JAERI-Conf
95-015
PROCEEDINGS OF THE SECOND WORKSHOP ON
RESIDUAL RADIOACTIVITY AND RECYCLING CRITERIA
JOINTLY SPONSORED BY
THE UNITED STATES ENVIRONMENTAL PROTECTION AGENCY
THE OFFICE OF RADIATION AND INDOOR AIR,
AND
THE JAPAN ATOMIC ENERGY RESEARCH INSTITUTE
November 9-11,1994, Tokai, Japan
July 1995
(Eds.) Hideaki YAMAMOTO and John A. MACKINNEY*
Japan Atomic Energy Research Institute
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°- Hi,
(T319-11
(=F3i9-n
This report is issued irregularly.
Inquiries about availability of the reports should be addressed to Information Division,
Department of Technical Information, Japan Atomic Energy Research Institute, Tokai-
mura, Naka-gun, Ibaraki-ken 319-11, Japan.
© Japan Atomic Energy Research Institute, 1995
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JABRI-Conf 95-015
Proceedings of the Second Workshop on Residual Radioactivity and Recycling Criteria
Jointly Sponsored by the United States Environmental Protection Agency,
the Office of Radiation and Indoor Air, and the Japan Atomic Energy Research Institute
November 9-11, 1994, Tokai, Japan
(Eds.) Hideaki YAMAMOTO and John A, MACKINNEY*
Department of Health Physics
Tokai Research Establishment
Japan Atomic Energy Research Institute
Tokai-iura» Naka-gun, Ibaraki-ken
(Received June 5, 1995)
On November 9 to 11, 1994, Japan Atomic Energy Research Institute and the Office of
Radiation aid Indoor Air of the United States Environmental Protection Agency together
sponsored the second workshop on residual radioactivity and recycling criteria in
Tokai-mura, Japan. This volume of proceedings includes the presented papers and the
record of the discussion at the workshop.
Keywords: Residual Radioactivity, Recycling, Criteria, Reuse, Cleanup, Radiation
Protection, Decommissioning, Radioactive Waste, Regulation
United States Environmental Protection Agency
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JABRI-Conf 95-015
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JAERI-Conf 95-015
PREFACE
The use of atomic energy has provided an enormous benefit to mankind: the
generation of power, medical diagnostics and therapy, industrial applications, and countless
other uses. These benefits have been accompanied by advancements in technology and as
well as our knowledge of protection against the potential hazards of radiation. The latter is
manifest in radiation protection criteria and standards in national regulations and international
guidance.
The United States and Japan have long cooperated in the development of radiation
protection standards. Such standards cover the wide range of radiation-related human
activities, from licensing of sealed sources to limiting worker and public exposures. One
current area of concern for both countries is standards for site cleanup and the subsequent
management of wastes from site cleanup.
Both the U.S. and Japan are addressing the pending closure and decommissioning of
nuclear power stations as first generation power plants reach the end of their useful life. The
U.S. has further entered into the decommissioning and cleanup of aging nuclear weapons
facilities. New technologies are needed to safety and cost-effectively decommission these
facilities, and new standards are needed to address the environmental challenges they pose.
One of the most daunting problems facing nuclear cleanup is the vast quantities of
soil and construction materials that are classified as low level radioactive waste (LLW).
Currently, anything even loosely associated with a nuclear process or material may be
classified as LLW, whether or not any radioactivity is present. National standards do not
exist in either Japan or the U.S. that define how much radioactivity constitutes LLW. As a
result, either everything may be classified as LLW, or materials are released based on site-by-
site decisions as to what is sufficiently protective.
As part of an ongoing agreement, the Japan Atomic Energy Research Institute
(JAERI) and the Office of Radiation and Indoor Air (ORIA) of the U.S. Environmental
Protection Agency (EPA), co-sponsored the Second Workshop on Residual Radioactivity and
Recycling, on November 9-11, 1994. Fifteen technical experts from the U.S., representing
federal agencies and the private sector, joined over twelve Japanese experts in presenting their
latest research, risk assessments, and technical studies on residual radioactivity criteria and
recycling at JAERTs Tokai Research Establishment in Tokai-mura, Japan.
These proceedings are a compilation of papers developed by the authors and
presented at the Workshop. JAERI and EPA hope that the ongoing research and analyses
for residual radioactivity criteria will help to advance the assessment of potential risks and the
technologies needed to meet the new demands of radiation protection.
Hideaki Yamamoto John A. MacKinney
Department of Health Physics Office of Radiation and Indoor Air
JAERI U.S. EPA
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JAEM-Conf 95-015
Contents
Opening Address ••••• • • • ]_
K. Bingo, JABRI and B. C. Durman, U. S. EPA
1. Extent of the Problems related to Residual Radioactivity and
Recycling Criteria •• — - * - • 7
1-1 Status of the Japan's Regulatory Policy on Radioactive Waste Management,
- Cleanup and Recycling Issues - 9
D. Takeuchi, Science and Technology Agency, Japan
1-2 Potential Impacts of Pending Residual Radioactivity Rules • 18
D. D. Burns, Fernald Environmental Restoration Management Corporation
1-3 Progress of JPDR Decommissioning Project 2?
IL Kiyota, S. Yanagihara, JABRI
1-4 The Decommissioning Program of JAERI's Reprocessing Test Facility • • 39
T. UchikosM, T. Mimori, Y. Iwasaki, A. I to, JAERI
1-5 The Decommissioning Plan of the Nuclear Ship MUTSU •••• 49
M. Adachi, R. Matsuo, S. Fujikawa, T. Nomura, JAERI
2. Cleanup and Residual Radioactivity Criteria 65
2-1 Surface Radiological Free Release Program for the Battelle Columbus
Laboratory Decommissioning Project QY
C. N. Morton, Battelle Columbus Operations
2-2 Cost-Benefit Analysis for U.S. NRC Proposed Radiological Criteria for
Decomissioning - -75
R.A. Meek, U.S. NRC
2-3 EPA*s Technical Methodology for the Development of Cleanup Regulations for
Radioactively-contaminated Soils and Buildings • 80
H.B. Hull, M. Doehnert, A. Wolbarst, U.S. BPA
J. L Mauro, L. Ralston, Sanford Cohen I Associates
2-4 Radiological Surveys: Methods, Criteria, and Their Implementation 92
G. Subbaraman, R. J. Tuttle, B.M, Oliver, Rockwell International
2-5 Development of Risk-based Computer Models for Deriving Criteria on
Residual Radioactivity and Recycling • •••••• • * 101
S. Y. Chen, Argonne National Laboratory
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JAERI-Conf 95-015
2-6 Unrestricted Release of Contaminated Lands and the Dose to
the General Public • - - •• — • • Ill
H. Yamamoto, S. Kato, JAERI
2-7 Status of Environmental Response Efforts at Radioaclively Contaminated
Sites in the United States Air Force Installation Restoration Program • 116
W. D. Rowe,Jr., I.E. McEntee, Jr., The MITRE Corporation
B. Johnson, L. Manning, U. S. Air Force
3. Recycling and Criteria • 123
3-1 Evaluation of the Costs and Benefits of Recycling Radioactively
Contaminated Scrap Metal — — • 125
E. C. Durman, P. Tsirigotis, J.A. MacKinney, U.S. EPA
3-2 Technical Issues Relating to the Recycle of Contaminated Scrap Metal 135
S. Warren, U.S. DOE,
D. E. Clark, Westinghouse Hanford Co.
3-3 The Prospect for Recycle of Radioactive Scrap Metals to Products for
Restricted and Unrestricted Use • * 150
A.L. Liby, Manufacturing Sciences Corporation
3-4 Economic Aspects of Recycling U.S. Department of Energy Radioactive
Scrap Metal • ••— • • 160
J. Harrop, N. J. Numark, Sanf ord Cohen & Associates,
J.A MacKinney, U.S. EPA
3-5 Summary of Industrial Impacts from Recycled Radioactive Scrap Metals 175
J. C. Dehmel, J. Harrop, Sanf ord Cohen I Associates,
J.A. MacKinney, U.S. EPA
3-6 A Methodology for Estimating Potential Doses and Risks from Recycling
U.S. Department of Energy Radioactive Scrap Metals •••• 191
J.A, MacKinney, U.S. EPA
3-7 Study on Safety Evaluation for Unrestricted Recycling Criteria of
Radioactive Waste from Dismantling Operation •— 207
M. Yoshimori, M. OhkosM, M. Abe, JAERI
3-8 Radiological Control Criteria for Materials Considered for Recycle
and Reuse •••••• — • - • 216
I.E. Kennedy, Jr., R. L. Hill, R.L. Aaberg, Pacific Northwest Laboratory,
A. Walla I, U.S. DOE
3-9 Effects on Radiation Sensitive Instruments from Recycling of
Contaminated Metal • 229
H. Yamamoto, S. Kato, JAERI
VI
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JAERl-Conf 95-015
3-10 Metal Recycling Technology and Related Issues in the United States,
A BNPL Perspective •-• —• 234
P. Bradbury, S. Dam, W. Starke, BNFL, Inc.
3-11 Melting Tests for Recycling of Radioactive Metal Wastes •• — 247
E Nakamura, K. Kanazawa, K. Fujiki, JABR1
3-12 Investigation on Recycling of Radioactive Waste ••••• 255
D. Sakurai, K. Takahashi, A. Umemura, K. Kimura, Mitsubishi Material
Corporation,
S. Abe, M. Yamamoto, Radioactive Waste Management Center
4 Compliance with Criteria •*••• ••• • —— 269
4-1 Radiological Surveys to Demonstrate Compliance with Decommissioning
Limits • - • — 271
D.N. Fauver, U.S. NRG
4-2 The Japan Power Deionstration Reactor Decommissioning Program
- Decontaiination and Radioactivity Measurement on Building Surfaces - •••••• 281
M. Tachibana, M. Hatakeyama, Y, Seiki, S. Yanagihara, JAERI
4-3 Measurement of Residual Radioactivity in the Facility Being
Decommissioned ••• 290
E Bzure, S. Miyasaka, E Kuroda, J. Koiatsu,
Research Association for Nuclear Facility Deconoissioning
4-4 Radiochemical Analysis of Homogeneously Solidified Low Level
Radioactive Waste from Nuclear Power Plants • 302
K. Sato, Y, Ikeuchi, E Higuchi, Japan Chemical Analysis Center
4-5 Classification of Solid Wastes as Non-radioactive Wastes • 317
M. Suzuki, E Tomioka, K. Kamike, J. Koiatsu,
Research Association for Nuclear Facility Decommissioning
Free Discussion and Summary ••*•• • - 327
Appendix 1 Workshop Participants — • - 341
Appendix 2 Workshop Program *••* 351
Vil
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JAERI-Conf 95-015
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B.C. Durman, P. Tsirigotis, J.A. MacKinney (U.S. EPA)
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JAERI-Conf 95-015
3-2 135
S. Warren (U.S. DOB), D.B. Clark (iestinghouse Hanford Co.)
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A.L. Liby (Manufacturing Sciences Corporation)
3-4 160
J. Harrop, N. J. Nunark (Sanford Cohen & Associates),
J.A. MacKinney (U.S. BPA)
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J. C. Dehmel, J. Harrop (Sanford Cohen k Associates),
J.A. MacKinney (U.S. EPA)
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I.E. Kennedy, Jr., R.L. Hill, R.L. Aaberg (PNL), A. Wallo, I (U.S. DOB)
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JAERI-Conf 95-015
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JAERI-Ccmf 95-015
Opening Address
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JAERI-Conf 95-015
OPENING ADDRESS
Kazuyoshi Bingo
Japan Atomic Energy Research Institute
Ladies and Gentlemen,
On behalf of the Department of Health Physics of JAERI, I would like to express our sincere
welcome to all of you who are attending the second workshop on residual radioactivity and
recycling criteria, jointly sponsored by the Office of Radiation and Indoor Air, the United
States Environmental Protection Agency, ORIA/EPA, and JAERI. In particular, I would
like to extend our special word of thanks to the participants who came over from the United
States to Tokai-mura. And we are also very honored to have the officials from the Science
and Technology Agency of the Japanese Government.
Since 1985, we at the Department of Health Physics have undertaken several research
projects together with our friends of ORIA, Five years ago, in September 1989, the first
workshop on residual radioactivity and recycling criteria was held at St. Michaels near
Washington, D.C.. The Proceedings of the first workshop was published by EPA, and this
document is very useful for us as one of the most important resource documents in the
development of criteria for cleanup and recycling of radioactively contaminated objects.
Last year, our agreement went to the second stage. And following the first successful
workshop in USA, we, JAERI, has decided to host the second one in Tokai-mura. The
workshop is designed to exchange up-to-date information which is relevant to the issues of
common interest. Fifteen presentations are scheduled from USA, and tirrteen from Japan;
total is twenty-eight. We also plan a free discussion at the end of the workshop.
I am sure that the meeting will bring useful results for both of us.
Lastly, I wish all of you a pleasant stay in Japan, especially for U.S. participants, as well as a
successful workshop.
Thank you very much.
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JAERl-Conf 95-015
Eugene C. Durman
United States Environmental Protection Agency
First of all, thank you, on behalf of the U. S. Environmental Protection Agency and all of the
agencies representing the U. S. delegation.
Although this is the official opening of the Workshop, I've already found participation here to
have been very useful and very informative. I've learned two things already before this
morning. First, that JAERI has many activities in the nuclear arena, and its staff performs
those activities with a great deal of professional sophistication. We saw that clearly on our
tour yesterday. Second, JAERI is a marvelous host, and we thank you for the hospitality
that you have shown us thus far.
We are beginning the formal phase of the Workshop. Dr. Bingo has summarized the history
of the past relationship between EPA and JAERI. His remarks bring us to this workshop
today which focuses on two very important policy questions, both for the United States and
for Japan. First, is the question of "how clean is clean?" at sites that need to be released.
The second question, is what we do with all the material that we remove from those sites; the
question of recycling, reuse or disposal.
As you may know, the Environmental Protection Agency is attempting at this time to address
both of those questions, and therefore we think the information we will receive at this
Workshop will be very helpful and relevant to work that we need to do.
Good answers to the many questions on cleanup we will be discussing throughout the
Workshop are important. In the United States, for example, the size of the cleanup problem
for the U. S. Department of Energy alone may be hundreds of billions of U. S. dollars. I am
sure that this problem is serious for Japan as well So, this is a very important public issue
for both countries.
There are a number of key issues we have to discuss:
• The level of protectiveness that the public demands,
• The question of land use; whether or not industrial land should be cleaned to the same
level as land released for unrestricted public use,
* The question of public participation; what is the role the public should play in making
these decisions,
* The question of ground water; should it be protected in addition to an overall level of
protection. It is a very key issue in the United States.
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JAEW-Conf 95-015
When we look at recycling, a fundamental question for us is whether or not the public will
ever accept large scale reintroduction of radioactive material into the economy. It's a very
complex and difficult question for us. I understand that this is also the case in Japan.
But, as I reflect on these issues, I am encouraged by the content of our program today. We
have an excellent combination of policy papers, papers dealing with actual case studies, and
papers dealing with the technical issues that we all need to solve. As I look over the agenda,
I am very hopeful that our presentations are going to provide us information we can use, and
that our work here will contribute to both countries being able to develop successful policies
in this area. Thank you.
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JAERI-Conf 95-015
1. Extent of the Problems related to
Residual Radioactivity and
Recycling Criteria
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JAEM-Conf 95-015
1 -1 STATUS OF THE JAPAN'S REGULATORY POLICY ON RADIOACTIVE
WASTE MANAGEMENT, - CLEANUP AND RECYCLING ISSUES-
DAffl TAKEUCffl
Deputy Director, Office of Radioactive Waste Regulation,
Nuclear Safety Bureau, Science and Technology Agency
1. Forward
Wastes from nuclear, facilities are very diversified concerning that have different
levels of radioactivity and include different kinds of radioactive materials. Besides some
of those waste is not assumed as radioactive waste. The basic policy of the radioactive
waste management is taking that diversity into full account for appropriate separate
management of different types of radioactive waste and treatment and disposal of each
type in a rational manner, including recycling.
From the point, the disposal methods are considered or under consideration to that
waste, (1) from nuclear reactor facility, (2) from nuclear fuel cycle facility — HLW, waste
contaminated TRU nuclides, or contaminated uranium, (3) from RI utilization or research
institute, and (4) from decommissioning of nuclear facility.
Now in Japan, regulation framework for some kind of LLW from reactor facility,
including waste from decommissioning of reactor is established.
2. Treatment of non-radioactive waste from nuclear facilities
47 nuclear power reactors (the capacity is 3947.6 Mkw) are under operation in
Japan. As for the nuclear fuel facilities, 8 fuel fabrication facilities, 2 reprocessing
facilities are under operation or construction. Another activities, for example R&D by
using research reactor, nuclear materials or radio-isotope, are done. Many kind of wastes
are generating by those activities.
Some of the waste from nuclear facilities, however, have not been contaminated.
Nuclear Safety Commission (NSC) recommended 'On the Understanding of "Nuclear Fuel
Materials or Substances Contaminated by Nuclear Fuel Materials"' in Oct. 1993 , to it
makes clear the concept of 'non-radioactive waste*. (Appendix 1)
The concepts are as follows;
(1) Those wastes that are apparently free from secondary contamination due to the
adhesion and permeation, or those contaminated parts are separated.
(2) Concrete wastes and metal wastes that need not be taken into account the activation
by the structure, the activation levels are calculated as of no significant difference
from the materials used in non-nuclear field, or those parts are separated.
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JAERI-Conf 95-015
According to this recommendation, some concrete waste generated by the work of
the exchange of steam generator in PWR and decommissioning of JPDR are treated as
non-radioactive waste. Those wastes are treated similar with industrial wastes, without the
regulation from the viewpoint of radiation protection.
3. Radioactive waste in Japan
The various types of radioactive wastes, generated during the course of nuclear
activities and they are stored in nuclear facilities. The amount of low-level radioactive
wastes stored in nuclear facilities as of the end of March 1994 (measured in 200 liter
drum-can equivalents), is about 824,000 cans. The amount increased by about 15,000 cans
during fiscal 1992 (See Table 1). The appropriate management and disposal of radioactive
wastes is an indispensable issue in promoting further development and utilization of
nuclear energy.
Table 1. Cumulative Amount of Radioactive Waste in Storage
Fiscal Year
Nuclear
Power
Staton
JAERI
PNC
Fuel
Fablication
Plants
Japan
Isotope
Association
total
emulative
amount as
of fiscal
1989
466,800
108,000
100,800
24,800
56,200
756,600
Cumulative
amount as
of fiscal
1990
471,800
113,000
106,000
27,400
62,100
780,300
Cumulative
amount as
of fiscal
1991
478,300
119,100
112,200
29,300
66,600
805,500
Cumulative
amount as
of fiscal
1992
479,000
124,900
116,700
30,300
69,600
820,400
Cumulative
amount as
of fiscal
1993
471,800
129,200
120,100
32,000
71,000
824,000
( 20,500 (cumulative amount is 26,600) from nuclear power station
were shipped to disposal facility)
The basic concept for the disposal of low-level radioactive wastes is the land
disposal. The Japan Nuclear Fuel Ltd. is operating a relatively shallow land burial facility
in which to dispose of about 200 thousands of 200 liter drum-cans of radioactive waste
(ultimate scale, 3 million cans). This facility began operation in December, 1992 .
As of the end of March 1994, high-level radioactive wastes generated by the
reprocessing facility amount to approximately 542 nP, and are stored in the facility under
strict control. Liquid high-level radioactive wastes generated by the reprocessing facility
are extremely small in quantity, they have a long half-life and high radioactivity. It is
therefore considered necessary to keep these wastes isolated from human living in order to
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JAERI-Conf 95-015
prevent environmental pollution and radioactive exposure of the public. Bearing this in
mind, the basic policy is to solidify the liquid high-level radioactive wastes into stable
forms and store them for 30 to 50 years for cooling. The Japan Nuclear Fuel Ltd. is
constructing a radioactive waste management facility which will receive and store high-
level radioactive wastes arising from overseas fuel reprocessing (iii France and in the
United Kingdom). This facility is scheduled to begin operation in Feb. 1995 . They are
then disposed of into geologic formations more than several hundred meters underground.
4. National policy for radioactive waste treatment
The basic policy for nuclear program in Japan is "Long-Term program for
Researchs Development and Utilization of Nuclear Energy"( Atomic Energy Commission
(AEC), June 1994.
The basic policy for radioactive waste management summarized as separate
management of different types of radioactive waste. Radioactive waste is very diversified
considering that it has different levels of radioactivity and includes different kinds of
radioactive material. That being the case, besides taking that diversity into fully account
for appropriate separate management of different types of radioactive waste and treatment
and disposal of each type in a rational manner, it is intended to consider reutilization
possibilities from the standpoint of effective use of resources. Furthermore, in addition to
carry forward in a steady manner necessary research and development, it is intended to
appropriately proceed with exemption and clearance from regulation when necessary for
the purpose, taking into consideration international trends in that respect.
(1) Waste from Nuclear Power Plants
Low level radioactive waste will disposed in shallow land disposal facility.
(2) Waste from nuclear fuel cycle facility
(i) High level radioactive waste
After cooling for 30-50 years, it is disposed in geological disposal facility. The
repository will start its operation around 2030-mid-2040.
(ii) Waste containing TRU nuclide
Waste from reprocessing facility or MOX fuel fabrication facility are treated as
TRU wastes. Low level TRU wastes may be disposed at the shallow depth in the
ground with engineered barrier, and specific measures. High level TRU waste will
be disposed in the ground but not at shallow depth.
(iii) Uranium Wastes
Most of the wastes from uranium conversion, fabrication or enrichment facilities
has a comparatively low uranium concentration and will be possible to dispose of it
at a shallow depth in the ground without engineered barrier by a simple method that
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JAERI-Conf 95-015
does not entail step-by-step control.
(3) RI waste and Research Institute and other waste
Waste contaminated by short-life beta and gamma nuclide will be disposed by
shallow land disposal or simplified shallow land disposal. Waste contaminated alpha
nuciide will be disposed by similar way of TRU tmclides.
Research institute treats many kind of nuclides, those waste from research
institute will be disposed by the way of its origin (mentioned above).
5. The Legal System related to Regulations OB Radioactive Waste Disposal
The management and disposal of radioactive wastes is regulated by either the
Reactor Regulation Law or the Radiation Hazard Prevention Law according to the
facilities of origin. The following is an elaboration of the Reactor Regulation Law as the
regulations of the Radiation Hazard Prevention Law are similar to those of the Reactor
Regulation Law.
When the reactor licenses, fabricating businesses, and other nuclear businesses
manage or dispose of radioactive wastes, generated as a result of the practice of their
business, within the plant site, the regulations governing such operations are part of the
regulations concerning each operative facility.
Regulations state that when nuclear businesses dispose of radioactive wastes
outside of the plant site, the following measures should be taken:
(A) The wastes must be placed in nuclear facilities, (other than the ones where the
radioactive wastes were generated) , which have radiation hazard prevention
functions.
(B) The imported wastes must be placed in radioactive waste management facility
The necessary measures for insuring the safety include,
i) enclosing the wastes in containers by the prescribed methods, or
ii) solidifying the wastes in containers, and
iii) making sure the radioactivity concentration does not exceed the prescribed level.
Disposal operations cannot be carried out without prior confirmation by the Prime
Minister that the operations comply with the above measures.
6. Regulations on the Business of Burial of Radioactive Wastes
The Nuclear Safety Commission had discussed basic concept of the safety
regulation of the burial disposal of low-level radioactive wastes. The report was complied
in October 1985 entitled basic Concept of safety regulations concerning land disposal of
low-level radioactive solid wastes".
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JAERI-Conf 95-015
Based on the prescribed concept, the Nuclear Safety Commission and the
Radiation Council have discussed the policy to summarize the Basic Concept of Safety
Regulations concerning Land Disposal (March, 19B8). The recommendation by the
Radiation Council is that in the case of establistaent of the Japanese own exemption of
radioactive solid wastes regarding shallow land burial disposal from a viewpoint of the
control of radiation protection, as the dose, which the exemption to be based on, it is quite
appropriate to adopt 10 micro-Sv per year.
Safety Regulation for Burial Disposal Business of Radioactive wastes will be
carried out on these concepts. The concept of safety regulations that apply to burial
disposal business can be summarized as follows in Table 2 and Fig. 1.
In Japan, Disposal business can be applied only when radioactive wastes below
" Upper limits of radioactivity concentration " decided by Cabinet Order. When
this order may be amended, it must be submitted to NSC and AEC.
7. Recycle of radioactive waste
Reutilization is also under consideration from the view point of effective use of
resource. The application of the concept of clearance or exemption to the radioactive
waste from nuclear facility is also under study taking into consideration international
trends. We are continuing some study to adopt those concepts in our regulation framework,
because very amounts of waste may be generated in nuclear power plant decommissioning
in future.
R&D for reutilization technology , for example by using JPDR decommissioning
waste, is under way.
From the regulative point, the dose criteria, 10 // Sv/y to individual, is
recommended for the reutilization by the Radiation Council. ( 'it will not be necessary to
review this regulatory exemption dose at the time of the implementation of recycling in
future.') It is same criteria for the waste burial facility after institutional controlled period.
Research is also under way to apply this recommendation to our regulatory framework,
because the scenario for safety evaluation may be different from waste burial and
reutilization.
For that purpose, from FY 1995 to FY 1997, the research to establish the criteria
for reutilization of LLW will start under the Safety Research Program decided by NSC in
March, 1994.
The use of grounds after decommissioning is also considered. When radioactive
wastes will disposed, the site will be used without regulation from the point of radiation
protection.
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JAERI-Conf 95-015
Appendix 1
On the Understanding of "Nuclear Fuel Materials or Substances Contaminated by
Nuclear Fuel Materials" - Nan Radioactive Solid Waste Generated from the
Decommissioning of Reactor Facilities -
1. Background
In Japan, radioactive wastes generally refer to nuclear fuel materials and/or
nuclear-fuel-contaminated substances which are to be disposed (as stipulated by the laws
and regulations related to the Law for the Regulation of Nuclear Source Material, Nuclear
Fuel and Reactors. Based on this definition, practically all solid wastes from control areas
in reactor facilities are regarded as being contaminated and treated as such (storage and/or
disposal etc.) under these laws and regulations.
These solid wastes, however, include wastes which have not been contaminated.
Moreover, it is expected that the dismantling of nuclear facilities will generate not only
radioactive wastes but also non-radioactive wastes exceeding in amount the former kind of
waste.
It is not reasonable therefore to regard those wastes which have no possibility of
having been contaminated by radioactive substances or those wastes with no significant
difference in their radioactivity levels from natural background as radioactive wastes.
Such being the case, an understanding on radioactive wastes has been framed to optimize
and rationalize the management of radioactive wastes. This understanding forms the basis
of discriminating non-radioactive wastes from radioactive wastes.
2. Basic Viewpoint in Discriminating Non-radioactive Wastes from Radioactive Wastes
The radioactive contamination of solid wastes from nuclear facilities is generally
classified into secondary contamination due to the adhesion and permeation of radioactive
substances and activation contamination due to neutron capture. With these classifications
of radioactive contamination being taken into account, "contamination" is defined, and the
basic viewpoint in discriminating non-radioactive wastes from radioactive wastes is
described in the following pages.
(1) Secondary contamination
i) Those wastes which are apparently free from secondary contamination due to the
.adhesion and permeation of radioactive substances, in view of their history of use
and their conditions of installation.
ii) Those wastes whose secondary contamination due to the adhesion and permeation
of radioactive substances is limited and whose contaminated parts are separated,
in view of their history of use and their conditions of installation.
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JAERI-Conf 95-015
Those wastes which fall under either one of the classifications mentioned above are
regarded as non radioactive wastes.
(2) Contamination by neutron capture (Concrete waste including reinforcing bars)
i) Those concrete wastes which need not be taken into account the activation by
neutron capture, in view of the structure of facilities, such as being sufficiently
shielded by shielding materials.
ii) Those concrete wastes for which the activation levels are calculated as of no
significant difference from the concrete used in non-nuclear field (including
reinforcing bars in them).
iii) Those concrete wastes whose activation levels are defined by calculation and
whose significantly activated portions have been removed.
Those concrete wastes which fall under either one of the classifications
mentioned above are regarded as non radioactive wastes.
(3) Metal wastes contaminated by activation
To metal wastes generated with the dismantling of nuclear facilities, the
classifications mentioned in (2) above can be applied for the activation by neutron
capture.
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Table 2 Safegy regulation for repository
Institutional Control Period
Stage
First stage; maintaining the
integrity of the artificial barrier
Second stage; Securing the barriers
performance
Third stage: Prohibition or restriction
of specific acts
Post closure
Concept of securing safety
Prevention of the leakage of
radionuclides outside the
artificial barriers and
confirmation that there is no
leakage
Prevention of the effects of
radioactive nuclides reaching
living environment by artificial
barriers. Confirmation of safety
by environmental radioactivity
monitoring.
Prevention of the effects of
radioactive nuclides reaching living
environment mainly by natural
barriers. Prohibiting or restricting
specific acts, i.e. digging out wastes.
There is no necessity to
take measures to prohibit
specific acts.
Safety
regulation
Concepts of
safety
evaluation at
each stage
On conditions that there is no
leakage from the artificial barrier
and that at repairs will be made
in the case damages occur, it is
satisfied that the effects of
radioactivity on the environment
will be prevented
On conditions that the specific
acts, i.e. digging out wastes are
prohibited or restricted, it is
satisfied that the mainly natural
barrier will prevent the effects on
the environment.
On conditions that the members of
the public don't enter the repository,
it is satisfied that the combination of
artificial barriers and natural barrier
will prevent the effects on the
environment.
It is satisfied that the dose
to public is extremely low
without having to
consider regulation of the
waste as radioactive
materials.
Concepts of
security
measures
Installation of building
structure : monitoring leakage of
radioactive materials from the
buildings and structures and
repairing if necessary.
Monitoring radiation at the
site: ,measures to limit access.
Monitoring of environmental
radiation at the site; patrolling and
inspection; measures to restrict
access, etc.
Prohibition or restriction of specific
acts; patrolling, etc.
2
o
o
-------
(managed in the
nuclear facilities)
Radioactivity
upperbounds
--a
I
Radioach
Stagel: maintaining
the integrety of the
barriers
concentration
for disposal
ve level of waste
Stage 2; Securing the
barrier performance
Stage3;
Prhibition
or restriction of
specific acts
Post closure
(level below regulatory
construction of repository
A (closure)
End of burial disposal buisiness
I
i
Fig. 1 Concept for disposal
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JAERI-Conf 95-015
1-2 POTENTIAL IMPACTS OF
PENDING RESIDUAL RADIOACTIVITY RULES
Daniel D. Burns
Manager, Recycle Programs
Fernald Environmental Restoration Management Corporation
ABSTRACT
The purpose of this paper is to present an overview of pending rules
governing residual radioactive release criteria and radioactive waste
management, as well as addressing the potential impacts on the Fernald Scrap
Metal program. More than 600,000 cubic feet of radioactively contaminated
waste will be generated during the dismantlement of 3 complexes at the
Fernald Site in the next year and a half. Under current regulations, as much
as 70% (5000 tons) will be either recycled or re-used in controlled
applications. Depending on regulatory developments, the ratios of recycling
to burial will range from 100% burial to recycling more than 90% of the
waste.
The lack of federal rules and regulations for classification of permissible
levels of residual radioactivity is one of the most troublesome issues in the
nuclear industry. The issue is growing in importance with the approaching
end of useful life for many nuclear power generating stations and the planned
remediation of the DOE nuclear weapons complex. Federal regulators have been
involved in the "Enhanced rulemaking" process for over two years. The DOE
Fernald site offers a good opportunity for understanding the potential
impacts of the pending regulations due to the maturity of the planned D&D
activities, aggressive recycling program, and simple nature of contamination,
and may offer a point of departure for many facilities engaged in D&D and
waste management.
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JAERI-Conf 95-015
BACKGROUND
The Fernald Site is a former uranium metal production facility which was
utilized for the conversion of UF6 to uranium metal, and subsequently used
for various applications within the Department of Energy. The production
mission commenced in 1952, and proceeded through 1989. In 1989, the
Department of Energy made a decision to.end the production mission at the
Fernald facility, and begin the remedial action dedicated to the cleanup of
the former production facility.
As a result of the production activities, uranium contamination was dispersed
throughout the facility, which is comprised of approximately 80 active acres
of operations. Two major areas are being addressed within the complex, the
former production facility, and a waste pit area to the east of the
production facility used for land placement of various process generated from
the beginning of operations until 1985.
In 1985, land burial at the facility was ended and process waste were either
stockpiled or packaged for transport and burial at the Nevada Test Site.
The mission was very straight forward with respect to the operations at the
Fernald facility. The primary contaminants associated with all areas at
Fernald are uranium, and thorium. No reprocessed fuel was used at the
facility, therefore, fission products and activation products are not
suspected at Fernald.
Various forms of Uranium were produced during the life of the project, which
included a depleted uranium metal, normal distribution metal and also low
enriched uranium up to approximately 2% Uranium 235. The goal of the
remedial action at the Fernald site is to excavate and stabilize the waste
that was previously placed in the ground, remove contamination from an
aquifer which underlies the entire facility, and to take to grade or demolish
all of the production facilities formerly used for uranium production.
As a result of the remedial actions, a large quantity of radioactive waste
will be generated. Table 1 depicts the volumes of the major categories of
waste. As can be seen, the total is nearly 3,000,000 cubic meters of waste,
two- thirds of which will be soil and clay. The next major contributor to
the volume of waste requiring remediation is 600,000 cubic meters of waste
pit contents, and the remaining portion, nearly 300,000 cubic meters, will
involve the management of the construction debris from the dismantlement.
TABLE 1
TYPES
Soil and Clay
Waste Pits
Construction Debris
Water Treatment Solids
Radium Bearing
TOTAL
CUBIC YARDS
1,700,000
780,000
320,000
35,000
8,000
3,844,000
CUBIC METERS
2,064,528
596,419
244,685
26,762
6,882
2,939,000
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JAERI-Conf 95-015
SCRAP HETAL MANAGEMENT
Historical Practice
The historical practice for the management of radioactive scrap metal
generated during the life of the production mission was to stockpile the
metal in the northeast section of the production area. By 1989, more than
6,000 short tons of radioactive scrap metal had been placed in the scrap
metal storage area. An aggressive project was initiated in 1991 to clean up
this area, and plans developed for the management of the radioactive scrap
metal.
All but 2,200 tons of the radioactive scrap metal were packaged into large 8
foot x 8 foot x 20 foot containers and transhipped to the Nevada Test Site
for burial. A project was initiated in 1991 to recycle or beneficially reuse
the remaining metal stockpiled at the scrap metal storage facility.
The radioactive scrap metal destined for recycling or beneficial reuse
consisted of both ferrous and non-ferrous metals (primarily ferrous metals)
with a nominal contamination level of 50,000 dpm per 100 centimeters squared
or 8.3 becquerel per centimeter squared with natural uranium.
A turnkey project was initiated to hire a subcontractor to provide
characterization, size reduction, packaging, transportation, surface
decontamination, metal melt, and secondary waste disposition. The end
product for the action was the fabrication of shield blocks which would be
transhipped to the Department of Energy for use as bioshielding in
accelerator projects within the medium energy physics program. At the
completion of the project, 90% of the material by weight had been
beneficially reused or recycled.
A cost assessment was performed for the activity in which the recycle and
reuse contracts was compared to the historical practice of disposal at the
NTS. Disposal of the E210 tons of scrap metal, would have cost approximately
$4 million as compared to the expenditure of nearly $4.8 million to contract
the services for beneficial reuse.
A net cost advantage was realized, given that the DOE avoided the expenditure
of more than $1.7 million for the purchase of virgin metal shield block for
the medium energy physics program. The benefit equated to nearly $1 million
savings within the DOE.
Future RSM Management
At the completion of the initial scrap metal recycling project, the
management at Fernald reviewed the performance of the contract and the
methodologies employed. A primary consideration was made to further
segregate any future generated radioactive scrap metal. The segregation
would occur primarily based on physical form, with the distinction being made
on not only the radiologically characteristics, but the presentation of the
substrate.
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JAERI-Conf 95^015
HETAL CLASSIFICATIONS
Metal waste at the FEHP is divided into two categories: refuse and
recoverable. Recoverable metal (scrap metal) is further divided into two
subcategories; recyclable and reusable. The distinctions are based on the
physical and radiological characteristics of the metal form. Disposition of
these materials can only be identified once the materials are appropriately
categorized. Appropriate segregation into these categories will facilitate
the most cost effective and timely final disposition of metal waste. The
following are descriptions of the categories:
1. REFUSE - Refuse metal waste is metal which is radiologically
contaminated or suspected of being radiologically contaminated. The
physical form os the metal is such that is excessively oxidized or a
bimaterial form where separation of the metal from the other materials
is not cost effective. Evaluation of cost effectiveness requires a
comparison of the cost of managing the material as refuse considering
the regulatory status of the material as a waste (a specific material
may be cost effective to recover if it would be regulated as mixed
waste whereas it may not be cost effective to recover if it would be
regulated as low level radioactive waste).
2. RECOVERABLE - Recoverable metal is metal which is radiologically
contaminated and can be processed for unrestricted release or
controlled reuse. Generally, this category includes all metal which
does not have the refuse characteristics.
A. Unrestricted Release metal is metal which can be
decontaminated and all potentially contaminated areas are
accessible for direct contamination survey. Generally,
unrestricted release scrap metal has a low surface area to
mass ratio. Examples of reusable scrap metal are
structural steel, tanks and decking. Metal forms may be
considered for unrestricted release even if there are minor
portions which cannot be cleaned or monitored if that
portion can be effectively removed from the form.
B. Restricted Release scrap metal is metal which cannot be
decontaminated or surveyed to verify that the release
limits have been met. Generally, restricted release metal
is light gauge or has inaccessible areas where
contamination may be present, such as ductwork, cabinets,
machinery, and odd sized forms. Restricted release scrap
metal may include unrestricted release metal when it is
determined that the restricted end-use is more cost
effective.
Fernald will be generating large quantities of radioactive scrap metal. It
is anticipated that during the demolition of the former production area, more
than 100,000 tons of radioactive scrap metal will be generated. Nearly one-
third of this will fit into the category of Unrestricted Release Recoverable
metal, while the remainder will be considered Restricted Release Material not
conducive to free-release in accordance with existing surface radioactivity
guidance.
Several projects are ongoing for the implementation of the radioactive scrap
metal recycling project, the first of which is the Plant 7 Project.
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JAERI-Conf 95-015
Plant 7
As a result of the demolition of the building of Plant 7, 710 tons of
structural steel and deck plate has been generated. All of this material
has been containerized into reusable containers, and is awaiting shipment to
an offsite facility for surface decontamination and free-release. The
contamination level of the structural steel is a nominal 30,000 dpm per 100
centimeters squared or 4.51 becquerel per centimeter squared.
Only depleted uranium was processed at this facility, and measurements were
taken to determine the thickness of lead base paint on the members. 8 mils
of lead base paint were discovered to be on the surfaces. A contract has
been let for the transportation, surface decontamination, survey, release,
and secondary waste disposal of the 710 tons of scrap metal. The end product
will be recycled scrap metal with no restrictions, and will be sold to a
commercial vendor. It is believed that 95% by weight will be recycled.
The cost of the activity is approximately $1.4 million, as compared to a
disposal cost of $1.5 million for this material. An important consideration
in conducting cost comparisons between recycle and reuse options vs. disposal
is an understanding of the packaging efficiency for this type of material.
Previous experience at Fernald has indicated that a density of 16 Ibs/cubic
foot can be obtained without exhaustive size reduction actions prior to
packaging. Given that no automated or methodized size reduction capabilities
exist at Fernald, it is appropriate to use this density in the disposal
analysis.
Material Release Facility
Another project initiated at the Fernald site is the utilization of a
previously unused facility as a Material Release Facility. The purpose of
this facility is to provide the necessary quality assurance, survey and
decontamination operations to release metal from the radiologically
controlled area. The candidate material identified for processing through
this facility is in general heavy gauge, lightly contaminated material that
is suspected of not requiring exhorbant decontamination technologies. In
fact, the only decontamination techniques which are employed are dry
vacuuming, scrubbing, scraping and low pressure steam with detergent
additives. It is anticipated in the future that additional decontamination
technologies (i.e. grit blast, close circuit grit blast) will be employed but
will not be complex from the perspective of either capital investment or
technology.
Through the first five months of the project nearly 180 tons of metals have
been released through the process and have been sold to local scrap dealers
for nominal scrap value. This facility operation will continue through the
life of the remedial project. As long as activities are ongoing in the
radio!ogically controlled area, there will exist a need for the controlled
survey and release of items that may become potentially contaminated.
Fernald believes that it will process approximately 600 tons of material
through this facility annually.
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JAERI-Conf 95-015
Waste Management Approach
The approach at Fernald Is to develop a portfolio of disposition options for
the waste generated as a result of the remedial action. At different times
within the life of the project, various needs will become priority. Most
notably, the needs will consist of economic evaluations and scheduler
concerns. It is felt that with a portfolio of options for the management of
the various types of radioactive waste the roost responsible disposition will
be able to be utilized.
Currently, for radioactive scrap metal the intention is to recycle all of the
material generated at the Fernald site. This can be accomplished through a
combination of use of the Materials Release Facility and individual contracts
for surface decontaminable material and beneficial reuse for volumetrically
contaminated or metal with inaccessible spaces. Additionally, the option of
land burial remains a viable option if the other methods can not be employed
in a reasonable manner.
Regulations
All of these management techniques have been developed to conform to the
currently existing regulations. Changes are anticipated in the regulation of
radioactive waste treatment storage and disposal. Most notably in the
definition of radioactive material itself and also recycling radioactive
scrap metal criteria.
At this point, the existing regulations only allow for the release of
material which can be demonstrated to conform to surface radioactivity
guidance. No regulatory foundation exists for the release of volumetric
contamination or material that has inaccessible contamination for surveying.
As a summary to the pending regulations within the United States, Chart 1 is
offered to depict the activity. When cleaning up a facility, it is easily
visualized that there are four lodes for releasing contaminants into the
environment which could result in potential exposures. Of the four exposure
pathways two are extremely well regulated. Air emissions resulting from the
operation and decommission of a facility are well regulated under the Clean
Air Act. Additionally, any water effluent associated with a facility are
well regulated under the Clean Water Act and the Safe Drinking Water Act.
The direct exposure associated with the facility, and the exposure associated
with the solid waste generated at the facility are less well regulated.
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JAERI-Conf 95-015
PENDING REGULATIONS
Radiological Criteria for
Clean Air Act
Decommissioning (USNRC)
FACILITY
(USEPA)
Air
Water
Clean Water Act
Safe Drinking Water Act
Radiation Site Cleanup
Solid Haste
Remediation Waste
Management
Recycle
Radioactive Recycling
Recycling Criteria
There are two pending regulations for the control of direct exposures as a
result of a facility being remediated. One regulation, Radiological Criteria
for Decommissioning issued by the United States Nuclear Regulatory
Commission, is designed to regulate facilities which operate under an NRC
license. A parallel regulation issued by United States Environmental
Protection Agency entitled "Radiation Site Cleanup Standards" is being
developed for implementation at facilities other than NRC license facilities,
such as federal facilities.
By definition if the air, water and direct exposure routes are regulated to
certain levels, this will dictate a certain amount of solid waste be
generated to conform to these standards. The industry is in great need of
regulations which will adequately address the issue of solid waste of which
recycling regulations would be a subset. The U.S. EPA is developing
regulations for solid waste. The overall program which is titled "Radiation
Waste Management" has been initially developed and will continue to be worked
on for the next several years. As a subset of this, a specific regulation
will be developed for the management of materials which may be recycled out
of this solid waste stream generated during facility cleanup.
DESCRIPTION OF PENDING REGULATIONS
The U.S. EPA Decommissioning Staff Draft
The scope includes setting standards for the remediation of soil,
groundwater, surface water, and structures at Federal facilities. A staff
draft is in review and comment resolution. Pathway analysis and modeling are
in progress, the most mature of which are the soil regulations. The major
element of this regulation is the establishment of a 15 millirem per year
effective dose equivalent exposure to the reasonably maximumly exposed
individual. If this level is met, the facility may be abandoned with no
restrictions based on its future use.
The 15 millirem per year value includes a four millirem per year component
dedicated to the groundwater associated with the facility. The basis for
this regulation has been developed from the International Atomic Energy
Agency, the International Council on Radiation Protection, and the National
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JAERI-Conf 95-015
Council on Radiation Protection recommendations. The risk based levels are
consistent with the CERCLA requirements for an excess cancer rate of 104 to
106. In actuality, the 15 millirera per year dose equates to a 3 x ID4, which
is considered to be within the range described by the U.S. EPA.
United States Nuclear Regulatory Commission Radiation Site Draft Cleanup
Staff Draft
The scope includes specific radiological criteria for decommission of soils
and structures at NRC license facilities. A.final rule is anticipated in
May, 1995. The major element of this regulation is the establishment of a 15
millirem/year total effective dose equivalent, distinguishable from
background and with ALARA considerations. The basis is the International
Council on Radiation Protection, and the National Council on Radiation
Protection recommendations for individual dose.
Solid Radioactive Waste Pending Regulations
The scope of regulations under the are being developed and have not been
fully determined, but may include source material, special nuclear material,
by-products, high-level waste, mixed waste, transuranic waste, and low-level
waste. The status of the regulation is that an issue paper has been
developed, and a proposed rule is anticipated late in 1995. The major
elements will include requirements for treatment, storage and disposal of
radioactive waste.
The most significant issue being discussed in the preliminary development of
this regulation is the inadequacy of current waste classification systems.
In essence, it may be appropriate for regulators to come up with a new
classification system based on hazard rather than the generating process. In
some cases low-level waste are more hazardous than some forms of high-level
waste, as well as some forms of NORM waste being more hazardous than mixed
waste.
U.S. EPA Radioactive Material Recycling
The scope for recycling radioactive material rules has not been determined,
but may include both restricted and unrestricted scenarios for regulation and
implementation. The current status is that an issue paper is being
developed to initiate the discussions, and identify the need for any future
regulations.
Impact of Potential Pending Regulations
It is premature to address the impact of the pending regulations on the
current waste management practices at the Fernald site. The desire is to
have a consistent and accepted rule governing the activities associated with
radioactive waste management, in particular, the release of materials for
reuse and recycling to reclaim their raw material value. At the Fernald site
the question is extremely significant because the issuance of any of these
rules will occur during the implementation of the Fernald cleanup.
Therefore, adjustments will have to be made as the work is conducted. At the
Fernald site it is not possible to wait for resolution of these issues and
issuance of these regulations prior to conducting the' cleanup activities.
Recycling radioactive scrap metal will continue to play an important role in
the remediation of the Fernald site.
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JAERI-Conf 95-015
Risk Based Regulations are welcome and it is felt that they can be
implemented at the site with little concern. The impact of any regulation
will primarily be associated with the cost of dispositioning the material.
In the absence of adequate regulation or with regulations that result in
levels of release which are indistinguishable from the background radiation
the affect may be that radioactive scrap metal will not be recycled and
disposal of land burial at an appropriate facility.
Summary
The Fernald site is an ongoing project. We are on the verge of implementing
large scale activities which will result in a generation of large quantities
of radioactive waste, including radioactive scrap metal. Under the current
regulations Fernald is able to recycle a majority of the radioactive scrap
metal being generated at a cost which is comparable to other viable options
such as land burial.
The pending regulations will be issued during the life of the Fernald
project, and may have a severe impact to the ability of Fernald to continue
beneficial reuse or recycling of its radioactive scrap metal, and may result
in the burial of this material along with the contaminated soils and other
radioactive waste residues. At this time, however, it is premature to
speculate on these impacts given the lack of scope definition and lack of
confidence in the ability to develop a widely accepted regulation concerning
release of radioactive scrap metal.
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JAERI-Conf 95-015
1 -3 Progress of JPDR Decommissioning Project
M. Kiyota and S. Yanagihara
Department of Decommissioning and Waste Management
Japan Atomic Energy Research Institute
Tohai-mura, Ibaraki-ken, 319-11, Japan
ABSTRACT
The Japan Power Demonstration Reactor (JPDR) decommissioning project is
progressively achieving its final goal; the project will be finished by March 1996 to release
the JPDR's site into unrestricted use in a peen field condition. The new techniques which
developed or improved in R&D, the first phase of this program, have been successfully applied
to the actual dismantling activities. Some decommissioning wastes have been managed as the
first case of onsite shallow land burial based on the new regulatory frame of radioactive waste
management. The experiences and the data obtained from the JPDR dismantling activities are
expected to contribute to future decommissioning of commercial nuclear power plants.
INTRODUCTION
The JPDR is the first Japanese nuclear power reactor, which constructed during
September 1960 to August 1963 for the purpose of obtaining an experience on construction
and operation of a nuclear power plant, studying on characteristics of a nuclear power reactor,
and contributing to development of domestic technology by irradiating nuclear fuels made by
Japanese industries.
The JPDR attained criticality initially August 2"2,1963, and it started power generation
of 45 MWt March 15,1965. After several years run of the JPDR, the reactor was shutdown
and it was modified by installing recirculation pumps to enhance nuclear capability to the
power of 90 MWt. The modified reactor, JPDR-II, attained first criticality February 1972
and operated for about four years. The JPDR was shutdown in March 1976 due to several
problems such as cracking on the nozzle of in-core monitor turbines, the failure of control rod
drive mechanisms and other complications. Consequently the possibility of re-starting the
reactor and the use for other objectives had been evaluated by experts in JAERI, then it was
finally decided to decommissioning the JPDR as a demonstration project for future
decommissioning of commercial nuclear power plants.
The JPDR decommissioning program was initiated in 1981 under contract with the
Science and Technology Agency (STA) in Japan,(l) The JPDR decommissioning program
consists of two major phases; Phase 1 began in 1981, aiming at developing the technology
necessary for reactor decommissioning. Phase 2, actual dismantling of the JPDR began in
1986 to reach green field condition, that is, stage 3 in IAEA definition, using the technology
developed in Phase 1. To date the dismantling activities are in progress successfully, and
various data on the dismantling activities have been collected and accumulated in the
decommissioning database.®
JPDR DECOMMISSIONING PROJECT
JPDR
The JPDR is a Boiling Water Reactor (BWR) designed by General Electric (GE) Co.
Ltd. It has the containment vessel with a hemispherical topdome, cylindrical body and
hemispherical bottom shape, which is called reactor enclosure. The total operation time and
output of electricity were about 17,000 h and 1.4 x 106 kWh, respectively.
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JAERI-Conf 95-015
The residual radioactive inventory in the JPDR facility is estimated to be approximately
130 TBq as of April 1988. Almost all radioactive inventory remains in the reactor internals,
the reactor pressure vessel (RPV) and the biological shield. Table 1 shows the major
specifications, operation history and radioactive inventory of the JPDR,
Japanese Policy for Decommissioning Nuclear Power Plants
The nation's policy for decommissioning nuclear facilities described first in the long
term program for research, development and utilization of nuclear energy (Long-Term
Program) issued 1982 by Atomic Energy Commission, Japan. Even the Long-Term Program
was revised since then, the basic nation's policy for decommissioning nuclear facilities has not
been changed so far. It described that in decommissioning commercial power reactors, as a
rule the nuclear reactor should be dismantled and removed as soon as possible after its
operation is terminated. It described about development of dismantling technology, as
development efforts and field testing will be continued using the JPDR and other facilities for
the purpose of obtaining dismantling technology and know-how that can be applied in actual
decommissioning of commercial power reactors and similar nuclear facilities.
Time Schedule
The project had started with R&D (Phase-1) on technology development such as
underwater plasma arc cutting, robotic manipulator, underwater arc saw cutting, shaped
explosives, rotary disk knife, mechanical cutting and water jet cutting, which included basic
studies and mockup tests using these techniques.
During the Phase-1, the JPDR spent fuels were transferred to Power Reactor and
Nuclear Fuel Development Company (PNC) for reprocessing. Actual dismantling activities
started 1986; components around the RPV were first dismantled to prepare the space for
subsequent dismantling activities using remote dismantling machines. Figure 1 shows the time
schedule of the JPDR decommissioning program. The major milestones conducted so far are
described briefly as follows.
The reactor internals were dismantled using the underwater plasma torch operated by
the robotic manipulator during 1987 to 1990. The piping connected to the RPV were then cut
and removed using the rotary disk knife, shaped explosives and conventional tools in 1990.
The RPV was dismantled using the underwater arc saw cutting system during 1990 and 1991.
After removing all steel components in the reactor cavity, the biological shield was dismantled
using three techniques: mechanical cutting (diamond sawing and coring), abrasive water jet
cutting and controlled blasting during 1991 to 1994. In parallel with the above dismantling
activities, the components in auxiliary buildings were removed using conventional cutting
tools. For example, the turbine was dismantled by using oxi-acetylene gas torch, in-air
plasma torch and conventional cutting tools. The dismantled components were put into mainly
1 m3 steel containers. The steel liner of the spent fuel storage pool was also removed to be
put into containers. The activities on decontamination of building inner surfaces and final
survey of radioactivities were on going for cancellation of radiation control area. The JPDR
decommissioning project will be finished by March 1996.
JPDR DISMANTLING ACTIVITIES
In the JPDR decommissioning project, various kinds of dismantling techniques were
developed and these were applied to the actual dismantling activities. Table 2 lists the
techniques developed in this program.(3H5) For example, the underwater arc saw cutting was
applied to dismantling the RPV by in-situ underwater cutting. This was the first case in the
world. In this section, the dismantling activities will be described focusing on the techniques
developed in the JPDR decommissioning program.
28 -
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JAERl-Conf 95-015
Dismantling Techniques to Steel Components
•Underwater plasma arc cutting : The plasma torch (1,000 A in max, AT + H2 for plasma gas)
is possible to cut 130 mm thick stainless steel components underwater. The torch was
mounted on two types of remote operation systems: master/slave robotic manipulator and mast
type manipulator systems. Both of them were controlled outside the R/E. The robotic
manipulator was applied to dismantling of the core spray, the feedwater sparger and the core
grid support bolts, and the mast type manipulator was applied to the other parts of reactor
internals.
•Shaped explosives : The shaped explosive is cylindrical with a copper-lined V-poove on
the circumferential surface. When the explosive is detonated, the copper liner collapses and
accelerates in the direction of the pipe wall to be cut. It is easy to use the remote setting on
a cutting portion to separate a pipe by detonation. The shaped explosives were applied to
cutting of the pipes in the core spray system and the emergency condensate system (26 inches
diameter, schedule 80 thick carbon steel pipes). Also, these were applied to cutting the control
guide tubes underwater. The in-air cuttings were successfully completed, however, the
underwater cutting was not successful due to decreasing the detonation force by the water
between the explosive and the pipe wall.
* Rotary disk knife : The cutter part of this system consists of a small disk knife and two
rollers. The edge of the disk knife is pressed against the inner surface of pipe to be cut, and
is rotated by the driving shaft connected to the hydraulic motor. Two machines were
fabricated for cutting the pipes in the recirculation system (318 mm diameter, 17 mm thick)
and the feedwater/unloading closed cooling system (114 mm demeter, 9 mm thick). The larger
disk knife is available for both straight line and curved line with the elbow. In the case of
cutting the elbow part, the cutter part could not be inserted into the elbow part due to the
projection which had been made by welding, resulting in failure of cutting.
» Underwater arc saw cutting : This system consists of the base support structure, the main
mast, the saw blade and the blade drive mechanism. The base support structure mounted mast
rotating/lifting mechanisms which controls the positions of the saw blade circumferentially and
vertically. The main mast is equipped with three outriggers which fixes their movement. The
saw blade and its driving mechanism are mounted at the lower end of the mast, and driving
mechanism rotates the saw blade by hydraulic pressure and changes the direction of the saw
blade horizontally and vertically. This system operated remotely outside the reactor building.
The maximum arc power was 40,000 kA at 50 V direct current, and could cut 250 mm thick
carbon steel plate underwater. The top flange portion of the RPV body was cut vertically into
9 pieces. The other parts were cut into 8 horizontal pieces and 9 vertical pieces.
Dismantling Techniques to Concrete Structures
Three techniques: mechanical cutting, abrasive waterjet cutting and controlled blasting
were developed to dismantling concrete structures. Figure 2 shows cross-sectional view of
the biological shield showing the area classification related to the dismantling techniques.
» Mechanical cutting : In the mechanical cutting, concrete blocks were removed by sawing
and coring. The machine consists of sawing blade and coring cylinder. It was installed into
the reactor cavity by wires from the bridge at the service floor. The vertical position of the
machine was changed depending on the height of dismantling portions. It was horizontally
fixed with three outriggers, and was rotated circumferentially by the driving mechanism. The
upper protrusion of the biological shield was dismantled using this machine. The averaged
cutting efficiency was 1.3 m2 per hour.
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JAERI-Conf 95-015
* Abrasive water jet cutting : The water jet cutting system consists of the nozzle of water jet,
the nozzle drive mechanism, the supersonic wave sensor to confirm the cutting completion,
the supporting assembly, pumps and abrasive supply mechanism, etc. The nozzle of water jet
could be moved horizontally and vertically by the drive mechanisms. The pressurized water
up to 2,000 kg/cm2 was projected into the concrete surface to cut the concrete structure with
reinforcing bars. It was cleared that the system could cut the reinforcing concrete structures
up to 60 cm in thickness. It was applied to dismantling of the lower protrusion of the
biological shield,
• Controlled blasting : In the controlled blasting, the explosives which installed in holes of
about 3 cm in diameter were detonated sequentially with time interval of blasting to control
the direction of the .mechanical force. The explosives were vertically charged to the inner
layer of the biological shield, while these were charged horizontally to the outer layer of the
biological shield. Urbanite, an improved form of Dynamite was used as the explosives due
to relatively slow detonation velocity.
WASTE MANAGEMENT
Waste Categorization
The wastes resulting from decommissioning are basically categorized into activated,
contaminated and non-radioactive wastes. Further the wastes of contaminated and activated
materials were classified into four levels as follows.
(a-i) >4KBq/g, (c-i) >400 KBq/cm2
(a-ii) 4K-4QBq/g, (c-ii) 400 KBq/cm2
(a-iii) 40 - 0.4 Bq/g, (e-iii) 4 K - 40 Bq/cm2
(a-iv) <0.4 Bq/g (e-iv) <40 Bq/cm2
Where the levels, (a-i) through (a-iv) are applied to the activated components and the
levels (c-i) through (c-iv) contaminated components and building surfaces. More detailed
levels were considered in some case.
Before starting the dismantling activity at each work area, the components to be
dismantled were classified into the groups based on the on-site measurement of radioactivity
levels. The wastes were put into 200 litter drums, steel containers (1 m3 or 3 m3), or shielded
containers depending on their radiation levels. The wastes were also separately packaged in
terms of material and component types. The containers filled with wastes were temporarily
placed at the temporal storage yard in the JPDR facility, then these were transported
periodically to the JAERI's waste storage facility. In the dismantling activities, various kinds
of by-products such as vinyl sheet, gloves, shoe covers were produced. The by-products were
put into carton boxes, of which size is approximately 30 cm in diameter and 50 cm in height.
The carton boxes were transported periodically to the waste treatment facility for incineration
and compaction, where these were treated together with the wastes produced by another R&D
studies of JAERI.
In addition to the solid wastes, gaseous and liquid wastes were also produced from the
dismantling activities. Gaseous wastes were filtered and exhausted through the stack after
confirming that the radiation level was less than 3 x 10~6 Bq/cm3. Liquid waste was treated
by the water treatment systems, which were used during the operation of the JPDR. After
confirming that its radioactivity was less than 0.4 Bq/cm3, it was diluted with water to be one
hundredth in radioactivity, then discharged to the Pacific ocean.
Wastes Arisings
Approximately 4,600 tons of waste have arisen as of March 1994. Among the wastes,
the radioactive waste weighed about 3,500 tons. The weight breakdown by materials are 1,100
tons in metal, 2,000 tons in concrete and 230 tons in by-products. The waste weight
- 30-
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JAERJ-Conf 95-4)15
breakdown by radioactivity levels are also listed as follows :
a-i) 41 tons, a-ii) 146,1 tons, a-iii) 410.2 tons, a-iv) 1872.8 tons,
c-i) 0 tons, c-ii) 34.3 tons, c-iii) 190.7 tons, c-iv) 1415.0 tons
Figure 3 shows the flow of the waste treatment in the JPDR decommissioning program.
ShaliowJLand Burial
Due to the revision of the measures for implementing the Nuclear Regulation Low, it
becomes possible to dispose concrete waste in extremely low levels into onsite shallow land
burial place. The most wastes resulting from the demolition of outer layer of the biological
shield and from decontamination of the building surfaces were put into flexible containers,
which were placed at temporary storage yard. These will be disposed at the shallow land
burial place.
The finai radioactivity survey has been conducted after decontamination of the building
inner surface. First, radioactivity was surveyed to each area by contamination detector (conta-
mat), then samples were taken from each area for confirmation of that the radioactivity of
cobalt-60 is low (3 Bq/kg) enough compared with the natural background. The final
confirmation of radioactivity will be conducted in such a way that one sample per each 10 m2
area at penetrated contamination areas or one sample per each room at surface contamination
areas will be taken and the radioactivity of the sample will be measured.
DATA COLLECTION AND ANALYSIS
Information about the JPDR dismantling activities have been collected and accumulated
in the decommissioning database. The data have been collected on data collection and
retrieval systems, that use the JAERI mainframe (FACOM - M780) and minicomputers. In
addition, information about machine performance and operability were also collected when
newly developed decommissioning techniques were applied to the dismantling activities. The
data on project management such as waste arising, manpower expenditure and radiation
exposure of workers are analyzed and reported as a periodical documentation. These are used
for managing the JPDR dismantling work. Figure 4 shows the concept of data collection and
retrieval systems developed for the JPDR decommissioning program. The breakdown of the
worker dose measured m the dismantling activities is shown in Fig. 5 as an example of the
data collected in the JPDR dismantling project.
CONCLUDING REMARKS
The JPDR decommissioning program is in the final stage; various dismantling
techniques were developed and demonstrated in the actual dismantling of the JPDR. Various
data on project management and performance of dismantling machines developed in Phase-1
have been collected in the dismantling activities, and the data were accumulated in the
decommissioning database.
These data and know-how obtained in the dismantling activities will be useful for
future decommissioning of commercial nuclear power plants.
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JAERI-Conf 95-015
REFERENCES
(1) S. Yanagihara, M. Tachibana, H. Ogihara and Y. Seiki: The Japan Power Reactor
Demonstration Decommissioning Program - Experience and data analysis on dismantling
procedures -, Proceedings of the 1994 International Symposium on Decontamination and
Decommissioning, April 25-28, 1994, Knovville, Tennessee, USA
(2) S. Yanagihara, et al.: Systems Engineering for Decommissioning the Japan Power
Demonstration Reactor, JSME Inter. J., Series B, Vol.36, No.3, 1993
(3) S. Yanagihara, et al.: Dismantling Techniques for reactor Steel Structures, Nucl. Tech.
Vol.86, Aug. 1989
(4) S. Yanagihara, et al.: Dismantling Techniques for Reactor Steel Piping, Nucl. Tech.
VoL86, Aug. 1989
(5) H. Nakamura, et al.: Cutting Technique and System for Biological Shield, Nucl. Tech.
Vol.86, Aug. 1989
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JAERI-Conf 95-015
Table I Major Specifications, Operation History
and Radioactive Inventory of JPDR
Specifications
Type of reactor
Thermal power
Pressure vessel
material
inner diameter
height
thickness
Biological shield
material
thickness
Reactor enclosure
inner diameter
height
Operation history
Operation time
Output of electricity
Radioactive inventory
BWR
90MWt (45MWt initially)
ASTM-A302-56GrB
2.1 m
8.1m
7 cm
reinforced concrete
1.5 to 3 m
15m
38m
17,000 hours
1.4 x 10fi KWH
131 TBq
(as of April, 1988)
Table 2 Dismantling Techniques Developed in R&D (Phase 1)
Technique
Components to be dismantled
Cutting ability
Underwater arc saw Reactor pressure vessel (RPV)
Underwater plasma arc Reactor internals
Rotary disc knife Piping connected lo RPV
Shaped explosive Piping connected to RPV
Mechanical catting Biological shield
Abrasive water jet Biological shield
Controlled blasting Biological shield
250 mm thick carbon steel in water
130 mm thick stainless steel in water
12 inch8, Sell 160 (stainless steel)
26 inch, Sch 80 (carbon steel)
cutting efficiency : 1.3 m2 per hour
450-600 mm thick reinforced concrete
Blasting effieienc : 0,1 m3 per hour
1 inch = 2.54 cm
-33-
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1i87
1988
1989
1990
1991
1992
Equipment in R.Build. Equipment in Reactor Building
I
Components in Fuel Pool
BBro^Mi
Reactor Internals ReactoMnternals
sSMSS™HiM§£l
Pipmgto RPV
Reactor Pressure Vessel
i
Bioloqical Shield
RPV Head
B I
1993
1994
199S
1996
Equipment in Other Buildings
1
pump-condenser)
pjiiig:ii;j Demolition
Itlil
Reactor Containment
(Radwaste)
(Control)
DemQlition
As of September 1994
I
r>
o
VO
I
Fig.l Time schedule of the JPDR decommissioning program
-------
EL 18.65m
00
en
40 Bq/g
dismantled by controlled blasting
max. radioactivity:4Bq/g
waste arising:metal 34ton
concrete 1190ton.
dismantled by controlled blasting
max. radioactivlty;4000Bq/g
waste arislng:metal 63ton
concrete 180ton
dismantled by mechanical cutting
. rradioactivity:40-7000Bq/g
[waste arising:9ton
dismantled by water jet cutting
rradioactivity:40-70QOBq/g 1
[waste arising:25ton j
* The waste will be disposed in
shallow land burial place.
JO
Flg.2 Area classification and applied techniques to dismantling biological shield
-------
waste arisings from
JPDR decommissioning
I
I
radioactive waste
I
I
1
non-radioactive waste
relatively high
radioactive waste
1
relatively low
radioactive waste
metal
metal
concrete
1
by-products
shielded containers
I
containers
I
recording
temporal storage in JPDR site
transportation
recording
storage
3m container
1 rrfeontainer
200 litter
drum, etc
packaging
recording
temporal storage
in JPDR site
transportation
recording
ra
yo
h
Fig.3 Flow of waste treatment in JPDR dismantling activities
-------
CO
I
JPDR
Dismantlement
Radiation Control
Minicomputer
Mainframe
M-780
(Fujitsu)
Dismantling Activities
Decommis
stoning
Database
/Terminal
(input by operator)
Waste Management
Output
/ Tables \
\ Figures/
Waste
Container
Storage
Transportation
I
h
§
* Alarm Pocket Dosimeter
Fig.4 Concept of data collection and retrieval systems
-------
components in
reactor building
10,83
biological shield
28,53
components in auxiliary buildings
14.67
supervision of
dismantling work
7.83 reactor internals
73.11
305.83
man-mSv
reactor pressure vessel
107.62
pipes connected to
reactor pressure vessel
63.24
as of end of March, 1994
max. individual dose ; 8.52 mSv
en
93
•Vmi
I
Fig.5 Radiation exposure of workers measured in various dismantling activities
-------
JAERI-Conf 95-015
1-4
THE DECOMMISSIONING PROGRAM OF JAERI'S REPROCESSING TEST FACILITY
Tadaaki Uchikoshi, Takeo Mimori, Yukio Iwasaki, Akira Ito
Nuclear Fuel Facility Decommissioning Technology Division
Department of Decommissioning and Waste Management
Japan Atomic Energy Research Institute
1. ABSTRACT
Decommissioning program of JAERPs Reprocessing Test Facility (JRTF) has
been carried out to establish decommissioning techniques for nuclear fuel
facilities. The project consists of 2 phases ; phase 1 is preparatory stage of
decommissioning project, and phase 2 is execution stage of the JRTF
deconmissioning. The project started in 1990 under a contract with the Science
and Technology Agency, and will be finished in 2001. Up to now, treatment of
some radioactive liquid waste and physical inventory estimation were carried
out. In addition to the technical development for dismantling, the design for
treatment of the unpurified uranium solution and high level liquid waste are in
progress steadily.
2. INTRODUCTION
JAERPs Reprocessing Test Facility (JRTF) was constructed in 1966 to carry
out the first fundamental reprocessing studies on PUREX process in Japan. JRTF
was operated from 1968 to 1969 and reprocessed the spent U/A1 metal fuel from a
research reactor JRR-3. In JRTF, early reprocessing tests of spent fuel were
carried out successfully, and about 200g of plutonium was recovered. The
reprocessing equipments were officially shut down in 1970. After the shut down,
a part of the facilities has been kept open and used for the studies on the
measurement of fuel burn up rate, treatment of liquid waste, and other tests.
Nuclear fuel facilities have not been decommissioned except for those
remodeling and dismantling of individual facility in Japan. However, in the
near future, it is indispensable theme that nuclear fuel facilities will be
decommissioned. Therefore, with a view to carrying out development of
decommissioning technology for actual nuclear fuel facilities immediately,
research and development of decommissioning technology for nuclear fuel
facility has been carried out using JRTF following JPDR decommissioning since
1990.
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JAERI-Conf 95-015
3. OUTLINE OF JRIF
JRTF consists of a main building for reprocessing and two annex buildings
for storing liquid waste, and these buildings are connected by a few ducts.
Fig.l shows a plot plan of JRIF.
3.1 Main Building
The main building is three stories above and one under the ground, and its
total floor area is about 3,000 m2. There are various kinds of cells such as a
main cell (about SmxSmxlOm, 1.7m wall thickness) in which dissolver and pulse
colimns are installed, a plutonium purification cell (about 5m xSmxfim, 0.4m
wall thickness) in which concentrator and mixer-settlers are installed, a
solvent recovery cell (about 4mx4mx4m, 0.5m wall thickness) in which solvent
washing decanters are installed, and 11 lead cells for analysis (1 m3 volume),
etc. Fig.2 shows the cross-sectional view of the main building.
3.2 Annex Building A
The annex building A is one story above and two under the ground, and its
total floor area is about 160 m2. In the building, there are 12 tanks which
store low level liquid waste generated from the reprocessing tests. Fig,3
shows the cross-sectional view of the annex building A.
3.3 Annex Building B
The annex building B is one story above and one under the ground, and its
total floor area is about 400 m2. In the building, there are 6 tanks which
store liquid waste such as Al-decladding liquid waste and high level liquid
waste generated from the reprocessing tests. Fig.4 shows the cross-sectional
view of the annex building B.
4. DECOMMISSIONING PROGRAM
The decommissioning program consists of 2 major phases which are preparatory
stage and execution stage. Phase 1 ; preparatory stage consists of treatment
of liquid waste, preparatory survey of facility, and technical development for
dismantling . Phase 2 ; execution stage is decommissioning of entire facility.
Fig.5 shows a schedule of the JRIF decommissioning program.
4.1 PHASE 1 ; Preparatory stage of decommissioning project
4.1.1 Treatment of Liquid Waste
Total amount of about 70m3 of different types of liquid wastes generated
- 40-
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JAERI-Conf 95-015
from the reprocessing tests was stored in 16 storage tanks. Liquid wastes were
classified in 5 categories as shown in Table 1.
In case of decommissioning of the JRTF, it is indispensable to treat liquid
waste. Therefore, these liquid wastes have been treated in parallel with
preparatory survey and technical development. The outline of treatment is as
follows.
Al-Decladding LiquidJfaste
Treatment of the Al-decladding liquid waste was completed by the
biturainization process in 1986.
a-Contaminated Waste
Treatment of the a-contaminated waste is in progress. The waste is
separated into supernatant and sludge. The supernatant is solidified by the
bituminization process and the sludge is vitrified by microwave heating.
• Separation into Supernatant and Sludge
The a -contaminated waste is separated into supernatant and sludge using
TRU Waste Treatment Apparatus (TWA) with coagulation and sedimentation
process. In this process, NaOH orUNO3 is added in the waste in order to
adjust pH, and at the pH of about 11, ferric nitrate and organic polymer are
added to accelerate the coagulation with them. Then the supernatant is
transported and solidified by bituminization process in the radioactive waste
treatment facility. The sludge is solidified as follows:
• Solidification of Sludge
The sludge generated from TwTA is solution in which main ingredients are
iron, sodium nitrate, and a nuclides. The a-contaminated solution should
be solidified for interim storage. Therefore, the sludge is vitrified with
TRU Sludge Solidification Apparatus (TRUSSAR). In vitrification process,
glass tablets which are the mixture of diboron oxide (B203), zinc oxide (ZnO),
and calcium oxide (CaO) are added in the sludge, the sludge is melted by
microwave and electrical heater using in-can melt method.
Spent Solvent
The spent solvent is washed by sodium carbonate in order to remove
plutonium. The washed spent solvent is added calcium octylate and incinerate.
The ash is solidified in cement.
The treatment apparatus has been manufactured and the treatment of spent
solvent started in middle of 1994.
_ 41 _
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JAERI-Conf 95-015
Unpurified Uranium Solution
The unpurified uranium solution was generated, because the JRTF had no
uranium purification process. Cesium and plutonium should be removed from the
solution in order to solidify and store in JAERI. Therefore, the unpurified
uranium solution is treated by adsorption process using inorganic adsorbent to
remove the main inpirities of plutonium.
The basic design of the treatment system was completed in 1991 and the
apparatus for this treatment will be manufactured in 1995.
High Level Liquid Waste
The high level liquid waste was generated from co-decontamination process.
It had the highest inventory of radioactivity in Tokai Research Establishment of
JAERI. Removal of plutonium, cesium, and strontium from the high level liquid
waste is also necessary for the interim storage. Therefore, the high level
liquid waste is treated by adsorption process using inorganic adsorbents.
The design for the treatment of the high level liquid waste has been
investigated in detail since 1992.
4.1.2 Preparatory survey
Before start of decannissioning of entire facility, it is indispensable that
survey of characteristic of facility and design of dismantling procedure are
carried out. Therefore, followings investigation have been carried out.
• Estimation of total weight of dismantling waste and its radioactivity
Based on the record of apparatus and contamination data in the JRTF, the
total weight of dismantling waste was estimated. The penetration of
contaminant in concrete was surveyed to measure radioactivity of the
concrete cores which were bored from concrete surface. The detail data of
inventory of radioactivity is scheduled to be accumulated in the future.
• Basic planning of JRTF decommissioning project
Considering characteristics of JRTF and the present situation for
dismantling techniques, basic dismantling procedure for facility was
investigated.
4.1.3 Technical development for dismantling
In the decommissioning of reprocessing facilities, following characteristics
which are different from that of nuclear reactors should be considered:
• There are wider variations in the facility structure and shape. Apparatus
- 42-
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JAERI-Conf 95-015
and equipment are complicated and different from each other.
• The facilities are contaminated by a-emitters such as plutoniura and many
other transuranium (TRU) nuelides.
• The chemical forms of nuclear fuel materials are in variety.
Considering above characteristics, the following technical developments have
been carried out to use dismantling of entire facility.
• Deeomnissioning program by 3D-CAD system.
Dismantling simulation program, waste control program, and database, have
been developed to make the design of dismantling techniques systematic and
efficient.
•Remote controlled data acquisition system
A robot acquiring three dimensional data of dismantled object is being
developed for the place where radioactivity is high. The acquired data
will be processed by the 3D-CAD system, and will be used for the design of
dismantling. Fig.6 shows the outline of this system.
• Protective equipment for a-contamination
Protective equipment has been developed to improve that of safety and work
efficiency ; protection for radioactivity, strength of materials, heat-
resisting property, to make easy putting on and taking off, etc..
• Remote control dismantling system for large-sized vessels
Compact remote control dismantling system which has the function of washing,
cutting, packaging, carrying, etc., is being developed. Fig. 7 shows the
outline of this system.
• Decontamination of concrete surface layer by laser treatment
By scanning high power density laser beam on concrete surface, followings
principles are appeared ; the constituents of concrete melt instantaneously
to form a glass, concrete surface tear off and hop-off owing to rapid
pressure of evaporation. Using these characteristics of laser, the
technology for decontamination of concrete surface layer will be developed.
4.2 PHASE 2 ; Decomnissioning of Entire Facility
The JRTF will be actually dismantled using the techniques developed in
phasel. The developed techniques will be verified and data and experience of
dismantling will be accumulated through the JRTF dismantling. The JRTF
decommissioning will start in 1995 from a part where dismantling is possible,
and will be completed in 2001. The budget for the JRTF decommissioning is about
100 million dollars including treatment of liquid waste and development of
dismantling techniques.
