o-EPA
United States
Environmental Protection
Agency
Office of
Radiation Programs
Washington DC 20460
EPA 520/3-79-002
May 1979
Radiation
Low-Level Radioactive
Waste Management
Proceedings of
Health Physics Society
Twelfth Midyear
Topical Symposium
February 11 - 15, 1979
v
Williamsburg, Virginia
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LOW-LEVEL RADIOACTIVE
WASTE MANAGEMENT
PROCEEDINGS OF
HEALTH PHYSICS SOCIETY
TWELFTH MIDYEAR TOPICAL SYMPOSIUM
FEBRUARY 11-15, 1979
WILLIAMSBURG, VIRGINIA
CO-HOSTED BY
NORTH CAROLINA CHAPTER
VIRGINIA CHAPTER
EDITOR
JAMES E. WATSON, JR.
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FOREWARD
The management of the Nation's radioactive waste is a matter
vitally important to not only the availability of electrical power but
also other beneficial uses of radioisotopes for medical and industrial
uses. Though not previously considered as radioactive waste, the
by-products of many mining and milling operations for other than
uranium, e.g., phosphates and metals, contain naturally-occurring
radioactive materials which must be addressed in a total waste
management program. Often the focus of attention is on high-level
radioactive waste management related to nuclear power generation.
While obviously this type of radioactive waste merits attention, the
Health Physics Society has provided a valuable service in providing an
opportunity for scientific and technical discussions to focus in this
Mid-year Symposium on the broad concerns of "low-level radioactive
waste management." The topics discussed in these proceedings clearly
demonstrate that low- level radioactive waste materials should and can
be managed for the protection of public health. The Office of
Radiation Programs in the Environmental Protection Agency is pleased to
have the opportunity to sponsor the publication of these proceedings
for the benefit of not only health physicists but also for those
interested in assuring that the public health is protected and our
environment respected.
William A. Mills, Ph.D.
Acting Deputy Assistant Administrator
for Radiation Programs (ANR-458)
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HEALTH PHYSICS SOCIETY ORGANIZATION
FOR 1978 - 1979
OFFICERS
Carl M. Unruh, President
Melvin W. Carter, President-elect
Ralph H. Thomas, Treasurer
Genevieve S. Roessler, Secretary
Richard J. Burk, Jr., Executive Secretary
BOARD OF DIRECTORS
Carl M. Unruh, Chairperson
John A. Auxier
Herbert E. Book
Stewart C. Bushong
Melvin W. Carter
Donald L Collins
Harold V. Larson
Richard V. Osborne
Genevieve S. Roessler
Keith J. Schiager
Jacob Sedlet
Jack M. Selby
Ralph H. Thomas
Edward J. Vallario
Robert G. Wissink
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SYMPOSIUM COMMITTEES
m
General Chairpersons
William P. Kirk
Sylvester L Meyers
NC
VA
Finance Committee
*John W. Cure III
*Elizabeth P. Katsikis
VA
NC
Joint Symposium
Executive Committee
Gary J. Adler VA
Lee S. Anthony VA
Worth B. Bowman III NC
John W. Cure III VA
Jewel C. Finch NC
A. Keith Furr VA
Philip E. Hamrick NC
James A. Hancock, Jr. VA
Stanton F Hoegerman VA
Elizabeth P. Katsikis NC
William P. Kirk NC
Conrad Knight NC
Sylvester L. Meyers VA
Dan Strom VA
Richard D. Terry NC,VA
James E. Watson, Jr. NC
Finley C. Watts NC
Audiovisual Committee
*Worth Bowman III NC
*Philip E. Hamrick NC
Robert D. Cross NC
Graham M. Hairr NC
George J. Oliver NC
Blair F. Rehnberg NC
Exhibits Committee
*A. Keith Furr VA
'Richard D. Terry NC,VA
*Committee Chairperson/Co-Chairperson
Local Arrangements
*Gary J. Adler VA
*Stanton F. Hoegerman VA
D. William Morgan NC
Daniel D. Sprau NC
* Daniel J. Strom VA
Ms. Cookie Little VA
(ex officio)
Program Committee
*James E. Watson, Jr. NC
* Fin ley C. Watts NC
Harold W. Berk NC
Mary L. Birch VA
Emil T. Chanlett NC
Fearghus T. O'Fogludha NC
Lester Seiden NC
Billy H. Webster VA
Publications Committee
*Conrad M. Knight NC
Publicity Committee
*Lee S. Anthony VA
Robert Mogle VA
Barry Parks VA
Special Events Committee
* James A. Hancock, Jr. VA
Jewel C. Finch NC
Stanton F. Hoegerman VA
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ACKNOWLEDGEMENTS
Appreciation is extended to all authors and chairpersons for their
participation in this symposium. In general, papers are presented in these
proceedings exactly as submitted by the authors and no editing of content
was undertaken. In a few cases figures and/or tables of a particular
manuscript were grouped for photo-reduction, and format changes were made
in papers 8 and 21.
The symposium proceedings were published through the Office of Radiation
Programs, U.S. Environmental Protection Agency. This method of publication
made the proceedings available to attendees and other interested persons in
a timely manner. Coordination between EPA and the program committee in
the preparation of these proceedings was very capably managed by Dr. Stephen
T. Bard.
It is impossible to acknowledge all of the many persons who unselfish-
ly gave of their time to work on the preparation for this symposium. Many
hours, days and weeks were worked by those persons shown in the section
"Symposium Committees" and by other members of the North Carolina and
Virginia Chapters. Any credit for this symposium is shared by all members
of these Chapters.
A special acknowledgement must be made to my secretary, Ms Frances L.
Dancy. Ms Dancy's work covered the preparation of the abstract form, the
preliminary technical program, the book of abstracts and these proceedings.
During this time she also handled a multitude of correspondence. Her
efforts made it possible for the program committee to meet all established
schedules.
James E. Watson, Jr.
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TABLE OF CONTENTS
PREFACE i
HEALTH PHYSICS SOCIETY ORGANIZATION FOR 1978 - 1979 11
SYMPOSIUM COMMITTEES 111
ACKNOWLEDGEMENTS iv
TABLE OF CONTENTS v
OPENING SESSION
LOW-LEVEL RADIOACTIVE WASTE MANAGEMENT - RETROSPECT AND PROSPECT,
KEYNOTE ADDRESS, P/l
J. A. Lieberman 1
SESSION A - ORIGINS OF LOW-LEVEL RADIOACTIVE WASTE
CHAIRPERSON - R. J. Stouky 10
SOURCES, AMOUNTS, AND CHARACTERISTICS OF LOW-LEVEL RADIOACTIVE
SOLID WASTES, P/2
A. H. Kibbey and H. W. Godbee 11
A PERSPECTIVE ON THE RELATIVE HAZARD OF LOW-LEVEL RADIOACTIVE
WASTE DISPOSAL, P/3
W. P. Dornsife 16
LOW-LEVEL RADIOACTIVE WASTES - SOURCES, CHARACTERISTICS AND VOLUME
PROJECTIONS, P/4
W. F. Holcomb, R. L. Clark, and M. F. O'Connell *
A PROFILE OF INSTITUTIONAL RADIOACTIVE WASTES GENERATED IN 1977,
P/5
T.J. Beck, M.R. McCampbell, L.R. Cooley 27
HEALTH PHYSICS CONSIDERATIONS IN ACCELERATOR DECOMMISSIONING
AND DISPOSAL, P/6
R.L. Mundis, M.J. Kikta, G.J. Manner, J.H. Opelka,
J.M. Peterson, B. Siskind 38
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VI
1 >
LOW-LEVEL RADIOACTIVE WASTE FROM RARE METALS PROCESSING FACILITIES,
P/7
J. Eng, J. Feldman, and P.A. Giardina - 49
SESSION B - WASTE HANDLING AND TRANSPORTATION
CHAIRPERSON - W. C. McArthur 61
DOE'S LOW-LEVEL WASTE MANAGEMENT PROGRAM, INVITED, P/8
S. Meyers 62
LOW-LEVEL RADIOACTIVE WASTE PACKAGING FOR CONTAINMENT OR
TRANSPORTATION, P/9
R. L. Clark and W. F. Holcomb *
A NUCLEAR POWER PLANT APPROACH TO TEMPORARY ON-SITE STORAGE OF
RADIOACTIVE WASTE, P/10
M.R. Buring and E.E. Gutwein 67
RADIOACTIVE SLUDGE TREATMENT BY THE CENTRIFUGAL DEHYDRATION
METHOD, P/ll
S. L. Hwang and C. M. Tsal 74
REDUCTION IN WASTE HANDLING AND TRANSPORTATION, P/12
J. E. Stewart -— 82
SOLIDIFICATION OF LOW-LEVEL RADIOACTIVE LIQUID WASTE USING A
CEMENT-SILICATE PROCESS, P/13
R. W. Granlund and J. F. Hayes — 91
COLLECTION AND DISPOSAL OF LOW LEVEL WASTE AT AN EDUCATIONAL
INSTITUTION, P/14 **
D.L. Andrews, J.R. Gilchrist, and H.W. Berk 101
COLLECTION AND HANDLING OF RADIOACTIVE WASTES FROM A LARGE
UNIVERSITY, P/15
J.G. Shotts, D.L. Spate, and P.K. Lee 107
AN ALTERNATIVE TO STOP THE PROLIFERATION OF LOW-LEVEL TRU WASTE
CONTAINERS, P/16
J. B. Peterson - — - - lie
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vn
VOLUME REDUCTION OF LOW-LEVEL RADIOACTIVE WASTE WITH A HAMMERMILL,***
W. D. Gregory - 121
SESSION C - WASTE DISPOSAL OPERATIONS AND ALTERNATIVES
CHAIRPERSON - A.A. Moghlssi 125
REGULATORY CONSIDERATIONS FOR RADIOACTIVE WASTE DISPOSAL,
INVITED, P/17
D.R, Fuhrman, S.A. Black and J.P. Pasinosky — 126
OPERATIONAL EXPERIENCE AT CHEM-NUCLEAR'S BARNWELL FACILITY,
INVITED, P/18
D, G, Ebenhack 133
MANAGEMENT AND SURVEILLANCE OF A UNIVERSITY RADIOACTIVE WASTE
BURIAL SITE, P/19
P.K. Lee, J.G. Shotts and D.L. Spate — 141
A PRELIMINARY IMPACT ASSESSMENT OF INSTITUTIONAL RADIOACTIVE
WASTE DISPOSAL, P/20
R. Andersen; T.J. Beck, L.R. Cooley and M. McCampbell - 151
THE STATUS OF LOW-LEVEL RADIOACTIVE WASTE DISPOSAL-HOW TO
PLAN A DISASTER!, P/21
W. C. McArthur 159
SHALLOW LAND BURIAL—WHY OR WHY NOT?, P/22
W.T. Thompson, J.O. Ledbetter, and G.A. Rohlich 168
WASTE MANAGEMENT OF URANIUM MINING AND MILLING OPERATIONS, P/23
N.P. Kirner, A.A. Moghissi and P.A. Blackburn 174
MANAGEMENT OF LOW-LEVEL NATURAL RADIOACTIVITY WASTES OF PHOSPHATE
MINING AND PROCESSING, P/24
C.E. Roessler, Z.A. Smith, W.E. Bolch, and J.A.
Wethington, Jr. 182
ACTIVITY MEASUREMENTS AT A WASTE VOLUME REDUCTION FACILITY,
P/25
J. Richardson and D. A. Lee 196
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vili
A COMPARISON OF ALTERNATIVES FOR LOW-LEVEL RADIOACTIVE WASTE
DISPOSAL, P/26
P.O. Macbeth ~ 203
SESSION D - REGULATORY ASPECTS
CHAIRPERSON - C. R. Price 213
DISPOSAL CLASSIFICATION OF LOW-LEVEL RADIOACTIVE WASTE, INVITED,
P/27
J. A. Adam 214
A HISTORICAL REVIEW OF FEDERAL/STATE ROLES IN REGULATING
COMMERCIAL LOW LEVEL RADIOACTIVE WASTE BURIAL GROUNDS, P/28
G. W. Kerr 219
REGULATORY ASPECTS OF LOW LEVEL WASTE DISPOSAL, P/29
H. G. Shealy - 231
BURIAL OF SMALL QUANTITIES OF RADIONUCLIDES WITHOUT PRIOR NRC
APPROVAL, P/30
J. W. N. Hickey 234
WASTE MANAGEMENT PRACTICES IN DECOMMISSIONING NUCLEAR FACILITIES,
P/31
H. W. Dickson 238
DECOMMISSIONING STANDARDS - THE RADIOACTIVE WASTE IMPACT, P/32
J. L. Russell and W. N. Crofford 252
MODELING AND ENVIRONMENTAL ASSESSMENT OF LAND DISPOSAL METHODS
FOR LOW-LEVEL RADIOACTIVE WASTES, P/33
G.L. Meyer, S.T. Bard, C.Y. Hung and J. Neiheisel 261
THE USE OF A RISK LIMIT AS AN ENVIRONMENTAL SAFETY STANDARD FOR
RADIOACTIVE WASTE DISPOSAL SITES, P/34
A.E. Desrosiers and E. Njoku 270
RADON - AN ENVIRONMENTAL POLLUTANT?, P/35
W.A. Mills 280
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PROPOSED FEDERAL RADIATION GUIDANCE FOR RADIOACTIVE WASTES, P/36
J,E. Martin, H.J. Pettengill and S. Lichtman 284
SESSION E - ENVIRONMENTAL - PUBLIC HEALTH ASPECTS I
CHAIRPERSON - S. V. Kaye — 293
PREVIOUS MANAGEMENT PRACTICES FOR NATURALLY OCCURRING RADIONUCLIDE
WASTES: CURRENT RADIOLOGICAL STATUS, INVITED, P/37
W.A. Goldsmith, D.J. Crawford, F.F. Haywood, and R.W.
