102 104 10b
Concentralion of U, Th (ppb)
10fc
Figure 2 Error rate versus alpha flux, and
concentration of U and Th.(Source: Hirai,
1988)
Q}
O
O
DC
o
UJ
10
x-2
A64K CCD
I6K CCD
16K- MOS RAM
MOS RAM
0 1x10b : 2x10b 3x10fc
Critical Charge (#Electrons)
Figure 3 Error rate versus critical change
for devices from several manufactures.
(Source: T.C.May and M.H. Woods, 1978)
276
-------
Cop / Coval + Au, \
^Fe-Ni, A1203'
Sheal (Au-Sn, Gloss
Lead (Au,Cu, Coval)
Ceramic (A1203)
Static Ceramic Type
Ceramics (AL203)
Lead (Au, Cu, Coval
Ceramics (AL203)
Press Hermetic Type
Plastics
Lead (Au, Cu,-Coval
Plastic Molding Type
10
15
CD
*- 10
14
ce
Q
CD
JQ
E
10
13
2= 10
m
12
o
H—
O
K* Total Bit Number '
Total Number of DRAM
264 K DRAM
64KDRAM
1MDRAM
10'
10
cc
o
a>
JQ
E
10° -
1980 1985 1990
Year
1995.
Figure 5 Trend of the production of DRAM
in the world. (Source: NIKKEI ELECTRONICS, 1987)
Figure 4 Package types and the constituent
materials. ('Source: K. Kudo et al. 1982)
-------
oo
1(W 570 1,752
_0
o
20-
0
4,520 (108yen)
'80
-^Auto mobile
Watch, Camera
Desk Computer
(VTR, VideodlsiQ
Stereo, Taperecorder
Home Computer
Electronic Register
Measurement Eg.
•(Computer Eg.
Medical Instruments
Communication Eg.
Others )
{
'85
'90 (Year)
Trend of Application of 1C to Industrial
Goods
Trend of Ratio-of the Cost of 1C to Total.
Cost of the Good
Figure 6 Application of 1C to industrial
: goods. (Source: .KAGAKUKOGYO No.51, 1987)
-------
1600;
400
~ 200
£ 100
^ 40
"I 20
| 10
QJ
80 4
2
1
o o
OOO
coo
00 Q,
00
"CD COv COO
0.,-0-b
- Q.-
r
o o
1940-1945 1950 1955 1960 1965 1970 1975' 1980 1985
Year .
Figure 7 Trend of sensitivity of
photographic negative color film. (Source:
S.Takada, 1986)
o
D±
0.40
0.30
0.20
0,10
0
INDUSTRIAL
X-RAY
MEDICAL
X-RAY
PORTRAIT
ROLL
FILM . CLASSIFICATION .
1.0
0.8
0.6
0.4
0.2
0
co
o
o_
CO
LU
or
o
to
i
o
o
LU
LU
CC
I
Figure 8 Comparison of normal storage fog
glowth per year of five essentially,
different photographic films, with their
corresponding response to Co-60 gamma
radiation. (Source: D.P.Jones et al, ,1964)
279
-------
Decommissioning
Scfap materiols
Constituent
Materials of LSI
Dilution
, Measures
Uncontominated
Materials
LSI
Soft Error
, Measures
Electronic
Equipments
C Damage
Figure 9 Flow diagram of the effect of
recycling materials on LSI
Residual radioactivity level
/derived from health effecs
Case 1
Residual
Radioactivity
Case 3
Figure 10 Optimum residual radioactivity
level(ORRL) in recycling of contaminated
materials and the relationship between ORRL
and the residual radioactivity level derived
from health effects.
280
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Development of International Exemption
Principles for Recycle and! Reuse*
W. E. Kennedy, Jr.
Pacific Northwest Laboratory
Richland, Washington, USA
ABSTRACT
For the past several years, the International Atomic Energy Agency (IAEA) has been
investigating the possibility of exempting certain radiation sources and practices from
regulatory control. Initial efforts were conducted under the general heading of de minimis.
Interest in this topic arises from international recognition that a significant fraction of the
wastes from industry, research, medicine, and the nuclear fuel cycle are contaminated to such
low levels that applying regulatory controls may be unwarranted. The IAEA evaluation has
been conducted by Advisory Groups considering two interrelated topics: to establish"
principles for exemption, and to apply the principles to various areas of waste management.
The IAEA Advisory Groups developed the criteria from modeled assessments of the potential
radiation exposure pathways and scenarios for individuals and population groups following
the unrestricted release of radioactive materials. Although the scenarios and models used by
the IAEA are necessarily generic, consultants to the Advisory Groups attempted to identify the
most important radiation exposure pathways based on available literature. This approach is
intended to provide the basic framework for the numerical derivation of generic exempt
quantities that would be conservative in most situations. In their evaluations to date, the IAEA
Advisory Groups considered: 1) disposal of exempt wastes in a sanitary landfill, 2) disposal
of exempt wastes by incineration, 3) recycle of contaminated steel, aluminum, or concrete,
and 4) reuse of concrete buildings, tools, or equipment. This paper discusses the current
status of the IAEA's efforts on the subject of exemption and presents the generic results
expressed as overall exemption limits for municipal wastes and materials for recycle and
reuse.
The International Atomic Energy Agency (IAEA) has been investigating the possibility
of exempting certain radiation sources and practices from regulatory control as an extension
This work was supported in part by the International Atomic Energy Agency and in
part by the U.S. Department of Energy under Contract DE-AC06-76LO 1830.
281
-------
of its earlier work in the area of de minimis. Because of the potential value of scrap materials
recovered during decommissioning of commercial reactors, and because of national and
international efforts to minimize radioactive wastes, exemption criteria for recycle and reuse
have gained attention. The IAEA has established basic principles for exemption that limit the
radiation dose that individuals or population groups may receive from exempted practices or
sources.
This paper discusses the recent IAEA Advisory Group's recommendations on
principles for radiation practices and sources in the recycling of retired components and
materials from nuclear facilities. The background of the Advisory Group's work is discussed,
then its methods and preliminary recommendations are summarized. Finally, a similar effort
sponsored by the Commission of the European Communities is summarized and compared
to the IAEA approach.