_ 43 _
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JAERI-Conf 95-015
5. Conclusion
Nuclear fuel facilities have not been decommissioned except for those
remodeling and dismantling of individual facility in Japan. Therefore, the wide
variety of the results obtained through the JRTF dismantling are expected to
make a valuable contribution to deconraissioning of nuclear fuel facilities.
44-
-------
Table 1 Radioactivity and Volume of Liquid Waste
Categories
Al-decladding liquid waste
or-eontaminated waste
Unpurified uranium solution
Spent solvent
High level liquid waste
Radioactivity (Bq/l )
ex
2.2x10*
3.7X101 ~ 1.1X107
1.4X107
4.8X104 ~ 1.4X106
3.6X108
0 (r)
3.2x10*
1.3X103 ~ S.2X107
9.8xl06
6.3xl04
7.4X109
Volume
(M3)
0.8
56.5
1.7
1.7
11.0
tn
2
^
o
B,
-------
JAERI-Conf 95-015
Annex building B
Annex building A
Fig. 1 Plot plan of JKTF
Fig.2 Closs-sectional view of the main building
Fig.4 Closs-sectional view of the annex building B
Fig.3 Closs-sectional view
of the annex building A
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JAERI-Conf 95-015
FY
PHASEl
(1) Treatment of Liquid Waste
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JAERI-Conf 95-015
Vision
Equipment
Pan/Tilt Reehanism
(Three-axis)
TV Camera for
Normal Viewing
Laser Pointer/Range
Finder
TV Camera for
Close up Viewing
Lower Leg (Shank)
Structure
Upper Leg (Tight)
Structure
Fig.6 Remote Controlled Data Acquisition System
Remote Exchange Devices
(for Decontanination, Cutting, and collection}
• Washing Device
• Plasma Torch Device
• Grinder Device
• Crabbing and Shearing Device
• Bacuin Pad Device
ITV Camera
Five-axis Movement System
Capturing Device
Tor Secondary Products
Cut Peace Transfer System
Large-sized Vessel
Remote Control System
Fig.7 Remote Control Dismantling System for Large-sized Vessels
-48 -
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JAERI-Conf 95-015
1-5 The Decommissioning' Plan of the
Nuclear Ship MUTSU
M,Adachi, LMatsuo, S.Fujikawa, T. Nomura
Department of Nuclear Ship Decommissioning
Japan Atomic Energy Research Institute
Abstract
This paper describes the review about the decaniiss'toning plan and present
state of the Nuclear Ship Mutsu. The decommissioning of the Mutsu is carried out by
Reioval and Isolation method. The procedure of the decommissioning works is presented
in this paper. The' decommissioning works started in April, 1992 and it takes about
four years after her last experiiental voyage.
1. Introduction
The Mutsu is the first nuclear ship in Japan. Her reactor reached criticaiity
in 1974,but due to the necessary modification on the radiation shielding and public
debate on the reactor's safety, full operation of the reactor was protracted for
several years. Full power operation finally realized in 1990, and the experimental
voyage started in July, 1990 and completed in December, 1991, collecting various
characteristic data.- The Mutsu sailed about 82,000km during the experimental voyage.
After completing her experimental voyage, the decommissioning program was commenced
in accordance with the scheme previously decided by the Japanese government. The
chronology of the Mutsu for research and development is shown in Fig,1. General
arrangement of the Mutsu is shown in Fig._2. machinery arrangement in Fig. 3 and
technical specifications in Table 1.
2. Decommissioning plan of Mutsu
In general, it is said that there are Mothball ing, Entombment and Dismantling/
Removal as the methods of the nuclear plant decommissioning. Various plans were
considered in JAERI with reference to the above methods. As a result, it was decided
that the reactor room (see Fig.2 or 3) will be separated from the hull of Mutsu and
stored in the building. This method is called Reioval and Isolation method or
One-Piece Removal method. u This method has some advantage compared with others in
safety, suit for reuse, short period,. little waste, low exposure and inexpensive.
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JAERI-Conf 95^015
The guideline for the decommissioning of the Hutsu is as follows,
(1) Cooling of the spent fuels
(2) Unloading the spent fuels and the neutron sources from the core
(3) Removing of the coiponents in the reactor auxiliary room
(4) Dredging and digging in the Sekinehana home port
(5) Construction of the storage building
(6) Separating the reactor room from the hull of the Mutsu on a semi-subaergible
barge
(7) Transferring the reactor rooi with the reactor by a floating crane to the
storage buiIding
Decoiaissioning procedure of the Mutsu is shown in Fig. 4. Removed reactor room
will be exhibited to the public at the storage building, auseum. After taking off
the reactor space, the Mutsu fill be delivered to JAMSTEC (Japan Marine Science and
Technology Agency) to convert her into a large sized ocean research ship.
3. Present stage of decommissioning
(1) Cooling of the spent fuels and (2) Unloading the spent fuels and the neutron
sources from the core finished in Noveiber, 1993. The each spent fuel was picked up
from the core with the fuel handling equipment and transferred to a storage building
by the spent fuel transport cask. The core is similar to that of a land-based PfR
with 32 fuel assemblies. To defuel, the enclosed space above the reactor was floored
and the vessel head and upper internals were renoved. These operations were performed
in a Uaporary structure placed on the ship's deck above the reactor compartment.
Spent fuel unloading forks are shown in Photo. I, 2. (3) Eemoving of the components in
the reactor auxiliary room was finished in September, 1994. The ton-exchangers,
sumpling and charging pumps, exhaust filters and other various equipment were reioved
from the reactor auxiliary room. Every parts were divided with separate activity
levels and waste types and enclosed into the drums. Removing works of the components
are shown in Photo. 3,4. (4) Dredging and digging in the Sekinehama home port finished
in October, 1994. The sea-bed was dredged and the coast was dug of some area in
Sekinehaia home port for transferring the reactor rooi to the storage building by a
floating crane. (5) Construction of the storage building is now in progress. Present
status of the coast and storage building is shOTO in Pho_to.,5.6.
The decoraission ing schedule of the Mutsu is shown in Fj&JL (6) Separating the
reactor room from the hull of the Hutsu on a semi-submergible barge and
(7) Transferring the reactor rooi with the reactor by a floating crane to the storage
building is scheduled to be carried out around the summer of 1995.
4. Conclusion
The decommissioning plan and present state of the nuclear ship Mutsu was
reviewed. Decomiissioniog works including construction, dredging and digging are now
going in accordance with our tentative schedule. In the result of this decoimission,
a large amount of radioactive waste was generated. A large portion of
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JAERI-Conf 95-015
this ?aste is stored in the storage building with the reactor room. Some remaining
taste including soie equipient components, resin, filters and other lot level
contaminated parts will be stored in another building or reused in our site. Spent
fuels will be reprocessed at Tokai reprocessing plant in near future. All of this
decommissioning forks fill be finished at the end of 1995.
REFERENCE
1) N.ONISHI, et al,- "Investigations Related to a One-PSece Removal of the
Reactor Block in the Fraie of JEE-3 Reconstruction Program," Int. Decoimissioning
Symposium, Pittsburg Pennsylvania USA, 4~8 Oct. (1987)
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in
(O
q •/ r
U I J,
N, 8.MUT8U was buill,
I1' ue Is were loaded.
Power-up to sis be/ian on lljc
nor Llicrn I'ac i f i c Ocean ;
Achieved cr i t'i ca I i ly (Aiij, r/,8) ;
K
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JAERI-Conf 95-015
N S " M U T S U'
2 Genera
and ?.
- 53 -
-------
(MQW) l'*illl!!i!!(.-)LESIlJ.!!£'lii1]U'!!
I—|—i Kitici'Wiiu:!' swilL-lilxjarii
I I -LU | ,
O|I|H:I- ili.'ili I Itnl wlr ik
tn
•U
inlrol anra vwili la!.imi cm
El!B ifl!LJ12 J—v—
H,iin switchboard
•••./• «[-i
"r«»liinjsL fillor
•-"-.'.I-, f^--—•'i f-lrl
norf.oncy walur sUinij;u Ik
•lih il«!i:
Qjiiiiwmmt cootinr.^C
;i wlr
Ion Kxcliiiti|'l(!r(V.)
^mtaHB
hwer ili.'ili 11 IN! *lr Ik
IV spray |nini|i(2} Ji
,'iUii':i|j(! lank
3 Machine;i:y ai.Tanp.onuMil: Inside N . !J. MU'i.'SU
-------
I
tn
7.4 „
"•'I""*
[j v^.---—- __.•_.-__.._
=r.^ Removal of Auxiliary System
Steel Containers
Cul of Piping
Shielding of Penetrations
RenewGd Ship
riiiiiilpili Hfliimr
Lifting of Ship by Barge
Separation ol Reactor Room
New Structures
Flonlincj Crane
"«*l*"«nr *""""" ~""~m— • -«-<*——: ~1"1-1— ~" ,...»—. .,**-.
nil.
Storage Building
Lifting of Reactor Room
"ranofor to Storage Building
a
\o
Fig. t\ Decommissioning Procedure of MUTSU Reactor
-------
"^--^ FY
ITEM^"*--*^_^
Dec omm i s s -
i o n i n g of
MUTSU
Construct-
ion of
storage
Bl d.
Dredging
and
Digging
1991
E x p e r i m- Cc
e n t a I
SI
voyage
111
r T
1992
>o 1 i ng of Ut
0
••ff
1993
t 1 o a d i ng Rer
SF of
. « a (
M
^m
ira
1994
i
9 9 5
loving Removing R e n e N
Compo- of Reactor of s'l
its i Room i t
^•f(_
••z
••I
V
y
A
V
(
N
Transfer
7 K
N
/
/
1996
/ i n s
\ i p
* i ng
en
g
Fig. 5 Dec omm issioning Schedule of Mu t s u
-------
Table 1 Specifications for "MUTSU1
Purpose
Navigation area
length overall
Breadth (molded)
Depth (molded)
Full-load draft
{molded!
Gross tonnage
Main engine
Typo x number
Output
Nuclear-powered research vessel
Ocean going areas
130.00 m
19.00m
13.20 m
6.90 m
8,242 t
Steam turbine x 1
10,000 pS
Reactor
Type x number
Heat output
Fuel
Service spued
Cruising range with
nuclear power
Structure
Equipment
Pressurized light-water moderated
reactor x 1
Appro*. 36,000 kw
Uranium oxide (low enriched uranium}
16.5 knots
ApproK. 145,000 nautical miles
Special requirements concerning subdivi-
sion and stability
Anti-collision and stranding structures
Automatic radar plotting aid
International maritime satellite system
m
I
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JAERl-Conf 95-015
Photo. 1 Unloading of the containment vessel cover and spent fuel
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JAERI-Conf 95-015
Photo.2 Transportation and Storage of the spent fuels
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Photo.3 Labeling for the pipes and Removing of the charge
pump
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JAERI-Conf 95-015
Photo. 4 Unloading of the gamma shielding gate and ion-exchange tank
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JAERI-Conf 95-015
Photo.5 Storage Building
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JAERl-CoEf 95-015
£*, ^"""SS"'—<**•• SfJ?
r%^£-i^-i^sS"^?-*;
mat•Z^* '* ¥•- T-v.*-
Photo, 6 Storage Building (from sea side)
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JAERI-Conf 95-015
2. Cleanup and
Residual Radioactivity Criteria
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JAERI-Conf 95-015
2-1 Surface Radiological Free Release Program
for the Battelle Columbus Laboratory Decommissioning Project
Presented by: Cort N. Horton, Battelle Columbus Operations
For: The Second EPA/JAERI Workshop on Residual Radioactivity and Recycling
ABSTRACT
This paper was prepared for the Second Residual Radioactivity and Recycling Criteria
Workshop and discusses decommissioning and decontamination activities at the
Battelle Columbus Laboratories Decommissioning Project (BCLDP). The BCLDP is
a joint effort between the Department of Energy (DOE) and Battelle Columbus
Operations to decontaminate fifteen Battelle-owned buildings contaminated with DOE
radioactive materials. The privately owned buildings located across the street from
The Ohio State University campus became contaminated with natural uranium and
thorium during nuclear research activities.
BCLDP waste management is supported by an extensive radiological free-release
program. Miscellaneous materials and building surfaces have been free-released from
the BCLDP. The free-release program has substantially reduced radioactive waste
volumes and supported waste minimization. Free release for unrestricted use has
challenged regulators and NRC licensees since the development of early surface-
release criteria. This paper discusses the surface radiological free-release program
incorporated by the BCLDP and the historical development of the surface radiological
free-release criteria. Concerns regarding radiological free-release criteria are also
presented.
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JAERI-Conf 95-015
SURFACE RADIOLOGICAL RELEASE PROGRAM
FOR BATTELLE COLUMBUS LABORATORIES
DECOMMISSIONING PROJECT
Introduction
The Battelle Columbus Laboratories Decommissioning Project is an extensive project to
remediate 15 buildings and associated facilities such that they can be released for unrestricted
use to the general public. This document establishes the technical basis by which present
radiological surface release criteria are applied to accomplish the remedial action.
Radiological release criteria for surfaces were first defined in ANSI N13.12 (draft), "Control
of Radioactive Surface Contamination on Materials, Equipment, and Facilities to Be Released
for Uncontrolled Use" and were later published in U.S. Nuclear Regulatory Commission
Regulatory Guide 1.86, "Termination of Operating Licenses for Nuclear Reactors". There is
little difference between these two standards, and both serve as guidance documents when
establishing surface radiological release criteria. The Department of Energy (DOE) Order
5400.5, "Radiation Protection of the Public and the Environment" also provides release
criteria for surface radioactivity and the criteria stated therein contain only subtle differences
from those in the other reference documents. For the BCLDP, DOE Order 5400,5 and
Regulation Guide 1.86 will be used to provide criteria that are followed as upper limits of
radioactive surface contamination for unconditional release of equipment, materials, and areas.
DOE Order 5400,5 requirements are mandatory for the BCLDP because the radioactive
material being removed is the property of the DOE and the DOE provides 90% funding for
the project. Regulatory Guide 1.86 requirements are mandatory because Battelle is an NRC
Licensee and the BCLDP is being conducted under an NRC required decommissioning plan.
Release Criteria
The surface radiological release criteria for the BCLDP are shown in Table 1, "Surface
Contamination Guidelines for BCLDP." These criteria are provided by DOE Order 5400.5
"Radiation Protection of the Public and the Environment," which reference Regulatory Guide
1.86. DOE Order 5400.5 does not define the release levels for nuclides such as transuranics
fTRU), Ra-226 and Th-230, therefore, BCLDP will adopt the guidance of Regulatory Guide
1.86 for these isotopes and these limits are also shown in Table 1. The criteria in Table 1 are
the maximum allowable quantities of radioactive material that may be left on surfaces of
equipment and buildings that are released to the general public for unrestricted use. The term
"Unconditional Free Release" is a generally accepted term in industry that is used
synonymously with this unrestricted use. It is the policy of the BCLDP to aggressively apply
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JAERI-Conf 95-015
the principles of As Low As Reasonably Achievable (ALARA) to release criteria. The
release criteria stated in Table 1 shall be applied as an upper limit of radioactive surface
contamination for free release of equipment, materials, and areas by BCLDP.
Release limits are grouped into several categories depending on the radiotoxicity of the
radioisotope as seen in Table 1. For each area, BCLDP will identify radioisotopes through
analytical techniques and determine their corresponding activity fractions. Release limits can
then be determined on the weighted activity fraction of each radioisotope. If radioisotopes or
activity fractions are not known or vary significantly, then release limits will be based on the
most restrictive nuclides to be encountered by the BCLDP. These release limits for gross
alpha and gross beta-gamma activity are shown in Table 2.
Radiological Release Logic for the BCLDP
Consistent with the requirements of DOE Order 5400.5, Section II.5, "Release of Property
Having Residual Radioactive Material," and the DOE Radiological Control Manual, Section
422, "Release to Uncontrolled Areas," ail facilities, areas, buildings, equipment, and
materials, having surface activity or activity concentrations in excess of applicable limits
(Table 1) shall require decontamination; and/or removal and disposal as radioactive waste.
Facilities, buildings, areas, equipment, and materials that do not have detectable
contamination, (i.e., above the Lower Limit of Detection [LLD]) and are not suspected of
potential internal contamination, shall be released without any further assessment or
evaluation. As required by DOE Order 5400.5, Section II.2.b, "ALARA Evaluations,"
formal ALARA evaluations and cost benefit analyses shall be performed as part of
Decontamination and Decommissioning Plans for facilities, buildings, and large volumes of
associated equipment and materials with residual radioactivity above the LLD.
As dictated in DOE Order 5400.5, Section II.5.C.1-4, "Release of Materials and Equipment,"
individual items not addressed by specific Decontamination & Decommissioning plans with
residual radioactivity above the LLD but below applicable limits will also be subjected to an
ALARA process and assessed for potential contamination prior to release by the BCLDP.
This ALARA process consists of a field assessment by a trained evaluator (usually a Health
Physics Supervisor) prior to releasing any materials or equipment with detectable
contamination below the limit. BCLDP Procedure HP-OP-Oll, "Release of Materials from
Radiologically Controlled Areas" further defines and establishes requirements for the
uncontrolled release of materials and equipment from radiological areas.
It is the practice of the BCLDP that all releases of buildings, areas, materials, and/or
equipment will meet or better the limits and criteria found in Table 1. In addition, for
buildings, areas, materials, and/or equipment with residual radioactivity below those in Table
1 but above the LLD radiation detection equipment, it is the practice of the BCLDP to use
sound ALARA principles and analyses to determine what, if any, decontamination actions are
warranted prior to release. It should be made clear this does not mean that all materials,
equipment, areas, or buildings will be released at or near the LLD since there may not be a
reasonable ALARA basis to do so.
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JAERI^Conf 95^015
Table 1. Surface Contamination Guidelines for BCLDP
> 'ii ?i*-WB»Sj*SB*:::S aSiSSS^iSS®";™;? wsSlxsw? « *§$™
= 1 1 HI
:! 1 lllll 1 ill;
Transuranics, 1-125, 1-129, Ra-226,
Ac-227, Ra 228, Th-228, Th-230, Pa-231
Th-Natural, Sr-90, 1-126, 1-131, 1-133,
Ra-223, Ra-224, U-232, Th-232
U-Natural, U-235, U-238, and associated
decay product, alpha emitters
Beta-gamma emitters (radionuclides with
decay modes other than alpha emission or
spontaneous fission) except Sr-90 and others
noted above. ^
III 111 Hi
ill 111! Il^^^iUf^^iMll^^^S 111
ill nil in
Reserved
(100)*
1,000
5,000
5,000
ii^i|||iiiips^?lii
Reserved
(300)*
3,000
15,000
15,000
lUiBleiit
^•:':':-:-:::':::":::-:--:::>:''::'>:::-:-'::t':"::::. ::-:•>••:•:•><:
Reserved
(20)*
200
1,000
1,000
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 measured by an
appropriate detector for background, efficiency, and geometric factors associated with
the instrumentation.
Where surface contamination by both alpha- and beta-gamma-emitting radionuclides exists,
the limits established for alpha- and beta-gamma-emittiag radionuclides should apply
independently.
Measurements of average contamination should not be averaged over an area of more than
1 m2. For objects of less surface area, the average should be derived for each such object.
The average and maximum dose rates associated with surface contamination resulting from
beta-gamma emitters should not exceed 0.2 mrad/hr and 1.0 mrad/hr, respectively, at 1 cm.
The maximum contamination level applies to an area of not more than 100 cm2
The amount of removable material per 100 cm2 of surface area should be determined by
wiping an area of that size with dry filter or soft absorbent paper, applying moderate
pressure, and measuring the amount of radioactive material on the wiping with an
appropriate instrument of known efficiency. When removable contamination on objects of
surface area less than 100 cm2 is determined, the activity per unit area should be based on
the actual area and the entire surface should be wiped. It is not necessary to use wiping
techniques to measure removable contamination levels if direct scan surveys indicate that the
total residual surface contamination levels are within the limits for removable contamination.
This category of radionuclides includes mixed fission products, including the Sr-90 which
has been separated from the other fission products or mixtures where the Sr-90 has been
enriched.
Regulatory Guide 1.86
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JAERI-Conf 95-015
Table 2. Release Limits for Gross Activity (Unknown Isotopes) - Regulatory
EMISSION
Alpha
Beta - Gamma
REMOVABLE
(dpm/100 cm2)
*20
**200
TOTAL
(Fixed and Removable)
(dpm/100 cm2)
*100
**1000
* Based on TRU, Ra-226 and Th-230
** Based on Sr-90 and Th-232
Natural and Electronic Background
The application of release criteria standards cannot be successfully applied without the
understanding of background. Two types of background exist: natural and electronic.
Natural and electronic background significantly impact the release criteria by the following:
• Natural background, by providing a quantity of radioactive material which is
available to be detected,
» Electronic background, by influencing the least amount of radioactivity that can
be measured by a particular instrument.
It is necessary to distinguish the difference of the two types of background. Obviously, the
term background could apply to either. The following are the terms as accepted by industry
practice.
Generic term - natural background, the amount of radioactive material that exists in a
substance, surface, or material as a result of nature. The quantity of natural background
is generally expressed in terms of picocuries/gram (pCi/g), femtocuries/liter (10'15 Ci/1),
milligrams/milliliter (mg/ml), disintegrations per minute (dpm), or other suitable
combinations of activity or quantity per unit mass or area. Cosmic radiation is also
considered a part of natural background. Natural background is detectable and must be
accounted for when making activity determinations. For example, field beta/gamma
type instruments, might have background that ranges from 100 to 500 counts/minute.
For a laboratory type alpha scintillation counter, the background might be 1 count per 2
minutes of counting time.
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JAERI-Conf 95-015
..Generic term - electronic, background^ the amount of electronic signal produced by
electronic noise which results in a meter or sealer deflection. Instrument background is
generally expressed in counts per minute (epm), picocuries/gram (pCI/g),
milligrams/liter (mg/1), or other suitable units. Electronic background (background) is
determined by measuring the signal output for a particular instrument when subjected to
an area or matrix that contains no radioactive material other than natural or cosmic
radiation.
Determining Background
Accurately determining both types of background must be accomplished before applying
release criteria. Two industry accepted practices exist for determining natural background in
materials. The first method is to accurately measure the naturally-occurring radioactivity in
materials with the appropriate analytical instrument. This is accomplished by collecting a
clean sample of similar material from an uncontaminated source. An example of this type of
natural background determination is to measure the radioactivity in a piece of lumber from the
hardware store or a quart of motor oil from Wal-Mart®. The expected results for such an
analysis would be 1 to 2 pCi/g in wood for natural uranium and less than 0.1 pCi/I for mixed
fission products in oil. This same process can also be applied to chemical contaminates in
various matrices.
The other type of material background analysis is a statistical procedure called Chauveneus
Determination. This process requires making a large number of radiation measurements in a
defined area and then casting out the larger measured results. The average of the smaller
remaining results is considered to be background for the defined population.
Both of these techniques are applied for determination of natural background for the BCLDP.
Instrumentation
Release surveys will be performed using suitable instrumentation and industry standards. It
should be noted that the upper end of the release criteria defined in the applicable regulatory
standards and being applied to the BCLDP were developed in part based on the detection
limitations of the field instruments available at the time the standards were published. The
BCLDP will utilize field instruments, laboratory techniques, and survey techniques capable of
achieving detection limits at or below the upper bounds of the release criteria stated in Table
1. Current BCLDP instruments have detection limits lower than the surface contamination
guidelines for the most restrictive nuclides shown in Table 1. However, the BCLDP will not
continue to upgrade with state-of-the-art detection systems simply to drive the lower limit of
detection (LLD) continually lower. Surfaces with detectable radioactive contamination levels
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JAERI-Conf 95-015
greater than the LLD but less than the stated release criteria will be evaluated based upon
ALARA analyses for decontamination, disposal, or free release. Materials greater than the
release criteria will be decontaminated or disposed of as of radioactive waste.
Program Implementation
The surface radiological free-release program is implemented through highly qualified health
physics technicians. The health physics technicians perform extensive radiological surveys for
fixed and removable contamination.
The BCLDP presently has 49 health physics technicians. Of the 49 health physics
technicians, 38 meet the American National Standard Institute (ANSI) requirements for Senior
Health Physics Technician. Additionally, 13 of the senior technicians are certified by the
National Registry for Radiation Protection Technologists. Only senior technicians are
qualified and allowed to perform radiological free-release surveys.
Each free-release survey is documented and reviewed for accuracy prior to materials' being
removed from a radiologically controlled area. Once materials are certified for free-release,
they must be removed from the radiologically controlled area within thirty days or the survey
becomes invalid. Materials not removed from radiologically controlled areas within thirty
days must be resurveyed prior to their removal. Similarly, building areas are isolated once
free-release surveys have been completed, to prevent cross-contamination.
Once building surfaces are surveyed for free-release, the Independent Verification Contractor
(I¥C) is notified. The IVC checks the survey work performed to ensure that the free-release
criteria have been met. The IVC for the Battelle Columbus Laboratories Decommissioning
Project designated by the Department of Energy is the Oak Ridge Institute for Science and
Education (ORISE) in Oak Ridge, Tennessee.
ORISE provides personnel and equipment on-site to perform independent surveys, the
duration of the survey work performed by ORISE, of course, depends on the size of the
facility being evaluated. For smaller areas, ORISE may choose to evaluate only the final
survey report and analyze samples collected by the BCLDP at its analytical laboratory.
Upon completion of the survey performed by the IVC, a formal free-release statement will be
issued by ORISE, releasing the property for unrestricted use.
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Conclusion and Recommendations
By successful implementation of current surface radiological release criteria, the BCLDP has
released in excess of 55,000 cu.ft. of materials that would have otherwise been disposed of as
low-level radioactive waste. In addition to these materials, several areas from Battelle-owned
buildings have been released for unrestricted use. To date, areas from King Avenue
Buildings 5, 6, 7, and 9 and West Jefferson Buildings JS-1, 10, and 12 have been free-
released for unrestricted use.
Although the stated release criteria have been implemented for the good of the BCLDP, new
surface radiological release criteria should be developed based on risk analysis. The new
release criteria should incorporate the industry developments in radiological monitoring
equipment and provide consideration for facility use after free-release. The new criteria
might also incorporate standards for conditional release for decontaminated facilities that
would not be occupied by specific groups of the general public.
The release criteria must be developed based on acceptable risk and cost-effective
implementation. Most importantly, surface radiological release criteria must be capable of
being implemented in the field.
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2-2 COST-BENEFIT ANALYSIS FOR U.S. NRC PROPOSED
RADIOLOGICAL CRITERIA FOR DECOMMISSIONING
Robert A, Heck, Ph.D..Environmental Policy Section Leader
U.S. Nuclear Regulatory Commission
ABSTRACT The U.S. Nuclear Regulatory Commission prepared cost-benefit
analyses in support of the proposed regulation on radiological criteria for
decommissioning. These analyses have been published in the Draft Generic
Environmental Impact Statement (GEIS), NUREG-1496, and in the Draft
Regulatory Analysis (RA). The method used was to first list the reasonable
regulatory alternatives that could be considered. Second, for each
regulatory alternative, we analyzed and compared the costs and the
incremental radiological and non-radiological impacts to workers and members
of the public.
The regulatory alternatives for unrestricted use of a site that were
analyzed and compared were: no regulatory change; a uniform risk based on
total effective dose equivalent (TEDE); use of "best" available technology;
and returning the level of radioactivity attributable to licensed activity
to background levels. We also considered the combination of remediation and
restricting the use of the site following decommissioning.
The analyses were performed for ten types of reference facilities, and
each facility was evaluated at low, medium, and high levels of
contamination. The reference facilities included: reactors; various
uranium and non-fuel cycle facilities; and independent spent fuel storage
installations. The levels of contamination were estimated on the basis of
available data and included surface contamination on structures and soil,
activated concrete, and also volumetric soil contamination.
Since both the radiological and non-radiological benefits were
considered, the benefits of the various alternatives were measured in terms
of "Estimated Mortalities Averted." Exposures to workers during
decommissioning and the public after decommissioning were combined with
industrial and transportation accidents, weighted by the appropriate risk
factors. This estimate replaced the collective dose calculation as would
be calculated in more traditional analyses.
Costs were estimated on the basis of removal of surfaces from
reference structures, soil removal and costs of radiological surveys and
measurements. Transportation and disposal costs of removed concrete and
soil were also included.
Conclusions supported by the analyses were that a risk limit,
expressed as a 15 mrem/y dose, is reasonable both as a level for protecting
public health and safety and with regard to its cost-benefit effects.
Further reductions on a site-specific application of the ALARA principle are
also supported in the context of accounting for both the radiological and
non-radiological effects in both the short and long terms. Restricted
release is also supported when the same level of protection is provided in
decommissioning those facilities that cannot reasonably meet the
unrestricted release criteria.
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JAERI-Conf 95-015
INTRODUCTION The U.S. Nuclear Regulatory Commission (NRC) is required by
law, the National Environmental Policy Act, to consider the effects of
major, new regulations on the environment. The analysis for the proposed
regulation on radiological criteria for decommissioning has been published
as a two volume draft report, entitled, "Generic Environmental Impact
Statement in Support of Rulemaking on Radiological Criteria for
Decommissioning of NRC-Licensed Nuclear Facilities," (NUREG-1496, August
I994)(6E!S). The GEIS first listed the reasonable alternatives for
regulatory approaches. Then for each type of reference facility to be
decommissioned, the environmental effects were analyzed on human health and
safety, specifically, radiation exposure resulting from occupancy of site
buildings and residence on site lands after license termination, and
radiation exposure during decommissioning and waste transport. In addition,
non-radiological impacts on humans, biology, economics, society, and land
use were considered. The results of these analyses are presented,
REGULATORY ALTERNATIVES The five broad categories of regulatory
alternatives that were considered were: no regulatory change; a uniform risk
limit; use of "best" available technology; return the site to background
levels; and restricted use of the site.
"No regulatory change" as an alternative means that the NRC would continue
to decommission sites on the basis of existing practices and policies on a
case-by-case basis or make a regulation stating those existing practices and
policies. This alternative was rejected and not analyzed in detail because
the criteria are inconsistent and have not been reassessed in light of
changes in radiation protection standards since the existing practices and
policies were established.
The "risk limit" alternative was expressed in terms of Total Effective Dose
Equivalent (TEDE) radiation dose and included a provision for implementation
of the ALARA principle to achieve doses below that limit. The costs and
benefits for this alternative were analyzed in detail as described below for
residual dose levels of 100, 60, 30, 10, 3, 1, 0.3, 0.1, and 0.03 mrem
TEDE/year distinguishable from background.
The "best effort" alternative implies that a site would be released for
unrestricted use only if the residual radioactivity cannot be removed or
measured using the best available technology. This alternative is clearly
technology driven and could lead to unequal levels of protection from site
to site and from time to time. It is not clear which technology would be
considered best. In some cases, the application of the best technology from
a radiological point of view could cause more harm than good to the
environment.
"Return to background" would require the removal of all radioactivity
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attributable to licensed activities. While it is desirable to reduce
residual radioactivity to a level indistinguishable from background, it
would be extremely difficult as a practical matter to demonstrate that such
an objective had been met. Evaluation of this alternative included
extrapolating the results of the analysis for the risk limit alternative
discussed above,
The "restricted use" alternative means that specific radiation exposure
pathways would be prevented by institutional controls. For example, a
special, local law could prohibit farming on a decommissioned site.
AFFECTED ENVIRONMENTS For the generic analysis, reference facilities
considered to be sufficiently representative of those licensed by the NRC
were described. The reference facilities were: reactors—power, research,
and test; uranium fuel fabrication plant; UF6 production plant; uranium mill
facility; independent spent fuel storage facility; sealed source
manufacturer; research and development facility; and a rare metal extraction
facility. Concrete in structures and soils, as appropriate for each type
of reference facility, were assumed to be contaminated with one or more of
the following reference radionuclides: Co-60, Sr-90, Cs-137, natural
uranium, or thorium in equilibrium with its progeny. The environment for
each of these facility types was considered with respect to impacts on human
health and safety, biology, physical resources, and socioeconomic factors.
Only the human health and safety impacts were quantitatively analyzed—the
remainder were qualitatively considered.
BENEFITS The incremental benefits of the decommissioning alternatives were
estimated by quantitatively analyzing the potential mortalities that could
be averted by decontamination to various residual radiation dose levels.
Potential mortalities from both radiological and non-radiological causes
were combined in the analysis. The radiological causes of potential
mortalities that were analyzed were radiation exposures of: persons
performing decontamination and decommissioning work on site before the
license is terminated; persons living on the site after decommissioning;
persons working on the site after decommissioning; and persons (both workers
and the public) exposed to radiation during the transport of waste to
licensed disposal sites. Non-radiological causes of potential mortalities
were: industrial accidents during the decommissioning work; and
transportation accidents of decommissioning workers and of the public
associated with the transport of radioactive waste to a licensed disposal
facility. Even though the amount of time between the injury and death can
be assumed to be quite different for radiological and non-radiological
causes, combining the potential mortalities provided a common basis for
comparison between different types of facilities. Dose conversion factors
and scenarios used in the analyses were consistent with those from the
screening pathway modeling described in, "Residual Radioactive Contamination
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From Decommissioning," (NUREG/CR-5512).
COSTS Each type of facility was analyzed using the assumptions that the
soil at each facility was contaminated in the "high, "medium," or "low"
range as judged from limited data and expert opinion. For each of these
three cases and for each of the ten facility types, costs were estimated
for: radiological characterization surveys; cleaning, removal, packaging,
transportation, and disposal of concrete, other building materials, and
soil; and final radiation surveys. The costs included estimates of both the
direct labor to decontaminate and the administrative costs during removal
of materials and soil, and also costs of materials used in the
decommissioning process and the costs of transportation and disposal of
wastes. The costs for radiation measurements were estimated in a similar
manner and were consistent with the methods described in the "Manual for
Conducting Radiological Surveys in Support of License Termination"
(NUREG/CR-5849).
COMPARISON OF COSTS AND BENEFITS The results of the analyses were compared
to ,a baseline of 100 mrem TEDE/year to the average individual in the
critical group exposed to residual radioactivity at the decommissioned
facility. The baseline dose rate, 100 mrem TEDE/year, was chosen because
it is consistent with the 100 rarem/year dose limit for members of the public
in NRC regulations. Thus, as the least restrictive regulatory alternative
to consider, each site would have to decontaminate to 100 mrem/year or less.
The benefits of requiring more restrictive decontamination were expressed
in terms of estimated mortalities averted compared to 100 mrem/year. The
costs for reaching the residual radiation dose levels listed above were
evaluated for each reference facility at different contamination levels.
The results of the analyses varied according to the type of facility
examined. As examples, for power reactors the benefit versus cost curve was
almost a step-function. For a sealed source manufacturer the curve was
increasingly more sloping and at low annual residual radiation doses leveled
to almost no benefit gained for additional cost. For the independent spent
fuel storage installation, the transportation and industrial accidents
dominated the benefits when residual radiation doses were lower than 10
mrem/year. The optimum annual dose for all facilities was in the same
general range, namely, 15 mrem/year. Examples of the results are shown in
the Figures from Chapter 7 of the GEIS as extracted for this paper.
PRELIMINARY RECOMMENDATIONS As a result of these analyses preliminary
recommendations were made in the GEIS that used characteristics of several
of the alternatives. A dose limit of 15 mrem/year was recommended for the
following reasons. The 15 mrem/year is a fraction of the public dose limit
recommended by the International Commission on Radiation Protection and the
National Council on Radiation Protection (U.S.) which is 100 mrem/year above
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background from all non-medical sources. The potential for radiation
exposure to the public from multiple sources implies that the dose from a
single source should be a fraction of the overall limit. The incremental
risk of a radiation-caused mortality at 15 mrem/year is roughly 4X10"4 per
lifetime, and is in the same range established by the U.S. Environmental
Protection Agency for the "Superfund" cleanups. The overall optimal limit,
based on a cost of benefits analysis and considering all reference
facilities, is roughly 15 mrem/year.
The analyses of costs compared to benefits also indicate that some specific
facilities may be abli to decontaminate to levels lower than 15 mrem/year
with minimal additional costs. Thus, the preliminary recommendation
includes a requirement for ALARA levels below the 15 mrem/year limit. It
is recognized that the ideal objective is to return the radiological
condition of the site to a condition that is indistinguishable from
background, however, ALARA implementation is expected to yield a range of
doses from indistinguishable from background to 15 mrem/year.
Some sites may not be able to be decontaminated to a 15 mrem/year level
without causing more harm than good to the environment or public health or
incurring a prohibitively large cost. For these sites restricted release
is a reasonable option provided that: the site is decontaminated to a level
not expected to cause more than 100 mrem/year even if there were no
restrictions; with institutional controls and restrictions in place the
expected dose would be no more than 15 mrem/year; and that ALARA be
implemented to reduce the predicted dose.
Figure 1. Estimated Averted Mortality and Additional Cost
Attributed to Reducing Residual Dose Rate Below 100 mrem/y
AH Reference Facilities*
Medium Soil contamination
o
=8
3
t3
O
3.0
2.5
2.0
1.5
t
0.5
0.0
03
Residual Dose Sate in tnrem/yr
$0
$10 $20 $30 $40
Additional Cost ($M)
$50
'NUREG-L496
$60
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2-3 EPA's Technical Methodology for the Development of
Cleanup Regulations for Radioactively-Contaminated
Soils and Buildings
H. Ben Hull, Mark Doehnert,
and Anthony Wolbarst
U.S. Environmental Protection Agency
Office of Radiation and Indoor Air
Washington, B.C. 20460
and
John J, Mauro and Lowell Ralston
Sanford Cohen & Associates, Inc.
McLean, VA
The total number of sites contaminated with radionuclides in the United States
is in the thousands. These sites range in size from corners of laboratories to
sprawling nuclear weapons facilities covering many square miles of land.
The U.S. Environmental Protection Agency (EPA) is proposing regulations for
the protection of the public from radionuclide contamination at sites that are to be
cleaned up and released for public use. The rule will apply to sites under the control
of Federal agencies, and to sites licensed by the Nuclear Regulatory Commission
(NRC) or NRC Agreement States.
In support of this rulemaking, EPA is conducting a comprehensive technical
analysis which will provide information on the volumes of soil requiring remediation
at various possible dose and risk levels. The analysis will also provide information at
different cleanup levels on the number of potential adverse health effects among
people living or working on or near a site following the cleanup of its radioactive
contamination. This paper summarizes the overall approach for the technical
analysis.
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EPA's Technical Methodology for the Development of
Cleanup Regulations for Radioactively-Contaminated
Soils and Buildings
by
H, Ben Hull, Mark Doehnert,
and Anthony Wolbarst
U.S. Environmental Protection Agency
Office of Radiation and Indoor Air
Washington, DC
and
John J, Mauro and Lowell Ralston
Sanford Cohen & Associates, Inc.
McLean, VA
The U.S. Environmental Protection Agency (EPA) is proposing regulations
that set cleanup standards for remediating soil and structures at sites contaminated
with radioactive material. These proposed regulations will apply to Federally-owned
or operated sites and to sites licensed by the Nuclear Regulatory Commission
(NRC)/NRC Agreement States.
In support of this rulemaking effort, EPA is conducting a comprehensive
technical analysis of benefits and costs of alternative cleanup criteria. The present
study discussed in this paper is part of that analysis, and presents the methodology
used to determine how radiological health impacts and volumes of soil to be
remediated vary as functions of the possible cleanup level. The analyses evaluate a set
of typical sites that are considered representative of the sites to which this regulation
will apply. For these "reference sites", the analytical work addresses: 1) the radiation
doses and risks to an individual resulting from exposure, via all environmental
pathways, to unit concentrations of radionuclides in soil; 2) the radionuclide soil
concentration, in units of pCi/g, that would have to be achieved in order to meet
various possible individual dose or risk levels; 3) the quantity of soil that contains
radioactivity in excess of any given radionuclide soil concentration; 4) the number of
potential radiogenic cancers, and cancer deaths,averted, by remediating the soil to the
radionuclide soil concentration corresponding to the various individual risk levels; and
5) the number of radiogenic health effects that might eventually occur among
remediation workers and the general public because of the remediation process itself.'
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In order to quantify these health and cost impacts, EPA has developed and
used a technical methodology which can be summarized by the following steps.
(1) Evaluation of the magnitude of the cleanup problem. Identification and
estimation of the number of sites and approximate volumes of soil in
the United States contaminated with radioactive materials.
(2) Development of a set of models, scenarios, and assumptions that may
be used to perform risk assessments in support of the regulations.
(3) Development and analysis of a generic site as a means to test the
models and derive generic risk factors which relate a given level of
radionuclide contamination in soil to potential public health risks.
(4) Development of a set of reference sites that encompass the
characteristic of the sites that may fall within the scope of the soil
cleanup rule. A reference site is defined in terms of the radionuclide
concentration contamination pattern and the environmental,
hydrogeological, demographic, and land use characteristics of the site.
(5) Analysis of the reference sites to determine 1) the volumes of soil that
must be remediated to achieve various levels of individual risk and/or
dose; and 2) the number of potential radiogenic cancers averted, or
caused, as a result of site cleanup to alternative risk-based cleanup
goals.
In addition, EPA analyses supporting the rulemaking consider the issue of
implementation. The analyses include the derivation of generic soil cleanup levels that
correspond to each of the alternative risk-based cleanup goals, and evaluation of their
practicality in light of the lower limits of detection of field and laboratory analytical
techniques and the presence of variable natural and manmade background radiation.
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MAGNITUDE OF THE CLEANUP PROBLEM
Number of Sites
EPA has estimated that there are about 5000 sites known to be contaminated
with radioactive materials in the United States. Included are sites on EPA's National
Priorities List (NPL); sites under the authority of various Federal agencies,
predominantly DOE and DoD; and sites licensed by the NRC and NRC Agreement
States. In estimating the number of sites with radioactive contamination, EPA has
adopted the following definition:
A "site" is any installation, facility, or discrete, physically separate
parcel of land, or any building or structure, or any body of water or
surface water, that is known to be contaminated with radionuclides in
concentrations greater then those naturally occurring. When a portion
of such an entity is contaminated, the entire entity is considered a
"site". For example, the Hanford Reservation, which has many
contaminated buildings, discrete release sites, and groundwater
contamination, is considered a single "site".
Therefore with this definition, when only a portion of a "site" is known to be
radioactively contaminated, the entire unit or location is designated a contaminated
"site."-Note that while, large DOE facilities are counted as single sites, they may be
extensively more complex then sites in other categories.
Categories of Sites
The identified radioactively contaminated sites have been placed into three
major administrative categories: Department of Energy (DOE), Department of
Defense (DOD), and NRC/Agreement State licensees. Most of the sites in the DOE
fall under the DOE Environmental Restoration Program and are large, complex, and
multi-functional facilities. These major sites encompass most of the contaminated soil
that falls within the scope of the rule.
To facilitate the process of identifying and characterizing reference sites, a site
categorization scheme was developed by EPA and representatives of DOE, DOD, and
NRC. Eighteen functional categories were identified which cover the full range of
sites containing radioactive materials. One of the eighteen categories is entitled
"Entire Sites", and was created to account for large, unique, complex sites that cross
functional category lines. Examples of such sites include the Hanford Reservation,
Oak Ridge National Laboratory, the Savannah River Plant, and the Idaho National
Engineering Laboratory.
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Volume of Soil Contaminated with Radioactive Material
The total national volume of radioactively contaminated soil is not known with
any degree of certainty, and will not be known until cleanup criteria are defined and
the sites are remediated. Based on preliminary information, however it is estimated
that approximately 108 cubic meters of contaminated soil are located at Federal
facilities and NRC licensees and fall within the scope of this rale. These estimates
are based on information provided in DOE's Integrated Data Base (DOE 94) and
ongoing EPA studies (EPA 94).
SELECTION/DEVELOPMENT OF PATHWAYS, SCENARIOS, AND MODELS
Two sets of mathematical pathway models and exposure scenarios were
selected/developed to perform risk assessment in support of the rulemaking: those for
assessing risks to individuals assuming reasonable maximum exposure (RME)
conditions and those used to estimate the cumulative health impacts over time in the
exposed populations (EPA 93). EPA also developed models and scenarios to compute
risks to workers.
Exposures and Risks to the RME Individual
The development of cleanup regulations for soil contaminated with radioactive
materials must be based on potential radiation dose and/or risk to the public from all
significant exposure scenarios, and pathways.
The selection of exposure scenarios, and pathways for deriving risks involved
a review of EPA guidance and standardized methodologies applicable to the
performance of risk assessment. After the identification of the significant scenarios
and pathways, potential candidate multimedia models were identified; EPA then
developed and used criteria for the pathway model selection.
The methodology for evaluating radiation-induced cancer risks was selected to
be essentially consistent and compatible with that used by EPA for evaluating cancer
risks from non-radioactive hazardous chemicals. As such, the methodology generally
follows the basic steps in the Superfund Remedial Investigation/Feasibility Study
(RI/FS) process described in the EPA manual, Risk Assessment Guidance for
Superfund (RAGS) (EPA 89, EPA 91), for baseline risk assessments.
Exposure scenarios are combinations of exposure pathways and exposure
assumptions that risk assessors use to evaluate site risks under different land-use
classifications. Each scenario describes actual or potential contaminant releases,
migration pathways, contaminated media, exposure point concentrations, and receptor
characteristics for a specific land use and assumed set of site conditions. The purpose
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of these scenarios is to ensure that every reasonable exposure pathway and assumption
is considered and that all individual exposures and risks are assessed in a consistent
and comprehensive manner.
For the purpose of this rulemaking, EPA has evaluated three land-use
scenarios: a rural residential scenario, a commercial/industrial scenario, and a
suburban scenario. Pathways assessed by the rural residential scenario are:
* External radiation exposure from photon-emitting radionuclides in soil.
• Inhalation of resuspended soil and dust containing radionuclides.
• Inhalation of radon (Rn-222 and Rn-220) and radon decay products
from soil containing radium (Ra-226).
« Incidental ingestion of soil containing radionuclides.
• Ingestion of drinking water containing radionuclides transported from
soil to potable groundwater sources.
* Ingestion of home-grown produce ( fruits and vegetables) contaminated
with radionuclides taken up from soil.
* Ingestion of meat (beef) containing radionuclides taken up by cows
grazing on contaminated plants (fodder).
* Ingestion of milk containing radionuclides taken up by cows grazing on
contaminated plants (fodder).
* Ingestion of locally caught fish containing radionuclides.
The suburban scenario makes use of the first six of these, and the commercial
industrial employs the first five. Exposure assumptions differ significantly among the
three land use scenarios.
Twenty-seven multimedia radioactivity transport models were reviewed to
determine the degree to which each model fulfilled modeling needs. Five criteria
were identified and used for the evaluation and selection of pathway models for use in
dose and risk calculations. These criteria require: that pathway models be capable of
addressing multiple exposure pathways and risks from radionuclides in soil; are
currently in use or planned for use at radiation sites; are validated, peer-reviewed, or
generally accepted by radiation risk assessors (with preference given to EPA-approved
or accepted methods); are user-friendly with a manageable number of input
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parameters requiring minimal or modest amount of site data; and are computer
encoded or amenable to simple hand calculations.
Three pathway models were selected to perform the risk assessment
calculations; DOE's RESRAD Version 5.19 (DOE 93), EPA PRESTO-CPG (EPA-
87), and a code based upon an expanded version of EPA's RAGS/HHEM Part B
equations, EPA determined that all three of these models would be used in order to
evaluate the sensitivity of the risk assessments to alternative models. The analysis of
reference sites, described below, were carried out with RESRAD.
Cumulative Population Impacts
To evaluate potential radiological impacts of a site on public health, it is not
sufficient simply to derive the risks to the RME individual prior to and following
cleanup. It is also necessary to derive the cumulative impacts to the population in the
vicinity of the site. The process used to select an approach for modeling cumulative
population impacts was similar to that used for modeling the RME individual.
It was determined that time integration periods of 100, 1000, and 10,000 years
needed to be considered. Alternative pathways were explicitly addressed in
consideration of future land-use scenarios. At this time analyses have been completed
using a conservative bounding model similar to the RAGS/HHEM equations.
ANALYSIS OF A GENERIC SITE
A generic site was defined and evaluated to test the models and derive generic
risk factors and soil-cleanup criteria. EPA also used the generic site to perform
sensitivity and uncertainty analyses of the models and assumptions. The generic site,
or some variation of it, may be used to create tables of soil cleanup levels for actual
implementation at real sites.
As defined, the generic site consists of a contaminated zone, an unsaturated
zone, and an aquifer. The contaminated zone is defined as a 10,000 m2 area of soil
with radionuclide contamination extending from the surface to a depth of 2 m. Each
radionuclide considered is assumed to be uniformly distributed throughout this soil
volume at a concentration of 1 pCi/g. All impacts are directly proportional to the
radionuclide concentration in the soil. The unsaturated zone extends down 4 meters
from the bottom of the contaminated zone to the top of the aquifer, and is initially
uncontaminated. The aquifer represents a groundwater resource that is assumed to
supply 100 percent of the daily drinking water intake for onsite residents (i.e., 2
L/day) or half of the drinking water intake for workers (i.e., 1 L/day). Also, it is
assumed to be a source for water used to irrigate crops and feed livestock. A well is
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assumed to be constructed onsite at the downgradient edge of the site, with an intake
3 meters below the top of the aquifer. The aquifer is assumed to be uriconlaminated
initially.
The selection of the parameters describing the generic site was based on (1) a
review of the range of similar features observed at actual radiation sites, (2) the
assumption that the site had to be of sufficient size to support the exposure pathways
and scenarios considered, and (3) the results of sensitivity analysis calculations; for
sensitive parameters, realistic, but conservative values were selected.
Derivation of Generic Risk Factors
Risk factors define the relationship between a given concentration of a
radionuclide in soil and the risk to individuals who may reside or work at the site.
For each radionuclide, risk factors are expressed in terms of lifetime risk of cancer
per pCi/g of a radionuclide in soil. This is a convenient relationship because the
individual risk is directly proportional to the radionuclide concentration in the soil.
As a result, once a risk factor for a radionuclide is determined for a site, the
radionuclide concentration corresponding to a given risk-based cleanup goal for the
RME individual can be readily derived by dividing the cleanup goal by the risk
factor.
Generic risk factors were derived for 67 radionuclides for the three selected
scenarios using RESRAD, HHEM Part B, and PRESTO. Studies of the differences
in results obtained with the three models (which tend to be relatively small for most
radionuclides) sheds light on their respective strengths and weaknesses.
Sensitivity Analysis
Among the most sensitive site parameters that can affect the values of risk
factors are the contaminated zone area and thickness, infiltration rates, distribution
coefficients, and unsaturated zone thickness. The values for these parameters were
varied individually to determine their affect on the risk factors.
In addition to posing a risk to the RME individual, contaminated soil at a site
could pose a health risk to the population living on or in the vicinity of a site.
Sensitivity analyses were performed to determine the cumulative population impacts
for the generic site, by varying the infiltration rate, the thicknesses of the
contaminated zone, the distribution coefficient, the thicknesses of the unsaturated
zone, and the time integration periods. Similar calculations were undertaken for the
reference sites, as described below.
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JAERI-Conf 95-015
DEVELOPiMENT OF REFERENCE SITES
It is not feasible to assess fully all the sites that may fall within the scope of
the rule because there are thousands of them, and many have highly complex
contaminant, environmental, demographic, and ecological characteristics. Moreover,
detailed site characterization data simply do not exist for most sites. Site
characterization supporting the ralemaking is limited to a representative sample of
sites. In addition, these representative sites or reference sites are only partially based
on real sites and cannot be taken as complete and accurate characterizations of those
sites.
Data on several hundred sites representing a broad range of administrative and
functional categories were used to select a set of reference sites that are representative
of: the major administrative categories of sites (i.e., DOE facilities, NRC licensees,
DOD facilities); the major functional categories of sites (e.g., weapons production
and R&D facilities, fuel cycle facilities, materials licensees); the major facilities with
unique characteristics (e.g., Hanford, Savannah River, Oak Ridge, etc.); the range of
source characteristics (e.g., radionuclides, concentrations, depth and area of
contamination, chemical and physical form); and the range of environmental settings
(i.e., climatology, hydrogeology, demography).
For DOE nuclear facilities, EPA obtained data from DOE Public Reading
Rooms and libraries and from EPA Regional Offices. Document review began with
acquisition of Federal Facility Agreements for each DOE facility, which identify the
waste management units at each site and the status of the remedial investigation, and
also indicate what reports are available concerning a site. Where available, Records
of Decision (RODs) and Remedial Investigation/Feasibility Studies (RI/FS) were
obtained. For sites where RODs and RI/FS materials have not yet been completed,
an attempt was made to obtain Preliminary Assessment/Site Investigation (PA/SI)
reports; Environmental Audit Reports; Environmental Assessment Reports;
Environmental Monitoring Reports; Environmental Data Packages; and Effluent
Reports. Significant use was made of the DOE's Integrated Data Base (IDB). Data
on DOD sites were obtained from similar sources.
EPA obtained data characterizing NRC licensed facilities from site descriptions
provided by the NRC in the preliminary draft of the Generic Environmental Impact
Statement (GEIS) for the NRC rulemaking on decontamination and decommissioning,
and documentation available on the NEC's Sites Decommissioning Management
Program (SDMP).
Soil volume versus radionuclide concentration curves for the reference sites
were derived using these data and a combination of three different analytical methods.
The method, or combination of methods, depended on the type and completeness of
the data characterizing the soil contamination at the actual sites upon which the
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reference sites are based. For sites where detailed soil contamination characterization
data were available, such as sites with completed Remedial Investigation/Feasibility
Study (RI/FS) reports, the data were used directly to derive the curves. For sites
with aerial radiological survey data, these data were used to supplement the soil
sampling data. For sites lacking adequate soil sampling data or aerial surveys, curves
were developed by extrapolating from data from other sites which were judged to be
generally applicable to the sites with limited data.
ANALYSIS OF REFERENCE SITES
The reference sites were analyzed to estimate (1) the volumes of soil that may
require remediation, (2) the number of potential cancers averted as a result of
achieving various site cleanup levels, and (3) the number of potential cancers caused
among workers and the public during remediation, to achieve levels of individual risk
to post-cleanup users of the sites.
Soil Cleanup Volumes for Reference Sites
EPA has calculated, for each reference site, the volume of soil that may need
to be remediated at each site to ensure that no individuals will receive radiation
exposures which could result in a lifetime cancer risk exceeding the alternative risk-
based cleanup goals ranging from lxlO~6 to lxlO~2. A three-step process was used to
estimate the volumes of soil at each reference site requiring remediation as a function
of the cleanup levels:
Step 1 - Construct curves which relate the volume of soil as a function of
contaminant concentration,
Step 2 - Determine the relationship between a given concentration of a
radionuclide in soil and risks to individuals. This relationship was established
for each radionuclide and each reference site and is defined as a site-specific
risk factor, which is expressed in units of lifetime risk of cancer per pCi/g,
Step 3 - Using the site-specific risk factors and the site-specific soil volume
versus contaminant concentration curves, determine the soil cleanup volume as
a function of the risk-based cleanup goals.
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Radiological Impacts Due to Soil Cleanup
One of the benefits associated with site cleanup is the reduction in the
cumulative exposure and associated health risks to the population residing on, or in
the vicinity of, the contaminated property following cleanup. For each reference site
it is assumed that in the future the site could be heavily populated, used extensively
for farming, and that the groundwater is used extensively for domestic purposes.
The specific population exposure pathways addressed include:
Direct radiation from living on contaminated soil,
Inhalation of suspended dust,
Exposure to indoor radon progeny,
Ingestion of crops raised on contaminated soil, and
Ingestion of contaminated groundwater
Cumulative population exposures and the adverse health effects attributable to
these exposures were derived for each pathway and for time integration periods of
100, 1000, and 10,000 years. These alternative pathways and time periods were
addressed explicitly for the consideration of future land-use scenarios and time periods
of interest for the rulemaking.
Four future demographic patterns were also considered in deriving the long
term impacts associated with radioactively contaminated soil: rural, intermediate,
suburban, and "most likely". In the rural demographic setting, one assumes that the
population density is 10 persons/km2, with and without farming. In the intermediate
demographic setting the population density is assumed to be 100 persons/km2, with
and without farming. In the suburban setting, the population density is assumed to be
1,000 persons/km2, without farming. Finally, the "most likely" scenario for each
reference site adopted the population density of the corresponding actual site upon
which it was based, and rounded off to the nearest hundred/km2.
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REFERENCES
EPA 87 Environmental Protection Agency, 1987. "Low Level and NARM
Radioactive Wastes, Model Documentation PRESTO-EPA-POP:
Volume 1 - Methodology Manual," EPA/520/1-87-024-1.
EPA 89 Environmental Protection Agency, 1989, "Risk Assessment Guidance
for Superfund Volume 1, Human Health Evaluation Manual, Part A, "
EPA/540/1-89-006-1.
EPA 91 Environmental Protection Agency, 1991. "Risk Assessment Guidance
for Superfund Volume 1, Human Health Evaluation Manual. Part B,
Development of Risk Based Preliminary Remediation Goals," PB92-
963333.
EPA 93 Environmental Protection Agency, 1993. "Issues Paper on Radiation
Site Cleanup Regulations," EPA 402-R-93-084.
EPA 94 Environmental Protection Agency, 1994. "Technical Support Document
for the Development of Radionuclide Cleanup Levels for Soil - Draft
Review".
DOE 93 Department of Energy, 1993. "Manual for Implementing
Residual Radioactive Material Guidelines Using
RESRAD," DOE/OR/21949-337.
DOE 94 Department of Energy, 1994. "Integrated Database for
1994: U.S. Spent Fuel and Radioactive Waste
Inventories, Projections, and Characteristics," DOE/RW-
006-Rev,9,
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2-4 RADIOLOGICAL SURVEYS: METHODS, CRITERIA, AND THEIR
IMPLEMENTATION*
G, Subbaraman, R. J. Turtle, and B. M. Oliver
Rockwell International, Roeketdyne Division
6633 Canoga Avenue, Canoga Park, CA 91303
ABSTRACT
A radiological survey provides an answer to the following question: Can a decontaminated or
remediated site or structure be released for use without radiological restrictions? The answer is
derived from considerations involving a host of site— and radionuclide—specific variables, pathway
analyses, and future use scenarios, of which the nuclide—specific data are obtained during the survey.
Deriving the answer also requires reducing the sample data to representative statistical parameters
for the entire site or structure, and, in turn, determining whether the statistical parameters compare
favorably with the corresponding regulatory acceptance criteria. Based on recent experience, this
paper provides some insights into performing radiological surveys, with examples to illustrate this
approach.
INTRODUCTION
Release of radioactively contaminated sites and structures following their remediation or
decontamination for unrestricted use requires a radiological survey to demonstrate compliance with
regulatory acceptance criteria, such as those specified in Regulatory Guide 1.86 of the U. S. Nuclear
Regulatory Commission (NRC) (1). In performing radiological surveys, detailed plans and analytical
and computational tools are used to guide the surveyor, the analyst, and the site owner toward
satisfactory regulatory compliance. Of these, the survey plan, based on the past operating history and
the decontamination efforts, qualitatively specifies the residual nuclides that may be present at the
site, the affected media, and the detailed scope of the survey effort. Survey data obtained from a
number of locations at the site are then statistically analyzed for application to the entire site, and may
be used as inputs to pathway analysis models (2,3) to determine potential exposure to current or future
occupants of the site. Results from the statistical analysis and the pathway analysis models for the site
can, in turn, be compared with the numerical regulatory limits to determine compliance or the need
for additional, perhaps localized, decontamination.
Although the approach stated above is simple in principle, its implementation in the performance of
radiological surveys requires careful consideration of a number of variables, which generally fall into
the following categories:
1. Acceptance limits or criteria
2. Physical survey parameters
3. Statistical methods and parameters to which the data are reduced
Paper originally published in the proceedings of "Waste Management '91" Conference,
Tucson, Arizona, USA, vol. 2, p. 121 (June 1991)
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4. Selection of natural "background" radiation data
5. Pathway analysis models and applicable inputs, including future use—scenarios
for the site or structure.
Based on experience gained from radiological surveys performed over recent years, this paper
presents some practical insights into performing such surveys and accomplishing the overall objective,
which is to release a site for use without radiological restrictions. Although the details presented here
apply to one geographical area, we found that choices among the above variables could not be made
uniformly, and the reasons for this are discussed.
BACKGROUND
Formerly used and adjoining areas of a nuclear test facility in Southern California were radiologically
surveyed for residual radioactivity. Both structures (buildings) and open sites within the 117—hectare
(290—acre) facility were surveyed. Suspected or potential contaminants included activation products
(e.g., Co—60), fission products (e.g., Cs—137), fuel isotopes (e.g., enriched uranium), and calibration
sources (e.g., Ra-226), Based on previous operating history and routine monitoring data, the
contamination was known to be minor and restricted to soil (surface and subsurface) and building
interiors. For purposes of discussion, this paper presents four cases of residual soil contamination, as
follows:
Case A. A storage yard (Cs-137)
Case B. A side yard adjacent to a building with a previous decontamination history
(also Cs-137)
Case C. A building which formerly housed a below—grade Ra-226 source with
breached outer encapsulation
Case D. A building drainage system with potential Cs-137 and enriched uranium
contamination
In all of these cases, through remediation efforts and surveys, residual radioactivity has been
determined to be well below acceptance limits for release without radiological restrictions.
SURVEY PLAN
The surveys were performed in several steps: First, a broad survey plan was established for the entire
test facility complex. Based on the operating history of the complex, the plan identified suspected
radionuclides and the media to be characterized during the field survey, and established the related
acceptance criteria. The plan divided the complex into 25 convenient areas and buildings. The plan
also specified the statistical design, techniques, and parameters (e.g., number and size of grid
locations for measurements, calculation of the test statistic—described below—•etc.) to be used to
reduce the data, and procedures for the calibration and use of survey instruments. Finally,, the plan
required performance of an interpretative analysis of the data and determination of compliance or
other recommended actions, all of which were documented in a survey report for each subdivision..
As directed by the plan, gamma exposure rate data were collected at random locations within each
subdivision and at background areas where no nuclear operations took place. If the field
measurements showed exposure rates above certain pre-established action levels, the surveyor was
instructed to collect additional gamma exposure rate data and soil samples for radiometric analyses.
As discussed below, the data, after corrections for background, were statistically analyzed and
compared with acceptance limits for compliance. Results from this first round of surveys typically
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eliminated most areas from further consideration. For those few remaining areas, the survey report
recommended specific actions, including decontamination efforts in localized areas, and additional
investigations by means of a second round of surveys of the affected area. The Cases A through D
mentioned earlier were all subjected to this second round of investigation.
For all cases, an evolutionary approach was needed to modify the acceptance criteria, treat the data,
and to demonstrate compliance on a case—by—case basis. Use was made of the U. S. Department of
Energy (DOE) computer code RES RAD to implement site-specific guidelines for residual
radioactivity (2). This approach, as it applied to establishing the acceptance criteria, the statistical
treatment of the data, and the use of the RESRAD code, is discussed in the following sections, with
illustrative examples from the four cases.
ACCEPTANCE CRITERIA
General. Federal agencies (NRC and DOE) and state regulatory authorities (e.g., State of California
Radiological Health Branch) specify the criteria for acceptance of remediated sites and structures for
their release and use without radiological restrictions ("unrestricted release" in NRC terminology).
Typically, the criteria are provided in terms of maximum limits for external exposure rate (gamma),
emission— (alpha or beta-gamma) or nuclide—specific surface contamination (removable and fixed)
levels, and nuclide—specific activity concentration in the media. The recently issued pathway analyses
documents by the NRC and the DOE also enable determination of an acceptance criterion for a
specific site on the basis of the combined presence of several nuclides and on the basis of a credible use
scenario for the site or structure (2,3). A generic acceptance limit is available for Ra—226 in soil (4).
Acceptance criteria for a given survey would, therefore, have to be chosen from among these
regulatory stipulations. Where the numerical value of the limit varies (from one agency to another or
from one time to another within the same agency), a conservative choice must be made, as discussed
below.
Gamma Exposure Rate. Although the DOE guidelines (4) recommend a value of 20 u.R/h (at 1 m)
above background for gamma exposure rates, a lower value of 5 (iR/h above background was chosen
for these surveys and was based on a previous NRC stipulation for the unrestricted release of a
dismantled test reactor facility in the complex. Also, the 5 uR/h above background corresponds to the
recently issued NRC limit of 10 mrem/yr (2000-h occupancy) under its "Below Regulatory Concern"
policy (5).
Although it is conceivable that the limit for above—background gamma exposure rate could be set
even lower than 5 u.R/h (e.g., 10 mrem/yr applied to year—round occupancy), practical difficulties are
encountered. In the case of the survey data discussed here, for example, a 3 to 4 \iR/h variability was
observed in the natural background in "clean" areas, which is close to the 5 \iRfh limit. To overcome
this difficulty, carefully selected "cohort" areas were used, which were adjacent to a subdivision being
surveyed and had gamma exposure rates with relatively less variability. The cohort areas were verified
by means of soil analyses to have only natural background radioactivity. Thus, the 5 jtR/h
above—background gamma exposure rate criterion was applied in all cases, with the background being
established on a case—by—case basis.
Surface Contamination Levels. For both alpha— and beta—gamma—emitting nuclides, the 5,000,
15,000, and 1000 dpm/100 cm2 average, maximum, and removable contamination levels, respectively,
were used as specified in the NRC Regulatory Guide 1.86 (1). These criteria were applied in cases
where interiors of buildings were surveyed.
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Soil Activity Concentrations. For open sites with residual Cs—137 contamination (Cases A and B), a
site—specific concentration limit was established using RESRAD. It was surmised that this
contamination had resulted from release of naked fission products and, hence, an equal activity
concentration for Sr-90 was also assumed in performing the RESRAD calculations. For the case of a
future residential use of the site, the RESRAD—derived soil concentration limit was calculated to be
60 pCi/g each of Cs -137 and Sr—90. Details of these RESRAD calculations are provided later in this
paper.
For the case of Ra - 226 (Case C), two generic limits were considered. The first, with a value of 5 pCi/g
above background, corresponds to the activity concentration over the first 0.15 m of soil and the
second, 15 pCi/g above background, is for soil at depths greater than 0.15 m (4).
For the case of enriched uranium that could have potentially migrated from a drain line into adjoining
soil (Case D), the ratio of activity concentrations of U—235 to U—238 was compared with the same
ratio for naturally occurring concentrations of the two isotopes. RESRAD—type activity
concentration limits for these isotopes (or the initially suspected Cs—137 nuclide) were not needed
because the findings showed only natural activity in soils adjacent to the drain line.
In summary, a 5 [iR/h above—background gamma exposure rate was used as a generic acceptance
criterion, the background value being established on a case—by—case basis. Acceptable surface
contamination levels for building interiors were the same as specified in the NRC Regulatory Guide
1.86. Soil activity concentration limits were established on a case—by-case basis and included a
RESRAD—derived value of 60 pCi/g each for the combined presence of residual Cs—137 and Sr—90
in soils.
STATISTICAL TREATMENT OF DATA
General. A statistical procedure is required to validate the applicability of data collected at random
locations to an entire area or region. Once a value for such a representative statistical parameter is
calculated for the data distribution, this value can then be compared with the acceptance criterion to
determine regulatory compliance. A representative statistical parameter will be required for a
corresponding acceptance criterion; that is, one each for the gamma exposure rate, the contamination
levels, and the soil activity concentration criteria. All criteria must be met together for compliance. To
our knowledge, generic regulatory guidance or standard practices (e.g., from the ASTM) for
statistically treating radiological data are not available. The techniques adopted from various other
sources for the surveys are summarized in the following paragraphs.
Sampling Inspection. When it is impossible, impractical, or uneconomical to measure the
characteristics of every item in a group (e.g., each grain of soil in a plot or square meter of a wall), it is
common to use a statistical technique called sampling inspection. This approach allows the
development of conclusions and decisions on the basis of statistically representative data. The method
has been widely applied in industry and military where destructive tests must be performed or where
the lot size is unpractically large.
Sampling inspection may be based on measurement of attributes (whether an item sampled is a reject
or not) or variables (the actual value of the characteristic being measured). The latter approach (6)
was most suitable for the present survey because it provides increased accuracy for the same number of
inspections and because it permits estimating the probability that the entire group from which the
samples are taken has items that exceed specified values.
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In sampling inspections by variables, the number of data points on which measurements are obtained
is first chosen to be reasonably large (greater than about 30) so that the distribution of the data should
be normal (i.e., Gaussian). The mean of the distribution, xm, and its standard deviation, s, are then
related to a test statistic, TS, as follows;
TS = xm + ks.
TS and xm are then compared with an acceptance limit (such as those described earlier) to determine
acceptance or other plans of actions, including rejection of the surveyed area. In the above expression,
k is known as the tolerance factor. The value of k is determined from the sample size and two other
statistical sampling coefficients that are related to a consumer's risk of accepting a lot, given that a
fraction of the lot has rejectable items in it. The values chosen for these coefficients corresponded to
ensuring with 90% confidence that 90% of the area has residual contamination below 100% of the
applicable limit (a 90/90/100 test). The choice of the values for the two coefficients is consistent with
industrial sampling practices and State of California guidelines (7).
Implementation. Data from the surveys were treated using this statistical approach. The reduced data
were plotted against the cumulative Gaussian probability function on a probability—grade scale.
Display of data in this manner permits clear identification of data with values significantly greater than
expected for the lot, based on the Gaussian distribution. Fig. 1 shows illustrative data obtained for
ambient gamma exposure rates obtained at one site (Case B—first—round survey). Data obtained
from a second—round survey of the same area following removal of soil from the affected locations are
shown in Fig. 2. Here, the data have been corrected for background. Fig. 2 also shows the TS value and
the corresponding 5 uR/h acceptance limit. As can be seen, TS is less than the acceptance limit, thus
satisfying the gamma exposure rate criterion for this lot. Similar calculations and comparisons were
applied to other data, such as soil activity concentration and surface contamination level distributions.
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tion)
For the cases cited, the above approach was used to treat the gamma exposure rate data in all the
first-round surveys. The approach was also used for Cases A and B during the second-round surveys.
Cases C and D did not require statistical treatment during the second round of surveys because they
required simple removal (for authorized disposal) of contaminated items from highly localized areas
without any effect on the previously obtained gamma exposure rate data from adjacent areas. For
similar reasons, soil activity concentration data from Cases C and D were also not treated using this
approach; instead, the individual datum was compared with a generic acceptance criterion (Ra-226
for Case C and the ratio of U-235 to U-238 with respect to their natural ratio for Case D).
RESRAD CALCULATIONS
For Cases A and B, which involved relatively large affected areas, site-specific soil activity
concentration acceptance limits were established using the RESRAD code. The technical approach to
using RESRAD for determining the soil activity concentration acceptance limit for Cases A and B and
for demonstrating compliance with respect to the DOE 100 mrem/yr "basic dose limit" are described
below.
Overview. RESRAD calculates the effective dose equivalent to an occupant (current or future) by
performing environmental and dietary pathway analyses resulting from the presence and transport of
radioactivity through terrestrial media (both living and inanimate). Reference 2 provides a
comprehensive description of the pathway analysis model and a users' manual for RESRAD. Similar
pathway analyses models are available from other sources (3,8).
RESRAD provides results both in terms of a calculated activity concentration limit (in pCi/g)
corresponding to the 100 mrem/yr effective dose for identified contaminant nuclides at a site and in
terms of the effective dose equivalent for specified concentrations of nuclides. Thus, only qualitative
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information on the contaminant (e.g., an arbitrarily chosen concentration for Cs—137) is necessary to
derive the acceptance limit. With specific data from a survey as input (e.g., the average concentration
of Cs—137 and Sr-90 shown in the distribution in Fig, 2), RESRAJD will provide the corresponding
dose to an occupant. In both instances, however, identical site—specific data and use—scenarios must
be employed to obtain comparable results.
The following categories of input data are required to implement RESRAD for a given site:
1. Soil activity concentration data
2. Site—specific geohydrological parameters
3. Dietary parameters
4. Scenario-specific parameters
In all, about 80 input parameters are required. The RESRAD manual provides ranges of input values
for the geohydrological and dietary parameters for the United States, from which the code employs a
set of default input values. The code further allows modifying or eliminating exposure pathways, as
necessary, for a given use scenario. For obtaining realistic dose estimates, the manual recommends
use of site—specific geohydrological parameter values whenever possible. Similarly, while the
RESRAD default scenario corresponds to a family farm occupant at the site, the parameters affecting
the scenario can be modified for considering other scenarios.
Implementation. For the sites surveyed, three credible scenarios (industrial, residential, and
wilderness) were considered (the family farm default scenario was determined not to be credible for
this suburban the area). The default occupancy and dietary parameters were modified for each
scenario. Site—specific geohydrological data were collected and used as much as possible. Where the
default RESRAD value had to be used, sensitivity calculations were performed to confirm that
variation of the default parameters did not significantly influence the results.
The dimensions of the contaminated zone do significantly influence results from RESRAD. In our
surveys, the area of the contaminated zone was measured, but the depth of the zone had to be
estimated. To be conservative, however, infinitely contaminated zone dimensions were used as inputs
(about 100,000 m2 area and 1 m depth) to establish acceptance limits. Actual dimensions, with
best—estimate values for depths, were used only to determine how the RESRAD—calculated dose to
an occupant compared with the basic dose limit.
With the above input data, the acceptance limits were first established for individual nuclides for each
of the three scenarios. For example, Cs—137 activity concentration values for the site were 239,71, and
3,830 pCi/g for the three credible (industrial, residential, and wilderness) scenarios, respectively. If TS
for the measured data is less than the lowest of the three values, then the site would be acceptably clean
for all credible scenarios. This lowest bounding value, 71 pCi/g, corresponds to the residential
scenario, which, therefore, corresponds to the credible—bounding scenario. Recalling that an equal
activity concentration of Sr—90 was assumed, a simple calculation showed the acceptance limit for the
combined presence of both nuclides to be 60 pCi/g each for this credible—bounding scenario. Fig. 3
shows a cumulative probability plot for the measured Cs-137 data (Case B) which compares the TS
for this data with the acceptance limit.
Using the average of the measured activity concentration for this data from Case B (4.9 pCi/g of
Cs-137) and assuming an equal activity concentration for Sr-90, RESRAD calculations were
performed to determine dose to a potential residential occupant of the site. Background activity
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o
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Single Nuciide Umit = 71 pCi/g
I I
Two Nuclide Limit = 60 pCi/g, each of Cs-137 and Sr-90
. Test Statistic p+ks) = 11.7 pCi/g
Data
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Calculated from Data
0.1
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Cumulative Probability (%}
90 93
99
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Fig. 3. Measured Soil Cs-137 Activity Data Compared with Acceptance Limit
(Case B—Second—Round Survey After Decontamination)
concentrations for these man-made nuclides were assumed to be zero. Also used in these calculations
were the measured area of the site and the estimated depth of the contaminated zone, which was
chosen to be equal to the depth to which surface soil was removed. Use of this estimate for the depth is
conservative because concentration profiles for these isotopes are likely to be decreasing with depth
(e.g., 9) (versus the constant value used here) and because shielding provided by near—surface soil
layers would effectively eliminate any further increase in the external dose rate (which was found to be
the major contributor to dose in these cases). The resulting RESRAD—calculated dose to an occupant
under the credible—bounding scenario was 5.2 mrem/yr during the first year, far less than the DOE
guidance value of 100 mrem/yr basic dose limit and about half of the 10 mrem/yr NRC limit. Similar
low annual doses were found in Case A as well.
SUMMARY AND CONCLUSIONS
Applicable generic criteria were used and site— and nuclide—specific acceptance criteria were
developed for determining acceptance of decontaminated sites and structures. Although all cases
considered in the survey were within a similar geographic location, a case-by—case determination of
the criteria was necessary.
The technique of sampling inspection by variables was applied to reduce the survey data and to
calculate the test statistic. Although this treatment was used in all first—round surveys and applied to
gamma exposure rate measurements, it was necessary only in two of the four second—round surveys
(both gamma exposure rate and soil activity concentration data).
The RESRAD computer code was used for two of the four cases to conservatively determine
site-specific soil concentration acceptance limits for Cs-137 and Sr-90. The generic limit for
Ra—226 was used for the case of contamination of a building with this nuclide. Comparison of
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measured U-235—to—U-238 ratio with their natural ratio was performed in an area where enriched
uranium contamination was suspect.
Based on the survey data (both first—round and second—round) and based on comparisons with the
established acceptance limits, the sites or structures surveyed in all four cases were determined to be
acceptably clean for release without radiological restrictions.