Leggett 294
RADIOLOGICAL IMPACT OF URANIUM TAILINGS AND ALTERNATIVES FOR
THEIR MANAGEMENT, P/38
M.H. Momeni, W.E. Kisieleski, S. Tyler, A. Zielen,
Y. Yuan and C.J. Roberts 307
DOSE MODELLING FOR RADON-PRODUCING ACTIVITIES IN HEAVILY POPULATED
AREAS, P/39
L. Bettenhausen and V. Burrows 329
RA-226 CONCENTRATIONS IN THE HYDROGRAPHIC BASINS NEAR URANIUM
MINING AND MILLING IN BRAZIL, P/40
A.S. Paschoa, G.B. Baptista, E.C. Montenegro, A.C.
Miranda, and G.M. Sigaud 337
EVALUATION OF THE ENVIRONMENTAL DOSE COMMITMENT DUE TO RADIUM-'
CONTAMINATED SOIL, P/41
J. Feldman, J. Eng and P.A. Giardina - 351
ASSESSMENT OF RADON PROGENY INHALATION EXPOSURE FROM LOW-LEVEL
WASTES OF PHOSPHATE MINING IN FLORIDA, P/42
D. R. Fisher and C. E. Roessler --- 356
RECOMMENDATIONS FOR REMEDIAL ACTION AND DECOMMISSIONING OF A
RADIOACTIVE WASTE BURIAL SITE, P/43
P.A. Giardina, J. Eng and J. Feldman 366
ACCEPTABLE RESIDUAL RADIOACTIVE CONTAMINATION LEVELS FOR SITES
OF DECOMMISSIONED NUCLEAR FACILITIES, P/44
W.E. Kennedy, Jr., R.B. McPherson and E.C. Watson 373
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DOSIMETRIC AND RISK/BENEFIT IMPLICATIONS OF AM-241 IN SMOKE
DETECTORS DISPOSED OF IN NORMAL WASTES, P/45
M. E. Wrenn and N. Cohen 385
BANQUET ADDRESS
MANAGING LOW-LEVEL RADIOACTIVE WASTES: BIOETHICAL CONCERNS, P/46
Margaret Maxey — 40°
SESSION F - ENVIRONMENTAL - PUBLIC HEALTH ASPECTS II
CHAIRPERSON - R. M. Fry — 420
A REVIEW OF ENVIRONMENTAL SURVEILLANCE DATA AROUND LOW-LEVEL
WASTE DISPOSAL AREAS AT OAK RIDGE NATIONAL LABORATORY, P/47
T. W. Oakes and K. E. Shank — 421
EVALUATION OF A DECOMMISSIONED RADWASTE POND, P/48
D. Paine, K.R. Price and P.M. Mitchell 442
RETENTION OF LOW-LEVEL RADIOACTIVE WASTE MATERIAL BY SOIL, P/49
E.H. Essington, E.B. Fowler and W.L. Polzer - 457
BIOLOGICAL INTRUSION - A LONG TERM PROBLEM, P/50
W.J. Smith and A.F. Gallegos - *
THE USE OF HANFORD WASTE WATER PONDS BY WATERFOWL, P/51
K.R. Price and R. E. Fitzner 471
CESIUM-137 IN COOTS (FULICA AMERICANA) ON HANFORD WASTE PONDS'
CONTRIBUTION TO POPULATION DOSE AND OFFSITE TRANSPORT ESTIMATES,
P/52
L.L. Cadwell, R.G. Schreckhise and R.E. Fitzner 485
RESPONSE TO A WIDESPREAD, UNAUTHORIZED DISPERSAL OF RADIOACTIVE
WASTE IN THE PUBLIC DOMAIN, P/53
F.A. Wenslawski and H.S. North, Jr. 492
HEALTH AND SAFETY IMPLICATION OF A WIDESPREAD, UNAUTHORIZED
DISPERSAL OF RADIOACTIVE WASTE IN THE PUBLIC DOMAIN, P/54
H.S. North, Jr. and F.A. Wenslawski 497
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AN ASSESSMENT OF THE ENVIRONMENTAL TRANSPORT OF RADIOIODINE IN
THE AIR-GRASS-COW-MILK PATHWAY USING REPORTED ENVIRONMENTAL
MONITORING DATA, P/55
J. C. Erb 504
USES AND MISUSES OF LOWER LIMITS OF DETECTION, P/56
W.E. Bolch and W. MacCready
LIST OF ATTENDEES 515
AUTHOR INDEX 535
* Withdrawn
** Accepted But Unable To Present In Person
*** Added
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KEYNOTE ADDRESS
Low-level Radioactive Waste Management - Retrospect and Prospect
J A Lieberman
Nuclear Safety Associates
There are probably two purposes of a keynote address. The first pre-
sumably is to set the stage or theme for the conference. However, after
looking at the program I think one can fairly argue that just the listing
of the papers and the authors does this almost automatically. The scope and
nature of this conference is already pretty well identified, and I am sure
we all can look forward to a very productive meeting. The second purpose
of a keynote address, at least the ones that I have watched, is more or less
a mechanism for "clearing the aisles," so to speak, and getting on with the
business of the conference so we can hear and discuss the papers. The aisles
seem reasonably clear. Nevertheless, I hope the remarks I have to make will
help serve both purposes in some small way.
Right at t_he outset I would state my conception of the keynote of this
meeting in fairly simple terms that I think are consistent with the "retrospect
and prospect" which is the title of my talk. It goes somewhat as follows.
In the field of management of low-level radioactive waste, we are the
beneficiaries of a record and a heritage, if you will, of which I think we
can be reasonably proud. This record and heritage are the result of work
by many dedicated and competent people who from the very beginning of this
industry responsibly recognized the public health and safety and environmental
impact potential of the waste associated with the industry. As a matter of
fact there are a number of these people who are in the audience this morning.
This does not mean that we can be allowed to rest on any self-acclaimed laurels.
We should continually direct our efforts in the future to doing the job of
managing these materials more effectively and more efficiently, but this
continuing effort must be rational as well as prudent. The rationality must
stem from a sound scientific and technical base and a realistic public cost-
benefit equation, that is, one that doesn't have zero risk or zero release
as one of its terms. The institutional control and regulatory framework
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should be established on a similar basis, and should also provide better
communication to the public, so that they can, if not be assured of, at least
be afforded an opportunity to understand the validity of the system. Having
stated this theme, two separate statements I ran across recently came to my
mind. The first was one that was made by Jim Coulter, Secretary to the Depart-
ment of Natural Resources for the State of Maryland. I think this statement
was made in connection with some problems the states were having in water
pollution control and, as I recall, the control of trihalomethanes. Jim said
something to the effect that he thought a distinct form of hell was being in
heaven, or close to it, and not knowing it. I think that might have some
pertinence to our low-level waste management situation. The other statement
that I ran across recently was in an article by Aaron Wildofsky, in the recent
issue of the American Scientist that some of you might have seen. He is now,
I believe, President of the Russell Sage Foundation. In the course of this
article he made the remark, that struck a responsive chord as far as I was
concerned, that "Chicken Little is alive and well in America."
There is also a little story that comes to mind that I think also has some
pertinence, and anyway I think it is sort of an interesting little story.
It concerns a Jewish grandmother. It seems that there was this grandmother
and her only daughter, who was either divorced or widowed but was alone, and
a young grandchild about four or five years old, who obviously was the apple
of his moi-her's and certainly his grandmother's eye. A lot of affection and
care was bestowed on this youngster. Now it turns out one day the grandmother
wanted to go to the beach and wanted to take the child along. Well, the
child's mother was quite reluctant—they were very careful about this youngster,
who was the only remaining line in the family—and said no, she didn't think
the grandmother ought to take the child, that it may not be safe, and he was
a very active youngster who might be too much to handle. But the grandmother
said it was a nice day and might be relaxing, and finally the mother relented
and grandma took the youngster to the beach. Well, the youngster was playing
at the water's edge and all of a sudden a great big wave came in, and lo and
behold, took the child out to sea. Well, you can imagine the weeping and
wailing of the grandmother and her very fervent supplications to God, to please
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please not take this youngster from her, and she prayed and she wailed and
sure enough, another wave came along and deposited the youngster at her feet
unharmed. Well, she quickly embraced the child, but then cast her eyes heaven-
ward and, in rather a stern voice and with a stern visage, she complained:
"He had a hat!"
Now back to a little reprospection—or perhaps even reminiscences. In
my view the roughly thirty-five-year history of the nuclear business is probably
categorized or broken down into three phases or milestones. The first one,
of course, began with the Einstein letter to Roosevelt that led to the Man-
hattan district, the weapons program, and subsequently to the establishment,
after World War II, of the AEG. The second phase was the passage of the
Atomic Energy Act of 1954 which, as we all know, was the National Policy basis
for the civilian commercial nuclear power industry. The third, in my view,
was the passage ten years ago of the National Environmental Policy Act. Some
people crudely refer to this as the Lawyers and Consultants Relief Act. It
has resulted, I think it's clear to all of us, in a much greater involvement
of the third branch of government, the judiciary, and, what is now increasingly
recognized by many as the fourth branch of government, (incidentally not called
for in the Constitution) the regulatory agencies, in nuclear related issues.
My first contact or exposure with low-level waste management activities
was in the old Atomic Energy Commission in the late 40's, just after the
first phase, and it was associated with what was then called the "particle
problem" at Hanford and Oak Ridge. At the time there was an organization
established called the Stack Gas Problem Working Group, a group ot outstanding
authorities in matters relating to the subject area, and out of the work of
that group came things like the high-efficiency partlculate filters, and the
publication Meteorology and Atomic Energy, a basis for assessment of atmospheric
transport and diffusion, and other developments that I think are still in
good stead.
With regard to solid low-level waste, one of my early exposures was to
the activities that went on at the Knolls Atomic Power Laboratory (KAPL)
and at Argonne. The activities I refer to had to do with the waste compaction
and incineration work that was going on there. The compaction work at KAPL,
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which involved shipment of the waste to Oak Ridge, I thought was quite effective
But I must say that the early experience on incineration, particularly at
Argonne, (I can say this because my colleague Walt Rodger was responsible for
making that incinerator work) led to my own general disenchantment with in-
cineration as a volume reduction technique. I can be convinced otherwise, how-
ever, and I expect there have oeen some later developments that might change
my conclusions from that early experience.
Shallow land burial, such as the operations at Hanford, Oak Ridge, Savannah
River, and Idaho, was carried out oy first-rate people, and in my view they
did a very good job. We have to remember that this was almost nineteen or
twenty years ago. I thinK it was in May of 1960 that the burial grounds at
Oak Ridge and Idaho were made available for commercial low-level waste, and it
was early in 1961 that the Atomic Energy Commission established the basis for
commercial shallow burial operation on federal or state-owned land. Of course,
we are all aware that between "62 and '67 there were five such commercial
establishments licensed at Beatty, Maxie Flats, West Valley, Hanford, and
Shettield, and the sixth, at Barnwell, was licensed in 1971. While much has
been made by some of releases or other so-called "problems" from some of these
operations, in my judgement, in terms of safety and environmental protection,
these activities rate pretty good marks.
Some of my early contacts with sea disposal were on the East coast, out
of what was then called the Naval Ammunition Depot at Earle, New Jersey and
ott the Farralone Islands on the West Coast. I recall taking a trip in an LSI,
which was used for transporting the waste from the NAD at Earle. It was in
the winter time and the weather was pretty bad but we took off early one
morning with the deck of the LSI loaded with 55-gallon drums and headed out
to the established, designated disposal area off about 100 miles the New Jersey
coast. It was a very straightforward operation. When we got to the place
where the skipper said his navigation chart showed him the disposal site was
located, the sailors proceeded to just roll the 55-gallon drums overooard.
They were categorized by source, i e originating institution, and there were
a few drums that came from an institution that shall remain unnamed that
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didn't sink, and that's how the sailors got their ritle practice that morning.
Again, I don't mean to imply that everything that was done in those days was
perfect, but nevertheless, as a result of many efforts by some thoroughly
competent people, the assessment and the investigative efforts related to
sea disposal, both nationally and internationally, I believe showed pretty
well that scientifically, technically, and environmentally sea disposal
for certain categories of waste was indeed acceptable.
Socially and politically, however, it was an entirely different story.
I recall with not much pleasure being a witness before the Joint Committee
on Atomic Energy with Senator Eastore the Chairman, and being interrogated—-
to put it mildly—on questions related to sea disposal. As a result of those
kinds of non-scientific or technical considerations which are certainly common-
place now, there was a moratorium placed on new licenses for sea disposal in
the early sixties, as I recall, and sea disposal was terminated in the United
States in the late sixties or early 1970. The future viability o± sea disposal,
at least in my mind, is questionable, again on socio-political grounds, although
I recognize that this technique is being practiced in other parts of the world.
With the shut-down of West Valley, Maxey Flats, and Shetfield, the volume
limitations in effect at Barnwell, and withdrawal of the license application
at Cimraaron, the current major problem, in my view, is the assurance of an
adequate disposal capacity, particularly east of the Rockies. The recent IRG
draft report to the President, which we are familiar with, estimates the total
acreage required tor Department of Energy low-level waste plus commercial
low-level wastes by the year 2000 ranges from about 450 to 1,650 acres depending
on an assumption of a volume reduction factor of 5 after 1985. The IRG
apparently did recognize the seriousness of the low-level waste problem, and
stated (and I quote) "That there presently exists neither a coordinated
national program for management of these low-level wastes or an institutional
mechanism to deal effectively with these issues." It then recommends that DOE
assume responsibility for developing and coordinating the needed national plant
for low-level waste management with active participation and advice from other
concerned federal agencies and input from the states, general public, and
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industry. I look forward to Sheldon Myers discussion of this subject this
afternoon. The IRG further recommends that states be provided the option to
retain management control of existing commercial low-level waste or to transfer
such control to the federal government. The draft report goes on to suggest
that future sites could be developed either Dy the individual states or by
the federal government, but such action should oe taken within the agreed-upon
framework of an overall low-level waste siting plan developed through a
joint Federal/State partnership.
This leads to the prospective part of my remarks.