DEVELOPMENT OF IAEA ADVISORY GROUP RECOMMENDATIONS
As the result of the recent IAEA Advisory Group efforts in exemption, in 1987 the IAEA
published interim general principles for exemption of sources and practices that may result in
both individual and collective doses of very low significance (IAEA 1987a). The principles
were quite general and applicable to any type of manmade radiation source that may give
rise to trivial risks. They did not, however, apply to natural sources of radiation. The Advisory
Group concluded that a trivial risk from radiation exposures would be in the range of 10 to
100 nSv (1 to 10 mrem) per year. Further, the IAEA interim exemption principles were
intended to provide a safety margin for selected individuals who may be exposed to radiation
from several exempted sources and to account for the uncertainty of future human activities.
With this safety margin in mind, the IAEA recommended that the individual doses from a
single exempted source or practice should not exceed 1% of the existing individual dose limit
for members of the public, or 10 nSv (1 mrem). This dose equivalent is less than 0.5% of the
annual effective dose equivalent from natural background radiation and is small compared
with the natural variation in background radiation. For skin doses, the IAEA Advisory Group
recommended a dose limit of 1% of the existing limit, or 500 nSv (50 mrem) (IAEA 1987a).
The 1987 IAEA Advisory Group statement also recommended the control of collective
dose. This would provide additional assurance that many small doses would not total a
significant dose. It would also guard against the possibility that exemption could occur
without the knowledge of controlling authorities. This is especially important for sources that
are exempt and, therefore, not subject to notification and registration. The Advisory Group
recommended that, as part of the "basic case" for exemption, the collective effective dose
equivalent commitment from the exempted source or practice be on the order of 1 manSv
(100 manrem) or less (IAEA 1987a). The 1987 interim guidance attempted to account for all
potential dose arising from an exempted source or practice over all time. National authorities
could exempt sources that give rise to larger collective dose commitments, but could also
establish a condition below which no further consideration need be given to the radiological
basis for exempting a source. The Advisory Group considered that sources and practices
complying with the conditions relating to individual and collective dose may be exempted
from the normal regulatory requirements of registration and notification, and treated as if no
radiation exposures were involved.
282
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The IAEA received numerous comments on the interim exemption principles from the
participating countries. Many of the comments related to the proposed collective dose
criterion, which normally was used only as a tool in the optimization process. Applying this
criterion was considered to be overly restrictive and was limiting for many of the cases under
review. Another major issue was the definition of a practice. This definition is directly related
to the collective dose criterion, which creates a problem, because the numerical criterion
takes on a very different significance depending on how a practice is defined. For example, if
the practice of exemption for recycle means application to the material from a single reactor,
the derived limits would likely be quite different than those derived where the practice is
defined to cover the material potentially recycled from all reactors in a country.
Because of these concerns, it became evident to the IAEA that further discussions
were needed if a firm international consensus on exemption principles were to result. In
response, the IAEA called an Advisory Group meeting in Vienna, Austria, in March, 1988, to
modify the interim statement. The Advisory Group was co-sponsored by the IAEA and the
Nuclear Energy Agency (NEA) of the Organization for Economic Co-Operation and
Development (OECD). The Advisory Group prepared a guidance document that recommends
a policy on exemption (Linsley and Salo 1988). The IAEA policy was later published as Safety
Series Report No, 89 (IAEA 1988). This policy calls for the system of notification, registration,
and licensing prescribed in the IAEA Basic Safety Standards for Radiation Protection (IAEA
1982). Safety Series No. 89 allows exemption on a generic or case-specific basis. The
revised definition of a practice for exemption purposes is "a set of coordinated and continuing
activities involving radiation exposure which are aimed at a given purpose, or the combination
of a number of similar such sets" (IAEA 1988).
i
The Advisory Group further clarified the basis for exemption by identifying two basic
qriteria to determine a candidate for exemption: 1) radiation protection must be optimized (as
shown through a cost/benefit analysis) and 2) individual risks must be sufficiently low. In
determining a trivial level of individual dose, a trivial risk level must be chosen, and average
natural background levels should be considered as a reference level. The Advisory Group
concluded that most authors identify trivial levels of risk to be in the range of 10"6 to 10~7 per
year (IAEA 1988). Using a risk factor of 10"2 per Sv (10"4 per rem), a trivial dose range of 10
to 100 [iSv (1 io 10 mrem) per year results. Comparison with natural background gave a
trivial effective dose equivalent range of 20 to 100 jiSv (2 to 10 mrem) per year. The Advisory
Group concluded that, regardless of the source, a trivial level of dose could be assured if it
djd not exceed the order of some 10s of [iSv per year (IAEA 1988). Accounting for the
potential for exposure to many individual exempted sources, critical group doses on the order
of 10 |iSv (1 mrem) per year would be reasonable.
The Advisory Group further concluded that optimization of protection must be
considered analogously to the requirements in the IAEA basic safety standard (IAEA 1988).
The optimal level of protection is achieved when the next spending level exceeds the health
detriment averted. Using the IAEA minimum value of $3,000 per manSv in 1983 dollars (IAEA
1985), a source-earelated trivial collective dose for exemption on the order of a few manSv
results. For continuing practices, the Advisory Group interpreted that a commitment of about
1 manSv (100 manrem) per year of practice would be reasonable. This revised approach
accounts for the 50-year dose commitment per year of practice instead of the dose
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commitment over all time. The Advisory Group's acceptance of the concept of optimization
allows a previously prohibited degree of flexibility if individual risks are appropriately low.
DESCRIPTION OF GENERIC IAEA METHODS
The main objectives of additional IAEA Advisory Group efforts to develop generic
exempt quantities have been:
i . •,
• to illustrate a methodology for developing practical radiological criteria through the,
application of the IAEA preliminary exemption principles,
• to establish generic criteria, and
• to determine the practicability of the preliminary exemption principles (IAEA 1987b).
Exempt quantities, expressed in units that relate to radiation-detecting instruments, are
a more practical expression of the general exemption principles. The steps used by the IAEA
Advisory Groups in deriving exempt quantities for a defined source or practice are:
• to establish a series of radiation exposure scenarios that account for various exposure
pathways and conditions,
• to estimate the resulting radiation doses to individuals and population groups for these
scenarios,
• to determine the limiting (highest dose) scenario for each radionuclide, and
• to determine the concentration of individual radionuclides that would result in the
exemption criteria (dose limits).
, .•(,••',.
In assessing the radiation doses, the IAEA Advisory Group advised careful selection of
parameters, assumptions, and data (IAEA 1987b). For their assessments, the IAEA Advisory
Group and their consultants judged scenarios on the likelihood of their occurrence leading to
human exposure and on the likely magnitude of those exposures. In addition, estimating the
potential exposure of a critical population group was necessary.