RECOMMENDATIONS
Although it is the survey team's and site owner's responsibility to conservatively establish acceptance
limits from values within the existing regulatory framework, it is recommended that a single document
listing all the relevant and available criteria be created by a standards—setting organization, such as
the ASTM or the American National Standards Institute (ANSI). This should include acceptable
limits or methods for distributed radioactivity, such as contaminated soil. A standard practice for
statistical techniques and parameters should also be developed for use in radiological surveys by
groups such as the ASTM or ANSI.
ACKNOWLEDGEMENTS
We thank J. A. Chapman (currently at the University of Tennessee) for coordinating, performing, and
documenting the numerous first—round surveys that formed the bases for the cases illustrated in this
paper.
REFERENCES
1. "Termination of Operating Licenses for Nuclear Reactors," U.S. Nuclear Regulatory
Commission Regulatory Guide 1.86 (June 1974).
2. T. L. Gilbert, et al., "A Manual for Implementing Residual Radioactive Material Guidelines,"
Argonne National Laboratory Report ANL/ES-160 (also DOE/CH/8901) (June 1989).
3. W. E. Kennedy, Jr., and R. A. Peloquin, "Residual Radioactive Contamination from
Decommissioning — A Technical Basis for Translating Contamination Levels to Annual Dose,"
NUREG/CR-5512 (also PNL-7212) (Draft, January 1990).
4. "Guidelines for Residual Radioactivity at FUSRAP and Remote SFMP Sites," U.S
Department of Energy (February 1985; Rev. 2 dated March 1987).
5. "Below Regulatory Concern; Policy Statement," U.S. Nuclear Regulatory Commission (July
1990), and "Below Regulatory Concern — A Guide to the Nuclear Regulatory Commission's
Policy on the Exemption of Very Low—Level Radioactive Materials, Wastes and Practices,"
NUREG/BR-0157 (July 1990).
6. "Sampling Procedures and Tables for Inspection by Variables for Percent Defective,"
MIL-STD-414 (June 1957).
7. "State of California Guidelines for Decontaminating Facilities and Equipment Prior to
Release for Unrestricted Use," DECON-1 (June 1977).
8. "Generic Models and Parameters for Assessing the Environmental Transfer of Radionuclides
from Routine Releases. Exposure of Critical Groups," International Atomic Energy Agency,
IAEA Safety Series No. 57 (1982).
9. E. C. Amaral, et al., "Distribution of Cs-137 in Soils Due to the Goiania Accident and
Decisions for Remedial Action During the Recovery Phase " Health Physics, Vol. 60 f 11
(January 1991).
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2-5 Development of Risk-Based Computer Models for Deriving
Criteria on Residual Radioactivity and Recycling1
S.Y. Chen
Environmental Assessment Division
Argonne National Laboratory, Argonne, Illinois, U.S.A.
ABSTRACT
Argonne National Laboratory (ANL) is developing multimedia environmental pathway and
health risk computer models to assess radiological risks to human health and to derive cleanup
guidelines for environmental restoration, decommissioning, and recycling activities. These
models are based on the existing RESRAD code, although each has a separate design and
serves different objectives. Two such codes are RESRAD-BUILD and RESRAD-
PROBABILISTIC. The RESRAD code was originally developed to implement the
U.S. Department of Energy's (DOE's) residual radioactive materials guidelines for
contaminated soils. RESRAD has been successfully used by DOE and its contractors to assess
health risks and develop cleanup criteria for several sites selected for cleanup or restoration
programs.
RESRAD-BUILD analyzes human health risks from radioactive releases during
decommissioning or rehabilitation of contaminated buildings. Risks to workers are assessed
for dismantling activities; risks to the public are assessed for occupancy. RESRAD-BUILD
is based on a room compartmental model analyzing the effects on room air quality of
contaminant emission and resuspension (as well as radon emanation), the external radiation
pathway, and other exposure pathways. RESRAD-PROBABILISTIC, currently under
development, is intended to perform uncertainty analysis for RESRAD by using the Monte
Carlo approach based on the Latin-Hypercube sampling scheme. The codes being developed
at ANL are tailored to meet a specific objective of human health risk assessment and require
specific parameter definition and data gathering. The combined capabilities of these codes
satisfy various risk assessment requirements in environmental restoration and remediation
activities.
INTRODUCTION
Decades of nuclear fuel cycle operations associated with both civilian and military
applications in the United States have left many installations contaminated with varying levels
of residual radioactivity. Cleaning up these installations and returning the sites to public use
have been objectives of the U.S. Department Of Energy (DOE) and regulators. Criteria for
establishing acceptable cleanup levels have been aimed at protecting human health; such
criteria are therefore "risk-based." For instance, the U.S. Environmental Protection Agency
(EPA) has specified an acceptable human health lifetime risk level of 10** to IQ"6 in the
Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA)
1 Work supported by the U,S. Department of Energy, Assistant Secretary for Environmental
Management, under Contract W-31-109-Eng-38.
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cleanup requirements. Likewise, DOE and the Nuclear Regulatory Commission (NRC) have
established dose limits for developing soil guidelines (DOE 1990) and proposed
decommissioning of nuclear facilities (NRC 1994), In the process of demonstrating
compliance with the prescribed risk (or dose) levels, sophisticated tools are needed to
accurately translate the human health risks into cleanup levels (such as radioactivity
concentrations) of the environmental media. To accomplish this objective, Argonne National
Laboratory (ANL) has developed several computer codes designed to assess human health
risks associated with residual radioactivity and recycling. These codes are based on risk-based
pathway models and are intended for use in assessing radiological health risks to humans and
in deriving cleanup guidelines for environmental restoration, decommissioning, and recycling
activities. Three codes have been developed for these purposes: RESRAD, RESRAD-
BUILD, and RESRAD-PROBABILISTIC.
RESRAD, designed to calculate site-specific residual radioactive material guidelines to an
on-site resident, was developed to implement the DOE's residual radioactive materials
guidelines for contaminated soils (DOE 1990). RESRAD-BUILD analyzes human health
risks, from contaminants during building dismantling and occupancy through decommissioning
or rehabilitation, as well as for material end-use scenarios (such as for reuse and recycle).
RESRAD-PROBABILISTIC is a separate model of RESRAD currently under development.
While RESRAD provides deterministic results of the analysis, RESRAD-PROBABILISTIC
is intended to perform uncertainty analysis for RESRAD by using the Monte Carlo approach
(Yu 1993). It is based on the Latin-Hypercube sampling scheme (Iman and Shortencarier
1984). This PROBABILISTIC code complements RESRAD in that it offers the capability
of generating PROBABILISTIC distribution of resulting risks from the RESRAD pathway
analysis.
APPROACH
The basic framework of the RESRAD code series has four major parts: (1) source analysis,
(2) environmental transport analysis, (3) dose/exposure analysis, and (4) scenario analysis.
Source analysis addresses the source terms that determine the rate at which residual
radioactivity is released into the environment. That rate is determined by the geometry of the
contaminated region, the concentrations of radionuclides present, the ingrowth and decay rates
of the radionuclides, and the removal rate by erosion, leaching, or resuspension.
Environmental transport analysis addresses the areas of (1) identifying environmental
pathways by which radionuclides can migrate from the source to a human exposure location
and (2) determining the migration rate along these pathways. Dose/exposure analysis
addresses the derivation of dose conversion factors for the radiation dose that will be incurred
by exposure to ionizing radiation. The parameters that control the rate of radionuclide release
into the environment and the severity and duration of human exposure at a given location are
determined by patterns of human activity referred to as exposure scenarios.
The RESRAD-PROBABILISTIC code incorporates additional uncertainty analysis capability
to the RESRAD code. It utilizes Monte Carlo simulation techniques to obtain a statistical
distribution of the results.
102 —
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JAERI-Conf 95-015
RESRAD: Nine environmental pathways are modeled by RESRAD. For each exposure
pathway, radionuclides can migrate from a source to a human exposure location by many
environmental pathways. Major pathways used to derive site-specific soil guidelines in the
RESRAD code are identified in Figure 1.
External gamma radiation from radionuclides distributed throughout the contaminated zone
is the dominant external radiation pathway and the only external radiation pathway taken into
account in calculating soil guidelines. The dose due to the external gamma radiation is first
calculated for an individual exposed continuously to radiation from an infinite contaminated
zone at a distance of 1 m from the ground surface. Correction factors are then applied for the
finite area and thickness of the contaminated zone, shielding by a cover of uncontaminated
soil, irregular shape, shielding by the floors and walls of a house, and less-than-continuous
occupancy.
Inhalation exposure results primarily from inhalation of radon decay products and
contaminated dusts. An inhalation pathway consists of two segments: (1) an exposure
segment linking the source (contaminated zone) with the airborne radionuclides at an exposure
location and (2) an inhalation segment linking the airborne radionuclides with the exposed
individual. Modeling the airborne exposure pathway segment consists of two steps:
(1) modeling the process by which radionuclides become airborne and (2) modeling the
process by which the airborne radionuclides are transported to a human exposure location and
diluted before inhalation. The first step gives the ratio of the airborne emission near the
source before it is dispersed and diluted to the concentration in the resuspendable layer of
dust; the second step gives the ratio of the airborne concentration at the point of exposure to
the airborne emission at the source.
Ingestion pathways consist of food, water, and soil ingestion pathways. Four food pathway
categories are considered: plant foods, meat, milk, and aquatic foods. Analysis of food
pathways involves rather complex radionuclide transport in the relevant environmental media
and the subsequent uptake via various food chains. Terrestrial pathways and water pathways
are included. The terrestrial pathway includes the following four plant food pathways:
(1) root uptake from crops grown in the contaminated zone, (2) foliar uptake from
contaminated dust deposited on the foliage, (3) root uptake from contaminated irrigation
water, and (4) foliar uptake from contaminated irrigation water. The water pathways include
surface and well water. Both well water and surface water can be used for drinking. The
fraction of well water blended with or supplemented by surface water is used to calculate the
total contribution from groundwater and surface water. The ingestion pathway also includes
direct ingestion of contaminated soil itself. The dose due to ingestion of soil depends on the
amount of soil ingested and the radionuclide concentrations in the soil
RESRAD-BUILD: The RESRAD-BUILD code (Yu et al. 1994) is based on a room
compartmental model that analyzes the effects on room air quality of contaminant emission
and resuspension (as well as radon emanation), the external radiation pathway, and other
pathways such as air immersion and indirect ingestion. The RESRAD-BUILD computer code
is a pathway analysis model that was developed to evaluate the potential radiological dose (or
risk) incurred by an individual (see Figure 2). Because of the proximity of human-to-
contaminant contact during building decommissioning or dismantling, RESRAD-BUILD
requires more precise pathway modeling and input data than does RESRAD. For instance,
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JAERI-Conf 95-015
Source
Environmental Pathway
Exposure Pathway
Dose
Residual Radioactive Material in So
Direct Exposure
On-site
Air Contamination
-
u
te
*
Livestock
.
-* Aquatic F<
Dust
Radon
i
DdS
Meat
Milk
sods
On-site Biotic Contamination
k
Ground
Dust ^
Radon ^
Plants ^
Meat ^
Milk _
Fish ^
On-site Water Contamination
Direct Ingestion
Water ^
Soil ^
External
Radiation
Inhalation
Ingestion
».
— *
ndividual
Effective Dose Equivalent to an Exposed
FIGURE 1. Schematic Representation of RESRAD Pathways
(Source: Yu et al. 1993)
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JAERI-Conf 95-015
External Exposure
Direct from
the Source
O)
c
3
m
Q)
2
"35
c
OJ
to
"5
CO
o
CO
QC
Radon
Diffusion
Release Rate
of Radon into
Indoor Air
indoor Air
Concentration
of Radon Progeny
Inhalation of
Indoor Radon
Progeny
Mechanical
Removal of
Material from
the Source
Surface
Release Rate
of Radionuclides
into Air
Indoor Airborne
Concentration
of Radionuclides
t
1
Resuspension
i
i.
5
Deposition
on Floor
Surface
\
F
Ingestion of
Deposited
Material
Inhalation of
Airborne
Radioactive Dust
External Exposure
Due to Air
Submersion
External Exposure
Due to Deposited
Material
Innactinn rrf
Ingestion of
Removable
Material
Radionuclides
FIGURE 2. Schematic Representation of RESRAD-BUILD Pathways
(Source: Yu 1994)
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JAERI-Conf 95-015
a detailed description of contaminant distribution is needed to better define the
source-to-receptor configuration. The same applies to the room air quality.
Several important site-specific parameters — including building structural material, air
exchange, room size, and contamination thickness - are considered in assessing risk
associated with building decontamination. Thus, detailed modeling of the transport of
contaminants inside the building and a comprehensive exposure pathway analysis have been
considered in RESRAD-BUILD evaluation of human health risks resulting from building
contamination and material end-use scenarios.
The radioactive material in the building structure can be released into the indoor air by
mechanisms such as diffusion (radon gas), mechanical removal (decontamination activities),
or erosion (removable surface contamination). The RESRAD-BUILD code consists of two
major model components: air quality resulting from removal and transport of radioactive
material inside the building, and external exposure based on various source-receptor
configurations. The air quality model considers the transport of radioactive dust particulates
and radon progeny due to (1) air exchange between compartments and with outdoor air,
(2) deposition and resuspension of particulates, and (3) radioactive decay and ingrowth. The
external exposure model is based on the SOELD model (Chen et al. 1991), which also has
incorporated the latest recommendations of the EPA. The external exposure model,
additionally, allows variation in source geometry, shape, and source material.
Seven pathways are considered in the RESRAD-BUILD code: (1) external exposure to
penetrating radiation emitted directly from the sources, (2) external exposure to penetrating
radiation emitted from radioactive particulates deposited onto the floors of the compartments,
(3) external exposure to penetrating radiation due to submersion in airborne radioactive
particulates, (4) inhalation of airborne radioactive particulates, (5) inhalation of aerosol indoor
radon decay products, (6) direct, inadvertent ingestion of radioactive material contained in
removable material, and (7) inadvertent ingestion of airborne radioactive particulates
deposited onto the surfaces of the building. The first three pathways would result in external
exposure, while the others would result in internal exposure due to internal contamination of
the exposed individual.
In the RESRAD-BUILD model, the building is conceptualized as a structure composed of up
to three compartments. It can be a one-room warehouse, a two-room office or apartment, a
three-room ranch house, a three-story office building, or a two-story house with a basement.
Air exchange is assumed to occur between compartments 1 and 2 and compartments 2 and 3
but not between compartments 1 and 3. All compartments can exchange air with the outdoor
atmosphere.
An air quality model was developed to calculate the contaminant concentration in each
compartment A coordinate system is used in RESRAD-BUILD to define the location of the
sources and receptor points inside the building. With the user specifying the locations of the
sources and receptors, as well as the time each receptor spends in each compartment, the
radiological dose to the receptors can be calculated for any type of building use, including
residential, commercial, or industrial. The analysis of the inhalation pathway also includes
consideration of the emanation of radon and concentrations of its progeny. Therefore, the
building model approach used in RESRAD-BUILD is quite flexible.
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JAERI-Conf 95-015
The building is assumed to be contaminated with radioactive materials located at a defined
number of places within the structure of the building. Each contaminated location in the
building is considered a distinct source, and as many as 10 sources can be specified in a single
run of RESRAD-BUILD. Depending on its geometric appearance, the source can be defined
as a volume, area, line, or point source. The distinction between these types of sources is
rather arbitrary and reflects the modeling objective of simplifying the overall configuration,
whenever justifiable. The proper classification of each source is left to the user's best
judgment.
RESRAD-PROBABDLIST1C: The RESRAD-PRQBABILISTie code employs an existing
Monte Carlo sampling algorithm as a driver to simulate statistic distribution of the RESRAD
output results. The algorithm, based on the Latin Hypercube Sampling (LHS) method (Iman
and Shortencarier 1984), has been developed by Sandia National Laboratories. Thus,
RESRAD-PROBABILISTIC basically represents an integration of the LHS driver and the
RESRAD code, plus a postprocessor for data compilation and treatment. Input to the LHS
sampling requires parametric characteristics such as data distribution characteristics, mean
values, as well as standard deviations. An effort has been initiated at ANL to compile data
on parameter distributions, distribution characteristics, and correlations among parameters.
SCENARIOS AND APPLICATIONS
RESRAD: Many parameters that determine the quantity of radionuclides or radiation to
which an individual is exposed are determined by exposure scenarios; that is, patterns of
human activity that can affect the release of radioactivity from the contaminated zone and the
amount of exposure received at the exposure location. For RESRAD, soil guidelines are
based on a family farm exposure scenario. This scenario includes all environmental pathways
for on-site or near-site exposure and is expected to result in the highest predicted lifetime
dose. Other scenarios, such as the industrial worker and recreationist, can be taken into
account by adjusting the scenario parameters in formulas for calculating the transport of
radionuclides through the pathways.
Soil guidelines are based on on-site exposure because on-site residents will receive a radiation
dose that is at least as large as (and generally larger than) the dose to off-site residents. The
external radiation dose will decrease rapidly with distance from the site, and secondary off-site
sources — such as surface deposits of airborne contaminated soil or water contaminated by
radionuclides leached from the soil - will have lower radionuclide concentrations. The
contributions from inhalation pathways will decrease with distance from the site for the same
reasons. The largest contribution from the groundwater pathway will be for drinking water
from the unconfined aquifer tapped by a well at the downgradient boundary of the
contaminated area. This contribution can be the same for on-site and near-site residents but
will decrease for wells at greater distances from the boundary. The situation is more
complicated for foodchain pathways because reconcentration can occur. However, the
predominant contribution is from on-site crops and domestic animals, and this contribution
will be greatest for on-site residents who raise food for their own consumption.
_ 107_
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JAERI-Conf 95-015
The assignment of appropriate values to the scenario parameters is based on existing patterns
of human activity that can be expected to persist for an indefinite period. For most scenario
parameters, this criterion enables a straightforward determination of parameter values on the
basis of data for current conditions.
Over the years, RESRAD has undergone major improvements in terms of both added
modeling capability and input database. The latest updates are described in the RESRAD
Manual, Version 5.0 (Yu et al. 1993). RESRAD is a participating code in the international
verification and validation (BIOMOVS) effort The code has been used widely by DOE and
its contractors and to a certain extent outside of DOE. To date, some 30 RESRAD workshops
have been conducted by ANL. Several major DOE programs have successfully utilized the
code in assessing human health risks and developing site cleanup criteria.
RESRAD-BUILD: The RESRAD-BUDLD code is designed with flexibility and simplicity
in mind so that it can evaluate diverse exposure scenarios for a contaminated building, such
as office work, building cleaning and maintenance work, building decontamination, building
renovation, building visits, and continuous residency. The receptors considered in the
RESRAD-BUILD model include office worker, resident, industrial worker, decontamination
worker, building visitor, or any other individual spending time inside the contaminated
building. The exact location (coordinates) of the receptor is required to calculate external
exposure. The receptor location should be the midpoint of the person. For example, if the
receptor is standing on a contaminated floor, the receptor location should be 1 m above the
floor. For other pathways, the code requires information only about the room in which the
receptor is located because the air quality model assumes that the air concentration is
homogeneously mixed in each compartment. To calculate the external dose, input parameters
to the code include the receptor location and the shielding material type, density, and
thickness. The orientation of the receptor to the source can also be selected, that is, rotational
or facing the source (anterior-posterior). The anterior-posterior orientation will result in a
higher direct external dose than the rotational orientation.
Up to 10 receptor points can be specified in the RESRAD-BUILD code. A time fraction spent
at each receptor point needs to be input. The total time can exceed unity, thus allowing a
single run of the RESRAD-BUILD code to evaluate total (collective) worker dose and total
individual dose.
RESRAD-BUILD is currently completed as draft and has been reviewed by DOE. The code
has been applied successfully for analyzing scenarios described in a separate DOE effort to
assess potential release standards by calculating human health risks from radioactive scrap
metal recycle and reuse (Murphie et al. 1993).
RESRAD-PROBABILISTIC: Scenarios used for RESRAD-PROBABILISTIC are similar
to those used by RESRAD, except RESRAD-PROBABILISTIC accepts input as distribution
functions and provides a range of output results. Figure 3 shows an example of
RESRAD-PROBABILISTIC output that provides a cumulative probability distribution curve
for the resulting dose.
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JAERI-Conf 95-015
1.0
•o
0)
•o 0.8 H
CU
O
X
w 0.6 H
w 0.4 -
O
o
O 0.2 -
CU
0.0
30
40 50 60
Peak Dose Rate (mrem/yr)
70
FIGURES. Sample of Cumulative Probability Distribution Calculated by
RESRAD-PROBABILISTIC
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JAERJ-Conf 95-015
SUMMARY AND CONCLUSIONS
A RESRAD code series is under development at ANL. These codes are designed to perform
human health risk analysts based on multimedia environmental pathway models. Each code
is tailored to meet a specific objective of human health risk assessment, requiring specific
parameter definition and data gathering. Through continued improvement and incorporation
of state-of-art methodology and data information, the combined capabilities of these codes
serve to satisfy various risk assessment requirements in environmental restoration and
remediation activities.
REFERENCES
Chen, S.Y., et al., 1991, "Calculating External Doses from Contaminated Soil with the
Computer Code SOILD," Transaction of American Nuclear Society, Vol. 64, pp. 75-76,
American Nuclear Society, LaGrange Park, 111.
Iman, R.L., and M.J. Shortencarier, 1984, A FORTRAN 77 Program and User's Guide for the
Generation of Latin Hypercube and Random Samples for Use with Computer Models,
NUREG/CR-3624, SAND83-2365, Sandia National Laboratories, Albuquerque, N.M.
Murphie, W., et al., 1993, "Assessment of Recycling/or Disposal Alternatives for Radioactive
Scrap Metal," ER'93, Meeting the Challenge, Environmental Remediation Conference
sponsored by U.S. Department of Energy, Augusta, Ga., Oct. 24-24.
U.S. Department of Energy, 1990, Radiation Protection of the Public and the Environment,
DOE Order 5400.5, Washington, D.C.
U.S. Nuclear Regulatory Commission, 1994, Proposed Federal Register Notice, Draft
Radiological Criteria for Decommissioning, 10 CFR 20, Washington, D.C.
Yu, C., 1993, "Development of Uncertainty Analysis Capability for the RESRAD Computer
Code," Transactions of American Nuclear Society, Vol. 69, pp. 30-31, American Nuclear
Society, LaGrange, Park, 111.
Yu, C., et al., 1993, Manual for Implementing Residual Radioactive Material Guidelines using
RESRAD, Version 5,0, ANL/EAD/LD-2, Argonne National Laboratory, Argonne, 11
Yu, C. et. al., 1994, unpublished information, Argonne National Laboratory, Argonne, 111.
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JAERI-Conf 95-015
2-6 Unrestricted Release of Contaminated Lands
and the Dose to the General Public
Hideaki Yamamoto and Shohei Kato
Department of Health Physics
Japan Atomic Energy Research Institute
Abstract
Japan has not encountered any environment restoration problems with serious
radioactive contamination in land. Thus, the issues on release of contaminated land
only relate to site release after normal termination of radiation-related practices. For
example, a criterion is established for unrestricted release of land which is
contaminated after the normal operation of radioactive wastes disposal. The
criterion is based on the dose to the general public.
This paper outlines the land release issues in Japan.
introduction
For the purpose of this paper, a land which is provided for a nuclear-related activity
is defined as a site where locates the facility for the activity including buildings (both
above-ground and underground structures). The definition of a land also includes a
sector of the earth where a radiation source exists, or where an effect of radiation
from another sector is found. In this case, the land may extend to the environment
outside of the nuclear site.
Release of a land means a withdrawal of a part of or all restriction on the use of the
land for radiation protection purposes. Release of a land follows 1) termination of a
radiation-related practice in the land or 2) remediation of the environment by
cleanup of a residual radiation source in the land. By the release the land will be
made available for use of the nuciear industry or public. When nothing is required
in the way of use from radiation protection standpoint, the land is under an
unrestricted release.
in -
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JAERI-Conf 95-015
The Japanese society has not encountered any serious environment remediation
problem. Accordingly, the paper only deals with release of a land in connection with
termination of radiation-related practices in the land.
Inventory
The radiation-related practices in Japan are restricted within those for peaceful
purposes. They can be categorized in the activities 1) in the nuciear fuel cycle, 2} for
industrial use, 3) for medical use, and 4} for the purposes of research/development
and education. Kato et al. summarized the inventory of lands where these activities
are conducted [1]. Initiation of these practices has to follow a legal procedure:
notification or licensing [2], [3]. The legal procedure also prescribes the
requirements which a licensee shall comply with at the termination of his practices;
1) complete removal of radioactive wastes from the land ,and 2) decontamination of
the land.
For most of the practices a criterion for unrestricted release of the land is that at the
termination of the practice there shail be no residual contamination in the land [4].
This criterion can be easily implemented because in most cases the contamination
remains in limited parts inside the facility which can be completely removed without
generating a large amount of subsequent radioactive wastes. An exception is the
practice of shallow land disposal of radioactive wastes: a dose-based criterion is
introduced in this case.
Regulations of shallow land disposal of radioactive wastes
A nuclear facility is usually designed to contain radioactive materials inside its
boundary. But this is not the case of a facility for shallow land disposal of radioactive
wastes: the facility is not required to contain radioactive materials which may reach
from the wastes. The repository allows the radioactive contamination to disperse
even outside its site. This brings a unique feature to regulation of shallow land
disposal, which includes land release issues.
The Nuclear Safety Commission of Japan has established a safety guide for shallow
land disposal of low-level radioactive solid wastes [5], [6]. The guide addresses
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fundamental guidelines to protect public health: radiation protection programs shall
be successively modified in accordance with the level of potential hazard of the
repository which generally reduces with time, and the burial land shali be eventually
released unrestrictedly.
As a part of the modification of radiation protection, release of land will be
implemented in two steps, A restricted release will be allowed in the first step when
the radiation level on the repository becomes sufficiently low. The general public can
freely access to the land of the repository, and use it for other purpose under certain
restriction. This step corresponds to a period of active institutional control.
In the second step the land will be released unrestrictedly: all of the radiation control
over the repository will be removed. The license of the operation of the facility will
terminate, and the land wili be exempted from radiation-related regulations.
The Commission also set the licensing guidelines that any radiation control of a near
surface repository shall be able to terminate within a period of 300 - 400 years after
the start of the operation. This means that any repositories shall be planned so as to
be exempted from regulatory control after about 400-year operation. This
requirement determines the initial radioactivity concentration and total radioactivity
of the wastes which can be accepted to a repository.
Criteria for unrestricted release of shallow land burial site
The Nuclear Safety Commission set the criteria by which release of shallow land
burial site are allowed. The annual effective dose equivalent to the general public
shall not exceed 1 mSv for restricted release, and 10 p,Sv for unrestricted release.
The latter criterion is the dose to be exempted from regulatory control recommended
by the Radiation Council [7]. Licenses for operation of shallow land disposal of
radioactive wastes shall be granted by the Japanese Government when licensees
can demonstrate the compliance with these criteria through their dose estimation.
The first commercial low-level radioactive wastes disposal was licensed, and the
repository is in operation by Japan Nuclear Fuel Industries Company Incorporation
since 1992. The site, Rokkasho Storage Center, locates in Shimokita Peninsula,
Aomorr Prefecture. The license is granted for buring 40,000 cubic-meter low-level
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JAERI-Conf 95-015
radioactive wastes in reinforced concrete vaults which are installed in depth of 14-19
meters underground. The site is permitted to accept wastes generated from nuclear
power plants only [8].
The licensee plans to release the burial land in unrestricted condition after about
300 years. Assuming the free access of the general public to the land after the
release, radiation exposures to the public through various pathways were evaluated
in the licensing procedure. The highest dose was estimated in the pathway of
residence in the burial land: annual effective dose equivalent of 0.15 p,Sv. This is
well below the Nuclear Safety Commission's unrestricted release criterion, 10
lASv[9],
Derived soil concentration
In addition to the dose estimation in the licensing procedure, the licensee of the
shallow land disposal shall demonstrate the compliance with the unrestricted
release criterion when he actually releases the land. The Nuclear Safety
Commission considers that a practical approach to demonstrate the compliance
would be to confirm that measurements of radioactivity concentration in the
contaminated soil from the site do not exceed a concentration limit which is derived
in advance from the 10 uSv criterion [5]. This is because after the release of the land
the contaminated soil in the site would be the environmental medium most likely
accessible for the general public.
In many exposure pathways with radiological importance after the site release,
radiation dose to the general public can be assumed to be directly proportional to
radioactivity concentration of the soil in the site. This simple relationship will make it
easier to derive a concentration limit corresponding to 10 jxSv/year. Pathways to be
considered will include;
- external exposure to the soil
- inhalation of the resuspended soil
- ingestion of radioactive material via food chain.
The licensee will select some land use scenarios which define the exposure
condition of the general public at his released land. Site-specific parameters will be
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JAERI-Conf 95-015
used in the dose/concentration calculation.
Summary
Japan has had no natural environment with serious radioactive contamination so far.
Thus, the issues on release of contaminated land only relates to the site release
after normal termination of the operation. A dose-based criterion for unrestricted
release of contaminated land is established for shallow land disposal of low-level
radioactive wastes. In this case, a practical way of implementing the criterion is to
develop radioactivity concentration limits for contaminated soil of the disposal site.
References
[1] Kato, S. et al.:" Site inventory of residual radioactivity in Japan", EPA 520/1-90-
013, pp. 44-56(1990).
[2] "The law concerning prevention from radiation hazards due to radioisotopes,
etc." (1957).
[3] "The law for regulation of nuclear material, nuclear fuel material and reactors",
(1957).
[4] Yamamoto, H. et al.: "Experience in decontamination and reuses of the large-
scale radiochemica! laboratory and the research reactor at the Japan Atomic
Energy Research institute", EPA 520/1-90-013, pp. 149 -156 (1990).
[5] Nuclear Safety Commission: "Regulatory Policy for land disposal of low-level
radioactive solid wastes" (1985).
[6] Nuclear Safety Commission: "Guidance for licensing of land disposal of low-level
radioactive solid wastes" (1990).
[7] Radiation Council: "Dose to be exempted from regulations concerned with
shallow land disposal of radioactive wastes" (1985).
[8] Simoda, H. et al.: "Shallow land disposal of low-level radioactive waste by
burying in concrete vault: design and construction of the facility at Rokkasho
Storage Center", in the Proceedings of "The Third International Conference on
Nuclear Fuel Reprocessing and Waste Management: RECOD '91", Atomic
Energy Society of Japan (1991).
[9] Nuclear Safety Commission: "Monthly report", Vol. 13, No. 12, Nuclear Safety
Commission (1990) (in Japanese).
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JAERl-Conf 95-015
2-7 STATUS OF ENVIRONMENTAL RESPONSE EFFORTS AT RADIOACTTVELY
CONTAMINATED SITES IN THE UNITED STATES AIR FORCE
INSTALLATION RESTORATION PROGRAM
William D. Rowe, Jr., Pb.D
Thomas E, McEntee, Jr., Ph.D.
The MITRE Corporation
Brent Johnson
United States Air Force
Environmental Quality Directorate
Capt Lonny Manning, PhD
United States Air Force
Low-Level Radioactive Waste Office
ABSTRACT
The United States Air Force has identified approximately 170 radtoactively contaminated sites at its
domestic installations. These sites contain a variety of low level radioactive and mixed wastes and are
classified as burial sites, landfills, buildings, and other disposal sites. Of these 170, approximately 70 are
presently being evaluated under the Air Force Installation Restoration Program (KP) in accordance with
applicable laws and regulations. Removal and/or remedial actions have been taken at specific sites using
site-specific residnal radioactivity criteria. The remaining sites are either under investigation to determine
the need for possible action or have been classified as response complete based on restricted or
unrestricted future use. This paper describes past Air Force operations that generated radioactive waste
materials; examines the current inventory of resulting radioactively contaminated sites in the Air Force
IRP; reviews criteria used to evaluate sites for removal and/or remedial actions; provides summary
information on actions taken at sites; and focuses on response actions and cleanup levels at two completed
sites. The paper concludes with an assessment of outstanding issues relevant to the remediation of
radioactively contaminated sites.
INTRODUCTION
The Air Force is in the process of planning and executing investigations, removal actions, interim
remedial actions, and cleanups at thousands of contaminated sites across the country. These sites are being
managed under the Air Force Installation Restoration Program {IRP) at both active installations and
installations scheduled for closure, known as Base Realignment and Closure (BRAC) installations. The
primary objective of the IRP is to complete response actions at contaminated sites so that risks to human
health and the environment from contamination are reduced to acceptable levels.
The IRP was formally established in 1984 with the creation of a fund known as the Defense
Environmental Restoration Account. The United States Congress funds this account each year at a level
that allows the Air Force to execute many of its IRP projects, including those at radioactively
contaminated sites.
Less than 5 percent of the 4100 sites addressed under the Air Force IRP are potentially contaminated with
radioactive materials. Management of response actions for these sites is accomplished at both the
headquarters and field level within the Air Force through several specific organizations and their
functional representatives. Key Air Force organizations and associated functional responsibilities are
listed in Table 1.
This paper was not presented at the Workshop because the authors were unable to attend the
meeting. (The Editors)
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JAERI-Conf 95-015
TABLE 1. ORGANIZATIONAL RESPONSIBILITIES FOR RADIOACTWELY CONTAMINATED SITES
ORGANIZATION PRIMARY AREAS OF RESPONSIBILITY
Office of the Civil
Engineer,
Headquarters United
States Air Force
(HO USAF/CEVR)
Air Force Major
Commands
(MAJCOM)
Installation
Commanders
AFBCA
Air Force Low-Level
Radioactive Waste
Office
(SA-ALC/EMP)
Air Force Medical
Operations Agency
(AFMOA/SGP)
Air Force
Radioisotope
Committee Executive
Secretariat
(AFMOA/SGPR)
Deputy Chief of Staff
for Logistics
(HQUSAF/LG)
Develops policy, allocates resources, and oversees execution of the
environmental restoration program throughout the Air Force at active (Le., non-
BRAC installations), including execution of response actions at radioactively
contaminated sites.
Provide execution guidance and oversee implementation of the environmental
restoration program, and related activities, at the installations and facilities
under their jurisdictions.
Responsible for the environmental condition of an installation, including all
environmental restoration activities. Restoration activities for all sites, including
those containing radioactive materials, are usually assigned to the
Environmental Management Office at an Air Force Installation.
Headquarters component is responsible for developing policy for radioactively
contaminated sites at BRAC installations. Headquarters allocates resources and
oversees execution of the environmental restoration program at all BRAC
installations, including execution and response actions at radioactively
contaminated sites.
Provides a contracting mechanism for executing environmental restoration
activities at most of the radioactively contaminated sites in the Air Force. Directs
wastes from restoration activities to cost-effective commercial burial sites.
Receives disposition requests from Air Force generators; generates letters of
instructions for packaging and shipping of low-level radioactive wastes; and
maintains an inventory of all low-level radioactive waste disposal. Operates a
recycling program for depleted uranium, Kr-85, and other isotopes. Provides
technical services to installations on radioactive waste investigation and disposal
Issues.
Develops radiation safety policy and policy for low-level radioactive waste
disposal in the Air Force.
The Radioisotope Committee oversees Air Force use of radioactive materials;
serves as the single point of contact for the Air Force Master Materials License
issued by the U.S. Nuclear Regulatory Commission; and sets up administrative
controls to acquire, receive, store, distribute, use, transfer, and dispose of
radioactive materials. The Radioisotope Committee Executive Secretariat
reviews and approves Air Force permit applications and other requests to use
radioactive materials; responds to radioactive materials incidents and accidents
to ensure that permittees comply with all rules and regulations; and approves
Statements of Work for invasive characterizations or exhumations of radioactive
materials at sites.
Maintains a program to manage and dispose of radioactive waste and
coordinates radioactive waste disposal among Air Force components and other
parties in accordance with Air Force technical requirements.
Roles and responsibilities of these and other Air Force organizations are described further in Air Force
Instruction 40-201, Managing Radioactive Materials in the USAF, 25 My 1994.
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AIR FORCE USE AND DISPOSAL OF RADIOACTIVE MATERIALS
The Air Force has used and continues to use naturally occurring, reactor produced, and accelerator
produced radioactive materials as part of its operations. Air Force uses of radioactive materials fall into
six general categories. These are listed in Table 2 along with examples of operations and equipment that
employ radioactive materials.
TABLE 2. AIR FORCE Usis OF RADIOACTIVE MATERIALS
USE CATEGORY EXAMPLES OF OPERATIONS/EQUIPMENT
Industrial
Weapon Systems
Research and
Development
Medical
Commodities
Environmental
Calibration, depot maintenance (e.g, aircraft counterweights), radiography, fixed
gauges, power sources
In-flight Blade Integrity System, target designations, airframe, munitions, dials
and gauges, engine exciter, counterweights, Astroinertial Navigation System,
aerospace payloads, weapon simulators
Gas chromatographs, biomedical studies, environmental tracers, irradiators,
radioanalysis, munitions testing, space launches, radiation effects studies
Nuclear medicine, radiation oncology, clinical laboratories, clinical
investigations
Chemical agent detectors, lead paint analyzers, static meters, troxler gauges, exit
signs, compasses, electron tubes, thoraited optics
Waste site excavations, contaminated buildings, training sites, ranges {e.g.
depleted uranium)
Most radioactively contaminated sites in the Air Force were created during the 1950s in accordance with
the Atomic Energy Commission (AEC) policy and general industrial practices at that time. Detailed
records of disposals were not generally required or kept Wastes authorized for disposal consisted
primarily of:
* Electron tubes containing both AEC and non-ABC regulated radioactive materials in solid form
• Non-AEC regulated low-level solid and liquid wastes from weapons maintenance
» Radioactive self-luminous dials, gauges, and circuit breakers containing non-AEC regulated
radium-based paint
» Wastes from non-AEC regulated radium dial painting operations
Radioactively contaminated sites resulting from the use and disposal of these materials include "pipe
sites," contaminated buildings, and landfills. Many onsite burials of licensed materials were made in
accordance with the conditions of a specific AEC license issued to an installation. Onsite burial, however,
was discontinued within the Air Force by 1965 unless granted on an exception basis.
Guidance on constructing and maintaining burial sites was published in technical order procedures which
included identifying site location on appropriate maps, and posting and fencing to prevent unauthorized
entry. The Air Force began radioactive waste disposal at licensed commercial sites in the late 1950s;
previous technical order requirements for waste burial/site maintenance were rescinded. No alternative
instructions were developed on radioactive site maintenance and a gradual loss of site records ensued.
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In 1971, the Air Force initiated an effort to find and consolidate existing site records and reestablish
maintenance requirements. In 1982, the U.S. Nuclear Regulatory Commission (NRC) approached the
Air Force with a single broad Icense concept for all source, byproduct, or special nuclear material which
eventually led to the issuance of a Master Materials License to the Air Force in 1985. Conditions of the
Master Material License provided tor the use of authorized materials as approved by the Air Force
Radioisotope Committee, and gave the Radioisotope Committee responsibility for maintaining the
Air Force Radioactive Materials Program (see Table 1).
Since issuance of the Master Materials License in 1985, onsite burials of radioactive materials have been
prohibited. However, many former "pipe sites", contaminated buildings, and landfills have been identified
for evaluation under the Air Force IRP. Pipe sites were constructed for the disposal of commodity items
such as check sources, electron tubes, and self-luminous parts (e.g., radium).
A typical pipe site consists of a concrete/metal culvert pipe approximately 20-30 feet long, and 18-30
inches in diameter with one end capped. The capped end was placed vertically in the ground, leaving the
open end flush with the ground surface. Commodities were placed inside the pipes until they were full, or
until the Air Force discontinued onsite burial, at which time the remaining end of the pipe was capped. To
date, no radiation levels above background (Le., local naturally occuring levels) have been detected on the
outside of these pipes.
In addition to pipe sites, there are a number of radium contaminated buildings that have Seen, or are
currently being, addressed under the Air Force IRP. Informally known as paint shops, these facilities were
used to repair and repaint luminous dials, guages, and in some instances, signs.
STATUS OF IUDIOACTTVELY CONTAMINATED SITES
To date, approximately 170 radioactively contaminated sites have been identified and evaluated for
inclusion in the Air Force IRP. These sites are distributed across 84 domestic facilities including former
Air Force properties and 14 BRAC installations. A review of Air Force databases for radioaetiveiy
contaminated sites revealed preliminary investigations to be ongoing at 12 percent of the sites; in-depth
investigations to be underway at 33 percent of the sites; remediation to be ongoing at 10 percent of the
sites; and no further action or undetermined action planned at the remaining 45 percent of the sites. As
mentioned, most sites are shallow burial or disposal areas and consist of radium dials, gauges, electron
tubes, aircraft components, hospital wastes, counter weights, paint residues, radioactively contaminated
clothing, and low-level radioactive wastewater. In some cases, these materials are mixed with other
hazardous or non-hazardous solid wastes. Most remedial actions to date have involved complete site
exhumation and disposal of waste materials in offsiie commercial facilities. Pipe site exhumations have
been conducted using background as the remedial action objective.
CLEANUP HIGHLIGHTS
Several successful remediation efforts at radioactively contaminated sites in the Air Force have been
documented. Two such efforts are described below, one for a pipe site and the other for a contaminated
building.
Bergstrom Air Force Base
Radioactive Waste Site No. 24 (RW-24) at Bergstrom Air Force Base, Texas was used for the disposal of
low-level radioactive materials such as luminous aircraft dial gauges and electron tubes. RW-24 is typical
of the many "pipe sites" that were constructed by the United States Air Force for the disposal of
commodity items such as check sources, electron tubes, and self-luminous parts.
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RW-24 consisted of three 20-feet long vertically emplaced cast iron pipes, two having 18-inch diameters
and one having a 12-inch diameter. Following each deposit of low-level radioactive material in the pipes,
a charge of concrete was reportedly added. The pipes were installed in the 1950s and were reportedly
sealed with lead-lined metal caps in 1971. The pipes were backfilled with dirt and a 4-inch thick slab of
concrete was poured to cover each pipe. Annual radiological surveys, conducted at the surface of the site,
found no activity above background levels. Drinking water wells within a one-mile radius of RW-24 were
tested by the Texas Department of Health in 1991 and all found to be below the U.S. Environmental
Protection Agency (EPA) established limits for public drinking water supplies.
Remediation activities at site RW-24 were undertaken by the Air Force Base Conversion Agency, in
coordination with the Air Force Low-Level Radioactive Waste Office (SA-ALC/EMP) and the U.S. Army
Corps of Engineers. The remedy selected was a controlled excavation of the pipes followed by waste
packaging in preparation for transport to a permitted low level radioactive waste disposal facility.
Materials were disposed of as Naturally Occurring Radioactive Material based on a determination
completed by the State of Washington. The site was designated for reiease to unrestricted use.
McCleUan Air Force Base
Building 252 at McClellan Air Force Base, California was built during the 1930s and operated as an
instrument repair facility until the late 1980s. From 1940 to 1960, radium paint was applied to instrument
dials. In early 1980, most of the existing operations were relocated and renovation activities were initiated
to convert the building into office space. During renovation, asbestos was found throughout the building;
mercury was found in portions of the building. Remediation of the asbestos and mercury was completed in
1992. A radiological characterization study in mid-1994 determined that several areas within Building
252 exceeded NRC Regulatory Guide 1.86 limits. Radiological decontamination of the facility was
initiated in late 1994.
The primary objective was to decontaminate Building 252 sufficiently to allow the eventual release of the
facility from licensing restrictions. Specifically, the decontamination activities were to; (1) decontaminate
interior and exterior surfaces within the basement and two floors of the building to "unrestricted use"
limits identified in NRC Regulatory Guide 1.86; (2) remove a contaminated drain pipe and surrounding
soil until soil levels of 5 pCi/g or less were achieved; and (3) package all generated radioactive waste for
disposal at a permitted low level radioactive waste disposal facility. Although NRC Regulatory Guide 1.86
was the decontamination standard, decontamination activities were conducted using the "As Low As
Reasonably Achievable" (ALARA) concept
Decontamination techniques employed at Building 252 included the use of vacuum blasters, shot blasters,
abrasive grinders, and brush hammers. Portable high-efficiency paniculate air vacuum systems were used
to remove contaminated debris generated. Decontamination of Building 252 was completed with
approximately 24,000 ft2 of concrete floor surface, 850 ft2 of wall and ceiling surface, and 500 Enear feet
of floor-wall junction remediated. In addition, approximately 30 feet of drain pipe and 1200 ft3 of
contaminated soil were removed and packaged for disposal. Typically, concrete floor and stair surfaces
required the removal of 1/16 to 1/4 inches of the concrete surface using the shot blaster, abrasive grinder,
and brush hammer. Floor-wall junctions were decontaminated by the removal of between 1/4 and 1/2
inches of material using a combination of abrasive grinders and brush hammers. Decontamination of wall
surfaces was achieved using a combination of vacuum blasters and brush hammers. The majority of wall
surfaces were remediated with approximately 1/32 inch of surface material being removed. Areas
requiring the use of the brush hammer were remediated to depths of approximately 1/4 inch.
Approximately 1400 ft3 of radioactive waste was generated during the decontamination activities. All
waste was packaged in U.S. Department of Transportation-approved 17-H 55-gallon drums and sent to the
permitted low level radioactive waste disposal facility.
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The goal of the remediation of Building 252 was to decontaminate building surfaces to background levels.
If this was not practical, surfaces were decontaminated until they met the project free release criteria of
100 dpm/lQO cm* for average alpha contamination. Following completion of the initial decontamination
activities in Building 252, a verification survey was conducted to determine the effectiveness of the
remediation. For this verification survey, more than 6400 stationary survey readings were taken. 104
points in the building were identified during the survey to be still contaminated above project free release
limits after primary decontamination was performed. All of these spots were marked and subsequently
further decontaminated to meet the project free release limits. The final results of the verification survey
confirm that the facility has been decontaminated to meet the free release limits of NRC Regulatory
Guide 1.86.
OUTSTANDING ISSUES
While the evaluation of radioactively contaminated sites will continue under the Air Force IRP, issues
related to radiation cleanup criteria and access to commercial disposal facilities are of particular concern
to Air Force organizations responsible for radiation site response actions. The question of risk-based
criteria versus cleanup to background or tabulated numerical levels by media is currently unresolved at the
national level. As indicated in the first ease history, restoration to background typically is achieved by
complete site exhumation and off-site disposal, as opposed to burial in place. Use of risk-based criteria
would allow sites with low risk to remain in place. The Air Force is actively supporting studies on the
benefits of using risk-based standards.
The question of access to commercial disposal facilities poses a potential waste disposal problem for the
Air Force. Current access to commercial disposal facilities is limited given that some have closed, while
the opening of new facilities has not occured due to significant delays in the permitting process. As a
result, the Air Force has resorted to a combination of storage and recycling of low-level radioactive waste.
Recycling of depleted uranium, Kr-85, and other isotopes has been successful. Other types of radioactive
waste must be stored or disposed of, although disposal may become increasingly expensive given the
limited access to current commercial disposal facilities and lack of new facilities in the near term.
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3. Recycling and Criteria
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Evaluation of the Costs and Benefits of Recycling
Radioactively Contaminated Scrap Metal
by
Eugene C. Dunnan
Deputy Office Director
Office of Radiation and Indoor Air
U.S. Environmental Protection Agency
Peter Tsirigotis
Economist
Office of Radiation and Indoor Air
U.S. Environmental Protection Agency
John A. MacKinney
Environmental Scientist
Office of Radiation and Indoor Air
U.S. Environmental Protection Agency
ABSTRACT: The U.S. Environmental Protection Agency (EPA) is evaluating the economic and
technical issues associated with the potential recycling of radioactive scrap metals (RSM). These
metals, usually only slightly contaminated, originate primarily from the decommissioning and
decontamination (D&D) of federal facilities, licensees of the Nuclear Regulatory Commission, and
certain unlicensed industries. EPA conducted a study entitled Analysis of the Potential Recycling
of Department of Energy Radioactive ScrqpMetftl, September 6, 1994, for the U.S. Department of
Energy (DOE) to provide information and tools to DOE for assessing DOE's problem with RSM
from the D&D of their sites.
EPA is now initiating an evaluation of RSM recycling to support a recycling regulation.
Although the study prepared for DOE will provide a useful start for the regulatory analysis,
additional information must be gathered to analyze the impacts of a recycling regulation that will
apply to all potential generators of RSM. This paper summarizes cost-benefit issues related to an
RSM recycling regulatory analysis, including: the quantity of potentially recyclable contaminated
metals; costs of disposal at federal and private waste repositories: all potential environmental,
health, and safety, and market impacts: and the potential for adverse effects on radio-sensitive
industries,
1. Radioactive Scrap Metal: Does Recycling Make Sense?
Under the Atomic Energy Act, the U.S. Environmental Protection Agency (EPA) is
authorized to develop federal guidance and regulations to protect public health and the environment
from the effects of radiation. EPA is now developing standards for the management and disposal
of low-level radioactive wastes (LLW), including those generated from decommissioning of nuclear
power plants and cleanup of contaminated sites.
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A major concern in managing wastes from power plant decommissioning and site cleanup
is what to do with the large amounts of radioactive scrap metal {RSM)—primarily steel, but also
nickel, aluminum, copper, lead, and other metals—from dismantled structures and equipment.
Historically, such materials have been disposed by burial or long-term storage (MUR), but these
methods have two disadvantages:
* Disposal costs are mounting rapidly. The cost to dispose of soil contaminated with
LLW at commercial disposal facilities is currently projected at about $1,500 to
$5.000 per cubic meter.1 Disposal costs are likely to increase dramatically as the
capacity of suitable disposal sites decreases. Additionally, disposal capacity is not
expected to increase soon because of the difficulties involved in siting new
facilities.
• RSM is a commodity that metal producers may be willing to buy. By disposing of
RSM, the federal government sacrifices the potential commodity value of RSM.
Much of the RSM that may be available, however, would need to be
decontaminated before sale on the scrap metal market. For some RSM, the
commodity value may not cover the cost of decontamination.
For these reasons, recycling of RSM appears to many observers like a favorable alternative to
disposal. Other observers of RSM recycling, however, question whether it is more cost-effective
than disposal and point out that recycling may have other economic and non-economic impacts that
also need to be considered.
EPA is considering including a recycling provision in its forthcoming Low Level Waste
Management rule. Before the Agency can adopt such a regulation, however, it must have a clear
idea of the likely costs, benefits, and other impacts of RSM recycling. EPA is now at the very
beginning of an investigation to gather that information, and a number of fundamental questions
must be answered before the costs and benefits can be identified.
In this paper, we outline some of the fundamental questions EPA is facing and the
approaches the Agency is considering to answer them. The principal issues to be investigated are
described in subsequent sections of this paper; we summarize them below.
• Compared to disposal, is RSM recycling cost-effective? If LLW disposal capacity
is limited in the future, or if disposal costs are sufficiently high, recycling may be
cost-effective even if RSM must be sold below the going price for scrap metal.
» How will RSM recycling affect the environment and the health and safety of the
public and workers? Disposal and recycling both have potentially adverse
environmental impacts (e.g.. environmental degradation from LLW disposal and
metal ore mining), and health and safety impacts {e.g.. radiation exposure and
occupational hazards from LLW disposal. RSM decontamination, and metal
products produced with RSM).
1 Low figure is based on disposal estimate from a commercial LLW disposal facility.
High estimate is based on NRC 94.
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» Will RSM recycling change the structure of the metal industiy and other related
industries? For example:
— Marfcet differentiation. Could introduction of large quantities of RSM into
the various markets for metal cause radio-sensitive industries (e.g.,
photographic products, electronics, and medical instruments) and industries
that cater to sensitive populations (e.g., schools, day care centers, and new
home construction) to demand metals made without RSM, or to incur new
costs to screen metals they purchase for residual radiation? If so. how
would this affect the structure of the metal markets? For example, could
separate markets for metals produced with and without RSM develop;
would industry create standards to grade metal by level of residual
radiation?
— Price and demand effects. If large quantities of RSM enter the metal
markets, what will be the effect to the price and demand of raw materials
(e.g., raw metal ore and scrap metal), intermediate products like iron ingots,
and final metal products?
— International considerations. What effect would RSM recycling have on
international markets? For example, if market prices for intermediate and
final metal products decline due to the increased supply of RSM. could the
U.S. increase sales abroad? Alternatively, would nations erect trade
barriers by restricting U.S. metal purchases fearing the health effects
associated with increased levels of residual radioactivity?
• Could RSM recycling disproportionately affect disadvantagcd or under-represented
groups in society? Depending on the potential effect of RSM recycling on the
structure of the metal markets (e.g.. development of separate markets for metals
produced with and without RSM, or industry grading standards), could certain
groups (e.g., minorities and the poor) be disproportionately exposed to greater
levels of radiation?
2. Types and Amounts of RSM Available for Recycling
To answer the questions presented above. EPA must first evaluate 1) the types of RSM
(e.g., lead, steel, copper etc.) and the quantities of RSM that could potentially be recycled and
2) the amount of radiation contained in the RSM before and after decontamination.
In a recent draft report, A nalysis of the Potential Recycling of Department ofEnei-gy
Radioactive Scrap Metal, September 6. 1994, EPA estimated the types and quantities of RSM from
DOE facilities potentially available for recycling. The draft report found about 154 thousand tons
of carbon steel, stainless steel, aluminum, lead, copper, nickel, and other scrap metals from DOE
facilities is currently available; while approximately 1.1 tons is expected to be generated through
DOE D&D activities. DOE is only one part of the RSM picture, however. EPA expects that
decommissioning and site cleanups initiated by EPA under the Superfund program, the Department
of Defense (DoD), and the Nuclear Regulatory Commission (NRC) and its licensees will generate
substantial quantities of RSM. EPA is conducting research and analyses to estimate the types and
quantities of RSM that may be generated by these activities. Note that as a preliminary estimate,
NRC has said that as much as 1.7 million tons of carbon and stainless steel scrap metal could be
available from D&D of 118 nuclear reactors licensed by the NRC and its "Agreement States"
(NRC 94).
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3. Recycling Options
It is too early in the evaluation process to provide more than a general definition of the
three options that EPA is considering for recycling. They are:
« Restricted recycling. Under this option, RSM would be decontaminated, melted,
and re-cast for reuse in the nuclear industry (e.g., as containers for radioactive
waste disposal or shielding blocks for accelerators). The RSM would be melted
and re-cast at facilities dedicated to RSM recycling.
• Semi-restricted recycling (may require some degree of regulation). RSM could be
decontaminated, melted, and re-cast for reuse in the nuclear industry and certain
controlled uses outside the nuclear industry (e.g., for building bridges).
Mechanisms would be needed to track metal products produced with RSM to make
sure that they are applied to appropriate controlled uses. If the quantity of RSM
available for recycling is sufficiently large, standards to grade metals by level of
radiation or separate markets for metals produced with and without RSM could
develop.
« Unrestricted recycling. Once decontaminated to prescribed levels, RSM could be
sold on the scrap metal market for any use. There would be no formal mechanism
to track RSM to its use in finished metal products. Even so. metal grading
standards or separate markets for metals produced with and without RSM could
develop if radio-sensitive industries or industries that cater to sensitive populations
become wary of using metals produced with RSM.
EPA will be able to provide more detailed definitions of the recycling options when more
information is available on the amount of RSM potentially available for recycling, the potential for
increased radiological exposure, the structure of the metal markets, and the likely effect of
recycling on the metals industry.
4. Summary of Approach to Evaluate the Feasibility of RSM Recycling
In Section 1, several questions were posed that need to be answered to determine if the
LLW rule should include a recycling provision. These questions are repeated below, and our
approach for answering these questions is summarized.
Compared to disposal, is RSM recycling cost-effective?
Advocates of RSM recycling maintain that the federal government would save money by
recycling RSM because it would avoid the cost of disposing of RSM in LLW disposal facilities
and at the same time be able to generate revenue by selling RSM on the scrap metal market. Note
that for some RSM, decontamination costs may be higher than the market value of scrap metal.
Even so, if the total cost of recycling is less than the alternative—LLW disposal—recycling would
be cost-effective.
To assess whether recycling is likely to be cost-effective, EPA will estimate the costs of
recycling RSM, taking into account its potential commodity value, under alternative options (i.e..
restricted, semi-restricted, and unrestricted recycling), and compare them to the cost of disposing of
RSM in LLW disposal facilities.
RSM recycling costs depend on several factors. Some of the categories of costs that will
consider are summarized below.
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* Screening, cutting, and sorting of metals. Some metals may be amenable to
recycling while others may not. The metal will need to be separated between
recyclable and non-recyclable metals. Recyclable metals will need to be cut for
shipment and sorted by type.
* Decontaminating. Decontamination costs depend on the type, condition, size, and
shape of the metal to be decontaminated; the quantity of metal to be
decontaminated; the decontamination process used; and the level of
decontamination (EPA 94), (Note that the level of decontamination may vary
according to recycling option—a higher level of residual radioactivity in RSM may
be acceptable for restricted use than for unrestricted use).
• Transportation. RSM would also need to be transported to decontamination sites.
Therefore, the cost of packaging and transporting will be included as part of the
cost of decontamination.
* Verification. Before RSM is released, it will be necessary to verify that the metal
falls below a specified level of radhtion. These radiation levels may vary by final
use {e.g., under the restricted and semi-restricted recycle options, different radiation
levels may be specified for different uses).
* Tracking RSM. Under the restricted and semi-restricted recycling options, RSM
may need to be tracked to its final destination. Depending on the final use, the
tracking procedure could be very costly.
If RSM is not recycled, it will eventually be disposed of in LLW disposal facilities.
Therefore, to determine the potential cost-effectiveness of RSM recycling, the RSM disposal cost
must be evaluated. Estimating LLW disposal costs will be difficult and controversial because
1) currently available LLW disposal capacity is limited—therefore applying current commercial
rates may not be appropriate, and 2) recent problems experienced when attempting to open new
LLW disposal facilities provide evidence that disposal costs in the future may be much higher than
they are today. In estimating the cost of LLW disposal in the future, EPA will attempt to address
these issues. EPA will also assess other relevant categories of cost (e.g., the cost of loading and
transporting RSM to a LLW disposal facility, and the cost of storing RSM until LLW disposal
capacity is available).
How will RSM recycling affect the environment and the health and safety of the public and
woitere?
Both disposing and recycling of RSM has adverse affects on public health, worker health
and safety, and the environment. This is because both options expose the general public, workers,
and the environment to radiation and potentially harmful chemicals and increase risks to workers
from industrial accidents. Environmental impacts associated with disposal of radioactive materials
come from creation of new disposal sites or expansion of existing sites. In many cases, potential
RSM may be classified as mixed waste and thus will require disposal in specially permitted sites
(only one of which currently exists, in Utah) (NRC 94). Because sufficient disposal capacity does
not now exist, large quantities of material may need to be stored for extensive periods of
time—either where they are produced or at other facilities (MUR).
Because the metal industry uses both virgin metal ore and scrap metal as raw material
inputs, some argue that introduction of RSM into the scrap metal market will supplant demand for
virgin metal ore. This of course would mean that some of the environmental degradation associaed
with virgin metal ore mining and refining would be avoided. For example, mining and refining
results in the generation of large quantities of mine overburden, tailing, and slag (some of which is
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classified as Naturally Occurring Radioactive Material [NORM]), increased use of scarce water
resources, degradation of natural resources, and threats to endangered species (HER 93). It is not
clear, however, that RSM recycling would have a measurable impact on the mining of virgin metal
ore. This is because:
* In relation to total scrap metal demand and demand for virgin metal ore, the
potential amount of RSM available may not be large enough to have a discernable
impact on the virgin metal ore market.
* RSM recycling may simply supplant demand for non-radioactive scrap metal, and
have no discernable impact on the demand for virgin metal ore.
* Metal-ore mining may be a highly inelastic industry; consequently, lower scrap
metal prices due to the introduction of RSM may not affect metal ore mining.
Recycling will also adversely affect the environment. It can produce concentrated waste
streams of radioactive and RCRA materials which will require treatment and disposal (LIL 92).
Some radiation which had been confined to contaminated materials could be released into the air or
to water during recycling operations.
A number of exposure scenarios will have to be evaluated to determine risks to the public,
workers, and the environment from disposal and recycling. The exposure scenarios should reflect
the entire life cycle of each option from recover)' of the contaminated material to either final
disposal or re-entry into commerce. Secondary impacts sucii as increased energy usage or mining
activity should also be considered. Risks to be considered should include:
* Worker exposure from handling material at radioaclively contaminated sites;
» Worker exposure from transportation, storage, treatment, and disposal:
* Worker exposure from recycling operations and from the manufacture and use of
RSM products;
« Nonradiological risks to workers from increased industrial activity;
* Public exposure from transportation, storage, disposal, treatment, recycling, and use
of RSM;
• Environmental impacts from treatment, storage, disposal, and recycling facilities.
At recycling facilities special attention should be given to production of
concentrated radioactive and mixed wastes and to transfer of contamination from
metal to water and air;
* Public, worker, and environmental impacts from increases or decreases in mining
activity brought on or prevented by recycling. Impacts of changes in energy use
should be considered.
Will RSM recycling change the structure: of the metal industry and other related industries?
In Section 1, economic impacts in the metal industry and related industries that need to be
considered were organized into three categories:
* Market differentiation.
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* Price and demand effects, and
• International considerations.
Results of our analysis under the first category—market differentiation (i.e., will introduction of
RSM into the metal markets cause industry grading standards or separate markets for metals
produced with and without RSM to develop)—will largely drive the analyses and results in the
other two categories. For example, if it is determined that the introduction of RSM into a scrap
metal market (e.g., scrap stainless steel) will not adversely affect radio-sensitive industries or
industries that cater to sensitive populations, then the likelihood of market differentiation is small,
and RSM recycling will probably have positive economic and social impacts (e.g., lower scrap
metal prices, and perhaps less demand for virgin metal ore). If, however, market differentiation is
likely to result, adverse economic and societal impacts are more likel)' to occur. For example,
sensitive industries may incur higher costs to screen the metals they purchase for acceptable levels
of radiation, or they may pay higher prices for metals produced exclusively with virgin ore.
Maiket Differentiation
Because market differentiation is such an important aspect of the analysis, we will assess,
for each relevant metal market (e.g., stainless steel, carbon steel, aluminum, copper, lead, nickel
etc.): 1) the potential for market differentiation, and 2) industry and market impacts assuming
differentiation does and does not occur.
It is important to note that the metal production industry is not homogeneous. Each type of
metal (e.g., carbon steel, stainless steel, copper, nickel, lead) has its own market, and some markets
are large (e.g., stainless steel) while others are small (e.g., nickel). Therefore, it is possible that
RSM recycling will affect the individual metal markets in different ways. EPA will assess the
likely impacts in each affected market.
To evaluate the likelihood of market differentiation, EPA must assess the potential increase
in the level or variability of residual radioactivity in metals used by radio-sensitive industries and
industries that cater to sensitive populations. Next. EPA must determine the tolerance of radio-
sensitive industries or industries that cater to sensitive populations to increases in level or
variability of residual radioactivity from current levels. If the increase in the levels or variability of
residual radioactivity in metals due to RSM recycling is expected to exceed the level of tolerance,
then it is likely that grading standards or separate markets for metals produced without RSM will
develop (i.e., market differentiation will occur).
Price and Demand Effects
As summarized above, the likely price and demand effects from recycling will depend
largely on whether markets become differentiated. Assuming differentiation does not occur, the
price and demand for virgin metal ores and scrap metal might, in principle, be lower due to RSM
recycling. This is because the addition of RSM to the scrap metal market would cause scrap metal
prices to decrease and thus the use of scrap metal in metal production would increase. Therefore,
more scrap metal and less virgin metal ore would be used as a result of RSM recycling. To
estimate how much less virgin metal ore might be used, EPA will estimate how much more scrap
metal will be used due to RSM recycling. To do this, EPA must 1) estimate how much RSM
would be available at different market prices, 2) the potential decline in the price for RSM, and
3) the potential increase in the use of scrap metal.
As discussed earlier, however, virgin metal ore mining might not be significantly affected,
because tn relation to the total demand for virgin metal ore, the amount of RSM available may be
too small; RSM may simply supplant demand for non-radioactive scrap metal; and virgin metal ore
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mining, which may be a highly inelastic industry for varying reasons, might not respond to lower
scrap metal prices. Therefore, these factors will also be assessed and taken into account in our
consideration of the potential impact on virgin metal ore mining.
If a separate market for metal without RSM develops, it is possible the use of virgin metal
ores could increase. This is because, under the unrestricted recycling option, metal producers may
not know if they are purchasing RSM or scrap metal that is not radioactive. As a result, to satisfy
demand for metal produced without RSM. they may decide to produce metals entirely from virgin
ore. To estimate the potential increase in the use of virgin metal ore. EPA must estimate 1) the
demand for metals from radio-sensitive industries and industries that cater to sensitive populations,
and 2) the amount of scrap metal that would have been used in metal production to satisfy demand
from these industries before the introduction of RSM.
International Considerations
Many of the issues described above in a domestic context—impacts on public health,
effects on volumes and prices, the possibility of market differentiation and impacts on sensitive
user industries—apply to international metals markets as well. But cross-boundary trade in
recycled RSM would raise a number of additional concerns. As we have stated, given that for
some RSM, decontamination costs may be higher than the commodity value of the RSM. it is
unlikely that RSM would enter the recycled metals market unless the cost of disposing of RSM in
LLW disposal facilities is sufficiently high. In fact, if disposal costs are sufficiently high, the
federal government may be willing to sell RSM below going scrap metal prices. If the RSM is
sold at the going market price, and if the volume of RSM introduced is small relative to the normal
level of trade, no market distortions should result. That would likely be the case, for example, in
the steel market. But it is possible that for specialty metals, such as those used for high-
performance alloys or advanced technological applications, the amounts of metal released into the
market by recycling of RSM might be a significant fraction of annual trade. In such cases, RSM-
exporting countries would stand to gain an advantage in those markets, because their influence over
export volumes would in turn affect the price.
Another possibility is that importing countries would require all shipments of metal
containing RSM to be identified as such. If this practice becomes common, a two-tier market
might develop in which RSM-contaiuing metals trade separately—and presumably at a lower
price—than the virgin or non-RSM recycled commodity. Such a development would, in principle.
increase the efficiency of the market, because buyers would have better information about what
they are getting and would alter their bid price accordingly. To operate such a market, however,
would require the creation of an extensive monitoring and tracking system to ensure compliance.
It is not at all clear that the current state of radiation measurement technology would allow for the
detection of low levels of radiation on a routine basis at ports of entry.
A two-tiered market could have implications for the equitable distribution of environmental
impacts. It is possible that poorer countries would import a disproportionate share of (presumably
cheaper) RSM-containing metals. In the absence of a well-developed public health protection
infrastructure in such countries, it is conceivable that RSM could be reconcentrated or improperly
used and that an environmental health hazard could develop. This runs counter to the principle of
environmental equity, which holds that poor and minority populations should not be affected
disproportionately by either environmental health hazards or by the burden of cleaning them up.
Countries frequently impose trade restrictions in the interests of environmental and public
health protection. Indeed, the United Stales has included the authority to impose labor and
environmental standards in its draft bill implementing the Uruguay round of General Agreement on
Tariffs and Trade (GATT). Such restrictions may be perfectly legitimate, if a hazard can be
demonstrated; if not. they may unreasonably hinder the free flow of commodities. It is possible
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that such restrictions may hinder or prevent the development of an international market in recycled
RSM to begin with.
The cases just cited assume honest intent on the part of exporters. But it is conceivable
that nations facing huge disposal costs for radioactively contaminated materials might seek to solve
their problem—and gain some revenue—by selling their RSM on the international market at a price
that is below the going market price. For low-volume metals, such "dumping" could affect prices
and the structure of the market as a whole. Furthermore, it is unlikely that anti-dumping measures
currently in place—-such as those provided in GATT and other conventions—would curb, or even
detect, such activity. One reason is that the vast majority of RSM is controlled by governments
and is not produced by normal economic activity. Another is that many potential countries of
origin do not participate in the international bodies set up to curb these abuses. Indeed, it is
possible, in the absence of a tracking and monitoring system, that recycled RSM is already finding
its way into international commerce.
To evaluate the international impacts of alternative policies. EPA needs more information
on the following topics:
* The worldwide volumes of RSM that might be introduced into international
commerce.
* Radiation levels of metals already in international commerce, and how the
introduction of recycled RSM might affect them.
• The acceptability of RSM-containing metals in international markets.
• The availability of technology for measuring radiation levels in metals on a routine.
large-scale basis.
* Projections, presumably using economic models, of the effects of recycled RSM on
prices, volumes, and the structure of various segments of the international metals
market.
Could RSM recycling disproportionately affect disadvantagcd or under-represented groups
in society?
As described above, separate markets for recycled scrap metal with and without RSM could
develop. Furthermore, the marketplace may value recycled material with full or partial RSM
content at a lower price than recycled or virgin metals without RSM. Because disadvantaged or
under-represented groups in the U.S. are typically associated with lower income levels and reduced
economic status, these groups may be disproportionately exposed to products containing RSM, and,
therefore, incur health risks at a level disproportionate to the population as a whole. Such
increased exposures could occur knowingly, as lower income individuals may choose to purchase
RSM-containing products at reduced cost, or unknowingly, if RSM content is not labeled in lower
cost products,
EPA will need to evaluate the following factors to determine whether disadvantaged groups
may be disproportionately affected by RSM:
* The likely differential in price for metals with and without RSM content and the
products that could be produced using metals containing RSM.
« The purchasing patterns of disadvantaged and non-disadvantaged groups for the
RSM-containing products.
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• The likely decline in prices for products produced with metals containing RSM.
« The exposure patterns and health risks associated with the use of the RSM-
containing products.
5. Conclusion
As in a number of countries, the question of recycling of radioactive materials in the
United States is potentially very controversial. In attempting to develop a publicly acceptable
approach to the recycling of valuable materials, EPA must carefully evaluate both the risks posed
by these materials, and the benefits that could accrue to society through their recycle and re-use.
This paper has detailed the considerations that EPA deems relevant to an assessment of the
economic impacts of-a policy on recycling. As our discussion has indicated there are a wide
variety of potential consequences and issues that must be reviewed. EPA is committed to
completing this review as quickly as possible as part of our overall effort to develop a policy that
addresses the difficult and long avoided question of the disposal/reuse/recycling of materials
lightly contaminated or formerly contaminated with radioactive substances.
6. References
HER 93 Hertzler, T., et al, Science Applications International Corporation, Recycle of
DOE Radiologicallv ContaminatedMetal': A Scoping Study (Draft), February
1993.
1AE 93 International Atomic Energy Agency. Exemption from Regulatory Control:
Recommended Unconditional Clearance Levels for Solid Materials Incorporating
Radiomiclides, Vienna, March 1993.
L1L 92 Lilly, M.J., et al., U.S. Department of Energy. Radioactive Scrap Metal
Recycling: A DOE Assessment, presented at Waste Management '92, Tucson, AZ,
March 1992.
LIN 91 Linsey, G., International Atomic Energy Agency, Exemption Principles Applied to
the Recycling and Reu.se of Materials from Nuclear Facilities, August 1991.
MUR Murray. Raymond. Understanding Radioactive Waste, Columbus: Battelle,
PET 91 Peters. Dale and Konrad Kundig. Copper Development Association, Inc., Copper
and Copper A Hoys as Containers for Radioactive Waste Disposal, presented at
Waste Management '91. Tucson, AZ. February 1991.
EPA 93 U.S. Environmental Protection Agency. Issues Paper on Radiation Site Cleanup
Regulations, September 1993.
EPA 94 U.S. Environmental Protection Agency. Analysis of the Potential Recycling of
Department of Energy Radioactive Scrap Metal: Executive Summary (Draft),
September 6. 1994.
NRC 94 U.S. Nuclear Regulatory Commission. Generic Environmental Impact Statement
in Support of Rule-making on Radiological Criteria for Decommissioning of
NRC-Licensed Nuclear Facilities, Draft Report. NRC NUREG 1496 Vol II,
Appendix G, August 1994.
TES 93 Testimony Before the House Subcommittee on Energy, May 17. 1993.
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3-2 TECHNICAL ISSUES RELATING TO THE RECYCLE
OF CONTAMINATED SCRAP METAL
by
Stephen Warren, U.S.Department of Energy, Office of Environmental Restoration (EM-43),
Quince Orchard, 19901 Germantown Road, Germantown, MD 20784
Donald E. Clark, Westinghouse Hanford Co., 601 Williams Blvd., Suite 2A,
Richland, WA 99352
ABSTRACT
A review was made of the literature on melting of radioactive metals that was published
in the 1980s and 1990s with attention to the resultant partitioning of radioactivities.
Various factors influencing the transfer of radionuclides from the melted ingot phase to
other phases such as the slag layer need to be considered both in optimizing the
partitioning of radioactivities and in assessing the radiation exposures received by
workers. Important technical issues relating to the recycle of radioactive scrap metal
(RSM) have been identified and will be discussed in this presentation. Of particular
interest is the modeling of radiation doses resulting from operations involving RSM and
with recycled materials resulting from the melting of RSM, with emphasis on
radioactively contaminated ferrous metals. Such doses result from exposure to the
radioactively contaminated metals as well as secondary wastes (i.e., slag, dust, and
aerosols/filter media) produced as byproducts of the melting operations.
Summary and Conclusions
The reported results for partitioning of radioactivities that are achievable by melting of
contaminated ferrous metals have been reviewed. The resultant redistribution and
stabilization of radioactivities is important for possible recycling of these materials. For dose
calculations of the various recycling options, it is essential to have credible partitioning data
for each treatment scenario. Such data exist for only a few radionuciides (e.g., of the
elements uranium, plutonium, and cobalt); the need for new or additional data has been
identified.
There are a variety of contaminated metals some of which require treatment with different
slags and different melting processes. Control of thermochemistry of such processes is
essential. Optimal sizing of melting heats is also required for cost-effectiveness.
In general, most of the reported melting work was not done under controlled conditions and
no useful thermodynamic data were obtained. There is a need for such data on contaminants
in iron and various slags. At the very least, the partition ratios for key radionuclides in given
slags must be determined.
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Data on microscopic distribution and speciation of radionuclides are lacking. Also, the effects
of subsequent treatments (e.g., rolling, milling, welding) on radioactivity remaining in the
remelted metals have not been investigated.
Pretreatment methods applied to radioactively contaminated metals prior to melting should
include decontamination of surfaces, segregation according to base metal, segregation
according to radionuclide contaminant chemistry, and size reduction/compaction as required
for furnace loading.
Studies culminating in the design of an integrated process for feed pretreatment/preparation
and control of gaseous effluents should be initiated. These studies must include attention to
mechanical methods capable of operation within containment to assure adherence to the as-
low-as-reasonably-achievable (ALARA) principle of radiological safety.
The key issues to be addressed in implementing a recycle program of any magnitude for
radioactively contaminated metals include the following:
establishment of a credible database concerning materials to
be recycled, including chemical and radioactive characteristics
determination of radioactive partitioning between the metal and
slag phases
assured operability of the process, subject to widely varying feed
chemistry and conditions
demonstrated ability to seal the candidate process to prevent the
release of hazardous species
effective modeling of radiation exposures to workers throughout
the recycling process
An integrated program for recycling radioactively contaminated metals is being developed
which will focus resources and address these issues in the near future.
Introduction
In the United States (U.S.), very large quantities of radioactively contaminated metals have
been generated as by-products of nuclear weapon materials production and the associated
research and development activities at federal sites operated by the U.S. Department of
Energy (DOE). When no longer available or useful for intended purposes, or when they are
the result of decommissioning of facilities, such metals are referred to as radioactive scrap
metal (RSM). In addition to the RSM produced at DOE sites, significant quantities of RSM
have arisen or will be generated in the future from activities in the private sector including
nuclear power production. For economic and safety reasons, the recycling and reuse of RSM
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is now receiving serious consideration by U.S. nuclear managers in their planning for waste
management, facility and site decommissioning, and environmental remediation activities.
In the case of some radioactively contaminated metals, where only the contamination of
accessible surfaces has occurred, decontamination or removal of radioactivity by chemical or
mechanical methods is readily achievable. Within the restrictions of existing regulations, such
decontaminated materials may then be reused or recycled into other products. However,
where volumetric or persistent contamination and inaccessible surfaces are involved, melting
is recognized as a desirable option that can effect volume reduction; production of useful
product sizes, shapes, and volumes; homogenization of radioactivities; reduced radiation
exposures; and partitioning of radioactivities, including an effective removal of radionuclides
for certain RSM.
Over the past two decades or so, many studies and demonstration tests of the melting of
radioactively contaminated metals have been conducted and reported on, particularly in the U.