Of course, there are a number or issues involved—technical, institutional,
and administrative. One of the more important technical issues in my judgement
is that of low-level waste classification. In my view the publication of
NUREG-0456, titled "A Classitication System tor Radioactive Waste Disposal -
What Goes Where" represents an important and valuable contribution to the
rationalization ot this subject. I believe we are to hear John Adam on this
later in the program. Nuclear Safety Associates, working with the Utility
Waste Management Group independently arrived at a similar approach and methodo-
logy and we believe it provides a sound basis for the development of comprehensive
regulations which are essential for consistent, proper management of low-level
wastes. While it is clear, at least to me, that the definition of waste
categories based on limiting concentration ot specific isotopes makes sense,
it is equally clear that such isotopic limits cannot De used in any practicable
way in the rield to individually classity each waste package. A program of
sampling and analysis ot representative wastes from particular sources as a
basis for generic classification of low-level wastes from particular sources
would appear to be in order. I might note that the Utility Waste Management
Group has taken some initial steps in this direction.
It should also De pointed out that the analyses ot NUREG-0456 and other
independent studies, including UWMG, have shown that the water migration pathway
is relatively unimportant in terms of shallow land burial ground pertormance.
Other factors that need to be carefully considered in connection with
their effectiveness and cost in terms of burial ground pertormance include
volume reduction and solidification. Considering the critical exposure pathways
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it is not immediately clear that the general need for such operations exists.
Systematic cost-benefit analyses are in order.
Another area ot possible technical consideration is that of alternative
disposal systems or methods. The recent NRC advance notice of proposed rule-
making in the low-level waste area begins with the statement that "recent
developments at the six commercially operated shallow land burial sites have
highlighted the need for an explicit regulatory program for shallow land disposal
of such wastes by alternative methods." While I am not sure what was meant
by "recent developments" I think it is important to make clear that they do not,
or should not, imply that any significant hazard to public health and safety
has resulted from the operation of shallow land burial grounds. I, of course,
fully agree with the need to upgrade the regulatory program for disposal of
low-level wastes. But that does not stem from any previous hazard to the
public, i e significant technical deficiencies. Instead, it arises because
the current regulatory tramework does not make sufficiently explicit the
information to be provided by an applicant proposing to undertake such an
operation, or the standards by which such application will De judged, and because
the public is not properly informed as to the bases for the regulatory programs
and thus lacks both a proper perception of the limited hazards and confidence
in the regulatory agencies and their programs. Accordingly, what is needed
is both an improved regulatory program and Better communication thereor to
the public. I am glad to see the NRC getting on with at least meeting the
first part of that need.
The statement in the advance notice appears to overemphasize the possible
need for alternative methods of disposal. I believe that the NRC studies
and the independent studies referred to reasonably demonstrate three major points,
a) Shallow land burial is an acceptable method ot disposing of radio-
nuclides up to determinable concentrations;
t>) The limiting concentrations for essentially all isotopes are determined
not so much oy the capability of a properly located , designed and
operated burial site to contain the isotopes once they are buried, but
by handling procedures upon receipt and burial of the waste and by
the possibility that intrusion into the waste by individuals may
occur at some later date; and
c) The limiting concentrations are higher than the concentrations in
wastes which have traditionally or historically oeen considered
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It follows that the need for alternative methods is marginal at best and
I strongly support NRC's stated position that "the regulations (10 CFR Part 61)
will initially cover only the currently practiced method of shallow land burial"
since this is clearly the primary need.
In this same vein, I have an additional general comment concerning
alternatives. The advance notice states, "The Commission believes that
development of those parts of its program dealing with regulation of alternative
methods of disposal of low-level wastes should be accelerated, since such
alternatives offer means of providing additional disposal capacity." Alter-
natives might indeed offer additional capacity, but so would expansion of
existing sites, reopening of closed sites, and opening of additonal shallow
land burial sites. I believe that any of these three options could be exer-
cised more practicably and more expeditiously than other alternatives.
I would urge that work on other alternatives not be permitted to slow the
schedule for putting into place the 10 CFR Part 61 regulations relating to
shallow land burial, which, incidentally, I understand will not be available
in draft form until April 1980.
In passing, I would also comment that the "stretching" consideration
of alternatives in NUREG-0308 might well be counter-productive. Even the
passing addressing of putting LLW into space and the atmospheric disposal
via a tethered balloon tends to imply a more difficult technical problem
than really exists and detracts from the credibility of the overall effort.
Other specific issues that have to be dealt with in the near future include
those of:
a) Federal-State relationships.
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This is, in my view, a critical area. While some useful ad hoc steps
have been taken recently to improve the working relationships I believe a more
specifically defined, and perhaps even a more formalized process for State
involvement must oe established in order to overcome this serious bottleneck
to orderly, effective low-level waste management. In any case, I believe the
States should have the option to retain control or existing commercial facilities
or transfer control to the Federal government, and that LLW disposal sites
oe licensed either directly Dy NRC or by the State through the Agreement
State process. (A State should not oe allowed to restrict use of a licensed
facility.)
b) DOE take-over of commercial waste burial grounds and opening ot
DOE facilities to commercial users.
This, in some respects, is connected to the Federal-State relationships
issue. There may well be some "sticky wickets" so to speak, for example,
the requirement for licensing of the DOE facilities, that would have to be
resolved before one might be able to arrive at conclusions on this issue.
Nevertheless, there are sufficient incentives to justity vigorous analysis
ot the pros and cons ot these possible actions. Perhaps Sheldon Myers will
also enlighten us in this area.
c) On-site management at reactor sites.
Because of the potential critical nature ot disposal capacity availability,
prudence would dictate careful examination of feasible actions in this are.
The National Environmental Studies Project ol AIF has sponsored an investigation
by NUS of possible on-site alternatives and a report should De out shortly.
There are, of course, a numoer ot additional pertinent issues or topics
that I know merit recognition but I will finish as I began.
We really haven't done all that uadly over the past uhree decades or more
in managing low-level radioactive wastes. Let's get on with doing the job
more efficiently and effectively but let's do it as I said—with prudence and
with some reasonable degree of rationality. Let's not ourselves be Chicken
Littles. Let's carry on with professionalism and integrity—after all, that's
what the Health Physics Society is all about.
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10
SESSION A
ORIGINS OF LOW-LEVEL RADIOACTIVE WASTE
Session Chairperson
R. J, Stouky
NUS Corporation
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11
SOURCES, AMOUNTS, AND CHARACTERISTICS
OF LOW-LEVEL RADIOACTIVE SOLID WASTES
A. H. Kibbey and H. W. Godbee
Chemical Technology Division
Oak Ridge National Laboratory
P. 0. Box X
Oak Ridge, Tennessee 37830
Abstract
Low-level radioactive solid wastes (LLW) are generated in the nuclear fuel
cycle, national defense programs, institutional (especially medical/biological)
applications, and other research and development activities. The estimated
total accumulation of defense LLW, 'vSO.S x 106 ft3 (VL.4 x 106 m3), is roughly
three times that estimated for commercial LLW, mill tailings excepted. All
nuclear fuel cycle steps generate some LLW, but power plants are the chief
source. From 1975 through 1977, reactor process stream cleanup generated
'VL x 106 ft3 (^2.8 x 101* m3) annually. Spent fuel storage (or reprocessing)
and facility decontamination and decommissioning will become important LLW
generators as the nuclear power industry matures.
The LLW contains dry contaminated trash, much of which is combustible and/
or compactible; discarded tools and equipment; wet filter sludges and ion-
exchange resins; disposable filter cartridges; and solidified or sorbed liquids,
including some organics. A distinguishing characteristic of LLW is a long-lived
alpha-emitting transuranic content of <10 nCi/g; this limit, however, is pres-
ently under review by NRC. If it is increased, the amount of LLW would also
increase. The nonfuel-cycle waste generation rate in 1975 was estimated to be
^7.6 x 105 ft3 (^2.1 x lO4 m3)/yr. The majority of these wastes, >6 x 105 ft3
(>1.7 x 101* m3), was medical and academic wastes which usually contained iso-
topes with induced activities of ^60-day half-life, neglecting 3H and lkC. The
remaining research and development LLW contained a broad spectrum of radioactive
species that was relatively small in total volume [probably M..5 x 105 ft3
0^4.2 x 103 m3)]. The amount of nonfuel-cycle waste that is generated annually
has been steadily increasing; thus, these estimates tend to be conservative.
The routine power-plant LLW contains varying amounts of activated corrosion
(e.g., 60Co, 59Fe, and 51tMn) and fission products (e.g., 13tt.137Cs and 90Sr);
the filter sludges and ion-exchange resins have the highest radiation levels
and normally require biological shielding and remote handling.
DISCUSSION
The problems that have been encountered at some shallow land-burial sites
for low-level radioactive solid wastes (LLW) have stimulated much interest in
finding more adequate ways of treating, handling, and disposing of these wastes.
There are several ongoing studies being made by various Governmental agencies,
among them the NRC, EPA, USGS, DOE, and some bodies at the state level. Nearly
all of these studies seem to point up the importance of characterizing all the
kinds of LLW with regard to source, volume generated, the type and amount of
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12
radioactivity contained, and the physical form of the waste and its container.
A knowledge of these parameters as they now exist is useful in future planning.
Low-level radioactive solid waste is usually described as having: (1) low
enough beta-gamma activity levels so that no special provision must be made for
heat removal, and (2) penetrating radiation levels such that only minimal or
no biological shielding or remote handling is necessary. In addition, it is
generally considered to contain <10 nCi/g of transuranic alpha-emitters. The
10-nCi/g value is currently under review by the NRC and may possibly be changed.
Obviously, an increase would instantly decrease the amounts of transuranic (TRU)
waste now in retrievable storage while simultaneously increasing the amount of
LLW acceptable for burial. Only those wastes containing <10 nCi/g transuranics
are addressed in this discussion.
An estimate of the accumulated volumes of LLW now buried in existing
burial grounds has recently been reported by the Interagency Review Group in
their Report to the President (IR78). The volume of buried LLW due to defense
operations is about three times that due to commercial operations [i.e., 50.8 x
106 vs 15.8 x 106 ft3 (1.4 x 106 vs 4.5 x 105 m3) ]. It should be pointed out,
however, that the collection time for the defense wastes is about double that
for the commercial wastes (i.e., approximately four decades vs approximately two
decades). The annual generation rate of defense LLW has leveled off at ^1 x 105
ft3 (2.8 x 103 m3).
The commercial wastes can be broken down into two categories, namely, fuel
cycle and nonfuel cycle. Some fuel-cycle LLW is generated in each step of the
cycle: mining, milling, conversion of U02 to UF6, enrichment, fuel fabrication,
and reactor operation. Spent fuel storage (or reprocessing if we should ever
choose that option) and facility decontamination and decommissioning will become
important LLW generators as the nuclear power industry matures. At present,
with mill tailings excluded from the LLW classification, the light water-cooled
reactor (LWR) power plants themselves are the largest generators of fuel-cycle
LLW, and the fuel fabrication plants are second.
It has been pointed out in a recent study at the University of Maryland
that most of the nonfuel-cycle LLW results from medical and academic (or insti-
tutional) applications (An78). The remaining nonfuel cycle wastes are gener-
ated in industrial or other research applications. The medical-type wastes
comprised the major fraction of the institutional wastes, with the academic
wastes representing about one-eighth of the total.
Of the total LLW shipped to commercial burial grounds, it is estimated
that, by volume, the fuel-cycle wastes represent ^60% and nonfuel-cycle wastes
make up the remaining 40%. The total volume of fuel-cycle LLW is ^1.2 x 106 ft3
(3.4 x 10^ m3), and ^80% of this waste, or M. x 106 ft3 (2.8 x 10^ m3) is
generated at LWR plants. Another estimated 2 x 105 ft3 (5.7 x 103 m3)'results
from fuel fabrication. Compared to these waste volumes, the other commercial
LLWs generated in the fuel cycle are almost negligible.
Based on 1975 data (An78), an estimated total annual volume of nonfuel-
cycle commercial LLW is in the order of 7.6 x 105 ft3 (2.2 x I0k m3). Of this
medical/biological-type wastes make up ^80%, or ^6.3 x 105 ft3 (1.8 x 104 m3) '
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13
Other academic-type wastes comprise ^3 to 4%, or ^2.7 x 104 ft3 (7.5 x 102 m3),
and industrial, research, and other miscellaneous wastes make up the remainder
[just under 15%, or VL.l x 105 ft3 (3.2 x 103 m3)].
Up to this point, the three generic classifications of LLW (defense, fuel
cycle, and nonfuel cycle) have been broadly described in terms of the estimated
annual volumes of each that might be expected. An understanding of the physica
chemical, and radiological characteristics of each waste type is also required
if improved waste management methods are to be developed.
All generators of LLW have some dry wastes that are compactible and/or
combustible (e.g., clothing, rags, paper, plastic, and wood). They also have
a relatively small fraction that is not compactible and/or combustible (e.g.,
contaminated equipment, tools, and glass). Most of the defense and fuel fabric
tlon plant LLWs that are buried fall in these categories. In the fuel cycle,
dry LLW probably represents between 30 and 40% of the total waste shipped to
burial sites; dry wastes are estimated to comprise between 40 and 50% of the
institutional LLW.
The wet LLWs generated at defense and fuel-cycle installations have many
similarities and frequently have been treated in similar ways. Most unit opera
tions commonly used in both defense and fuel-cycle process stream cleanup are
the same, namely, ion exchange, filtration, and evaporation. The spent ion-
exchange resins and filter sludges have, for the most part, been merely de-
watered by decantation, filtration, or centrifugation, placed in drums (or
shipping cask liners), and sent to the burial ground. Sometimes absorbent mate-
rials such as vermiculite are added to take up any free liquid that may remain.
The spent resins and filter sludges together probably account for 10 to 20% of
the fuel-cycle LLW sent to commercial burial grounds. Recently, more stringent
requirements have been imposed on nuclear power plants, the largest generators
of these types of waste within the fuel cycle. The current trend is toward
immobilization of resins and sludges by incorporating them in a solidification
agent such as cement or urea-formaldehyde resin. In the future, asphalt or
unsaturated polyester resins may be used as solidification agents. In any case
solidification of the resins and sludges will increase the LLW volume to be
shipped by a factor of ^1.2 to 2 (or greater), depending on the type of solidi-
fication agent.