In their evaluations to date, the IAEA Advisory Groups considered:
• disposal of exempt wastes in a sanitary landfill (IAEA 1987b);
• disposal of exempt wastes by incineration (IAEA 1987b);
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**
• _ recycle of contaminated steel, aluminum, or concrete ; and
• reuse of concrete buildings, tools, or equipment.**
Because of the potential value of scrap materials recovered during decommissioning
and because of national and international efforts to minimize wastes, exemption criteria for
recycle and reuse have gained recent attention.
The radiation exposure pathways included in the generic dose estimates were external
exposure to penetrating radiation, inhalation of airborne material, and ingestion of
contaminated foods or removable surface contamination (through secondary transfer from
hands to the mouth). A variety of representative radionuclides were considered to explore
fully the radionuclide-dependence of the resulting, exemption limits. These radionuclides were
chosen to represent alpha emitters (239Pu and^41Am), high-energy photon emitters (60Co),
low-energy photon emitters (^Fe), and pure beta emitters f°Sr and 99Tc).
The potential radiation exposures resulting from different generic scenarios have a
probability of occurrence that may range from zero to one. Thus, the IAEA used their
collective judgment to select scenarios for the derivation of exempt quantities. They paid
most attention to those scenarios where individuals could have direct contact with the
radioactive materials. These scenarios included workers at landfills, incinerators, smelters, or
recycle centers, and consumers who may use recycled materials or who reuse released
buildings, tools, or equipment. Additional scenarios, such as use of ground water near a
landfill or release of volatilized material through the stack at a smelter, were also included to
provide an estimate of the likely collective dose. .
IAEA RESULTS AND DISCUSSION
Example results for the groupings of reference radionuclides and the various types of
exemption considered by the IAEA Advisory Group are summarized in Table 1. The results
are presented in terms of reasonably expected ranges, based on the various radionuclides in
each group and on the expected variation among exposure scenarios. Control is lost in the
fate of exempted materials through unrestricted release. Thus, material exempted for recycle
or reuse could be disposed of in a landfill, or material exempted to a landfill could be re-
cycled or reused. Because of the lack of future control, the proposal has been made that a,
single exempt quantity should be established that would cover all alternative future
conditions, without specifying limitations for landfill disposal, incineration, recycle, or reuse
(Kennedy et al. 1988). This overall limit appears to be possible because of a close grouping
of the results shown in Table 1 across most radionuclide and unrestricted release categories.
**
As described in a draft working document on 'The Application of Exemption Principles
to Wastes from Decommissioning and Recycle of Materials from Nuclear Facilities," by
the International Atomic Energy Agency, Vienna, Austria, 1988.
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TABLE 1. Example Exempt Quantities for Various Exemption Categories
Alpha
1-10 0.1-10
0.1-100 0.5-10
Exemption Category
Sanitary Landfill
(Bq/g)
High-Energy Low-Energy
Photon Photon Pure Beta
Emitters Emitters Emitters Emitters
Incineration
(Bq/g)
Recycle^
(Bq/g)
Building Reuse
(Bq/cm)
(a)
(a)
300-600
io*-io3
1 -10
1 -5
Reuse of Tools and 10-100
Equipment (Bq/cm)
1-10 104-105 40-300
0.004-1 10-100 60-500
10-100 60-500 102-103
(a) No radionuclides were considered for the scenarios shown.
(b) For recycling of steel, aluminum, or concrete rubble.
Three potential groupings of released material are: 1) mass concentrations (in units of
Bq/g), 2) surface contamination in buildings (in units of Bq/cm), and 3) surface contamination
on reused tools and equipment (in units of Bq/cm). Further, it appears that a set of
radionuclide groupings could be made by combining the high-energy photon emitters and
alpha emitters into a single grouping across all categories. The overall exemption limits
resulting from such re-grouping are summarized in Table 2 (Kennedy et al. 1988).
Again, ranges of values are shown to denote the potential variations of radionuclides
and exposure conditions. For mixtures, the sum-of-fractions rule could be applied. The net
result for mixtures is that the limit is controlled by the most restrictive radionuclides present. It
may be noted that for all scenarios considered, the individual dose criterion is limiting in
relation to the exempt mass concentration and surface contamination limits. However, the
collective dose criterion js the determining factor in assessing the total quantity of material
that may be buried, incinerated, or recycled. The IAEA work will continue because further
studies are needed to determine whether additional practical considerations (including costs
and detectability) will change results or basic conclusions.
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TABLE 2. Preliminary Exemption Limits for all Release Categories
Limit Category
Mass Concentration
(Bq/g)
Building Surfaces
(Bq/cm)
Reuse of Tools and
Equipment (Bq/cm)
High-Energy
Photon and
Alpha Emitters
1 -10
0.1 -1.0
101-102
Low-Energy
Photon and Pure
Beta Emitters
102-103
102-103
102-103
OTHER INTERNATIONAL EFFORTS
In parallel with the IAEA efforts over the past few years, European countries have
adopted national policies regarding exemption. In 1988, the Commission of the European
Communities (CEC) published Radiation Protection No. 43: Radiological Protection Criteria
for the Recycling of Materials from the Dismantling of Nuclear Installations (CEC 1988). This
document is a summary of the recommendations from a group of experts set up under the
terms of Article 31 of the Euratom treaty. The CEC identified two approaches to the
establishment of radiological protection criteria for the recycling of materials from the
dismantling of nuclear installations. One is based on setting acceptable individual and
collective dose levels (similar to the approach adopted by the IAEA), and the other is based
on setting clearance levels for the activity concentration of the materials considered so that
the radiological consequences of recycling are insignificant from a health protection point of
view. The CEC concluded that the second approach, setting clearance levels, had the
advantages of regulatory simplicity* practicability, and consistency and afforded the same
level of health protection as the first approach (CEC 1988). For more complicated cases, it
would still be possible for national authorities to carry out a case-by-case assessment.
Because of these considerations, the CEC adopted the clearance-level approach; however,
they also provided that the IAEA dose limits would be met for specific cases where alternative
criteria would need to be established. In addition to the IAEA individual and collective dose
limits/the CEG also considered the IAEA transportation safety regulations limiting the surface
activity for materials.
The group of experts to the CEC considered the results of a study on radiological
protection criteria for recycling carried out by an ad. hoc working party. This working party
considered radiation exposure scenarios for steel scrap and equipment from dismantling
retired nuclear power reactors. The recommended clearance levels for recycling steel scrap
and equipment are summarized in Table 3. It should be noted that the 0.4-Bq/cm2 limit for
non-fixed beta-gamma emitters is derived from the IAEA transportation safety regulations.