S. Germany, France, United Kingdom, and Japan. Worcester and co-workers (Worcester,
1993) reported that they had identified nearly 300 publications related to the application of
melting to RSM. A summary of large-scale melting programs for ferrous RSM is given in
Table 1. The present review is focused on the reported work with melting of ferrous metals,
and especially on the results obtained for partitioning of radioactivities.
TaMe 1, Sumimry of Large-Sate Melting Programs for Radioactively
Contaminated Ferrous Melals
Eeftrence/Country:
Oomer(1985yUK
Menon(19!>OySweden
Pealue(!992)/Ftance
Sappok (I992yOermany
Mfes(199iyGeniuny
Ttatia ( 1 990yGermany
Nakamora(J992yjipai
Mautz(l9?5VUSA
Uu£e,SEG
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JAERI-Conf 95-015
The radioactive contaminants include fission and activation products, transuranic and uranium
nuclides, and radioactive daughters. From a DOE perspective and considering the principal
facilities involved in environmental cleanup operations, important radionuclides include the
following elements; hydrogen, carbon, manganese, iron, cobalt, nickel, zinc, strontium,
cesium, technetium, cerium, zirconium, ruthenium, tellurium, uranium, plutonium, and
americium.
For considering the viability of melting RSM for recycling, the potential radiation doses that
would be received by workers and others exposed to the scrap, primary and secondary wastes,
and the final products (if radioactive) must be known. An early study of radiation exposures
resulting from recycle of smelted radioactive metals was reported by O'Donnell and co-
workers (O'Donnell, 1978). Their generic methodology provides a framework that is
applicable with modification to the melting of RSM. The dose calculations needed for
recycling decision making will be accomplished through computer modeling and use of
reported partitioning ratios for the radionuclides of interest. Radiological safety concerns
could then be addressed through development of appropriate specifications for the recycled
RSM. This review is intended to assess the present state of understanding and to identify
data needs that should be addressed to evaluate the recycling of RSM for containers,
Melting of Radioactive Scrap Metal (RSM)
The melting of RSM for recycling into containers involves heating the metal to a molten state
and permitting phase separation to occur. Depending on the thermochemical conditions,
during the melting process, radioactivities will be redistributed in the melt, the insoluble
(oxide) slag, the ceramic lining, dust, and vapor phase, as well as on other accessible
surfaces. On cooling and separation from the other phases, the metal ingot is available for
manufacture of the recycled metal products (e.g., a waste container). The ingot will contain
lower concentrations (if any) of the initially present radionuclides since they would be
homogeneously distributed throughout the metallic mass and would serve to attenuate any
remaining radiation. The slag is the residue of the smelting procedure (typically, a few
weight-percent of the total) and may contain larger or smaller concentrations of the nuclides
than those present initially in the RSM, depending on the ability of the slagging process to
remove small quantities of the radioactive contamination. The slag residue is characterized
by having a homogeneously distributed radioactivity and attenuated radiation fields. Other
radioactively contaminated products of the smelting would include the dust generated by the
process, the refractory lining and other surfaces associated with the smelter, vapors (e.g.,
tritiated water), aerosols, scrubber liquors, and filter media. These materials, which typically
constitute only 1-2 weight-percent of the total, would have to be disposed of as secondary
wastes.
Several excellent reviews of melting of radioactive metals have been published (Worcester,
1993; Mautz, 1975; Reimann, 1991). A number of laboratory scale and large scale melt
consolidation programs for RSM have been conducted over the past thirty years or so. A
large degree of success has been achieved by the melting of radioactively contaminated
ferrous metals for reducing concentrations of lanthanides, actinides, and most other fission
products that are easily oxidized. On the other hand, melting to remove troublesome
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transition elements from stainless steels, such as cobalt and technetium, has met with little
success.
The technological community has selected two technologies for melting of ferrous RSM,
coreless induction and electric arc furnace. (EAF) melting. There are advantages and
disadvantages for these technologies both of which are widely used in the steelmaking
industry.
The induction melting technique is accounting for increasing tonnages of steel in the U.S. as
improvements in solid-state frequency converters have increased operating efficiencies of the
smaller units and as environmental standards have become more stringent. A coreless
induction furnace contains a crucible or refractory lining surrounded by a water-cooled power
coil through which electrical energy is applied. Very rapid heating and high melting rates are
attainable with this design. Also, the induction currents stir the bath vigorously, which
assures more uniform composition and temperature. Furnace capacities may range from a
few pounds to 75 tons or more, with higher frequencies being required for the smaller
furnaces. The main factors favoring coreless induction are the following: it offers better melt
agitation; it offers easier fume control; and it allows rapid heatup. Also, an induction furnace
reportedly produces only 20 percent as much effluent dust as an EAF of similar capacity
(Reimann, 1991). This reduction of effluent dust can be a major factor in furnace selection
for the melting of RSM.
Most of the stainless steel in the U.S. is produced by the three-phase direct EAF process.
The EAF process also accounts for about one-third of the total U.S. raw steel production.
The geometry of the furnace is such that it contains a large diameter, shallow melt with a
large surface area which facilitates charging of scrap and the evolution of gases. The large
surface area also contributes to increased bath oxidation. Larger arc furnaces have internal
diameters of 30 feet or more and capacities of over 350 tons. Advantages of the EAF process
include the following: it provides lower cost as heat sizes increase; it accommodates larger
scrap section sizes; it allows for easier modification of melt composition; and it provides a
greater margin of reliability and safety because of the absence of the water-cooled induction
coil.
A modified melting technology, electroslag refining, has also been used to melt uranium and
plutoniurn contaminated RSM (Ochiai, 1983; Uda, 1987). In this method, contaminated metal
generally is melted in a molten slag in a mold by Joule heat generated with large amounts of
electrical current and then gradually solidified in the same mold. For selected applications,
success in partitioning of radioactivity between the metal and slag has been reported for this
technology.
Successful melting of RSM requires fundamental understanding of the metallurgical and
thermodynamic phenomena that undergird the process. Each type of metal or alloy requires a
different treatment. Also, the crucible or refractory lining and slag composition, as well as
the melting and pouring practice, must be specified. The slag chemistry selected for
preferential removal of contaminants is achieved by small additions of fluxing agents,
whereas the viscosity of the slag at operating temperature is controlled by the use of
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surfactants and diluents. Deslagging may be performed either manually or automatically with
the aid of a flux trap during pouring; best results usually are obtained with the trap.
Many factors can influence the removal by slag of radionuclides from molten metal, including
chemical stabilities of the radionuclides and their compounds, rates of compound formation,
ease of entrainment of impurity compounds in the slag, and transportability of the compounds
to the slag. An excellent discussion of the thermodynamic basis for removal of contaminants
by melting has been given by Copeland and co-workers (Copeland, 1978). The Gibbs free
energies of formation can be used as a guide for predicting removal of radionuclides during
smelting. Figures 1 and 2 show the free energies of formation as functions of temperature for
the oxides of groups VII and Vffl and for Group V and the actinides, respectively. Elements
for which the free energy of formation of the oxide is more negative than that for iron (c.f., -
57 kcal/gram-atom of oxygen for iron oxide at 300 degrees Kelvin) can be separated from
steel by oxidation and partitioning into the slag layer under conditions of melt refining. Thus,
smelting is effective in removing uranium from iron (as shown in Figures 1 and 2, the free
energies of formation of the oxides of uranium and iron range from -124 and -57 kcal/gram-
atom of oxygen, respectively).
Slagging agents (e.g., silica sand, lime, and magnesia) are usually added to the scrap metal
during melting. The weight of added ingredients is typically about ten percent of the charge
(scrap) weight, although lesser quantities may be used. Thus, if a radionuclide is removed
completely from the scrap metal and is entrained in the slag, the concentration of the
radionuclide in slag will be at least ten times the concentration in the original metal. The
use of slags and experience with the various chemical and metallurgical factors affecting the
partitioning of impurities between molten metals and slags has been reviewed by G, A,
Reimann (Reimann, 1991).
Slags are classified as "acid" or "basic" according to the silicate degree or the basicity,
depending on the slag's ability to react with the refractories as well as its ability to refine the
underlying metal by selectively removing impurities. Slags high in silica are acid slags, while
those high in metal oxides (e.g., CaO) are basic slags. These terms are rooted in observations
of similarities in the behavior of aqueous solutions and oxide melts. In chemical terminology,
an acid is generally considered to be any species that may accept an electron pair, while a
base is any species that may donate an electron pair.
The effect of slag basicity on the residual uranium concentration of a mild steel ingot while
other conditions were held constant (melt temperature of 1650°C; 30 minutes of melting; and
a sample contamination level of 500 ppm) is shown in Figure 3. These results indicated a
maximum decontamination effect around a basicity of 1.5 and were interpreted in terms of
ionic interactions at the interface between the molten metal and the slag (Abe, 1985).
Published Results for Partitioning of Radionucljdejjjy Melting of
Radioactively Contaminated Ferrous Metals
Over the past couple of decades, a considerable amount of work has been reported that could
be applicable to campaigns for the partitioning of radionuclides by the melting of RSM.
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c c
o <"
— C1
O X
g o
, o
en t
si
tu O
«- O
O 300 5OO
1000 !50O 2000
Temperature (*K]
2500
Figure 1. Free Energies of Formation for Oxides or
Elements of Groups VII and VIE (Adapted from Copeland, 1978).
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c c
o s>
g O
O ^_
IL. o
-130
-150
ThO
•ThO2
Oxides of Elements
of Group 3t ond
of the Actlnides.
0 300 50O
1000 1500 2000
Temperature (*K)
25OO
Figure 2. Free Energies of Formation for Oxides or
Elements of Groups V and of the Activities (Adapted from Copeland, 1978).
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JAERI-Conf 95^015
E
CL
c
o
Q
f 10"
o
c
O)
10
I I I
FLUX
O SiO2— CaO— Alp3
SiOg— CaO— Ai203— CaF2
D Si
0
\ X
b'
D
0.5 1.0
1,5
2.0
2.5 3.0
Slag Basicity (-}
Figure 3. Effect of slag basicity (CaO+Nio or CaF2/SiO2+A!2C>3) on uranium
concentration in a mild steel ignot (Abe, 1985).
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Pertinent radionuclides for which results have been published include tritium, cobalt, cesium,
uranium, and plutonium,
One problem that is encountered when reviewing the RSM melting literature is failure of
authors to provide sufficient details on testing conditions or to use consistent terminology
when reporting experimental results. Some authors report partitioning results in terms of a
decontamination factor (DF) which is defined as "amount" (mass) of impurity in the slag
divided by "amount" (mass) of impurity in the metal ingot at equilibrium, as shown for
impurity i in (1) below:
This is
(1) DFj = masSj in slag/mass; in metal ingot
also commonly referred to as the partition coefficient, lambda.
Other authors report the reduction factor (RF) which is defined as the initial concentration of
impurity i in the metal to its final concentration in the metal, as shown in (2) below:
(2) RFj = initial concentration; in metal/final concentration; in metal
A scientifically acceptable set of terms should be used by all workers in the field so that
results could be readily interpreted in the context of underlying merrnodynamic/kinetic
descriptions.
A discussion of the theoretical basis for partitioning of impurities between slag and molten
metal was given by Copeland and co-workers (Copeland, 1978). Based on thermodynamic
data and using several assumptions, they calculated partition ratios for several elements in
iron melts at 1530°C. Their results are given for four elements of interest in Table 2. Due to
uncertainties in the calculations, these results can be taken only as order-of-magnitude
numbers but the relative values are significant. The results predict that uranium and
plutonium impurities should be readily separable from iron and steels by melting with slag,
while cobalt and technetium impurities would tend to remain in the metal ingot.
Table 2. Partition Ratios for Oxide Slags for Several
Contaminant in Iron Melts at ISSO'c1
Uranium
Plutonium
Cobalt
Technetium
10"
I08
10"5
10
r8
'Partition ratio is defined as mass of contaminant in the slag divided
by the mass of contaminant in melt at equilibrium. Results of
calculations by Copeland and co-workers (Copeland. 1978).
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The expected behavior of tritium during melting of ferrous metals is to escape as an aerosol.
Tritium may be present in either the oxide layer or be incorporated into the metallic lattice of
contaminated metals. On melting stainless steel containing tritium, it was observed that
tritiated water was quantitatively transferred to liquid nitrogen traps (Shenker, 1986).
Cobalt-60 is a principal activity in steelwork from decommissioned nuclear power plants. Its
behavior in melting of radioactively contaminated steel was studied by several groups in the
first Commission of European Communities (CEC) program on decommissioning (CEC,
1986). The findings of several groups was that cobalt partitioned almost entirely to the steel,
which is consistent with general experience and the thermodynamic properties of cobalt in
steelmaking. However, some investigators reported that, in induction furnaces, 25 percent or
more of the eobalt-60 was present in the slag under some circumstances. Testing by British
Steel (Harvey, 1990) confirmed that eobalt-60 could be present in the slag initially but that its
concentration tended to decrease with time and the transient presence of eobalt-60 was likely
to be dependent on slag viscosity. For the viscous or semi-solid slag found in the induction
furnace, metallic globules might be retained until the end of the process. However, in the arc
furnace and basic oxygen steelmaking vessel, slags are fluid so that residence times of
metallic globules would be short and cobalt-60 should not normally be found in the slag.
The behavior of cobalt-60 in basic oxygen steelmaking was investigated in the United
Kingdom (Harvey, 1990). Steel plates containing cobalt-60 were included in the raw
materials and the oxygen blowing process was performed in the normal way. Cobalt-60 was
found to remain with the steel, demonstrating that even under powerfully oxidizing non-
equilibrium conditions the formation of cobalt oxide is not favored. Kinetic factors
apparently do not outweigh thermodynamic factors and this process could be used to melt
steel contaminated with cobalt-60 with confidence that the slag and fume would not become
secondary wastes.
The reported experiences with cobalt-60 emphasize that practical considerations are important
in the melting of ferrous RSM. The precise separation of slag from the steel is generally not
possible. Rather, the physical actions involved in pouring the steel from the furnace will
inevitably result in some mixing of the steel and the slag. And, therefore, the steel produced
from the process will contain some traces of slag, and the slag will contain some traces of the
steel. Because of these problems, there are practical limits to the retention of radioactivity in
the steel ingots.
Cesium radionuclides are another main contaminant in steel from decommissioning. Its
behavior in RSM melting is dependent on the particular conditions (Harvey, 1990). Cesium
can be partially absorbed by an acidic slag, but very little is absorbed by a basic slag. These
characteristics are evident in the arc furnace, but in the induction furnace where the slag is
not fully melted the results tend to be erratic.
Over 30,000 tons of uranium contaminated steel were melt-slag processed up to the mid-
1970s. Uranium concentration varied but was generally less than 10 parts per thousand.
Residual uranium concentrations were about 15-45 percent of the starting concentrations.
Approximately 2.5 kg of plutonium contaminated steel was melt-slag processed in small
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JAERI-Conf 95-015
batches (less than about 200 g per batch). Residual plutonium content varied but reduction
factors of up to 200 were achieved. Important details (e.g., slag chemistry, starting
contaminant concentration, furnace design, operating parameters) necessary to assure
reproducibility and to establish process capability limits are generally lacking in the
referenced documentation. Using various slags, reduction factors of uranium in ferrous
metals of 1,000 or more have been reported (Uda, 1982; Heshmatpour, 198la), and similar
results have been reported for plutonium in ferrous metals (198 Ib).
Based on our review of the literature, it appears that melting conditions can be established to
effect the separation of H-3; C-14; Sr-90; Cs-134,137; Zn-65; U-235,238; and Pu-239.
Radionuclides expected to remain in the melt include Mn-54; Ni-63; Co-60; and Fe-55. In
general, major contaminants can be separated into four groups (Bechtold, 1993): 1) elements
that remain in the melt, 2) elements that form high-temperature intermetallics in metal and
slag (e.g., Ce-144), 3) elements that oxidize and partition to the slag (e.g., Sr-90; U-235,238),
and 4) elements that vaporize and report to the off-gas system (e.g., Cs-134,137; Zn-65).
Partitioning and Other Data Needed for Deciding Recycling Options for Radioactively
Contaminated Ferrous Metals
Our review of the literature on this subject has led to the conclusion that certain gaps remain
concerning our knowledge of partitioning of radioactivity when melting ferrous RSM. The
existing data base on RSM at DOE sites, including projected values for decommissioning of
facilities and the environmental remediation of sites, is incomplete. The present effort should
be supplemented by a more comprehensive review based on revisions in the DOE's data base
on RSM.
An understanding of the redistribution and stabilization of radioactivities in RSM and slags
during melting is important for the possible recycle of contaminated materials. For dose
calculations of the various recycling options, it is essential to have credible and applicable
partitioning data for each treatment scenario. Such data exits only in sketchy form and only
for a few radionuclides (e.g., of the elements uranium, plutonium, and cobalt). There is need
for additional data on most of the key radionuclides in DOE's inventory of RSM, including
data on partitioning, speciation, and applicable chemistries, and such data need to be obtained
under controlled conditions so that thermodynamic/kinetic predictions can be made. The
acquisition of thermodynamic data will permit optimization of the candidate processes.
In a recycling program of national scale, there will be a large variety of contaminated metals
some of which will require treatment with different slags and different melting processes. It
will be essential to control the thermochemistry of such processes and the effects of the
chemical makeup of feed materials are crucial in this regard. Also, the optimal sizing of
melting heats to minimize radionuclide characterization and analytical costs while casting the
material into suitable forms for recycling will be required in order for the program to be cost-
effective.
Data on the microscopic distribution and speciation of radionuclides in RSM are lacking.
Furthermore, the effects of subsequent treatments after melting (such as the rolling, milling,
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JAERI-Conf 95-015
and welding processes associated with manufacture of new items) on the radioactivities
remaining in the purified metals have not been studied at all,
RSM will require pretreatment, some of which could be extensive, prior to melting. These
may include chemical/physical decontamination of surfaces, segregation according to base
metal, segregation according to radionuclide contaminant chemistry and other properties, and
size reduction/compaction and sorting as required for furnace loading.
Studies culminating in the design of an integrated process for feed pretreatment/preparation
and control of gaseous effluents should be initiated. These studies must include attention to
mechanical methods capable of operation within containment to assure that there is adherence
to the as-low-as-reasonably-achievable (ALARA) principle of radiological and worker safety.
The key issues to be addressed in implementing a recycle program of any magnitude RSM
include the following:
establishment of a credible database on DOE's inventory of RSM
(current and projected), including chemical, physical, and
radioactive characteristics
determination of radioactive partitioning between the metal and
slag phases under controlled conditions
operability of the candidate process, subject to widely varying
feed chemistry and conditions and environmental protection
requirements
effective modeling of radiation exposures to workers and others
throughout the recycling process
While many reports on the melting of RSM and subsequent partitioning of radioactivities
have been published, most of the reported work was not done under controlled conditions and
no useful thermodynamic data were obtained. These is a need for such data on the expected
contaminants in ferrous metals and various slags which would be used in melting campaigns.
At the very least, the partition ratios for key radionuclides in given slags must be determined.
A credible and comprehensive data set will be required if a large-scale recycling program for
ferrous RSM is to be implemented.
Along with the needed thermochemistry studies, it would be desirable to construct a pilot
facility equipped with the appropriate gas cleaning and handling system to demonstrate the
candidate processes. The system of choice would probably consist of an inductive heating
furnace equipped with a gas tight system.
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References
M. Abe, T, Uba, and H. Iba, "A Melt Refining Method for Uranium Contaminated Steels and Copper," pp. 375-
379 in Volume 3 of the Proceedings of the Symposium on Waste Management at Tucson. Arizona. March 24-28.
1985. Roy G. Post, Editor, 1985,
T. E. Bechtold (Editor), Winco Metal Recycle Annual Report FY 1993. Westinghouse Idaho Nuclear Company,
Inc., Idaho Falls, Idaho, December 1993.
CEC Report, "Melting and Recycling of Radioactive Metals from Decommissioning of Nuclear Installations,"
Proceedings of a Commission of European Communities Workshop. Karlsruhe. Federal Republic of Germany.
May 27-28. 1986 [referenced by Harvey, 1990}, 1986.
G. L. Copeland, R, L. Heestand, and R. S. Mateer, "Volume Reduction of Low-Level Contaminated Metal Waste
by Melting - Selection of Method and Conceptual Plan," ORNL/TM-6388, Oak Ridge National Laboratory, Oak
Ridge, Tennessee, 1978.
R. Echols, SEG, Oak Ridge, Tennessee [referenced by Worcester (1993)], 1993.
C. R. Gamer and J. T. Lambley, "Melting of Contaminated Steel Scrap Arising in the Dismantling of Nuclear
Power Plants," EUR-10188, Commission of the European Communities, 1985.
D. S. Harvey, "Melting of Contaminated Steel Scrap from Decommissioning," pp. 473-481 ia Decommissioning
of Nuclear Installations, Proceedings of the 1989 International Conference. Brussels. Belgium, October 24-27.
19.89. Elsevier Applied Science, Barking, United Kingdom, 1990.
B. Heshmatpour and G. Copeland, "The Effects of Slag Composition and Process Variables on Decontamination
of Metallic Wastes by Melt Refining," ORNL/TM-7501, Oak Ridge National Laboratory, Oak Ridge, Tennessee,
January 1981a.
B. Heshmatpour, G. Copeland, and R. Heestand, "Decontamination of Transuranic Waste Metal by Melt
Refining," ORNL/TM-7951, Oak Ridge National Laboratory, Oak Ridge, Tennessee, December 1981 b.
D. Large, SEG, Oak Ridge, Tennessee [referenced by Worcester (1993)], 1993.
M, M. Larsen, J. N. Davis, and J, A. Logan, "Sizing and Melting Development Activities Using Contaminated
Metal at the Waste Experimental Reduction Facility," EGG-2411, Idaho National Engineering Laboratory, Idaho
Falls, Idaho, February 1985.
E. W. Mautz, G. G. Briggs, W. E. Shaw, and J. H. Cavendish, "Uranium Decontamination of Common Metals
by Smelting: A Review (Handbook)," NSA3204, National Lead Company of Ohio, Cincinnati, Ohio, February
1975.
S. Menon, G. Hernborg, and L. Anderson, "Meiting of Low-Level Contaminated Steels," Pecommissioning
Nuclear Installations - 1990. Studsvik AB, Sweden, 1990.
H. P. Mies and W. Stang, "Decommissioning of Nuclear Power Plant Gundremmingen Unit A," Kernkraftwerke
Gundremmingen Betriebsgesellschaft, Germany, 1991.
H. Nakamura and K. Fujiki, "Radioactive Metal Melting Test at Japan Atomic Energy Research Institute," Japan
Atomic Energy Research Institute, Tokai-mura, Naga-kun, Ibaraki-ken, Japan, 1992.
A. Ochiai, K. Kitagawa, Y. Sawada, S. Izuhara, and K. Ohtsuka, "Treatment of Plutonium-Contaminated Metallic
Waste by the Electro-Slag Melting Method," pp. 177-190 in Conditioning of Radioactive Wastes for Storage and
-148-
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Disposal. Proceedings of an International..Symposium.Utrec.ht. Netherlands. June 21-25..J982. IAEA-SM-261/20,
1983.
F. R. O'Donnell, S, J. Cotter, D, C. Kocher, E. L. Etnier, and A. P. Watson, "Potential Radiation Dose to Man
from Recycle of Metals Reclaimed from a Decommissioned Nuclear Power Plant," ORNL/NUREG/TM-215, Oak
Ridge National Laboratory, Oak Ridge, Tennessee, December 1978.
J. Peulve, "Treatment of Dismantled Materials by Fusion," Techno-992-022 [referenced by Worcester (1993)],
1992.
G. A. Reimann, "Technical Assessment of Processes to Enable Recycling of Low-Level Contaminated Metal
Waste," EGG-MS-9S79, Idaho National Engineering Laboratory, Idaho Falls, Idaho, October 1991.
E. Schenker, W, Francioni, and G. Ullrich, "Laboratory Scale Melting of Contaminated Metallic Scrap,"
Proceedings of Commission of European Communities Workshop on Melting and Recycling of .Radioactive
Metals from Decommissioning of Nuclear Installations. Mav 27-28. 1986. Karlsruhe, Federal Republic of
Germany, 1986.
T. Uda, Y. Ozawa, and H. Iba, "Melting of Uranium-Contaminated Metal Cylinders by Electroslag Refining," pp.
328-337 in Nuclear Technology. Volume 79, December 1987.
S. A. Worcester, L. G. Twidwell, D. J. Paolini, and T. A. Weldon,"Decontamination of Metals by Melt
Refining/Slagging - Art Annotated Bibliography, W1NCO-1138, Westinghouse Idaho Nuclear Company, Inc.,
Idaho Falls, Idaho, July 1993.
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JAERI-Conf 95-015
3_3 The Prospect for Recycle of Radioactive Scrap Metals to Products for
Restricted and Unrestricted Use
Alan L. Liby
Manufacturing Sciences Corporation
Oak Ridge, Tennessee, USA
ABSTRACT
Large quantities of radioactive scrap metals will arise from decontamination and
decommissioning of nuclear power plants and DOE facilities. Much of this metal
can be easily decontaminated and released to the existing secondary metais
industry for recycling. For metal that can not be readily released, recycle into
restricted-use end products is an economically attractive alternative to burial as low
level radioactive waste.
This paper wilf examine sources and types of scrap metal, technical approaches,
potential products, and economics of metals recycle. Construction, licensing,
environmental compliance, and possible reuse of existing nuclear facilities for
metals recycling will be discussed.
INTRODUCTION
Three avenues of disposition of radioactive scrap metal (RSM) from
decontamination and decommissioning (D&D) of nuclear facilities are evident: 1)
release to unrestricted commercial use after first removing contaminants from the
surface or refining bulk contaminated metals; 2) consolidate by compaction or
melting then store or bury; and 3) recycle to useful new products that would be
restricted in end use. Decontamination and release can provide an economical
approach in cases where there is no internal contamination and surfaces can be
easily accessed. Unrestricted release of bulk-contaminated metals is presently not
an option in practice in the U.S. due to the lack of an accepted standard for
unrestricted release of these metals. Consolidation and storage or burial will
become increasingly less attractive as the cost of repository storage or burial
increases. Even if acceptance of a standard for release of internally contaminated
metals is adopted in the near term, expense and difficulty of decontamination and
quality assurance will make recycling of RSM into products with restricted end use a
key element in the strategy for D&D of contaminated sites and facilities.
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JAERI-Conf 95-015
SOURCES AND TYPES OF RADIOACTIVE SCRAP METAL
RSM arises from a number of government and commercial sources as listed below:
• Government Sources
- Gaseous Diffusion Plants
- Production and Research Reactors
- Nuclear Laboratories
- Nuclear Weapons Manufacturing Facilities
- Military Test Ranges
- Nuclear Navy
» Commercial Sources
- Nuclear Power Plant Maintenance
- Nuclear Power Plant Decommissioning
- Nuclear Fuel Manufacturing Facilities
- NORM Contamination from Oil Drilling
Estimates of RSM expected to come from the decommissioning of U.S. Department
of Energy (DOE) facilities range from 1 to 2.5 million tons. Two estimates of DOE
RSM inventories have been documented. Whitfield, 1991, found in excess of
130,000 tons of RSM are in storage at various DOE sites and an ongoing
generation rate estimated at 15,000 tons/year that was expected to increase
dramatically to over 90,000 tons per year when major D&D projects begin. The
need to dispose of a total of 1.5 million tons of RSM was projected. Duda, 1993,
reported an inventory made from public documents and telephone survey of all DOE
sites and field offices. A total of 991,000 tons of potential RSM was identified,
including that from the survey shown by site in Table 1 and by metal type in Table 2
and projected arisings from D&D of the gaseous diffusion plants shown in Table 3,
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JAERI-Conf 95-015
Table 1. DOE RSM Inventory by Site
lSMfl^^Scinri^^;^^^
.:--.-..;.;.:.;,:,'.;-:.; ; ; :-:-;-:.;.v.;.v;.>:*^U |' J1 I,,,"' :•:•:•"•'•:•> •:•>. .;•;-•-• -Xs •:•:•:• •*
Fernald
Hanford
INEL
NTS
Oak Ridge Naf 1 Lab
K-25
Y-12
Paducah
Portsmouth
Rocky Flats
TOTAL
8,099
416
9,515
157,319
25,002
116,429
5,000
40,265
34,191
70
396,306
2.04%
0,10%
2.40%
39.69%
6.31%
29.38%
1.26%
10.16
8.63%
0.02%
100.0%
Table 2. DOE RSM Inventory by Metal Type
Aluminum
Brass
Copper
Lead
Monel
Nickel
Steel
Mixed
TOTAL
16,250
10
11,215
747
1,745
47,524
143,221
175,594
396,306
4.1%
<.01%
2.83%
0.19%
0.44%
11.99%
36.14%
44.31%
100%
Table 3. Arisings from D&D of Gaseous Diffusion Plants
lilifllSpIli
Aluminum
Coppet
Monel
Nickel
Steel
Mixed
TOTALS
!:i!lulNlTl; :,•;•:•; :•:-.'•.-.•••. ;-: •'••;•
.:"^:-^:::^':;>:;>.:j:•^l:^'-:^':V:-;v^Vx';•;^•;••;':;•:•:::;:::::•^^^::;:^X':":•'^•^::::: .:::x::-::-:::':'.-.-
iPilclflilllKlll
3.73%
7.45%
0.74%
9.7%
45.24%
33.14%
100.0%
152-
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JAERI-Conf 95-015
About 15,000 tons of RSM are generated annually from ongoing maintenance of
nuclear power plants in the U.S. based upon estimates from business activity in the
low-level waste processing industry. Decommissioning of power reactors and
systems will produce large quantities of RSM. Mizia and Wahnschaffe, 1994,
identified 89,670 tons of recyclable stainless steel from eventual decommissioning
of 122 reactors. Atteridge, 1993, found the total for all reusable metal from U.S.
reactor decommissioning to be about 760,000 tons. A summary of arisings from
known utility industry and DOE sources is shown in figure 1.
140000 j
120000 --
v, 100000 --
2 80000 --
1 60000 --
w 40000 --
20000
0
O'**-<»CSItf)OireOtM<0Oto>ooT-t-T-«McsicococO'j-'a»a>a>oooooooooooooo
Year
i NPP Operations m NPP Decommissioning • DOE Arisings
figure 1. RSM Arisings from DOE and Electric Utilities
PRODUCTS FROM RECYCLED METALS
The challenge ahead in recycling of contaminated metals from DOE D&D activities
lies in the resolution of regulatory standards for release of contaminated metals to
both restricted and unrestricted use. The reuse of scrap metals requires remelting
which evenly distributes in the newly cast metal any remaining contamination that is
not removed either by cleaning prior to melting or by removal during the melting
process. Metals that are contaminated only on the surface can often be
economically cleaned and released to the commercial secondary metals market.
While numerous studies have shown that melt refining is an effective method of
removal of some types of radioactive contamination from some metals (Worcester,
et al, 1993), the lack of a standard for release of volume-contaminated materials
makes use of this approach for recycling problematic for unrestricted end uses.
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JAERI-Gonf 95-0! 5
Even if the regulatory bodies could easily agree to standards for unrestricted reuse
of bulk contaminated metals, the most economic recycle of these metals will be
substantially aided by identification of products that may be made from these
metals. The process of melting and refining adds value to contaminated metals that
will be lost if the metals are then returned to the secondary metals market as if they
were scrap. Intermediate products such as billets or slabs for extrusion, forging, or
rolling by existing commercial metaiworking industries are perhaps the best
candidates for metals that may be released for unrestricted use. For surface
contaminated metals that cannot be economically decontaminated, and for metals
that are contaminated in the bulk, metals may be recycled only to restricted end
uses. Products such as containers for radioactive waste and materials of
construction for hew nuclear facilities where such materials will become
contaminated are good examples of end products with restricted use. The
beneficial reuse of thousands of tons of contaminated metals could be brought
about by a commitment on the part of governments and the nuclear industry to
make maximum use of previously contaminated metals in all applications in which
the metals will become contaminated.
Some candidate applications for restricted reuse of RSM in waste containers are
discussed below:
Spent Reactor Fuel. A multi-purpose canister (MPC) for dry storage and eventual
disposal of spent nuclear fuel will be developed starting in 1995. This concept will
require manufacture of 10,000 stainless steel canisters, each weighing about 21
tons, for containment of spent fuel from U.S. reactors.
High level waste. About 381,000 m^ of HLW are in storage at four DOE sites:
Hanford, Savannah River, INEL, and West Valley. Some of the HLW will be
concentrated to about 10 percent of its existing volume, vitrified using borosilicate
glass, and poured into stainless steel containers. The vitrification program plan
calls for construction of plants at Hanford, West Valley, and Savannah River. The
need for 7680 containers over 17 years is projected (DOE, 1991), The containers
are about 3 m long and 0.6 m in diameter and weigh 500 KG, The first 500
containers have been purchased for the Savannah River vitrification plant at a price
of $7000 each. Using a 25% waste concentration in glass and 0.8 nr»3 effective
volume of the container, this approach will only accommodate 4% of the existing
HLW, leading to the conclusion that nearly 200,000 such containers may be
required for DOE high-level waste.
TRU waste. TRU waste is defined as that containing more than 100 nanocuries per
gram (37 bq/g) of alpha-emitting transuranium radio nuclides with half-lives greater
than 20 years. DOE has about 37,000 m^ of such waste in storage and generates
additional waste at the rate of 2000 m^ per year (DOE, 1991). The Waste Isolation
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JAERI-Conf 95-015
Pilot Plant (W1PP) in New Mexico is designed to contain up to 170,000 m3 of TRU
waste, but its opening has been repeatedly delayed and temporary storage of TRU
waste at various DOE sites has been a crisis-level problem over the past several
years. Waste would be stored at WIPP in drums and boxes. Much of the existing
waste will have to be repackaged before shipment to meet the WIPP requirements.
Because some of the existing waste consists of acid and caustic residues, and
corrosion and leakage of drums is common, use of corrosion resistant alloys for
waste containers is under active consideration.
Low-Level Waste. The total amount of low level waste that will come from DOE
D&D activities is not known with any precision. The estimated volume of LLW
accumulated at the Fernald site, not including waste expected to be generated
during cleanup, is 230,000 m3 (DOE, 1991). The volume of burial of commercially-
generated LLW at the Barnwell, Beatty and Richland sites for 1991 was about
67,000 m3. The overall volume of LLW will continue to decrease as waste
reduction practices are refined.
A large quantity of carbon steel is used in packaging of low-level waste. Waste is
most often shipped to the LLW repositories in 100 ft3 boxes with a tare weight of
about 660 Ib. The contents of the boxes are super compacted waste that was
originally placed in drums or boxes for compaction. In the case of the compaction
boxes, each has a precompaction volume of 38.5 ft3 and a tare weight of about 220
Ib. Assuming compaction to 25% of original volume and 60% efficiency in packing
the large box, there would be 6 compacted boxes inside of the 100 ft3 box. The
steel usage would therefore be 1980 lb/100 ft3 of LLW disposal. The current usage
of steei for commercial LLW burial is thus about 21,000 tons per year. A like
quantity of usage would be expected for burial at DOE sites.
TECHNICAL APPROACH
Manufacturing Sciences Corporation is engaged in demonstration of stainless steel
RSM recycle, including melting, casting, roiling, and fabrication of boxes and drums
in its plant in Oak Ridge, Tennessee. The technologies chosen for demonstration
are directly relevant to the economical recycle of contaminated stainless steel on a
small scale, dedicated basis and will be used to establish design and operating
parameters for a commercially viable plant to recycle radioactively contaminated
metal (RSM). Some of the concepts represented by the chosen technologies that
are felt to be of fundamental importance with respect to successful recycle of RSM
into new products include:
Precleaning of RSM. RSM may require precleaning prior to melting,
depending upon the level of contamination from radio nuciides and other
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JAERI-Conf 95-015
foreign material. Reduction of the level of radioactive contamination prior to
melting can reduce the level of radioactivity in the final product and can
benefit the overall process efficiency by reducing personnel exposure levels
from the fabrication operations. It is possible that all operations subsequent
to melting could be done with very minimal radiation safety concerns if most
of the radio nuclide contamination is removed by precleaning combined with
separation in the melt. Removal of other foreign materials before melting will
greatly reduce the difficulties in maintaining control of the alloy chemistry of
the final product. While it is possible to adjust the chemistry of the melt
additions and reactions to achieve the desired alloy chemistry, MSC has
demonstrated that stainless steel may be remelted and to make new product
within the same specification if careful precleaning and melting practices are
followed.
Melting. The contaminated stainless steel will be melted in an enclosed, gas-
tight induction furnace under reduced pressure with an inert gas backfill.
Such melting practice has two important advantages: 1) the vacuum or inert
atmosphere prevents oxidation of the melt and thereby significantly reduces
the secondary radioactive waste generated during the melting process, and
2) melting in a gas-tight vessel allows positive containment of volatile radio-
nuclides such as tritium and cesium and provides the opportunity for
treatment of a very low volume of off-gas to remove volatile radioactive
constituents if found necessary.
The induction melting furnaces in use at MSC have successfully
demonstrated containment of radio nuclide and beryllium contamination while
melting depleted uranium, beryllium, and radioactively-contaminated scrap
metals. While the design of MSC's furnaces can be modified and optimized
for melting stainless steel for recycle, the concept of melting in an enclosed
vessel is fundamental to a clean operation with minimal personnel exposure
and low risk of uncontrolled release of radio nuclide contamination.
Roll ing. Demonstration of the fabrication of boxes and drums requires that
stainless steel sheet metal of 12 gage and 16 gage thickness be produced as
the starting material. In a modem high-volume commercial operation,
stainless steel sheet metal in this thickness range would probably be
produced as coiled hot-rolled product from a multiple-stand tandem mill and
would be further cold reduced by a second multiple-stand rolling operation or
on a reversing mill. Operations of this type produce large tonnage quantities
of sheet metal on a commodity basis. The investment required to install such
a rolling operation would not be supported by the low volume of RSM to be
recycled in the proposed plant. Furthermore, producing coiled material
followed by uncoiling and flattening prior to fabrication into final product
would serve practically no purpose in the contemplated operation.
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JAERI-Conf 95-015
MSC's approach to rolling for metals recycle has been to use a 4-high
reversing mill to produce plate or sheet to meet a specific product
requirement. In other words, the objective of the roiling operation is different
from commercial plate and sheet production in that the end product is not a
commodity to be shipped for use in a variety of undefined products, but
instead the plate and sheet is produced on a custom basis according to end
use. The typical approach is to produce discreet pieces that are tailored in
size to minimize waste and internal recycle. For the fabrication of boxes and
drums, pieces will be rolled to specific widths and gages to produce drum
body, box sides, ends, tops, bottoms or other components. In essence, the
rolling operation is treating in much the same manner as any other metal
forming operation in the fabrication sequence. This approach is less efficient
in terms of quantity of metal rolled, but the lack of efficiency is offset by low
capital investment and the appropriate match to the real job of recycle on a
low-volume basis.
Internal Recycle. A key issue in recycle of RSM is internal recycle of
trimmings and other scrap. The overall objective of the processing is
maximized process throughput, minimized secondary waste, and virtually
complete consumption of the RSM in useful product. Internal recycle of RSM
scrap and trimmings that arise from fabrication of boxes and drums will be
demonstrated within the proposed project
REGULATORY COMPLIANCE
Recycle of contaminated metals that can not be released to the commercial
secondary metals market must be done in licensed facilities with capabilities for
assuring compliance with a broad range of employee health and safety and
environmental issues. Such a facility is subject to regulation under applicable
federal, state and local environmental, health and safety regulations, including:
» Clean Water Act
» Clean Air Act
• Resource Conservation and Recovery Act
« Superfund Amendments and Reauthorization Act
• Comprehensive Environmental Response, Compensation, and Liability Act
* Toxic Substance Control Act
• Occupational Safety and Health Act
» Hazardous Materials Transportation Act
• Atomic Energy Act
» Low Level Radioactive Waste Policy Act
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JAERI-Conf 95-015
Handling of secondary waste streams is an extremely important issue from the
economic and regulatory perspective. All effluents, including air, liquid, and solid
will be monitored either continously or by batch before release. For a commercial
entity engaged in such business, failure to comply with all applicable regulations
can lead to immediate and total financial disaster, whereas an overly conservative
approach can lead to slow financial death.
ECONOMICS OF RECYCLE
First Law of Metal Recycle: "Economic viability of metal recycle increases in
proportion to intrinsic value of the recycled metal"
First Law of RSM Recycle: "Economic viability of RSM recycle increases in
proportion to cost of disposal"
The market potential for recycling of RSM is the sum of the cost avoidance of
disposal and the value of products that may be made from these metals. The
comparitive benefit of recyle for several metal types is shown if Table 4. The
combined effects of intrinsic value of the metal and the cost of disposal are readily
shown.
The current commercial price for disposal of radioactively contaminated scrap
metals, not including any value that might be recovered, is a minimum of $1.50 per
pound. The 1,5 million ton DOE RSM inventory therefore represents a $4+ billion
disposal liability. All other sources of RSM, the nuclear utility industry being the
largest among these, will create a disposal need of similar magnitude. When
viewed as a business opportunity, including products that might be manufactured for
restricted use, several businesses of moderate scale should be attacted to provide
RSM recycle services and products.
Table 4. Comparison of Recycle Benefit for Several Metals
Carbon Steel
Stainless Steel
Stainless Steel (Ni
Feed)
Nickel Alloy
||||;(§S|||||
:Commerciat
-iHSMetai
0.23
1.32
1.32
5.17
lihellWall
1.80
2.20
2.32
2.60
|l|||||||i;
IDispos'iiS
1.50
1.50
1.50
1.50
IIBenefifl
•mm^&*z&tm&.
•xttx:&sft&g:.iy£,<$i
'-::'.'"': :;.V::l:":"'::- • ..x^.T^.Vt':"1':'
iRecyclel
-0.07
0.62
0.50
4.07
- 158-
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JAERI-Conf 95-015
SUMMARY
The recycle of RSM to both restricted and unrestricted uses is economically viable
and highly desirable in an environment of high disposal cost. Businesses that
recycle RSM will be few in number, highly regulated, and vulnerable to nuances of
the regulatory process. The successful business approach will likely be outside of
the mainstream of conventional metals industry practice. The sources and uses of
RSM are roughly in balance. Restricted reuse could completely consume all
sources of RSM, however, this would not be the most economically advantageous
approach. A combination of decontamination and release, production of
intermediate metallurgical products for unrestricted use, and manufacture of final
products for restricted use is indicated.
REFERENCES
Whitfield, R.P., 1991, "Radioactive Metal Recycling: A DOE Assessment", Draft
Report, U.S. Department of Energy.
Duda, John, 1993, "U.S. Department of Energy's Weapons Complex Scrap Metal
Inventory", Morgantown Energy Technology Center, Draft Report.
Mizia, R.E. and Wahnschaffe, S.D., 1994, "Radioactive Scrap Metal Market Study;
Use of RSM by Electric Utility Industry for Spent Fuel Storage", WINCO-1190, Idaho
National Engineering Laboratory, U.S. Department of Energy.
Worcester, S.A., Twidwell, L.G., Paoiini, DJ,, and Weldon, T.A., 1993,
Decontamination of Metals by Melt Refining/Slagging - An Annotated Bibliography,
WlNCO-1138, Idaho National Engineering Laboratory, U.S. Department of Energy.
U.S. Department of Energy, 1991, "Environmental Restoration and Waste
Management Five Year Plan (1993-1997)", August 1991.
- 159 -
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JAERI-Conf 95-015
3-4 Economic Aspects of Recycling U.S. Department of Energy
Radioactive Scrap Metal
John Harrop & Neil Numark
Sanford Cohen & Associates
John MacKinney
U.S. Environmental Protection Agency
ABSTRACT
The U.S. Department of Energy (DOE) has substantial quantities of scrap metal contaminated with
low-levels of radioactivity. What is more important, current DOE Decommissioning and
Dismantlement (D&D) plans will generate even more radioactive scrap metal. Disposition of this
radioactive scrap metal could result in substantial costs to the DOE, but if certain options are
exercised, could result in an economic gain. This paper outlines five basic options the DOE could
follow for disposition of its radioactive scrap metal, and then examines the economic
consequences of each option.
A cost-benefit analysis was used to evaluate each of the five options. Real costs, derived from
DOE studies and private industry, formed the basis for all analysis. These include transportation,
packaging, processing (melt-refining) prices charged by industry, and burial fees and scrap metal
storage facility operating and surveillance costs faced by the DOE, Other potential costs, such
as the avoided costs of mining, and other less-well defined factors are assumed imbedded in the
prices charged by industry for processing radioactive scrap metal. The results of this analysis
show that burial cost is the most significant factor to consider in deciding which RSM disposition
option to pursue. Moderate variations in burial costs can dramatically change the outcome of the
cost-benefit analysis.
1. INTRODUCTION
This paper reports the results of a study of the various disposition options available to the U.S.
Department of Energy (DOE) for its existing and future radioactive scrap metal (RSM). Practical
and cost-effective methods exist for disposing DOE's current and projected inventory of RSM.
Some options, however, may prove significantly more cost effective and impose much less risk
to workers and the general population. Disposition options include storage, burial, and recycling.
DOE has extensive experience with RSM burial and storage. DOE has not attempted, but is
considering a comprehensive program to recycle RSM. A decision to recycle RSM must address
at least four interrelated questions: (1) Does technology exist to reduce radioactivity to a level
sufficiently low to allow the metal to be used in any credible application? (2) Does processing
and reusing the material pose a human health or environmental hazard? (3) What are the effects
on sensitive industries? (4) What is the economically preferable RSM disposition option? This
paper considers only the economic aspects of RSM disposition options.
The economic aspects of this study addressed two primary objectives. The first compares relative
costs of dispositioning DOE RSM through various options. Life cycle costs are required for each
- 160-
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JAEM-Conf 95-015
option. The second develops a workable system for analyzing cost effects for different disposition
options, given variations in the RSM inventory. Accordingly, this study addresses the following
questions: (1) Is there a significant cost difference among the various options for disposition of
DOB RSM? (2) Which decontamination technology is most cost effective? (3) Does a demand
exist, either within the nuclear industry, or within general commerce, for DOE RSM? (4) What
are the technical, cost, and schedule risks of each disposition option?
Scrap metal is a major waste stream for DOE, and presents both a potential resource and a
potential liability. According to one estimate, the current inventory of DOE RSM is 1.5 million
tons (DOE 91a). DOE's D&D programs are expected to generate additional quantities of RSM,
as much as 775,000 tons from the Oak Ridge Gaseous Diffusion Plant alone, over the next ten
years (EBA 91),
Recycling RSM has potential economic advantages. First, recycled RSM can be used in place of
virgin materials, thereby saving scarce resources. Second, provided that health risks are
acceptably low, recycling presents an immediate solution for disposition of radioactive materials,
thereby avoiding maintenance and monitoring costs for abandoned facilities. Third, recycling can
avoid possibly high land disposal costs.
On the other hand, recycling RSM presents potential economic problems. First, to recycle its
RSM, the DOE must find a market for this material, either within the nuclear industry or in
general markets in the United States and worldwide. Second, processing radioactive materials
entails costs that might exceed the costs of other options such as land disposal. Third, the cost
of achieving safe residual levels may be excessive.
2. RSM DISPOSITION OPTIONS
This analysis considered the following five RSM disposition options:
No Action. Under this option, DOE's existing inventory of RSM remains essentially as is, in
existing scrap yards, storage buildings, or other locations. As D&D generates more RSM,
additional material is added to existing or newly constructed scrap yards.
Safe Storage. All DOE RSM is removed to engineered facilities including existing warehouses
or new buildings. These facilities are designed to (1) enhance protection of human health and the
environment and (2) protect the economic value of the RSM. In general, the RSM would remain
at the site where it is generated. Safe Storage is an intermediate solution. The RSM is eventually
buried—at a central facility such as the Nevada Test Site—or placed into permanent storage.
Alternatively, the DOE could recycle the RSM. Delay in ultimate disposition may impose
additional maintenance and monitoring costs, but would allow retention of an economically
valuable resource until either (I) the short-lived radionuclides contaminating the metal decay to
less than the residual radioactivity level required for unrestricted release, or (2) facilities are
constructed to recycle the RSM.
Disposal (Immediate Burial). For this option, all DOE radioactively contaminated scrap is
buried as it is generated.
Restricted Recycle. DOE's current and future RSM is processed in a multi-metal recycling
- 161-
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JAERI-Conf 95-015
facility for controlled uses such as radioactive waste burial containers or shielding blocks for
accelerators. Radioactive carbon and stainless steel is prepared for its new use by melt refining.
Unrestricted Recycle. This option is very similar to restricted recycle, in terms of operations
required to process metal from scrap to a refined ingot. Unrestricted recycle assumes the scrap
is first decontaminated (by melt-refining or mechanical decontamination, for example) and then
sold to a commercial mill where it is remelted and ultimately used for industrial or consumer
products. Despite initial decontamination, the metal still contains residual radioactivity.
3. FUTURE DOE NEEDS FOR RADIOACTIVE SCRAP METAL
One possible recycle alternative available to the DOE involves processing its RSM and producing
components that are used internally to DOE. For DOE to decide whether to pursue a strategy of
internal recycling the quantities of materials required for such applications must be assessed.
Proposed internal uses include:
- High-level waste canisters - nickel, stainless steel, copper
• Concrete reinforcing bar - carbon steel
• Waste boxes - carbon steel
• Shielding blocks for high energy accelerators - carbon steel
The DOE Office of Technology Development is funding research programs to explore some of
these options in greater detail. For example, the Manufacturing Sciences Corporation and the
Colorado School of Mines are studying the decontamination of nickel from the gaseous diffusion
plants. The decontaminated nickel and decontaminated carbon steel would be remelted with other
alloying elements and cast into stainless steel ingots. The stainless steel could then be used to
produce containers for storing vitrified high-level waste (DOE 93a).
A considerable amount of work has focused on the use of radioaetively contaminated stainless
steel for high-level waste canisters. As much as 26,000 tons of stainless steel would be required
to handle the needs of high-level waste producers (SAI 93). A Waste Immobilization Facility is
scheduled for construction at the DOE's Idaho Chemical Processing Plant. The facility will
produce a glass ceramic waste for final disposal. This waste form is densified in process waste
cans. Three of these cans are then placed in a repository waste canister. Westinghouse Idaho
Nuclear Company estimated that about 9,000 tons of stainless steel are required for the process
cans and about 14,000 tons are required for the repository canisters (ICP 93).
Probably the least complicated recycling application involves remelting carbon steel scrap to cast
shielding blocks for accelerators. This application represents an immediate market for about
37,000 tons of steel scrap (SAI 93, SEG 93). The Scientific Ecology Group, Inc. projects a long-
term requirement of 200,000 tons for shield blocks (SEG 93a).
- 162 -
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JAERI-Coof 95-015
Table 1 compares the current DOE Inventory,1 plus quantities of radioactive scrap metal to be
generated from D&D activities, with potential internal recycle applications.
Table 1. Comparison of Potential DOE Recycle Applications with Inventory
and Expected D&D Generation of Radioactive Scrap Metal (tons)
ilSpi^tefiallilll
Copper
Carbon steel
Stainless steel
ll&iilil^l^plii^Mllllllll
Accelerator use
Shield blocks
Rebar and mats
Transport and disposal boxes
HLW canisters
Universal dry fuel casks
lllcSffliplll
:ll*Required»
1,000"
200,000"
100,000*
500,000"
23,000C
282,000"
1 •: : A\^' ^' ; v ,;. :'. ::;S. y <• v V: :y:':x-*'.',-:-Y
'.-.•::-, •::-M-.':-.-.-:>X>' •:-,>:•->:>"• ,•;-'-,-:-;;•'
tmzf^f-XZfSfm^
wmventory-:?-*':
24,000
59,000
7,400
;ilH:F«iSre1it
";M"f':D&Dl;:s;.
49,000
470,000
27,000
'SAI93
b SEO 93a
e ICP 93
d Another 500,000 tons of metals, other than copper, carbon steel, and stainless steel, are projected from
future D&D (EBA 91).
4. ANALYSIS OF COSTS AND BENEFITS '
This analysis shows the present worth of various options DOE could use for disposition of
radioactive scrap metal. Some benefits are not explicitly included in the analysis. For example,
one benefit not included is the energy savings between mining replacement ores and restricted or
unrestricted use of DOE RSM. This is because DOE scrap metal inventories, and DOE's potential
demand for metal products, are insignificant— less than 0.1% compared to annual U.S. production
and demand. DOE's limited demand for metal would be quickly met by the U.S. scrap metal
market, rather than resulting in additional iron ore mining. Thus, counting the energy savings
from using scrap rather than mining virgin ore is inappropriate.
Energy savings are also excluded here because this factor is already considered in pricing scrap.
By similar logic, accounting for many other specific cost elements is not required to produce a
valid benefit-cost analysis. For example, several options analyzed involve transportation of RSM
to various DOE facilities, some at considerable distances from the originating point of the RSM.
One might be tempted to quantify, from an economic perspective, additional costs associated from
vehicular accidents during transport. But the rate charged to haul RSM ($0.10 per ton-mile)
already includes this consequence. This rate includes payment of the trucking company's
insurance premiums and other liability mitigating factors that either regulations or market
pressures dictate. Assigning a value to each truck accident, and attempting to include this factor
in the analysis, would again artificially double costs.
To perform the cost analysis, we assumed that the DOE would process approximately 65,000 tons
per year for each of the five options. This number represents approximately the amount of RSM
1 The inventory number was obtained from "Analysis of the Potential Recycling of DOE Radioactive
Scrap Metal," a draft report for the U.S. EPA Office of Radiation and Indoor Air. The inventory quantity
in Table 1 is much smaller than DOE estimates of RSM (DOE 91b, 92a).
- 163-
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JAERI-Conf 95-015
in storage at the DOE Oak Ridge Tennessee facilities (DOE 9 la), and is about the amount of
carbon and stainless steel projected annually from D&D of the Oak Ridge area gaseous diffusion
plants (DOE 91a, 91b). This quantity of steel can easily be handled by the Nevada Test Site for
burial (BEC 93, REC 93) but represents about 400% of the capacity of one firm, Scientific
Ecology Group, Inc., that currently recycles RSM (SEG 93),
4.1 Cost Analysis
This section estimates the economic costs for each option primarily in relation to the disposition
of radioactive steel currently in DOE scrap yards. For purposes of these estimates, all operations
occur in one year. Furthermore, decontamination processing is conducted by a private contractor
so there are no direct DOE capital investment costs in establishing the recycling facilities.
No Action. The costs associated with the No Action option are those for continued maintenance
of scrap yards in their current form. These maintenance costs include physical security (lights and
fences), radiation surveillance and monitoring, and regular inspections. In addition, there are
significant potential costs associated with regulatory compliance under the No Action option.
There is also the possibility that some improvements in storage conditions may be required. The
continued maintenance costs for the K-25 site are presented in Table 2. The values for K-25
(Table 2) are used to derive an annual RSM maintenance cost of $2.00 per ton.
Table 2. Maintenance Costs for Continued Storage of Radioactive
Scrap Metal at the K-25 (Oak Ridge) Site
Security
Surveillance/Monitoring
Inspection
Total
Annual Cost*
$ 8,000
31,500
3.500
543,000
* Operating and maintenance costs are annualized over a 25-year period.
Safe Storage. The costs involved in Safe Storage include those associated with sectioning and
packaging the metal, and storage maintenance costs. The maintenance costs are identical to No
Action above (Table 2), plus building maintenance costs. DOE could store the scrap metal in
surplus buildings and thus avoid building construction costs.
Sectioning scrap metal is required for all options involving transportation of the RSM from the
scrap yard. The unit cost for sectioning most materials (e.g., structural components) is about
$0.10 per pound ($200 per ton) based on estimates from firms that provide this service. The unit
sectioning cost for sectioning more complex components (e.g., heat exchangers) is $0.15 per
pound ($300 per ton). Based on a mix of 58,000 tons of carbon steel and 7,000 tons of stainless
steel components, only 36,000 tons of which require sectioning, the average sectioning cost is
$221 per ton.
The RSM must be placed in containers for storage or transport. A special container, or B-25 box,
is used for all storage and transport containers. The box accepts 92 cubic feet of material and
costs $350 (SEG 93). We assume each 50 to 100 pounds of sectioned material occupies one
cubic foot. Ingots and size-reduced RSM are estimated to occupy one cubic foot of volume per
_ 164 _
-------
JAERI-Conf 95-015
240 pounds of material. Unit cost estimates for labor and equipment needed to package the
radioactive metal in full compliance with all DOT and DOE regulations was obtained from firms
providing these services. Costs range from $50 to $400 per ton and average $100 per ton. The
packaging costs for the DOE radioactive scrap material inventory considered in this analysis
averaged $156 per ton.
Building maintenance is estimated at an annual cost of $0,77 per ton. The costs for Safe
Storage—by metal type—by summing sectioning costs, packing costs, continuing maintenance
costs, and building maintenance costs was $227 per ton.
Disposal (Immediate Burial). The costs associated with the Disposal option include the initial
characterization required by the site waste acceptance criteria, sorting, sectioning, plus compacting,
packaging, and shipping. This analysis assumes all waste is sent to the low-level radioactive
waste {LLW) repository at the Nevada Test Site for disposal. The sectioning and packing costs
are the same as for Safe Storage. The waste acceptance criteria cost $15 per ton and the current
disposal cost is $282 per ton (EEC 93).2 Transportation costs are estimated to be $0.10 per ton-
mile. Total costs for low level radioactive waste disposal are $787 per ton.
Restricted Release Recycle. The costs associated with recycling RSM for restricted release
include packaging, sectioning, and transporting the material to the recycling facility and are
additional to the decontamination cost at the recycle facility. The net cost for the process is the
total of these costs less the value of the recycled metal to DOE (the cost of the material it
displaces). The estimated sectioning and packaging costs are identical to those of the Safe Storage
and the Disposal options. Transportation costs are estimated at $0.10 per ton-mile based on the
distances from remote storage locations to the processing locations in the Oak Ridge area. A cost
of $0.20 per ton-mile is used for the Oak Ridge scrap yards to adjust for fixed costs distributed
over such short travel distances.
Table 3 presents unit decontamination costs based on price estimates from commercial firms that
decontaminate scrap metals.3 The prices cover all costs associated with facility operation; such
as compliance with all environmental, safety, and health regulations, and all fixed and maintenance
expenses, such as disposal of the radioactive waste produced during decontamination. The choice
of decontamination process is determined by the nature of the contamination and the restricted use
end product. Carbon steel will probably require remelting. The resulting ingots are ready for use
as shielding blocks or for fabrication into disposal casks. Stainless steel also requires melting
before fabrication as disposal containers or storage casks. Table 4 lists possible uses for restricted
recycled metals. The cost ranges of the materials that would be displaced* are presented along
with the Net Recycling Cost — the appropriate decontamination costs from Table 3 less the
displaced material value.
2 The $282 per ton assumes a disposal cost of $7.05 per cubic foot and a density of 50 Ib per cubic
foot. At 100 Ib per cubic foot and S5.33 per cubic foot, the disposal cost at NTS is $107 per ton,
1 Scientific Ecology Group Inc. and the Quadrex Corporation.
4 The displaced material cost is the cost of obtaining a given quantity of a specific grade of metal
minus the scrap dealer price.
-165-
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JAERI-Conf 95-015
Total costs for Restricted Recycling of the present DOE inventory of radioactive scrap steel are
shown in Table 5, The associated recycling unit costs are SI,250 to 53,050 per ton of carbon
steel. The unit costs for stainless steel range from $900 to $1,050 per ton, showing a possible net
savings from Restricted Recycling.
Table 3. Costs of Decontaminating Radioactive Scrap Metal by Process and Metal Type
•:;:SEH5Sfe;™£J^
^mm^'^^e<^^lo^<^-^ocessMetai--:rim-:?^^y:
Mechanical Decontamination Process - Carbon
Steel
Mechanical and/or Chemical Decontamination
Process - Copper
Mechanical and/or Chemical Decontamination
- Aluminum
Chemical Decontamination Process - Stainless
Steel
Metal Melt - Carbon Steel
Metal Melt - Stainless Steel
Metal Melt - Copper
Metal Melt - Aluminum
Metal Melt - Nickel
Electro-Purification Process - Nickel*
:licSl:l®i:;:Tonlt
illiip;!fl9935$|i
$1,580 - 53,500
SI, 600 - $7,600
$2,800 - $6,700
$4,800 - 56,400
$1,980 - 53,280
$2,000 - $3,200
$2,000 - $3,300
$2,000 - $3300
$2,000 - $3,300
$3,500 - $4300
WjiWi&i^M^i^^SM
Surficial contamination and
uncomplicated geometry.
Surficial contamination.
Surficial contamination.
Surficial contamination.
Surficial or volumetric
contamination.
Surficial or volumetric
contamination.
Surficial or volumetric
contamination.
Surficial or volumetric
contamination.
Surficial or volumetric
contamination when there are
not high levels of
technetium-99.
Surficial or volumetric
contamination when high
levels of technetium-99 exist.
* This is not a proven technology. Estimate based on preliminary research.
Unrestricted Recycle, The cost elements associated with the Unrestricted Recycle option are
the same as those for the Restricted Recycle option. There are, however, two differences in the
recycling costs. The first is the value of the recycled scrap metal. This metal is valued at scrap
dealer prices listed in Table 6 instead of the value of displaced materials used in Restricted
Recycle. Second, the recycling and total costs are more likely to be at the high end of the range
shown due to the need to meet the presumably more stringent free-release standards.
The costs for recycling the current DOE inventory of radioactive steel for unrestricted release are
presented in Table 7. The unit costs range from $1,775 to $3,475 per ton for carbon steel, and
from $2,100 to $3,400 per ton for stainless steel.
Cost Comparisons. The total unit costs for steel are shown for each option in Table 8. The No
Action and Safe Storage options do not consider the costs for the final disposition of the material.
166 -
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JAERI-Conf 95-015
These costs may be increased by delaying action to either bury or recycle the RSM, Of the final
disposition options, Disposal as low level radioactive waste is the least expensive option. The
possible net savings shown for the restricted recycling of stainless steel depends on: 1) the quality
of the remelted stainless steel allowing for the displacement of 316 grade stainless steel for use
in making containers and casks and; 2) the decontamination costs being at the low end of the
estimated range shown in Table 3.
Table 4. Restricted Recycle Uses and Displaced Material Cost
'>£";¥„-£ ;vi*. >;>'" :>:>.-:>x'x::;<:<;:;:<):;:;^y •:•:<:•; •
•';>:•:': ';*'£V'":-:'":: ;;!|Xv>;|' ;::%&£<;£:&;£:£•'•
llff|e1ofMe1ail;:':
Carbon Steel
Stainless Steel
Copper
Nickel
Lead
Shielding Blocks
LLW Disposal Containers
LLW Disposal Containers,
Storage Casks
Wiring or Electrical Circuitry
Produce Stainless Steel
Shielding
li-D'isplalcelJJltfatenal.fe
lli^ue^f^nllf
^»ffiRwg£3&f%v;»i;.:-'
500 - 1,000
100
2,640 - 3,440
up to 1,600
4,400
180
l^NeffRfc'wie^Ocjsiifl
^j;s*i:::ssi:¥pef^fci:;;l5li::l
980 - 2,780
1,880-3,180
(1,440)' - 560
400 - 1,700
(2,200)* -(1,100)"
NA
NA - Not Available
* Values in parentheses represent a net savings
Table 5. Costs of Recycling Current DOE Inventory of Radioactive
Steel for Restricted Use
: •-"-'<' ::'-;v>'':':">.':-,:':::"-;-;-'1:-;:;':-:-: ";•.•,•"• '.':"
/;:''"•':' ";-:C V;vS':"-;:x'-:v:v:''-::.:' •';',' :;":'-;
- '.v;:':;'':'-x':v:;;-:"::;':";;::-rV:::":"::: --'•'-'-':•'---
':-.';-;-: ••::::":->;'-:v:«.:/>:v::VrV:". .': •:"•"• :.:''.:::
WSM^AK&:
Carbon Steel
Stainless Steel
All Steel
•jgi Sectioning'*
tiEa^gingll
wrsitooowi
14,868
3346
18,214
;':::'".": : ->;-.'>••/:•:••,•„':'•,';":-:-:-;•:': ':;:'""'
•:":-.: "•:•;• •-::;",-.-;-;';-, -•••';-•"•'•:••:-;•:•;•:-:•
llPCoi-olil
•:i:SHipmng?:?i
:|K$liOOOli^
1,137
208
1345
'sl>JerRecycliiig':Cost':-:'
mm^(i^Qyp^»^
58,000 - 162,000
(10,200f - 4,000
48,000 - 166,000
-' :•; ' ,-: .-:- -.-;•: ; :•'- •• ••>•- ,-. .-.-.•. i. : ••:-: '.•:-:•" -.' • ; .-'
-••;•;-: :-; -" .-; :-:-; :-••' -••'•'••• •' ,:•'••:•'•,:-':•'••• *-. : :
>,-••„•;-• ,•.,-:>,-,-:-,' ,--*-.-;•-' "•;-*..-. :-.•:•••••"•'•,-:•'••'••
'::, •.•:".'• •• i'" ;'-;.-.-'"-:.-:-; --"• '-',•:••/:•':";"•':": '>.":•."•
SiToii'NeY'CosiPi
^•-••- .:'-:-:-':'« :>Type"of 'Me tal'w^;^;?'^
Carbon Steel
Stainless Steel
Copper
Aluminum
Nickel
Lead
>iSTvpical Scrap Dealer "": •*•
::;€-:- Pries Per Ton ($) :^ :'
80
400
1,500
500
4,400
164
- 167-
-------
JAERI^Conf 95-015
Table 7. Costs of Recycling Current DOE Inventory of Radioactive Steel
for Unrestricted use
lllilllllK
llls|^t&:|l:f:?
Carbon Steel
Stainless
Steel
All Steel
•:Mi;Se6tionin|:w;
IIPacKagmgli
mmmffm
14,868
3346
18,214
Milillili
lllC'c^-ofjl;
igSMppiriglt
«TSl;M05-P-
1,137
208
U45
:ii^iiiliiii,¥;:-*Sfl^OO^::SSiK:!S«*-:
80,000- 188,000
11,000-20,000
90,000 - 210,000
iirlliilef feBtl I
ippimcoooif « £
103,000 -203,000
15,000 - 24,000
118JGO- 227,000
Table 8. Comparison of Unit Costs by Option for Radioactive Steel
;llib^&on1tjpe-;Steeril|-
No Actioa
Carbon Steel
Stainless Steel
Total Steel
Safe Storage
Carboa Steel
Stainless Steel
Total Steel
LLW Disposal
Carbon Steel
Stainless Steel
Total Steel
Restricted Recycle
Carbon Steel
Stainless Steel
Total Steel
Unrestricted Recycle
Carbon Steel
Stainless Steel
Total Steel
ll^peMffi(Gost!l
IWlet:Ton?!(S¥«l
0
0
0
250
470
280
760
990
780
1,300 - 3,100
(94Q)1- 1,100
1,000 - 2,800
1,800 - 3,500
2,100 - 3,400
1,800-3,500
m»^vmiK:mi«mmm
* AnnualMaMteriance -&
mcm^sr^m^m
2
2
2
3
3
3
0
0
0
0
0
0
0
0
0
>m««:m*mmKmm:mf
;«?First:YeMGdst»;:
::«l^efJT6n;iS)w«i
2
2
2
260
470
280
760
990
780
1,300 - 3,100
(940)*- 1,100
1,000 - 2,800
1,800 - 3,500
2,100 - 3,400
1,800-3,500
* Values in parentheses represent a net savings.
4,2
Basic Assumptions of the Cost-Benefit Analysis
DOE currently has 65,000 tons of radioactive scrap carbon and stainless steel to be dispositioned
in the first year of each option (except the No Action option). An additional 95,000 tons of other
scrap metal is not considered in this analysis. Other assumptions include:
• Each year, for ten years, DOE will generate an additional 65,000 tons of
radioactive carbon and stainless steel in the Oak Ridge area. This material will
be used for all options. All cost-benefit analyses will be conducted over 22 years.
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JAERI-Conf 95-015
All disposal (land burial) will occur at the Nevada Test Site at a total cost of
$297 per ton in 1994 dollars. Distance to the site is 2,000 miles. The
transportation cost is $0.10 per ton-mile.
« All metal recycling occurs in a multi-metal melt facility in the Oak Ridge,
Tennessee area. Transportation to this facility costs $0.20 per ton-mile and the
average distance is ten miles.
The distance to a commercial steel mill for the unrestricted recycle option is 300
miles. All costs associated with unrestricted recycle are based on free on board
(FOB) the DOE Oak Ridge facilities.
* A discount rate of 7% was used for present worth analysis, as specified by the
U.S. Office of Management and Budget (OMB 92). Because this discount rate
may not accurately reflect the historic time-value of money, the analysis is also
presented at 3.5% and 10% discount rates.
43 Summary of Cost-Benefit Analysis
Table 9 shows the present worth of dispositioning 65,000 tons of RSM annually, using different options.
The values are rounded to the nearest $1,000,000. None of the options shown in Table 9 have a benefit
minus cost of greater than or equal to zero (i.e., B-C > 0). By strict cost-benefit analysis criterion, none
of the options qualifies as a favorable alternative. However, we assume DOE must do something, if
nothing other than taking no action. In fact, the No Action option may be unacceptable for a variety of
overriding policy issues. Disregarding the No Action alternative, the next lowest cost option is Safe
Storage. This may seem surprising since, for our analysis, Safe Storage includes ultimate burial at the
Nevada Test Site. The reason is that the data presented in Table 9 represents the present worth of the
stream of economic consequences of each option. Delaying disposal means that the present worth of the
burial costs is discounted significantly.
Both recycling options show significantly higher present worth costs compared to every other option. Two
points need to be made here. First, only DOE, or some representative of the Federal government,, can
decide what all the objectives are when analyzing RSM disposition options. This analysis does not
include all consequences, and cannot until DOE specifies its objectives more completely and precisely.
When additional, difficult-to-monetize consequences are considered, recycling may be the "best" option.
The second point about recycling options requires not confusing melting of RSM with production of
normal steel in an average steel mill. These are two vastly different processes, and cannot logically be
compared on an economic basis. One need only visit the Scientific Ecology Group, Inc. facilities in Oak
Ridge, Tennessee, and an average steel mill to appreciate the difference.
The data in Table 9 suggests that Safe Storage is the "best" option. However, other clearly relevant
consequences, not monetized here, must be considered when deciding which option to follow. For
example, human health effects, such as the total dose and risk of cancer are important factors.
Some monetary costs and health risks associated with each of the five disposition options are easy to
quantify. Other costs present greater difficulty because DOE has not exactly specified their objectives in
pursuing a particular course of action. For example, DOB could completely remediate the Oak Ridge area,
and return the federal lands used for the scrap yards to the state of Tennessee. One benefit from this
course of action is the land value available to the state government or private enterprise. Of course, one
of the costs is the pro rata share of the costs of remediation. These costs and benefits are not included
169 -
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JAERI-Conf 95^015
in the No Action scenario, but might be for the other four options. Therefore, if one of the DOE's
objectives in dispositioning RSM is to clean up and eliminate scrap yards, there are additional costs and
benefits to consider. Table 10 summarizes costs and benefits for each disposition option.
Table 9. Present Worth Comparison for Five Disposition Options
*>*i::y'>:::;.":v::>//':V;7£,^^
:•;•;•.•' •:-•-'.•.••-:-:-: :-.••- -.-: :-:-:---.-;---.-, :*• -.-.-:-:•.• • •••:•-. ; -•-•: :• •• :-••: :•;•:-:- • •:•:•: :-: :•:•:
•';• 'l-^Vx^x..^-'^; x;:]:^
No Action
Safe Storage
Disposal (Immediate Burial)
Restricted Recycle
Unrestricted Recycle
/:;::|::2^
-x'riv:1;-.-:;^"'.'" ';"':'::':'i,"::".»:i"'i;:":':;f';':""":':':''':>-':'-*:'::1'-'
.;-.,-;"••:' -"-, :•-:•:• •>;:••.•> 2 "-:\OA:'--"'^:-t x-x '.-' •--.•-;--.--•---;:
.*-:•-.-- v. •;•-•-,-;•:• ,:•::•;• :-^).^J /Or"<^--'-'-,-:--'>-'---t-'-:-:--:^>
$16
$379
$458
$638 - 1,227
$766- 1,840
KEtiWxwrtbSfin ^millions);"
.,,;.,. :-..;,'. ..-"'.%>;._,;.;.: 'o .v-^. -. :-. .- • ..-'..-v-; .-.•.-.-•-'-.••^•-' '
$11
$262
$381
$531- 1,022
S638 - 1,532
• •,-:•?> :••• •:• •:•:-.•:-:-:,-: :-.-,". .•.:.:-.."..--:-; ,-.""•<•"• ."
:•:•••;•• .'.•: •• •'•'•'. "•:•:•,"•: v,-:,. • . • :-:• :-,-t;-.-, I-.-":.-:*;
•:•:•:;••:• •.•: ";:•:.:;:>;.<>,;;>:•-> • • :-:;:;;:.:r r;;f-:c;"\_:-
":•:•:•>" ';";•. •;-,'-• .-wx-r"'":-'-'' ' -v;:>'w:-:<;:'i"r;x ::::::.:
$8
$200
$330
$460 - 885
$553 - 1,327
* Present worth values have been rounded to nearest million dollars.
No Action Summary
With respect to maintaining radioactive metal scrap yards, the No Action cost-benefit analysis is predicated
on DOE's compliance with applicable federal regulations. Our analysis indicates current practices at these
scrap yards may meet all required regulations. If additional regulations are imposed on DOE, the costs
associated with No Action could rise dramatically, particularly if all the RSM is relocated.
Presently, maintenance and surveillance of DOE's scrap yards is relatively inexpensive, approximately S2
per ton-year. Additional scrap yards must be built, or existing scrap yards expanded, to contain the RSM
expected from D&D of various DOE facilities. This analysis assumes the per ton cost, however, to remain
the same S2 per ton. The No Action option could involve many other costs. No Action provides no long-
term solution to managing RSM. Eventually, the RSM will most likely have to be moved, buried, or
recycled. Therefore, No Action is simply delaying the inevitable. However, as indicated by a present
worth analysis, delay is not necessarily bad presuming monies withheld from the RSM disposition are
invested at an appropriate opportunity cost off capital.
Safe Storage Summary
Safe Storage is another alternative that delays the ultimate disposition of RSM. After some years in Safe
Storage, the RSM must be buried or recycled. One obvious benefit of leaving RSM in Safe Storage is
that radionuclides with short or moderate half-lives decay to lower levels of activity, perhaps below
residual radioactivity limits. More likely, the radioactive decay will simply reduce the activity levels to
the point where handling RSM is less costly, or decontamination is possible. As an example, if carbon
steel contaminated with activated cobalt-60 were maintained in safe storage for fifty years, the activity
levels associated with the cobalt-60 would be reduced significantly. The present worth cost of maintaining
one ton of carbon steel in Safe Storage for fifty years is approximately $28. Assuming the radioactive
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JAERI-Conf 95-015
decay reduced the steel's activity to less than the residual limit, presumably the steel could then be sold
at its scrap value, $80 per ton in 1994 dollars, without any further costs associated with decontamination
or surveying.
Another advantage of Safe Storage is—if one accepts the assumption some form of ultimate disposition
of the metal is required—-delaying the final disposition reduces the present worth cost of that disposition.
Conversely, burial rates could increase at a rate much higher than historical values presently indicate,
offsetting any gains made through time value of money.
Finally, Safe Storage allows DOE to protect the economic value of certain metals until decisions are made
regarding the ultimate disposition of those metals. For example, DOE currently owns over 9,000 tons of
nickel which are volume contaminated with technetium-99. No technology exists to decontaminate this
nickel, but at least one technique, electro-purification or inducto-slag melting, is being evaluated for this
purpose. Because of the relatively high per ton value of nickel, a Safe Storage option for this particular
metal may make good economic sense.
Table 10. Summary of Costs and Benefits for DOE RSM Disposition Options
No Action
Requires maintenance of existing and
new scrap yards.
Results in long-term exposure to
workers.
Requires eventual disposal of RSM.
Requires continued compliance with
federal and state environmental
regulations.
Is aesthetically unattractive.
Based on current scrap yard sizes, over
50 acres of land will be required.
Newly constructed scrap yards may
require environmental restoration after
RSM is removed.
Possibly exposes the general population
to radioauctides from groundwater,
^surface water and air pathways.
Scrap yards are relatively
inexpensive to construct and
maintain.