The evaporator concentrates at both defense installations and LWRs contain
decontamination solutions and laboratory wastes. The main chemical differences
appear to be the rather high concentrations of nitrates in the defense wastes,
the borates that are the dominant constituent of pressurized-water reactor
(PWR) wastes, and the sodium sulfate that characterizes boiling-water reactor
(BWR) wastes. Evaporator concentrates are incorporated into a solidification
agent prior to burial. The defense LLWs are usually solidified with cement.
Up to the present, in the United States, either Portland cement or urea-
formaldehyde resin has been used for the LLW fuel-cycle liquids. Cement chemi-
cally binds the water into the solid matrix, whereas urea-formaldehyde resin
merely encapsulates it within the pores of the matrix material. Solidified
evaporator concentrates account for an estimated 40 to 50% of the total LLW
from the fuel cycle. Any fuel-cycle liquids that are solidified at the burial
site are not included in this estimate. Before continuing, it should be
mentioned that borates in high concentration may inhibit the setting of cement,
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14
whereas high concentrations of sodium sulfate can interfere with urea-
formaldehyde resin solidification.
The cartridge filters, predominantly used for stream cleanup at PWRs,
are noncompactible, noncombustible solids that are treated as wet wastes.
They are usually collected in a shielded shipping container, and after several
have accumulated, the solidification agent is added. This type of waste
probably represents <5 vol % of the total fuel-cycle waste shipped annually.
The cartridge filters used in nuclear power plants are fabricated from such
materials as cotton, synthetic fibers, polypropylene, matted paper, or porous
ceramic.
A summary of the estimated proportions of the various types of LLW
currently being shipped to burial grounds from LWRs each year shows that
solidified concentrates are the largest fraction (^40 to 50%), closely fol-
lowed by the dry wastes (^30 to 40%). The filters, filter sludges, and
spent resins together comprise <25% of the ^1 x 106-ft3 (2.8 x lO^-m3) total
waste volume shipped annually.
Historically, the BWRs have quite consistently generated more LLW per
unit of thermal power output than the PWRs. A recent ORNL study of solid
radwaste practices at nuclear power plants (Ki78) showed that the overall
average, through 1977, for PWRs is VL.l x 10~3 ft3 (3.1 x 10~5 m3) per MWh(t);
the average for BWRs is slightly more than double this value [i.e., ^2.3 x 10~3
ft3 (6.4 x 10~5 m3) per MWh(t)]. The number of operating PWRs is about double
the number of operating BWRs.
As mentioned earlier, the dry solids in institutional LLW represent an
estimated 40 to 50% of the total volume. Another 50% or so is comprised of
analytical laboratory wastes, about half of which is solidified and absorbed
liquids. Probably <10% is biological waste.
The industrial, research, and other nonfuel-cycle LLW, which is estimated
to be slightly <15% of the total volume [i.e., VL.l x 105 ft3 (3.2 x 103 m3)/yr],
is not so well defined. It is made up of a miscellaneous mixture of assorted
debris.
The radiological characteristics of the LLW vary in accordance with their
point of origin. The activity in defense and fuel fabrication LLW is mostly
due to uranium and its daughters although, in some cases, mixed fission
products may be present. For the most part, however, the distinguishing
characteristic of these wastes is the dominance of the naturally-occurring,
long-lived alpha emitters.
The nuclear power plant wastes are characterized by their beta-gamma
activity. Corrosion products circulating in the coolant are activated in the
reactor core. The minute traces of uranium remaining on the fuel cladding
after fabrication and/or failed fuel during reactor operation introduce mixed
fission products into the coolant stream. The activities of most concern in
these wastes are those which have half-lives of several years for example
Co-60, Cs-134 and -137, and Sr-90. The filter sludges and ion exchange resins
that arise from coolant cleanup operations are the most radioactive of these
wastes, and they require biological shielding when being handled. The solidi-
fied evaporator concentrates and dry wastes generated at nuclear power plants
are generally much lower in activity level.
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The radioactive species associated with medical/biological LLW usually
have relatively short half-lives, generally being <60 days if 3H and 14C are
neglected. The other academic, industrial, research, and other miscellaneous
LLWs contain a broad spectrum of radioactive species which are not easily
classified.
Over the last five years or so, the characterization of LLW has been re-
ceiving increased attention that now seems to be culminating in a new National
Low-Level Waste Management Program. Several institutions and agencies are now
developing data bases that should provide a substantial foundation for making
the necessary decisions we face in the near future regarding the treatment,
storage, transport, and disposal of low-level radioactive solid wastes.
REFERENCES
An78 Andersen,R. L.,Beck, T. J., Cooley, L. R., and Strauss, C. S., Institu-
tional Radioactive Waste, NUREG/CR-0028, University of Maryland at Baltimore,
March 1978.
IR78 Report to the President by the Interagency Review Group on Nuclear Waste
Management, TID-28817 (Draft), Washington, D.C., October 1978.
Ki78 Kibbey, A. H., Godbee, H. W., and Compere, E. L., A Review of Solid Radio-
active Waste Practices in Light-Water-Cooled Nuclear Reactor Power Plants,
NUREG/CR-0144 (ORNL/NUREG-43), October 1978.
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A PERSPECTIVE ON THE RELATIVE HAZARD OF LOW-LEVEL RADIOACTIVE WASTE DISPOSAL
William P. Dornsife
Bureau of Radiation Protection
Pennsylvania Department of Environmental Resources
Abstract
Low-level radioactive waste disposal is presently one of the major prob-
lems facing most of the industries and institutions which use and produce
radioactive materials. A more widespread but less publicized waste disposal
problem may, however, be facing almost all major industries. This problem is
the disposal of other hazardous wastes.
This paper will examine the problem of radioactive waste management and
the management of other hazardous wastes by comparing their relative toxicities.
The relative risks of the total problem will then be assessed by comparing
the generation rates of each. In addition, the radioactive wastes produced by
the nuclear fuel cycle will be compared with the potentially hazardous waste
produced by alternate energy sources, namely, the coal fuel cycle and the
manufacturing of equipment for the collection of solar energy.
These comparisons suggest that the consequences of the disposal of other
hazardous waste could result in risks to current and future generations which
are comparable with or which may exceed that due to radioactive waste disposal.
Other aspects of the problems inherent in the safe management of hazardous
waste compared to radioactive waste will also be explored.
Introduction
The safe disposal of radioactive waste has been called such things as the
problem of centuries, a dilemma, unprecedented, the problem without solution
and the monster. All these descriptions suggest that the problems involved
with radioactive waste disposal are unique and most difficult to solve. To
the contrary, there may indeed be a more difficult waste disposal task facing
many industries other than those which use and produce radioactive materials.
This task is the safe disposal of non-radioactive industrial hazardous waste.
What Is Hazardous Waste?
Almost all industries produce byproduct materials which have no commer-
cial value and therefore must be disposed of as waste products. What makes
this a very serious problem is the fact that about 10% of these industrial
wastes are considered to be a hazard to both public health and the environment.
These hazardous wastes include toxic heavy metals, primarily chromium, lead,
arsenic and cadmium, and also highly dangerous chemicals and pesticides.
Table 1 lists those industries which were the major contributors to the
hazardous waste disposal problem in 1977. The highly dangerous chemical and
pesticide wastes were produced primarily by the organic chemical industry,
while the waste from the other industries was considered hazardous mainly'
because of its toxic heavy metal content.
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The heavy metals and chemicals that make the waste hazardous can cause
adverse effects upon human health in a variety of ways which are much the same
as the adverse effects of the radioisotopes in radioactive waste. Some of
these toxic substances, if ingested in sufficient quantities, can cause imme-
diate or acute harm, much like the hazards of ingesting or being exposed to
high-level radioactive waste. Furthermore, many of these same toxic substances,
if ingested in lesser quantities, can cause long term chronic effects, such as
cancer and genetic damage, which is similar to the hazards of low-level radio-
active waste and uranium tailings.
Another measure of the hazard of a substance is its persistence in the
environment. All of the toxic heavy metals are stable elements and therefore
remain hazardous forever, assuming that a possible change in chemical form will
not affect their toxicity. Likewise, many of the highly dangerous chemicals
are also extremely stable and their decomposition to less harmful substances
may be uncertain over the long period of time that containment is necessary.
Radioactive waste, on the other hand, does obey the law of radioactive decay.
It will systematically decay to a relatively harmless material in a known period
of time. For low-level radioactive waste, this period is a few hundred years.
For high-level radioactive waste, the transuranic isotopes which are present
may extend this period to many thousands of years. Therefore, from the stand-
point of its persistence, much of the hazardous waste is comparable to the
transuranic isotopes in high-level radwaste or the long lived naturally occur-
ring isotopes in uranium tailings.
Toxicity of Hazardous Waste Versus Radioactive Waste
When comparing hazardous with radioactive waste, one of the most difficult
questions to be addressed is the identification of an appropriate basis for
comparing the relative toxicity of the two. At least one other study (Co77)
has compared the relative toxicity of naturally occurring toxic heavy metals to
high-level radioactive waste. That study concluded that the EPA safe drinking
water regulations (En76) for toxic heavy metals was equivalent to the 10 CFR 20,
Appendix B, Table 2, Col. 2 (MPC^) limits for radioactive isotopes.
Since limits for radioactive isotopes are also given in the EPA safe
drinking water regulations and since this is one of the few Federal regulations
which addresses both, it may be more appropriate and definitely much more con-
servative to use this standard exclusively for comparing radioactive and haz-
ardous waste. These regulations are also appropriate because they address the
problem of public drinking water contamination, which is the most likely route
of ingestion by man of any buried toxic material.
When looking in detail at the EPA drinking water regulations, it becomes
apparent that the limits for heavy metals and chemicals are based on criteria
which is at best vaguely defined. The limits are typically set at background
concentrations or at levels where no adverse effects are known to have occurred.
This is basically due to the fact that relatively little reliable scientific
data exists on the adverse effects and allowable levels for heavy metals and
chemicals in the environment.
On the other hand, as you are well aware, the effects of ionizing radia-
tion are probably the most studied and best understood of all environmental
insults. The allowable concentrations of radioactive isotopes, as given in
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18
the EPA drinking water regulations, are based on a maximum allowable whole
body or internal organ dose of 4 mrem per year for most isotopes. The excep-
tions to this are the bone seekers, such as radium-226, whose limits are based
on a cancer risk of approximately 1 x 10~6 per person per year. This risk is
about the same as that due to a whole body exposure of 4 mrem per year.
In order to establish a firmer basis, mainly for the toxic heavy metals
and chemicals, the National Academy of Sciences (NAS) recently completed a
study (Na77) which reviews in detail the current interim EPA drinking water
limits. The major conclusions of that study which are of interest here is that:
(1) the limits for at least arsenic and lead are probably too high to protect
the public adequately, (2) the limit for chromium should probably be based
only on the Cr+^ or chromate ion, and (3) the limits for radioisotopes are prob-
ably adequate to protect the public. This NAS study also confirms that for at
least one of the chlorinated hydrocarbons, which has a limit set by the EPA
drinking water regulations, the expected cancer risk from that limit is compar-
able with the cancer risk for the allowable levels of radioisotopes. Consider-
ation of these additional factors tends to confirm the assumption that the limits
prescribed by the EPA drinking water regulations provide equal protection from
toxic heavy metals, chemicals and radioisotopes. They are therefore probably
the best currently available guidelines with which to compare the relative
toxicity of hazardous and radioactive waste.
By using the limits as set by the EPA safe drinking water regulations and
by extending the concept of a radiotoxic hazard index to include toxic heavy
metals and chemicals, we can develop a relative toxicity index, expressed as nr
of water required to dilute a given quantity of toxic material to safe drinking
water levels. With this index we can make a direct comparison of the relative
potential hazard of radioactive versus hazardous waste. This comparison per
metric ton of the various types of waste is shown in Figure 1. It should be
noted that this is only a comparison of the potential for groundwater contamin-
ation, and as such does not include possible mitigating factors which may retard
or enhance its ultimate uptake by humans.
When examining Figure 1 in detail, a few important points should be men-
tioned. In order to provide a baseline for comparisons with background, the
average toxic heavy metal and naturally occurring radium concentration (Co77)
per metric ton of the earth's crust is shown in broken lines on the figure. The
radioactive waste potentials are taken from expected concentrations as given in
various NRC reports (Nu78a) (Nu76). The transuranic potential of spent fuel/HLW
was developed by comparing its cancer risk with that of radium. The increase in
the potential long term toxicity of low-level radwaste above the essentially
stable component primarily due to iodine-129 is caused by the ingrowth of the
daughter products of uranium-238 which is disposed of as source material. The
hazardous waste potential is a composite of various EPA reports (Ba76) (Ca75)
(Gr75) (Ja76) (Ve75) which have reported the toxic material concentration of the
waste products of those industries listed in Table 1. The toxic heavy metal
content accounts for the non-decaying portion of the hazardous waste curve. The
The radiotoxic hazard index is defined as a measure of the amount of water re-
quired to dilute a certain quantity of radioisotopes to permissible concentration
and is determined by dividing the initial quantity in curies by the permissible
r* r\n f* AT^ ^ ^*o +• 4 *%*\ A ** f* A /TnJ I \Ti * "74*. *V
concentration in Ci/m^ (Nu76).
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10
10
10
10
13
12
11
1 0
10s
10'
I
W
106
\ Hazardous Waste
U Tailings
Natural ToxicVHeavy Metal Background
Low Level Radwaste (LLW)
Natural Radium Background
102
I
10
Figure 1:
102 103 10* 105 106
TIME SINCE DISPOSAL, YEARS
Relative Toxicity of A Typical Metric Ton of Hazardous
Versus Radioactive Waste
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20
broken line decaying portion is an upper bound to account for the highly dan-
gerous chemicals which do not currently have safe drinking water limits. The
exceptions to this are a few chlorinated hydrocarbons, whose limits are used
as representative of all the highly dangerous chemicals.