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TABLE 3. Clearance Levels Recommended by the CEC for Recycling
Steel Scrap and Equipment
Type of
Radiation
Beta-Gamma
Alpha N/R®
Mass Activity
Clearance Level
1 Bq/g averaged over
a maximum mass of
1000 kg(a)
Surface Level
Clearance Level
0.4 Bq/cm2 for non-
fixed contamination on
accessible surfaces^
0.4Bq/crh2(b)
(a) No single item may exceed 10 Bq/g.
(b) Averaged over any area of 300 cm2 of any part of the surface.
(c) N/R means that no value is recommended because of potential detectability problems
and because most alpha activity is assumed to be present as surface activity only.
SUMMARY •-
The IAEA's individual dose criterion is selected based on a consensus of what
constitutes a trivial risk and a comparison with natural background. The collective dose
criterion established by the IAEA is intended to serve as the basis for an optimization
assessment that can help justify exemption decisions. The IAEA is in the process of
developing exempt quantities for recycle arid reuse using a radiation exposure scenario
analysis. The exempt quantities will be useful to illustrate a methodology for developing
practical radiological criteria, establish generic criteria, and help determine the practicability of
the preliminary exemption principles.
As an extension of the IAEA exemption principles, the CEC has derived clearance
levels that define the activity concentration of the materials to be recycled, based on a
generic radiation exposure scenario analysis. Both the IAEA and the CEC efforts indicate that
national authorities have the flexibility to modify the overall approach to account for special
situations that may require additional regulatory actions.
REFERENCES
[1] Commission of the European Communities (CEC). 1988. Radiological Protection
Criteria for the Recycling of Materials from the Dismantling of Nuclear Installations.
Radiation Protection No. 43, Luxembourg.
[2] International Atomic Energy Agency (IAEA). 1982. Basic Safety Standards for
Radiation Protection, 1982 Edition. IAEA Safety Series No. 9, IAEA, Vienna, Austria.
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[3]
[4]
[5]
[6]
[7]
[8]
International Atomic Energy Agency (IAEA). 1985. Assigning a Value to
Transboundary Radiation Exposure. IAEA Safety Series No. 67, IAEA, Vienna, Austria.
International Atomic Energy Agency (IAEA). 1987b. fxempf Quantities of Low-Level
Radioactive Wastes for Disposal to Municipal Landfill or by Incineration: Methods for
Their Derivation and Generic Values. Part II of Exemption of Radiation Sources and
Practices from Regulatory Control - Interim Report. IAEA-TECDOC-401, IAEA, Vienna,
Austria.
International Atomic Energy Agency (IAEA). 1987a, General Principles for the
Exemption of Radiation Sources and Practices from Regulatory Control. Part I of
Exemption of Radiation Sources and Practices from Regulatory Control - Interim
Report. IAEA-TECDOC-401, IAEA, Vienna, Austria.
International Atomic Energy Agency (IAEA). 1988. Principles for the Exemption of
Radiation Sources and Practices from Regulatory Control. IAEA Safety Series No. 89,
IAEA, Vienna, Austria. .
Kennedy, W. E., Jr., C. R. Hemming, F. R. O'Donnell, and G. S. Linsley. 1988.
Application of Exemption Principles to Low-Level Waste Disposal and Recycle of
Wastes from Nuclear Facilities. In Radiation Protection Practice, Vol. Ill, Proceedings
of the Seventh International Congress of the International Radiation Protection
Association, Sydney, Australia.
1 " • ' • • ' :'"..• ' < " •
Linsley, G. S., and A. Salo. 1988. Exemption from Regulatory Control -International
Developments. IAEA-CN-51/04, Presented at the International Atomic Energy Agency
International Conference on Radiation Protection in Nuclear Energy, Sydney, Australia.
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Summary & Panel Discussion
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Summary & Panel Discussion
Moderator: Anthony B. Wolbarst (EPA)
Session chairmen: Donald A. Cool (NRC)
Masao Oshino (JAERI)
Allan C.B. Richardson (EPA)
Andrew Wallo, III (DOE)
DR. WOLBARST: The panel will be soliciting comments and questions from the audience.
But before that, we are scheduling something extra which will take a few minutes, but which I
think will be of general interest.
In his opening address, Rich Guimond stressed the importance of effective risk
communication in dealing with the public. Risk communication is an issue that is taken seriously
by us at EPA. We have established a Risk Communication Program, for example, involving
people with expertise in the subject, to provide assistance and guidance for programs with public
relations problems. And we have set up a two-day course on risk communication that 10 percent
of all EPA staff eventually will take.
An important contribution to this effort is a book, "Improving Risk Communication,"
recently published by the National Academy of Sciences (NAS). This document was funded
partly by EPA and several other Federal agencies. The primary staffer at NAS who handled it
was Rob Coppock. So I am going to ask Rob to take a few minutes to describe the book, and
perhaps tell us how we can get it.
DR. COPPOCK: Rich Guimond pointed out that in the rule-making process it is important
not only to provide the technical means to ensure protection, but also to instill public confidence
in it. And that is very much what this report, "Improving Risk Communication," is about.
Yesterday I saw a memorandum signed by Secretary Watkins of the Department of Energy
that is going to the heads of operating units, divisions, and programs in DOE, accompanying
copies of this report. Secretary Watkins stated that he intends to make improved risk
communication a Department priority, and asks for their response as to the changes in
Much of the tape recording was barely audible, and we had to do a fair amount of heavy
editing. Hopefully there are no serious mis-representations. The Editors.
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procedures they would expect to implement as a result of the contents of the report. That is a
pretty powerful statement and endorsement from his Office.
Also, Terry Davies, the Assistant Administrator at EPA for Policy and Evaluation, is
distributing about 250 copies of the book within EPA. The Toxic Substance Disease Registry,
the Public Health Service organization responsible for preparing the hazard profiles at Superfund
sites, purchased enough copies of the book for all its field offices.
So this is a document, I think, that you will be seeing around from time to time. The
reviewers have stated that they expect it to be the standard reference in the area.
I want to take 2 minutes to summarize this 200 page document, and to point out a couple
of the conclusions that are relevant to the issues we have been discussing here.
First of all, the book emphasizes that there is no single risk communication problem, so
there is no simple solution. There is no silver bullet. Further, improvement will take much effort
over many years, not months. We are in it for the long haul.