Allows delay in decision
whether to bury or recycle.
Takes advantage of natural
radioactive decay to reduce
activity levels.
Safe Storage
Results in long-term exposure to workers
who must monitor the RSM.
Requires long-term institutional controls.
The RSM must still be either buried at a
later date, or recycled.
Buildings used for Safe Storage may
eventually require decontamination and
expensive remediation.
Provides enhanced protection
for human health and the
environment by isolating the
RSM from its surroundings.
Allows delay in decision
whether to bury or recycle.
Takes advantage of natural
radioactive decay to reduce
activity levels.
Preserves the economic value
of metals until a suitable
decontamination technique
exists.
Disposal
At current burial rates, immediate
disposal is more expensive than either
Safe Storage or No Action.
Loss of potentially reusable metals
which will require replacement from
uncontaminated sources.
Provides for immediate
disposition of RSM as it is
generated during D&D.
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JAERI-Conf 95-015
Table 10 (Continued)
Option
Restricted
Recycle
Costs
Recycled RSM, not used for waste
containers, could become contaminated
and may eventually require burial.
Special facilities are needed to
decontaminate the RSM and/or fabricate
new components from the RSM.
Creates secondary waste streams.
Benefits
Provides for immediate
disposition of RSM as it is
generated during D&D.
Increased employment
opportunities at RSM recycle
facilities.
Avoids costs of burial.
Avoids costs of obtaining
nuclear components from
clean metals.
Unrestricted
Recycle
Special facilities are needed to
decontaminate the RSM.
Potentially exposes workers and the
general population during processing of
RSM.
Exposes the general population from use
of RSM in consumer products.
Radioactivity in certain metals could
impact sensitive industries.
Provides for immediate
disposition of RSM as it is
generated during D&D.
increases employment
opportunities.
Avoids costs of burial.
Recycling uses less energy
and raw materials than does
production from virgin ore.
Safe Storage requires a commitment to institutional controls. Buildings must be maintained, arid
radiological, security, and safety requirements must be met. Furthermore, regulatory requirements could
become more stringent in the future and the land occupied by the storage buildings is not available for
other, potentially more valuable purposes. Finally, the buildings used for Safe Storage may themselves
require expensive remediation as a result of being used to store radioactive materials.
Disposal Summary
The Disposal option involves burial at a long-term disposal facility such as the Nevada Test Site. Once
interred, the RSM is non-retrievable, hence the metal's economic value is lost forever. From an economic
perspective, Disposal seems fairly attractive. Current disposal costs for low level radioactive waste at the
Nevada Test Site are about $7 per cubic foot. The burial rate would have to increase to almost $15 per
cubic foot before the present worth cost of burial of the 65,000 tons considered in this analysis exceeds
the most optimistic present worth cost of Restricted Recycle.
Restricted Recycle Summary
The major advantage of Restricted Recycle is that metals are available for reuse where they either become
contaminated again, after being decontaminated, or will be further contaminated. In either event, virgin
materials are not needed. The DOE may prefer to bury other forms of radioactive waste, such as spent
fuel, in containers fabricated from RSM. Some current recycling applications, specifically shielding
blocks, may simply delay the ultimate need to dispose of the RSM. That is, when shield blocks are no
longer needed, they may be classified as RSM and then require storage or burial.
Restricted Recycle is very costly due to the radiological controls required to fabricate various components.
This analysis considered only decontamination costs. Chemical, mechanical, and metal melt
decontamination costs all greatly exceed the scrap value of the metal. Other costs of fabrication, such as
rolling ingots into shape, were not considered. If, as a result of metal melt, for example, the RSM was
still contaminated above a residual radioactivity limit, the rolling process could also cost considerably
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JAERI-Conf 95-015
more than rolling uneontaminated metal,
Unrestricted Recycle Summary
Unrestricted Recycle has all the costs of Restricted Recycle plus additional costs attendant with the fact
the RSM must be decontaminated below a residual radioactivity limit Current technologies are inadequate
to decontaminate RSM with certain radionuclides, such as technetium-99. Furthermore, decontamination
below specific low levels, such as 10 pCi/g, may not be possible,
4,4 Sensitivity to Cost Variations
The current Nevada Test Site disposal cost of approximately $7 per cubic foot could increase substantially.
If burial costs rise sufficiently, both restricted and unrestricted recycling could become the least cost
option. Likewise, if RSM decontamination costs (i.e., the cost of melt refining) were to decrease
substantially, restricted and unrestricted recycling could be the lowest cost option.
The cost-benefit analysis shows that if burial costs increase to approximately SI 5 per cubic foot, Restricted
Recycling would cost less than Disposal. At approximately $65 per cubic foot, the Unrestricted Recycling
option would cost less than the Disposal option.
If the cost of melt refining is reduced, both recycling options tend to be more attractive. Table 11 shows
the affect of reducing the cost of melt refining to 50% and 25% of the value shown in Table 3. For
example, if the cost of melt refining is reduced to about $1,200 per ton, then Restricted Recycling is equal
in cost to Disposal. At about $600 per ton for melt refining, Restricted Recycle is superior to all other
options including the No Action option.
Table 1L Present Worth Comparison for Five Disposition Options,
With Varying Melt-Refining Costs (Discount Rate = 7%)
J: ;•£.; 'S :te:::;::SsOptioii f :! ?&1 111 ?:-: •
No Action
Safe Storage
Disposal (Immediate Burial)
Restricted Recycle
Unrestricted Recycle
mK^Xlliii-,
I Melt-Refining Cost
$11,000,000
$262,000,000
S38 1,000,000
$870,000.000
$1,360,000,000
fiiijimSmt^
?lMelt-Rermin£ Cost =::•
$11,000
5262,000,000
$381,000,000
$382,000,000
$750,000,000
•I*1S;: -:' %< 7 so/. •viSSiSsiW;
•-;v-: i-.-x; - "-:,:- Zj /O, ::':-;-• •:•¥. :':.:v.;.-\ •
•: MeltiRefminfe Costs
$11,000,000
$262,000,000
$381,000,000
$ . 44,000,000*
$440,000,000
The negative value indicates that the benefit derived from reuse by DOE, of its own RSM, exceeds
the cost of processing.
5 REFERENCES
BEC 93 Personal Communication with Bruce Becker, REECO, December 28, 1993.
DOE 9 la "Radioactive Scrap Metal Recycling: A DOE Assessment," Draft White Paper, U.S.
Department of Energy, prepared by Office of Technical Services/Weston and H&R,
August 1991.
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JAERI-Conf 95-015
DOE 91b "U.S. Department of Energy Decontamination and Decommissioning Planning and
Operations Experience (Draft)," W.E. Muiphie, U.S. Department of Energy, (Document
prepared by M.L. McKeman, Roy F. Weston), International Conference on Nuclear
Engineering-1, Tokyo, Japan, November 8, 1991.
DOE 92a "Radioactive Scrap Metal Recycling: A DOE Assessment," MJ. Lilly et al, U.S.
Department of Energy, presented at Waste Management '92, Tuscon, AZ, March 1992.
DOE 93a "DOE Selects 19 Projects for Recycling and Disposing of Materials from Closed
Facilities," DOE News Release, January 28, 1993.
EBA 91 "Environmental Restoration of the Gaseous Diffusion Plants," Ebasco Services, Inc.,
October 1991.
ICP 93 "Five-Year Program Plan for Metal Recycle/Waste Minimization (ADS ID-10008-
WN)," BJ. Frazee, Westinghouse Idaho Nuclear Company, Inc., January 1993.
LAN 93 Private Communication from J. Donahue, Los Alamos National Laboratory, April 1993.
OMB 92 "Discount Rates to Be Used in Evaluating Time-Distributed Costs and Benefits," Office
of Management and Budget, Circular No. A-94, Revised, Washington D.C., October 29,
1992.
REC 93 "Private Communication from T. Holmes," Waste Minimization Project Office - REECO,
March 1993.
SAI 93 "Recycle of DOE Radiologically Contaminated Metal - A Scoping Study (Draft)," T.
Hertzler et al, Science Applications International Corporation, February 1993.
SEG 93 Private Communication from Dewey Large, Scientific Ecology Group, March/May 1993.
SEG 93a Testimony before the House Subcommittee on Energy, May 17, 1993.
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JAERI-Conf 95-015
3-5 SUMMARY OF INDUSTRIAL IMPACTS FROM
RECYCLED RADIOACTIVE SCRAP METALS
Jean-Claude Dehmel, CHP
John Harrop
Sanford Cohen & Associates
and
John A. MacKinney
U.S. Environmental Protection Agency
ABSTRACT
During operation, decontamination, and dismantlement, nuclear facilities in the United
States and foreign countries are generating significant quantities of radioactive scrap
metal (RSM). Future decommissioning will generate even more RSM. The petroleum
industry also generates RSM in the form of equipment contaminated with naturally
occurring radioactivity. Finally, the accidental melting of radioactive sources in steel
mills has generated smaller amounts of contaminated metals. Steel mills, smelters, and
foundries could recycle these materials, which might then appear in finished products or
as feedstocks used by other industries.
If introduced in this manner, residual radioactivity can adversely affect the performance
of certain products. Such products include computers and other devices that rely on
integrated circuits. The most important effect of residual radioactivity on integrated
circuits is a phenomenon known as "single event upsets or soft errors." Soft errors are
characterized by the disruption of the logic state caused by the passage of radiation in
nodes or memory arrays. Soft error rates increase with increasing circuit density. Since
the trend is to increase circuit density, the soft error phenomenon may become more
significant.
Radioactivity can also adversely affect the performance of products such as
photographic film and components designed to measure the presence of radioactivity.
Highly sensitive photographic films, such as those used by the art and medical fields or
industrial radiography, are prone to damage by the presence of radioactivity. One effect
is creation of a fog or shadows. The longer the film is exposed, the more pronounced
this effect becomes. Radioactivity that raises background count-rates to higher levels
could affect the performance of radiation monitoring systems and analytical equipment.
Higher background count-rates would lead to reduced sensitivity and lower resolution
in spectroscopic systems.
The computer, photographic, and radiation measurement industries have taken steps to
minimize the impact of residual radioactivity on their products. These steps include
monitoring manufacturing processes, specifying material acceptance standards, and
screening suppliers. As RSM is recycled, these steps may become more important and
more costly. This paper characterizes potentially impacted industries and vulnerability
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JAERI-Conf 95-015
and effects due to the presence of residual radioactivity. Finally, the paper describes
practices used to limit the impact of residual radioactivity.
COMPUTER AND ELECTRONICS INDUSTRY
Overview of the Computer and Electronics Industry
In 1988, over 45 million computers and 27 million computer-related components,
including printers, plotters, monitors, and modems, were in use in the United States.
Over 33 million homes have at least one computer. Sales of computers used in business
and industrial applications exceeded 10,8 million units in 1991. Businesses use about
70 percent of all computers sold. The balance consists of science and technology users
(15 percent), educational users (10 percent), and other users (5 percent).
Electronic equipment includes individual components and subassemblies made both in
the United States and in foreign countries. Circuit boards hold electronic chips, e.g.,
central processing unit, memory, video display, disk controllers, and data input/output
devices. Internal subassemblies contain a variety of discrete parts, including solid state
products and passive components, such as resistors, capacitors, and connectors.
Materials Usage in the Computer and Electronics Industry
Computers and components contain a broad range of materials. Large components such
as casings and frameworks usually consist of steel and aluminum or plastic.
Connectors, wires, and chip sockets are made of various metals and alloys consisting of
tin, nickel, gold, silver, lead, copper, chromium, cadmium, bronze, and brass. The
circuits are a very thin film, typically a fraction of a mil thick. Certain metals are used
to make electrical leads (e.g., nickel, gold), while other metals are functional elements
(e.g., tantalum in capacitors). Peripheral equipment such as printers or video monitors
have the same types of components used in computers.
Computer chips consist primarily of silicon, silicon oxide, and aluminum, applied in
layers or regions. The n-regions are made with phosphorous diffused in acid etched
areas and the silicon substrate forms the p-regions. In a process called metallization,
aluminum from a sputtering target is vacuum-deposited as a film over the various n- and
p-regions to form the chip's internal circuitry. Leads out of the chips are made of tin
and nickel alloys or gold. The metallization layer may also be coated with a protective
film made of nitrides and oxides. A plastic or ceramic package encases the entire chip.
Ceramic packaging is used for about two percent of the total chip production. Metallic
cladding is no longer used as an outer package.
The amount of materials used in a typical chip are minuscule. A typical chip contains a
square piece of silicon (called a die) with dimensions of 0.25 x 0.25 inches and 0.015
inches thick. The metallization film forming the electronic circuit is about 1 urn thick.
Assuming total die area coverage, such a die contains 0.04 grams of silicon and 10
ugrams of aluminum film.
_ 176 _
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JAERI-Conf 95-015
The lead frames making the connections from the metallization film to the external
connectors consist of very thin, hairlike wires or metal strips. Table 1 presents typical
compositions of lead frames and other components (POW 93).
Table 1. Compositions of Lead Frames and Other Components
•"^yWxv:- :•:-:- ^:i:S:::::';"i:;:"v:-; •*.
': :.:::x -::.':-,-'.--,-,-r-:v.-r-:-.-:-:--; ";'
•:- ; :-;•:- '-;-:-; •" :-:•: ;-,-"•>:-; -',->:-. '--_
. . - -.-,- .-.•--,- v.-.v.-.v.- .•-• :- :-
:: -:-:•>: :-;-»i-: :-••;.:-'-:•> :x-: -i-
'•- '''Jf'-i- "'T^(SVfltT"'"K:''^: '~'">
Fe 54%
Co 17%
Ni 29%
;>•*?•¥ Si;?; S' SxixSg «;
Fe68%
Ni 42%
Cu 97.5%
Fe 2.35%
ZaO.12%
P 0.03%
11P:1IB1I
Cu 99.96%
lillililiill
Cu 10%
W 90%
1 |||l|||l@en
1 l|S|r||]|a|| f
BeO 99.5%
Cu 0.5%
AJ,0,
100%
Kovar, Alloy-42, Alloy-194, and the other metallic alloys are primarily used in making
lead frames and heat dissipating devices. The beryllia ceramic is a heat sink, while
A12O3 ceramics serve as packaging. External leads comprise most of the bulk weight of
metals contained in finished integrated chips.
Table 2 presents assumed quantities of materials used by the electrical and electronic
industries, as a whole. The majority of the materials listed in Table 2 are used to make
components and parts, such as housings, instrument racks, and frames. Computer chips
require only a small percentage of these materials.
The aluminum used in metallization comes from special suppliers and is purified and
prepared to customer specifications (POW 93; COM 93). Sputtering targets contain
small amounts of aluminum (about two to six kg) and provide enough aluminum to last
for several months' worth of production. Sputtering targets are usually replaced when
they no longer meet production criteria (uneven deposition of aluminum on silicon die),
not because of aluminum depletion.
Table 2. Estimates of Primary Resources Used in Electronics and
Electrical Industries* -1990 and 1991 Data
Metal
Al
Sn
Ni
Pb
Cu
Cd
Be
Ta
Primary Use"
Electrical
Solder
Nonferrous
Solder
Electrical
Batteries
Electronics
Electronics
Percent of Total'"'.
U.S. Consumption
8
6.7
17
1.3
24
50
65
60
-:- Estimated Use - '&:,>
Quantity & Units v';"-
952 MTxlO3
2.9 MTxlO3
21.7 MTxlO3
14.8 MTxlO1
670 MTxlO3
1.6 MTxlO3
3.2 MTxlO1
223 MT
;:-':;" Cost "($)& Units""
0.50- 0.6 1/lb .
2.49 - 3.43/lb
2.39 - 3.72/lb
0.32 - 0.40/lb
0.94- 1.15/lb
1.90-2.05/lb
160/lb
27.5 - 29.0/lb
- 177-
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JABRI^Conf 95-015
Pt
Au
Ag
Electronics
Industrial
Electronics
13
19
21
3.9 MT
701 ounce x 103
29 ounce x 106
342 - 379/ounce
332 -36 I/ounce
3.63 - 4.32/ounce
1 Data extracted from AMM 93.
b Category assumed to best characterize industry practices. Unless otherwise noted, electrical, as a
category, includes electronics.
c Prices represent the cost of the feedstocks, finished product, or resources. The cost for recycled
products or recovered metals could be very different from those tabulated here,
Impacts of RSM on the Computer and Electronics Industries
Over the past several decades, the Department of Defense, NASA, and the electronics
industry have assessed equipment reliability in the presence of radiation. Other studies
involve the qualification of equipment for use at nuclear power plants and particle
accelerators. Essentially all of the studies focus on the effects of external radiation
(gamma and x-rays, neutrons, and charged particles). Collectively, damage to
semiconductors, electronic systems, and passive components is characterized by
overlapping doses, ranging from one kilorad to one gigarad. Other studies evaluated the
indirect effects of radiation exposures, such as the generation or liberation of hydrogen
within the chip and materials displacement or responses (e.g., scission, polymerization,
and embrittlement).
The industry has not evaluated to the same extent the impacts associated with materials
containing residual amounts of radioactivity. The computer and electronics industries
have already experienced problems with elevated levels of naturally occurring alpha
particle emitters in chip packages and processing materials (MAY 79). Elevated levels
of uranium (2.5 - 17 ppm or 1.7 -11.5 pCi/g) and thorium (3-6 ppm or 0.33 - 0.66
pCi/g) were found in A1203 ceramics with zirconia and clay fillers (MAY 79). The
corresponding average surface alpha emission rates ranged from 5.2 to 45 particles per
cm2-hour, due to both uranium and thorium. Normally, alpha particles are easily
attenuated because of their mass and high charge. However, when uranium and thorium
are directly in contact with the metallization and oxide layers, alpha particles irradiate
the internal circuitry of the chip, with little or no attenuation. Furthermore, all the alpha
particle energy is deposited in a very short distance, maximizing the deleterious effect
on the chip circuitry.
For damage to occur, the radioactivity must deliver energy to sensitive areas or
components. The alpha particle creates a trail of charged ion-pairs as it passes through
the chip's cell. In turn, the ion-pairs neutralize the cell charge. The range of alpha
particles in silicon is about 25 micrometers (|im), disregarding the presence of the
aluminum metallization layer. Assuming a track length and diameter of 25 and I um,
respectively, a 5 MeV alpha particle will generate 1.3 x 10s ion pairs and 6.6 x 104 ion
pairs per ^m3. Energy deposition of about 102 to 103 electron volts (eV) per um3 is a
threshold above which performance is affected.
A 5 MeV alpha particle can deposit as much as 225 femtocoulomb (fC) in silicon,
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JAERI-Conf 95-015
assuming minimal attenuation (BUE 90). Thisxharge is about 10 times more than that
needed (about 20 to 80 fC) to cause an upset (SAI 82). The presence of a charge
determines the logic state of a transistor, a logic state of "1" for a differential voltage
and "0" for none. A logic state change from a" 1" to a "0" or "0" to " 1" state affects the
performance of the transistor.
These effects, called "soft-errors," "single event upsets," "iatch-ups," and "single event
transients," are collectively known as "single event phenomena." Since the single event
phenomena process is totally random, its occurrence does not depend on the location of
a cell within an integrated circuit and one cell is not more susceptible than another.
Soft-errors do not induce any permanent physical damage to the silicon structure and
are corrected during the next write-cycle. Latch-ups, on the other hand, require the
system to be reset to clear the problem and can induce permanent damages.
The sensitivity of a semiconductor to alpha particles is dependent upon several factors,
including the following:
• Semiconductor operating parameters, e.g., voltages, clock speed,
temperature, read/write frequency cycles, etc.
• Surface area of the die, i.e., area of chip with transistors, amplifier nodes,
memory cells, etc.
• Alpha particle energy
» Relationship between source of alpha particles and sensitive portion of the
die, i.e., incident angle
» Alpha particle flux at the surface of the die
• Presence of mitigating measures, such as protective films and thicknesses
• Charge dissipation, e.g., drift, diffusion
Newer semiconductor designs might be even more sensitive as they incorporate much
higher transistor densities, e.g., the recently released Pentium holds 3.1 million
transistors on a 16.2 x 16.2 mm chip and DRAMs typically have 16 million transistors.
Such designs are much more susceptible to soft-errors given that the travel path of a
particle is more likely to interact (and adversely affect) a greater number of transistors.
As computer designs are also changing to lower operating voltages (from 5.0 to 3.3 V),
systems operating at a reduced voltage are more susceptible to ionizing radiation
because charge differentials between a logical state of "1" and "0" are lower still.
Together, these factors make computers relying on newer technology increasingly
susceptible to the presence of radiation. The current strategy incorporates soft-error
correction circuitry or software to verify DRAM read-write cycles. The use of such
correction methods, however, is limited to simpler DRAM applications, such as data
storage and retrieval, rather than in complex applications in DRAM logic circuitry, such
as monitoring aircraft systems status. In complex applications, the use of self-
correcting software is not feasible as the logic state is unique in the context of the data
being processed.
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Computer and Electronics Industry Practices and Safeguards
The increased potential for the introduction of radioactivity into integrated circuits is a
concern to the electronics industry. Individual companies are conducting studies to
assess the impact of radioactivity on specific components by focusing on methods to
make chips less sensitive to radioactivity and reducing the associated soft-error rates.
The problem of alpha contamination in ceramics has been "engineered out" by
manufacturing changes and selecting new ceramics suppliers. Work has been done in
response to other types of radioactive contaminants, i.e., beta and gamma emitters, by
introducing packaging materials with low- and high-Z (atomic number) materials.
Low-Z materials are used to stop beta particles and high-Z materials are used to
attenuate gamma or x-rays. The industry is also assessing methods to further reduce the
presence of uranium and thorium in sputtering targets by developing new aluminum
refining techniques (COM 93).
There are currently no specific standards for radioactivity in integrated circuits.
However, the industry is beginning to address the issue by establishing a standard. A
few standards, namely ASTM and MEL-STD, address only testing procedures to
determine total dose, dose rate induced upsets, response, aging, memory degradation,
induced displacement damage in silicon, semiconductor recovery and annealing, and
estimating alpha particle induced soft-error rates (MIL 91; ASTM 88a, 88b). Vendors
supplying the industry have applied "limits" based on the results of studies conducted
by individual member companies. These "limits" have not been officially promulgated
as standards (e.g., with ASTM, JEDEC, ANSI, ISO). The "limits" often cited in
technical papers include:
W&;JIZ%$i.y ^:;--:DesigiiBti6n'::p::"':;^:^i^:H.'^
Surface alpha activity emission rate
Concentration (U-238)
Purity grade
:ittMi&K-!^tv-'--. " Criteria ;r;^5:;-y?.|::.-::.:-fl;|?:f:::i:o.'
s 0.001 alpha particles per cm2/h
<; lOppb (-0.003 pCi/g)
i 99.999% purity
(Sources: RIL 81; ML 82; MAY 79}
The "limits" listed above are applicable to materials used in making chip packages,
primarily for ceramics and fillers used with ceramics. Specific "limits" may be
similarly developed for other types of media. Such "limits" have yet to be identified or
published.
THE PHOTOGRAPHIC FILM INDUSTRY
Overview of the Photographic Film Industry
Besides photographic uses, films meet a broad range of applications, including those of
the medical, industrial, graphics, cinematography, and radiation dosimetry
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communities. Furthermore, for each application, various grades of films exist to meet
specific needs. The number of still photographs used has steadily increased over the
past decade, from 12.8 billion exposures in 1982 to 20.4 billion in 1991. In 1991, the
United States alone accounted for nearly 38 percent of all worldwide still photographic
exposures. Of this total, instant photography comprises a small fraction of conventional
films, 4.4 percent in the United States and 3.5 percent worldwide for 1991. In the
United States, recreational or amateur uses account for about 80 percent of film sales.
Professional applications in scientific and industrial sectors and the arts account for the
remaining sales. The sale of conventional films totaled nearly 850 million rolls in 1990,
with the 24-exposure roll accounting for about 70 percent of sales. The 35-mm format
controls the film market, accounting for 85 percent of the sales in 1992. The balance of
sales is associated with disc, sheet, and other formats, including films in instant
cameras.
Materials Usage in the Photographic Industry
The photographic industry typically uses about 50 percent of the United States
industrial silver production. From 1985 to 1990, the annual consumption was stable,
averaging 58.9 million troy ounces (a troy ounce equals 32.1 grams) (AMM 93). In
1991, the United States used 66 million troy ounces of silver (WOL 92), Silver is also
recovered from used photographic films and spent processing solutions. Between 1986
and 1992, the annual recycling rate of 999-grade silver ranged from 42.2 to 49.1 million
troy ounces, averaging 44.6 million troy ounces (AMM 93).
Currently, the only packaging that includes metal is the 35-mm film cassette. A typical
35-mm cassette consists of three parts, the main body that is a single piece of soft steel
(7.5 cm x 2.8 cm and 0.3 mm thick), and two stamped sides, about 2.5 cm in diameter,
with 1.2 cm holes to hold the spool on which the film is rolled. The main body forms a
right cylinder with the two sides crimped in place. The total weight of the metal
cassette alone is about seven grams. As was noted earlier, 850 million rolls were used
in 1990, with the 35-mm format making up 85 percent of all of the film sold. Based on
this information, the photographic industry is assumed to use annually about 5,400
metric tons of steel.
Currently available data does not provide the means to estimate the amounts of steel
and alloy products used to make film processing and manufacturing equipment.
Manufacturers will not release such information because they consider it proprietary.
Impacts of RSM on the Photographic Industry
Silver halide crystals are the sensitive elements of the photographic emulsion. When
exposed to light or ionizing radiation, halide grains inject electrons into conduction
bands where they become trapped, forming neutral silver atoms. When enough neutral
silver atoms form, a stable latent image results. A photon of visible light will only
liberate one electron. To fully develop one silver grain requires successive photons.
On the other hand, a single ionization event can achieve the same effect, as radiation
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imparts thousands of times more energy than light. Depending on how the radioactivity
is distributed, exposure to ionizing radiation may cause uneven latent image centers,
resulting in a nonlinear film response. This response, called a high or low intensity
reciprocity failure, depends upon the rate of exposure. Reciprocity refers to the
relationship between optical density and exposure rate. Major differences between
ionizing radiation and visible light exposures are that:
• radiation delivers far more energy to the silver crystals than light does
and,
• latent images are caused not by the radiation, but by the electrons
liberated in its path.
Film response also depends on the energy of the incident radiation whether electronic
equilibrium occurs during film exposure. This dependence is due to the energy
absorption characteristics of silver halide crystals and electronic equilibrium. This
process is dependent upon the electronic structure (shell electron densities) of the
material with which the radiation interacts. High energy particles or rays are more
effective in releasing electrons. The response is most pronounced below 0.1 million
electron volts (MeV), typically 5 to 30 times that at 0.6 MeV. At lower energies, the
attenuation properties of the emulsion govern the response and at higher energies, the
response is relatively flat from 0.3 to 10 MeV.
The presence of residual amounts of radioactivity in manufacturing and processing
equipment can have detrimental impacts on film quality, as residual radioactivity could
also be present in the metal making up the components of film assembly lines. These
extraneous exposures may appear as diffuse shadows, general fogging, or result in the
projection of images of parts or components. These effects can vary depending upon
where the radioactivity enters the film making equipment (e.g., evenly distributed as
surface or volums contamination or present as randomly located hot-spots).
Metals contained in camera bodies potentially present a source of radiation. Most
cameras are made of aluminum and steel. The exposure would occur while the film is
loaded in the camera. Another concern is whether film packaging material (metal
cassettes) could contain radioactivity from recycled scrap metal. The radioactivity
contained in the steel could expose the film, which from packaging to processing could
be as much as two years.
Practices and Safeguards in the Photographic Industry
The photographic industry has well-established procedures to routinely monitor the
presence of radioactivity during all manufacturing stages. These procedures reflect
several decades of operating experience and practices developed when environmental
radioactivity was at its highest during atmospheric testing of nuclear weapons (over 400
tests were conducted from 1945 to 1980). More recently, the industry has focused on
evaluating potential impacts associated with environmental radiation from reactor
accidents (e.g., Chernobyl in 1986) and the inadvertent introduction of man-made
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sources.
Routine surveillance includes monitoring airborne and waterbome radioactivity in
plants and process streams. Currently air and water are filtered before being introduced
or used in manufacturing processes or plants. The presence of radioactivity is also
routinely monitored in wood pulp used to make printing paper and emulsion materials.
Any radioactivity detected above statistically defined limits, based on background
levels, is evaluated to determine if the amounts or concentrations could have an adverse
impact on film or paper products. A review of earlier industry practices (early 1960s),
however, does provide some insight about radioactivity thresholds (NUS 62). The
threshold values noted below represent activity (gross beta) levels, believed to cause
some impact on film products.
Raw Water (Insoluble Materials) 1,5 pCi/L
Air (Particulates) 0.1 pCi/m*
These limits reflect film performance characteristics of the early 1960s. Since that time,
film sensitivity has increased several fold, especially with newer high speed films.
The photographic industry has not formally published standards addressing residual
radioactivity, specifications, or requirements that specifically prevent the introduction
of radioactivity into photographic components and materials (FAA 92; ASTM 91; MCI
93; ANSI 83, 80a). However, it is clear that individual companies have developed their
own internal standards.
The photographic industry has evaluated sources of radiation that could affect a film
product after its sale. As an example, the industry actively evaluates the performance of
airport x-ray units used to inspect carry-on luggage. The industry is concerned that the
new high speed films could be damaged by these x-ray units. The industry, in concert
with the Federal Aviation Administration (FAA), has successfully recommended
radiation exposure limits for airport x-ray inspection units (FAA 92). FAA regulations,
under 14 CFR 108.17, now limit radiation exposures from x-ray systems to one
milliroentgen during inspections.
THE MEASUREMENT AND ANALYTICAL INDUSTRY
Overview of the Measurement and Analytical Industry
The measurement and analytical equipment industry includes a broad range of
instrumentation divided into two major categories, general instrumentation, and
radiation monitoring or analytical equipment. General instrumentation comprises
electronic components (such as chips, circuit boards, etc.), and electro-mechanical
devices (such as relays, solenoids, motors, etc.). In many industrial applications,
radiation monitoring or radio-analytical equipment function with other types of
instrumentation. Both rely on the extensive use of computers. Some computer designs
are for specific applications, while others use off-the-shelf systems. In both cases, the
range of electronic components is essentially identical to that found in computers and
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related peripheral equipment.
Radiation measurement and analytical equipment also include a broad range of
instruments, from portable survey meters to complex laboratory and industrial systems.
Portable survey meters vary in complexity from simple analog to more complex digital
circuitry. The major components of portable survey meters include a detector, a high
voltage power supply, pre- and full-amplifiers, count-rate circuitry, and readout
displays.
Other systems exist in a broad array of applications, including the production of radio-
pharmaceuticals, food processing, industrial process control, scientific research and
development, and educational fields. Many laboratory radioanalytical systems rely on
lead shielding to minimize the influence of background radiation from terrestrial or
industrial sources. Typically, the detector is placed in a shielded container made of an
inner core of lead, clad in a steel jacket. Consequently, the use of even very slightly
contaminated lead and steel might affect analytical results, capability to resolve low
levels of radioactivity, and identify radionuclides.
The health care industry also makes extensive use of electronic equipment, for patient
monitoring, diagnostics, x-ray and nuclear imaging, irradiation equipment, and life and
surgical support systems,
Materials Usage in the Measurement and Analytical Industry
As was noted earlier, the same types of components and materials are used by the
measurement and analytical industry, and the electronics industry. However, while
both industries use aluminum, the radiation measurement and analytical equipment
industry uses lead, steel, and other alloys.
Manufacturers use light gauge steel and aluminum to make equipment cabinets and
enclosures. Stainless steel and aluminum are used to make discrete components such as
detector holders and fixtures. Some steel predates World War II ensuring that it is free
of long-lived radionuclides from weapons testing fallout. Pre-WW II steel is used to
make shielding components for radiation detection systems with very high sensitivity.
Lead is used in the form of bricks, sheets, and machined components. Lead shielding
consists of 4 to 6 percent antimony. Other lead alloying elements include copper,
silver, cadmium, and indium. Counting shields contain an inner core of lead and are
clad in steel. The lead is poured into shields or machined to meet specific design needs
(LIA 85). In 1990, nuclear applications, including shielding, required the use of about
13,000 metric tons of lead. Most of the lead is used in sheets and cast products use the
majority of lead.
Table 3 presents typical metal usage by four radiation instrumentation manufacturers.
Individual manufacturers of radiation monitoring equipment do not publish, for
proprietary reasons, the amount of lead they use. The same is true for other metals,
such as steel, aluminum, and stainless steel.
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Table 3. Metals Usage by Radiation Monitoring Equipment and Shielding
Manufacturers and Suppliers*
• •:•,• •:•.-:•:-:•:•.•; :,-.•,-. .•>.:•;-,:•:•' ,;->.•;-,•;•.-,-. ;-:•;• ;-,•:•'-:-"•'-,- ••••
• :•••:• •»••:•:•:-.-:• •:-,-:•;-.•:•>:-:•:•;-:•:-;-;•;-;-;•>;-;•>,-:•>;-;•;-:•;••-•-•••-;-•.
:•:•••:•:•'• :oX-: •::;'. .;;-,•:•>:•:•>:-:-; :;.•:;••:•:•;•;••;•;•;•:•;•;-;•;•;-:•:•:-:•:;-;-
x-"x; ;>:;>. ":x-;":-''x""1'i::;'-:::::$':'-:;;4:J;;;:"~;'': •'•'"'"••'•' -'''^ •••'-•'•
|
Lead and Alloys
Steel and Alloys
Al and Alloys
Al/Fe/Stainless
Pre-WW II Steel
~£?£';::"-i:::;: ;-:x::::::^xx::K:,x'^x?"Xvl.:^
••-:-:•-,-:•:•;:•: •--v.*.x^^x:-.-:^^x-xo:•.-^:^'::-:: '::::r:-'::-:;''::i':':-:-:-;':-:::-.-:'".-;':'.-:'; :
-I;-;.,;:';':-1;- ';-;:->-;-•' - -x- X- :•:;:• :v,vX_;;:;:v:;
15
—
—
_
..
„
-
-
30
~
90
5
3
—
—
* Information obtained via an informal telephone survey (May/June 1993), Data supplied
by Ludlum, Sweetwater, TX; Canberra, Meriden, CT; Majestic Metals, Denver, CO; and
J.L. Shepherd, San Fernando, CA.
The data indicate that the radiation equipment manufacturing industry uses relatively
small amounts of steel, aluminum, and other alloys. For comparison, U.S. steel mills
and smelters produced, in 1992, four million metric tons of aluminum (primary) and 83
million metric tons of raw steel (AMM 93),
Impacts of RSM on the Measurement and Analytical Industry
Modern instrumentation and analytical systems rely heavily on computers and related
peripheral components. Some types of instrumentation also use electro-mechanical
devices to control process equipment, gather data from external sensors, trip alarms,
actuate servos, etc. As was noted earlier, integrated circuits, and passive and active
components, have different response characteristics to the presence of radioactivity.
Typically, electro-mechanical systems and passive components (e.g., resistors,
switches, motors, relays, etc.) are relatively immune to external radiation, unless
exposed to very high levels, on the order of several megarads.
The relative immunity of electro-mechanical systems to external radiation means that
the discussion addressing the sensitivity of computer equipment should be equally valid
for general instrumentation and analytical equipment. One exception is analytical
equipment and instrumentation used to measure the presence of radioactivity in samples
or characterize radiation fields.
The introduction of radioactivity in materials incorporated in radiation monitoring and
radio-analytical systems could have major impacts on their performance. Radioactivity
could be present in materials making up nonessential peripheral systems (e.g., racks or
frames), or in critical components such as detector housings and shielding.
Radioactivity in detector housings and shielding materials would have the most severe
impact. Other components (e.g., pre-amplifiers, cold fingers, and Dewars) that directly
connect to, or that are collocated near the detector, would also present problems if
contaminated with radioactivity.
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Practices and Safeguards in the Measurement and Analytical Equipment Industry
Several standards and specifications address testing procedures and performance
characteristics of radiation monitoring and analytical equipment, including the presence
of radioactivity in detectors and shielding materials (ANSI 91). However, the standards
do not provide any numerical criteria for allowable levels of radioactivity. For
example, ANSI N42-14 (ANSI 91) notes that shield materials should be made of
radiation free, or low activity, lead or steel and does not define the terms "radiation
free," and "low activity." The intent of the standard is that radioactivity in materials
should be low enough that the component fabricated from these materials can perform
as intended. The ANSI N42-14 standard also states that steel and aluminum should be
free of natural thorium and uranium, and potassium-40, cobalt-60, and cesium-137.
Similarly, ANSI N42-14 states that structural materials used to manufacture pre-
amplifier casings and "cold-fingers" should also be radiation free. Furthermore, the
standard also recommends that in some instances components known to have elevated
radioactivity levels be installed outside of the counting shield, or be shielded
themselves to protect the detector from extraneous radiation.
If a completely new system were purchased, the initial setup and calibration procedures
might provide the opportunity to identify the presence of residual radioactivity.
However, the detection of radioactivity in peripheral equipment would be more
difficult, unless specific survey procedures were initiated.
In addition, manufacturers perform operational checks before shipping systems to
customers. As part of this investigation, the manufacturer can trace the source of
radioactivity to its suppliers and manufacturers. Accordingly, it is expected that the
presence of radioactivity would ultimately be detected.
CONCLUSIONS AND OBSERVATIONS
The introduction of residual radioactivity in commercial materials, metals in particular,
could have negative effects, depending upon the industry. Ultimately, the potential
impacts are dependent upon how the radioactivity is incorporated in products, the
amount of radioactivity, and the radio-sensitivity of the products.
External penetrating radiation does not begin to affect products until exposure rates or
doses are in the thousands of rads. This is the case for most electronic products and
systems, such as computers, integrated circuits, and passive components, including
those that are not radiation-hardened. However, electronics industry representatives
have pointed out the need to evaluate all aspects of a system. In other words, limits
designed to protect 1C chips might not be adequate to protect other peripheral
components. For example, the dose from external radiation might be controlling when
compared to self-irradiation for a component that is also internally contaminated.
Another potential impact involves internal contamination and the interference of
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radioactivity when materials contaminated with residual radioactivity become integral
parts of a product. The presence of radioactivity in materials used to make or package
films, electronic components, and radiation detectors and shields could have severe
impacts. For example, photographic films might fog, radiation monitoring and
analytical systems' detection limits and resolution might degrade, and integrated circuit
chips might experience various types of single event upsets. Besides the amounts of
radioactivity, the severity of these effects would be dependent upon the types of
emissions (e.g., alpha, beta, gamma and x-rays), distance from the source of
radioactivity to radio-sensitive components, inherent internal attenuation, and duration
of exposures.
The few examples gleaned from the literature on residual radioactivity provide only
limited technical insights about isolated or indirectly related cases. From this
information, it is not possible to scale up a general characterization of each industry and
potential cost implications. The literature implies some companies have instituted new
requirements (not published), but it is not possible to determine if such standards reflect
isolated cases or new industry-wide practices in recognition of a problem.
In some instances, apparent "limits" have been proposed by the industry, but it is not
possible to determine if they incorporate the benefit of peer reviews. Also, it is not
possible to determine if these apparent "limits" reflect a point, from production to the
end-user, at which products appear to be the most vulnerable to damage from residual
radioactivity. Other than assessing the effects of external penetrating radiation, existing
standards and specifications are silent on the presence of radioactivity in materials used
by the electronics industry. Further constraints in making this information available
involve proprietary matters, market competitions and, possibly, product liability issues.
The available information provides no means to thoroughly assess other impacts, such
as cost of lost products, domestic and international market disruptions, and the
implementation of corrective actions required to account for residual radioactivity.
Such actions might include products or materials monitoring, implementing new
manufacturing processes, revising procurement and specification procedures for
materials and equipment, and developing new standards.
There is enough circumstantial information, however, to determine that the introduction
of radioactivity might have detrimental impacts. Radioactivity at naturally occurring
concentrations is known to cause detrimental effects on the performance of
semiconductors. Similarly, the photographic industry is already monitoring air, water,
and feedstock materials used in manufacturing film and paper products. The
manufacturers of radiation monitoring and analytical equipment are also selecting
materials and products with the lowest levels of radioactivity.
These actions are taken to protect state-of-the-art technology, however, current trends
indicate future products or equipments might become even more sensitive to the
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presence of residual radioactivity. Accordingly, a RSM recycling standard
implemented to protect existing products or systems might prove inadequate for future
technologies. The past few decades have revealed a tremendous growth in technologies
with intense international competition. Indications are this trend will continue. In
effect, the rapid succession of technological breakthroughs makes it very difficult to set
standards that would remain valid over the long-term. In part, for this reason trade
organization representatives are objecting to the practice of recycling radioactive scrap
metals. In addition, U.S.-based companies may be required to obtain materials and
products from foreign countries that do recycle radioactive scrap metals or impose
much more restrictive specifications to avoid it. In either case, such impacts couid
affect the competitive edge of U.S. companies.
Standards being developed by other countries and international agencies are proposing
residual radioactivity levels that are significantly above those levels causing problems
for the electronic and photographic industries. For example, the Commission of
European Communities published a limit of 27 pCi/g (1 Bq/g) for alpha, beta, and
gamma contamination (CEC 88). The International Atomic Energy Agency is
proposing a limit of 8 pCi/g (0.3 Bq/g) (IAEA 93, 88). The limits might prove
detrimental to the industries of concern, especially if recycling were practiced on a large
scale. Since the United States has very large potential quantities of recyclable RSM, the
possibility of future problems is so much the greater. These limits also present a
challenge to these industries as many companies have manufacturing plants in foreign
countries.
Predictions of thresholds, below which radiation exposures or presence of radioactivity
cease to have impacts on product quality or performance, are impossible to define given
limited information. The trade organizations that were contacted were not able or
willing to provide specific details describing or characterizing potential impacts on the
electronics, photographic, and instrumentation industries.
We must assume that residual radioactivity in all materials will be increasingly
problematic to these industries. As technological developments lead to further
miniaturization, greater speed, greater sensitivity, and greater economic potential for the
United States, our best interest would be better served by proceeding with extreme
caution when considering the deliberate introduction of residual radioactivity.
REFERENCES
AMM 93 "Metal Statistics - 1993," American Metal Market, 85th edition, New
York, NY.
ANSI 91 "American National Standard Calibration and Use of Germanium
Spectrometers for the Measurement of Gamma Ray Emissions Rates of
Radionuclides," ANSI N42.14-1991, New York, NY.
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ANSI 83 "American National Standard for Protection - Photographic Film
Dosimeters - Criteria for Performance," ANSI N13.7-1983, New York,
NY, 1983.
ANSI 80a "American National Standard Dimensions for Paperboard Cores for
Radiation Sensitive Photographic Films," ANSI PH1.62-1980, New
York, NY, 1980a.
ASTM 88a "Standard Guide for the Radiation Testing of Semiconductor Memories,"
American Society for Testing and Materials, Fl 191-88, Philadelphia,
PA, 1988a.
ASTM 88b "Standard Guide for Ionizing Radiation Effects Testing of
Semiconductor Devices," American Society for Testing and Materials,
F867-88, Philadelphia, PA, 1988b.
ASTM 91 "Standard Test Method for Determining Low-Level X-Radiation
Sensitivity of Photographic Films," American Society for Testing and
Materials, F947-85, reapproved 1991, Philadelphia, PA., 1991.
BUE 90 "Alpha Particle Sensitive Test SRAMs," M.G. Buehler, et al., IEEE
Transactions on Nuclear Science, Vol. 37, No. 6, December 1990.
CEC 88 Commission of the European Communities, Radiation Protection No.
43: Radiological Protection Criteria for the Recycling of Materials
From the Dismantling of Ntf clear Installations, November 1988.
COM 93 "La Haute Technology, Perspective/Technology," R. Comerford, IEEE
Spectrum, pp. 55-60, March 1993.
FAA 92 "Code of Federal Regulations," Title 14, Part 108.17, Use of X-Ray
Systems, Federal Aviation Administration, 1/1/92 edition.
IAEA 93 "Exemption from Regulatory Control: Recommended Unconditional
Clearance Levels for Solid Materials Incorporating Radionuclides,"
International Atomic Energy Agency, Draft Working Document, Vienna,
March 1993.
IAEA 88 "Principles for the Exemption of Radiation Sources and Practices From
Regulatory Control," International Atomic Energy Agency, Safety Series
No. 89, 1988.
LIA 85 "Lead Shielding for the Nuclear Industry," Lead Industries Association,
Inc., New York, NY, 12/1985.
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MAY 79 "Alpha Particle Induced Soft-Errors in Dynamic Memories," T.C. May
and M.H. Woods, IEEE Transactions on Electron Devices, Vol. ED-26,
No. 1, January 1979.
MCI 93 "Safety Considerations for Photographic Film in the X-Ray Screening of
Air Travelers' Luggage," V.G. Mclninch, Journal of Testing and
Evaluation, American Society for Testing and Materials, November
1993, pp. 368-375, reprint, Philadelphia, PA.
MIL 91 Test Method and Procedures for Microelectronics, Test Methods No.
1017.2, 1019.4, 1020.1, 1021,2, 1023.1, 1032.1, Military Standard 883D,
November 15, 1991, Defense Printing Service, Department of the Navy,
Philadelphia, PA.
NUS 62 "The Significance to the Photographic Industry of Radiation and
Radioactive Materials in the Environment," Nuclear Utility Services,
prepared for the Office of Atomic Development, State of New York,
NUS-106, March 1962.
POW 93 Texas Instruments, Dallas, TX, T. Powell, letter to Mr. J-C. Dehmel,
SC&A, Inc., June 14, 1993.
RIL 82 "Ultrahigh Sensitivity Uranium Analyses Using Fission Track Counting:
Further Analyses of Semiconductor Packaging Materials," J.E. Riley, Jr.,
Journal of Radioanafytical Chemistry, Vol. 72, No. 1-2, p. 89-99, 1982.
RIL 81 "Determining Trace Uranium in Ceramic Memory Packages Using
Neutron Activation with Fission Track Counting," J.E. Riley, Jr.,
Semiconductor International, p. 109-120 May 1981.
SAI 82 "Alpha Particle Induced Soft Error Rates in VLSI Circuits," G.A. Sai-
Halasz et al, IEEE Transactions on Electron Devices, V. ED-29, No. 4,
April 1992.
WOL 92 " 1991 -1992 Wolfman Report on the Photographic and Imaging Industry
in the United States," L, Wolfman, Popular Photography Magazine,
New York, NY, 1992.
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JAERI-Conf 95-015
3-6
A Methodology For Estimating Potential Doses And RisksFrom
Recvcling U.S. Department Of Energy Radioactive Scrap Metals
by
John A, MacKinney
Environmental Scientist
Radiation Studies Division
U.S. Environmental Protection Agency
ABSTRACT: The U.S. Environmental Protection Agency (1PA) is
considering writing regulations for the controlled use of
materials originating from radioactively contaminated zones which
may be recyclable. These materials include metals, such as steel
(carbon and stainless), nickel, copper, aluminum and lead, from
the decommissioning of federal, and non-federal facilities. To
develop criteria for the release of such materials, a risk
analysis of all potential exposure pathways should be conducted.
These pathways include direct exposure to the recycled material
by the public and workers, both individual and collective, as
well as numerous other potential exposure pathways in the life of
the material. EPA has developed a risk assessment methodology
for estimating doses and risks associated with recycling
radioactive scrap metals. This methodolgy was applied to metal
belonging to the U.S. Department of Energy. This paper will
discuss the draft EPA risk assessment methodology as a tool for
estimating doses and risks from recycling.
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JAERI-Conf 95-015
INTRODUCTION
With the down-sizing of the United States' nuclear weapons
program, and the closure of commercial nuclear power stations,
large amounts of radioactively contaminated, or suspect,
materials are expected to be generated. These materials may be
comprised of old process equipment, piping, and structural metal
and concrete, from the dismantlement of surplus buildings and
structures in radioactively contaminated zones.
Materials originating in a controlled zone are generally
treated as contaminated, until they can be declared "clean."
Given the public sensitivity regarding radioactivity, however,
the question "how clean is clean?" is especially problematic.
The underlying question is how much risk, imposed or otherwise,
is acceptable, and how is that level of risk determined?
EPA is now addressing the problem of management and
disposition of wastes generated during site cleanup, including
the demolition of buildings and structures. No national
standards exist for the release and recycling of such materials
containing residual levels of radioactive contamination.
However, if some of these materials could be recycled, the burden
on waste disposal facilities would be lessened.
EPA has performed a study for the U.S. Department of Energy
(DOE) on the recycling of radioactive scrap metals (RSM)
belonging to the DOE. These metals come mostly from the
dismantlement of obsolete nuclear fuel facilities, weapons
development and testing facilities, research facilities, and
other operations where radioactive materials were used or
processed. The study, titled "Analysis of the Potential
Recycling of Department of Energy Radioactive Scrap Metal"
(draft, September, 1994)', was undertaken to provide decision-
making tools for DOE. A key component of the study is an
assessment of the potential risks from recycling DOE's metal.
EPA developed a methodolc v for evaluating doses and risks
from recycling RSM, and appliec it to DOE RSM. If appropriate,
EPA could further develop this r cycling risk assessment
methology for use in a future reg lat ~>ry effort.
EPA RISK METHODOLOGY
Since no national standards exist for recycling
radioactively contaminated materials, the EPA recycling study for
DOE was constrained from making any implications as to what an
appropriate recycling residual contamination level would be.
Instead, the methodology is intended to be a tool for determining
dose and risk for a set of scenarios by inputting the
concentrations of the radionuclides in question for a quantity of
metal. The study then estimated risks from the recycling of a
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specific quantity of DOS-owned steel with known radioactive
contaminants. All data given are draft data.
The study focused on carbon and stainless steel. A
recycling route for the steel was established, detailing the
likely path that the metal would take from the scrap yard,
through various recycling and manufacturing operations, to the
hands of the consumer in the form of a product. Doses (effective
dose equivalent) and risks were modeled at multiple points in the
scrap route, up to the consumer who purchases a product.
All exposures to workers and the public come from the three
products of the steel melting process: metal, slag, and offgas
emissions. Radionuclides partition fairly consistently among
these three phases. For example, most alpha-emitters partition
thermochemically to the slag phase, while Co-60, Ni-63, and Tc-99
tend to stay in the molten metal. Others, such as Cs-137 and
isotopes of polonium and lead, volatilize readily and must be
captured in the offgas system. All major pathways were
considered in each scenario, including:
* direct exposure (workers and public)
* inhalation of contaminated dust (workers)
* plume dispersion and consequent inhalation (workers and
public)
* plume dispersion and consequent immersion (public)
* plume dispersion and consequent contaminated ground/soil
leading to external exposure and food pathway (public)
* ingestion of contaminated water from run-off (public)
* ingestion of food contaminated with dust, or hand-to-mouth
ingestion (workers)
* ingestion of food prepared in cook ware from recycled
metal (public)
The scenarios chosen were judged to represent significant
potential doses and risks, while still being probable.
Scenarios, or operations, represent points in the process that
include transportation workers, scrap yard workers, furnace
operations and casting, bag house and slag pile workers, product
manufacturing, and finished product users. (Table 3, at the end
of this paper, shows a listing of all the scenarios/operations.)
RSM decontamination operations were not modeled. These workers
would be radiation workers and subject to DOE exposure limits and
ALARA (as low as reasonably achievable) procedures.
All scrap metal in the study was assumed to be processed via
melting in an 80 ton electric arc furnace (EAF), or a basic
oxygen furnace (the EAF was considered to be more likely) .-
Melting is an effective means of decontamination for many
radionuclides, as well as producing an ingot or slab which is a
feedstock for manufacturing. An EAF was assumed to operate
continuously processing 110,000 MT of RSM in a year.
Initially, the modeling assumes a radionuclide concentration
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JAERI-Conf 95-015
of 1 picocurie (37 mBq) per gram in metal for each radionuclide
at each stage in the process. (The exceptions are airborne
releases from the furnace, and concentrations in the slag.) This
approach was chosen for the following reasons: the actual
contamination in the scrap is not well characterized (with some
exceptions); the melt decontamination partition coefficients are
not precisely known; the results would not depend on the degree
of dilution; and, doses and risks can easily be scaled up or down
for a given scenario, depending on the actual radionuclide
concentration.
Partition factors were used to determine the slag and offgas
emission concentrations. For example, Am-241 goes 99% to the
slag, with the remaining 1% left in the metal, while 90% of Mn-54
goes to the metal and 10% to the slag. Cs-137 varies from 0-50%
going to the slag, and 50-100% to the offgas (100% was assumed to
go to the offgas in the study).
Thus, doses and risks were assessed on a per picocurie per
gram basis for 46 radionuclides (or 59, including separate
calculations for those with short-lived progeny). The
radionuclides considered are listed in Table 1.
Individual and population doses and risks per picocurie per
gram were calculated for a total of 67 worker and consumer
exposure scenarios (see Table 3 at the end of this paper for list
of scenarios). One can then multiply the dose and risk values by
any radionuclide concentration in metal to calculate the doses
and risks for all applicable scenarios.
TABLE 1
AC-228
Am-241
Ba-137m
Bi-210
Bi-212
Bi-214
Co-58
Co-60
Cs-134
Cs-137
Mn-54
Ni-63
Np-237
Pa-233
Pa-234
Pa-234m
Pb-210
Pb-212
Pb-214
Po-210
Po-212
Po-216
Po-218
Pu-239
Pu-241
Ra-224
Ra-226
Ra-228
Rh-106
Ru-106
Sr-90
Tc-99
Th-228
Th-229
Th-230
Th-231
Th-232
Th-233
Th-234
Tl-208
U-232
U-234
U-235
U-236
U-238
Y-90
Cs-137+P
Np-237+P
Pb-210+P
Pb-212+P
Pb-214+P
PO-216+P
PO-218+P
RU-106+P
Sr-90+P
Th-232+P
U-232+P
U-235-J-P
U-238+P
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JAERI-Conf 95-015
DOSE AND RISK CALCULATION
A number of computer codes were evaluated dealing with
recycling and pathway analysis, including CONDOS2 and IMPACTS-
BRC3, which treat recycling explicitly, but none were found
entirely suitable for this analysis. Therefore, the modeling
strategy was developed. The intent was to produce a methodology
that would accurately evaluate the potential doses and risks, and
be flexible enough to be applied to other recycling situations.
Each operation was represented by one or more spreadsheets
using Microsoft EXCEL4. The EXCEL spreadsheet approach was
chosen to take advantage of the program @RISK3 to perform Monte
Carlo uncertainty analyses along with the point estimates of dose
and risk for each operation. Risks are presented as total
lifetime cancer incidence risk per picocurie per gram per year of
exposure.
The one exception to the EXCEL approach was the method used
to estimate the means and uncertainties in the furnace emissions
used to calculate individual and population exposures. An
uncertainty method used in the development of the National
Emission Standards for Hazardous Air Pollutants (NESHAPs6)
regulation was used.
Table 2, below, gives sample input parameters for two worker
exposure scenarios, and one consumer exposure scenario. Note
that not all pathways factor in each scenario.
TABLE 2
Input Parameters for Selected Scenarios
a. Direct Exposure Pathway Calculation Input Parameters and Distributions
Scenario/Operation
1.1.1.8 Worker near a small scrap
pile
6.2.1 .a Exposure from EAF during
melt
1 1 .2.1 .b End user of a large home
appliance
Scenario/Operation
1 .1-.1 .a Worker near a small scrap
pile
6.2.1.a Exposure from EAF during •
melt
11.2.1,b End user of a large home
appliance
Distance (m) most value
distribution min likely max expected
uniform 1.0 20.0 10,5
uniform 2.0 10.0 6.0
triangular 0.3 0.3 1.0 0.5
Exposure Period
days/yr years hours
250 1 1,083
250 1 1,750
350 1 525
Exposure !hrs/d) most value
distribution min likely max expected
triangular 1 .0 4,0 8.0 4.3
triangular 6.0 7.0 8.0 7.0
triangular 1 .0 1 .5 2.0 1 .5
Number of Workars most value
distribution min likely max expected
triangular 1.0 3.0 5.0 3.0
uniform 3.0 5.0 4.0
uniform 1.0 1.0 1.0
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JAERI-Conf 95^015
TABLE 2, Cont'd
b. inhalation Exposure Pathway Calculation Input Parameters and Distributions
Scenario/Operation
1,1.1.8 Worker near a
smalt scrap pile
6.2.1.a Exposure from EAF
during melt
Scenario/Operation
Respiratory Protection Reduction Factor |
Distribution min max expected value c
uniform 11 1 t
uniform 11 1 t
1 .1 .1 .a Worker near a small scrap
pile
6.2.1 .a Exposure from EAF during
melt
Scenario/Operation
1.1.1.8 Worker near a
small scrap pile
6.2,1.9 Exposure from EAF
during melt
>ust Loading ig/ro3) value
Istrib. min likely max expected
riangular S.OE-4 5.0E-3 1.0E-2 5.17E-3
riangular 5.0E-4 5.0E-3 1 .OE-2 5.17E-3
Exposure Periods Number of Workers most value
days/yr years hours distribution min likely max expected
250 1 1,083 triangular
250 1 1,750 uniform
Breathing Rate {m3/hr)
distrib. geo. mean GSD expected value
lognormal 1.83E-1 1.3 1.89E-1
lognormal 1.83E-1 1.3 1.89E-1
1353
3 54
Exposure Rate (hrs/d) value
* dtstrib, min likely max expected
triangular 148 4.3
triangular 6 7 8 7.0
c. Secondary Ingestion Exposure Pathway Calculation Input Parameters and Distributions
Scenario/Operation
1.1-l.a Worker near
a small scrap pile
Scenario/Operation
CF for Fe
in Fea03
0.7
ingestion Rate flm/hr)
distribution min max value expected
log-uniform 4.17E-7 4.17E-4 1.32E-5
1 .1.1 .a Worker near a small scrap
pile
Exposure Periods
davs/yr years hours
250 1 25O
Exposure Rate jhrsfd) value
distrib. min max likely expected
triangular 0.5 1.0 1,5 1.0
Number of Workers value
distrib. min likely max expected
triangular 135 3
The computer code MICROSHIELD7 was used to calculate dose
conversion factors for external doses from all operations, for
their particular geometry, at varying distances, and for each
radionuclide. Dose and risk factors for other pathways were
taken from published sources (see references 8-11). The doses
and risks were calculated using standard dose/risk algorithms.
For example, dose and risk from direct exposure are calculated as
follows:
Dose,, (x) = CiMtlKum * FE(ifx) * t
Risk, (X) = C,^^ * FDR * FE(i,x) * t
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where,
Dosej(x)
Riskj(x)
FE(i,x)
X
t
= dose, 1 year exposure to radionuclide i; mrem/yr
- total cancer incidence risk for 1 year of
exposure to radionuclide i; risk/yr
= concentration of radionuclide i in the medium of
interest; pCi/g
= dose factor relating exposure to dose for
radionuclide i at distance x; mrem/hr per pCi/g
= risk factor relating dose to risk; total cancer
incidence risk per mrem
= radionuclide of interest
— distance from source to receptor; meters
= rate of exposure; hours per year
As already noted, estimates of dose and risk are based on a
concentration of 1 pCi/g in metal for each radionuclide, for each
scenario. The values in the dose/risk table generated in the
report {hereafter called the DR Table) are used to estimate dose
and risk once the true radionuclide concentrations are known.
This is a simple scaling procedure. For example, consider a
batch of metal to be recycled that contains 12 pCi/g of Cs-137+P,
and 15 pCi/g of Co-60, for the operation, "Worker near a small
scrap pile (operation 1.1.1,a in Table 3}." From the DR Table,
the estimates of dose and risk per picocurie per gram for an
individual from these radionuclides are as follows:
Nuclide
Cs-137+P
Co-60
Dose (mrem/yr)
3.2xl(r3
l.SxlCT2
Risk/year
-9
2.2x10
l.OxlO'8
The dose to the worker would be,
12 X 3.2X10'3 + 15 X 1.5X10'2
and the risk would be,
12 x 2.2X10'9 + 15 x l.OxlO'8
0.26 mrem/yr
= l.SxlO"7 lifetime risk of
cancer incidence per year of exposure.
The above calculation presumes all assumptions that went
into producing the values in the DR Table are unchanged. Here,
if desired, one can easily account for two of the assumptions.
This worker scenario is based on five days per week of exposure
to the pile. If the exposure time is reduced to one day per
week, the above results need only be divided by 5. This scenario
also assumes the pile is 100% RSM. If the pile is only 50% RSM,
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JAERI^Conf 95-015
and the rest is "clean" scrap (evenly distributed), the initial
results need only be divided by 2.
Each dose and risk value in the DR Table is a sum of dose
and risk for each applicable pathway. The methodology affords
certain parameters in the scenarios to be easily varied to
estimate doses and risks with different inputs. For example, the
"Worker near a small scrap pile" consists of direct exposure,
inhalation and secondary ingestion pathways. One can vary
exposure parameters for each pathway.
The concentration in the slag and offgas, however, is not
taken to be 1 pCi/g as it is in the metal. To obtain the
concentration in the slag and offgas dust, partition and
concentration factors are applied to the scrap metal radionuclide
concentration.
Some modifications must be made to estimate population doses
and risks from consumer products. Since it is not known how many
people will use a consumer product, the population doses and
risks were calculated per curie of radioactivity in the metal
used to make that product (curie values are for melted metal).
The population of product users need not be known to calculate
these values - the values represent dose and risk spread over a
group of individuals. If some of the scrap is used for making
kitchen cook ware, and some for large appliances, the total
activity must be adjusted for the fraction going to each use.
The impact on the population, then, is the number of cancers
that would be expected. For example, in the DR Table, the large
home appliance scenario (Table 3, #11.2.1b) gives a population
risk of 0.66 per curie from Co-60. If the population exposed is
1000 persons from 1000 appliances with a total of one curie of
Co-60 in them, the effects would be 0.66 cancers in the
population of 1000 (given the parameters of the scenario). If
one curie is divided into 100,000 appliances with 100,000 users,
the effects would be 0.66 cancers in that population.
DOSE AND RISK FROM DOE'S STEEL
EPA evaluated doses and risks for 5 options presumed to
represent what DOE could do with its RSM. The options evaluated
are: No Action, or keeping the metal in piles or storage; Safe
Storage, which allows for improved storage of the RSM and delayed
disposal; Disposal, which entails immediate shipment to a waste
facility; Restricted Recycling, which would allow the metal to be
recycled and fabricated into uses within the DOE or nuclear
industry; and, Unrestricted Release, which could result in
consumer products being made with recycled RSM.
In the study, an existing quantity of RSM at the Oak Ridge
Reservation in Oak Ridge, Tennessee, which is a candidate for
recycling, was used to estimate potential risks from recycling
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DOE steel. A total of 60,000 metric tons of steel currently
exists at Oak Ridge, The Oak Ridge metal has been fairly well
characterized, including its radioactive contamination. The
analysis is, therefore, specific to this particular quantity, its
location, and suite of radioactive contaminants, and does not
necessarily characterize risks from all DOE RSM. The measured
contaminants are:
Radionuclide Concentration in Metal
U-238 600 pCi/g (22 Bq/g)
Tc-99 85,000 pCi/g (3,160 Bg/g)
Np-237-HP 0.035 pCi/g (0.0013 Bq/g)
Pu-239 0.30 pCi/g (0.011 Bq/g)
The estimates of dose and risk are based on the DR Table
containing dose and risk per picocurie per gram, adjusted to
account for the actual concentrations of the radionuclides found
in the Oak Ridge metal. To obtain the concentrations of the four
radionuclides in each melt phase, they were given the following
partition factors (note that no fraction of these radionuclides
partitions to the offgas):
Radionuclide Metal Offgas/dust
U-238 0.01 0
Tc-99 1.0 0
Np-237+P 0.01 0
Pu-239 0.01 0
After melting and decontamination, dose and risk factors
from the DR Table were multiplied by the resultant concentrations
of the four radionuclides, given below, and summed to estimate
the doses and risks in the various scenarios involved in each DOE
option.
Radionuclide Resultant Concentration - Metal
U-238 6 pCi/g (0.22 Bq/g)
Tc-99 85,000 pCi/g (3,160 Bq/g)
Np-237+P 3.5XW4 pCi/g (1.3xlO'5 Bq/g)
Pu-239 0.003 pCi/g (1.1x10"* Bq/g)
Worker and public exposure scenarios were adjusted to
accommodate the specific parameters for the Oak Ridge metal
recycling options. For example, the Disposal option involves
shipping 60,000 MT of RSM from Oak Ridge, TN, to a waste facility
in Nevada (3,200 km). Worker and population doses were
calculated according to the number of shipments, the number of
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JAERI-Conf 95-015
workers necessary, and the population that would be exposed
during the shipments.
The individual and population doses and risks for each
scenario from recycling 60,000 MT currently retained at the Oak
Ridge Reservation are listed in Table 3 at the end of this paper,
In the restricted recycling option, all the sceanrios would
presumably occur except that no metal would be made into consumer
products. In the unrestricted recycling option, all the
scenarios would presumably occur.
Population risk from products made with the recycled metal
must also adjusted for the quantity of metal recycled. The
population risk depends on the total curies of each radionuclide
that is recycled. Thus, the doses and risks for scenarios were
multiplied by the mass of metal (60,000 MT), and the
concentration of each radionuclide (given the above
partitioning), and then summed over the four radionuclides.
Population doses and risks for the end product scenarios are also
a function of the fraction of the total quantity given to each
product. Therefore, a value must be given describing what
percentage of the RSM may be used for each product, and
multiplied by the dose and risk. For example, if 50% of the
metal is presumed to go into automobiles, and 50% into
appliances, then the dose and risk for the automobile and
appliance scenarios are each divided by two.
LIMITATIONS
One limitation of this methodology is that only the
scenarios given can be modeled. New scenarios can be developed,
but they are time consuming to research and generate. Also, some
parameters of these scenarios are not easy to change, resulting
in inflexibility in the use of the scenario; for example,
doubling the EAF size from 80 tons to 160 tons will not result in
twice the emissions, nor will it lead to twice the dose to the
furnace worker. To truly evaluate the risks from RSM recycling,
a more comprehensive analysis of the most appropriate and
representative scenarios should be done.
It should also be noted that many of the scenarios have
large uncertainties. For many scenario parameters, the mean
value and probability distribution are not known. In some cases,
the distributions reflect the subjective judgement of the
analysts, and are not based on subjective evidence. Others have
large inherent uncertainties, for example, scenarios which
require breathing rates as a factor, and those requiring dust
settling velocities, will have large uncertainties associated
with them.
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CONCLUSIONS
The EPA has completed a draft study of recycling
radioactively contaminated scrap metal to provide information and
decision-making tools for the U.S. Department of Energy. A key
component of the study is a methodology for assessing doses and
risks which may result from the recycling of DOE-owned steel.
The methodology is based on picocurie per gram dose and risk
factors calculated for each radionuclide considered for one year
of exposure. Individual and population risk factors for a wide
range of worker and public exposure scenarios attempt to
characterize the potential doses and risks that could be received
if metal containing residual radiation were recycled. This
methodology was then applied to a specific quantity of DOE-owned
steel with known radioactive contamination, to estimate the doses
and risks that could result.
The methodology affords flexibility in modeling recycling
situations. The modeler only needs to know the starting quantity
of metal and the nature of the contamination to estimate
individual and population doses and risks for the given
scenarios. Some of the parameters in the exposure scenarios can
be easily modified to accommodate specific exposure situations to
which the model is applied.
Additional work that is needed on the risk assessment
methodology includes the development of more exposure scenarios,
and improving on uncertainties. This modeling methodology should
be evaluated for its usefulness in any regulatory development
process concerning residual radioactivity in recycled materials.
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Table 3. Annual Doses and Risks from Recycling 60,000 MT of Scrap Metal
Starting Concentrations, pCi/gm
U-238 5.99E+02 TC-99 8.52E+04 NP-2374P. 3.51E-02 PU-239 2.97E-Q1 :"
OPERATION
1.1. la Working near small (25 ion) scrap pile
I, Lib Working near large (250 ton) scrap pile
2,1. la Loading /Unloading from small (25 ton) scrap pile
2.1. Ib Loading /Unloading from large (250 ton) scrap pile
2.1.2a Driver: Beside vehicle
2,l,2b Driver: Inside cab of vehicle
2.2,1 Population: Truck transporting scrap
4.1. la Unloading the truck - sraal! (25 ton) scrap pile
4. Lib Unloading the truck - large (250 ton) scrap pile
4.1.2 Working around 1,500 ton scrap pile
4,1.3 Cutting/sizing scrap for furnace charge
5.1.1 Loading /Unloading truck
5.1. 2a Driver: Beside vehicle
5.1.2b Driver: Inside cab of vehicle
5.2.1 Population: Track transporting processed scrap
6.1. la Exposure from scrap pile
6.1.1b Exposure from moving the scrap by magnet
6. He Exposure from moving the scrap by small charging bucket
6.1. Id Exposure from moving the scrap by large charging bucket
INDIV.
DOSE
3.29E+02
3.29E4-Q2
3.29E+02
3.29E+02
5.69E-05
4.32E-Q4
0.00
1.24E-04
3.69E-04
3.29E+02
L56E+Q2
1.70E-05
4.14E-05
3.72E-04
0.00
5.31E+02
5.31E4-02
5.31E+02
5.31E+02
INDIV.
RISK
6",42E-05
6.42E-05
6.42E-05
6.42E-05
4.04E-11
3.07E-10
0.00
8.81E-11
2.61E-10
6.42E-05
3.05E-05
1.20E-11
2.96E-11
2.65E-10
0.00
1.04E-04
LQ4E-04
L04E-04
1.04E-04
POP,
DOSE
9.88E-01
9J8E-01
9J8E-01
9.88E-01
5.69E-08
4.32E-07
6.84E-05
2.49E-07
7.34E-07
1.32E+00
4.67E-01
5.09E-08
4.14E-08
3.72E-Q?
1.04E-04
7.94E-01
7.94E-01
7.94E-01
7.94E-01
POP.
RISK
1.93E-04
1 .93E-04
1.93E-04
1.93E-04
4.04E-11
3.07E-10
4J8E-08
1.77E-10
5.23E-IO
2.57E-04
9.19E-05
3.62E-1 1
2.96E-11
2.65E-10
7.41E-08
1.56E-04
L56E-G4
1 .56E-04
1.56E-04
P8
-------
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' • •''':,• ' •' '• .'. ... ' ..'• "••' • • ,"••• • . • . • • .':.. . .' -; '". .•.•„ '•-.;'•...•-... .. •-...'• . .' --:*1 -..'"'..•••...•"...•-..•',.';-;-.':-.':•:-•. °-- -•". •"-•"•; •:•".•.-:-'-.-'-' :.'-::•.';:>;*. -:v:-:V:V :E:I "•:•'•'•''•'•'••
. .' : :'•:,. •-:.:'. •: ••:':.-. .- ' ' • ' . '• •• ::;.-:.". -•.•-• • ' •• ..:•:.:. V v.-"..v -:' •:••:••: •--,..,•,•.-'. 'V -"' * I'V^iV^. V^*^ ^:-"^
:• . -•:•-• -... ;.. •: •• •:.. . '• :.'.••.-.-•• •-.•-. ::- .. ..-:v.- ••,••• .- •;• • -;^ . •' ^ -y
-------
Slatting Concentrations, pCi/gm
U-238 5.99E+02 TC-99 8.52E+04 NP-237+P 3.51E-02 PU-239 2.97E-Q1 '
OPERATION
" • • . '.••''- • • • "
9.1.2 Exposure from use of slag in road construction
9.2,1 Population: Truck transporting slag for disposal
9,2.2 Population: Exposure from use of slag in road construction
9.2.3 Population; Ingeslion of water contaminated by slag pile
10,1. It Manufacturingcars
10. Lib Manufacturing large home appliance
lO.l.lc Manufacturing large industrial equipment
10.1. Id Manufacturing eye glasses
10. Lie Manufacturing pans
10, 1 .2a Distribution of cars
10. 1 .2b Distribution of large home appliance
10.1 .2d Distribution of eye glasses
10.1,4a Point of sale of car
10.1.4b Point of sale of large home appliances
10.1.4(1 Point of sale of eye glasses
10. 1 .4e Point of sale of pans
10.2. la Population: Truck transporting cars
10.2. Ib Population: Truck transporting large home appliances
10.2. Ic Population: Truck transporting large industrial equipment
INDIV,
DOSE
4.13E+04
0,00
2.76E-04
0.00
2.37E-06
2.70E-06
4.67E-06
3.86E-07
6.56E-06
7.65E-06
8.11E-06
3.58E-07
5.93E-07
4.07E-07
7.79E-07
L17E-06
0.00
0.00
0.00
INDIV.
RISK
7,79E-03
0.00
L96E-10
0.00
L68E-12
L92E-12
3.31E-12
2.73E-13
4.67E-12
5.44E-12
5.76E-12
2J5E-13
4.21E-13
2.90E-13
5.55E-13
8.28E-13
0.00
o.oo
o.oo
POP,
DOSE
3.72E+02
2.39E-OS
3.31E-04
1.40E+02
9.48E-09
1.08E-08
L86E-08
L54E-09
2.63E-08
7.65E-09
8.11E-09
3.58E-10
5.93E-10
4.07E-10
7.79E-10
L17E-09
5.12E-08
7.93E-09
9.27E-09
POP.
RISK
7.01E-02
1.70E-08
2,35E-07
3.19E-06
6.74E-12
7.65E-12
1.32E-11
1.09E-12
1.87E-11
5.44E-12
5.76E-12
2.55E-13
4.21E-13
2.90E-13
5.55E-13
8.28E-13
3.65E-11
5.62E-12
6.60E-12
en
s
Ul
I
o
-------
Wl
I
Starting Concentrations, pCi/gm
U-238 5.99E+02 TC-99 8.52E+04 NP-237+P 3.51E-02 PU-239 2.97B-01
OPERATION
10.2. Id Population: Truck transporting eye glasses
10.2, le Population: Truck transporting pans
11. 2. la Population: End user of a car
1 1.2,lb Population: End user of a large home appliance
ll.Z.lc Population: End user of large industrial equipment
1 1.2, Id Population: End user of eye glasses
1 1 .2.2a Population: Fe from iron utensils
11.2.2b Population: Fe from stainless steel
1 1 .2.2c Population: Ni from stainless steel
1 1 .2.2d Population: Al from aluminum utensils
INDIV,
DOSE
0.00
0.00
3.65E-06
3.28E-06
6.14E-06
1.25E-05
4.76E+00
1.85E-02
4.76E-02
9.53E-03
IHDIV,
RISK
0.00
0.00
2.59E-12
2.33E-12
4.35E-12
8.88E-12
4.02E-06
1.55E-08
4.02E-08
8.02E-09
POP,
DOSE
2.54E-09
9.41E-09
1.36E-03
1.51E-03
4.97E-05
6.56E+Q1
4.16E-03
4.16E-03
4.16E-03
4.16E-03
POP.
RISK
1.81E-12
6.67E-12
9.69E-07
1.08E-06
3.53E-08
4.66E-02
2.94E-06
2.94E-06
2.94E-06
2.94E-06
-------
JAERI-Conf 95-015
References
1. "Analysis of the Potential Recycling of Department of Energy
Radioactive Scrap Metal," Sandford Cohen & Associates, for U.S.
Environmental Protection Agency, Contract No. 68D20155, Draft,
September 6, 1994,
2. "CONDOS-II-A Tool for Estimating Radiation Doses from
Radionuclide-Containing Consumer Products," F.R. O'Donnell, et
al., U.S. Nuclear Regulatory Commission, NUREG/CR-2068, November
1981.
3. "De Minimis Waste Impacts Analysis Methodology," O.I. Oztunali
and G.W. Roles, U.S. Nuclear Regulatory Commission, NUREG/CR-
3585, Volume l, February 1984.
4. "Microsoft EXCEL, Version 4.0", Microsoft Corporation,
Redmond, WA, 1992.
5. "@RISK: Risk Analysis and Simulation Add-In for Microsoft
EXCEL," Windows Version Release 1.1, Palisade Corporation,31
Decker Road, Newfield, NY 14867, February 6, 1992.
6. "Environmental Impact Statement, NESHAPS for Radionuclides,
Background Information Document, Volume 1," U.S. Environmental
Protection Agency, EPA 520/1-89-005, September, 1985.
7. "MICROSHIELD, Version 3", Grove Engineering Inc., Shadey Grove
Rd., Rockville, MD, April 12, 1988.
8. "External Dose-Rate Conversion Factors for Calculation of Dose
to the Public," U.S. Department of Energy, DOE/EH-0700, July
1988.