Since Figure 1 only compares a typical metric ton of each type of waste,
the total annual production rate of each must also be considered to give a
truly indicative perspective as to the total potential for groundwater contam-
ination from radioactive and hazardous wastes. These annual production rates
for 1977 are shown in Table 1. Including this consideration, the total rela-
tive toxicity of hazardous versus radioactive waste is shown in Figure 2. This
comparison shows that based on current production rates the total long term
potential of hazardous waste is comparable with that of spent fuel and several
orders of magnitude higher than the long term potential of low-level radwaste.
Comparison of Wastes Produced by Alternate Energy Sources
It is well known by the public that most of the radioactive waste is pro-
duced by the nuclear fuel cycle. On the other hand, it is a little known fact
that most other alternate energy sources also produce large quantities of haz-
ardous waste. This can be shown, by looking at the two major short and long
term competitors with nuclear, namely coal and solar.
The coal fuel cycle can produce hazardous waste due to the fact that during
combustion most of the toxic heavy metals which are contained in coal in essen-
tially background concentrations are carried over into the fly ash and bottom
ash. Here they are concentrated by about a factor of ten and become much more
available for leaching due to chemical changes. In addition, large quantities
of sulfur dioxide scrubber sludges are generated by most modern coal fired
plants. This sludge is produced in a form that is not only difficult to dispose
of properly, but may be hazardous due to its sulfur content or from the carry-
over of some toxic heavy metals.
Solar energy facilities, and for that matter almost all renewable energy
sources, do not generate any waste products during operation. However, because
these energy sources are very dilute, much larger quantities of materials com-
pared to the more conventional energy sources are required for the manufacturing
of equipment which is necessary to collect this diffuse energy. With this
requirement for materials, mainly primary metals, comes the generation of large
quantities of hazardous waste. This waste is produced mostly from the smelting
and refining of the ores and the finishing of the metal surfaces.
Table 2 compares the total lifetime radioactive waste production of a
1000 MW nuclear plant with the lifetime hazardous waste production of a 1000 MW
coal plant and 1000 MW of solar power. The solar contribution is assumed to
be half flat plate collectors (heating and cooling) and half solar thermal
(electric). The potential relative toxicities of these wastes are then compared
graphically in Figure 3. This figure suggests that the wastes from coal and
solar power are of comparable hazard to the wastes from the nuclear fuel cycle
over the long term.
It should also be pointed out that the above indicated waste potential is
maximized for nuclear since the quantity of uranium tailings is the greatest for
the once through LWR fuel cycle. However, since the coal and solar potentials
-------
21
Table 1: Total Quantity of Hazardous and Radioactive Waste Produced in the
U.S. in 1977
Type of Waste
Hazardous (by type of industry)
Primary metals production
Organic chemicals and pesticides
Inorganic chemicals
Electroplating and metal finishing
Petroleum refining
Other industries
Total Hazardous Waste
Spent Reactor Fuel^2'
Low-Level Radwaste'^)
Uranium Mill Tailings
(4)
(3)Mu76
(4)Assumed to be 0.!
Quantity Produced
(Metric Tons/Year -
Dry Weight)
13.
9.
1.
7.
80
40
09
30
x 106
x 102
x 105
x 106
4.
3.
2.
1.
0.
1.
73
50
30
32
72
23
x
X
X
X
X
X
10
10
10
10
10
10
6
6
6
6
6
6
uranium ore (Nu76)
Table 2: Quantities of Lifetime Waste Generated by 1000 MW of Various
Alternate Energy Sources
Type of Waste and Energy Source
(2)
Nuclear
Spent Fuel'
Low-level radwaste^J'
Uranium tailings
Coal(*)
Fly ash and bottom ash
Scrubber sludge
Solar
Lifetime^ Generated Quantity
(Metric Tons - Dry Weight)
1.05 x 10°
5.67 x 10"
8.16 x 106
2.03 x 106
3.57 x 106
Hazardous waste'
of solar collectors1
the manufacturing
4.28 x
(l)Assumed to be 30 years in all cases
(2)Nu76
(3)Mu76
(4)Assuming Northern Appalachian coal which has been washed (Dv77)
(5)Hazardous waste production rate and constituents (Ba76) (Ca75)
(6)Material required for 1000 MW solar installation (In78)
-------
22
O
H
X!
O
H
W
Low-Level Radwaste
10s
TIME SINCE DISPOSAL, YEARS
Figure 2: Relative Toxicity of the Total Quantity of Hazardous
Versus Radioactive Waste Produced in 1977
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23
10
16
H
U
H
g
H
io15L
10
131
io12L
'10'
io10H
10E
Low-Level Radwaste
10
10-
10*
TIME SINCE DISPOSAL, YEARS
Figure 3:
Relative Toxicity of the Lifetime Radioactive Waste From A
1000 MW LWR Nuclear Fuel Cycle Versus the Lifetime Hazardous
Waste From A 1000 MW Coal Fuel Cycle and the Manufacturing
of Collectors for 1000 MW of Solar Energy
-------
24
only account for toxic heavy metals, the additional consideration of naturally
occurring radioactive isotopes could make the relative toxicity of their
wastes at least equal to the uranium tailings potential. This follows from
the fact that some coals may contain over 40 ppm uranium (Nu78a) and by assum-
ing all copper solar collectors and noting that some copper ores may contain
50 ppm uranium (Co78).
Other Factors Which Make the Hazardous Waste Problem Difficult to Solve
In 1976 Congress passed the Resource Conservation and Recovery Act (RCRA)
which mandates that EPA develop regulations which will insure proper handling
and disposal of hazardous wastes. These regulations, which have just recently
been published in draft form, (Er.78a) are about nine months behind the Congres-
sionally mandated deadlines, and are not scheduled to be finalized until Jan-
uary 1980. In the meantime, EPA has estimated (En77) that about 90% of the
total quantity of hazardous waste which is currently generated is being handled
and disposed of in a manner which may not be adequate to protect public health
and the environment. In fact, it has recently been reported by the EPA that
there are about 32,000 sites in the U.S. which contain potentially dangerous
amounts of hazardous waste, of which at least 638 may contain quantities which
could cause "significant imminent hazards" to public health (En78b).
There are many potentially troublesome problems inherent in the management
of hazardous waste which may make it a more difficult task than the safe manage-
ment of radioactive wastes. A few of these potential problems are the following.
Hazardous waste is much more difficult to account for and control at the
source. This is mainly due to the fact that hazardous waste generators will
not be regulated as stringently as are all radioactive waste generators. Also
complicating this is the fact that many industrial processes are proprietary
and therefore even information concerning the composition of the waste products
may be very difficult to obtain. This problem with accountability then leads to
inexpensive solutions such as illegal dumping in waterways and sewers.
The constituents of hazardous waste are much more difficult to detect and
measure accurately than are radioisotopes. Very sophisticated and time consuming
procedures are required to achieve low detection thresholds, and even then they
do not compare with the thresholds which are achievable for radioisotopes. In
addition, some toxic chemicals may change to other toxic compounds and therefore
an individual analysis must be initiated for each suspected contaminant. There
is no simple screening procedure such as a radioactive gross alpha or beta meas-
urement, which is very useful for quickly determining the effectiveness of radio-
active waste management.
Much of the industrial waste, particularly from primary metals production,
was considered nonhazardous by the EPA reports on hazardous waste production if
it did not leach toxic materials in appreciable quantities when mixed with dis-
tilled water. This was true even if the industrial waste contained large quanti-
ties of toxic materials. In addition, there were many other toxic constituents
of the industrial and hazardous waste, such as cyanide, phenol, oils and greases,
which were not considered in the comparisons of potential toxicities because they
currently are not included in the EPA drinking water regulations. By contrast,
the radioactive waste potentials include all the radioisotopes which are gener-
ated regardless of their chemical or physical forms.
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25
Like the situation with low-level radioactive waste disposal there cur-
rently is a very limited number of licensed hazardous waste disposal sites in
the U.S. This fact is proving to be very troublesome to hazardous waste gener-
ators since currently they produce about 200 times more waste by weight than
the generators of low-level radwaste. In addition, the publicity of recent
problems with some of the abandoned hazardous waste disposal sites may make
future licensing of these sites extremely difficult.
Even considering all of the above, probably the most disturbing inequal-
ity in the situation is that, in the opinion of many, the proposed EPA regula-
tions for hazardous waste disposal will be much less stringent than even those
which are being developed for low-level radioactive waste disposal. These
inconsistencies follow from a long standing tradition whenever radioactive
materials are involved.
As a prime example, the proposed hazardous waste regulations consider
solid waste to be hazardous only if the contained toxic materials can be leach-
ed from the wastes in concentrations exceeding 10 times the EPA safe drinking
water standards. The regulations go on to exempt those producers which gener-
ate less than 100 kilograms of waste per month from all the regulations except
those dealing with disposal in approved facilities. (The waste producers them-
selves are to determine if they exceed this threshold.) In contrast with this
are the recently proposed EPA criteria for all radioactive wastes (En78a) which
precludes the establishment of any'tie minimus" concentration of radioisotopes
for a waste to be deemed radioactive. Furthermore, a recent NRC proposed rule
change to 10 CFR 20 (Nu78b) would discontinue the practice of burial of small
quantities of licensed materials at sites other than licensed disposal facilities.
Causing the situation to be even more inconsistent is the fact that radium-
226, which has the lowest NRC permissible drinking water limit of any radio-
isotope, will be regulated under the proposed EPA rules as a hazardous waste.
Incredibly, the EPA is proposing as a'tie minimus"level for radium-226 (a concept
already disallowed for radioactive wastes), the same limit which the NRC has
deemed unacceptable for unlicensed disposal of this same radioisotope.
Summary
In the past radioactive waste disposal has been viewed by a majority of
the public in a complete vacuum because it was felt that the problems involved
were not comparable to any other environmental insult. However, when consider-
ing some of the problems which are inherent in the safe disposal of hazardous
waste, it must be concluded that radioactive waste disposal is not a unique
problem for this country to solve. In fact, when directly comparing the two
problems, radioactive waste seems to be the more manageable and therefore the
easier to implement successfully.
The responsible Federal agencies should also take note of the inconsisten-
cies which prevail in their regulatory efforts and begin treating equal hazards
equally. The intent here is not to ease the strict requirements that will be
necessary for safe radioactive waste disposal, but to determine if the standards
for hazardous waste disposal will be adequate in comparison with those strin-
gent standards. This will assure that public health and safety is being equally
protected from all toxic waste, regardless of the source.
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26
References
Ba76 Battelle-Columbus Laboratories, 1976, "Assessment of Industrial Hazardous
Waste Practices; Electroplating and Metal Finishing Industries," USEPA
Ca75 Calspan Corporation, 1975, "Assessment of Industrial Hazardous Waste
Practices in the Metal Smelting and Refining Industry," USEPA
Co77 Cohen, J.J. and Tonnes«en, R.A., 1977,"Survey of Naturally Occurring
Hazardous Materials in Deep Geological Formation: A Perspective on the
Relative Hazard in Deep Burial of Nuclear Waste," Report UCRL-52199,
Lawrence Livermore Laboratory.
Co78 Conference of Radiation Control Program Directors, Inc., 1978, "Natural
Radioactivity Contamination Problems," Report EPA-520/4-77-015.
Dv77 Dvorak, A.J., 1977, "The Environmental Effects of Using Coal for Genera-
ting Electricity." Report NUREG-0252, Argonne National Laboratory.
En76 U.S. Environmental Protection Agency, 1976, "National Interim Primary
Drinking Water Regulations," Report EPA-570/9-76-003.
En77 U.S. Environmental Protection Agency, 1977, "State Decision-Makers Guide
for Hazardous Waste Management," USEPA Report SW-612.
En78 U.S. Environmental Protection Agency, 1978, "Proposed Hazardous Waste
Regulations," Federal Register, Vol. 43, No. 243, December 18, 1978.
En78a U.S. Environmental Protection Agency, 1978, "Proposed Criteria for Radio-
active Wastes," Federal Register, Vol. 43, No. 221, November 15, 1978.
En78b Environmental Reporter, 1978, Vol. 9, Number 30, p. 1342, The Bureau of
National Affairs, Inc.
Gr75 Gruber, G.I., 1975, "Assessment of Industrial Waste Practices, Organic
Chemicals, Pesticides and Explosives Industries," USEPA Report SW-118c.
In78 Inhaber, H., 1978, "Risk of Energy Production," Report AECB 1119/Rev. 1,
Canadian Atomic Energy Control Board.
Ja76 Jacobs Engineering Co., 1976, "Assessment of Industrial Hazardous Waste
Practices, Petroleum Refining Industry," USEPA
Mu76 Mullarkey, T.B., Jenty, T.L., Connelly, J.M. and Kane, J.P., 1976, "A
Survey and Evaluation of Handling and Disposing of Solid Low-Level Nuclear
Fuel Cycle Wastes," Report AIF/NESP-008, Atomic Industrial Forum, Inc.
Na77 National Academy of Sciences, 1977, "Drinking Water and Health," Safe
Drinking Water Committee Report.
Nu76 U.S. Nuclear Regulatory Commission, 1976, "Environmental Survey of the
Reprocessing and Waste Management Portions of the LWR Fuel Cycle " Report
NUREG-0116.
Nu78 U.S. Nuclear Regulatory Commission, 1978, "Draft Generic Environmental
Impact Statement on Handling and Storage of Spent Light Water Power Reactor
Fuel," Report NUREG-0404.
Nu78a U.S. Nuclear Regulatory Commission, 1978, "A Classification System for
Radioactive Waste Disposal - What Waste Goes Where?" Report NUREG-0456
Nu78b U.S. Nuclear Regulatory Commission, 1978, "Proposed Standards for Pro-
tection Against Radiation," Federal Register, Vol. 43, No. 233, Dec. 4, 1978.
Ve75 Versar, Inc., 1975, "Assessment of Industrial Hazardous Waste Practices,
Inorganic Chemicals Industry," USEPA Report SW-104c.