Secondly, improving risk communication will not necessarily smooth management, making
things easier. In fact, it may make things more controversial. Many of the issues of concern are
ones on which people have different opinions, use different vocabularies, hold different values
regarding the possible outcomes. That being the case, clarifying the issues may make people
realize that they have very strong positions, and are in conflict. Helping people understand the
issues may increase the level of conflict. The report states that improving risk communication
may make management issues more difficult. But we are obliged to make things clearer, even
if it makes our jobs more difficult.
Lastly, the report points out that both technical competence and communication skills are
needed in communication programs. You will shoot yourself in the foot if you get either one
wrong - there needs to be a balance.
Good communication skills need to be applied early in the life of a project. Very often
the communication and presentation problems can only be solved by adjusting analysis and
management steps along the way. Unless you have included what people are worried about in,
your analysis, you may present your conclusions, and find them rejected because they don't
include the things that people are really worried about. If people do not see the issues that they
believe to be important in the analysis, they are likely to dismiss the whole analysis.
The book is available, and can be ordered from:
National Academy Press
2101 Constitution Avenue, NW
Washington, D.C. 20418
paperbound $29.95
hardbound $39.95
(5-24 copies 15% discount, 25-499 copies 25% discount)
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. DR. WOLBARST: Thank you, Rob. We are running a bit late, so I feel we should start the
Panel discussion. Let's begin with Andy Wallo, of DOE, who chaired the session on "Impacts of
Cleanup Technologies and Economics on Criteria".
MR. WALLO: We have been asked first to summarize the sessjon that we chaired, and
so I will mention the papers that were covered in the session on impacts of clean-up
technologies and economics on criteria. There were seven informative papers.
The first three papers covered limitations of clean-up technologies. They talked about
procedures and methods of decontamination, particularly of portions of facilities, buildings, and
surfaces, and for measuring the levels of residual radioactivity.
A fourth paper discussed the processing of low-level radioactive waste, and criteria for
release. This one looked mainly at reducing volume, again on wastes from buildings and
decontamination projects going for disposal.
The fifth paper discussed the disposal capacity and projected waste volumes within the
compacts, and another described the Superfund volume reduction process for specific sites in
New Jersey.
The final paper, on residual radioactivity cost impacts evaluation looked at various options
for controlling contamination and for decontamination standards, again of structures, and at the
effects they might have on the cost of cleaning up the structures. The costs of these will be
impacted by the criteria, the options selected for decontamination - whether it be
decommissioning a building or totally destroying and removing it. Cost could be impacted by
the availability of a BRC limit and the amounts of the wastes.
I think an important message in Bob Dyer's paper on volume reduction was that, for soil
criteria clean-up, we have two options: either remove the contaminated soil or leave it in place.
Volume reduction was looked at on a very site-specific basis. It was for one radionuclide and
one site, and would have to re-evaluated again every time one wanted to do it. It is not really
a generic process for volume reduction; its applicability depends on the criteria that apply for
cleanup, and the dose levels.
It is clear that both the actual clean-up and the measurement processes are going to be
impacted by the criteria selected, and that a site must be characterized before the remedial
action. When you are deciding on your final cleanup standard, you really want to have a
complete characterization of the sites.
It is essential, at the end of a project, to document the clean-up and the procedure used
in verifying or certifying that the site meets standards. Measurement techniques and detectable
limits vary, and the standards and associated certification procedures should be tied to the risks
and benefits of the project. In almost all instances where statistical sampling procedures are
used for certification, another survey could measure something that is not measured with the
procedure used. While these differences in results are not likely to have any significance in the
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overall risk associated with the site, they could cause questions in the validity of the sources.
A certification process has to be tied to a procedure for measurement.
And again, the issue of whether BRC standards affect clean-up, and whether volumes can
be averaged over mass, or surface contamination averaged over mass, can impact decisions and
the costs of clean-up.
Some of the discussion that we might want to follow up on concerned the general
tendency in a lot of States or communities to say that anything that ever was touched with
radioactivity is radioactive waste. We have seen that in a number of our DOE sites. Some
barrels, dug up near one of our sites, were filled with sludge and petroleum-based residues that
were actually associated with an airport construction action. An analysis was done on the
wastes, and some radium and uranium showed up. It was clearly part of the petroleum-based
material, and was not radioactive waste. But the analysis showed radioactivity, and the headlines
said that this was radioactive waste that approaches or exceeds NRC unrestricted use standards.
So here we have a solvent or a petroleum-based material that nobody is going to drink. But we
are looking at drinking water standards or unrestricted release standards and calling it radioactive
waste, because it has this taint of radioactivity in it. Clearly, the hazard associated with this
material was chemical.
DR. COOL: The last session that we had yesterday afternoon was entitled "Health Effects,"
although it ran the gamut on a number of different issues.
The first paper by Bill Ellett talked about health effects predictions and analysis. It
provided us with a clear view of the very unclear view we have of radiation risks. Although we
have a lot of information, there are great uncertainties. And the doses that are involved with
residual radioactivity assessments, and the levels that we are actually finding, are well below the
levels where effects have been demonstrated. In fact, the risks are likely not to be measurable
in the population with the statistical power we have in the groups available.
Mr. Yamamoto provided a paper on JAERPs experience in decontamination and reuse of
a large-scale radiochemical laboratory, indicating some of the criteria that were actually applied
in a live situation and some of the doses actually resulting. One of the things that impressed me
was how low -the doses were when they were finished with the decontamination, on the order
than less than a tenth of a micro-Sv. That corresponds to an extremely low level of potential
health effect and risk.
We then had a couple of papers dealing with exposure models, the first by Don Lee on
applied exposure modeling for residual radioactivity, and the second by Andy Wailo on the DOE
guidelines.
Modeling is an area that we end up using whether we like it or not, because we cannot
empirically know all of the things that are going to happen prior to actually releasing the site.
The point was made that modeling can be supportive of rulemakings, compliance determinations,
and research, and that those three activities have very different goals and require very different
types of models.
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Site-specific models can provide an accurate means for dealing with case-by-case
determinations, but they have a potential drawback of discrepancies between sites, and between
different actions at different times. The alternative to that is generic models. These are useful
in rulemakings by a regulatory agency, like the NRG, doing a regulation which would involve a
release for unrestricted use. But models of that type are not very realistic when you start looking
at the details of an individual site. A model which is generic is not going to represent the western
desert of the United States very accurately, just as it is not going to represent the eastern United
States very well'. The appropriate level of detail may.be difficult to achieve with a generic model.
And one of the underlying themes in this conference is that for models to be useful, they need
good input data. And that requires an accurate characterization of sites before anything else is
done.