9. "External Exposure to Radionuclides in Airf Water, and Soil,"
Federal Guidance Report No. 12, U.S. Environmental Protection
Agency, EPA 402-R-93-081, September 1993.
10. "Health Effects Assessment Summary Tables," U.S.
Environmental Protection Agency, OHEA ECAO-CIN-821, March 1992.
11. "Limiting Values of Radionuclide Intake and Air
Concentration, and Dose Conversion Factors for Inhalation,
Submersion, and Ingestion," Federal Guidance Report No. 11, U.S.
Environmental Protection Agency, EPA-520/1-88-020, September
1988.
- 206 -
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JAEM-Conf 95-015
3-7
Study on Safety Evaluation for Unrestricted Recycling Criteria
of Radioactive Waste from Dismantling Operation
M. YOSHIMORl, M, OHKOSH1, M. ABE
Department of Decommissioning and Waste Management,
Japan Atomic Energy Research Institute
ABSTRACT
The study on safety evaluation was done, under contracting with the
Science and Technology Agency, for recycling scrap metal arising from
dismantling of reactor facilities. An object of this study is to contribute to the
examination of establishing criteria and safety regulation for unrestrected
recycling steei scrap.
To define amount of market flow of iron material in Japan and the amount
of radioactive waste generated from dismantling of reactor facilities, investigation
had been carried out. On basis of these investigation results and data in several
literature, individual doses to workers and to the members of the public have
been calculated as well as collective doses.
1, Introduction
The issue about e'xemption of very low level radioactive materials from
the regulatory control has investigated by some organs to the goverment
of Japan.
The Atomic Energy Commission and the Nuclear Safety Commission
have pointed out the possibility of the exemption from the regulatory
concern. For example, very low level radioactive materials, which cause
the trivial dose to the members of public, are able to dispose of as usual
municipal trash or to reuse.
In 1987, the Radiation Council has established the exemption criteria
for solid radioactive waste in shallow land burial as follows:
(1) The annual individual dose to the critical group shall be less than 10
{j.Sv from each source;
(2) The disposal methods of solid radioactive waste should also be
examined from the viewpoit of optimization.
It is expected that this criteria can be applied for recycling and reuse of
materials slightly conataminated with radionuclides.
Establishment of activity concentration limits for exemption is under
investigation by the Nuclear Safety Commission. And some scientific
investigations for the recycling of contaminated steel scrap are also being
carried out by JAERI under contract from the Science and Technology
Agency.
2. Amount of market flow of iron material
Crude steel output is shown in Fig. 1. Annual output of crude steel is
about 100 Mt in Japan. There is a economical fluctuation in total output
and in output of steel converter.Crude steel output of electric furnace
which is used relatively large amount of scrap increses to over 30 %.
The supply and demand of scrap is shown in Fig.2. There are also
- 207-
-------
JAERI-Conf 95-015
economical fluctuation in total amount of consumption and consumption
for steel-making. 50 Mt of scrap steel is dissipated, and about 80 % of
scrap steel is used for steel making. Therefore on the calculation, it is
assumed that 80 % recycled steel is used to manufacture a reinforecement.
The remainder is assumed to be used to manufacture each, items of
evaluation.
The market flow of iron material in 1990 is shown in Fig.3. Most of
scrap iron will be melted by electric furnuce and will be manufacturing
steel material. The avelage capacity of electric furnace is 50 t. A daily
throughput of steel products is 150 t. An annual working days of steel
woks are 250 days. Consequently the steel plant has a yearly throughput
37,500 t. Therefore on the calculation, the total activity of the recycled
steel is assumed to be spred uniformly over the whole production, namely
37,500 t.
O-
t—
=3
O
1/5
UJ
Q
196S
1970
Open-hearth Furnace
l=fca
19TS
1930 1985
FISCAL YEAR
1990
1995
Fig.l. Crude steel output in JAPAN,
Q_
U.
O
O
-<
60
50
40
30
.(Mt)
Q_
Q_
1965 19TO 1975 1980 1985
FISCAL YEAR
1990
19§5
Fig-2. Supply and demand of scrap.
- 208 -
-------
JAEM-Conf 95-015
IRON ORE
I
PIG IRC
1
-,_,_. „ _ •
,w SCRAP UNIT: M ton
A43'4
Hil
IB sl
2.0 •31. 5 1 1.2 7.6
| STEEL CONVERTER ELECTRIC FURNACE INDUCTION FURNACE
CAST IRON
104.4
0.5
5.9
Fig.3. Flow of Iron Material in JAPAN
3. Contaminated scrap metal arising from dismantling of reactor
facilities
In Japan, commercial nuclear power reactor has been increased since
the first reactor was coastracted in 1966. Over 40 reactors are operating
now. It is assumed that these reactor had been operating for 30 or 40
years and mothbaliing 5 or 10 years after shutdown. And it is expected
that two reactors will be dismantled in a year and will start in the year
2000.
The amount of recyclable scrap arising from dismantling reactor
facility is assumed to be from 1,200 to 1,600 t. Thus, in the calculation,
it is setuped 1,500 t of steel scrap generating from dismantling operation
of one reactor. This value is based on litelature data shown in Table l.(l)
It is assumed that average concentration of steel scrap is less than 1 Bq/g.
Consideration of the composition of contaminated steel and activated
steel scrap led to the selection of 30 nuclides for the calculation of
potential doses due to recycling. The radionuclides composition used in
the evaluation of activity concentration for recycling steel scrap is shown
in Table 2.
Table 1. Typical masses and activities in steels from a 1000 MWe PWR
containing very low level of activity
Activity Range
Surface
Contamina
-tion
(Bq/cm2)
37-370
3.7-37
0.37-3.7
Average
Coneentra
-tion
(Bq/g)
10
1
0.1
Time from Reactor Shutdown
5 Years
Steel
Mass (t)
800
1600
3200
Total
Activity
(Bq)
8x109
1.6x109
3.2x108
25 Years
Steal
Mass (t)
440
880
1760
Total
Activity
(Bq)
4.4 x109
8.8x108
1.8x108
100 Years
Steei
Mass (t)
240
480
960
Total
Activity
(Bq)
2.4x109
4.8x108
9.6x107
209
-------
JAERl-Conf 95-015
Table 2, Radionuclides composition considered in surface contaminants
and radi©activated carbon steel.
(5 Years after Reactor Shutdown)
Nuclide
55Fe
Surface Rddioactivated Carbon
Contaminants (%) Steel (%)
6.5 95
60Co 65.3 5
63Ni
90Sr
0/r Emitters 125Sb
134Cs
1.3
3.3
6.5
2.6
13/£s 6.5
147pm
241 pu
238pu
239pu
a Emitters 240Pu
241Am
244Cm
6.5
Z.6
48.1
5.8
7.7
16.0
22.4
-. Exposure Scenarios
In order to evaluate the radiological impact of recycling of steel scrap,
individual doses to workers and to the members of the public have been
calculated as well as collective doses.
The scenarios have been chosen assuming that Steel Scrap will be
recycled by means of melting in steel works. Therefore workers could be
exposed during the pre-treatment process of steel scrap, before the
melting process, the melting process .and the casting process.
The members of the public could be exposed to items made of recycled
steel. In addition, the public could be exposed to slag arising from the
melting process.
4.1 Compaction and Cutting Process
The steel scrap are classified, cut and compacted in the p re-treatment
works of steel scrap. The average monthly capacity of in these pre-
treatment works is from 4000 to 4500 t.
It is assumed that 1500 t of contaminated Steel Scrap could be treated
for 10 days, and the daily labour hours in these works are 8 hours.
The exposure scenarios and parameters consided in. compaction and
cutting process is shown in Table 3-1.
- 210 -
-------
JAERJ-Conf 95-015
Table 3-1 Exposure Scenarios and Parameters
for Compaction and Cutting Process
Exposure Scenarios
1 ) Load Trucks
2) Drive to Compaction and Cutting
Works
3) Unload Trucks
4) Store Scrap
5} Selection of Scrap
6} Compaction and Cutting
7) Transfer to Stock Yard
8) Loading Trucks
9) Drive to Steel Foundry
Number
of
Pepote
Exposed
Z
-1
2
4
Z
1
Z
Z
1
Exposure
Duration
(hr/year)
15
10
15
80
80
80
80
IS
10
Source
Geometry
P
A
P
P
P
P
P
P
A
Dimention of
the Source
(m)
Radius
1.1
1.1
1.1
10.0
1.7
1.7
0.3
1.1
0.6
Height
6.8
6.8
6.8
2.0
3.0
3.0
0.7
6.8
5.8
Distance
from
the Source
Cm)
0.5
1.0
O.S
10.0
1.0
1.0
1.0
0.5
1.0
Density
of the
Source
(g/crrf)
0.5
O.S
0.5
0.5
0.5
0.5
1.5
1.5
1.5
Shielding
None
None
None
None
None
None
None
None
None
Inhalarior
o
o
o
o
o
o
o
o
logestion
O
o
o
0
0
o .
o
o
Note of the source geometry
P:Cylindrical, homogeneous self absorbing volume source with or without shielding, dose point on trie perpendicular bisector
of the axis.
A:Cylindrical, homogeneous self absorbing volume source with or without shielding, dose point on the axis.
4.2 Smelting Process
Each 1000 t of steel-throughput is assumed to produce 100 t of slag
and 10 t of dust (3),(4). The distribution coefficients of radionuclides
between steel and slag or dust during the smelting process are set as
shown in Table 4 based on several literatures data (4),(5),(6). And the
parameters used in internal dose are set as follows:
- The Dust Concentration in Working Area is 1 mg/m3. It is the
recommendation value for the dust concentration of ferro-oxide in
the woking atomosphere in Japan,
- The respiratory flow of woker is 1,2 m3/h,
- The rate of secondery ingestion of removable surface contamination
is 0.01 g/h (7).
The internal exposure duration is the same as the external exposure.
Table 4. Distribution of the Contaminants during the Smelting Process.
Element
Fe
Co
Ni
Sr
Sb
Cs
Pm
a Emitters
Distribution Coefficient
Ingot
1
1
1
1
1
1x10-1
1
1 X 1 O-1
Slag
1X10-2
1X10-2
IXIO-2
1
1 X 1 0-2
1
1
1
Fume
5X10-3
5X10-3
5X10-3
1 X 1 O-1
1
1
1
5X10-3
-211 -
-------
JAERI-Conf 95-O15
The exposure scenarios and parameters consided in smelting process
3wn in Table 3-2.
is
shown
Table 3-2 Exposure Scenarios and Parameters
for Smelting Process
Exposure Scenarios
1 ) Unload Trucks
2) Store Processed Scrap
3) Transfer to Furnace
4) Charge Furnace
5) Operate Furnace
6)- Sampling
7) Tap Fumance
8} Store Slag
9) Transfer Slag
10) Treatment of Dust
Number
of
Pepote
Exposed
2
2
2
2
2
2
3
3
3
1
Exposure
Duration
(hr/year)
275
2200
825
412.5
1650
412.5
275
27S
27S
187,5
Source
Geometry
P
P
P
A
P
A
A
P
P
P
Dimention of
the Source
(m)
Radius
1.1
10.0
1.0
2.4
2-4
2.4
2.4
1.5
1.S
1.5
Height
6.8
0.7
1.1
0.4
0.4
0.4
0.4
1.0
1.0
0.7
Distance
from
the Source
(m)
0.5
10.0
0.5
5.0
5.0
O.S '
O.S
5.0
1.0
0.5
Density
of the
Source
fa/errf)
1.S
1.5
1.5
7
7
7
7
3
3
3
Sruetdimj
None
None
None
None
Concrete*
None
None
None
None
None
Inhabtior
o
o
o
o
0
o
o
o
o
o
Ingestion
O
o
o
o
o
o
o
o
o
o
Note of the source geometry
P: Cylindrical, homogeneous self absorbing-volume source
of the axis.
A : Cylindrical, homogeneous seff absorbing volume source
•1 : Density Z,3g/crn?, Thickness 0-ESm
with or without shielding, dose point, an th« perpendieuiar bisector
with or without shielding, dose point on the sxb.
Table 3-3 Exposure Scenarios and Parameters
for Casting Process
Exposure Scenarios
1 ) Transfer of Melted Steel
2) Pour Ingots
3) Strip Molds (1) "2
4) Strip Molds (2) *3
5) Transfer in Steel Works (1 ) 2
6) Transfer in Steel Works (2)*3
7) Grind Ingots
8) Stock Ingots
9) Load Trucks
10) Drive to Manufacturer
Number
of
Pepote
Exposed
3
2
2
2
1
1
1
4
2
1
Exposure
Duration
(hr/year)
96H.5
550
550
SSO
550
550
1100
2200
275
250
Source
Geometry
P
A
P
P
A
A
P
P
P
A
Qimention of
the Source
Cm)
Radius
1.25
1.25
0.34
0.57
0.34
0.57
0.57
2.50
0.57
0.57
Height
1.5
1.5
2.0
2.2
2.0
2.2
2.2
2.2
2.2
2.2
Distance
from
the Source
(m)
3.0
0.5
O.S
0.5
0.5
0,5
2.5
10.0
O.S
1.0
Density
of the
Source
, , 3.
(gftm )
7
7
7.2
7.2
7.2
7.2
7.2
7.2
7.2
7.Z
Shreldhg
Concrete*^
None
None
None
None
None
None
None
None
None
nhajation
O
o
o
o
o
o
o
o
o
ingestiot
O
o
o
o
o
o
o
o
o
Not* of the source geometry
P: Cylindrical, hQwogeneciys self absorbing vokime source with of without shielding, dose point on th« pmrpendscular bisector
of th« axis.
A : Cylindrical, homogeneous saif absorbing volume source with or without shielding, dose point an the axis.
*t ;0ensity 2.3g/ert£. Thickness 0.2Sm *z:5 tors casting mould ~3 :16tan casting rnouW
- 212
-------
JAERl-Conf 95-015
4.3 Casting Process
The exposure scenarios cotisided in casting process are shown in Table
3-3. We considered two types of casting mould.
4.4 Use of Items Made with Recycled Steel
(1) Use of mnforcement in building concrete
It is assumed that reinforced concrete contains 80 kg of steel per cubic
meter of concrete and the reinforcement is located 2 cm depth in walls,
floor and ceiling.
We assumed that the floor area of the room is 15 m2, the height is 3 m,
and the thickness of concrete is 18 cm.
and 30,000 t of steel used in the building induslory to construct about
30,000 rooms each housing 4 person.
120,000 person will be exposed in this scenario.
(2) Use of car,made with recycled steel
The body of the car is represented by a sphere, having 1.5 m radius
and 2 mm thickness, and the driver is assumed to be located at the center
of the sphere.
And 7,500 t of steel used in the manufacturing of cars,
About 420,000 cars will be made and 420,000 persons will be
exposed.
(3) Use of railroad c_arriajge_ma,d_e_ with recycled steel,
The body of the railroad carriage is represented by a rectangular form,
having 3 m width, 20 m length, 2.5 m height, and 0.5 mm thickness, and
a passenger is assumed to be located at the center of the carrige and at the
1 m heigt of the floor.
7,500 t of Steel used in the manufacturing of about 2,200 railroad
carriages.
In fiscal 1986, Japanese avrage railroad passenger number is 440,000
persons per carriage. Thus 970 Million persons will be exposed. But in
the calculation for collective dose, an average railroad bording time in
fiscal 1987, which is about 25 minutes, is used.
(4) Use of furniture
The furniture concerned is assumed to be a 50 kg cylinder having 100
cm height and 45 cm radius.
7,500 t of steel used in the manufacturing of furniture. About 15,000
pieces of furniture will be made and 15,000 persons will thus be exposed.
(5) Use pf steel desk.
The steel desk is assumed to be made from three steel plates having,1
mm thickness.
An office woker is assumed to use it 1,500 hours per year.
7,500 t of steel used in the manufacturing of steel desks.
About 250,000 desks will be made and 250,000 persons will thus be
exposed.
(6) Use of flying pan
The frying pan concerned is auumed to be a 3 kg cylinder having 0.5
~213 -
-------
JAERI-Conf 95-015
cm height and 15 cm radius.
A housewife is assumed to use it 300 hours per year for meal
preparation at 30 cm away.
The frying pan is assumed to corrode during meal preparation.
The corroded steel could be released to foodstuffs.
The housewife could ingest contaminated steel.
In the calculation we assumed that corrion rates of cast iron is 0.127
cm/y and 1 % of corroded iron is ingested.
This is equivalent to ingest 0.167 g contaminated steel per year by her.
150 t of steel used in the manufacturing of frying pans.
About 250,000 pans will be made and 250,000 persons will thus be
exposed.
(7) Use of slag as a base material of pavement
In the calculation, slag is assumed to be used as a base material of a
paved parking lot.
The slag layer, having a thick of 35 cm, is located under the surface
asphalt layer 25 cm thickness.
The manager of the parking lot is assumed to spend 2,000 hours per
year.
In the calculation, the parking lot is represented semi-infinite volume.
. Results
Table 5 show the calculation results of the activity concentration which
corrsponds to annual individual-dose level 10tiSv. In the derivation of
the surface activity from the mass activity, it is assumed that the average
ratio of contaminated surface to the mass is 10 m~lt. Consequently, the
activity concentration for $/y nuclides is similara to the value, 1 Bq/g
for $ I y nuclides, which is indicated in the CEC Radiation Protection
No.43.
Table 6 shows the calculation results of collective dose for typical
exposure scenarios. The total number of the collective dose is the sum of
the result of the reinforced building scenario and the biggest numerical
value in the other scenarios concerned. This results exceed 1 man-Sv/y,
but it can be considered as thin board manufactured by scrap are almost
nothing. Because it is weak in bendingd. Therefore, we think this results
is not a serious problem.
Table 5. Recommend Clearance Level for Recycling Steel Scrap.
0/r Emitters
a Emitters
Surface Contaminants
(Bq/g) (Bq/cm?)
0.7 7.0
0.8 8.0
Radioactivated
Carbon Steel
(Bq/9)
2.7
-
214
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JAERI-Conf 95-015
Table 6. Calculation Results of Collective Dose for Typical Exposure
Scenarios. (Using Recommended Clearance Levels)
Exposure Scenarios
Reinforcement in Building
Concrete
Car Driving
Boarding on Railroad
Carriage
Use of Furniture
Use of Steel Desk
Use of Flying Pan
Number of
Peoples
1.2X1Q5
4.4X104
9.7X1 0s
1.5X104
2.5X105
2.5X105
Total
Surface
Contaminats
(man Sv)
1.2X10 °
1.4X10-1
2.9X10 °
1.1X10-2
1.9X10'1
1.2X10'3
4.1X10 °
Activated
Carbon Steel
(man Sv)
3.4X10-'
4.2X1Q-1
8.9X10°
3.0X10-2
5.8X10-'
3.4X1Q'4
9.2X10°
Referace
(1) G.M.Smith.C.R.Hemming.M.J.Ciark.A.M.Chapuis and H.Garbay,
"Methdology for Evaluating Radiological Consequences of the
Management.of Very Low Level Solid Waste Arising from the
Decommissioning of Nuclear Power Plants", EUR 1058-EN(1985),
(2) J.C.Evans, et ah, "Long-Lived Activation Products in Rector Materials",
NUREG/CR-3474, (1984).
(3) F.R.O'Donnell.S.J.Cotter.D.C.Kocher.E.L.Etnier and A.P.Watson,
"Potential Radiation Dose to Man from Recycle of Materials Reclaimed
from a Decommissioned Nuclear Power Plant"., U.S.Nuclear Regulatory
Commission (NRC),NUREG/CR-Q134(1978)
(4) Commission of the European Communities (CEC), "Radiological
Protection Criteria for the Recycling of Materials from the Dismantling of
Nuclear Installations", Radiation Protection No.43.rLuxembourg, (1988).
(5) A.M.Chapuis,P. Guetat and H.Garbay, "Exemption Limits for the
Recycling of Materials from the Dismantling of Nuclear Installations",
CONF-871018, (1987)
(6) D.S.Harvey, "Research into the melting/refining of contaminated steel
scrap arising in the dismantling of nuclear installations", EUR
12605(1990).
(7) International Atomic Energy Agency (IAEA), "Working Document on
Exemption Principles Applied to the Recycling of Materials from Nuclear
Facilities", (1990).
_ 215 -
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JAERI-Conf 95-015
3_8 RADIOLOGICAL CONTROL CRITERIA FOR MATERIALS
CONSIDERED FOR RECYCLE AND REUSE
W. E. Kennedy, Jr.*, R, L, Hill*, R. L. Aaberg*, and A. Wallo, III*
'Pacific Northwest Laboratory^
+U.S. Department of Energy, Office of Environmental Guidance
ABSTRACT
Pacific Northwest Laboratory (PNL) is conducting technical analyses to support the U.S.
Department of Energy (DOE), Office of Environmental Guidance, Air, Water, and Radiation
Division (DOE/EH-232) in developing radiological control criteria for recycling or reuse of metals
or equipment containing residual radioactive contamination from DOE operations. The criteria,
framed as acceptable concentrations for release of materials for recycling or reuse, are risk-based
and were developed through analysis of generic radiation exposure scenarios and pathways. The
analysis, includes evaluation of relevant radionuclides, potential mechanisms of exposure, and non-
health-related impacts of residual radioactivity on electronics and film. The analysis considers 42
key radionuclides that DOE operations are known to generate and that may be contained in
recycled or reused metals or equipment. The preliminary results are compared with similar results
reported by the International Atomic Energy Agency, by radionuclide grouping,
1.0 INTRODUCTION
Pacific Northwest Laboratory (PNL) is collecting data and conducting technical analyses to
support efforts by the U.S. Department of Energy (DOE) to develop radiological control criteria for
the recycling and reuse of scrap materials and equipment that contain residual radioactive
contamination. The initial radiological control criteria are the concentrations in or on materials
considered for recycling or reuse that meet the individual or industrial (electronics/film) dose
criteria. The analyses include determining relevant radionuclides, potential mechanisms of
exposure, and methods to determine possible non-health-related impacts from residual radioactive
contamination in materials considered for recycling or reuse. The data and models described in this
paper may be considered by DOE (in coordination with other U.S. Federal agencies) with other
information to set radiological control criteria for recycling that are as low as reasonably achievable
(ALARA) and to support environmental regulations.
To determine if recycling is the "preferred" action or approach for management of material,
DOE has identified two criteria. The action must be 1) environmentally acceptable and cost
effective or 2) environmentally preferred. Under this approach it is recognized that some situations
exist under which the direct costs may be higher for recycling than for burial, but environmental
costs avoided by recycling (e.g., the recycling option reduces environmental insults associated with
certain secondary impacts and reduces the overall need for disposal space) balance the short-term
costs associated with the recycling activity (i.e., the costs recovered from the recycled materials may
not compensate for overall program expenses).
(a) The Pacific Northwest Laboratory is operated by Battelle Memorial Institute for the
U.S. Department of Energy under Contract DE-AC06-76RLO 1830.
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JAERI-Ccmf 95-015
The preliminary results described in this paper are based on generic exposure scenarios and
pathway analyses using 42 radionuclides determined to be potentially present as residual
contamination in metals or equipment from DOE operations. These radionuclides were identified
from input provided across the DOE complex that considered all aspects of the defense nuclear fuel
cycle and research activities including the operation of accelerators. The scenarios and information
developed by the International Atomic Energy Agency (IAEA) in Safety Series No. 111-P-Ll
(1992), Application of Exemption Principles to the Recycle and Reuse of Materials from Nuclear
Facilities [1], were considered as a primary reference in developing the initial radiation exposure
pathway and scenario analysis. Additional analyses were conducted to determine the potential non-
health-related impacts industry may experience from residual contamination in recycled metals, such
as those used in industries producing and using electronics and X-ray film.
Although alternative public dose limits were considered, the initial control criteria in this
report are based on 1) a dose of 10 /u.Sv y"1 (1 mrem y"1) to a worker in a smelter or an individual
who uses consumer products made from recycled materials, 2) a dose of 1 ju.Sv y"1 (0.1 mrem y"1) to
an individual downwind from a smelter used to process recycled metals, and 3) minimizing non-
health impacts associated with potential radiation effects on electronics or film.
2.0 DOSE ASSESSMENT METHODS
To determine if radioactively contaminated materials can be released from regulatory
controls, it is necessary to first determine the potential future uses for the materials and then the
potential radiation doses resulting from those future uses. Generic radiation exposure scenarios
were used to conceptually model likely future uses for materials released for recycle or reuse.
While these scenarios may not exactly match existing or projected future conditions, they are
designed to serve as the basis for conducting a dose analysis for the average member of a critical
population group. These scenarios are a combination of radiation exposure pathways that contain
specific exposure conditions. This section contains a summary of the basic radiation exposure
pathways, scenarios, and methods used to estimate the preliminary control criteria for recycle or
reuse of materials.
2,1 General Assumptions
For the preliminary calculations that follow, it is necessary to assume that 100 t of
contaminated steel, aluminum, and concrete and 10 t of copper with a normalized unit of initial
activity per unit mass are recycled during a year. This assumption allowed a normalized calculation
that leads to the development of bulk contamination control criteria. For the development of
surface contamination control criteria, individual tools or pieces of equipment for reuse are
considered. A unit concentration of each radionuclide is assumed and control criteria in terms of
Bq g*1 for volume and Bq cm"2 for surface contamination are derived. For the scenario calculations
that follow, 42 reference radionuclides (plus two additional concrete activation products) were
selected. The radionuclides considered and their physical half-lives are listed in Table 1.
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JAERI-Conf 95^015
TABLE 1. Radionuclides Considered in the Recycle and Reuse Analysis
Nuclide
3H
14C
36a
41Ca
^Mn
55pe
57Co
60Co
' 63Ni
^Zn
^Se
90Sr + Y
93Zr
^Nb
"Tc
106Ru
I10mAg
125Sb
129!
134Cs
137Cs
Half-Life (y)
12.3
SJxlO3
SxlO5
UxlO5
0.86
2.7
0.74
5.3
99.9
0.67
6.5xl04
28.5
l.SxlO6
2.0xl04
2.13xl05
1.01
0.68
2.8
1.6xl07
2.06
30.1
Nuclide
144Ce
l47Pm
mSm
mEu
154Eu
226Ra
228Th
229Th
230Th
232Th
232TJ
233U
234U
235rj
238|j
237Np
238pu
239pu
240Pu
241Pu
241Am
Half-Life (y)
0.78
2.62
87
13.6
8.8
1.6xl03
1.91
7.34x1 03
7.7xl04
1.4xl010
72
l.SxlO5
2.47x1 05
7.1xl08
4.51xl09
2.14xl06
87.6
2.4X104
6.57xl03
14.4
4.34xl02
The choice of the 42 radionuclides also takes into account other considerations including
the origin of the radionuclides; whether natural uranium (238U), uranium activation products
(B9Pu, 241Pu, and 24lAm), fission products (mSt, 99Tc, and 137Cs), or activation products (XC\,
41Ca, 54Mn, 55Fe, "Co, 63Ni, ^Zn, 94Nb, "Tc, and I52Eu)
the half-life of the radionuclides; whether relatively short (65Zn) or very long (94Nb or 239Pu)
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JAERI-€onf 95-015
the importance of the radionuclides in ,the context of bulk activation or surface
contamination; that is, over the short term (55Fe, ^Zn, and MCo), long term (63Ni, 137Cs, and
1S2Eu), or very long term pU, ^'Pu, MNb, and wTc)
the mode of decay and internal dose conversion factors (DCF) including alpha emitters with
large DCFs (228Th, 23<>rh, *»Th, ^U, 233U, ^U, ^U, ^U, 226Ra, ^No, ^pu, «°Pu, M1Pu,
and 241Am); beta/gamma emitters with large DCFs (^Co, ^Zn, MNb, *Sr, 110mAg, ml, 134Cs,
137Cs, 144Ce, 147Pm, 151Sm, 152Eu, and 154Eu), non-photon emitters with moderate DCFs ("Sr,
106Ru, and ^Pu), beta/gamma emitters with low DCFs (*O, ^Mn, 55Fe, 57Co, "Tc, and
125Sb), or non-photon emitters with low DCFs (3H, 14C, 41Ca, 63Ni, and
• the behavior of the radionuclides during recycle operations; that is, whether they are
volatilized and escape, are concentrated in the metal product, or partitioned in slag or ingots
(products).
Early daughter products in equilibrium with parent radionuclides are assumed in all cases,
For smelting, it is probable that the majority of some radionuclides, such as ^Co, remain in the
ingot. However, a fraction of material will remain in the slag, and another portion will likely
volatilize and be released with fumes and gases. The behavior of a specific radionuclide will depend
on the chemistry of the radionuclide in question and the type of smelting process considered.
Because the partitioning is not known for most radionuclides during smelting, the dose calculations
that follow are based on the conservative assumption that, for each radionuclide, aE of the activity is
retained in each of the three phases of smelting: the metal (steel, aluminum, or copper), the slag,
and gases released out of the stack. The slag is assumed to equal about 10% of initial mass of the
steel, or about 10 t in the steel and aluminum analyses and 1 t in the copper analysis. This triple
accounting approach will overestimate the true doses; however, it will maximize the potential
importance of the scenarios and should serve as an adequate basis for the initial development of
radiological control criteria for recycling and reuse.
22. Radiation Exposure Pathways
Humans may be exposed to radiation in three main ways:
* exposure to external radiation
» inhalation of radioactive gases or small particles
* ingestion of radioactive material.
The following paragraphs describe the specific ways in which these pathways have been used
as part of the assessment methods in this study,
2.2.1 External Radiation Exposure
The radioactive sources considered in this study are generally represented by a self-absorbing,
homogeneous, cylindrical volume or surface contaminated source with the dose point on the central,
longitudinal axis of the cylinder [1, 2], Except for exposure conditions that represent exposure to
molten metals contained in a furnace, external absorbers and shields are ignored. This procedure
tends to maximize the estimated dose equivalents from exposure to external radiation. In some
situations, a source can be represented better by a half cylinder than by a full cylinder. For these
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JAEW-Conf 95-015
situations, a full cylinder is defined such that the area of its flat surface is twice that actually
needed; the effective dose equivalent is then calculated by using the full cylinder and dividing the
resulting dose by two. The external dose calculations are performed using the EXTDF module of
the GENII Software System [3].
2.22 Inhalation Exposure
Committed effective dose equivalent factors are taken from International Commission on
Radiological Protection (ICRP 1977-1982) Publication No. 30, and its supplements [4). The
concentration of respirable dust in the air will vary depending upon a variety of factors including the
physical condition of the- material being handled, the quantity of the material present, and the ,
building ventilation. Thus, it is difficult to predict the concentrations that may be present during
any recycle step. However, so that a complete analysis may be performed, air concentrations have
been assumed based on the information in IAEA Safety Series No. 111-P-l.l [1] for those recycle
steps where the potential for inhalation is most likely. In general, the air concentrations were
assumed to vary between about 1Q"3 and 1CT5 g m"3.
2.2.5 Ingestion Exposure
For this study, ingestion is assumed to occur by one of three separate routes:
* ingestion of removable radioactive materials on surfaces
* ingestion of corroded material from using frying pans or water pipes
• ingestion of food products contaminated by airborne plumes released from a smelter.
Ingestion of removable radioactive contamination found on recycled metals or reused
equipment can occur when workers inadvertently transfer contamination from a surface to hands,
foodstuffs, cigarettes, or other items that enter the mouth. Since very little information exists on
the estimated radiation doses associated with this pathway, the methods outlined by the IAEA for
recycle and reuse [1] are used for this study. A quantity of 10 mg of contamination per hour of
direct contact exposure is assumed for ingestion by adult workers at a smelter. Ingestion of
contaminated metal corroded from frying pans during cooking or from copper water pipes are
considered as separate ingestion pathways. O'Donnel et al. [5] studied the potential impact of
recycle considered cast iron pans using an assumed corrosion rate of 0.127 cm y4 [5]. Since the use
of stainless steel and aluminum pans with a much lower corrosion rate is perhaps more consistent
with current domestic practice, a lower value of 0.13 mm y"1 is used for this study.
2.2.4 Downwind Exposures
A potential source of public exposure from metal recycling materials that may volatilize and
released through the stack during smelting. The potential radiation doses to the downwind public
are estimated using the CAP88-PC [6] computer code. This software was developed by the U.S.
Environmental Protection Agency to perform dose and risk assessments for demonstrating
compliance with the National Emission Standards for Hazardous Air Pollutants (NESHAPS) rules
in 40 CFR 61.93a [7]. The exposure pathways considered in the analysis included inhalation of
airborne material, external exposure to penetrating radiation, and ingestion of contaminated foods.
The default meteorological data files for Chicago contained in the CAP88-PC code are used in this
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JAERI-Conf 95-015
study. Meteorological data for Chicago are selected because they were felt to be representative of
many midwestern U.S. industrial settings with a large population in the nearby vicinity,
3.0 RADIATION EXPOSURE SCENARIOS AND ASSUMPTIONS
For this analysis, six separate categories of contaminated (or activated) materials and future
conditions were considered:
* recycle of steel
* recycle of aluminum
* recycle of copper
* recycle of concrete (as aggregate)
• reuse of a contaminated room within a facility
• reuse of tools or equipment (with surface contamination).
These six categories are further subdivided into various exposure scenarios, describing the
activities of specific individuals or groups of individuals. The range of scenarios evaluated is based
on previous dose estimates for recycling and reuse [1, 5, 8, 9] to adequately represent those
scenarios likely to be of generic importance and relevance to all DOE nuclear facilities. The
scenarios presented here yield the highest potential doses for each category of recycled material and
radionuclide grouping, as determined from the IAEA [1] study. Details of the scenarios considered,
the relevant assumptions, and the values assigned to the important parameters are discussed in the
following sections.
3.1 Scenarios for Steel Recycle
Recycled steel may contain both activation products and surface contamination from reactor
coolant or other sources. The three most limiting scenarios identified by the IAEA [1] are for 1) a
slag worker at the smelter, 2) consumers who drive an automobile, and 3) consumers who work with
a piece of large equipment made of recycled steel.
3.2 Scenarios for Aluminum Recycle
Although the long-lived activation of aluminum is negligible, surfaces may become
contaminated through contact with reactor coolant or other sources. The scenarios described by the
IAEA [1] for recycle of aluminum were evaluated and the three most limiting ones are used in this
study. The limiting scenarios are 1) an operator at a furnace, 2) consumers using automobiles, and
3) consumers using frying pans.
33 Scenarios for Copper Recycle
The IAEA [1] study did not consider copper recycle. However, within the DOE complex,
copper recycle may be quite significant; thus, it is included in this analysis. By analogy with steel
and aluminum, three scenarios for recycle of copper are identified for this study. They are doses to
1) a furnace operator, 2) individuals who use copper pans, and 3) individuals who live in houses
with pipes made from recycled copper.
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JAERI-Conf 95-015
33 Scenarios for Concrete Recycle
Large quantities of activated or contaminated concrete will be encountered during
decommissioning of DOE facilities. Because there is an economic incentive to avoid the costs of
transport and disposal of radioactive concrete as radioactive waste, recycling of concrete as
feedstock for further concrete manufacture has been considered by European countries [1]. Before
such reuse could be authorized, it is clear that any existing building candidate for release would
have to pass an extensive radiation survey to assure compliance with existing national regulations.
For the IAEA [1] study, the recycled concrete is assumed to be used to build a new structure in
which individuals live or work for 6000 h y"1. For calculational purposes, the initial concrete is
assumed to be contaminated to an undiluted, unit concentration. Although a very large dilution
could occur during the manufacture of new concrete structures, for this analysis a 1:10 dilution is
assumed. The limiting scenarios used are a concrete worker and a resident in a room made from
recycled concrete. The concrete activation products ^Cl and 41Ca are included in the analysis to
account for the potential activation of concrete.
3.4 Scenario for Reuse of Contaminated Rooms
Concrete buildings may be decontaminated and reused for other purposes after
decommissioning. The scenario considered for building reuse is intended to account for normal
occupancy, as described in an evaluation of residual radioactive contamination conducted for the
U.S. Nuclear REguIatory Commission by Kennedy and Strenge [10]. For the building occupancy
scenario, individuals are assumed to work in a building after unrestricted release. Although the
residence time could vary, a normal work year of 2000 h y"1 is assumed. Because decontamination
efforts before release focus on the removal of surface sources, the air concentration was assumed to
be 10~5 g m"3 and the ingestion rate was assumed to be 1.0 mg h"1 of exposure. These values are
10% of the values assumed for workers at a smelter.
3.5 Scenarios for Reuse of Equipment or Tools
During decommissioning, discrete pieces of contaminated equipment (including hand tools,
pumps, small motors, furniture, and storage tanks) may be salvaged and released for unrestricted
use if they can meet radiological control criteria. For the IAEA [1] study, it is assumed that the
fixed contamination present on the surfaces of the tools or equipment is ten times higher than the
removable fraction as measured by swabbing. For this study, the radiation exposure scenarios that
may be most limiting are used. These involve the use of hand tools that incorporate a small motor
(i.e., an electric hand drill or saw) because of the potential presence of contamination on the inner
surfaces of the motor which is difficult to monitor. A high exposure duration of 600 h y"1 is
assumed because of the relative proximity of power tools to workers under construction conditions.
The exposure pathways considered by the IAEA included exposure to external radiation, ingestion
of contamination transferred from the surfaces of the tool to hands and then to the mouth, and
inhalation of localized airborne material from the hand tool. In addition to reuse of small items,
the IAEA considered that larger items could also be candidates for reuse. These items are likely to
contain surface contamination; thus, the same exposure considerations as for hand tools apply, with
modifications accounting for the size of the item.
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JAERI-Conf 95-015
4.0 POTENTIAL EFFECTS ON ELECTRONICS
In addition to human radiation doses, another concern is the potential effects of unrestricted
use of radioactively contaminated recycled metals on electronic components. The threshold range
for damage to electronic components from radiation varies with the type of component. In
addition, selected electronic components can be "hardened" against radiation effects when it is
anticipated that they may be used in high radiation fields (such as space applications). In general,
for non-hardened components, the damage thresholds for electronic components range from about
5 Gy (500 rad) to about 500 Gy (50,000 rad) [11]. Assuming a 10-y lifetime for electronic
components, this translates to a dose-rate range of about 5 x 10"5 Gy h"1 (5 x W3 rad h"1) to 5 x 10"3
Gy h"1 (0.5 rad h'1) [11]. For comparison, natural background radiation is about 1 x 10"3 Gy y"1
(0.1 rad y"1, or about 1 x 10"7 Gy h*1 (1 x 10"5 rad h*1). Since the dose limits considered for the
development of control criteria are a fraction of annual background, the development of a special
control criteria for electronic components is deemed unnecessary.
5.0 POTENTIAL EFFECTS ON FILM
One potential concern related to recycling of metals and concrete containing residual
radioactive contamination is that these recycled materials may be used as material for making film-
storage boxes. It is well known that film is sensitive to exposure to radiation and that two of the
major uses of film are in the fields of medical and industrial radiography. To help prevent
undesirable darkening or fogging of films prior to use, the National Council on Radiation Protection
and Measurements (NCRP) recommended that radiographic film stored in darkrooms or storage
areas should not be exposed to more than 2 /u.Gy (0.2 mrad) of radiation prior to developing [12].
For design specifications for film-storage areas, the NCRP recommends assuming a one-month
storage time as an average, if the exact time is not known. In this analysis, an estimate of the
potential doses to film resulting from storage was made for storage in four different types of
containers constructed from recycled materials.
6.0 RESULTS AND DISCUSSION
The results of this preliminary study are based on generic exposure scenarios and pathway
analyses using 42 radionuclides determined to be potentially present as. residual contamination in
metals or on equipment from DOE operations that may be considered for recycling or reuse.
Although alternative public dose limits were considered, the initial control criteria in this report are
based on 1) a dose of 10 /u.Sv y"1 (1 mrem y"1) to a worker in a smelter or to an individual who uses
consumer products made from recycled materials, 2) a dose of 1 /iSv y"1 (0.1 mrem y"1) to an
individual downwind from a smelter used to process recycled metals, or 3) non-health impacts
associated with potential radiation effects on electronics or film.
Table 2 summarizes the limiting concentrations based on individual radiation dose for residual
contamination in (or on) recycled materials. For the radionuclides in Table 2, doses to smelter
workers or to users of consumer products provided the most restrictive (i.e., the smallest) derived
residual concentrations. This table shows the initial radiological control criteria for bulk materials,
in units of Bq g"1, for steel, aluminum, copper, and concrete, and the initial control criteria for
surface contamination in units of Bq cm"2.
-------
JAERI-Conf 95-015
TABLE 2. Draft Radiological Control Levels Based on an Individual Dose of 10 ptSv y'1
for Recycling and Reuse of DOE Metals or Equipment Containing Residual
Radioactive Contamination^
Radionuclide
3H
Mc
36Q
41Ca
MMn
55Fe
"Co
^Co
63M
65Zn
79Se
90Sr
93Zr
MNb
"Tc
106Ru
110mAg
I25sb
129l
134Cs
137Cs
l44Ce
147pm
151Sm
152Eu
154Eu
226Ra
228Th
229Th
23ftrh
Steel
2.1E+05
7.0E+03
NA«
NA
4.8E-01
5.6E+02
3.3E+QG
1.6E-01
1.9E+04
6.3E-01
1.6E+03
9.3E+01
1.4E+03
2.6E-01
4.8E+03
1.6E+OQ
1.4E-01
8.9E-01
4.8E+01
2.3E-01
7.0E-01
2.7E+01
2.9E+03
3.7E+03
3.4E-01
3.3E-01
7.4E+00
4.1E-01
7.0E-02
4.8E-01
Bulk
Contamination
(Bq g1)
Aluminum
3.3E+05
9.6E+03
NA
NA
1.3E+00
9.3E+02
8.9E+00
4.8E-01
3.7E+04
1.8E+00
2.5E+03
1.6E+02
7.0E+03
7.4E-01
UE+Q4
4.8E+00
4.1E-01
2.4E+00
4.4E+01
6.7E-01
2.0E+00
5.9E+01
1.5E+04
1.9E+04
9.6E-01
9.2E-01
1.9E+01
2.0E+00
3.6E-01
2.3E+00
Copper
3.3E+05
9.6E+03
NA
NA
7.0E+00
7.8E+03
5.2E+01
2.5E + 00
3.7E+04
9.6E+00
2.5E+03
1.6E+02
7.0E+03
4. IE -l-OO
1.6E+04
2.5E+01
2.1E+00
1.3E+01
7.4E+01
3.6E + 00
l.OE+01
4.4E+02
1.5E+04
1.9E+04
5.2E+00
4.8E+00
1.9E+01
2.0E+00
3.6E-01
23E+00
Concrete
6.3E+06
2.1E-I-04
6.3E+02
3, IE +02
2.2E-01
1.4E+02
2.9E+00
6.3E-02
2.0E+05
2.5E-01
2.7E+04
1.2E + 03
5.2E+03
1.3E-01
5.2E+03
8.9E-01
6.3E-02
4.8E-01
2.0E+01
1.1E-01
3.6E-01
2.2E+01
9.3E+03
1.3E+04
1.4E-01
1.4E-01
4.8E+01
1.3E + 00
2.3E-01
1.5E+00
Surface
Contamination
(Bq cm'1)
Tools and
Equipment
9.6E+04
2.9E+03
NA
NA
1.2E+02
5.6E+03
6.7E+02
4.1E+01
1.1E4-04
1.2E+02
7.4E4-02
4.8E+01
2.5E+03
6.3E+01
4.4E+03
1.6E+02
3.6E+01
2. IE +02
2.2E+01
3.6E+01
7.4E+01
2.4E+02
4.4E + 03
8.9E+03
8.1E+01
7.4E+01
5.2E+00
1.5E+00
2.6E-01
1.7E+00
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JAERI-Conf 95-015
TABLE 2. (Cont'd)
Radionuclide
232Tn
232y
233U
234TJ
2351J
238u
237Np
238pu
2»pu
240pu
241pu
241Am
Steel
1.1E-01
1.8E-01
9.6E-01
9.6E-01
1.0E400
l.OE+00
2.4E-01
4.1E-01
3.7E-01
3.7E-01
2.1E+01
2.2E-01
Bulk
Contamination
(Bq g1)
Aluminum
5.5E-01
9.2E-01
4.8E+00
4.8E+00
5.2E+00
5.2E+00
1.2E+00
2.0E+00
I.8E+00
1.8E+00
1.1E+02
1.1E+00
Surface
Contamination
(Bq cm'1)
Copper
5.5E-01
9.2E-01
4.8E+00
4.8E+00
5.2E+00
5.2E400
1.2E+00
2.0E+00
1.8E+00
1.8E+00
1.1E+02
1.1E+00
Concrete
3.5E-01
5.9E-01
3.0E+00
3.0E+00
3.2E+00
3.2E+00
7.8E-01
1.3E+00
1.2E+00
1.2E+00
6.7E-1-01
7.4E-01
Tools and
Equipment
3.7E-01
7.8E-01
4.1E+00
4.1E+00
4.1E+00
4.1E+00
6.3E-01
L7E4-00
1.5E+00
1.5E+00
8.9E+01
5.9E-01
(a) Calculations were made using the EXTDF module from the GENII
Software System [3] and selected scenarios based on the methods in IAEA
Safety Series No. III-P-U [1].
(b) "NA" indicates that this concrete activation product was Not Applicable to
this scenario and was considered only for concrete recycle scenarios.
Doses to the public downwind of a smelter are estimated using the generic data on
atmospheric dispersion and medium-high population density in the U.S. Environmental Protection
Agency's CAP88-PC software. Doses were calculated to the maximally exposed individual (MEI)
downwind of a smelter, assuming a unit release. For ali radionuclides considered (except 238U), the
individual doses were more restrictive than the collective doses to the downwind public.
Also evaluated were non-health-related impacts industry may experience from residual
contamination in recycled metals, such as those used in the electronics and film industries. Upon
investigation, we found that most electronic components can withstand doses well in excess of the
DOE individual dose limit. Thus, recycling the materials considered in this report at or below the
contamination levels indicated Table 2 would have little impact on the electronics industry. On the
other hand, use of recycled metals were found to have potential impacts on the film industry.
Table 3 summarizes the limiting concentrations in recycled materials based on a 2/u-Gy (0.2 mrad)
exposure to film stored for one month [12] in a box constructed of either undiluted steel or concrete
for each of the 42 radionuclides considered. This table shows the initial radiological control criteria
(Bq g"1) for bulk materials for steel and concrete both with and without a non-contaminated 0.5-cm
lead lining. The initial control levels for film are more restrictive than those derived from doses to
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JAERI-Conf 95-015
TABLE 3. Draft Radiological Control Levels Based on 2 ^.Gy Exposure to Film Stored
for One Month
1.3E-01
1.3E+02
9.6E-01
4.4E-02 •
l.OE+05
1.8E-01
l.OE + 04
2.8E+02
1.3E+05
7.4E-02
1.5E+03
4.4E-01
4.1E-02
2.5E-01
3.2E+01
6.7E-02
1.9E-OI
8.1E+00
3.0E + 03
1.9E+04
9.6E-02
9.2E-02
2.7E+01
6.7E+01
1.8E + 00
2.6E+02
3.3E+02
2.3E+02
3.4E+02
Lead-lined
Steel
8.5E+10
NA(e)
NA(e)
2.5E-01
2.4E+02
7.4E-02
2.8E-01
5.6E+04
1.6E-01
2.5E+07
1.3E+00
7.8E-02
7.8E-01
SpA
1.4E-01
4.8E-01
8.9E+04
2.0E+08
SpA(d>
1.8E-0!
1.7E-01
3.7E+04
l.OE+04
7.0E+02
1.4E+07
L3E+07
5.2E + 06
Concrete
6.3E+07
4.4E+03
2.0E+02
4.4E+01
l.OE-01
2.0E+01
1.1E+00
3.1E-02
4.8E+04
1.2E-01
5.6E+03
3.6E+02
5.6E+04
5.9E-02
1.2E+03
4.1E-01
3.0E-02
2.1E-01
5.5E+00
5.2E-Q2
1.6E-01
8.1E + 00
2.2E+03
3.5E + 03
6.7E-02
6.7E-02
3.2E+01
3.6E + 01
1.6E+00
5.9E+01
6.3E+01
4.4E+01
1.2E+02
Lead-lined
Concrete
3.0E+1I
1.7E+04
1.9E-01
2.0E+02
4.8E-02
1.8E-01
1.2E+05
1.3E-01
6.7E+07
1.1E+00
5.5E-02
6.3E-01
1.1E-01
4.1E-01
1.2E+05
3.7E+OS
1.2E-01
1.2E-01
3.6E+04
1.1E+04
7.8E+02
1.9E+07
1.8E+07
7.0E+06
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JAERI-Conf 95-015
TABLE 3. (Cont'd)
Initial Radiological Control Levels for Bulk Contamination
(Bq g'1)
Radionuclide
234LF
235TJ
238y
^Np
238pu
239pu
240pu
241pu
241Am
Steel
2.8E+02
1.1E+00
3.7E+02
7.4E+00
3.0E+02
5.9E+02
3.2E+02
4.1E+07
l.SE+01
Lead-lined
Steel
2.4E+07
4.1E402
SpAW
1.5E+G4
SpA^
2.QE+G7
SpA^
...(»)
SpA
Concrete
5.2E+Q1
1.2E+00
6.3E+01
4.1E+OQ
4.8E+01
1.2E+02
5.2E+01
1.9E+Q7
5.9E+00
Lead-lined
Concrete
3.2E+07
4.4E+02
SPA^d>
1.6E+04
SpA(d)
2.7E+07
SPACd)
...(b)
SPA(d)
(a) Calculations were made assuming that the film was stored for one month in
a rectangular container made from either steel or concrete, with or without
lead shielding lining (0.5 cm thickness) the box. The radiological control
levels were determined based on the 2 fj, Gy (0.2 mrad) limit recommended
by the NCRP [12] for diagnostic x-ray film,
(b) For radionuclides having no gamma emissions, the lead lining reduced the
dose to zero resulting in initial control levels that approached infinity. This
is represented by (—) in the table.
(c) "NA" indicates that this concrete activation product was Not Applicable to
this scenario and was considered only for concrete recycle scenarios.
(d) "SpA" indicates calculated control level exceeds the specific activity possible
for the radiormdide shown.
smelter workers or to consumers for the photon-emitting radionuclides. This result is considered to
be preliminary because of the highly conservative assumptions used in the analysis and because it is
unlikely that storage areas for film would be constructed exclusively of undiluted (i.e., 100%),
recycled steel or concrete. Further evaluation of the assumptions and data associated with the film
scenario are underway.
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JAERI-Conf 95-015
REFERENCES
[1] International Atomic Energy Agency (IAEA), 1992. Application of Exemption Principles to
the Recycle and Reuse of Materials from Nuclear Facilities. Safety Series No. 111-P-l.l.
Vienna, Austria.
[2] Blizzard, E. P., A, Foderaro, N. G. Goussev, and E. E. Kovalev. 1968. "Extended Radiation
Sources (Point Kernel Integrations)." In Engineering Compendium on Radiation Shielding,
Voi. 1. pp. 363-386 ed, R. G. Jaeger, New York.
[3] Napier, B.A., R. A. Peloquin, D. L. Strenge, and J. V. Ramsdell. 1988. GENII - The
Hanford Environmental Radiation Dosimetry Software System. PNL-6584, Vol. 1-3, Pacific
Northwest Laboratory, Richland, Washington,
[4] International Commission on Radiological Protection (ICRP). 1977-1982. Limits for Intakes
of Radionuclides by Workers. ICRP Publication 30 Part 1 (and subsequent parts and
supplements), Vol 2 No. 3-4 through Vol. 8, No. 4. Pergamon Press, Oxford.
[5] O'Donnell, F. R., D. C. Kocher, O. W. Burk, and F. H. Clark. 1981. CONDOS-II - A Tool
for Estimating Radiation Doses from Radionuclide-Containing Consumer Projects.
NUREG/CR-206 (ORNL/NUREG/TM-454), U.S. Nuclear Regulatory Commission,
Washington, D.C.
[6] Parks, B. S. 1992. User's Guide for CAP88-PC. Version 1.0. 402-B-92-001. U.S.
Environmental Protection Agency, Las Vegas, Nevada.
[7] 54 FR 51695. 1990. "National Emission Standards for Radionuclide Emissions from
Department of Energy Facilities." 40 CFR 61, Subpart H, Federal Register. U.S.
Environmental Protection Agency, Washington, D.C.
[8] 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. Commission of the European Communities, Luxembourg.
[9] U.S. Nuclear Regulatory Commission (NRC). 1980. Draft Environmental Impact Statement
Concerning Proposed Rulemaking: Exemption from Licensing Requirements for Smelted
Alloys Containing Residual Technetium-99 and Low-Enriched Uranium. NUREG-0519.
Washington, D.C.
[10] Kennedy, W. E, Jr., and D. L. Strenge. 1992. Residual Radioactive Contamination from
Decommissioning: Technical Basis for Translating Contamination Levels to Annual Total
Effective Dose Equivalent. NUREG/CR-5512 (PNL-7994). U.S. Nuclear Regulatory
Commission, Washington, D.C.
[11] Messenger, G. C., and M. S. Ash. 1986. The Effects of Radiation on Electronic Systems.
Van Nostrand Reinhold Co., New York.
[12] National Council on Radiation Protection and Measurements (NCRP). 1989. Medical
X-Ray, Electron Beam and Gamma-Ray Protection for Energies up to 50 MeV (Equipment
Design. Performance and Use). NCRP Report No. 49. National Council on Radiation
Protection and Measurements, Washington, D.C.
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JAERI-Conf 95-015
3-9 Effects on Radiation Sensitive Instruments
from Recycling of Contaminated Metal
Hideaki Yamamoto and Shohei Kato
Department of Health Physics
Japan Atomic Energy Research Institute
Abstract
The authors are developing a computer code aiming to analyze industrial impacts
from recycling radioactively contaminated scrap metal (RSM). The code comprises
mathematical models simulating possible adverse phenomena in radiation sensitive
instruments whose components are assumed to be made of RSM. Microelectronics
devices (large scale integrated circuit (LSI) chips) was identified as a radiation
sensitive instrument for the analysis,
The computer code predicts frequencies of LSI malfunction, and provides the bases
for developing radiation criteria for recycling of RSM. A preliminary result is
compared with health-related radiological criteria.
1. Introduction
It has been observed that natural background radiation can induce various adverse
effects on the performance of certain instruments which are used in
scientific/industrial activities or daily life. These radiation sensitive instruments might
also be exposed to radiation when radioactively contaminated scrap metal (RSM) are
released from radiation control and subsequently introduced to their production. RSM
may be applied directly to the component material of these instruments. Otherwise
residual radioactivity in RSM may be transferred to these instruments as
contamination if the RSM exists in their manufacturing process.
In addition to human health and environmental protection criteria, acceptance
radiation criteria should be developed in order to prevent adverse effects of residual
radioactivity in RSM on radiation sensitive instruments and subsequent deleterious
impacts on human society.
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2, Large scale integrated circuits (LSI) as Radiation Sensitive Instrument
The authors indicated that radiation sensitive instruments include large scale
integrated circuits (LSI) [1]. LSI as a component of computers is indispensable to the
human activities today, thus any adverse effects on LSI from residual radiation from
RSM could affect largely our society.
The basic construction of LSI is: a silicon chip, which generates the function of the
LSI, a lead frame connecting the silicon chip to external circuits, and a package
protecting the inner circuits on the silicon chips. Lead frames are made of metallic
alloys, typically those of copper, iron or nickel. Ceramics (Ai20a) is used in making
packages [1]
A radiation effect on LSI due to the internal exposure to residual radioactivity has
been studied. Naturally-occurring uranium or thorium as impurities in ceramic
packages irradiate silicon chips with alpha particles, and induce malfunction of the
LSI [2].
Dynamic random access memories (DRAM) could be the most sensitive LSI to
alpha-particle irradiation because their memories are stored in memory cells of their
silicon chips as electric charges. This malfunction is called a "soft error"; an upset
of memory data by the passage of alpha particles through the chips. The upset is
caused by electrons flowing into the memory cell from ion-pairs generated by the
alpha particle.
Essentially, a rate of soft errors is proportional to a flux of alpha particles at the
surface over the memory area. In addition, the soft error rate increases as the
integration density of DRAM increases [1], [3]. Soft error rates are often measured
g
by a unit of FIT (Failure unlT), One FIT means one failure per 10 hours per DRAM
chip. The relation can be expressed in the form;
SER = c(D) Fa (1)
where SER = soft error rate [FIT],
c(D) = constant depending on DRAM integrity density D
[FIT/(n/(cm2h))],
Fa = alpha-particle flux [n/(cm2h}].
To assume the value of c (D), the authors extrapolated the data reported in the
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JAERI-Conf 95-015
8 9 10 2
reference [3]. For example, values of 10 , 10 and 10 FIT / (n / (cm h )} are
assumed for 4 Mbit, 16Mbit and 64Mbit DRAM, respectively.
3. Impacts on DRAM from Recycling Contaminated Metal
In order to evaluate hypothetical impacts from recycling contaminated metal, lead
frames, the major metallic component of DRAM, were assumed to be contaminated
with residual radioactivity: nickel contaminated with uranium-238 was assumed to be
recycled for making the lead frames.
The geometry of DRAM structure was simplified to estimate alpha-particle flux over
the surface of a memory celi: a lead frame was expressed by a disk (diameter; r)
having the equivalent area to the original DRAM, and a memory cell - a square over
the disk with a distance of H. The alpha-particle flax of this geometry was given by
the expression:
Fa = 1800 Sa In (sec arctan (r/H)) F
(2)
where Fa = alpha-particle flux [ n / (cm h}]
Sa = residual radioactivity concentration [Bq/ cm ]
r = equivalent diameter of lead frame [ cm ]
H = distance between lead frame and memory cell [ cm ]
F = penetration factor [-]
Assuming the contamination of the lead frame in levels of 1 Bq/g and 10 Bq/g, SER
of the DRAM with various integrated densities was estimated as follows:
Table 1 Soft Error Rate of DRAM
Contamination
Pq/g]
1
10
4 Mbit DRAM
[FIT]
4
40
16 Mbit DRAM
[FIT]
50
500
64 Mbit DFIAM
[FIT]
500
5000
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JAEW-Conf 95-4)15
4. Acceptance criteria
It seems practical to accept impacts from recycling contaminated metate if the levei
of the impacts is comparable to that from a similar naturally-occurring phenomenon.
In this context, the SER due to the contaminated nickel recycling could be acceptable
if the SER is comparable to that induced from natural uranium or thorium in the
package material.
The LSI industry had made efforts to reduce the concentration of natural alpha-
emitters in the LSF materials to such a concentration level as to induce an alpha-
particle flux of less than 0.01 n / (cm h) at the surface over the memory area [4],
This level of flux corresponds to a SER of 100 to 1000 FIT in 64 Kbit DRAM, which
used to be the major product [1], [2], Accordingly, it would be reasonable to assume
a SER of 100 FIT as the acceptable SER criterion for recycling contaminated nickel
in the lead frame production.
The estimation in Table 1 inspires that for 16 Mbit DRAM, coming major product in
the DRAM manufacture, the radioactivity concentration in the lead frame of an order
of one Bq/g could be an acceptable residual radioactivity for protecting the DRAM
industry against impacts from recycling contaminated nickel.
5. Discussion
The estimated acceptance criterion for nickel recycling can be compared to the
release criteria or the clearance levels of radioactivity concentration derived from
the human health analysis. For example, the International Atomic Energy Agency
(IAEA) reports the clearance level of 0.1 to 10 Bq/g for alpha-emitters in recycled
metal to derived from an annual individual radiation dose of 10 micro Sv [5].
In this preliminary comparison, the acceptance level for the DRAM industry is
higher than the IAEA clearance level. This means that the DRAM industry would
not be affected by the health-related unrestricted release of contaminated metal.
6. Acknowledgment
The authors wish to thank Dr. Kentaro Umeda from WOODLAND corporation for his
help in many aspects of this work.
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JAERI-Conf 95-015
7. References
[1] Kato, S. et al.: "Effects of residual radioactivity in recycled materials on
scientific and industrial equipments", EPA 520/1-90-013, pp. 266-280 (1990).
[2] May, T. C. et al.: "A new physical mechanism for soft errors in dynamic
memory", IEEE Trans. ED-26 (1), pp. 2-9 (1979).
[3] Hirai, S.: "Determination of uranium, thorium and alpha emitters in
semiconductor", BUNSEKi 9, p. 639 (1988).
[4] Shimohigashi, K. et al.: "Alpha-particle-induced soft error in semiconductor
memories", Denkitsushin-gakkaishi 63.7, pp. 742-744 (1988).
[5] International Atomic Energy Agency: "Application of exemption principles to the
recycle and reuse of materials from nuclear facilities", Safety Series No.111-P-
1.1 (1992).
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JAERl-Conf 95-015
3-10 Second EPA/JAERI Workshop
on
Residual Radioactivity and Recycling Criteria
METAL RECYCLING TECHNOLOGY AND
RELATED ISSUES IN THE UNITED STATES, A BNFL PERSPECTIVE
By
Phillip Bradbury, Vice President
Scott Dam, Program Manager
Wayne Starke, Manager, Oak Ridge Operations
Facilities Ownership, Management, and Operations
BNFL Inc.
Fairfax, Virginia, U.S.A
ABSTRACT
Radioactively contaminated metallic materials comprise a large part of the potential waste
products which result from nuclear facility repair, refurbishment, and decommissioning.
United States Government (Departments of Energy and Defense) facilities, U.S. nuclear
power plants, and other commercial nuclear fuel cycle facilities have large inventories
of radioactive scrap metal which could be decontaminated and recycled into useful
radioactive and non-radioactive products. Residual radioactivity and recycling criteria is
needed to avoid the high cost of disposal and the waste of natural resources.
In the United Kingdom, BNFL has decommissioned the gaseous diffusion plant at
Capenhurst and has recycled a large fraction of the metallic scrap into the metals market.
Other structural materials have also been released as uncontaminated scrap. U.K. release
criteria for residual radionuclide contamination have been applied to these operations. A
variety of techniques were utilized to size reduce large components, to remove
radioactivity, and to survey and release these materials. These methods and the
application of release criteria has a direct relationship to methods which would be
applicable in the U.S. and in other countries.
This paper will describe the specific U.K. technology and experience in the
decontamination, recycle, and release of scrap metal. It will also describe the U.S.
environment for metal recycle, including the volumes and levels of contamination, and
the current and proposed release criteria. Comparisons will be presented between the
U.S. and U.K., both in technology and methodology for recycle and in regulatory criteria
for residual radioactivity and material release and for ultimate decommissioning. The
paper will then provide suggested approaches and criteria for U.S. recycling and
decommissioning.
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INTRODUCTION
BNFL Inc. is a wholly owned subsidiary of British Nuclear Fuels (BNF) pic.
BNFL Inc. is actively engaged in offering over four decades of experience in the
engineering, construction, operation, and decommissioning of facilities that comprise the
total nuclear fuel cycle, BNFL Inc. 's staff of professionals represent many years of
experience working on programs for the Department of Energy (DOE) and its
management and operating contractors, as well as with the private sector nuclear power
industry.
A major part of BNFL Inc.'s expertise is total waste management. The
Company's owner, BNF pic, is a company concerned with providing world leading
nuclear fuel cycle services. Managing radioactive waste so as to minimize hazards to
people and the environment is a key competence and the company has significant
capability in this important area of the fuel cycle. BNF pic manages and operates five
major sites in northwest England and southern Scotland. These sites generate various
forms of radioactive waste, including liquids, gases, and solids in various levels of
radiation from low-level to high level. All the wastes generated at these sites are
collected, treated, and stored or disposed. BNF pic has over three decades of experience
in waste management and treatment resulting hi state-of-art technologies that not only
minimize the final waste, but also package it in a form that is safe to the environment
and to the public. BNF pic actively engages in transferring its experience and technology
to BNFL Inc. for use in North America.
BNFL Inc. and BNF pic are actively engaged in metal recycle projects taking
radioactive contaminated scrap metal (RSM) and decontaminating it for reuse or
manufacturing products from slightly contaminated metal. Experience gained at the
Capenhurst gaseous diffusion plant decommissioning and the metal recycle activities is
being applied in the U.S. In the U.S., BNFL Inc. is actively involved in metal recycle
projects teamed with Manufacturing Sciences Corporation in their National Conversion
Pilot Project at the DOE Rocky Flats facility, in a demonstration of the beneficial reuse
of stainless steel for the DOE Savannah River facility, and in proposed work for the Oak
Ridge Gaseous Diffusion Plant.
The objective of this paper is to assist in criteria development for release of
radioactive scrap metal and in the development of radiological standards for release of
decommissioned facilities.
U.S. METALS PROBLEM
In the US there is a large quantity of radioactive scrap metal (RSM) which could be
recycled. The sources of this scrap metal are presented below along with approximate
quantities of these metals.
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METAL SOURCES
There are three main sources that represent significant quantities of metals suitable for
recycle,
1. Nuclear Power Plants represents the largest current source of RSM. Nuclear
power plants currently generate from regular plant overhauls radioactive scrap
metal such as old heat exchangers, piping, pipe hangers, and spent fuel racks.
Since this metal is not generally free-released directly by the plants to the open
market, few choices exist:
(a) Bury the metal at an approved low-level waste landfill such as Barnwell
where the current cost is from $3.28/lb to $8,00/lb depending on status
as either a sited or unsited waste generator with the Southeast Compact.
(b) Send the metals to a licensed radioactive scrap metal processor such as
SEG or Quadrex (American Ecology).
(c) Dispose directly after specific approval from the NRC (10 CFR 20.2002 -
formerly 20.302). About 25 such approvals have been obtained
The total volume of metal generated from the 109 operating nuclear power plants
last year was 29 million pounds,
2. DOE Facilities, There currently exist stockpiles of metal at various DOE sites
across the country. Significant quantities of additional metals will come from
facilities as they are decommissioned and dismantled. The current estimate of the
DOE facilities for all metal including that which is in stockpiles is 1.7 to 3.6
billion pounds! As facilities start to be decommissioned/dismantled and the metal
is sent to be processed, the annual quantity should increase to 50-90 million
pounds/year. DOE disposal costs vary from $5 to $10 per cubic foot at Nevada
(not the life cycle cost) to about $50 per cubic foot per year at Idaho for low-level
waste. The latter reflects the continual costs of maintenance of the
disposal/storage facility.
3. Nuclear Plant Decommissioning. As power plants reach the end of their current
40 year operating licenses, they will be decommissioned. The metal that arises
from these decommissionings will be spread over the next 50+ years to the year
2053. There are currently 109 nuclear power plants. The first plants to be
decommissioned include Yankee Rowe, Trojan, and Shoreham, This market
might be stretched if some of the plants successfully extend their NRC operating
licenses beyond the 40 years of if the power plants get acceptance to put the
majority of the facility into mothballs. It is currently estimated that 1.5 billion
pounds of steel exist in the nuclear power plants that could be processed. The
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JAERI-Conf 95-015
quantity of this material should start to be significant by 2003. It is estimated that
the average annual flow of metal should be over 25 million pounds per year.
TIMING
The sources of RSM market is currently increasing in all three segments. The nuclear
power plant RSM is expected to level off in five years and will start to decrease in the
year 2003 as more plants are decommissioned, eliminating the generation of metal waste
from operations. RSM from nuclear plants will cease altogether by 2042 if no
change/extension occurs in the present 40 year license requirement.
Metal from the DOE is a difficult source to predict because of the problem DOE has
with free release and the fact that the Nevada Test Site is currently charging between $5 -
$10/cu ft to dispose whereas the actual life cycle cost is more like $35/cu ft. Until the
DOE decides on a workable free release policy (including liability issues for their
management & operating contractors), the material for processing will be limited to
acceptable internal end uses, such as containers, and to limited free-release from specific
sites.
The third segment, RSM from the decommissioning of nuclear power plants, will be
erratic in nature untE about 2014. The flow of metal should continue until 2053 when
the final plants will be decommissioned completely.
DOE SCRAP METAL VOLUMES AND LOCATIONS
Figures 1 and 2 show the quantities of DOE scrap metal by type and by location. The
specific quantities actually in inventory vary depending on which survey you believe.
Total quantities range in estimates from 400,000 tons to 1,800,000 tons. The figures
showing the breakdowns are from a quick, telephone survey and may not reflect the
actual quantities nor the correct distribution of metals or locations. It is included in this
paper for illustrative purposes only to show a wide variety of materials and locations.
U.K. EXPERIENCE IN METAL RECYCLE
CAPENHURST DIFFUSION PLANT DECOMMISSIONING
The Capenhurst Diffusion Plant was built in the early 1950s at which time it was the
largest industrial building in Europe under a single roof, measuring 1,000 yards in length
and 160 yards in width. It was originally built to produce highly enriched uranium for
military purposes but this came to an end in the early 1960s when it was converted and
extended for low enriched uranium production for civil use. After a further twenty years
of life, the plant was shut down in 1982. By this tune centrifuge enrichment plants were
built and operating at Capenhurst and the diffusion plant was no longer economical.
Since that time a program of decommissioning and dismantling has been in progress
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with an process plant material having been recycled or put in the appropriate form for
disposal. Part of the building has already been demolished and a new centrifuge plant
constructed on the same land. The remainder will remain in tact so long as there is
economic use for it.
Post operational work to empty and cleanup the plant was carried out to leave as little
residual contamination as possible by the use of a fluorinating agent to convert solid
deposits to volatile fluorides which were pumped away. Further cleanup operations were
carried out on the static plant to locate and deal with any significant buildup of solid or
gaseous pockets of. contamination remaining within the 4,800 process stages and 1,200
miles of interconnecting pipework.
A safe system of working was then established for the dismantling of the plant, which
included survey measurements throughout the plant to ascertain what amounts of activity
were left which would have to be dealt with at the decontamination and disposal stages.
The initial phase of dismantling involved the cutout, removal, and storage of large
numbers of components, including compressors, coolers, valves, large diameter
pipework, and the large process stage units.
Since 1984, up to 6,500 tons of major plant components have been successfully stored
outdoors, including 850 tons of steel stage units. The 18,000 tons of structural steel have
been dismantled and disposed of as clean scrap. Process pipework and the large process
stage units have been cut up into lengths and sections suitable for the decontamination
process. A variety of methods have been employed in cutting up the process plant.
These include both hot and cold cutting, automated wherever feasible. Robotically-
controlled plasma cutting has been employed to volume reduce a total of 5,300 tons of
large aluminum stage units. Steel shells have been separated from a further 1,000 tons
of stage units using remotely-controEed oxyacetylene methods. Both processes have been
carried out in specially developed ventilated enclosures. The bulk of cold cutting has
been done by automated metal cutting techniques (including band sawing).
DECONTAMINATION TECHNOLOGY
Chemical treatment for the removal of uranium and uranium breakdown products is a
well established process and significant quantities of aluminum and steel have been
cleaned and recycled to the metals market. However, a large section of the plant had
been exposed to reactor recycled feed material and contaminated with Np-237 and Tc-99,
the latter being particularly difficult to remove. One disposal option was land burial at
Drigg (near BNFL's reprocessing plant at Sellafield), the only currently available low-
level waste site in the U.K. The large volumes of material involved (some tens of
thousands of cubic meters) which was equivalent to several years of remaining Drigg
lifetime made this option unacceptable. Furthermore, costs would have been very high.
It was decided to develop a procedure for effective decontamination to very low,
deminimis levels at which it could be sold as uncontaminated scrap metal. An extensive
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JAERI-Conf 95-015
laboratory and pilot plant investigation was successfully carried out and a foil scale
decontamination plant has been built and is in operation. Over 9,000 tons of aluminum
and steel has been decontaminated to date and recycled to the scrap metal market. The
flowsheet design for the plant was based on the plant discharges having a negligibly small
impact on the environment and on the U.K. statutory regulations for recycling scrap
metals to the open market. The process is one of successive stages of decontamination
in process liquors which are subsequently treated by ion exchange methods. This results
in the activity being transferred to relatively small and manageable volumes of solid and
liquid residues and low activity solid and liquid wastes.
U.K. SCRAP METAL RELEASE CRITERIA
Criteria for free release of scrap metal was established for the U.K. based on the UK
Radioactive Substances Act, 1960 and revised 1988. The following are these criteria:
All radionuclides - (total alpha, beta and gamma) = less than 0.4 Bq/gm (0.011 nCi/gra)
Exemption for naturally-occurring uranium (alpha) of 11.1 Bq/gm (0.3 nCi/gm)
For decommissioning, the buildings and site must be returned to the local area naturally-
occurring background radiation levels. Rubble and other non-recyclable scrap must also
be at natural background before releasing it to commercial landfills.
RESULTS FROM THE CAPENHURST METAL RECYCLE PROGRAM
In summary, the major achievements to date are as follows:
• Over 99.5% of the decommissioning materials (160,000 tons) was recycled as
clean material.
• Disassembled a complete diffusion plant comprising 4,800 process stage units,
1,200 miles of pipework, and associated plant items, including 3,500 tons of
electric motors.
» Management and control of the outdoor storage of 6,500 tons of contaminated
plant components for periods of time up to nine years.
• Developed hands-on and remotely controlled systems for volume reducing 8,000
tons aluminum, 23,000 tons steel, 70 tons copper, 300 tons aluminum bronze,
200 tons cupro-nickel, and 320 tons of nickel.
• Since 1983, over 5,000 tons of aluminum (including 950 tons from the highly
enriched section of the plant) have been prepared, decontaminated, and sold for
recycle into the metals market. Some 23,000 tons of steel (including 1,100 tons
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JAERI-Conf 95-015
from the highly enriched section) "have been prepared and sold dkectly to the
metals market as "clean" scrap, without the need for full decontamination.
» The 3,500 tons of contaminated electric drive motors (including 800 tons from
the highly enriched section of the plant) have been successfully treated and sold
for recycle in the metals market.
* Developed a process for separation of contaminated aluminum heat exchange
sheaths from cupro-nickel liner cooling tubes.
* Ancillary buildings, including 11 large cooling towers (13,000 tons), pumphouses,
electrical substation, and part of the main diffusion plant building were
completely raised to the ground producing 46,000 tons of concrete rubble for off-
site disposal.
* Five bays of the plant have been demolished to greenfield status. The recovered
site (of approximately six acres) was made available for the construction of a new
centrifuge enrichment building.
* The 850 tons of low-level contaminated waste have been dispatched for land
burial at the Company's open cast landfill site at Drigg.
» Over 88,000 cubic yards of material was able to be disposed of by recycling or
free release thus avoiding disposal as radioactive waste. At $2000 per cubic yard
($74 per cubic foot, much less than U.S. commercial rates), this resulted in a
savings of over $180 million. Additional savings came from not having to use
radioactive materials rules for much of the dismantling activity and not having to
package and transport materials as radioactive. This savings is more than the cost
of the actual decommissioning activity.
U.S. / U.K. REGULATORY COMPARISON
The rules and regulations under which the nuclear industry operates in the U.K. have
both similarities and differences with the rules and regulations in the U.S.
One area of significant difference is that in the U.K. the same agencies regulate both
radioactive and hazardous wastes. This eliminates areas of conflict of jurisdiction such
as occurs in the U.S. with the handling of mixed wastes. In the U.K. there is no
classification such as mixed waste. Also in the U.K. the nuclear waste category called
intermediate-level waste is between low-level waste and high-level waste. This
intermediate-level waste is currently being cement grouted and stored for the eventual
long term burial in an underground repository.
There are also some differences in the definition of requirements and their application.
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In general, the U.S. requirements are prescriptive and extremely detailed, while in the
U.K., the onus is placed on the owner of the waste to provide detailed protective criteria
to the satisfaction of the regulator. In some cases, these risk-based criteria are more
extensive than equivalent U.S. legal requirements.
Quantitative criteria for nuclear protection, upon which requirements in the two nations
are based, are the same. They are the International Committee on Radiological Protection
(ICRP) and the International Atomic Energy Agency (IAEA). However, these
organizations are applied differently. In the U.S. isotopic discharges are regulated by
concentration at the source, while a U.K. site is limited to an annual total discharge
quantity. In the U.K., deminimis levels .of radioactivity are defined which allow
innovative clean-up and recycling of materials; while in the U.S., deminimis is
effectively limited to that which can be shown to result in a dose to an individual of less
than 0.02 mSv/hr. Table I lists the discharge limits by radionuclide for gases and liquids
in the U.K. and the U.S.
On the environmental side, the U.K. has certain environmental release regulations based
on the use of surrounding arable lands for food production and water for drinking and
fishing. Since the U.K. has a relatively small land mass, it is more extensively populated
and farmed in the U.K, than in the U.S. This makes the environmental requirements
more stringent, but reasonably based on real risk rather than concepts in which no risk
is acceptable.