-------
27
A PROFILE OF INSTITUTIONAL RADIOACTIVE WASTES
GENERATED IN 1977
Thomas 3. Beck, Margaret R. McCampbell, Leland R. Cooley; University of Maryland
at Baltimore.
Abstract
A national survey of radwaste volumes and characteristics generated by large
medical and academic institutions in 1977 was performed. This is a followup to a
survey which obtained 1975 data. The estimated total waste volume generated by the
survey population in 1977 was 7771 m3 (274,400 ft3). These data and those from the
previous survey show that the volume is increasingly linearly and consistently
accounts for approximately 11% of the total volume of low level radwaste buried
commercially. Most of the waste shipped by respondents (78%) was shipped to the
shallow land burial site at Barnwell, South Carolina.
Included in this profile of institutional radwastes are: a report of the principal
radionuclides present as waste contaminants, and a breakdown of waste by form; an
estimate of the effect of mechanical compaction of dry waste and a review of the
extent to which alternative disposal methods are used by the study population for the
various waste forms.
Performed under USNRC Contract #NRC 04-76-0344.
Introduction
Among the most significant sources of non fuel cycle low level radwastes, are
medical and academic licensees. In 1975 we showed that the larger medical and
academic institutions accounted for approximately one-third of the non fuel cycle
wastes buried that year in the commercial shallow land burial sites (An78).
We are at present concluding a followup to the 1975 survey, after having
obtained waste data from the same population for the calendar year 1977. As in the
preceding survey, the study population was selected from NRC and agreement state
licensee lists, to meet the following criteria:
* Large hospitals with -450 beds or more, excluding mental health and
other extended care facilities.
• Schools of medicine (hereafter referred to as medschools).
• Four year colleges and universities within excess of 5000 students
(hereafter referred to as colleges).
The data that we sought from these institutions was generally compiled by the
individual within the institution who managed the license. Therefore, we actually
surveyed radiation control programs.
-------
28
Radiation control programs are often, but not always, consolidated within the
various parts of an academic institution and also with neighboring institutions. To get
a better idea of the fraction of the study population represented by the data, we
divided the population into entities. We define a population entity as a college, a
medschool or a hospital. Some members of the study population may contain more
than one entity due to the consolidation of radiation control programs. Based on
responses, the population was categorized into the following groups of entities:
• Hospitals only
• Hospitals and medschools
• Hospitals, medschools and colleges
• Medschools only
• Medschools and colleges
• Colleges only
The numbers of entities within each category for the total population and for
respondents are shown in Table 1. The percent responses in the right most column
were used for data extrapolation. Overall the reponses accounted for 59% of the
hospitals, 66% of the medschools, and 56% of the colleges for which the totals are
348, 116, and 323, respectively. Geographically, we obtained data from 48 states and
the District of Columbia, only Nevada and Alaska were not represented.
Results
The estimated total waste volume shipped for burial by the study population in
1977 was 7771.1 m3 (274,400 fts). This constitutes approximately 11% of the total low
level waste volume shipped during the study year (Ho79). The institutional waste
fraction reported in 1975 was also 11% suggesting similar growth rates (An78). We
also obtained shipment volume data for 1976 in the current study; and have data for
the years 1972 to 1975 from the previous survey. The latter volume estimates have
been recomputed using the present population breakdown scheme, and differ
somewhat from that previously reported (An78). These six data points are plotted in
Figure 1. The waste volume appears to be increasing linearly: the correlation
coefficient for the least squares fitted line is .96. The equation of the line is:
V= m x + b
where:
V = volume in m
m= 639.74
X = (year - 1900)
b = - 41621.63
-------
29
Five commercial burial sites were in operation during the study year. The
breakdown of the data reported by respondents by destination is shown in Table 2.
The dominant burial site is again Barnwell, South Carolina, which received 78% of the
volume. This contrasts with 1975 when 30% of the volume was shipped to this site
(An78).
In 1977, respondents shipped a total of 372 Ci of non sealed source activity for
burial. Of this total, 99.1% was identified; the breakdown of the identified activity by
half-life category is shown in Figure 2. Most of the activity in the category with half •
life of 90 days or more is Tritium (91.5%). Of the remaining activity in this category,
5.7% is Carbon 1*. Again, it can be concluded that the significant nuclides
contaminating the waste of the study population are 3H and 14C. The major part of
the remaining activity shipped by this population in 1977 would be undetectable if one
assayed that waste today.
Sealed sources shipped for burial were treated separately. The total reported
sealed source activity is broken down by nuclide in Table 3. Again, the majority of
the activity shipped is Tritium, generally consisting of spent neutron generator
targets. Most of the remaining sealed source activity is 13 ts,192Ir and^Sla.
The types of waste shipped by the study population were categorized by waste
form:
• Dry, solids - syringes, vials, test tubes and other disposable labware,
absorbent papers, gloves, etc.
• Adsorbed liquids - aqueous and organic liquids dispersed in an adsorbent
media.
• Liquid scintillation vials - full scintillation vials packed in adsorbent
material.
• Biological wastes - predominantly carcasses of laboratory animals, also
including animal tissues, bedding, excreta and labeled culture media.
The breakdown of the wastes by type are shown in Figure 3. It is noteworthy
that the scintillation vial waste fraction has increased from 29% in 1975 to 43% in
1977 (An78X
Volume reduction techniques are frequently touted as an important adjunct to
waste processing in this population. In 1975, 23% (N = 29) of the respondents who
shipped waste compacted dry waste prior to shipment (An79)y in 1977 24% (N = 47)
compacted.
To obtain an estimate of the impact of the use of mechanical compactors for
reduction of dry waste volume, the dry waste fractions of those institutions
compacting waste was multiplied by the reported compaction ratio. A hypothetical
total volume was then obtained by summing the "expanded" dry waste with the other
waste fractions. The total volume thus obtained is 41% greater than that without
compaction. This is undoubtedly an optimistic assessment of the impact of waste
compaction, but could be considered to be an upper limit. It can be stated, however,
-------
30
that those compacting dry waste are the largest waste producers; therefore, a
relatively small number of compactors will have a major effect.
Of the 342 coded responses, 316 or 92% received unsealed sources of radioactive
material in 1977 and could be considered to be potential waste producers. Of these,
292 indicated that radioactive wastes of some type were diposed of in the study year.
Although we only quantitated wastes shipped for burial, data regarding the use of
alternative disposal methods was also obtained. Table 3 reports the numbers and
percentages of respondents disposing of the following waste types:
• Liquid scintillation vials and fluids
• Other (non scintillation fluid) organic liquids
• Aqueous liquids
• Biological wastes
Dry solids
The percentage breakdowns, showing the disposal methods utilized for each waste
type, are shown in Table 4. The alternatives shown are not exclusive, often more
than one disposal method was used for a given waste type.
The disposal of scintillation vials and fluid, generally contaminated with11* C and
*i, perhaps present the greatest disposal difficulties. Although shipment for burial is
the most prevalent alternative, the use of other methods is common. A relatively
common method is to dispose of scintillation fluids via the sanitary sewer and dispose
of the empty vials in the common refuse. Much of the dry wastes which appear in the
common refuse are apparently contaminated with short-lived nuclides; they are
thrown there after the nuclide is "decayed off.
The use of incineration as a disposal alternative is most common with biological
wastes, and is used to a lesser extent with the other waste types.
In the survey questionnaire we asked respondents to give the total cost, excluding
labor, for shipping waste in 1977. The total cost for the 197 respondents who shipped
waste in 1977 was approximately $1.4 million. To put this figure in perspective the
costs per cubic foot of wastes averaged $10.90 (±$13.20) and ranged from a maximum
of $125 per cubic foot to $0.38 per cubic foot shipped. Obviously, the shipment costs
vary widely depending on location of the institution, its shipment frequency, volume
and other factors.
This presentation is essentially a very broad stroke profile of the data in the 1977
survey. In the near future, we expect to examine closely the differences in waste
type among the various types of institutions within the population. We are currently
pursuing a waste stream approach which we hope will correlate the waste tvoes and
disposal alternatives to the type of institution and the specific uses of radioactive
materials.
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31
References
Andersen, R.L. Beck, T.J., Cooley, L.R., Strauss, C.S., 1978, Institutional Radioactive
Wastes (NUREG/CR-0028).
Andersen, R.L., Beck, T.3., Cooley, L.R., Strauss, C.S., 1979, Unpublished data from
1975 survey.
Holcomb, W.F., 1978, "Total Waste Volume Shipped for Burial - 1977, " Personal
Communication.
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32
TABLE 1.
BREAKDOWN OF SURVEY
POPULATION BY CATEGORY
HOSPITALS ALONE
HOSPITALS AND MEDSCHOOLS
HOSPITALS AND MEDSCHOOLS
AND COLLEGES
MEDSCHOOLS ALONE
MEDSCHOOLS AND COLLEGES
COLLEGES ALONE
NUMBERS OF ENTITIES
TOTAL RESPONDING
267
77
124
13
52
254
142
55
103
7
22
134
% RESPONSE
53.0%
71.4%
83.1%
53.8%
42.3%
52.8%
TABLE 2. BREAKDOWN BY DESTINATION
OF WASTE VOLUMES SHIPPED
DESTINATION
SHEFFIELD, IL
MAXIE FLATS, KY
BEATTY, NV
BARNWELL, SC
RICHLAND, WA
OTHER SITES
PERCENT OF
TOTAL VOLUME
8.42%
1.06%
6.66%
77.64%
5.50%
0.71%
-------
TABLE 3. SEALED SOURCES
SHIPPED FOR BURIAL
SEALED
SOURCE
3H
137Cs
I92lr
226Ra
Others
N
10
11
7
17
33
TOTAL
ACTIVITY
(Ci)
71.550
11.080
1.194
3.435
1.207
PERCENT
TOTAL
ACTIVITY
80.9%
12.5%
1.3%
3.9%
1.4%
TABLE 4. WASTE TYPES DISPOSED
BY STUDY POPULATION
WASTE TYPE
SCINTILLATION VIALS
AND FLUIDS
OTHER ORGANIC
(NON SCINTILLATION) FLUIDS
AQUEOUS LIQUIDS
BIOLOGICAL WASTES
DRY SOLIDS
^DISPOSING*
71%
48%
78%
59%
*Percentof those who disposed of radwaste in 1977
(total N=292).
-------
TABLE 5. DISPOSAL ALTERNATIVES*
WASTE TYPE
SCINTILLATION VIALS
AND FLUIDS
OTHER ORGANIC
LIQUIDS
AQUEOUS LIQUIDS
BIOLOGICAL WASTES
DRY SOLID WASTES
PERCENT OF RESPONDENTS USING DISPOSAL METHOD**
SEWER COMMON INCINERATION SHIP FOR BURY ON TRANSFER TO
REFUSE BURIAL SITE OTHER INST.
29.0%
46.0%
68.9%
7.0%
-
23.7%
-
-
4.1%
27.5%
41 °/
.3%
2.9%
-
25.7%
12.0%
80.7%
74.8%
54.4%
71.9%
72.1%
7.7%
7.9%
5.3%
9.4%
7.2%
1.0%
1.4%
1.8%
2.3%
6.4%
* Alternatives are not exclusive; many institutions disposed of a waste
type by more than one method.
** Percent of those who disposed of that type of waste.
-------
9-
8-
7-
m3 X 1000 6-
5-
4-
3-
72
1
74
1
76
1—
78
?
80
FIGURE 1. INSTITUTIONAL RADIOACTIVE WASTES SHIPPED FOR BURIAL 1972 - 1977
CO
in
-------
< 7 DAYS 6-X-&&S21 7.8 %
X ' fc^ *"t I %-* ••••••••••••« ^ "
7-90 DAYS
\\\\\\\\\\\\\\\
25.7%
> 9O DAYS
66.5%
H
I I Others
FIGURE 2. NON SEALED SOURCE ACTIVITY SHIPPED FOR BURIAL
BROKEN DOWN BY HALF-LIFE CATEGORY
OJ
-------
43.4%
10.1%
BIOLOGICAL
WASTES
11.6%
ADSORBED
LIQUIDS
34.9%
DRY SOLID
WASTES
LIQUID
SCINTILLATION
VIALS
FIGURE 3. BREAKDOWN OF WASTES SHIPPED FOR BURIAL BY FORM
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38
HEALTH PHYSICS CONSIDERATIONS IN ACCELERATOR DECOMMISSIONING AND DISPOSAL
R. L. Mundis, M. J. Kikta, G. J. Marmer, J. H. Opelka,
J. M. Peterson, B. Siskind
Argonne National Laboratory
Argonne, Illinois 60439
ABSTRACT
There are perhaps as many as 1,200 particle accelerators in the United
States, ranging in size from the very small Cockroft-Walton and electron
linear accelerators to the multi-GeV research synchrotrons. When an
accelerator has reached the end of its useful life the radioactivity induced
in the components then presents several disposal problems. At least fifty
accelerators produce significant induced activation, and several hundred more
are capable of producing fluxes of neutrons which could result in activation
of various components of the accelerator facility. This is generally of low
level except for the very high energy/high intensity machines.
In most cases, the induced radioactivity is confined to relatively few
parts of the machine and the disposition of these is through the normal radio-
active waste channels without complication. However, some machines leave a
legacy of low level induced radioactivity in massive components. Examples
of massive items are large magnets, shield blocks and beam stops of concrete,
earth or iron, and even the walls and floors of the building itself. These
need to be dealt with in a manner so as to pose no potential health hazards
to persons in the vicinity of the public at large.
The disposition of radioactive waste from major past decommissionings,
including the Cambridge Electron Accelerator (CEA) and the Brookhaven
Cosmotron, will be discussed. The extent of the induced radioactivity
problems in decommissioning of smaller accelerators will also be discussed.
INTRODUCTION
Over the past several years, there has been increasing concern over the
accumulation of radioactive materials at various scientific, industrial and
other facilities in the United States, and increasing pressure to ascertain
what is being done to assure that.any potentially serious waste disposal
problems are not being overlooked. Of course, the nuclear power industry has
the major share of the problem and, therefore, commands most of the attention.