MR. RICHARDSON: The next session was a monstrous one, with 12 papers. I hope I
don't neglect any of the major points of the various authors, because we have so many to talk
about.
Rob Coppock started it off, pointing out that there are four kinds of national regulatory
styles. He called them evidential, consensual, authoritative, and corporatist. What I found most
interesting is that even though all these four styles may exist, all of the countries end up with
basically the same kinds of answers to radiation problems. They all use the same basic decision
criteria. So the conclusion for me is that it does not matter how you get to the decision. If you
use the right principles, you will probably end up with pretty close to the right decision.
After Rob's paper, I reviewed the basis for such decisions. They are the old radiation
protection principles we nave had with us for many years. There are two new factors that are
important, however. One is national (in this country) mandates under legislation for Superfuindi
for ground water protection levels. The other is more conceptual, the need for source-related
categorization of the individual dose limit. You cannot use the whole 100 millirems for residual
radioactivity. Then I raised the issue of whether we should be looking for dose limits or for
concentration and quantity limits for our actual criteria.
Vern Rogers carried that much further in his paper, and suggested categories of nuclides
- a number of other authors did the same thing. He also emphasized the importance of
individual dose scenarios and their large uncertainties. c
Mr. Yamamotb and Mr. Oshino reviewed the Japanese situation. They are using some
of the IAEA recommendations for exemptions, and are at least headed in the direction of using
those same criteria for residual radioactivity standards. (The IAEA exemption criteria are on the
order of a millirem (0.01 mSv) for individual dose, and 100 persons - rems (one man-Sievert) for
collective dose.) They are not using the collective dose criterion.
Lynn Wallis described the decommissioning of Shippingport. I think the most important
thing that came out of that experience is the finding that they would pass most of the criteria that
have been discussed as site release criteria, and are well below the levels of residual radioactivity
that they might have to meet. None of these levels have been set yet, of course, but
Shippingport would have passed almost any of the proposals. So it can be done. One of .the
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interesting sidelines of the paper was the discussion of how ALARA was implemented at that site,
and the importance of social factors.
Stan Neuder discussed methods for deriving surface contamination limits, Reg Guide 1.86,
and associated problems. A number on papers touched on the difficulties with Reg Guide 1.86.
Clearly, there is lots of work to be done on surface contamination criteria.
Tim Johnson and Don Cool described NRG activities. The NRC is moving ahead
aggressively with initiatives based on their legislative mandates for low-level waste, and also on
initiatives for "below regulatory concern" or "exemption levels", depending on what lexicon you
use. Their exemption policy, however, is at variance with that of the international community.
It is an order of magnitude higher - 10 millirems per year, as currently proposed. Similarly, the
criterion that was proposed for exemptible collective dose - a bottom line, as it were, for when
you have to look at ALARA in detail - is a factor of five higher than the IAEA recommendation.
So that is something that, I am sure, will be a discussion item in the future.
Steve Adams talked about the tortuous history of ANSI N13.12. I was glad that we had
his presentation, because perhaps the regulated community, after having gone through that
exercise, has gained some perspective on why regulatory agencies take so long to get the
answers out. And when they are finally but, why they are generally perceived by the public and
the industry as arbitrary and capricious. [Laughter]
Bill Holcomb described EPA's draft proposed low-level waste standards. These standards
have two significant new things in them. One is that for the first time they reflect national ground
water protection policies, which the current standards do not. Also they reflect EPA's (at least
up to now, EPA's) approach to BRC matters, and that is to do it on a case-by-case basis and
feel our way along on the question of BRC.
The last two papers, by Jim Vance and Joe Ray, dealt with BRC. Jim distinguished three
related concepts: de minimis, BRC, and generic BRC. He pointed out the difficulties with generic
BRC. They are not unrelated to the kinds of difficulties that caused. EPA to be cautious in
approaching BRC, and described their petition on BRC for specific waste treatments.
I was interested in the extremely low population impact of the proposed exemptible waste
streams. Three hundred man/rems is only about a 10 percent chance of one cancer in 10,000
years.
Joe Ray also talked about the problems of Reg. Guide 1.86 for uses down the line, and
possible alternate uses for decommissioned facilities. He pointed out the need for basic dose
limits and accepted pathway models.
To summarize the whole session, at least for this listener: We don't have the answers
with respect to residual radioactivity criteria, and we need them. There seems to be general
agreement on the need for a dose limit, if not as a single standard, at least as a guidepost that
guides all the other standards.
This dose limit could possibly be BRC, and we need to explore carefully the possibilities
for implementing such a limit in a sensible way. Alternatives that have been suggested are
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standardized models, pathway models, or standardized results based on conservative scenarios
expressed in terms of concentration levels.
MR. OSHINO: The last session covered the recycling of materials and equipment. In this
session, four papers were presented. The first, by Mr. Nakamura, concerned a research program
on the recycling of decommissioned materials at JAERI. The decommissioning of JPDR started
to demonstrate real reuse of materials. Such a program is now in progress.
The second paper was on the effects of residual radioactivity in recycled industrial
materials. Mr. Kato presented this paper, and concentrated on the effects of residual radioactivity
on LSI and photographic films.
The third, by Mr. Kennedy, was on the development of international exemption principles
for recycle and reuse on the sites. He presented IAEA's exemption principles.
The last one, by Dr. Lichtman, discussed DOE's practice of using the National
Environmental Policy Act, until cleanup criteria are established, to decide recycling issues.
DR. WOLBARST: Thank you. I want to thank the chairmen for their summaries, and open
the proceedings for discussion, comments, questions.
QUESTION: Would you say that EPA people here today have come away from this
workshop with a consensus or some conclusions on the next steps, and if so, what are your
reflections on this?
DR. WOLBARST: I think it is not a matter only of what EPA people think, and I would like
to ask all the members on the panel to respond.
MR. RICHARDSON: We need a dose number, and we need a way to implement it.
We also need to explore the question of generally acceptable models. That is going to
be a key to getting numerical criteria. There are a number of models out there now, and one of
the things that is on my list is to see if we cannot get some agreement on some common agency
models from EPA, NRC, and DOE.
And we should explore the idea of BRC as a basis for numbers. It is not clear to me that
BRC is the answer. One of the problems with it is that there are going to be some obvious
situations where BRC limits cannot be reached, not the kinds of BRC limits that the general
public can accept. So we have to look at those situations and see how we are going to handle
them.
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One alternative is a relatively high individual dose limit, high compared to the BRC, at
least, and the application of ALARA. That is the current DOE approach, as I understand it, but
that needs to be tested, too.