CURRENT U. S. REUEASE CRITERIA
The Energy, Environment, and Resource Center at the University of Tennessee was
directed by the DOE to research the establishment of effective environmental standards
for RSM. In a draft report published in May 1993, they gave a summary of the current
status.
There currently are no nationally applicable standards that facilitate the treatment and
recycle of RSM. Existing standards by the NRC do not appear to be adequate guidance
in the treatment of RSM.
The following standards by various agencies exist:
The International Atomic Energy Agency (IAEA) - Basic Safety Standards (BSS)
for specific isotopes of radioactivity-follow ALARA principles. In addition there
is 'The Application of Exemption Principles to the Recycle and Reuse of
Materials from Nuclear Facilities' which provides safe levels for the recycle and
re-use of steel, aluminum, and concrete. This set of principles includes
volumetric and surface contamination levels, but are based on an individual
dosage of one mrem per year, which is below what is considered practical and
achievable on an economic basis.
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JAERI-Conf 95-015
The International Commission of Radiological Protection (ICRP) - original
advocates of ALARA.
The Environmental Protection Agency (EPA) has developed standards for the safe
levels of radioactivity, as they apply to drinking water, and air quality. There are
also some recent guidelines pertaining to the treatment of soils and debris which
might have a bearing on RSM.
The Nuclear Regulatory Commission (NRC) developed the Regulatory Guide 1.86
in 1974. .This guide deals with surface contaminated materials that are
nonactivated. It is this guideline that the nuclear waste processors have used in
their licensed approval to free release metals from nuclear power plants. The
values are in DPM/lOOcm2. The Energy Act of 1992 was to encourage the NRC
to initiate the Enhanced Proposed Rule (EPR), an effort to establish cleanup
standards for the decommissioning of nuclear power plants. This effort could
result in changes or at least clarifications in how RSM is to be processed and
released for unrestricted release.
The DOE has Order 5400.5 which in part is based on NRC Regulatory Guide
1,86. It deals only with surface contamination. Unfortunately, when Leo Duffy
was with the DOE, he issued a memo stopping all release of DOE metals for
recycle. Even though he is no longer with the DOE, the sites are hesitant to
resort back to Order 5400.5. The only exception is Fernald which has been
sending steel to SEG to be made into shield blocks and to Quadrex for free
release. The RFP currently on the street for Building 7 at Fernald will result in
a contract for the decontamination and release or fabrication into boxes of over
700 tons of mild steel.
PROPOSED U.S. REGULATORY CHANGES AND COMMENTS/SUGGESTIONS
EPA and NRC are currently in the process of ralemaking actions to set radiological
standards for decommissioning. No effort is currently underway to the authors'
knowledge to develop volumetric contamination release standards for recycled materials.
Such standards are necessary to fully implement a scrap metal recycle policy.
Reasonable standards are needed for recycling and decommissioning. Reasonable
standards will promote their use. These standards must be measurable with current
technology cost-effectively and must be understandable. Where possible, the regulatory
agencies should establish release numbers which are directly measurable, e.g. nCi/gm
or DPM/100cm2. This approach is preferable than requiring each radioactive material
user to develop their own release numbers from site specific analyses for exposure to the
public. The experience of others should be utilized in developing these limits, especially
current scrap metal recycle faculties and decommissioned facilities. A separate set of
criteria which is for "restricted use" should be developed and allowed which will permit
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JAERI-Conf 95-015
a decommissioning to restrict the use of fhe facility/site without having to maintain an
active license for the site. For example, allowing the use of a former nuclear facility for
an industrial plant, with certain obvious restrictions. Material recycling and free release
standards are needed for volumetric contamination in addition to surface contamination.
Much of the DOE scrap metal has been volumetrically contaminated,
DERIVED BENEFITS
This paper has discussed the waste forms that are generated in the U.K. and how they
are handled. The paper also discussed the specific handling of radioactive scrap metal at
Capenhurst. Next a discussion was presented on the regulatory requirements in the U.K.
and how they are similar and different from the U.S.
Now that these subjects have been discussed, it is appropriate to tie them all together in
a discussion of what benefits can the U.S. derive from the U.K.'s nuclear industry?
First of all the U.K. is actively collecting, treating, and at least properly storing for
disposal all of its nuclear waste as it is being generated. In the U.S. this is not the case,
especially in the DOE communities. For the U.S. to get to this practice they need to
expedite the colective process of standardizing the discharge limits for all nuclear waste
materials as their release is related to the public health and the protection of the
environment. Risk based analysis can result in risk based criteria that can be restrictive
relative to the personal exposure and not on the absolute values.
Secondly, the U.K. has been able to actively recycle significant quantities of metal back
into society. The DOE is currently stalled with this effort because of their policies on
release of DOE metals. Huge quantities of metals are not being processed and recycled.
In the U.K. the Capenhurst facility has been able to generate jobs and revenue for the
nuclear industry while protecting the public and the environment.
Thirdly, in the U.K. land is at a premium and the country cannot afford to contaminate
and not clean up for reuse the land and facilities that have ceased their intended
operation. The Sellafield complex is on only 700 acres and is surrounded by
communities. This is in comparison to say the Savannah River Site which is on 192,000
acres of land and its closest community is Aiken, South Carolina, about 20 miles away.
Therefore, in the U.K. at the Sellafield Complex, when a new facility is to built, it is
designed with a specific life expectancy. It is also designed so as to be more easily
decommissioned and dismantled at the end of its useful life. When a new facility is to
be built, an old facility must be dismantled completely including any soil and ground
preparation. The cost to do the eventual D&D is factored into the facility as a life cycle
cost and is accrued over the life of the facility so that when the facility ceases to operate
there are sufficient funds available to perform the proper D&D.
Lastly, there is in the U.K. an integrated effort to treat the public and the environment
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JAERI-Conf 95-015
in a proper perspective. Rules and regulations are designed along the principles
recommended by the ICRP. These basic principles are:
* Any practice involving radiation exposure must be justified in relation to its
benefits.
« Any necessary exposures must be kept as low as reasonable achievable, economic
and social factors being taken into account.
* Radiation doses must not exceed recommended dose levels.
In addition to the U.K. experience, there are benefits demved from U.S. recycling and
decommissioning experience. The cost savings of having rules that can be complied with
is true not only for the owner but also for the public. We have numerous occurrences in
the U.S. of companies which declared bankruptcy and thereby forced the public to pay
for the cost of cleanup, whether it was low-level radioactive, or high-level waste. Having
reasonable rules will also avoid or reduce dismantling, packaging, transportation and land
disposal, increasing the safety to the public and to the workers.
There are additional benefits to recycling which should be considered. Most of these
benefits are true whether we are talking about radioactive scrap metal or domestic scrap,
such as aluminum soft drink cans. These benefits include the overall economic benefit,
reduces energy consumption, and minimizing the environmental impact of new mining
and material processing.
CONCLUSIONS/RECOMMENDATIONS
Recycling is an important part of the overall safe management of radioactive materials.
Reasonable standards for material recycle and free release, including radiological
standards for decommissioning, will assist in compliance and will be an overall benefit
to the public, workers, and industry. It is also be best use of our natural resources.
Specific volumetric and surface contamination standards for recycle and decommissioning
should be established such that they are definitive, can be measured with current
capabilities cost-effectively.
REBHRENCES
S.G. BAXTER and P. BRADBURY, "Decommissioning of the Gaseous Diffusion Plant
at BNFL Capenhurst," Presentation at WM92 Conference, Tucson, Arizona
(March 1-5, 1992).
PHDLJJP BRADBURY, MARCO S. COLALANCIA, JAMES R. CROSS, and JOHN
GRAHAM, "British and U.S. Regulatory Practices in Waste Management,"
Presentation at ER93 Conference, Augusta, Georgia (October 24-28, 1993).
IJLLY, M. JUDSON, TH, USDOE, et al., "Radioactive Scrap Metal Recycling in the
United States Department of Energy," Presentation atER93 Conference, Augusta,
Georgia (October 24-28, 1993).
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TABLE I
DISCHARGE LIMITS
Radionuclide
Total Alpha
Total Beta
Tritium
Carbon 14
Sulphur 35
Cobalt 60
Argon 41
Strontium 90
Krypton 85
Zirconium
Technetium 99
Ruthenium 106
Iodine 129
Iodine 131
Cesium 134
Cesium 137
Cerium 144
Plutonium 241
Plutonium Alpha
Americlum 241
Curium 242
Critical Group
Dose
GASEOUS DISCHARGES
U.K. Site
Authorization
(GigaBequerel/yr)
10.5
210
1,100,000
30,000
220
18
3,700,000
12
10,000,000
69
88
220
76
47
3,4
4.5
4.5
450 microSieverts
U.S. Discharge
Limit
(GBq per in3)
5.55 x 10*
3.7 x 10-10
4.81 x 10-"
6.29 x 10-'3
6.29 x 10*
7,03 x ID'13
3.7 x 10-s
1.26 x 10-"
3.37 x 1043
1.67 x 10*
7.03 x lO'13
3.7 x 10-'2
3,0 x 10-iS
1.96 x 1042
LIQUID DISCHARGES
U.K. Site
Authorization
(TeraBequerel/yr)
10
500
3,500
4
8
35
ISO
10
170
0.4
10
110
22
170
7
3
219 microSieverts
U.S. Discharge
Limit*
(TBq per liter)
1.11 x 10-7
2.96 x 10*
1.11 x 10*
1.11 x 10-"
2.96 x 10-*
7.4 x 10*
2.96 x 10*
2.22 x 10-'2
3.33 x lO40
7.4 x lO40
3.7 x 10-'°
7.4 x 10"»
1.48 x ID40
U.S. Liquid discharge are given for the more restrictive soluble limit!, except for Cobalt whose limit is slightly more restrictive for its insoluble
form.
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JAERI-Conf 95-015
S(eel (35.1%)
(12.0%) Njcke(
\
\
Copper
^ (oe%) Brass, Lead,
Monel
(4.1%)
Aluminum
(4i.3%) Mixed
Figure 1 Quantities by Type
9.7%)
Oak Ridge
Paducah (10,2%)
Portsmouth
(2.0%) Ferna|d
(1.4%)
Other
(29.4%) K-25
Figure 2 Quantities by DOE site
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3-11
MELTING TESTS FOR RECYCLING OF RADIOACTIVE METAL WASTES
Hisashi Nakamura, Katsuo Kanazawa, and Kazuo Fujiki
Department of Decommissioning and Waste Management
Japan Atomic Energy Research Institute
Tokai-mura, IbaraM-ken 319-H Japan
ABSTRACT
To allow the future recycling of
decommissioning wastes to promote smoothly,
melting tests were conducted using metal wastes
and simulated wastes with radioisotopes. The test
results indicate that the transfer behavior of
radionuclides during melting is basically understood
by considering the volatility and oxidizable
tendency of each radiomielide. The partitioning of
some radionuclides into products was influenced by
the melting process of wastes. The radioactivity
distribution in ingots was uniform regardless of the
kinds of radionuclide.
1. INTRODUCTION
In some European countries and the United
States, recycling of wastes arising from
decommissioning, converting wastes to beneficial
resources, is being conducted to reduce the amount
of wastes and to ease the problem of disposal [1-4].
The criteria development for the recycling of wastes
has also been addressed by some organizations [5,6]
to make a internationally consistent criteria, because
the number of nuclear facilities to be
decommissioned is being increased.
In Japan, decommissioning of some nuclear
power plants currently operating is expected to
begin early in the next century. At this juncture, a
large volume of low-level radioactive wastes will
be produced by such decommissioning and
recycling of these waste will be very important.
However, research and development on the
recycling of radioactive wastes in Japan has not
been addressed enough, and there is a .little data
concerning the transfer behavior of radionuclides
during melting.
With this incentive, the Japan Atomic Energy
Research Institute (JAER!) has been conducting
melting tests in preparation for the recycling of
radioactive metal wastes under-the contract with the
Science and Technology Agency [7-10]. The
objectives of the tests are the followings: 1) to
determine the transfer behavior of radionuclides
during melting process, 2) to accumulate basic data
for metal recycling through melting tests, 3) to gain
experience on melting of actual metal wastes.
This paper describes the partitioning of
radionuclides among the ingot, slag, and off-gas,
some influencing factors on it, and the radioactivity
homogeneity of ingots.
2. MELTING TESTS
2.1 Progress of Melting Tests
As shown in Fig.l, the test program for melting
radioactive metal wastes began in 1987. Installation
of the test facility was completed in 1990.
Following several cold tests to establish a safe
melting procedure, hot melting tests were conducted
using metal wastes generated from the
decommissioning [11] of the Japan Power
Demonstration Reactor (JPDR) and simulated
wastes with radioisotopes. By the end of March
1994, a total of 33 hot tests had been completed.
(Qleadar Year)
"87
DE
OF
88
SIGNS
EQUlPd
••••
89
=ABU&a
«ENT
••li^l
"90
TION
'91
INSTALLATION OF EQU
8 FUNCTIONAL TEST
§•
COLD
••
TEST
'92
PMENT
HOT re
'93
3T
REMO
EQUIP
'84
VALOF
WENT
Fig.l Schedule of melting test program
2.2 Test Conditions
Melting tests were carried out under
conditions as shown in Tables 1 and 2.
the
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JAERl-Conf 95-015
casting equipment were installed in the steel
containment.
Photo.l snows the induction furnace. The furnace
has a capacity of 500 kg of steel per batch, and can
process about 500 kg of metal materials in 90
minutes.
Photo.l Induction furnace at JAERI
t
1,873 -
All auxiliary activities such as slag removal
(Photo.2), temperature measurement and sampling
of molten metal, were carried out through the hatch
in the ceiling of the containment. Off-gases
generated during melting were filtered through
cyclone, bag-house, and HEPA filters. About
8,000m3/h of ventilation air was processed during
melting.
Photo.2 Slag removal
2.4 Melting Process
The melting process for the simulated waste tests
is schematically shown in Fig,3. To investigate the
difference in the transfer behavior of radionuelides
by an melting process, melting of simulated wastes
was conducted by two process. The first is that
simulated waste was heated and melted at the
SAMPLING OF
MOLTEN METAL
AND SLAG
(REMOVE SLAG |
IADDFL
ESLAG 1
| ADD FLUX |
SAMPLING OF
MOLTEN METAL
AND SLAG
&
p
-B 1,723 -
m
D.
I
298
T: TEMPERATURE
MEASUREMENT
CHARGE OF
SIMULATED WASTE
90 100 110
Time (m!n.)
Fig.3 Melting process for simulated wastes
120
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JAERI-Conf 95-015
beginning of the tests together with large amounts
of commercial steel. The second is that simulated
waste was directly fed into the molten commeieial
steel, and this process was mainly applied for the
melting tests.
Molten metal was poured directly into the ingot
mold by tilting the furnace body. Photo.3 shows an
product ingot. Hie size of the ingot is about 900
mm long and about 300 mm wide. According to the
sectional inspection of ingots, no blow holes were
detected.
Photo.3 Product Ingot
2.5 Sampling and Measurements
In order to assess the transient radioactivities of
molten metal and slag, samples were taken by the
use of cylindrical metallic cups or slag removal
tongs: one sample before melting the simulated
wastes and three samples after melting of these
wastes. Six to ten samples of an ingot were taken
from the three sections, the upper, middle, and
lower portions of each ingot. Off-gas dust was
collected by paper and charcoal filters at several
points in the off-gas system.
The germaniom-semiconductor detector and the
liquid scintillation counter were used for measuring
radioactivity of samples. Samples of the ingot were
dissolved and samples of the slag were powdered.
Seventy-five to ninety radioactivity measurements
were made for each test. It required 5,000 to 10,000
seconds to measure each sample. Chemical
compositions of the product slag were determined
by inductively coupled plasma ernission-
spectroscopic device.
3, TEST RESULTS AND DISCUSSION
3.1 Partitioning of radionuclides
The melting of decommissioning wastes showed
that more than 99% of Co-60 in the wastes was
partitioned to the ingot, and the partitioning of Co-
60 was independent of the waste material.
The partitioning of radionuclides among the
ingot, slag, and off-gas is shown in Fig.4 regarding
five typical gamma-emitting radionuelides. These
data were obtained in tests in which simulated
wastes were fed direetiy into molten carbon steel.
As was expected from its thermodynamic and
120
100
£
c
o
ts
5
u.
o>
SO
60
g
¥ 40
20
Mn-54
Co-60
Zn-65
(CaO/BiO2)
BSSSf ingot (1.0) — Rl-2,3
stag (3.03 ~- RI-6,7
Dust (0.33 *• RI-10,11
Sr-85
Cs-137
3 6 71011 3 6 71011
236 71011
Test No.
3 6 71011 236 71011
Fig.4 Partitioning fraction of radionuclides
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JAERI-Conf 95-015
Table 1 Summary of melting tests using decsamnussioEing wastes
Table 2 Summary of melting tests using simulated wastes
TEST NO.
JP-1
JP-Z
JP-3
JP-4
ac-s
jf-e
W-7
JP-i
WASTE TVP£
Pipes
Pipes
Pipas
Pipes
Stuci votts
: Keainm-nK^a*d i
Stabilizers
{ Nwron-irndisud)
Pipes
Pi pas
MATERML
SUS 304ZI
SUS 304
ASTM-«3S
ASTM-ASJS
ASTM-A193
ASTM-A6SS
ASTM-A33S
SUS 304
RAOKMCTIVrtY"
2.0x!Q~3- 2.1x10"* Bt^om2
aextO"3- 2.4K10"* Bqfcm2
4.9xtCT3-4.18q*mJ
4.4X1 O^-S^Bq/on8
1.0-3JB*g
35 - I2i Bcj'g
7.7x10^- 4.&105 Bq/on!
3.7X10-3- 2.4x10' Bq/an*
1} Cobah~60 3 sbeoulfdaaa'aaai ndiooadide.
2) SUS : Stainless steel
Contaminated pipes and neutron-irradiated
components such as stud volts of pressure vessel
head were used in the decommissioning waste tests.
The wastes applied for the tests included Co-60
only as dominant radionuclide.
Simulated wastes were prepared by evaporating
radioisotope solutions onto three types of steel in a
vessel at 358 K using a heating furnace,
Radioisotopes applied to elucidate their transfer
behavior are Mn-54, Co-60, Zn-65, Sr-85 (a
substitute for Sr-90), Cs-137, and Ni-63. TTiese
radioisotopes were selected as typical radionuclides
found in contaminated components of nuclear power
plants. The total radioactivity applied to each test
was about 15 MBq for gamma-emitting
radionuclides. Depending on the materials, melting
occurred in the range of 1,593-1,923 K. The fluxes
used in the tests are CaO, Si02, Al (18-27wt%),
CaF2 (10-19wt%) and the basicity of (CaO/SiO2) of
Jhe flux compounds was varied from 0.3 lo 3.0 to
TESF
NO.
BI-1
HI-2
RI-3
RI-4
RI-5
ni-e
m-r
ni-a
RI-9
RI-10
BI-11
HI-12
HI-13
H1-14
RI-15
RI-16
HI-17
BI-18
RI-19
m-zo
RI-21
RI-22
RI-23
m-m
R1-2I
MATERIAL
ASTM-A33S
ASTM-A33S
MELTING
TEMP.
• (K)
1.913
1.873
1,803
1,823
1.843
1,893
1.S93
1,768
1.783
1.903
1.908
1.913
1,913
1J923.
1,913
1,613
1 593
1.763
T.793
1.829
J,873
1 esa
1.B43
1.858
1.B43
BASICITY
Of FLUX
(CaCVSiCy
1.0
1.0
1.0
3.0
3.0
3.0
3.0
0.3
0.3
0.3
0.3
1.0
1.0
1.0
1.0
3.0
3.0
0.3
0,3
_
1.0
1.0
03
03
RAKONUCUOES USED
Mn-S4,CO-60,Sr-8S
Zn-6S,Cs-137
SiadanuctidM1*
Ni-63
Mn-54,Zn-8i,Sf-85
Cs-137
5 radtonucSdes''
Hn-84Zn-6S,Sr-83.Cs-137
a-es
Mn-S4,Sr-8S,Cs-1 27
Mff-S4<2i>-8S,Sr-85
Cs-137
J) Mn-54, Co-60, Zn-65, Sr-85, Cs-137
2) SUS : Stainless steel
examine the influence of it on the partitioning of
radionuclides among the ingot, slag, and off-gas.
2.3 Test Facility
A small industrial-scale induction furnace
system located at JPDR was used for the melting
tests as shown in Fig.2. The system includes feed
equipment, a high-frequency induction furnace
(350kW, l.OOOHz), casting equipment, local
ventilation equipment, off-gas treatment equipment,
and so on. To prevent dust and fumes from
spreading during melting, the furnace and the
OFFGAS SYSTEM
CYCLONE
(Unitmm)
Fig.2 Schematic of Melting Test Facility
250 -
-------
JAERI-Conf 95-015
physical properties (change in free energy of
oxidation, boiling point of 2,877 °C), Co-60 was
partitioned nearly 100% to the ingot.
A laige amount of Mn-54 and Zn-65 were
found in the ingot. However, 7 to 20% of Mn-54
was oxidized into the slag and about 15% of Zn-65
volatilized into the off-gas. In spite of the high
volatility of zinc (boiling point; 1,179 K), Zn-65
remained in the ingot to a peat extent. Two
possible reasons are considered for these results.
The first is that the amount of Zn-65 applied for
the tests was less than ppb, and therefore, the
Zn-65 could be enough within the solubility limit
in molten metal bath. The second is that the
presence of the slag layer prevented Zn-65 from
vaporizing from molten metal bath.
Strontium-85 was found in the slag only, and
this tendency was not influenced by applied flux.
Since strontium is expected to be a less soluble
element in molten iron bath due to the size effect
(larger atom diameter than iron) and is also a highly
oxidizable element due to the lower free energy of
oxidation, it is considered that Sr-85 was oxidized
into the slag.
The measured radioactivities of Cs-137 were
under the detection limit for all samples taken from
ingots. Cesium-137 was oxidized into the slag
(2-40%) or volatilized into the off-gas (30-60%).
The partitioning of Cs-137 to the off-gas indicated
a tendency to be increased by the use of the flux
composition that its basicity is higher than one.
These results suggest that oxidation and
volatilization of Cs-137 is happening during
melting due to its lower boiling point (963 K) and
higher affinity to oxygen than iron.
As can be seen from Fig.4, there was imbalance
of radioactivity between detected and applied
radioactivities. Especially, the unaccounted
radioactivities were significant to the radionuclides
such as Sr-85 and Cs-137, which showed greater
partitioning to the slag or the off-gas. It is
considered that the unaccounted radioactivity was
mainly caused by the retention of some amount of
radionuclides in the furnace refractory and the off-
gas system. According to the radioactivity
measurements of samples taken from the furnace
refractory, Sr-85, Cs-137, Mn-54, and Zn-65 were
detected. As shown in Fig.5, the estimated
radioactivity of Sr-85 was nearly equivalent to the
unaccounted radioactivity during 7 melting tests.
However, other radionuclides were within 30% of
each unaccounted radioactivity. From the
10
(MBqj
0) oo
Radi
E23 Sr-85 applied per test OU Estimated residual activity
ra Activity found in slag rt Sr'as in *» «**»«¥
dl Unaccounted E
** i!
?
/
/
*** >iifcm '
* .f.ns... ;
1'
;
V
V
\Decay
\
**
T«$t No. RI-1 RI-2 Rl-3 W-2 RI-4 RI-5 W-3
Data 10/1510/2911/1211/1911/2613/121017
Refractory
As Of 4/21/92
Fig.S Residual Sr-85 in furnace refractory compared with
the unaccounted radioactivity during 7 tests
preliminary survey concerning contamination in the
off-gas system, there was only found 5 to 10% of
the unaccounted radioactivity to Cs-137 and Zn-65.
As described above, although smaller amounts of
some radionuclides applied for the tests are still
missing, the test results demonstrate that the
volatility and the affinity to oxygen of radionuclides
are important factors to predict the transfer behavior
of each radionuclide during melting.
3.2 Transfer behavior of radionuclides
(1) Variation in partitioning fraction by melting
process
The transfer behavior of Cs-137 and Zn-65 is
strongly influenced by the melting process of
simulated wastes. Figure 6 shows the influence of
a melting process on the partitioning of
radionuclides among the ingot, slag, and off-gas.
When the simulated wastes were heated and melted
at the beginning of the tests, the partitioning of
Cs-137 and Zn-65 to the off-gas increased by far
compared to those of when wastes were fed directly
into molten steel bath. This suggests that the
radionuclides with lower boiling points and higher
volatility such as Cs-137 and Zn-65 have a
120
Mn-MCo-W Zn-«S Sr
Instantaneously melted
Gradually melted
Fig.6 Effect of melting method on transfer behavior
of radionuclides
-251-
-------
JAERJ-Conf 95-015
tendency to transfer easily into the off-gas in the
process of melting. On the contrary, it is obvious
that the partitioning of Co-60, Mn-54, and Sr-85
is hardly influenced by the melting process,
(2) Effect of slag on partitioning fraction
As shown in Fig.7, test results applied no flux
showed a great increase of the partitioning of
Zn-65 to the off-gas. About a half of partitioned
Zn-65 was detected in the slag, leaving from
molten metal bath. At the same time, higher
amounts (20-25%) of Mn-54 was partitioned into
the slag compared to the results for which tests
were conducted in sufficient slag layer. These
results suggest that the presence of slag layer
prevents escaping and oxidizing of volatile elements
from molten metal bath to some extent.
Accordingly, it seems to be probable that there
happens no complete escaping from molten metal
bath even for the radionuclides that have higher
volatility.
120
I100
I 80
I
"- 60
0. 20
ingot
1
I
Oust
Carbon steel
Mn-54Co-6ClZn-6S Sr-SSCl-137
Mn-54Cc-«OZn-BB Sl-SSCl-137
Add flux
» t-Q
Fig.7 Effect of flux on partitioning fraction
(3) Temperature dependence of partitioning
fraction
There was found a little temperature dependence
on the transfer behavior of some radionuclides.
Figure 8 shows the temperature dependence of
partitioning fraction as a function of the melting
temperature. The melting temperature was evaluated
as the mean melting temperature before tapping
molten metal.
As shown in Fig.8(a), the partitioning of Co-60
to the ingot was almost constant within tested
melting temperatures. The partitioning of Mn-54
and Zn-65 to the ingot showed no clear tendency
due to the scatter of the test data.
On the other hand, as shown in Fig.8(b), slight
increasing of Cs-137 to the off-gas was observed
with melting temperatures. But, the partitioning of
s£.
c
"u
s
en
_c
"c
_o
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ro
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.riU
00
80
1?5
-------
JAERI-Omf 95-015
1.3
e
IB
0)
1.2
1.1
S 1
1
0,9
~ 0.8
0.7
*•*
*****
*****
*****
*****
*****
*****
•*
•
Mn
•**
*****
*****
*»*
-54
1,3
e
CO
S 1.2
1 1-1
0)
ts
1
g 0.9
o
U
o
•s 0.8
cc
0 5 10 15
Number of samples
0,7
**
••
****
*****
*****
*****
*****
*****
*
*
*
Co-
*
*****
*****
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-60
»*
*
1.3
•o
0)
11.1
0 5 10 IS
Number of samples
§ 0.9
o
u
o
~ 0.8
-------
JAERI-Conf 95-015
10) H. Nakamura, el al.,"Fundamental Research on
Melting of Radioactive Metal Materials," Proe.
Nucl, and Hazardous Waste Management Int.
Topical Meeting, Atlanta, USA, Vol.1, pp.206-210
(1994).
11) S.YANAGIHARA, et al.,"The Japan Power
Demonstration Reactor Decommissioning Program
," The 1994 Int. Symp. on Decontamination and
Decommissioning (1994).
- 254 -
-------
JAERI-Conf 95-015
3-12
Investigation on Recycling of Radioactive iaste
Mitubishi Materials Corp. Dr.D.Sakurai,K.Takahashi,
A,Uiemura,K.Kinura,
Rad-faste Management Center S.Abe.M.Yananoto
1. Introduction
Nuclear power generation is steadily increasing in Japan ,and the
quantity of the radioactive waste is also increasing according to the
aiount of nuclear power.general ion. Decoimisstoning of JPDR had been
conducted, and decomissioning of commercial nuclear plant will be held
in the near future. It will bring the explosive increase of radioactive
waste.
Japan has the very narrow land space,and it seeis to be very difficult
to store or to dispose all of the waste. Mill has been trying to
investgate the recycling system of such low level radioactive waste.
This is the intermediate report of the activities of the work.
2.Schedule of the investigation
The investigation started in 1988, at first the preparatory study of
the scenarios for recycling was carried out. And then the conditions of
the investigation were decided. In parallel with the study,the safety
studies and the conceptual investigation of recycling system were
perforied. After that, the conceptual design of the demonstration test
was carried out. From 1990, the demonstration test facilities were
designed and constructed. Since 1991 the demonstration test has been
continued. The investigation of the practical recycling systei was
done in 1993. The test is scheduled that it will finish in March,1997.
The detailed schedule is shown as Fig.l.
- 255 -
-------
JAERI-Conf 95-015
3. Investigatin of the recycling systei
3-1.Preparatory study of the recycling scenarios
25 scenarios were preparatorily studied for recycling radioactive
waste from nuclear power plants, and each scenario was divided in four
steps. 12 scenario were chosen in accordance with materials of the
wastes, the places to be used, the final usages of recycling, radio
activities, the degree of decontaiination, the exemption level of radio
activities and recycling facilities. A scenario was separated into four
steps according to the tine flow.
According to the scenarios, the quantities of the waste and the usage
of- the products were studied. The mass of the waste and the usages were
shown in Table-1 and Table-2.
As the result, concrete bars, casks for LLI, and pipes in nuclear
power plants were selected as the most promising recycling products.
3-2.Preparatory safety studies of the recycling
According to the scenarios some processes were settled. And some
foreign safety analysis and foreign regulations were followed.
The Measuring method of radio activities were also studies. And safety
evaluation was performed. Then, it could be possible to keep safe
during recycling,
3-3.Conceptual investigation of recycling system
Before the study on the system, the technologies for recycling were
followed, that is, the decontamination, melting, finishing etc..
After the studies of technologies, according to the scenarios various
system flows were investigated. The typical system flow is shown in
Fig.2. The conceptual design of the typical recycling plants was
brought out, and according to the design the econoatc evaluation was
carried out. The result was that recycling was lore econoiical than the
- 256-
-------
JAERJ-CoHf 95-015
disposal and it is shown in Fig.3.
4.Demonstration Test
4-1.Test Faci1 ities
The demonstration facilities for manufacturing 2001 drum inner
shield-ing material as a recycling product were constructed from April
1991 till September 1992.
The lay out is shown in Fig.4 and the main equipment are listed as
Table-3.
4-2.Test Products
The specifications of the test products are as follows.
*20Q-liter drum shielding laterial
*Size:full scale; 560mi^0uter Diameter, 8lOii Might
*fall ThSckness:20Miii-40am
^Product MaterialslGray Cast Iron, Ductile Cast Iron,
Cast Carbon Steel, Cast Stainlss Steel
4-3.Cold Deionstration Test No.1
4-3-1.Purpose of the test
The purpose-of the test was to confirm the technology to produce the
drum inner shielding laterial using various scrap.
^Control of chemical compositions by the induction furnace
^Processing technology of the material by centrifugal casting
*Reaote operation technojogy to produce the material
4-3-2.Test Procedure
The test procedure of the cold deionstration test No.1 is shown in
Fig.5.
- 257
-------
JAERI-Conf 95^015
4-3-3.Test Results
The typical shape of the products are shown in Fig,6, and it shows the
shape satisfies the purpose. Chemical compositions and mechanical
properties are shown in Table-4 and Table-5,
All of the results show that the technologies of this recycling system
run satisfactorily well.
4-4.Hot Demonstration Test No.1
4-4-1.Pur poses of the test
The lain purpose of the test was the confirmation of the safety during
pr-oeessing the radioactive waste. Additionally the confirmation of the
behaviour of radio-nucI ides and the uniformity of then in the products.
Moreover the function of off-gas treatment equipment was confirmed.
4-4-2.Test Procedure
The test procedure of the hot demonstration test was the sane as that
of the cold demonstration test,but for simulating the radio-active
waste from the nuclear power plants, 60Co, 5*Mn, 65Zn, 85Sr, and 137Cs
were added to the activity level about 25 Bq/g. The radio isotopes
were added into the molten metal with steel boxes in which dried metal
plates with the radio isotope hydro-oxide were put.
4-4-3.Test Results
Test conditions are shown in Table-6. The material balance is shown in
Table-7. According to this result, the recovery ratio 93.1% was very
good. And the ratio of secondary waste,that is slag and dust, 0.41 of
the charged laterial. The transpot ratio of radio nudities is shown in
Fig.7. 60Co was found only in metal. 54Mn was found mostly in metal and
merely in slag except H-l. The roost of 65Zn was found also in metal and
slightly in off-gas but for H-l. 85Sr was caught only in slag except
- 258 -
-------
JAERI-Conf 95-015
H-l. 137Cs was found in slag and off-gas except H-4. These results were
siiilar to the results of other investigations. The distribution of the
nuclides in the products is shown in Fig,8, and the location nuabers
are shown in Fig.9. From these results the distribution of the nuclides
was confirmed to be uniformed, and the analysis of a molten letal
sanple can represent the radio-activity level in the final product. The
results of the contaiination level of the test facility after the hot
test are shown in Table-8 and Table-9. As the result, the test facility
was not contaminated at all. And this means that the off-gas treatment
system was excellent in,the function.
5.The problem to be solved for recycling
As Mentioned before, soie technical investigations are proceeding, but
as- for the legal matters, there was no decision at all yet. The body
who can recycle the waste,the procedure to get the legal autholization ,
the regulation level of activities in the waste and so on should be
legally decided.
Concerning the technical latters, the following subjects are remained.
l)The effect of impurities like as paints, non-ferrous letals, and some
other gas containt should be investigated.
2)For improving safety, some automization of recycling process had
better to be developed.
3)After the legally decision, some development to get the autholization
nust be necesary.
6.Conclusions
In Japan, the studies on recycling have been started,and technically
they proceed quite smoothly by the assistance of the government. And
decomiisionlng of commercial power plant will be performed in the near
future. The legal maintenance to realize recycling are eagerly expected.
- 259 -
-------
JAERI-Conf 95-015
1. System investigation
1) Preparotory Study
2) Study Condition
3) Safety Study
4) System Concept
5) Econontc Study
^.Demonstration Test
1) Test Concept
2) Test Plant Design
3) Plant Construction
4) Cold Test SI
5) Hot Test $1
6) Interned late Suiaary
7) Cold Test S2
8) Hot Test 82
3. Summary
88
89
90
91
92
93
—
94
95
96
97
Fig.l Schedule of the investigation
- 260-
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JAERI-Conf 95-015
Hsiycli.-g products
I.ir.s.- shielding =ai.
Recycling c.-sdcsls
Inner shielding sat.
Vasts co.itair.ar
Stainl«ss slat! sip*
Ssai-prsd-sts
'a 3
Is st3rz;s
sxaasl* C
Fig.2 Typical Systei Flows of Recycling
5. 000
c
o
DISPOSAL COST(150Kg/Drun)
DISPOSAL COST
(300Kg/Drun!)--
^-—. OfSPOSAL COST-
(500Kg/Drun)
RECYCLIHC COST
20 40 60 80 100
DISPOSAL UNIT COST(10,OOOYen/Drui)
Fig.3 The result of econooic investigation
261-
-------
JAERI-Cotif 95-015
l*Wng!ji^| OO r=vlj! p '
- ?gg|jr ; E||r s f
o
'O
COO
o
-262 -
-------
JAERI-Conf 95-015
A
t~r~T-r-j
)Melting and Tapping
©Pouring
©Ingots
=> To Machining;
Secondarily
Melting
©Centrifugal
Casting
©Product
(as casted!
Fig.5 The test procedure of the cold test
Operation Ho.; Run 32
Material; Cast Stainless Steel
(SCS5)
Casting Tsmp.; 15SQ*C
Weight; 537*3
46-
t
Operation No.; Run 44'
Material; Cast Carbon Stes!
(SC41Q)
Casting Temp.; U?05C
W-ight; 363kg
U £54,5
40
c
-£473-
/X
SS4
27
24
Jfc
<—5i492
-p492'
7T
SS2
^.
27 34
Fig.6 A typical shape of the product
- 263
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JAERI-Conf 95-015
OJ
OS
c:
o
o
a
w
:H-2. A'H-3.
'-H-S, O'H-S
SUj otf-tts Htul
or(-«»a **i»l Sl«t Off-(»s
Off-Ill
Transportation Location
Fig. 7 The transortat ion ratio of the radio nuclides
"O
C3
.^
,
1
.
•
i -«- C3
: -- -I
;,"?~?— "•-«
i~~?~*~"*: \
....i .•' L J- t.
--r-rr"]™
: : ; ; :
: -: ; ; i
0! a: nj 0-= cj
3*
— •-
«s.
- Mn
K_
T--
Si s
H
*-*».
-i r
'*'
-«•
•*i
- 2-
«t
*
f
ij.fi.
*
*
••
*n!
£A«. )d,|tell j|ttt)
Locat ion
Test No. H-3
O
««e j
03
at ct »i (Si o>
Slettl
E 2 t 3 £4
01
D2
03
.,?4
05'
Locat ion
Test No. H-6
Fig. 8 The distribution of the nuclides Fig. 9 The location numbers of Fi:
in the products
- 264-
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JAERI-Conf 95-015
Table-l The mass of the radio active waste
(Tons/Year)
Radio Activity
Ci/t
1Q~1 ~10"2
10~2 ~-10~3
io~3-~io"*
-------
JAERI-Conf 95-015
Table-4 The chemical conpositions (%)
Material
JIS SC410
Cast Carbon
Steal
JIS FC100
Gray Cast
Iron
JIS SCS5
Cast
• Stainless
Steel
JIS FCD4QQ
Ductile
Cast Iron
Run No.
JIS std
Run. 38
Run, 44
JIS std
Run. 35
Run. 42
Run. 43
JIS std
Run. 28
Run. 30
Run. 32
JIS std
Run. 40
Chemical Composition (Ladle Analysis :wtS)
C
<0.3
0,17
0.26
-
3.47
3.53
3.53
<
o.os
0.03
0.05
0. 10
>2.5
3.48
Si
•-
0.31
0,58
-
1.86
1.90
1.95
<
1.00
0.23
0.39
0.72
-
2.24
•Mn
-
0.90
0.89
-
0.32
0.17
O.OS
<
1.00
0.61
0.80
0.93
-
0.25
P
<0.040
O.Oil
0-010
-
0.010
0.010
0.010
<
0.040
0.018
0-019
0.019
-
0.010
s
<0.040
0.008
0.005
-
0.001
0.016
0. 004
<
0.040
0.005
0.005
<0.02
0.001
Ni
-
0.03
0.03
-
0.03
0.03
0.03
3.50
4.50
3.7?
5-05
4.28
-
0.15
Cr
-
0.14
0.07
-
0.13
0.04
0.02
11.5
14.0
10.7
9.69
10.9
-
O.OS
iMg
-
-
-
m
-
-
-
0.034
Table-5 The mechanical properties
Material
JIS SC410
Cast Carbon
Steel
JIS FClOO
Gray Cast
Iron
JIS SCS 5
Cast Stainle
ss Steel
JIS FCD400
Ductile
Cast Iron
Test Nd
JIS Standard
Run. 38
Run. 44
JIS Standard
Run. 36
Run. 42
Run. 43
JIS Standard
Run. 30
JIS Standard
Run. 40
Yield
(Proof)
Strength
(N/mm2)
> 205
215
293
—
_
> 540
713
> 250
266
Tens i 1 e
Strength
(N/mmJ )
> 410
458
531
> 100
167
183
151
> 740
919
> 400
354
Elonga-
tion
(%}
> 21
27.4
11.9
—
— •
>13
15.9
> 12
10.1
Brinell
Hardness
(HBS)
—
138
150
<201
168'
158
157
217-277
303
<201
146
Charpy
Impact
Value
EJ/cm1]
—
20
18
—
3
5
3
—
14
—
12
NOTE SC410 ; Cast Carbon Steel SCS 5 : Cast Stainless Steel
FClOO : Gray Cast Iron FCD400 : Ductile Cast Iron
- 266 -
-------
JAERI-Conf 95-015
Table-S The test conditions of the hot test
Test No.
H-I
H-2
H-3
H-4
H-5
H-6
Objective
Ingot Making
Casting
Casting
Ingot Making
Casting
Casting
Material
FC
FC
FC
SCS
SCS
SCS
Weight
(Kg)
803
798
790
733
824
590
Casting
Teapsrture
CO
1,340
1,300
1.300
1,640
1,568
1,572
7ot=! of
Radioactivity .
OBq)
28.74
11. S3
29.26
25.4!
14.36
15.17
Notes: FC—Gray Cast iron, SCS—Cast Stainless Stsel
Table-? The material balance of the hot test
Test
No.
H-l
H-2
H-3
H-4
H-5
H-6
Ttl
Carged Weight(Kg)
Mate
rial
82S
798
831
779
824
590
4645
Alloy
20.5
2,6
4,5
82.0
6.0
85.8
201.4
Total
844
801
838
861
830
878
4846
Product tfeight(Kg)
Ingot
Produ
cts
827
790
794
822
•823 •
669
4725
Slag
0
0
4. 1
3.5
5,5
1.6
14.7
Dust
0.32
2. 15
1.00
0.45
0.26
0.60
4.78
Oth
ers
8.5
5.9
15.0
18.0
9.2
4.2
60.8
Total
838
798
814
844
838
675
4791
Ratio
of
Prod./
Charg
99. 1
99.7
97.4
98.0
101.0
100.0
98.9
- 267-
-------
JAERI-Conf 95-015
Tabte-8 The contaainat ion of the test factli'ty(l)
-
Iff™
1
2
3
4
5
6
7
0
9
10
(Hi
74
53
72
61
52
59
§f
3?
a. 5
36
20.5
5
29.S
6? | 33.5
72
£6
73
;s
33
36.5
Sf
4.3
-
3.3
-
—
-
o.a
3.3
0.3
3.0
(!&
ff.O
N.D
ft.O
N.D
H.O
N.O
N.O
W.O
N.O
fl.O
tt
11
12
13
11
IS
15
17
ia
19
20
(C/a;n.)
5?
59
^3
52
6t
|q£f
23.5
y..s
z^
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20. 5
©—-A: f-t- >'
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-
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-
-
-
•iriii]
N.O
N.O
W.O
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,'t-y.'l -, I-2SW:
Table-9 The contanination of the test facility(2)
jft. «s»,
i*r "-j
I
2
3
ii
5
6
7
8
9
10
35*1
ttttoa.
(c/ lain
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23
20
35
35
«S
3?
«3
29
35
iHW.
(cp)
IE',8
,!!•»?:
CCK)
9.3
5.3
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2.3
2.3
12.3
4.3
7.3
—
2.3
«£
(Sq/d]
N.D
K.O
N.O
H.O
N.D
M.D
H.O
fl.O
H.O
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I'f J
11
12
13
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18
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(c/ain. )
45
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20
35
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(q*>)
44 |
31 '
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12.3
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13.3
3.3
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11.3
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- 268
-------
JAERI-Conf 95-015
4. Compliance with Criteria
- 269
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Page Intentionally Blank
-------
JAERI-Conf 95-015
4-1 RADIOLOGICAL SURVEYS TO DEMONSTRATE COMPLIANCE
WITH DECOMMISSIONING LIMITS
David N. Fauver
U.S. Nuclear Regulatory Commission
ABSTRACT
In 1992, the U.S. Nuclear Regulatory Commission (USNRC)
provided guidance to USNRC licensees on acceptable methods
for surveying remediated facilities to demonstrate that
residual radiological contamination is below the
decommissioning limits. This guidance includes survey design
criteria, acceptable instrumentation, methods for averaging
small areas that exceed the limits (hot-spots), and the
statistical methods for demonstrating compliance at a given
confidence level. An overview of this guidance is presented.
The implementation of this method at a decommissioned
commercial nuclear power plant is discussed. NRC is
currently developing additional guidance on survey design and
statistical methods. The current research in this area is
reviewed.
INTRODUCTION
The goal of decommissioning a nuclear facility is to
decontaminate or remove contaminated equipment and reduce the
residual contamination on building surfaces and in soil to
acceptable levels. The acceptable levels of residual
contamination (decommissioning limits) are determined at the
beginning of a decommissioning project. At the end of the
project, a comprehensive radiological survey of the facility
must be performed to demonstrate compliance with the selected
decommissioning limits (final survey).
This paper discusses the method for performing final surveys
that is recommended by the United States Nuclear Regulatory
Commission (USNRC) and examines the implementation of this
final survey method at a commercial nuclear power plant.
Work in progress at the USNRC to upgrade the current final
survey method is also presented.
DISCUSSION
Final Survey Objectives
The objectives of the final radiological survey are 1) to
demonstrate that the average residual radioactivity is below
the decommissioning limits, and 2) to demonstrate that the
distribution of residual radioactivity, i.e., the number and
size of hot-spots, is acceptable.
The decommissioning limits for soil and building surfaces, in
units of Bq/g or Bq/cm2, are derived from dose or risk goals
- 271 -
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JAERI-Conf 95-015
set by regulatory agencies using computer based environmental
pathway/dose assessment models. The pathway analysis/dose
assessment models currently available assume that the
contamination in soil, and on building surfaces is random,
and uniformly distributed. If the assumption of uniformity
is valid, then the objective of the final survey is to
demonstrate that the average level of residual contamination
that remains in the facility is less then the decommissioning
limit.
However, in practice, a final survey will identify areas that
contain contamination above the decommissioning limit (hot-
spots) . In some cases, the distribution of hot-spots may
invalidate the assumption of uniformity and randomness. Past
experience has shown that the evaluation, remediation, and
resurvey of identified hot-spots tends to be a significant
effort and can be expensive. In addition, the presence of
hot-spots can be a primary concern to members of the public
who are interested in the decommissioning of a given
facility. When the average contamination is below the limit,
but hot-spots have been identified, the final survey must
also demonstrate that the distribution of the hot-spots is
acceptable.
Final Survey Method Recommended by USNRC
The final survey method recommended by the USNRC is described
in NUREG/CR-5849, "Manual for Conducting Radiological Surveys
in Support of Decommissioning."1 The major components of
this survey method are discussed below in the context of
designing a final survey plan.
The first step is to divide the facility into "survey units."
A survey unit is an area with similar operating history and
contamination potential. Dividing a facility into survey
units is necessary for proper statistical treatment of the
resulting data. But in practice, dividing the facility into
survey units also serves as the foundation for designing and
managing the actual survey, and reporting the survey results.
Concurrently with the effort to divide the facility into
survey units, each survey unit is categorized as either
having a high potential for contamination or a low potential
for contamination. Low potential survey units are designated
as unaffected; high potential survey units designated as
affected. The proper classification is important since the
recommended frequency of surveys, and survey cost, in
unaffected areas is significantly less than that recommended
for affected areas. The classifications are based on
characterization surveys, operational surveys, surveys
performed during decommissioning, and operational history.
Current USNRC decommissioning limits require buildings and
structures to be surveyed for direct (fixed plus removable)
and removable contamination, and .^or exposure rate at 1 m
from surfaces. For soil, the survey must measure the
272 -
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JAERI-Conf 95-4)15
concentration of residual contamination in the soil and the
exposure rate at 1 m from the ground surface. Scan surveys
of building and ground surfaces are also recommended.
The recommended frequency of soil and building surface
surveys depends on the classification of the area, i.e.,
affected or unaffected. However, for building surfaces, the
survey frequency in each affected area is further dependent
on whether the surface is a floor or wall (or other
structure) below two meters, or a wall (or other structure)
above two meters or a ceiling. Therefore, there are
essentially three different recommended survey frequencies
for building surfaces:
1. Affected area: floors and walls below 2 m
- scan survey of 100 percent of surface,
- one direct contamination measurement per 4 mz of
surface area, and
- one removable contamination measurement per 4 m2 of
surface
- one exposure rate measurement per 4 m2
2. Affected area: walls above 2 m and ceilings
minimum of either 30 direct measurements
or
one measurement per 20 mz
- removable contamination and exposure rate
measurement at each direct measurement location
- scan survey in the immediate area around each
direct measurement location
3. Unaffected area: floors, walls, and ceiling combined
- scan survey 10 percent of surface area
minimum of either 30 direct measurements
or
one direct measurement per 50 m2
- removable contamination and exposure rate
measurement at each direct measurement location
- investigation surveys for any measurement exceeding
25% of the decommissioning limit
There are two different survey frequencies recommended•for
open land:
1. Affected area
- scan survey 100% of area
- soil sample at corners of 5 m triangular grid
exposure rate measurement at each soil sample
location
273 -
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JAERI^Conf 95-015
2. Unaffected area
scan survey 10% of area
minimum of 30 soil samples
exposure rate measurement at each soil sample
location
investigation surveys for any measurement exceeding
25% of the limit
The actual net results (after background subtraction), both
positive and negative, should be reported. Any result
greater than the critical level, L02, should be identified in
the survey report as being greater than background. The data
from each survey unit are subjected to several tests to
determine if the residual contamination is acceptable.
First, the average residual contamination levels, in soil and
on building surfaces, are calculated and statistically
tested, to determine if the averages are below the
decommissioning limits at a 95% confidence level. A standard
one-tail students t-test is used to test the average.
Second, the hot-spots are tested.
There are three tests for hot-spots in soil. First the hot-
spot concentration (Bg/g) should not exceed the following:
Hot-Spot (Bq/g) < [ (100/A) °-5] [decommissioning limit]
where: A = area of hot-spot (m2)
For example, if the final survey identifies a 25 raz hot-spot,
and the limit is 1 Bq/g, the concentration in the hot-spot
could be up to 2 Bq/g.
The second soil hot-spot test, requires the average
contamination level over any 100 mz to be less than the
decommissioning limit. The third hot-spot test in the
current method requires the concentration to be limited to 3
times the decommissioning limit, regardless of size. If the
hot-spot concentration exceeds three times the limit the area
should be remediated.
There are two hot-spot tests for building surfaces, 1) the
average contamination over any 1 m2 of surface should not
exceed the decommissioning limit, and 2) the contamination
level should not exceed three times the limit, regardless of
size. Areas above three times the limit should be
remediated.
Implementation of Final Survey Method atDecommissioning
Commercial Nuclear Power Station
The final survey method described above was used to perform
the final survey at the Shoreham Nuclear Power Station (SNPS)
in Brookhaven, New York. SNPS was an 849 Megawatt Electrical
Boiling Water Reactor that operated between 1985 and 1987 for
274-
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JAERI-Conf 95-015
low-power testing. During this time, the facility operated
for less than two effective full power days. The station
permanently shut-down in June 1990, and completed the final
survey, the last step in the decommissioning process, in
October 1994.
During the final survey of SNPS, approximately 230,000
measurements were made on building surfaces and the exterior
and interior of remaining equipment and pipes. Very limited
soil sampling was required since there were no significant
spills or leaks during plant operation.
Fixed and removable contamination measurements were made on
building and equipment surfaces. The exposure rate was also
measured at 1 m from the surfaces. For soil, samples were
collected to determine radionuclide concentrations and
exposure rate measurements were made at 1 m above the ground.
The decommissioning limits for building and equipment surface
contamination, in 'units of disintegrations/minute/loo cm2
(dpm/100 cm2) , for the SNPS decommissioning were;
Fixed Plus Removable
Nuclide/ Average Maximum Removable
Radiation Type (dpm/100 cm2}
Beta(Co-60) 5,000 15,000 1,000
Alpha 5,000 15,000 1,000
Fe-55,H-3 200,000 600,000 1,000
The exposure rate limit for building and equipment surfaces
was 1.3E-09 C/kg/hr (5 uR/hr), above background, at 1 m from
the respective surfaces. The limit for Co-60 in soil was
0.3 Bq/g (8 pCi/g). The exposure rate limit for open land
areas was 1.6E-09 (10 uR/hr), above background, at 1 m from
the ground surface.
The SNPS facility was divided into 385 survey units. Survey
units were designated as structures, systems, or outside
areas. The contamination potential of each survey unit was
classified as being either affected or unaffected to
determine the required survey frequency. Table 1 was
extracted from the SNPS final survey report3. .Table 1 shows
the breakdown of the SNPS facility into affected and
unaffected survey units. Note that the majority of the
survey units in the Reactor Building and the Radwaste
Building were classified as affected, i.e., potentially
contaminated. The survey units in the Turbine Building,
other support buildings, and the outside areas were
classified as unaffected. Dividing the facility into survey
units was an effective tool for managing the final survey
project. Survey units were used to organize the survey, from
the planning stage to the reporting of final survey results,
to develop work plans, and to track progress.
Examples of the summary data that was provided in the SNPS
275 -
-------
JAERI-Conf 95-0 IS
Table 1 - Classification of survey units for the
final survey of SNPS. Excerpted from the SNPS
final survey report.
DESCRIPTION
STRUCTURES
Reactor Bldg
Dryweil
Suppression Pool
Turbine Bldg
Radwaste Bldg
Control Bldg
0 & S Bldg
0 & S Bldg Annex
Other Site Bldgs
Structure Totals
OUTSIDE AREAS
Site Grounds
Soil Samples
Structure Exteriors
COL-i
RB
PC
SP
TB
RW
CB
OB
AB
OS
SG
SS
SE
Outside Area Totals
PLANT SYSTEMS
TOTALS
SU
Total No.
of Survey Units
85
14
5
106
50
4
4
4
_S
281
7
1
14
22
82
385
No. of Affected
Survey Units
84
14
5
8
48
_L
160
0
39
199
No, of Unaffected
Survey Units
1
98
2
4
4
4
_1
121
7
1
14
22
43
186
-276-
-------
JAERI-Conf 95-015
final survey report are shown in Tables 2 and 3. Table 2
shows the results for total surface activity (fixed plus
removable) in a given survey unit. The results of the
individual measurements are reported. Note that negative
values and the results that were above background, i.e,
exceeded Lc, are included in the Table 2. Table 3 provides a
partial summary of the results for each survey unit. The
number of measurements, maximum result, and the upper 95%
confidence value is reported for each survey unit, and for
each survey type. Each reported value is compared to its
respective decommissioning limit.
Overall, the final survey method employed at SNPS was
effective in providing final survey data that demonstrated
that the average contamination level was less then the
decommissioning limit and that the distribution of the
contamination was acceptable. However, after two years of
experience implementing the current method, at SNPS as well
as a number of other decommissioning facilities, the USNRC
has identified several components of the method that can be
improved.
Work in Progress at USNRC to Improve the Final Survey Method
There are several components of the current method that are
under revision by the USNRC. The first is the statistical
test used to evaluate the average contamination level. The
current method uses the students t-test, which requires the
data to be normally distributed. In practice, the
distribution of final survey results are often not normally
distributed. To reduce the dependence on the data
distribution, a statistical method is being developed that
uses non-parametric statistics. Non-parametric statistics do
not require any assumptions regarding the distribution of the
data and are therefore more generally applicable. In
addition, the statistical tests being developed provide
estimates of the Type II error, as well as the Type I error.
The Type II error provides the probability of deciding that
an area meets the decommissioning limit when in fact it has
not. The Type I error provides the probability of deciding
that the area does not meet the decommissioning limit when in
fact it has. The current method estimate the Type I error
only.
The method for evaluating hot-spots is also being revised,
The actual dose consequences from hot-spots, as well as the
conditions where the assumption of uniformity can be assumed,
are being more quantitatively evaluated. The revised method
should provide a technique for demonstrating the
acceptability of hot-spots that is more clearly based on
potential dose. Other research areas include the use of in-
situ gamma spectrometry for final surveys, a method for
evaluating subsurface contamination, and a method for guiding
remediation, and performing followup surveys, when an area
fails one of the compliance tests.
- 277 -
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JAERI-Conf 95-015
Table 2 - Example of final survey report for total
surface activity in a given survey unit at SNPS,
page No. 1
Date: 06/25/93
DECOMMISSIONING PROJECT
Termination Survey Data Heport
Detail Data Reccrz
survey Unit ID. :SE001 System Code:
Name: SECURED AREA. - SOUTHEAST
Survey data 04/05/93
Location
V-2
9-3
W-5
W-6
W-7
w-is
W-16
W-17
W-21
K-23
W-24
W-26
W-27
W-28
H-31
W-33
W-34
W-37
W-39
W-40
W-41
W-49
K-50
W-5 2
H-S3
W-5 6
H-57
W-5 8
W-61
W-63
W-65
H-66
W-68
W-70
W-7 3
W-7 4
H-S2
W-83
tot-1 sxirface activity (beta-gaiama]
point
1
2
3
4
5
6
7
a
9
10
11
12
13
14
15
16
17
18
19
20
21
22
23
24
2S
26
27
28
29
30
31
32
33
34
35
36
37
38
gcpo
183
201
198
141
189
151
iOi
133
179
196
169
132
139
159
151
168
145
155
170
151
120
191
198
181
14 S
151
147
14S
133
140
199
136
1S5
176
1S4
168
179
171
, —
bk_cpm
159.0
159,0
159. 0
159.0
159.0
159,0
159.0
159.0
159.0
159.0
159.0
159.0
159.0
159. 0
132.0
132.0
132.0
132.0
132.0
132.0
132.0
159.0
159.0
159.0
159.0
159.0
159.0
159.0
159.0
159.0
159.0
159.0
159.0
159.0
159.0
159.0
1S9.0
159.0
eff
0.049
0.049
0.049
0.049
0.049
0.049
0.04?
0.049
0.049
0.049
0.049
Q.04S
0.049
0.043
0.049
0.049
0.049
0.049
0.049
0.049
0.049
0.049
0.043
0.049
0.049
0.049
0.049
0.049
0.049
0.049
0.049
0.049
0.049
0,049
0.049
0.049
0.049
0.049
dpm/100cm2
490
857
796
-367
612
-163
857
592
40S
755
204
-SSI
-408
0
388
735
265
469
776
388
-245
653
796
449
-224
-163
-245
-265
-429
-388
816
-469
122
347
S10
184
403
245
Lc
740
7S9
7=5
693
74S
704
7S9
745
735
754
724
632
631
713
673
653
666
678
695
S73
635
743
755
73S
701
704
700
699
639
692
757
637
720
732
741
723
735
727
>Lc
X
X
X
X
X
X
X
X
+Fe55
1029
955
1029
906
882
931
955
980
Lr
1162
1157
1162
1154
1060
1064
1157
1153
>LS
- 278
-------
JAERI-Conf 95-015
Table 3 - Excerpt from summary table in final
survey report for SNPS.
Compilation of Individ *l Survey unit Results
SETA-GAMMA TOTAL SETA-GAMMA REMOVABLE CAMKA EXPOSURE RATE ALPHA REMOVABLE
dpi!/100 atf dpm/iOO cm2 nicroft/hrilit. dp™/100 onZ
SWVEt UNIT 10 MEASURE MAXIMUM UCL NEASUSE MAXIMUM UCl MEASURE NAXtMUN
PHASE -MENTS (15,0003(5,000) -MENTS (1,000) (1,000) -MENTS (10>
REACTOR BUILBlNG-AFFiCTED
RB001 REACT 8L0G-8'N C/A
RS002 REACT BLDG-S'NE G/A
R800S REACT BLDG 8'E G/A
BB004 REACT BLDG-S'SE C/A
RBOQ5 REACT ilPG-8'S C/A
RB004 REACT aiDG-8'SV G/A
RS007 REACT BLDG-S'W G/A
118008 REACT BIDG-8'NU C/A
88009 REACT BLDG-40'K C/A
RB010 REACT ILOG-40'NE G/A
RSOIt REACT 81.06 40 '€ G/A
RB012 REACT BIDG-40'SE G/A
R8013 REACT BLOG-40'S C/A
RBOH REACT BLCO-40'SV G/A
RBOtS REACT BLOG-40' V G/A
RB016 REACT BLOG-40'NII G/A
RB017 PAS BLOC 1FL S C/A
R8018 RB ACC LOCK ENCt
«B020 >W BUJ5 ZfL S C/A
RB021 PAS 8LOG UL N G/A
RB02J REACT 8LDG-63'« G/A
R8024 REACT 8LDC-63'II£ G/A
R802S EAST U1KC RM
•8026 PAS BLOC 2FL N G/A
RB028 WEST WING UN
RSQ29 REACT flt.DG-63'lfll C/A
HO JO MM STEAM ISOL VLV RM
RB031 REACT 81. DC- TO 'NE G/A
R8032 TIP KuON
RB03S E ACCUMULATOR A1SIE
MOV, REACT BL06-7B'S£ G/A
RB036 CM REBUILD SCON
R«037 REACT ILOG-TS'SU G/A
RS03S U ACCUMULATOR AtSLE
H8039 REACT BLOG-TfS'MU G/A
RS042 REACT BLDG-9S1 G/A
(RMS REACT BlDG-101* G/A
18044 REACT ILDG-112'K G/A
RB045 REACT BtDG-112'NEG/A
RB046 XUCU PIPE SIH.O EKCl
RB04? A RUCU RECIRC PHP RM
RB04& n RUCU RECIRC FM> RH
RB049 REACT 81DG-EASI CORR
RB050 EAST MCC ROOM
R80S1 REACT BLDG-112'SEG/A
RB052 RB VENTIl INTAKE RM
RBOS3 REACT BLDG-112 -SUG/A
RBOJ4 WEST HCC ROCM
RB055 S K2 RECOHB1HER AREA
4
4
4
4
4
4
4
4
4
4
4
4
4
4
4
4
«
4
t
4
4
4
4
4
4
4
4
4
4
4
4
4
4
4
4
«
4
4
4
4
4
4
4
4
4
4
4
4
4
233
204
190
213
225
202
215
268
188
206
20Z
261
207
«5
209
263
244
274
198
isr
192
226
2 TO
173
238
205
207
192
161
250
-,2
152
197
264
269
237
229
176
294
146
140
146
1S3
140
283
171
198
136
126
988
988
1224
93V
1059
1298
1371
3551
1059
1249
1592
2449
1694
1714
955
1592
1102
1151
2541
4871
857
4306
2188
2612
1445
1976
98C
1592
10S3
2057
i.94
1347
1200
205?
1371
isn
1482
833
1249
1396
95S
163
988
673
31 76
1322
H12
833
1129
-183
-207
10
-17
-142
-174
17
-99
-112
-163
22
89
83
101
88
94
105
91
90
77
-1
182
194
512
63
119
-179
46
122
103
64
31
93
125
95
-39
79
85
56
138
33
-447
152
-399
58
94
58
16
237
223
204
WO
213
225
202
21S
268
183
2%
202
261
207
155
209
263
244
274
198
W
192
226
270
in
238
205
207
192
161
250
112
152
197
264
269
237
229
176
294
146
140
140
153
140
283
171
198
136
126
226
16S
Rt
13J
197
225
nr
197
85
21
84
48
48
46
47
70
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JAERI-Conf 95-015
CONCLUSION
Final radiological surveys should be. performed at the end of
a dcommissioned project to determine compliance with the
decommissioning limits. The final survey should be capable
of demonstrating that the average contamination level is
below the limit and that the distribution of hot-spots is
acceptable. The USNRC has developed a final survey method
that has been successfully applied by licensees at a number
of decommissioned facilities. Work is in progress at USNRC
to improve several aspects of the current final survey
method.
REFERENCES
1. Berger, J.D., Oak Ridge Associated Universities, Draft
"Manual for Conducting Radiological Surveys in Support
of License Termination," MUREG/CR-5849, June 1992.
2. Currie, L.A., "Limits for Qualitative Detection and
Quantitative Determination - Application to
Radiochemistry," Anal. Chem. 40, 586, 1968.
3. Long Island Power Authority, "Shoreham Decommissioning
Project, Termination Survey Final Report Phases 1,2,3,
and 4," October 1994.
280
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JAERI-Conf 95-015
4-2
The Japan Power Demonstration Reactor Decommissioning Program
- decontamination and radioactivity measurement on building surfaces -
Mitsuo Tachibana, Mutsuo Hatakeyama.Yoshihiro Seiki and Satoshi Yanagihara
Japan Atomic Energy Research Institute
ABSTRACT
After dismantling the components in the facilities of Japan Power Demonstration
Reactor (JPDR), decontamination on concrete surfaces and final survey of radioactivity
have been started as the last step in the JPDR dismantling activities.
At the first step'for the decontamination on concrete surfaces and the final survey
of radioactivity is as follows; The contamination on the concrete surfaces in the JPDR
facility was characterized on the basis of radioactivity measurements of samples taken
from the buildings. The contamination in the JPDR facility was categorized into two
groups: fixed or removable; deep penetrative contamination was not found in the JPDR
facility. The distribution map of the contamination was made based on the
characterization. Decontamination activities were planed according to the distribution map
of the contamination.
The all buildings will be demolished and the site will be landscaped after finishing
the final survey of radioactivity by March 1996.
INTRODUCTION
The Japan Power Demonstration Reactor (JPDR) decommissioning program is
underway. This program has two phases; Phase-l is to develop reactor decommissioning
techniques, Phase-l I is to demonstrate the techniques developed and to obtain
experience and data on actual dismantling of the JPDR. The scope of the JPDR
dismantling project (Phase-l I) activity is to remove all radioactive materials from the
facility.
The decontamination of the concrete surfaces in the JPDR facilities was
successfully performed by a scabbier, a shot-blaster, a sand-blaster and a needle gun.
Data on actual use of these tools were also collected through the decontamination
activities.
This report describes the JPDR dismantling activities focusing on the procedures
of decontamination and measurement of radioactivity on the concrete surfaces.
JPDR DECOMMISSIONING PROGRAM
JPDR
The JPDR, which was the first reactor to generate electricity in Japan, is a boiling
water reactor that began generating electricity in 1963 and ceased the operation in 1976.
The total operation time and the output of electricity during this period were
approximately 17,000 hours and 1.4 x 10s kWh, respectively. The thermal power was
initially 45 MW (JPDR-I). which was later increased to 90 MW (JPDR-II) for the
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JAERI-Conf 95-015
enhancement of its neutron irradiation capability.
Overview of JPDR decommissioning
The JPDR decommissioning program was initiated in 1981 under contract with the
Science and Technology Agency (STA) in Japanci). It consists of two major phases:
Phase-! began in 1981, aiming at developing technologies necessary for reactor
decommissioning®-'31. Phase-ll began in 1986, actual dismantling of the JPDR to reach
green field condition. The objectives of the program are (1) to gain actual experience
of nuclear power reactor dismantling, {2} to verify and confirm the developed techniques
and (3) to collect the data on the dismantling activities.
Actual dismantling of the JPDR has been progressing since 1986, aiming at
making green field condition as the'"final goal. The main parts of JPDR facilities including
the reactor pressure vessei(4) and the biological shield® were dismantled by January
1994.
The JPDR decommissioning program is in final stage; the major activities in the
stage are decontamination and survey of radioactivity for cancellation of radiation
controlled areas.
PROCEDURE OF JPDR SITE RELEASE
Figure 1 shows flow of site release procedure on the JPDR decommissioning
activities. The procedure of the site release was divided into 4 steps. At the first step,
contamination on concrete surfaces of the buildings was characterized by sampling and
measurements. On the basis of the measurements, contamination map was created,
then the contaminated concrete surfaces were removed according to the plan which was
made by reflecting the contamination map. The building surface has been
decontaminated in each area, then the radioactivity on the decontaminated areas has
been surveyed to confirm that the radioactivity is less than the planed value. After all
the activities will be finished, the site on the JPDR will be released.
Characterization of contaminated areas
At the first step toward releasing the facilities for unrestricted use, the
characteristics of the contaminated areas on the facilities were evaluated by the
procedures shown in Fig.2. About 1,800 samples were taken from every 2 m by 2 m
area of the buildings; the size of the samples were 1 cm in depth and 4 cm in
diameter®. Gross gamma-ray of each sample was measured by a Nal (TI) detector with
a single channel analyzer. Figure 3 shows sampling points and results of measurement.
In addition to the measurement, previous records of contamination during the JPDR
operation were surveyed using the log book. The contamination on the concrete surfaces
in the JPDR facilities was roughly characterized on the basis of the radioactivity
measurements and the previous records. The map of the contaminated areas was made
from the data. Figure 4 shows distribution of radiDactivity contamination in the JPDR
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JAERI-Conf 95-015
facilities. It was found that about 47 % of total radioactivities of contamination (1.3 x 108
Bq) in the JPDR facilities was existed in the rad-waste building, and that about 37 %
of total areas ( 20,000 m2 ) was contaminated.
The following two sampling methods were applied to getting more exact data,
especially on contaminated depth at the points where contamination was detected. In
the areas where only surface contamination was found, surface layer (2 mm in depth)
was removed until contamination was not detected. In the areas where penetrated
contamination was found, cores (10 cm in depth) were taken from concrete surface.
Samples were made in such a way that every 1 mm thickness layer was taken from the
surface to 10 mm in depth, and 5 mm thickness layer below 10 mm in depth. Gamma-
ray spectrum of each sample was measured by a Ge-detector with a multichannel pulse
height analyzer.
Figure 5 shows characteristics of the contamination in the JPDR facilities. About
86 % of contaminated area of the JPDR facilities was the surface contamination in
which the radioactivity existed within a very thin surface layer of only 2 mm in depth.
It is therefore enough to remove thin surface layer slightly for decontamination of the
building surfaces. Plans for decontamination on the contaminated areas were made
based on the above data. Figure 6 shows a sample of the plan for decontamination in
the precoat charge tank 1A and the control room of the rad-waste building.
Decontamination on buildings
As described above most contamination in the JPDR facilities was of two types,
surface and penetrated contamination. For this reason, the decontamination procedures
were roughly divided into two groups. Figure 7 shows the decontamination procedures
on the JPDR.
The penetrated contamination was removed in the following procedures. First, holes
were drilfed in every 1 m by 1 m area to a specified depth. Then the holes were
colored by paint. The concrete surface in each 1 m by 1 m area was decontaminated until
the color disappeared.
Decontamination of surface contamination was performed as follows. Contaminated
concrete surface areas were colored by paint. The concrete surfaces were removed until
no color remained on the concrete surfaces. Decontamination was performed by a sand-
blaster, a shot-blaster, a scabbier and so on. Photo 1 shows decontamination tools.
Final survey of radioactivity
After finishing the decontamination, the final survey of radioactivity was conducted
with the procedures shown in Fig.8. At first, the radioactivity was directly measurement
at each room in the JPDR facilities using gas flow counter type and scintillation counter
type survey meters, which has 160 cm2 or 1,800 cm2 and 900 cm2 of the sensitive
window area, respectively. These survey meters are capable of 0.4 Bq/cm2 for beta-
emitters under normal background conditions. The final survey is performed on every 0.8
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JAERI-Conf 95-015
m x 0.8 m square blocks moving the survey meters, and the highest counting rate in
each block was recorded.
Furthermore, after the decontamination activities some concrete samples were
taken from the decontaminated areas. The gamma-ray spectra of the samples were
measured by a Ge-detector with a multichannel pulse height analyzer to confirm that
there was no contamination in the concrete samples.
After finishing the final survey by workers, the survey was conducted by the
personal in the department of health physics of JAERI.
Site release
Based on the report of site release procedure, which was submitted to the STA,
the confirmatory survey will be conducted by the STA. Then the site on the JPDR will
be released. The turbine building, the control building, the fuel building, the reactor
building, the rad-waste building and the fan building will be demolished. The site on the
JPDR will be landscaped.
CONCLUDING REMARKS
On the basis of the procedures of site release on the JPDR, the decontamination
of the contaminated concrete surfaces was performed efficiently and safely. It was very
useful for making the decontamination plans to evaluate the characteristics of the
contaminated areas.
REFERENCES
(1)M. Ishikawa, M. Kawasaki, and M, Yokota; JPDR decommissioning program - plan
and experience, Nucl. Eng. Vol.122, pp.358-364, 1990
(2)S. Yanagihara, Y. Seiki, and H. Nakamura; Dismantling techniques for reactor steel
structures, Nuclear Technology, Vol.86, pp.148-158, August 1989
(3)S. Yanagihara, F. Hiraga, and H. Nakamura; Dismantling techniques for reactor steel
piping, Nuclear Technology, Vol.86, pp.159-167, August 1989
(4}M. Tachibana, T. Hoshi, and I. Miki; Underwater cutting of JPDR reactor pressure
vessel and core internals, Proceedings of the 1st ASME/JSME Nuclear Engineering
Conference, pp.81-84, November 1991
(5}K. Kozawa, M. Kan, S. Yanagihara, and K. Fujiki; The progress of the Japan Power
Demonstration Reactor (JPDR) Dismantling Activities: Dismantling the Biological Shield
Concrete by controlled Blasting, Proceedings of the 2nd ASME/JSME Nuclear
Engineering Conference, pp.821-826
(6)H. Yasunaka, M. Hatakeyama, T. Sukegawa, T. Kozaki, S. Yamashita, and T. Hoshi;
Evaluation of contamination on concrete of JPDR building, Proceedings of the 1989 joint
international waste management conference, pp.183-187, October, 1989
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JAERI-Conf 95-015
Characterization of contamination
Decontamination
Final survey of radioactivity
Site release
Fig. 1 Flow of site release on JPDR
Measurement of
contamination
Survey of previous j
Records of contaminationj
Mapping of contaminated area
Characterization of contamination by
measurement (depth of penetration)
Planning for
decontamiation activity
Fig,2 Procedure for characterization of contamination
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JAERl-Conf 95-015
Control and Turbine Building
Bad Waste Building
Filter Sludge
Storage
Tank Room
Purap Room
Waste Tank Roan
Sampling point
wall floor
p O non-contamination
I • contamination
Fig.3 Sampling points and results of measurement
6.2%
8.5%
Fuel Storage Building
3.7%
Fan Building
Turbine Building
21.4%
46.7%
Rad-waste Building
Reactor Building
Fig,4 Distribution of contamination
Q
[Total activities of contamination 1.3 x 10 Bq)
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JAERI-Conf 95-015
5.9%
l Penetrated contamination
62.6%
Non-contamination
31.5%
Surface contamination
Fig.5 Characteristic of contamination
(Gross areas of buildings: 20,000m2)
Room
"\^
Floor- 1
Floor-2
Precoat charge tankIA and control room
Decontaminated area
7m2
63m2
Decontaminated depth
1.5cm
Surface
Fig.6 Sample of plan for decontamination
(Rad—waste building)
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JAERI-Conf 95-015
Penetrated contamination
Holes were drilled in every 1 m
x 1 m area to a specified depth.
Surface contamination
Contaminated surfaces
were colored by paint.
Hole surfaces were colored by paint.
Decontamination until
color is disappeared
Final survey
Fig.7 Decontamination procedure
(Direct measurement (all areas)]
Sampling and measurement
(gamma—ray spectrum)
by workers
Survey by Department of
health physics of JAERI
Confirmatory survey by S.T.A.
Site release
Fig.8 Final survey of radioactivity
288
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JAERI-Conf 95-015
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JAERI-Conf 95-015
4-3
Measurement of Residual Radioactivity in the Facility Being Decommissioned
H. Ezure, S. Miyasaka, H, Kuroda, J, Komatsu
Research Association For Nuclear Facility Decornmissioning(RANDEC)
Abstract
After decommissioned, the site of a nuclear facility is necessary not to
be contaminated with any radioactivity for site radioactive-free release.
Major nuclides deposited on components, building floors, sites etc. are
found to be 60Co, 137Cs and etc. Because the latter nuclides usually accompany
BOCo, the detection nuclide characterizing contamination can be deteriined to be
s°Co. The termination survey for the site release must be carried out on the
condition that the residual radioactivity is a very low level, and the site is a
very large area. In addition there is possibility that the detection of e°Co is
disturbed by the background level due to 40K and other natural radioactivities.
Therefore, the basis of the measuring system consists of several Nal detectors
and electronic circuits mounted on a vehicle in order to enhance the
efficiencies of detection and of measuring operation. Further, the position of
measuring points will be deteriined by an auto-positioning apparatus.
Our development program of the measuring system is going on and will be
outlined in the presentation.
1. Introduction
Generally, after decommissioned the release of a nuclear site from
radioactivity is necessary not to be contaminated with any radioactivity. In
order to confiri the existence of radioactivity in the site a termination
survey should be conducted.
The nuclear site is characterized by a very wide area. After final
decommission operations, the residual radioactivity is very low, even if it
would exist. These characteristics originate two problems for measuring residual
radioactivity in the site.