However, there exists, in addition to the power industry, a sizable variety
of users of radioactive materials in medical products, industrial products, as
sources for radiation processing and sterilization and, as by-products, from
the operations of particle accelerators. The subject of nuclear facility de-
commissioning has recently been addressed by the Comptroller General in a
June 2, 1977, report to the Congress entitled "Cleaning Up the Remains of
Nuclear Facilities - a Multi-billion Dollar Problem" (Co77) . The primary
thrust of the report is toward the nuclear power industry; however, other
aspects of the problem, which include isotope usage and accelerator facilities,
are recognized as potential problems.
The Department of Energy has initiated a comprehensive study of the
quantities and types of radioactive materials in existence both at its
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39
existing facilities and at the facilities formerly utilized as part of the
Manhattan Engineer District/Atomic Energy Commission (MED/AEC) program.
The Division of Environmental Impact Studies at Argonne National Laboratory
was requested to perform a comprehensive study of the portion of the de-
commissioning problem that concerns the dismantling and disposal of all types
of particle accelerators in the United States.
The questions that this study relates to are contained in the report to
the Congress and are as follows (Co77):
"What is the extent of the decommissioning problem for accelerators?
"Are standards needed for induced radiation?
"What should be the limits on acceptable radiation levels?"
Only the first of these questions is directly pursued in the review effort.
In order to develop information relative to accelerator decommissioning,
the following tasks have been initiated. Since some of these tasks are not
yet complete, only progress can be reported in some areas. Also, the aspects
of the study that do not relate to the waste problem will not be discussed.
The tasks are:
1. Compile a census of the accelerator population in the United States
and categorize the machines according to their potential inventories of active
material.
2. Review the history of past accelerator decommissionings in regard to
technological, environmental, health and economic aspects,
3. Survey the quantity of radioactivated material at existing particle
accelerator facilities with emphasis on the high energy machines (e.g., ACS,
Bevatron, FNAL, LAMPF, SLAG, ZGS, etc.).
Lower energy machines have not been ignored, since there is a concern
about the neutron production by medical electron accelerators.
4. Develop a generalized decommissioning scenario for the various
categories of accelerators. Included in this effort will be such things as
estimate of costs, determination of the volume of radioactive waste to be
expected, and the estimation of the degree of reusability of various classes
of components.
5. Study the alternative methods of decommissioning as they relate to
accelerators. The alternatives are:
a) complete dismantling and removal of the accelerator and
its building to a waste burial site, temporary storage,
or immediate recycle.
b) removal of the accelerator only to a waste burial site,
temporary storage or immediate recycle, leaving the
building for research or office space.
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40
c) mothballing in place with eventual performance of a)
or b) above.
d) entombment in place.
Mothballing and entombment may, in fact, be necessary in cases where
radiation levels around accelerator components are comparable with the
nuclear reactor environment.
PRODUCTION OF INDUCED RADIOACTIVITY
Induced activity is due to both the interactions of the primary accelera-
ted beam with a target and to interactions of any secondary particles produced
by the primary beam interactions. The governing factors for material in the
vicinity of the target, generally are related to the neutron production
capabilities of the accelerated beam. The pertinent factors are as follows:
1. Species of particle(s) accelerated. Generally deuterons and tritons
generate more secondary neutrons than protons, while electrons generate much
fewer, other factors being equal.
2. Energy of the accelerated particles. The induced activity is a
function of the particle energy and the appropriate reaction cross sections
once the threshold energy is exceeded.
3. Beam intensity or current. Induced activity is proportional to the
number of particles accelerated.
4. Duty factor. The ratio of operating time to shutdown time determines
the actual value of the effective long-term average beam intensity.
5. Primary usage of the accelerator. This has a direct impact on the
other parameters, e.g., a high energy research accelerator or an isotope
production facility is generally run at maximum attainable currents and as
continuously as operational and fiscal constraints allow. At the other
extreme, some research accelerators are used exclusively for short-term
intermittent sample irradiations or for low intensity scattering experiments.
Table 1, found in the National Bureau of Standards Handbook 107 (NBS70),
summarizes the potential for induced radioactivity expected for various
accelerated particles and energies.
Until the last few years, medical and industrial accelerators (almost
exclusively electron accelerators) have been of energies less than
10 MeV. However, the use of medical linacs in the energy region above 10 MeV
is now on the increase. The rate of growth of medical linac installations is
presently in the range of 200 to 250 units per year with an estimated eventual
population of approximately 2000 units in the United States (Ro78) Heavy
ion and neutron therapy are both receiving increased attention from medical
researchers and may eventually add significantly to the neutron producing
accelerator population. The number of compact neutron generators used as
analytical tools now exceeds 200 in the United States. Neutron generators
were specifically not part of this review because they are generally of very
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41
small size with very low levels of activation.
PAST DECOMMISSIONINGS
A number of accelerators of all types have already been decommissioned.
Some of the earliest cyclotrons and betatrons were simply disassembled and
the components reused for other purposes or sold as scrap metal. The beam
energy and intensity of the early machines were generally very low so that
any induced radioactivity would have been essentially undetectable except by
very sensitive survey techniques. There are essentially no records regarding
these very early decommissionings.
The major decommissionings which have been reviewed in some detail are
the BNL Cosmotron, the Cambridge Electron Accelerator at Harvard, the 142
inch Carnegie-Mellon synchrocyclotron, and the University of Rochester
synchrocyclotron. The Brookhaven Cosmotron, a 3-GeV proton synchrotron, was
shut down December 31, 1966. The machine was kept in standby condition for 1
year after shutdown during which time the experimental area was dismantled and
much of the equipment was transferred to the Alternating Gradient Synchrotron
(ACS) facility at BNL. After the year had elapsed authorization to proceed
with the dismantling of the Cosmotron itself was granted by the AEC, the
owner of the machine. The removal of reusable equipment and components was
performed on a spare-time basis by BNL personnel. The actual disassembly of
the synchrotron ring magnets was done by contract technician labor over a 3
or 4 - month period. The one-year waiting period resulted in a significant
reduction of the induced activity levels. The iron magnet segments, copper
windings, vacuum chambers and vacuum pumps were placed in the radioactive
material storage yard where most of them remain today. The presence of
induced radioactivity in these items precluded their release to scrap dealers.
A number of the magnet blocks have been used as shielding at the ZGS and more
recently by the Fermi National Accelerator Laboratory. Because of the
difficulty in removing the epoxy resin and fibre glass insulation bonded to
the copper magnet windings, very few of these have been reused.
The Cambridge Electron Accelerator operated for the last time on May 31,
1973. The initial decommissioning plan, based upon the conditions of the
contract between Harvard and the AEC, included complete removal of the
accelerator along with all underground structures and other buildings and re-
storation of the land to its original status. A modified plan was agreed
upon and only the underground accelerator tunnel was scheduled for demolition
and removal. The accelerator itself was dismantled and the magnets sent to
BNL for possible reuse in an ACS experiment. They are still in storage at
BNL. The rest of the experimental equipment was either absorbed by other pro-
grams at Harvard or by other laboratories and universities. The demolition
of the underground tunnel was more difficult than expected because of the
heavy reinforcing and was not completely accomplished. Since the shock waves
from the wrecking ball were endangering the nearby private residences,
segments of 6 foot thick reinforced concrete retaining walls were left buried
in place. The site presently has the appearance of a vacant lot. Radio-
active waste volumes and radiation intensities were minimal from this effort
since this was an electron accelerator with low average beam current. A
radiation survey made within 24 hours after the final shutdown indicated only
one area with radiation intensity as high as 100 mR/h at contact (Sh73).
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42
The Carnegie-Mellon Synchrocyclotron decommissioning is an example of a
case in which it was desired to save the building, belonging to the
University, and offer it for sale on the open market. The complete accelera-
tor and all associated equipment were disassembled and removed from the
building, and the building itself decontaminated to levels which did not
exceed two times natural background. In order to remove the magnet iron
(- 310 tons) from the accelerator vault, it was cut up in place with torches
using appropriate contamination control measures. Approximately half of the
iron was transferred to MIT and the other half to the Los Alamos Meson
Facility to be used for shielding. The coils are in storage at BNL. After
removal of all the accelerator hardware, it was found that in certain areas
of the concrete walls and floor, there were detectable levels of induced
radioactivity which necessitated the removal of a layer of concrete several
inches thick. This effort generated several truck loads of concrete and
several hundred barrels of dirt and rubble which required transportation to a
commercial low-level waste burial ground.
RADIOACTIVITY LEVELS AT ACCELERATOR DECOMMISSIONING
In discussing the types and quantities of radioactive materials that are
generaged by an accelerator D&D effort, this report will focus on accelera-
tors with beam energies of tens of MeV and higher. There are a limited number
of components in any given accelerator that will become highly radioactive.
These will be portions of the primary beam transport systems, target stations
and beam stops which are directly struck by the accelerated beam as part of
normal operations. For the very high energy and high intensity accelerators,
those components and structures in the vicinity of points of primary beam
interaction will also be highly activated by secondary particles. In addition
to these localized "hot spots", the main structure of the accelerator,
primarily the magnet iron along with its copper or aluminum windings for
circular machines, and copper drift tubes and tanks for linear accelerators
will all contain a highly variable volume distribution of activation products.
It can generally be assumed that all the permanent components within the
primary shield enclosure or vault will have some degree of induced radio-
activity. Additionally, the walls of the shield vault itself may contain
significant quantities of induced radioactivity. Other components can
contribute to the radioactive waste volume. For example, cooling water
systems may accumulate and concentrate radioactivated corrosion products.
Vacuum systems, ventilation systems, and target transport systems also could
present additional radioactively contaminated material.
The detailed distribution of isotopes produced in accelerator materials
is a complicated function of the type and energy of the incident particle(s),
beam intensity, beam transport efficiency, target elements, and the cross-
sections for the various reactions involved. To illustrate the situation,
Figure 1 shows production cross-section curves for a Bi target bombarded with
protons of three different energies (Pa73). For machines of particle energy
in the 10's MeV, the isotopes produced are clustered very near the mass number
of the target material. For machines with higher energies (100's of MeV)
spallation and high energy fission processes result in a much broader spectrum
of product isotopes. The curve for 500-MeV particles shows two humps in the
production curve. The hump at the higher mass numbers is due to spallation
reactions and the hump at approximately one half of the target mass number
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43
results from high energy fission. In the GeV region, the production of all
isotopes becomes almost equally probable, with all types of production re-
actions becoming energetically possible. Similar distribution curves would
be obtained with any other heavy element. In this report, the term "target
material" is used in the broad sense meaning any material with which the
primary beam interacts.
Most of the major components of particle accelerators consist of either
iron or copper with minor amounts of other materials. Major exceptions to
this are the use of depleted uranium and lead for certain shielding and
collimation applications, and the use of aluminum for magnet windings.
Activation products in iron and copper are primarily short-lived with half-
lives of less than a few days. Table 2 summarizes the important long-lived
isotopes found in activated accelerator components. It is seen that 60Co,
22Na and 54Mn will be the controlling isotopes.
The longer an accelerator is operated, the closer will be the approach
to saturation activity for the long-lived products. The total quantity of
the long-lived activities present at shutdown depends on the gross long-term
average operating conditions of the accelerator. Short-lived activity is due
only to the operations during that operating period just prior to shut off.
Qualitatively, there is an initial rapid decay of the short-lived
components in the mix followed by a slower decay governed by the long-lived
isotopes. Some generalizations can be made in regard to the shape of the
decay curve. For positive ion accelerators with beam energies less than
approximately 100 MeV, the initial decay is very rapid and essentially lasts
for about a week after shutdown. The long-lived decay tail is then controlled
by the decay of the 59Fe and 65Zn with half-lives of 45 and 245 days re-
spectively. In higher energy accelerators, i.e., above 100 MeV, the slow de-
cay component is controlled by such additional isotopes as 51*Mn, 57Co and
60Co, with half-lives ranging from 270 days to 5.26 years. 60Co is the longest
lived gamma emitter and is generally found in the iron and copper, which make
up the bulk of the mass of most large particle accelerators. A few accelera-
tors have used aluminum windings for the main magnet coil and 22Na, with a
2.6 year half-life, may be the controlling long-lived activity.
A generalized decay curve of accelerator-induced radioactivity can be
derived using an analytical expression developed by Sullivan and Overton
(Su65) for high energy accelerators.
D(t) = a G In
where D(t) - the dose rate
a - machine dependent parameters
G - cross section and other physical constants
T - length of irradiation time (age of the accelerator)
t - length of decay time after shutdown
Figure 2 shows the relative value of D(t)/aG for two values of T as a
function of decay time t. The upper curve is for T = 25 years and the lower
is for T = 100 days.
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44
Two empirical decay curves are also plotted for comparison. These curves
were obtained as follows. Samples of reagent grade copper and iron were
exposed to the particle flux near a high beam loss area of the ANL 12.5-GeV
synchrotron for a three-month period. The decay of the gamma activity of the
samples was then monitored using a large Nal gamma counter for a period of
four years. The theoretical 100-day irradiation curve is seen to be in
reasonable agreement with the decay curves observed for the copper and iron.
Data obtained from the LBL 184 inch synchrocyclotron on the decay of radia-
tion levels during an 11-day undisturbed shutdown in 1971 falls between the
curves for Cu and Fe (Ri71). It appears, then, that it would be reasonable
for planning purposes to use a decay curve derived using the actual age of
an accelerator in the Sullivan and Overton formula. It can be seen from the
upper curve, for an assumed 25-year old accelerator, that ~30% of the radio-
activity will remain two years after shutdown. From this point on, the de-
cay could be assumed to be due primarily to the 60Co in the material.
Based on this model, a block of magnet iron from a high energy accelerator
reading 100 mR/h at 1 day after shutdown will read 30 mR/h after two years
of decay. Then, using the 5.26-year half-life of 60Co, it will take 61 years
to reach a radiation intensity of 0.02 mR/h, which is on the order of two
times natural background. It can, therefore, be argued that 100 years would
be a reasonable upper limit on the lifetime of accelerator-induced activity.