DR. COOL: The general approach which is being pursued is to try to establish some
dose numbers (although we do not necessarily know what the numbers are going to be) and
from there move to generic criteria that would be useful in screening situations and site-specific
criteria, to allow for individual licensees or users who wish to be more sophisticated. This
approach seems to have a great deal of support around this room.
This feeling may be coming more from the discussions we have had over breakfast, lunch,
and dinner than from the papers, but I have heard a.great deal of support for the general
approach which the NRC is taking, which Japan is taking, and others, in setting up a structure
which starts from the basic dose and risk situation and then moves to values which can be
implemented, measured, and complied with on a site-specific basis.
So from my perspective (and not necessarily that of the NRC), I come away from this
conference with some encouragement that we are moving at least in the same general direction.
MR. WALLO: I was glad to hear what Allah Richardson said about BRC. I did have a bit
of concern when we started talking about BRC. Bill Holcomb talked about it with respect to low-
level waste, and we then automatically looked at applying a low-level waste BRC to
decontamination and decommissioning of structures.
I think we must be careful when we consider BRC or some other number plus ALARA.
BRC is basically, or should be, I think, a case-specific or source-specific determination. A generic
BRC for everything does not seem to me to be a thing that we can sell very readily. BRC for low-
level waste should be different than BRC for cleanup. Ratios of individual risks and population
risks (collective dose) are likely to differ. We need a BRC approach to balance these.
When you are doing decontamination and decommissioning, you are left with residual
radioactivity. You have soil and buildings and equipment that you have to clean up, and those
may be disposed of in a landfill, or released for reuse or to be salvaged, and you have no control
over them once they are gone.
These are all very different scenarios (exposures associated with the site being cleaned
and exposures associated with the waste disposal), and they may not necessarily have the same
BRC risk associated with them. It may not be possible to get soil criteria for remedial action
down to the same level that you might get releasable, recyclable material.
So I think we want to be careful, at least in the beginning, in considering BRC. As an
alternative, we could look at an upper limit, as Allan suggested, and then try to do site-specific
analyses. That is another option. I think it is good to step back and look at all of those before
we decide on where we are going.
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The screening criteria approach is always interesting, and it is desirable to have a
screening level so that you can say, 'If I am below this, I am okay, and if I am not, then I have
to do a site-specific evaluation, so that I can clean up to meet a limit or get below it.' The
drawback we are faced with is that those screening numbers many times get written in concrete.
Can we afford to set screening levels effectively and still offer the alternatives of doing site-
specific evaluations where it is necessary?
COMMENT: In preparing the closure plan for our low-level waste sites, a technique like
what you just described was used. We used the environmental impact statement and 10 CFR
61, where it gave a generic scenario for both a humid and a dry site, and then defined within
those scenarios what would be allowable limits for contamination on the buildings, equipment,
and the soil. If your contamination levels are at those concentrations or below, you don't have
to go through any further site analysis; you can go ahead and release your site for closure.
However, if you have greater than those concentrations, then you do further analysis. So there
is one instance where there are dose limits and derived concentrations, and with just a few
criteria you could compare your site with a standard.
MR. WALLIS: I might mention a problem we have run into several times. We closed the
cooling towers down at a site in California, and because there wasn't any BRC number we could
use, we simply had to declare everything low-level waste, although we knew it was not low-level
waste. The cooling tower material was eventually sent to conventional burial ground; it was not
declared LLW. At Shippingport, we had a lot of asbestos, and we knew it was not contaminated,
but we had no choice but to declare it low-level waste.
So there is a need for BRC in those kinds of situations. And without it, you cannot take
any chance at all. You just simply declare it low-level waste, We are talking about large
volumes.
DR. LEE: I have a whole list of things to comment on. I heard the call for a generally
accepted model. As a modeler, I am very sensitive to the fact that the scenarios and the
approaches must be agreed upon, but I think that, as part of a competent risk management-risk
communication process, the burden of proof is always on the person attempting to do the study
or to run the model. That requires adequate documentation of all steps; adequate QA of the
model and its development, adequate QA of each of the scenarios and adequate QA of the
simulations that are run. This is necessary to build the credentials of the results in a manner
necessary to have them believable both to the technical community and to the public.
I think that selecting a single model in some regards may be a way to streamline that, but
it may also have the pitfall of trapping everybody into having a single model that may be
inadequate for specific situations, and I would caution that I am not an advocate of a single-
model idea.
DR WOLBARST: Do you support the idea of the Federal model that we raised a couple
of years ago? We held a meeting of a number of agencies at EPA to discuss the possibility.
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DR. LEE: Modelers have a variety of different attitudes. I view most models as being a
collection, sort of a bunch of little models put together to form one giant black box. I am not an
advocate of that approach because I think that the individual components need to be tailored
to match the physical or chemical situation encountered. If you have one black box, it is very
difficult to prepare a site-specific analysis that can accomplish that.
I think the message that I like, that I heard over dinner last night, revolves around the risk
communication points. Maybe it would be simpler to try to communicate a single exemption
number that would cover BRC and all types of releases. That is not to ignore the fact that might
justify separate practices at different levels of risk because of the associated cost. My point of
view, rather, is to look at the assignment of different levels of risk for separate practices very hard
before you attempt to do it. In saying that 4 millirem is okay for drinking water and for BRC
waste, and 10 millirem is okay for recycle/reuse, it may be very difficult to communicate the basis
and the health risk impacts of each of those kinds of practices, sources or activities. It may be
better to have a single level that is communicated across the board in a uniform manner. Allow
exemptions for whatever X millirem equals for waste stream A or waste stream B or recycled
material components D or E, and let the analysis determine the screening levels or the derived
criteria based on a single dose standard.
MR. RICHARDSON: One of the pitfalls that I think we need to be careful to avoid is to set
the BRC level so high that the public does not accept it as BRC. That is a real danger. If you
do the arithmetic and calculate the risk factor for a number like 20 milNrem, you come out with
risks that are on the order of 10"3, and nobody is going to accept that as a tolerable risk.
So if we are going to have something which will be accepted by the public as being a
trivial risk, so that we can throw the stuff away and walk away from these little tiny bits of
contamination, it has got to be truly trivial.
On the other hand, if we choose numbers that are truly trivial, there are going to be cases
that will not fit, that it will not make sense to clean up to that level. I think the problem that is
before us is to figure out how to deal with those situations in a rational way.