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JAERI-Conf 95-015
The first problem is a large amount of measuring time and work though the
existing methods, because the site is wide. In addition, the residual
radioactivity is very low. Unfortunately this fact increases measuring time and
measuring works. The second problem is the great difficulty in distinguishing
whether radioactivity is residual or not in the site, because the residual
radioactivity is very low.
These problems have been solved usually through the expense of a large
amount of time and work on measuring. Furthermore it increases with the
requirement on the measurement of release level.
Therefore, the reduction in measuring time is the first step for
rationalization in measurement of residual radioactivity. The subjects for this
solution includes two. The first subject is the improvement of the efficiency of
the measurement. The second subject is the establishment of a proper method for
determining whether radioactivity is residual or not in the site.
Up to the present, we have studied three items for developing the measuring
technology. But today, we have not yet manufactured measurement apparatus or
system. Our development program is going on. Accordingly, our study will be
outlined.
2. Outline of our study
Our study includes as follows.
(1) Investigation on natural radioactivities
(2) Concept Design of measuring system
(3) Fundamental experiment on response function method.
Next these items will be explained in order.
2. 1 Investigation on Natural radioactivities
First, we made a small survey on natural radioactivity and nuclides of the
soil in Ibaraki prefecture. From this survey we can see that
(D "°K contain is 160~460 Bq/kg.
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JAERI^Conf 95-015
© 214Bi is 6,3-13 Bq/kg.
(D 228Ac is 11-23 Bq/kg,
® 13TCs is 12—88 Bq/kg(not natural)
® The other contain is below 10 Bq/kg.
It is remarkable that 60Co does not exist in the soil. Furthermore, these
nuclides indicate the amount of scatter in the radioactivity, but there is a
mutual relation of radioactivity between 2I4Bi and 228Ac. However, there is no
mutual relation between 40K and the other nuclides. This fact is very important.
If 40K has a relation between with the other nuclides, we can adopt the
relation for reduction of background level or disturbance by 40K in measuring
the radioactivity of an interesting nuclide, 60Co.
2. 2 Concept design of measuring system
Our final concept design of measuring system is shown in Fig. 1. This system
is composed of two systems, namely, "Residual radioactivity measurement system
and Residual radioactivity evaluation system". The measurement system includes
four Nal detectors, multi-channel analyzer, location detector, temperature
compensator and so on. This system is loaded on a vehicle. The another system,
namely evaluation system mathematically processes and arranges the data
measured by the measurement system. This system determines the existence of
residual radioactivity and its amount. We are planing to apply statistical
method.
In order to make a good concept design, we investigated:
(D Detector selection
(D Background reduction
(D Detection efficiency increase
@ Function of measuring system
2.2. 1 Detector selection
For concept design of measuring system, we investigated what kind of
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JAERI-Conf 95-015
detector is most suitable and what nuclide is interesting for measuring
residual radioactivity. Then, we evaluated the characteristics of Nal detector,
Ge detector, plastic scintillator and ion chamber.
As for detector efficiency, plastic detector and ion chaiber are very large,
but Ge detector is small. Nal detector is comparatively large, but it is not
good for resolution. The weigh and size of each detector are almost same. As
understood, these detectors have characteristics each other. Therefore, it is
difficult to determine what kind of detector is most suitable for measuring
residual radoactivity. Accordingly, we selected Nal detector first, because this
type is comparatively cheaper and easily buyable in comparison with the others,
and detection efficiency is large.
2.2.2 Background reduction
Generally, the reduction of the background of the detector increases
detectablity of radioactivity, that is, signal-to-noise ratio increases with
background reduction. As for Nal detector, there are many method for this
reduction, such as method for shielding the detector with lead block,
application of electronic circuit for the reduction, and others. We adopted the
method for shielding the detector with lead block which is widely used in
radioactivity measurement
Figure 2 shows the shielding characteristics of Nal detector in measuring
natural radioactivity. Detector diameter is 2 inches. Shielding material is lead,
and thickness is 3 cm. A kind of shielding is five, as shown in Pig. 2, that is,
lower side shield, middle side shield, upper side shield and window shield, in
addition to none shield, bare type. The count rates shown in Fig. 2 are the
values measured in the case that each shield is mounted on the Nal detector. In
these shielding experiments, the lower side shield and middle side shield is
most effective. Furthermore, it is possible to neglect the upper side shield by
priority of reduction of the weight of the measurement system.
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JAERJ^Conf 95-015
2.2.3 Detection efficiency
Above-mentioned, we assumed to adopt Nal detectors. Upon this adoption, we
learned how much detection increases with a number of detectors and how we
arrange detectors for improvement of detection efficiency.
Figure 3 shows detection efficiencies of four detectors which are arrayed
in 2 by 2. These efficiencies are the results calculated on the assumption that
one array distance is 50 cm and another array distance is a variable. In
addition, height difference between detectors and source is 60 cm and the points
calculated are on the center of array. Similarly, Fig. 4 shows the detection
efficiencies calculated in the case of 20 cm height difference. As shown, the
latter detection efficiency is larger than before, because the height difference
is small.
2.2.3 Function of measuring system
We learned the following items in order to determine the specification of
measuring system.
(1) Measurement speed
(2) Movability without a dead angle
(3) Detectable limit
(4) Determination of measuring location
As for measurement speed, we adopted two type, that is quick measurement
and ordinary measurement. The quick measurement method measures radioactivity
in a wide area per one measurement by expanding a measuring area with
increasing detector height. This method is a quick measurement, but detection
efficiency is not good. On the other hand, the ordinary measurement method is
an object of a narrow area by taking the detectors close to the object.
Accordingly, detection efficiency is high, but measuring time increases.
When we load the measuring system on a vehicle, there is a dead angle at
the area corner in measuring. Therefore, we can absolutely not make a measuring
area nothing
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JAEM-Conf 95-015
Our goal is to be below or equivalent to the level of natural
radioactivity in detectable liiit and to be as short as possible in measuring
time.
We are planning to adopt an optical distance measurement for determining
measured locations, because this method is widely used as an auto-survey
instrument in the field of the civil engineering and the accuracy is
satisfactory.
2. 3 Experiment on response function method
We tried to make some experiment on the response function method in order
to estimate radioactivity from measured count rates. As the positions of
detector and source illustrated in Pig. 5, we divide a square area of 4 m x4
i into array 4 by 4 and put a point source at a mesh center. Di, D2> D3 and so
on are detector position, and Si, 82. 83 and so on are source position.
Now that we put a point source at Si mesh center, we measure count rates
above each mesh center. Similarly, we change the position of the point source
from Si to 82, we measure count rates above each mesh center. From this
measured results, we estimate the activity of the point source by the response
function method. The resulting values are shown in Fig. 6. Figures are total
counts in 120 sec. 12%, 2.71 etc. are errors of the estimated source. It is
possible to apply the response function method to the source estiiation.
3. Conclusion
Our conclusions are conducted on measuring residual radioactivity:
(1) It is possible that an interesting nuclide is 60Co. Natural radioactivity
scatters. There is no mutual relation between 40K and the other nuclides.
(2) Measuring time and work increase enormously. It is difficult to
distinguish the existence of residual radioactivity.
(3) We made a concept Design of measuring system. The system is equipped with
four Nal detectors in array and the side shield for reducing background. In
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JAERI-Conf 95-015
addition, it can make quick measurement and ordinary measurement.
(4) The response function method is applicable to deteriination of
radioactivity.
3. Future plan
Our program is going on development. Therefore, some study concerned was
outlined. Our future plan is as follows:
(1) To generate a program to process and arrange the measured data
(2) To manufacture measuring system with concept design
(3) To make field test of measuring system
- 296-
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Detector Assembly
-Col 1imator
Nal Detector
Nal Detector
Nal Detector
Nal Detector
Temp. Detector
P re-amp.
P re-amp.
P re-amp.
P re-amp.
Temp. Compensator -> Hi-tension
MCA
Personal
Computer
A
Interface
A
StabiIized Power Source
Location detector
Residual Radioactivity Measurement System
Residual Radioactivity
Evaluation System
<=>
Floppy
Personal
Computer
Printer
Vehicle
i—4
is
Fig. 1 Measuring System
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JAERI-Conf 95-015
1 0 0 T
:. Lower side shield
1
Fig. 2 Shielding characteristics of Nal detector
for natural radioactivity
-298-
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/ \y*'
/.
,••
. ': .0-
.O • •
lay ui
/
Zr e» /
- /
.
.-+''*
"K '
t "?
O U
^
r'
v "
1 **"
N,..-<
3MUC •• -~
g-:~^ ' ^^^-g
* m
i Array dista
B
K.
*. ' (3
% * v
: ID UVI
• • *
>' t ''' ° '« 0 : '
150 CM* * 9 '•<
nee
a • .
;\
5 0 cm
3
r--. :X
*%4, :H
> :.
: O
-75.0 -60.0 -45.0 -30,0 -15.0
O.o 15. 0 30. 0 45. 0
Distance from four detectors' center
o
Ut
k
Lrt
60. 0 75. 0
Fig. 3 Detector efficiency of four detectors (Heigh 80cm)
-------
O
O
o
c
(D
O
CD
c
o
o
(D
4-»
(D
Q
0.004
0.0035 -
0.003
0, 0025
0.002
0.0015 -
Array distance
a^^?--g...-g^N4
W -71- ' ^<,
77 • ;• •: • 75 cm ; -?V
0.001
Array distance 0 cm
o.
150 CM
-50.0 -40.0 -3Q.T -20.0 -10-0 0.0 10.0 2070" 30.0 40.0. 50.0
Distance from four detectors' center
w
73
o
Fig. 4 Detector efficiency of four detectors (Heigh 20cm)
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JAERI-Conf 95-015
Detector position
Di
V
•s,
A
DM On On
o o
s.
S,,
Du
Su
S,
/ / / '-V
S,,
s,,
Si,
s,
IB la IB
X
/
Source position
Fig. 5 Experiment on response function method
Si a
3.34B+6
(12%)
S0
2.90B + 6
(-2. 7X)
S5
2.90E+6
(-2. W
S,
3.34B+6
(121)
SM
2.90B+6
(-2. 7X)
Si o
3.22E+6
(8. IX)
S.
3.22E+6
(8, 11)
S2
2.90E + 6
(-2. 7X)
Si.,
2.90B+6
(-2. 7X)
Su
3.22E+6
(8. IX)
S7
3.22E+6
(8. m
s,
2. 90B+6
(-2. 7X)
S]g^_
3. 34B+6-*-
(121)^.
S[ 2
2.90B+6
(-2. 7X)
S8
2.90B+6
(-2. 71)
S4
3.34E+6
(121)
•Source position
Counts
~Error
Fig. 6 Source strength estimated by
the response function method
- 301
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JAERI-Conf 95-015
4-4 Radiochemical Analysis of Homogeneously Solidified
Low Level Radioactive Waste from Nuclear Power Plants
Kaneaki SATO, Yoshihiro IKEUCHI and Hideo HIGUCHI
Division of Research and Training
Japan Chemical Analysis Center (JCAC)
295-3, Sanno-cho, Inage-ku Chiba-shi, Chiba 263
1. INTRODUCTION-.
In Japan we have U8 nuclear reactors in the 18 nuclear power
plants which generate about 30$ of all electricity. As the result of
the operation of those plants, the electric power companies have
stored a lot of solid waste package in their sites. The Japanese
Government has permitted the Japan Nuclear Fuel Ltd.(JNFL) to dispose
such solid waste packages into shallow land, according to several
safety controlled manner laid down the law. The shallow land disposal
in Japan was started by JNFL at Rokkasho-mura, Aomori prefecture, on
December 8, 1992, and planned to bury 200,000 drums of low level
radioactive waste(LLH) from nuclear power plants until 1998. The
inspections, such as visual inspection and non-destructive analysis
by 7 -ray spectroraetry prior to buring LLW are carried out according
to the law.
As one of JCAC1s projects entrusted by the Science and Technology
Agency(STA), we have been studying to establish the radiochenical
analysis methods in order to verify the radioactivity in LLW
calculated by the correlation factor since 1986, and carrying out the
reliability test on the methods since 1990. The radiochemical
analysis methods have been already prepared for long-lived or
critical nuclides in homogeneously solidified LLW from nuclear power
plants.
In this paper, the analytical methods developed by JCAC and
results of reliability test are described.
302
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JAERI-Conf 95-015
2. TYPE OP WASTE AND RADIONUCLIDES FOR ANALYSIS.
The LLW generated from the nuclear power plants are normally
classified into homogeneous and miscellaneous waste. At present, it
has been permitted by the Government to dispose 6 types of
homogeneous waste listed in table.!.
Long-lived nuclides such as 3H,1«C,5?Ni,63Ni,*°Sr,*»Nb,99Tc,l29I,
a. emitters such as 238Pu, 239Pu, 2»°Pu, 2*1Am, 242Cra, 243Cm and 2**Cm
and key nuclides such as 60Co and 137Cs as critical radionuclides in
waste for the analysis have been selected by the JNFL. The STA
commissioned JCAC to establish the radiochemical analysis methods in
order to verify the radioactivity of these nuclides in waste.
3. PREPARATION AND ANALYSIS.
3.1 Determination of 3H and '*C.
The combustion method is used for the preparation of the analysis
of 3H and l*C. The combustion apparatus is shown in figure 1. The
sample is heated in quarts reaction tube with hopcalite as catalyst
while passing oxygen, and then 3H and '*C are released from the sample
as water and as carbon dioxide, respectively. Water released is
collected in a cold-trap, and is distilled until dryness. Distillate
is mixed with emulsion scintillator for measuring 3H activity. Gases
released are passed through wash bottles filled with phenethylamine-
methyl alcohol mixture for '*C trapping. Toluene scintillator is
added to the mixture for measuring '"C activity. The recovery of more
than 90% for 3H or '*C was obtained. The analytical procedure for 3H
and '*C is shown in figure 2.
3.2 Determination of 5»Ni,*3Ni,*°Co,90Srf'»Nb and 137Cs.
The analytical procedure for 59Ni,63Ni,60Co,'0Sr,91
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JAERI-Conf 95-015
sample In order to determine the chemical yield, and the sample is
decomposed with acid. Ni,Co,Sr and Nb are precipitated as acid soluble
carbonate group.
Activity of J37Cs in the solution is determined using a Ge
detector after the collection of Cs by ammonium phosphomolybdate.
Nb is precipitated with dilute hydrochloric acid, and purified by
anion exchange method. Then the '"Nb activity is measured using a Ge
detector.
Activity of &0Co is measured using a Ge detector after
purification by anion exchange method.
Ni is purified by the extraction of Ni-dimethylglyoxime into
chloroform and back extracted with the hydrochloric acid. A portion
of aqueous layer containing Ni is mixed with emulsion scintillator
for measuring 63Ni activity by liquid scintillation counter, and
residual layer is electrodeposited on a copper plate for measuring
59N1 activity by a low energy photon Ge spectrometer.
Sr is mainly purified from other elements by the precipitation
method. The 90Sr concentration is determined from 90Y activity
measured by a low back ground GM counter(LBC).
3-3 Determination of "To and 129I.
The analytical procedure for "Tc and I29I is shown in figure U.
9 9 •" T c and iodine carrier are added to the sample for the
determination of chemical yield, and the sample is decomposed by
alkaline fusion technique, because both nuclides are highly volatile
elements.
129I is extracted with carbon tetrachloride and determined by
using neutron activation analysis.
"Tc is mainly purified with anion exchange method. And "Tc
activity is measured with LBC.
3.tt Determination of a activities.
The most a nuclides are considered to be transuranium elements
304 -
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JAEM-Conf 95-015
such as 238Pu,239+2i|0Pu,2l"Am,2«2Cm and 2*3*2»*Cm. According to the
results of calculation by computer software developed by ORNL, the
total amount of these nuclides covered more than 99% of total a
activities in LLW. These elements are mainly purified with anion
exchange method, then are electrodeposited on stainless steel plate
for counting with a silicon surface barrier detector(SSB). Pu.Am and
Cm are analyzed individually, and the values of each nuclide obtained
are summed up. The analytical procedure for 238Pu,23*+2*°Pu,2i|1An),
z%2Cm and 2*3+2**Cm is Shown jn figure 5.
M. TIME REQUIRED FOR ANALYSIS AND DETECTION LIMIT.
Time required for analysis and detection limits for each nuclides
are shown in table 2. It was found that the time required was proper
for routine analysis. The detection limits are very low and
considered to be very useful for verification the radioactivity in
LLW.
5. DECONTAMINATION FACTOR
Decontamination factor is very important key factor to analysis of
ft emitter. .
Typical decontamination factors are shown in table 3. -
Most of the decontamination- factors are greater than 10s for many
radionuclides investigated, except 137Cs and 60Co which are 7
emitters. • • . •
6. ANALYSIS OF MOCK SAMPLE
6.1 Preparation of mock sample
Mock samples for the verification of the reliability of the
-305
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JAERI-Conf 95-015
radiocheraieal analysis are prepared by using plastic, chemical
reagents and radiotracers. The preparation procedure of mock sample
for BWR evaporator concentrates is shown in figure 6 as an example.
Other types of mock samples are prepared in a similar way as this
one.
6.2 Results of radiochemical analysis
Table tt shows an example of the results of mock samples by
radiochemical analysis .
The reliability of each radiochemical analysis was estimated by
comparing the analyzed value of each nuclide with the spiked one. The
reliability of 59Ni and **Nb analysis were estimated on the recovery
of amount of added carrier, because these nuclides were not available
in our laboratory. The analyzed value agrees with the spiked value
for each nuclide, and heigh chemical yields were obtained for 5*Ni and
**Nb analysis. The results of other types of mock samples also are
satisfied.
7. CONCLUSION
As mentioned above, we have reliable radioanalytical methods for
all kinds of homogeneously solidified wastes.
We are now under studying an analytical method for pellets which
are made from evaporator concentrates or resin.
And we are going to study to establish new analytical method for
the rad-waste including metal, cloths and so on in near future.
- 306 -
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JAERI-Conf 95-015
Table 1 Type of waste.
1) Evaporator concentrates solidified with cement
2) Ion exchange resins solidified with cement
3) Filter sludge solidified with cement
4) Evaporator concentrates solidified with asphalt
5) Evaporator concentrates solidified with plastic
6) ion exchange resins solidified with plastic
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JAEM-Conf 95^015
Table 2 Time required for analysis and detection limit.
Nuclide Time required for Counting Time Detection Limit
Analysis (day) (rain.) ( Bq/ g )
H- 3 3 SO 8x10-3
C-14 30 4 x 1 0 ^3
N i - 5 9
N i - 6 3
C o - 6 0 40
S r - 9 0
Nb- 9 4
C s - I 3 7
T c - 9 9 12
1-129
P u (a)
Am- 241 14
Cm (a)
3
3
3
3
1 2 0
3
1 2 0
3 0
3
3
3
0
0
0
0
0
0
0
0
0
0
0
4
3
2
1
7
6
2
2
7
7
7
x
x
X
X
X
X
X
X
X
X
X
1
1
1
1
1
1
1
1
1
1
1
0
0
0
0
0
0
0
0
0
0
0
-3
-2
-2
-3
_,
-3
-,
-5
"
-4
-n
-308 -
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JAERI-Conf 95-015
Table 3 Typical decontamination factor.
Decontamination Factor
Added nuclides
(chemical form). H-3 -C-14
Analysis
Nh59»63 Co-60 Sr-90 Nb-94 1-129 Tc~99 Cs-137
H-3 CH20)
C-!4(Na2eQ3)
4xiOJ
Fe-59CFeClj)
Co-eOCCoCh)
Zn-SSCZnCU
Sr-85($rClt>
Ru-106(RuCI,)
S6-125(SbCI,)
Cs-l57(CsCl)
Ce-I44(CeCl3)
8X102
>7X105 6X105
>2X10S >2X10S
>3XJOS JXJOS
5xlOs >5X105
ixiO6 1X10S
>8xiOs >8xiQs
>2Xi05 >1X10S
>1X105 >1X105
>2X10$ >ixiQ5
>2xi05 >2xiOs
>JX10S >1XJ05
>2X105 >5xiOs
JXJO5 IxJO5
>7X105 5X102 1X105 1X103 8X105 >2X10S 4X104
1X107 3X102 >3X10S 1XJ05 1X105 >6X105 >2X103'
>6xjQ5 3X1Q2 >6XJOS 9X103 >JX10S >1XJ05 >2X103
5X10S - >3X105 1X105 >4X10* 4X10S 2x]03
3X10* 5X102 >5x305 2X105 2X10" >4X10J 2xi04
>5xi05 >3XJ03 >4X105 1X104 >3X10S >4xiQ3 >3X103
5xiOs 8X10' >4X105 }X1Q3 >2X105 2X105 3xl02
>2X10S 3X10' >3xi05 IXIO" >9X10" >!X105 ixfo2
>1X10S >4X102 >4X10S 1X10° >2X1QS >2xi05 >2xio2
>ixl05 ixiO5 >3X10S >2xiO'! — — 2X10S ?X103
>JX|05 >6XI02'>3XI05 6XJ03 >3XIOS >3X10J >3xl02
4XJ05 >2xl03 >5X10S ixjO4 ixiO4 ixi0s -
3XJ05 >ixiOJ >2XJ05 9XI03 >2X105 4XjQ5 3XJQ2
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JAERJ-Conf 95-015
Table 4 Results of mock samples by chemical analysis.
BWR evaporator concentrates solidified with pIastic(radionuclide-spiked)
Radionucl Sde
Analyzed value
(Bq/sanrple)
H-3
C- 1
N i -
C o-
S r -
Cc
5> —
TT r»
1 _ -J
p u_
A m —
Cm-
4
6 3
6 0
9 0
1 3 7
9 9
2 9
242
2 4 1
244
34 ±
96 ±
53 ±
130 ±
49 ±
160 ±
50 ±
51 ±
0. 020±
0,51 ±
0. 096±
Recove
1
1
2
10
1
3
1
1
0.001
0.02
0.004
ry of
Spiked value
Okt/sarnple)
34.4
101
48.8
125
53.9
148
51.2
50.6
0.0185
0. 548
0.102
carrier
Ana ! yzed
Spiked
0. 99±0. 03
0. 95±0. 01
1.09±0.04
1.04±0. 08
0.91 ±0.02
1.08±0. 02
0. 98±0. 02
1.01 + 0.02
1.08±0. 05
0. 93±0. 04
0. 94±0. 04
Radionucl ide
N i —
Nb-
5 9
94
86
92
- 310-
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boat-
reaction tube (quartz)
oxygen
flow meter
I
catalyst (hopcalite:Mn02,CuO»Co203,Ag20)
electric furnace
quartz wool
U-tube
gas washing bottle
oxygen bomb Defar> s bottle
I
o
a
freezing mixture (ice+NaCl) phenethylamine-methanol
Figure 1 Combustion apparatus.
-------
JAERI-Conf 95-015
(3H)
condensed (H20)
distillation
Sample ( 5— 10g )
«- 02 gas
heat
cold trap
not condensed (C02)
emulsion scintillator
absorption in
phenetylamine-
methyl alcohol
L S C counting
toluene scintillator
L S C counting
Figure 2 Analytical Procedure for 3H and ]*C.
312
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JAERI-Conf 95-015
Saraple(lOg)
<- carrier(Ni,Co,Sr,Nb,Cs)
heat
acid
sodium carbonate
precipitate
(Ni,Co,Sr,Nb)
filtrate
CCs)
(Nb)
(Ni»Co,Sr)
precipitate
filtrate
G e counting
anion exchange
resin
G e counting
(Ni.Sr)
anion exchange
resin
dimethylglyoxime — chloroform
(Co)
(Sr)
Y precipitation
(Ni)
G e counting
9 0
electrodeposit 59Ni
& 3
Ni
L B C counting
(low background GM)
L E P S counting
(low energy photon spc.)
L S C
counting
Figure 3 Anlytical procedure for 59Ni,63Ni,6°Co, *°Sr, 9*Nb and
137Cs.
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JAERI-Conf 95-015
Sanple(IQg)
carrier(99mTc,I)
fusion
extraction
carbon tetrachloride
aqueous layer
(TC)
ion exchange
resin
electrodeposit
7 -ray spectrometry
of 99inTc with Ge-detector
("TC)
L B C counting
(low background GM)
organic layer
(I)
backextraction
neutron activation analysis
G e counting
'2*I+n=I30I
Figure U Analytical procedure for "Tc and 129I.
- 314 -
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JAERI-Conf 95-015
Sample(lQg)
heat
acid
precipitate
filtrate
(Am,Cm)-
(T.h)
hydrogen peroxide
anion exchange
resin
[Pu* *
electrodeposit
S S B counting
anion exchange
resin'
(x3)
electrodeposit
(Am, Cm)
S S B counting
Figure 5 Analytical procedure for 238Pu,2 39 *
2»2Cm and 2 »3 t2 « "Cm.
,2 * 1 Am,
- 315-
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mix •$-
JAERI-Conf 95-015
plastic UOwt.Jg
sodium sulfate 5H.6%
ferric chloride 2.7%
sodium chloride 2.7%
solution including radiotracer
stir (for Ihr)
evaporate to dryness
grind (homogenize)
sample (10g)
Figure 6 Preparation procedure of mock sample.
(BWR evaporator concentrates solidified with plastic)
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JAERI-Cotrf 95-015
4-5 Classification of Solid Hastes as ion-radioactive Hastes'
M. Suzuki, H, Tom i oka, K, Kara ike and J. Komatu
Research Association for Nuclear Facility Decommissioning
ABSTRACT
The radioactive wastes generally include nuclear fuels,materials
contaminated with radioactive contaminants or nuetron activation to be
discarded. The solid wastes arising from the radiation control area in nuclear
facilities are used to treat and stored as radioactive solid wastes at the
operation of nuclear facilities in Japan. However, these wastes include many
non-radioactive wastes. Especially,a large amount of wastes is expected to
generate at the decommissioning of nuciear facilities in the near future. It is
important to classify these wastes into non-radioactive and radioactive wastes.
The exemption or recycling criteria of radioactive solid wastes is under
discussion and not decided yet in Japan. Under these circumstances, the Nuclear
Safety Committee recently decided the concept on the category of non-radioactive
waste for the wastes arising from decommissioning of nuclear facilities.
The concept is based on the separation and removal of the radioactiveiy
contaminated parts from radioactive solid wastes. The residual parts of these
solid wastes will be treated as non-radioactive waste if no significant
difference in radioactivity between the similar natural materials and materials
removed the radioactive contaminants.
The paper describes the procedures of classification of solid wastes as
non-radioactive wastes.
BACKGROUND
The development and utilization of nuclear energy in Japan have been
promoted since the starting of Japan Research Reactor(JRFH) operation and more
than thirty years has passed. As of January 1994, 46 nuclear power plants are
operating with about 37,000 Mff electricity generation in order to supply
electvic power to industry and the public in Japan as the most stable and
secure energy.
* The work is carried out under the contract of the Science and Technology
Agency.
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JAERI-Conf 95-015
fhile, the decommissioning of nuclear power plants and fuel cycle facilities
at the end of their lives has become an important issue. In addition, it is
expected that renewal and refurblishraent work will increase by the technology
developments. In particular, a large amount of wastes is expected to generate
at the decommissioning and renewal and refurblishment work.
The radioactive wastes generally include nuclear fuels and/or substances
contaminated with radioactive contaminants or neutron activation to be
discarded. In Japan, the solid wastes arising from the radiation controlled
areas in nuclear facilities are used to be treated and stored as radioactive
solid wastes at the operation of nuclear facilities. However, these wastes
include many non"-radioactive wastes. It is an inevitable matter from the
economic and safety points of view to establish the effective effective waste
minimization, rational treatment and disposal of dismantled wastes. Therefore,
it is important to classify these wastes into non-radioactive wastes from
radioactive wastes.
'On the other hand, the exemption or recycling criteria of radioactive solid
wastes is under discussion and not decided yet in Japan. Under these
circumstances, in June, 1992, the Nuclear Safety Committee decided the concept
on the category of the non-radioactive wastes for the wastes arising from
decommissioning of nuclear facilities.
Based on the concept , the technically available procedures on the
classification of solid wastes as non-radioactive waste have been studied and
recommended.
BAB IC CONCEPT
The radioactive contamination of solid wastes from nuclear facilities is
generally classified into Secondary contamination due to the adhesion and
permeation of radioactive substances and activation contamination due to
neutron capture.
According to the concept by the Nuclear Safety Committee, the following
wastes are regarded as non-radioactive wastes.
1. In the case of Secondary contamination
1) Those wastes which are apparently free from Secondary contamination
due to the adhesion and permeation of radioactive substances, in the
light of their history of use and their conditions of installation.
2) Those wastes whose secondary contamination due to the adhesion and
permeation of radioactive substances is limited and whose contaminated
parts are separated and removed in the light of their history of use
and their conditions of installation.
2. In the case of Activation contaminationCConcrete waste including
reinforced bars)
1) Those concrete wastes which need not be taken into account the
activationby neutron in the light of the structure of facilities, such
-318-
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JAERI-Conf 95-015
as being sufficiently shielded by shielding materials,
2) Those concrete wastes for which the activation levels are calculated
of no significant difference from the concrete used in non-nuclear
field.
3) Those concrete wastes for which the activation levels are calculated
and whose significantly activated portions have been removed.
PROCEDURES OF CLASS IFI CAT I ON
There are secondary contamination due to the adhesion and permeation of
radioactive substance and activation contamination due to neutron capture in the
nuclear reactor facility. Basded on the concept of the Nuclear Safety
Committee, we have studied the technically available procedures on the
classification of solid wastes as non-radioactive wastes and have recommened the
practical identification method as shown in Figure 1.
The solid wastes will be technically judged and evaluated as Non-radioactive
or not in accordance with the following criteria of the judgement and
evaluation.
1. The solid wastes from decommissioning of nuclear facilities are judged to
be free from the radioactive substances or not by the past records and da
ta.
2. In the case of secondary contamination, if the contaminated parts are lim
ited, the residual parts after separation and removal of contaminated par
ts are regarded as non-radioactive wastes. The contaminated parts to be r
emoved are classified into radioactive wastes. If the limitation of the c
ontammated parts can not be identified, the wastes are treated and store
d as radioactive wastes.
3. In the case of activation contamination, if the radiation level of neutron
activated wastes is equal to(or less than) natural background, the wastes
can be categorized as non-radioactive wastes.
ITEMS ON PAST RECORDS AND DATA
The items on the past records and data available to judge the solid wastes
as non-radioactive wastes are based on the following:
1. Structural design feature of reactor facility ( shielding wall, trench for
the contaminated water etc.)
2. Radiation control records ( routine survey data of the surface and the
air in the controlled areas)
3. Operational history of the reactor
1) Surface contaiination measurements(Bq/cmz)
2) Concentrations of radioactive substances in the air of the controlled
areas(Bq/m3)
3) Experience of decontamination
4) Experience of accident
5) Decision or release of the controlled areas
- 319-
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JAERI-Conf 95-015
The items mentioned above are useful in judging whether the wastes from the
areas being dismantled is apparently free from the radioactive substances or
not. Especially, the records on the experiences of accidents or the data
related to contaminations and decontaminations are important to judge and
evaluate the wastes as non-contaminated wastes. Therefore, It is desirable that
these records and data are reserved until the beginning of decommissioning of
reactor.
LIM1 TAT I ON OF CONTAMINATED PARTS
1. MetaI
For the metals cI added or coated with some materials, the contamination of
radioactive substances is limited at the surface of the materials covered.
Therefore, It is easily possible for the metal to be regarded as non-radioactive
waste by removing the contaminated cladding or coating materials without cross
contamination.
In the case of existense of defects on the layer of the matal surface
without the cladding materials, there is a strong possibility that the
contaminated substances may be permeated into some depth of metal. In this
case, there is a need to identify the depth of the contaminated parts. An
example for settling on the limitations of the contaminated parts in the metal
without the cladding materials is shown in Figure 2. In this figure, by
removing the contaminated surface piece by piece from the original surface, no
radioactivityt was detected at a certain depth. The surface contamination level
at this depth represents the lower level of detection(LLD) for 30 min. counting
times, and then, the measurements for this surfase were continued to increase
the detection sensitivity for 1 day, 2 days, and 3 days, respectively. The
results of the measurements were less than the lower lever of detection without
exception. Therefore, this depth is recognized as the limit of the contaminated
parts in the metal. If the contaminated layers can be separated and removed
without cross-contamination, the residual parts are regarded as non-radioactive
wastes.
2. Concrete
The radioactive contaminants adhered to the surface of the concrete seem to
penetrate into lower layer by the diffusion phenomenon. However, in any case the
limitations of the contaminated portions in the concrete are recognized
actually at the operation in the nuclear facility because the contaminated
substance, at the surface if there are, are removed as soon as possible, from
the point of safety central.
In Figure 3 is shown an example related to the limitation of the
contaminated portions in the concrete without the coating materials. In this
case, no detection was appeared at a certain depth from the surface. In common
with the case of Figure 2, the measuments for the same surface at this depth
were carried out for about 8 h, 20 h, 40 h and 140 h, respectively. All of the
results were less than the lower level of detection. Therefore, this depth is
- 320-
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JAERI-Conf 95-015
regarded as the limit of the contaminated portions sn the concrete.
CONCEPT ON THE CLASSIFICATION OF NON-
RADIOACTIVE WASTES IN CONCRETE OF
REACTOR
As the basic concept mentioned above, neutron-activated concrete wastes of
which the radiation level is equal to ( or less than ) the natural background
are regarded as the non-radioactive concrete wastes. The nuclides in the
natural concrete consist mainly of K-40,Ra-226 and Th-232. According to the
reports on the radiation levels of the natural concrete in Japan, there are
considerable deviations in the background levels of the natural concrete
available.
Based on the basic concept of Non-radioactive waste the classification of
non-radioactive portions in the concrete can be defined as shown in Figure 4,
The A point in this figure is regarded as the radiation level corresponding to
be equal to the natural background. In general, the neutron-activated levels
can be calculated by using the available computer cords such as ANISN, DOT3. 5
and ORIGEN etc.. and then, the A point is resulted from the cross points of
calculated activation values and the background level of natural concrete. As
the reasonable width of the deviation in the background level, we adapted three
times the standerd deviation (3
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JAERI-Conf 95-015
CONCLUS!ONS
The practical methods of the classification of solid wastes as non-
radioactive wastes for executing the concept of non-radioactive solid wastes,
are recommended.
1. The items on the pasts records and data to judge the solid wastes as
non-radioactive wastes are recommended.
2. The technical criteria and procedures for judging and evaluating the
solid wastes as non-radioactive wastes are recommended in the case of
secondary or activation contaminations.
ACKNOWLEDGMENTS
The authers express thanks to the members of study group on this matter and
the staffs of the Nuclear Safety Bureau of the Science and Technology Agency for
thei r works.
REFERENCES
[1] The Radioactive Concentration Upperbounds for the Safety Regurations on
the Shallow-Land Disposal of Low-Level Solid Radioactive lastes. ( The
Second Interim Report), Special Committee on the Safety Standards of
Radioactive lastes. Nuclear Safety Commission, 1992.
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O9
to
OJ
Secondary
Contamination
The [imitation of
the contaminated
parts
Wastes from
decommissionig
of Nuclear
facility
Judgementof
no contaminated
materials by
the post records
NO
YES
YES
The separation
and removal of
the contaminated
parts
NO
YES
NO
Radioactive Wastes
Final
survey
NO
The separation
and removal of
the activated
YES
Non-radioactive
wastes
§
k
Ul
Activation
contamination
Difference fi
natural back
level
"om
ground
-------
1000
t\j
E
o
•x
Q-
CO
"55
>
CD
— 1
C
TO
C
TO
.4_J
C
o
a
CD
O
TO
4—
S—
13
00
100
10
0.1
0.01
LLD 1 Days
Radioactive Wastes
LLD ZDays
LLD 3 Days
^on-Radioactive Wastes
w
50
10
Depth^m)
10°
1000
Fig.2 Limitation of the Contaminated Parts in Metal
-------
10'
to
£*
i
O"
QQ
c
_o
4_>
CO
i_
•M
c
0)
o
c
o
u
CD
u
TO
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CO.
10s-
10"
TO2
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o > n
Redioactive
wastes
-------
OJ
O"
CQ
ID
>
0)
_J
4)
O
C3
O
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TO
a:
Neutron-Activated Level
Neutron-Activated Level
.4-
B.G.Level of Natural Concrete •
B.G.Level of Natural Concrete
Radioactive Wastes
Non-radioactive Wastes
Deviation
in
B.G.Level of
Natural
Concrete
Distance from the Reactor Core (cm)
Fig.4 Concept on the Classification of Non-radioactive
Portions in the Concrete of Reactor
w
I
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Free Discussion and Summary
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JAERI-Conf. 95-015
FREE DISCUSSION AND SUMMARY *
Mr. MacKinney: This session is being recorded so that the proceedings can include this
final discussion. There will be a microphone which will be passed around, two microphones
actually, and if each individual who would like to share a comment or observation would raise
his hand, then we can send the microphone to him. Bob [Meek]?
Mr. Meek: I have a suggestion on the agenda. Do you want to expand the agenda beyond
the residual radiation criteria to include recycle-reuse? Or was that intended?
Mr. MacKinney: That is the intent. The intent is to include residual radiation and
recycling and reuse, but in particular, recycling. Both countries have a bad problem at this
point with recycling, and the Workshop is somewhat more focused on recycling. And, the
question that I have placed here [on the chalk board] is: "what does it take to achieve residual
radiation criteria in Japan and the United States" in terms of some issues, which I have stated
below.
What does it take to achieve residual radiation criteria in Japan and
the U.S. ?
• risk assessment, tools, approaches
• costs vs. benefits
* technologies
• other impacts analyses
- industries
- environment, etc
• other ?
If there are other things I did not put up here which would aid our two countries in achieving
criteria, then we will discuss those also. Gene [Durmanj?
Part of the tape recording was barely audible, and we had to do a fair amount of editing. Hopefully
there are no serious mis-representations. The Editors.
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Mr. Durman: John, I would like to ask a question. I would be very interested in a reaction
either from my colleagues in the United States, or from my colleagues in Japan, The
question is: what role does metal melting have in a potential policy for recycle and reuse? It
seems like a very promising technology; it seems it can achieve significant reduction in
radionuclides. Should it be a central part of a recycling-reuse policy? Should metal-melt be
assumed to be a part of such a process? What technical issues remain with that, and what are
your views on its role in a policy?
Mr. Burns: This is an area that I've looked at quite a bit at Fernald. And, I want to pose a
reservation, or caution, about targeting an individual technology. I think it is more
appropriate for the people involved in setting standards and regulations to realize what metal-
melt does, that is the desired effect, and what makes it promising. It results in
decontamination and reduction in hazard level of a raw material, and it also homogenizes the
contaminants so that they are more easily measured, understood, and the effects of the
contaminants can be analyzed and quantified for pathway analysis. So I am in total
agreement that technologies that do those two things need to be at the core of what we are
talking about. I would just express some reservation about specifying a discreet technology,
because others may come out with innovative thinking and creativity that may accomplish the
same task, and may even accomplish it better.
Mr. Dam: That is a good point. We do not need to specify an exact method, if we know
what the technology can do, and maybe support some R&D efforts to make it better. For
example, better partitioning with different kinds of fluxes, different ways of getting
radioactivity out of metal, so we don't end up with too homogeneous of a melting, but
recognize there are other techniques that also work well, particularly for surface
contamination. It is more important we get reasonable standards set for both surface and
volumetric contamination, recognizing there are a lot of different techniques.
Mr. Subbaraman; Again, following up on the same topic, I think rather than on the policy
level, the technologies and their limitations should be addressed during implementation.
Regarding implementation, one of the things I see our colleagues in Japan have accomplished
that we need to look into further, is they already have the scrap metal industry, that is willing
to work with recycled radioactive scrap. Mitsubishi, for example, has done melting tests.
Whereas contrary to that, in the United States the scrap metal industry is very adverse right
now, because inadvertently contaminated materials are introduced into scrap where they don't
belong. The problem is a significant concern. Thus, as part of this implementation it would
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behoove us to bring in that industry and make them aware of what is happening, for example,
in Japan. Thank you.
Mr. Durman: I am interested in the Japanese perspective on the issue of melting in recycle-
reuse. How is it a factor in your thinking at this point? I don't know, for example, what
Mitsubishi is planning to do as a part of its future activity in this area, and the kinds of issues
they see as being important to examine.
Mr. Fujiki: Fujiki, JAERI. My opinion is somewhat limited, and some parts are personal at
present. Melting is not a universal way. It is quite an important method in the course of
recycling or reuse of the metal So, for example, even decontaminable surface contamination
gets into the metal after melting. So I still think that in a future system of recycling of metal
wastes, some optimization of a combination of the surface decontamination and direct melting
will establish a reasonable system. That is one point. And another point is, the Japanese
people, or the Japanese nuclear society, is somewhat behind in establishing such criteria for
free release and reuse, but we should not wait until some people decide on such criteria, that is,
we need to proceed continuously, so even before the criteria are established, we will take
steps to advance toward future recycling. For example, JAERI has a lot of metal scrap,
radioactive metal scrap, so our very urgent task is how to reduce the volume of RSM and
store it rationally, until we have some criteria, or we have a way to dispose of this waste in
some site. It's a very urgent task. So, to partly resolve the issue we will take some steps to
reduce the volume by melting. It is now planned in our Department, but we need very good
data for future reprocessing of metal for some reuse purpose, or to dispose. So I think that
we need a very sophisticated way to track the activity and inventory, and measurement for
such an approach.
Mr. MacKinney: Scott [Dam].
Mr. Dam; I think what industry, anyway - I may be one of the only people speaking for
industry at the moment - really wants the Agency [U.S. Environmental Protection Agency] to
do is to set standards, as you are proceeding to do, which allow and/or encourage an
appropriate recycling and reuse, but don't get so prescriptive on the methods, for exampie,
working with steel companies on how to do recycle. Industry can do that. I think we can
do that well. If you work very hard on getting standards set, industry will implement them.
There are criteria on the street now, and industry has implemented it. When the next set of
criteria come out, industry will implement it.
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Mr. MacKinney: Dan [Burns].
Mr. Burns: I would like to introduce a different subject. It occurred to me after listening
for the last several days that a couple of realizations have occurred. I don't believe that we
are. going to reach a consensus between Japan and the United States, or anywhere else within
the international community on what the cleanup level will be. Each country will have its
own needs. And that also goes for cost. In the United States, we are probably much less
willing to spend the amount of money that the Japanese may spend on the same type of waste.
Getting rid of the discussion of those two things [clean up levels and cost] and trying to find
some common ground, I think the real opportunity is going to lie in developing a consensus
among the methodologies to come up with those levels and those costs that we would each be
comfortable with. Having a common tool such as a code for computing hazard, no matter
what hazard level you adopt, the code would still be the same. I think there is a great
opportunity for this group of people to do that, and also the entire international community,
and the same thing can be said for cost also. The end product may be different, but the
methodology employed to come to the decision could be common, I would like to see that
as the emphasis between our two countries. I think we have a great opportunity to do that.
Mr. MacKinney: Are there any further comments about Mr. Burns' comments, or any
additional comments about the technology which we were talking about previously?
Mr. Yamamoto: I would like to make some comments about international trade in
connection with worldwide distribution of the radioactively contaminated metals. It is
necessary to reach a common standard level, or common criteria to be established in the
international market. I think it is also necessary for us to reach a consensus to establish such
a standard. So it should be encouraged to make a forum, an international forum, such as
OECD, or IAEA meetings, and we should concentrate our efforts on such international
activities to make the world cleaner. Thank you.
Mr. MacKinney: I would like to follow on that and ask our Japanese hosts; do you think
the people of your country would react strongly if the United States were to establish
radioactive scrap metal recycling criteria? Strongly in favor, or strongly against?
Mr. Yanagihara: I think that everything in Japan is very influenced by the American people,
so if there are some criteria made in America, especially, the [Japanese] Government will be
strongly influenced by that criteria. So I think we need a push from the American people to
the [Japanese] Government. I think that Japan is a very small island. We don't have so
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much travel to foreign countries, and also we do not have too many people from other
countries in Japan. I hope, if possible, people have a chance to come to Japan, and speak
loudly what you are doing in America. This is my opinion, my hope, for our job to go
forward.
Mr. Fujiki: I would like to address the question from Mr. MacKinney. Your question is:
"Can we support the reaction to free release criteria in the United States? " It's a fragile
situation at present, so it's somewhat difficult to answer, but even at present European
countries have already established criteria for free release. And the actual situation in Japan
is almost nobody knows that such criteria exist. So it is a problem of propaganda. To
receive such foreign criteria smoothly, or introduce such a situation smoothly into Japan, we
need some very cautious and careful studies, for release of basic information to the public, but
not to hesitate, to cover [up the issue]. That means some kind of public acceptance program.
For example, JAERI has been successfully promoting its program very smoothly in this village
[Tokai], by cooperation between the Institute and the local community. We need continuous
efforts to establish basic, exact, and valuable data and information to be released to the public.
So probably the next ten years will be the critical period, or a very important period to
establish such criteria, so we should perform the basic efforts step by step. And if so, it will
not be so difficult to receive internationally established criteria.
Mr. Suga: I saw an excellent factory in the United States, and I thought that Japanese
industrialization of melting metals from reactors was not developed so much, I think it's
behind about 10 years for industrialization. Now, Japan is doing JPDR decommissioning as
a pilot. But, many Japanese reactors are being decommissioned in a few years, and I think
Japan will see the necessity of melting and recycle-reuse. And then I think we will have to
introduce the United States' knowledge and continue such programs.
Mr. Durman: I have another general question on the slightly different topic of risk
assessment. It's a two-part question. The first is: are we close to achieving any consensus
on the appropriate scenarios that should be used. This is important because it is clear that if
you pick a certain scenario there is not much risk at all with recycling, but if you pick another
scenario there could be a great deal of risk. So, the scenario you choose is very critical to
the outcome that you receive. Is there any emerging consensus on what is appropriate in
terms of scenarios. And then secondly - perhaps for my Japanese colleagues - how
important is this risk assessment approach to Japanese policy? You heard the debate back
and forth from the U.S. delegates to this Workshop. Which scenarios do you use? Is the
whole risk assessment approach a major part of the way you would like to proceed?
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Mr, MacKinney: Yes, Mr. Kennedy.
Mr. Kennedy: I try to view this as an observer, because I was involved in the IAEA efforts
and [their] metal recycle document. But, the observation I had last year in the meeting in
Vienna, was that a number of countries read that work, Safety Series No. 111-P-l.l, and it
seemed that they were somewhat in agreement, and in consensus on the approach and
scenarios outlined. The careful position of IAEA was to set that information as guidelines
with several caveats - that national authorities would need to review their national situation
and develop criteria within the framework IAEA established, but specific to each country. I
think clearly U.S. efforts with EPA, NRG [U.S. Nuclear Regulatory Commission] and DOE
[U.S. Department of Energy] are doing that - looking at the national interpretation of how to
proceed with some guidance. Other countries are simply adopting the IAEA guidance. To
me that is a little unfortunate, because I think the situation improves the more minds and the
more talents are put on the problem. And, to end the debate too soon, I think, is a disservice.
But perhaps also carrying the debate too far is a disservice as well, because of the need for a
balanced criteria that's well reasoned. So, I guess my hope is that there is information to
provide a good starting point and that there will be a reasonable process in each country that
occurs.
Mr. MacKinney: Mr. Subbaraman.
Mr. Subbaraman: Thank you, John. On the question of the scenarios, I recommend that
generally, [when] scenarios [are] being examined, make sure of the variables in the integrated
risk assessment, through the selection of scenarios are understood, and the uncertainty of
those variables are modified as far as possible. The RESRAD code allows the user to pick
scenarios. For my part, I observed the subjectiveness in selecting those scenarios and I
recognized personally that I was making a quasi-judicial decision in selecting scenarios for my
site. I subsequently thought that in having to determine a whole host of things, dozens, and
finally put some scenarios together, there was no quality control, or quality assurance, or
uncertainty measurement that I could put in. It was not clear what the outcome of my
analysis was going to be. So, my recommendation is "yes", reference scenarios should be
created and variables for analyses be examined, and somehow quantify the uncertainty in those
scenarios. Thank you.
Mr. Warren: I would like to rephrase the question, if I may, and direct it based upon
information we heard in the last couple of days. In your presentations you seem to indicate
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your cleanup level is three slgma above background. Our cleanup levels are based on a risk
analysis. Which is better? Why? I guess, maybe you seem to be going on non-risk based, at
least for natural radiation. I know for plutonium it's going to be different. We didn't talk
about that since you cannot find it in the natural background unless you can use fallout. I'm
curious, why, even though it's a conceptual study, your emphasis seems to be on three sigma
over background versus a risk based standard. What is pushing you that way versus we, in
America, who are doing risk based?
Mr. Yanagihara: We presented in our presentation, that temporarily three sigma is some
kind of clearance level. Now, in Japan, discussion is ongoing what level is better. So to
determine the final clearance level, this is my opinion, 1 need a risk assessment approach. It's
very important. So at this moment we don't have a clearance level, but temporarily we use
three sigma for the clearance level. The basic concept, or idea, at this moment, is no artificial
radioactive materials will be released to the environment.
Mr. Fujiki: First I would like to make sure your clearance means changing the material to be
free from regulation. The concept of natural background plus three sigma, that Satoshi [Mr.
Yanagihara] mentioned, is a temporary solution, and also it is only applied to distinguish what
is not radioactive from radioactive. So, it's not free-release criteria for materials, that is
wastes. In your terminology, free release criteria is usually for radioactive waste. Your
question is on the site or materials? [from the audience "Both."]. Both. On the site, I'm
not sure, but on the materials, we bypassed [went ahead] and determined free-release criteria.
And before that, we developed the concept of non-radioactive material and radioactive
material criteria. So, there remains a very important task to establish, or adopt, or avoid the
free release criteria. Yes, we have a very important task still remaining. So, we should go
further.
Mr. Warren: We had a controversy a couple of years ago over "below regulatory concern".
You have not had that controversy, but you have an answer - three sigma over background.
Even though it is just a guideline, you are doing that. If we came forward in the United
States and said "three sigma over background is below regulatory concern", the answer would
be "No". So, even though you have not done it formally, you are doing it, and that is a non-
risk base decision. The British are doing the same thing. Why? Because at the starting
point you could have said [a risk of) "ten to the minus sixth" would be your starting point.
But, you chose three sigma as a starting point. What drove you to that decision?
Mr. Fujiki: The risk based concept is behind the decision in establishing the guide.
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Because, naturally occurring radioactivity inside the concrete will not be distinguished from
such a low level of contamination or activation. And so, I would ask the people in health
physics to confirm my opinion later. But, the basic philosophy of "below the naturally
occurring level" will not affect so much the health effect. I believe there is such a philosophy.
So, the guidelines of natural background plus three sigma is consistence with such a
philosophy. So, STA [the Science and Technology Agency, Government of Japan] adopted
this one, I believe that. So, the risk based concept is not quantified as a rigid number but it is
there.
Mr. Warren: Even for radium, and radon? That seems to counter your argument.
Mr. Kumazawa: I'm Kumazawa from JAERI. I'm not an expert in the field [of risk
assessment]. What Mr, Fujiki said, three sigma plus background, is reasonable based on the
level of the Japanese background. So, if the Japanese background is high, I think he cannot
say, we cannot adopt such a way of thinking. Anyway, I think the important thing is a
balance. If we have a lack of metal, we need metal even if the metal is contaminated.
Before, Mr. Yamamoto said we need a common criteria of contamination level for worldwide
trade. But according to my opinion, we cannot have a common level, we can only have a
regional level. Some countries want to have a very pure [low] level, and other countries do
not mind slightly contaminated level. So, I think [in setting criteria] we consider the real
situation - what we need, what is best for us to set. So, anyway, as general criteria we can
set one level. But, as I said, we need various other situations. We should consider other
various situations. The resource is very important for our life. So, anyway, the level of
contaminant is not so high, so, we should take a flexible approach, flexible way of thinking.
Mr. MacKinney: Let me intervene briefly. Because we are being recorded, and this will
be translated into script, please give your name, so we can attribute each of your comments to
you. We have only a few minutes left before two o'clock.
Mr. Suga: I am Suga. We control radioactivity dependent on the radionuclide. Some
radionuclides, such as cobalt, we control, health-physically, to about five percent of natural
radioactivity. Some contamination was found around the site of JPDR. We removed these
soils and we buried them, and as such, we act depending on the radionuclide.
Mr. Burns: Dan Burns from FERMCO. I want to speak to Mr. Warren's comments. In
my mind, I saw no correlation between the scientific definition of radioactive material versus
non-radioactive material, to BRC. It just didn't click at all. The reason being, is that it
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doesn't sound "risk-based" to me, it sounds like a definition. This is Japan's ability to say
something is natural or not, or artificial. I think that my interpretation of what they were
saying, was that we are going to great extents to show that we have not added to what is
naturally there, and they have defined that in some quantified, scientific, statistically-based
protocols. And three sigma to me sounds a little bit above and beyond the call of duty.
Mr. Warren; I agree that maybe the intent was not to do that. But, to use an American
colloquialism., "if it walks like a duck, and it looks like a duck, and it sounds like a duck, it is
a duck." And even if the intent was not to do that, in reality, you .have. I think it's a good
thing. We could not do that, but we probably should.
Mr. Burns: I'll make this quick. This is Dan Bums, again. I had the same analogy in my
mind. That is: if it walks like a duck, smells like a duck, looks like a duck, it's a duck. That
was what they were doing. If it walks like a duck, smells like a duck, is not radioactive; it's
naturally occurring. I guess it's just perspective.
Mr. Dam: I would like to change subjects, and go back to the original subject which Gene
[Durman] asked us about. Let's go back to the first part of Gene's question: are we close to
achieving consensus on the scenarios. From the presentations I heard today, yesterday, and
the day before yesterday, and the comments from the audience, I don't think we're close. I
think there's more work to be done. And I think, some sensitivity on some of the existing
scenarios, as well as, perhaps, looking at whether we covered the range, is appropriate.
Certainly the conclusions, whether they were thought to be conclusions or not, were
presented as such. I don't think we're there yet.
Mr. MacKinney; Let me just briefly follow on that question, and maybe we can get a few
more comments in. Is there a means by which we, meaning the United States and Japan,
could arrive at scenarios which are truly representative, or demonstrably representative, of
doses which might, or would, occur if metals were to be recycled, or concrete were to be
recycled?
Mr. Fujiki: Setting or establishing the scenarios reasonable for all cases is very difficult. So,
it's region wide, or country wide. Even in one country, the complete spectrum, of scenarios
is very difficult to establish. I was engaged in the safety field for light water reactors. At
that time, the PRA [Probablistic Risk Assessment] methodology became very popular and
familiar in the late 80's. And so, such a quantifying method is just now becoming popular in
the Japanese industry. The concern to adopt such a method will increase in the next few
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years, even in waste management. At present, only the safety people are concerned about
such a methodology, risk-based methodology. So Japanese utilities or the Japanese industry
will organize a very vast spectrum of work related to the risk-based waste management issue.
So I expect that in the near future the Japanese people will recognize the importance of
scenario setting, even to set scenarios. We may adopt some PRA-like approach, such as
event-tree, etc. I understand the importance of such a sophisticated way. But, we need
some time to prepare to adopt such a formal method.
Mr. MacKinney; Maybe just one or two more comments. It's just passed two o'clock.
Mr. Warren, briefly.
Mr. Warren: I'm Stephen Warren, [U.S.] Department of Energy. I would like to simplify
the previous comments, and use two examples from a scenario standpoint. We eat a lot of
beef, if you will, a lot of burgers, a lot of fried food. But, I think in Japan it's a little bit
different. As an example, for the dose to be higher in America; if you look at living
conditions -1 think houses and rooms are smaller, the distances between concrete walls are a
lot smaller in Japan than in the United States. So, the dose from concrete, or re-bar
[concrete reinforcing bar], if you will, in Japan will be a lot higher than in America. So,
standard scenarios on an international bases, I think, would be difficult, unless you have
multipliers to make up the difference.
Mr. MacKinney: OK, is there one final comment? Gene [Durman].
Mr. Durman: This is by way of my closing comments for this workshop. I think we all
have had many experiences that have been very good professionally. We've also had many
experiences that were very good personally. I've been very pleased, and I know that I speak
for my colleagues, that this has been a marvelous combination of the professional experiences,
and cultural and social experiences. I am very pleased that JAERI was such a kind host, and
we look forward to such an activity again in the near future. Thank you very much.
Mr. Katagiri: My name is Katagiri, Department of Health Physics , JAERI. On behalf of
the secretary group, I cordially express many thanks to all participants who made this second
EPA/JAERI Workshop a success. I believe, and expect, that all the discussion and
presentations will contribute to rulemaking and other technical work on radioactive wastes,
especially for residual radiation criteria. Thank you very much.
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Appendices
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Appendix 1
WORKSHOP PARTICIPANTS
U. S. Participants
Daniel D. Burns
Manager
Recycling Programs
Feraald Environmental Restoration Management Corporation
P. 0. Box 398704
Cincinnati, OH 45239-8704
Tel: (513)738-8439
Fax: (513) 738-7388
Shih-Yew Chen
Group Leader
Risk Assessment and Safety Evaluation, Einvironmental Assessment Division
Argonne National Laboratory
9700 South Cass Ave. EAD/900
Argonne, IL 60439-4832
Tel: (708) 252-7695
Fax:(708)252-4611
Scott Dam
Program Manager
Facilities Ownership, Management & Operations
BNFL Inc.
9302 Lee Highway Suite 950
Fairfax, VA 22031-1207
Tel: (703)385-7100
Fax: (703) 385-7128
Eugene C. Durman,
Deputy Director
Office of Radiation and Indoor Air
United States Environmental Protection Agency
401 M Street, S.W. (6601 J)
Washington, D.C, 20460
Tel: (202) 233-9340
Fax:(202)233-9651
David Fauver
Senior Project Manager
United States Nuclear Regulatory Commission
Washington, D.C. 20555
Tel: (301)415-6625
Fax: (301) 415-5397
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Cort N, Horton
Field Services Manager
Decontamination & Decommissioning Operations
Battelle Columbus Laboratory
505 King Avenue
Columbus, OH 43201-2693
Tel: (614) 424-7521
Fax: (614) 424-3954
Harris B.Hull
Geologist, Office of Radiation and Indoor Air
United States Environmental Protection Agency
401 M Street, S.W. (6603J)
Washington, D.C. 20460
Tel: (202) 233-9382
Fax: (202) 233-9650
William E. Kennedy, Jr.
Technical Group Leader
Environmental Health Physics Group, Health Physics Department
Pacific Northwest Laboratory
P. O. Box 999 (M/S K3-54)
Richland, WA99352
Tel: (509) 375-3849
Fax:(509)375-2019
Alan L. Liby
President
Manufacturing Sciences Corporation
804 Kerr Hollow Rd.
Oak Ridge, TN 37830
Tel: (615)481-0455
Fax:(615)481-3142
John A. MacKinney
Environmental Scientist
Radiation Studies Division
United States Environmental Protection Agency
401 M Street, S.W. (6603J)
Washington, D.C. 20460
Tel: (202) 233-9487
Fax: (202) 233-9650
Robert Meek
Section Leader
Environmental Policy Section, M/S T-9C24
United States Nuclear Regulatory Commission
Washington, D.C. 20555
Tel: (301) 415-6205
Fax:(301)415-5385
342-
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JAERI-Conf 95-015
Neil J. Numark
Associate
Sanford Cohen & Associates, Inc.
1608 Twentieth St. NW, Suite 400
Washington, D.C. 20009-1081
Tel: (202) 462-5860
Fax: (202) 462-5879
David Rast
Environmental Engineer
Fernald Area Office
United States Department of Energy
P. 0. Box 538705
Cincinnati, OH 45253-8705
Tel: (513)648-3138
Fax: (513) 648-3077
Ganesan Subbaraman
Technical Program Manager
Energy Technology Engineering Center
Rockwell International Corporation
P. O. Box 7930
Canoga Park, CA 91309-7930
Tel: (818) 586-5625
Fax:(818)586-5118
Stephen W. Warren
National Decommissioning and Recycle Program Coordinator
Office of Environmental Restoration (EM-43)
United States Department of Energy
Quince Orchard, 19901 Germantown Road,
Germantown, MD 20784
Tel: (301)427-1664
Fax:(301)427-1881 .
Japanese Participants
Seiji Abe
Deputy Division Manager
Third Research Division
Radioactive Waste Management Center
15th Mori bldg.,
Toranomon 2-8-10^ Minato-ku
Tokyo, 105
Tel: (03) 3504-1081
Fax: (03) 3504-1297
343-
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JAERI-Conf 95-015
Mamoru Adachi
Director
Department of Nuclear Ship Decommissioning
Mutsu Establishment, JAERI
Sekine, Mutsu-shi,
Aomori, 035
Tel: (0175) 23-4211, Ex. 4201
Fax:(0175)23-6194
Kazuyoshi Bingo
Director
Department of Health Physics
Tokai Research Establishment, JAERI
Tokai-mura, Naka-gun,
Ibaraki,319-ll
Tel: (029)282-5311
Fax: (029) 282-6063
Hideo Ezure
Director
Department of Research and Development
Research Association For Nuclear Facility Decommissioning
Funaishikawa 821-100, Tokai-mura, Naka-gun
Ibaraki,319-ll
Tel: (029)283-3010
Fax: (029) 287-0022
Mikio Fujii
General Manager
Health Physics Administration Division
Department of Health Physics
Tokai Research Establishment, JAERI
Tokai-mura, Naka-gun,
Ibaraki, 319-11
Tel: (029) 282-5895
Fax: (029) 282-6063
Kazuo Fujiki
General Manager
Decommissioning Technology Laboratory
Department of Decommissioning and Waste Management
Tokai Research Establishment, JAERI
Tokai-mura, Naka-gun,
Ibaraki, 319-11
Tel: (029) 282-5554
Fax: (029) 282-5998
- 344 -
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JAERI^Coof 95-015
Yoshiro Hattori
General Manager
Planning and Coordination Division
Department of Decommissioning and Waste Management
Tokai Research Establishment, JAERI
Tokai-mura, Naka-gun,
Ibaraki,319-ll
Tel: (029) 282-5691
Fax: (029) 282-5998
Hideo HBguchi
Director
Division of Research and Training
Japan Chemical Analysis Center
Sanno-cho 295-3
Chiba,281
Tel: (043) 423-5325
Fax: (043) 423-5326
Yoshihiro Ikeuchi
Deputy Manager of Research and Development Section
Division of Research and Training
Japan Chemical Analysis Center
Sanno-cho 295-3
Chiba, 281
Tel: (043) 423-5325
Fax: (043) 423-5326
Hiroshi Katagiri
General Manager
Radiation Control Division I
Department of Health Physics
Tokai Research Establishment, JAERI
Tokai-mura, Naka-gun,
Ibaraki,319-ll
Tel: (029) 282-5876
Fax: (029) 282-6063
Yutaka Kawakami
Deputy Director
Office of Safety and Control, JAERI
Fukoku-Seimei bldg.
Uchisaiwai-cho 2-2-2
Chiyoda-ku, Tokyo, 100
Tel: (03) 3592-2360
Fax: (03)3592-2199
-345-
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JAERl-Conf 95-015
Shigeru Kumazawa
General Manager
Radiation Dosimetry Division
Department of Health Physics
Tokai Research Establishment, JAERI
Tokai-mura, Naka-gun, Ibaraki, 319-11
Tel: (029) 282-5205
Fax: (029) 282-6063
Takeo Mimori
General Manager
Nuclear Fuel Facility Decommissioning Technology Division
Department of Decommissioning and Waste Management
Tokai Research Establishment, JAERI
Tokai-mura, Naka-gun, Ibaraki, 319-11
Tel: (029) 282-5425
Fax: (029) 282-6366
Kentaro Minami
Deputy Director
Department of Health Physics
Tokai Research Establishment, JAERI
Tokai-mura, Naka-gun, Ibaraki, 319-11
Tel: (029)282-5191
Fax: (029) 282-6063
Satoru Morishita
Deputy Leading Engineer
Third Research Division
Radioactive Waste Management Center
15th Mori bldg.
Toranompn 2-8-10, Minato-ku
Tokyo, 105
Tel: (03)3504-1081
Fax: (03) 3504-1297
Chikara Nakamura
Deputy General Manager
Radiation Control Division I
Department of Health Physics
Tokai Research Establishment, JAERI
Tokai-mura, Naka-gun
Ibaraki, 319-11
Tel: (029) 282-6324
Fax: (029) 282-6063
- 346-
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JAERI-Conf 95-015
Hisashi Nakamura
Research Engineer
Decommissioning Technology Laboratory
Department of Decommissioning and Waste Management
Tokai Research Establishment, JAERI
Tokai-mura, Naka-gun
Ibaraki, 319-11
Tel: (029) 282-6348
Fax (029)282-5998
Hiroshi Noguchi
Senior Scientist
Radiation Control Division I
Department of Health Physics
Tokai Research Establishment, JAERI
Tokai-mura, Naka-gun
Ibaraki, 319-11
Tel: (029) 282-5242
Fax:(029)282-6063
Daihachiro Sakurai
Consultant Engineer (Metal)
Nuclear Energy Center
Mitsubishi Material Corporation
Koishikawa-Daikoku bldg.
Koishikawa 1-3-25, Bunkyo-ku
Tokyo, 112
Tel: (03) 5800-9320
Fax: (03) 5800-9375
Kaneaki Sato
Manager of Research and Development Section
Division of Research and Training
Japan Chemical Analysis Center
Sanno-cho 295-3
Chiba,281
Tel: (043) 423-5325
Fax: (043) 423-5326
YoshiMro Seiki
General Manager
Reactor Decommissioning Technology Division
Department of Decommissioning and Waste Management
Tokai Research Establishment, JAERI
Tokai-mura, Naka-gun
Ibaraki, 319-11
Tel: (029) 282-5689
Fax: (029) 282-5998
- 347-
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JAERI-Conf 95^015
Kenji Shimooka
Deputy General Manager
Radioactive Waste Technology Division
Department of Decommissioning and Waste Management
Tokai Research Establishment, JAERI
Tokai-mura, Naka-gun
Ibaraki, 319-11
Tel: (029) 282-5581
Fax: (029) 282-5227
Shinichi Suga
General Manager
Department of Health Physics
Tokai Research Establishment, JAERI
Tokai-mura, Naka-gun
Ibaraki, 319-11
Tel: (029) 282-5023
Fax: (029) 282-6041
Masahiro Suzuki
Director
Department of Investigation & Planning
Research Association For Nuclear Facility Decommissioning
FunaisMkawa 821-100, Tokai-mura, Naka-gun
Ibaraki, 319-11
Tel: (029)283-3010
Fax: (029) 287-0022
Mitsuo Tachibana
Research Scientist
Reactor Decommissioning Technology Division
Department of Decommissioning and Waste Management
Tokai Research Establishment, JAERI
Tokai-mura, Naka-gun
Ibaraki, 319-11
Tel: (029) 282-6739
Fax: (029) 282-5998
Daiji Takeuchi
Deputy Director
Office of Radioactive Waste Regulation, Nuclear Safety Bureau
Science and Technology Agency
Kasumigaseki 2-2-1, Chiyoda-ku
Tokyo, 100
Tel: (03) 3581-5271 Ext, 869
Fax: (03)3581-0774
- 348 -
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JAERI-Conf 95-015
Tadaaki Uchikoshi
Deputy General Manager
Nuclear Fuel Facility Decommissioning Technology Division
Department of Decommissioning and Waste Management
Tokai Research Establishment, JAERI
Tokai-mura, Naka-gun
Ibaraki, 319-11
Tel: (029) 282-6084
Fax: (029) 282-6366
HGdeaki Yamamoto
Research Scientist
Radiation Control Division I
Department of Health Physics
Tokai Research Establishment, JAERI
Tokai-mura, Naka-gun
Ibaraki, 319-11
Tel: (029) 282-6324
Fax: (029) 282-6063
Satoshi Yanagihara
Principal Engineer
Decommissioning Technology Laboratory
Department of Decommissioning and Waste Management
Tokai Research Establishment, JAERI
Tokai-mura, Naka-gun
Ibaraki, 319-11
Tel: (029) 282-5682
Fax: (029) 282-5998
Mitsuo Yokota
Director
Department of Decommissioning and Waste Management
Tokai Research Establishment, JAERI
Tokai-mura, Naka-gun
Ibaraki, 319-11
Tel; (029) 282-5410
Fax:(029)282-5998
Michiro Yoshimori
Chief
Radioactive Wast Management Division
Department of Decommissioning and Waste Management
Tokai Research Establishment, JAERI
Tokai-mura, Naka-gun
Ibaraki, 319-11
Tel: (029) 282-5226
Fax: (029) 282-5899
- 349 -
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JAERI-Conf 95-015
Secretarial staff
T. Hattori
M- Ishii
H. KobayasM
T- Maejima
A. Mihara
K. Ogawa
N. Okuyama
J. Onodera
M. Osawa
Y. Sasa
H. Tachibana
S. Yokoyama
- 350 -
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JAERI-Conf 95-015
Appendix 2
WORKSHOP PROGRAM
TUESDAY, NOVEMBER 8, 1994
10:00 -12:00 Technical Tour - Japan Power Demonstration Reactor (JPDR)
Radioactive Wastes Storage Facility
12:00 -13:30 Lunch
13:30 -15:15 Technical Tour (continued) -- JT-60
WEDNESDAY, NOVEMBERS, 1994
9:30 - 9:45 Opening Address - K. Bingo, JAERI and E. C, Durman, U.S. EPA
9:45 - 10:45 Extent of the Problems related to Residual Radioactivity and Recycling
Criteria
Chairman: K. Fujiki, JAERI
- "Status of the Japan's Regulatory Policy on Radioactive Waste Management,
- Cleanup and Recycling Issues -",
D. Takeuchi, Science and Technology Agency, Japan
- "Potential Impacts of Pending Residual Radioactivity Rules",
D. D. Burns, Fernald Environmental Restoration Management
Corporation
10:45-11:00 Break
11:00 -12:00 Extent of the Problems related to Residual Radioactivity and Recycling
Criteria (continued)
- "Progress of JPDR Decommissioning Project"
M. Kiyota, S. Yanagihara, JAERI
- "The Decommissioning Program of JAERI's Reprocessing Test Facility",
T. Uchikoshi, T, Mimori, Y. Iwasaki, A. Ito, JAERI
12:00-13:45 Lunch
13:45 -14:15 Extent of the Problems related to Residual Radioactivity and Recycling
Criteria (continued)
Chairman: E. C. Durman, U.S. EPA
- "The Decommissioning Plan of the Nuclear Ship MUTSU",
M. Adachi, R. Matsuo, S. Fujikawa, T. Nomura, JAERI
14:15-15:45 Cleanup and Residual Radioactivity Criteria
- "Surface Radiological Free Release Program for the Battelle Columbus
Laboratory Decommissioning Project",
C. N, Morton, Battelle Columbus Operations
- 351 -
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JAERI-Conf 95-015
- "Cost-Benefit Analysis for U.S. NRC Proposed Radiological Criteria for
Decommissioning",
R. A. Meek, U.S. NRC
- "EPA's Technical Methodology for the Development of Cleanup Regulations
for Radioactively-Contaminated Soils and Buildings",
H. B. Hull, M. Doehnert, A. Wolbarst, U.S. EPA
J. J. Mauro, L Ralston, Sanford Cohen & Associates
15:45-16:15 Break
16:15-17:45 Cleanup and Residual Radioactivity Criteria (continued)
Chairman: H. Noguchi, JAERI
- "Radiological Surveys : Methods, Criteria, and Their Implementation",
G. Subbaraman, R. J. Tuttle, B. M. Oliver, Rockwell International
- "Development of Risk-Based Computer Models for Deriving Criteria on
Residual Radioactivity and Recycling",
S. Y. Chen, Argonne National Laboratory
- "Unrestricted Release of Contaminated Lands and the Dose to the General
Public",
H. Yamamoto, S. Kato, JAERI
19:00 - 20:30 Reception (at "Grand Hotel Takeda" in Katsuta city)
THURSDAY, NOVEMBER 10, 1994
9:30 - 11:00 Recycling and Criteria
Chairman: S. Yanagihara, JAERI
- "Evaluation of the Costs and Benefits of Recycling Radioactively
Contaminated Scrap Metal",
E. C. Durman, P. Tsirigotis, J. A. MacKinney, U.S. EPA
- "Technical Issues Relating to the Recycle of Contaminated Scrap Metal",
S. Warren, U.S. DOE
D. E. Clark, Westinghouse Hanford Co.
- "The Prospect for Recycle of Radioactive Scrap Metals to Products for
Restricted and Unrestricted Use",
A. L. Liby, Manufacturing Sciences Corporation
11:00-11:15 Break
11:15-12:15 Recycling and Criteria (continued)
- "Economic Aspects of Recycling U.S. Department of Energy Radioactive
Scrap Metal",
J. Harrop, N. J. Numark, Sanford Cohen & Associates
J. A. MacKinney, U.S. EPA
- "Summary of Industrial Impacts from Recycled Radioactive Scrap Metals."
J. C. Dehmel, J. Harrop, Sanford Cohen & Associates
J. A. MacKinney, U.S. EPA
- 352
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JAERI-Conf 95-015
12:15-13:45 Lunch
13:45-15:45 Recycling and Criteria (continued)
Chairman: Y. Kawakami, JAERI
- "A Methodology for Estimating Potential Doses and Risks from Recycling
U.S. Department of Energy Radioactive Scrap Metals",
J. A. MacKinney, U.S. EPA
- "Study on Safety Evaluation for Unrestricted Recycling Criteria of
Radioactive Waste from Dismantling Operation",
M, Yoshimori, M. Ohkoshi, M. Abe, JAERI
- "Radiological Control Criteria for Materials Considered for Recycle and
Reuse",
W. E. Kennedy, Jr., R. L. Hill, R. L. Aaberg, Pacific Northwest
Laboratory
A. Wallo, HI, U.S. DOE
- "Effects on Radiation Sensitive Instruments from Recycling of Contaminated
Metal",
H. Yaroamoto, S. Kato, JAERI
15:45-16:00 Break
16:00 -18:00 Recycling and Criteria (continued)
Chairman: W. E. Kennedy, Jr, Pacific Northwest Laboratory
- "Metal Recycling Technology and Related Issues in the United States,
A BNFL Perspective",
P. Bradbury, S. Dam, W. Starke, BNFL, Inc.
- "Melting Tests for Recycling of Radioactive Metal Wastes",
H. Nakamura, K. Kanazawa, K. Fujiki, JAERI
- "Investigation on Recycling of Radioactive Waste",
D, Sakurai, K. Takahashi, A. Umemura, K. Kimura, Mitsubishi Material
Corporation
S. Abe, M. Yamamoto, Radioactive Waste Management Center
- Video "Investigation on Recycling of Radioactive Waste"
presented by Mitsubishi Material Corporation
FRIDAY, NOVEMBER 11,1994
9:30 - 11:00 Compliance with Criteria
Chairman: D. N. Fauver, U.S. NRC
- "Radiological Surveys to Demonstrate Compliance with Decommissioning
Limits",
D. N. Fauver, U.S. NRC
353 -
-------
JAERI-Conf 95-0)5
- "The Japan Power Demonstration Reactor Decommissioning Program
- Decontamination and Radioactivity Measurement on Building Surfaces - ",
M. Tachibana, M, Hatakeyama, Y. Seiki, S. Yanagihara, JAERI
- "Measurement of Residual Radioactivity in the Facility Being
Decommissioned",
H. Ezure, S. Miyasaka, H. Kuroda, J. Komatsu,
Research Association For Nuclear Facility Decommissioning
11:00-11:15 Break
11:15 - 12:15 Compliance with Criteria (continued)
- "Radiochemical Analysis of Homogeneously Solidified Low Level Radioactive
Waste from Nuclear Power Plants" ,
K, Sato, Y. Ikeuchi, H. Higuchi, Japan Chemical Analysis Center
- "Classification of Solid Wastes as Non-radioactive Wastes",
M. Suzuki, H. Tomioka, K. Kamike, J. Komatsu,
Research Association For Nuclear Facility Decommissioning
12:15-13:15 Lunch
13:15 - 14:00 Free Discussion and Summary
Chairman: J. A. MacKinney, U.S. EPA
354-
-------
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PROCEEDINGS OF THE SECOND WORKSHOP ON RESIDUAL RADIOACTIVITY AND RECYCLING CRITERIA JOINTLY SPONSORED BY THE UNITED STATES ENVIRONMENTAL PROTECTION
AGENCY, THE OFFICE OF RADIATION AND INDOOR AIR, AND THI JAPAN ATOMIC ENERGY RESEARCH INSTITUTE November 9-11, 1994, Tokai, Japan
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