One of the alternatives that we have not talked about at all is various forms of institutional
controls. Another possibility is an exception clause where we have to dig too deeply into the
public pocketbook: 'It is not worthwhile doing this, we cannot afford to do it.'
There are a variety of ways to handle these problems, but I think we have to recognize
that there are two classes of problems: One class of problems is getting public acceptance for
below regulatory control, and the other class of problems is dealing with the hard cases.
COMMENT: I want to pick up on that, if I may. De minimis is one thing; below regulatory
concern is another, and there are still some controls associated with below regulatory concern
as viewed in some way. It is not an uncontrolled situation. The phrase is a terrible phrase, as
we all have said. I don't know how to communicate to the public when we are using a phrase
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that they will not be able to understand. Perhaps we need to do some work on that in its own
right to help communicate what we are really trying to achieve with a phrase of that nature.
From the electric utility perspective, we are concerned about the large volumes of
materials that would normally go to a disposal ground that do not have high levels. We are not
interested, necessarily, in the recycled part. Our interest is more the volume aspect, and being
able not just to eliminate any control, but to control it in a way that is acceptable.
Andy Wallo expressed some concern about a situation in which you take material to a
sanitary landfill, and have to go back and clean up the sanitary landfill at some later time if the
ground rules change. We have certainly been exposed to that under the Superfund area. That
would suggest that maybe we don't go to sanitary landfills, but maybe we do have to have other
options. ,
COMMENT: If you think about the whole subject and these last 2 days of discussion and
of trying to set either exemption or BRC levels, it runs counter to what people have viewed as
prudent health physics judgment, if you will, associated with control of potentially radioactive
materials.
If you cannot measure the contamination oh thb surface of a component that you want
to release, or you have such a large bulk of material that it is very difficult to certify it as not being
contaminated, the prudent health physics practice has been to declare it low-level waste,
independent of the volume or the cost or whatever, and send it to disposal. So what we are
attempting to do now is to reverse what in the past we have called this prudent health physics
judgment, so as to allow a different regulation to come into effect, to reduce the volumes of
material and make decisions that may in fact run counter to historical practice. Of course, there
are going to be problems in implementing it, and of course there are going to be these questions
about 'Well, gee, does this mean you have to go and clean up sites that you are going to have
a tough time identifying from the records that are pretty much non-existent? Will we always be
stumbling over new sites to be cleaned up?' And the answers are likely to be 'Yes'. We are
going to have to commit to some additional remedial action kind of program if this is going to
fly.
DR COPPOCK: As the most extreme "outsider" in the room, not being from the health
physics profession at all, I am rather astonished that you think it possible to dismiss sites that
had been dealt with in some way in the past. With regard to hazardous chemicals, we have
embarked on revisiting old decisions. Just because it was the best we could do 20 years ago
does not mean it is the best we can do today. We ought to treat old hazards with no lesser
standards of safety than we treat new hazards today.
COMMENT: With respect to depending on Superfund to solve the future problems that
we are seeing, faced with triple damages under Superfund, I am sure that the prudent decision
would probably be to work with the NRC to try to resolve it without having to go through that
kind of process.
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DR. NEUDER: I was going to ask one question, but let me combine it with another. I was
going to ask whether particularly the EPA group had thought any more about whether standards
or guidance were appropriate for these kinds of situations.
Now I got led back to something that we also thought about a few years ago that was
just brought up. Grandfathering is a major issue. If you set stringent requirements now, it is very
difficult to keep people from the demand for grandfathering. It is very hard to write words that
will convey the notion that it does not pay to go back to many of these sites, that it is close
enough; let them alone. I am not talking about ones that are lousy or where things were not
addressed that we have subsequently found to be hazardous, but rather where the radioactivity
was addressed, and we just cleaned up at one level, and we would all, by consensus, agree that
it does not make sense to remount the effort to bring it down to a new level.
So, unless you can solve that problem, that becomes a compelling argument for
guidance, because guidance can deal with things like this, very different from standards. In
particular, guidance can work the way DOE is. You can set up a 100-millirem limit, and you can
say do ALARA and take account of the following kinds of things, because any of these old sites
will comply, almost all of them, comply with a 100-millirem limit, and so they don't raise this issue
of compelling, of creating an urgent situation on something that doesn't demand it.
COMMENT: I cannot see a situation in which a site was delivering - let's take the worst
case -, in perpetuity, 100 millirems a year to anybody who happened to live on it, not getting on
the National Priority List and having to be cleaned up to some arbitrary level, in the absence of
more specific criteria than that kind of guidance would provide.
COMMENT: I think the point Stan was trying to make is that there are a lot of these little
sites out there that maybe, if we were doing it today, we would have cleaned up a lot neater and
left in a lot nicer package, but on the other hand, it is not worth a lot of effort to go back and
spend big bucks to even characterize or look at these sites if you have enough documentation
in the file to really let them know that we are in the noise-level type of situation.
COMMENT: The federal risk management process is a process by which regulatory
decisions and actions are made by balancing a number of factors. Those factors include the
health risk potential of the situation, the costs, regulatory compliance, the achievability of
technology with keeping people working, and all of these things are weighed and a decision is
made in support of a licensing or a regulatory decision.
Now, in the case of an old site, it may be that there is high uncertainly about the actual
potential of that site to cause a public health risk. But there may be well-documented factors like
the risks of workers digging up old sites that may become a compelling argument in favor of the
decision to stabilize it in place, rather than dig it up and move it again. Of the potential
alternatives for dealing with that situation, the worst may be to dig everything up, repackage it,
and move it someplace else - and something lesser, like stabilizing it in place, may be in fact
an acceptable risk management-based regulatory decision.
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COMMENT: If you go back and look at the records at old sites to evaluate whether,
because of new and stricter standards, perhaps they need to be cleaned up, you will notice that,
invariably, all the records are going to show that the measurements were less than whatever the
standard is now.
COMMENT: There are maybe a couple of factors that we have not discussed much, but
which impact on all of this. We talked about 4 mrem per year as being different from 10, as
being different from 20, as being different from 1. And I am not at all sure that we can support
the differences between those numbers when we actually go out and do surveys, simply on the
basis of counting statistics, let alone on the basis of the differences in the risks. We talk about
people understanding what 4 means. I am not convinced I know what 4 means. Does anybody
know what 4 means?
We would all admit that we don't know what the absolute risks are for these dose ranges,
which are very small increments over 100 millirems. Whatever the risk of 100 millirems, I think
we can agree that the risk of 104 millirems is 4 percent bigger, and the risk of 101 millirems is
1 percent bigger, and so on.
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