ESTIMATION OF QUANTITY OF RADIOACTIVE MATERIAL
Proton Accelerators
Estimates can be made of the total quantity of radioactivity contained
in a proton accelerator by using approximations discussed in a paper by
Gollon (Go76).
The method for estimating the total radioactivity in an accelerator is
based on the fact that an equilibrium for constant conditions of operation,
the decay rate of radioactivity is equal to the production rate. The pro-
duction rate is related to the accelerated beam intensity and energy. As a
first approximation, for accelerators of energy on the order of a few
hundred MeV, the saturation activity is numerically equal to the beam
intensity. Using the basic relationships of
1 ya = 6.025 x 1012 protons/sec and
1 Ci = 3.7 x 1010 dis/sec,
we can calculate a value of 160 Ci/yA. This radioactivity is distributed
among the various machine components and the experimental apparatus which
intercepts the beam. For example, the fraction of the beam that results in
the activation of a cyclotron magnet is probably in the range of 1% to 10%.
Therefore, we could expect to see approximately 1.6 to 16 Ci of saturation'
activity from operations with a 1-yA beam of protons. For machines of lower
energies, this relationship will overestimate the long-lived activity present.
V
For higher energy accelerators, the phenomenon of spallation, or star
production, in which a multiplicity of secondary particles are produced, can
result in activation products. For example, a beam of 1012 400-GeV protons/
sec is estimated to produce approximately 10 kCi of long-lived activity
(Go76), which yields a ratio of about 1600 Ci/yA of beam.
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45
Electron Accelerator
The estimation of radioactivity induced by high energy electron beams
and the associated bremstrahlung has been reviewed in a paper by Swanson
(Sw79). The photons produced by electron beams produce radioisotopes by
photonuclear reactions, the quasi-deuteron effect and photo spallation at
high energies. The components which absorb most of the electron and photon
beams are subject to the highest activation levels.
Table 3 is a summary of the long-lived isotope production data calcula-
ted for electron beams of energy 35 MeV and greater. The percent of satura-
tion attained for each isotope after an assumed 25-year operating life has
been calculated and is shown along with the expected production in terms of
curies per kilowatt of beam power. The long half-life isotope 26A1 is
included to illustrate the fact that products with very long half-lives will
generally not be a problem.
SUMMARY
Table 4 partially summarizes the results of this study to date on the
estimation of the eventual quantity of radioactive material that will re-
quire proper disposal. Emphasis is upon the largest accelerators with the
specific goal of arriving at the total masses and volumes of radioactive
waste that they represent. Most of the accelerator material involved is
iron or copper. The shielding included in these totals is primarily steel or
concrete. While the total quantities involved are considerable, it must be
remembered that all accelerators will not be decommissioned and dismantled
as waste at one time.
In keeping with the philosophy of maintaining all radiation exposure to
levels which are As Low As Reasonably Achievable (ALARA), it can be argued
that any radioactivity which would add to natural background should not be
released to the world at large for unrestricted use. It can also be argued
that dwindling natural resources, including metals such as copper and iron,
need to be reutilized to the maximum extent possible for both ecologic and
economic reasons. Consequently, if the recycling of these materials is
contemplated, all potential exposure pathways need to be considered, including
the internal exposure possibilities (e.g. from melting, grinding, cutting,
or welding operations) as well as the external exposures potential from
usage of the material. In addition, there are a number of other potential
detrimental effects aside from health effects which would need careful
consideration. Two examples of these are the possible effects on the photo-
graphic film industry and the disruption of analytical techniques used in
geology, archeology, medicine, crime detection, etc.
It seems that it would be possible to mothball in place or temporarily
store the accelerator components long enough to allow the induced activity
to decay to levels which are essentially indistinguishable from natural
background by sensitive detectors such as gas proportional counters or viR
meters. Based on the results of this study, it is probable that this period
of time would not be longer than 100 years for the major components of the
present high intensity machines. Millions of tons of material would be a
definite burden to bury and perhaps be a treasure worth eventual recycle.
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46
TABLE 1
RELATIVE PRODUCTION OF INDUCED RADIOACTIVITY IN PARTICLE ACCELERATORS
Accelerated Particle
Electrons
Protons,
Helium ions
Deuterons,
Tritons
All ions of light
atomic weight
Energy Range
(MeV)
<1.67
1.67 - 10
>10
<1
1-10
Any Energy
>10
Induced
Target
Radioactivity In
Vicinity
None None
Limited Very slight
Probable Suspect
Limited
Limited
Limited
Certain
None
Suspect
Suspect
Suspect
TABLE 2
RADIOACTIVE ISOTOPES OF CONCERN IN ACCELERATOR DECOMMISSIONING
Isotope
22Na
57Co
60Co
65Zn
Half Life
2.6 a
303 d
270 d
5.26 a
245 d
Produced In
Aluminum
Iron
Iron
Stainless steel
& copper
Copper
TABLE 3
INDUCED RADIOACTIVITY FROM ELECTRON BEAMS
Target
Material
Concrete
Aluminum
M
Iron
Nickle
it
Copper
ii
Isotope
Produced
22Na
22Na
26A1
5ltMn
57Co
6uCo
60Co
63N±
Half
Life
2.6 a
2.6 a
.4xl05 a
303 d
270 d
5.26 a
5.26 a
92 a
Saturation*
Activity,
Ci/kW
0.1
0.28
8.8
0.59
5.9
0.1
0.65
0.45
% of Saturation
for 25 Year
Operation
99.9
99.9
0.0023
100.0
100.0
96.2
96.2
17.0
Actual
Expected
Activity,
Ci/kW
0.10
0.28
2.04x10""
0.59
5.9
0,096
0.63
0.077
*Data from (Sw79)
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47
TABLE 4
ESTIMATED MASS & VOLUME OF ACCELERATOR RADIOACTIVE WASTE
Type of Accelerator
Cyclotrons <20" diam.
& "~ linacs
Cyclotrons
20" to 50" diam.
Cyclotrons
50 to 100" diam.
e~ linacs >_ 600'
Cyclotrons diam.
> 100"
Proton synchrotrons
60' - 200* diam.
Proton synchrotrons
> 200' diam.
proton linac
> 500' length
Approximate
It of Machines
1000
15
16
4
3
3
Estimated Mass
of Active Waste,
tons per
1 30
30 - 50
250 800
800 - 2300
2300 16,500
16,500 220,000
Estimated
Volume ,
ft3/accelerator
£ 60
260 2200
2200 - 7000
7000 - 20,000
20,000 - 140,000
140,000 - 2 * 106
REFERENCES
Co77 - Comptroller General's Report to the Congress "Cleaning Up the
Remains of Nuclear Facilities - A Multi-billion Dollar Problem" EMD-77-46
June 16, 1977
G076 - P. J. Gollon, "Production of Radioactivity by Particle
Accelerators", IEEE Transactions on Nuclear Science, Vol. NS-23, No. 4
August 1976
NBS70 - National Bureau of Standards Handbook 107, "Radiological Safety
in the Design and Operation of a Particle Accelerator", June 1970
Pa73 - H. W. Patterson and R. H. Thomas, Accelerator Health Physics
(New York: Academic Press), 1973
Ri71 - A. Rindi and L. D. Stephens, 184-inch Cyclotron Residual
Activity Decay Measurements, Internal Report, January 1971
Ro78 - Dr. David Rpmer, Siemens Corp., private communication
Sh73 - W. A. Shurcliff, "Survey of Highest Readings of Radioactivity
in CEA Synchrotron", Internal Report, June 1, 1973
Su65 - A. H. Sullivan and T. R. Overton, "Time Variation of the Dose
Rate for Radioactivity Induced in High Energy Particle Accelerators",
Health Physics 11: 1101, 1965
Sw79 - W. P- Swanson, "Radiological Safety Aspects of the Operation
of Electron Linear Accelerators, Ch.2.6., IAEA, Vienna (1979 to be published)
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48
1000 C
100
Fig. 1. Mass Yield Curves for Protons on Bismuth
(Pa73)
1.0
1.5
2.0
Fig.
Decay Time (yrs)
2. Decay of Accelerator Induced Radioactivity
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49
LOW-LEVEL RADIOACTIVE WASTE
FROM RARE METALS PROCESSING FACILITIES
Jeanette Eng, New Jersey Department of Environmental Protection
Donald W. Hendricks, ORP-Las Vegas Facility, U.S. EPA
Joyce Feldman, Radiation Branch, U.S. EPA Region II
Paul A. Giardina, Radiation Branch, U.S. EPA Region II
Abstract
In the past year, problems of disposal of byproducts, tailings, and
wastes from rare metals and thorium producing facilities have received the
attention of radiological agencies. Many of the raw ores used by these pro-
cessing facilities were rich in natural radioactivity and the residues of
production were often not disposed of properly. Mill tailings from the
uranium mining industry have only recently come under federal regulation.
It can be expected that similar attention will be focused on the environmental
impacts of the rare metals processing industry as illustrated by interest in
the problems of Parkersburg, WV, and Albany, OR. One other site in Akron, NY,
does not appear to be an immediate problem, but its situation is typical of
the many yet unsurveyed inactive rare metal facilities in the country. The
radiological problems presented by and tydecontamination activities which may
be required of these rare metal facilities are examined.
Introduction
The federal government has recognized that companies which process thorium
and uranium ores require regulatory controls in order to protect man and the
environment from unnecessary radiation. The recent passage in November, 1978
of the Uranium Mill Tailings Bill (H.R. 13650) demonstrates the government's
recognition that the front-end of the uranium fuel cycle, i.e., mining and
milling of uranium, had been'neglected. The bill defines procedures for a re-
medial action program at inactive mill sites and regulations for active mill
sites.
Companies which provide titanium, phosphorus, rare earths, and rare metals
for industrial and chemical use are not normally regarded as possessors of
large quantities of radioactive materials. In fact there appears to be a his-
torical laxity in documenting the processing and waste disposal activities of
these industries. A recent EPA publication reviews the available literature
on technologically enhanced natural radiation due to mineral extraction in-
dustries (Bl 78). It is only recently that phosphate industrial wastes have
been listed as hazardous radioactive wastes in the U.S. Environmental Protec-
tion Agency's proposed Hazardous Waste Guidelines and Regulations (Co 78).
This paper will review the situations at the existing Teledyne Wah Chang, Co.,
Inc. located at Albany, Oregon, and the former Carborundum Corp./Amax Specialty
Metals, Inc., facilities located at Parkersburg, West Virginia, and Akron, New
York, in order to show the extent of the radioactivity problem at rare metals
processing facilities and the need to identify for radiological review other
rare metal and rare earth processing sites.
As shown in Figure 1, the unusual grouping of rare earth and rare metal
processing industries stems from their common ore origin. The ores used in
rare earth and metals processing are byproducts of mining for titanium ores,
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50
since the ores for the specific processing are seldom found in economically
mineable rock. The principal domestic areas for raw materials are Florida
and Georgia, although mining has occurred in western and other southeastern
states. Outside of the U.S., major deposits are located in Australia, Canada,
Brazil, South Africa, Sri Lanka, India, and Mexico. Often ores with higher
specific mineral content were imported, such as Nigerian zircon sand for hafnium
processing and Australian zircon sand for zirconium processing.
The beach and fluvatile sand deposits from these areas are rich in mar-
ketable ilmenite, rutile, monazite, and zircon. Monazite commonly incorpo-
rates thorium and uranium as well as rare earths due to similarity in geo-
chemistry and electronic structure. Ilmenite and rutile are ore materials
for titanium processing, monazite is the principal ore for rare earth proces-
sing, and zircon the principal ore for zirconium and hafnium processing. Gen-
erally, these beach or placer sands are treated to produce heavy mineral con-
centrations containing the zircon, rutile, ilmenite, monazite, and other
marketable minerals. The concentrates may then be treated by various combina-
tions of gravity, electrostatic or electromagnetic methods to separate the
individual minerals. Monazite being slightly magnetic can be separated from
zircon by electromagnets. The purity of the zircon product (or conversely the
degree of monazite contamination of the product) is obviously a function of
the degree of separation effort. Initial treatment usually is provided at or
near the mine site. As a rule of thumb, the sand deposits are usually but
not always processed primarily for the titanium content in the form of rutile
and ilmenite. The zircon and monazite fractions are then byproducts which
are treated separately to extract zirconium and rare earths, respectively.
Thorium is then a further byproduct of the rare earth processing of the
monazite portion. This has been the major source of thorium up to the present.
For zirconium metal production, zircon sand is usually processed to min-
imize the monazite content since the phosphate content of the monazite has a
deleterious effect on the metallurgical process. This in turn should mean a
lower thorium and uranium content in the metallic zirconium wastes than in
foundry wastes where the monazite content of the zircon sands should be of
less importance to the process. However, Wagstaff has reported levels of
radium-226 from the uranium decay chain to be about 100 pCi/g in incoming
zircon sands at both foundries and metallic zirconium production facilities
(Wa 78). As Table 1 shows, the uranium and thorium content of monazite con-
centrates varies depending on where the ore is mined. The amount of monazite in
the zircon sand also depends on how well the separation facility removed the
monazite before shipping to the use point. At the use point (such as a zir-
conium metal manufacturing plant), the manufacturer may find it necessary to
further separate monazite from the sand. Low-level radioactive wastes may be
generated at each separation point. The natural concentration of uranium and
thorium decay series products in the sands are low but the industrial proces-
sing to obtain the specific minerals concentrates in the waste residues of
these normally occurring radioactive materials. The disposition of these waste
residues is the subject of this paper.
Case Studies of Three Facilities
The zirconium and hafnium processing method was developed by W J Kroll
for the U.S. Bureau of Mines. The bureau established a pilot plant'in'l947
at Albany, Oregon, to extract zirconium and hafnium using the Kroll process
and a purification plant in 1951 at Oak Ridge, Tennessee, to produce high puri-
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51
ty, low hafnium, reactor grade zirconium. The zircon sand is mixed with graph-
ite or coke and is fused in an electric furnace to produce a mixture of car-
bonitrides of zirconium and hafnium. The carbonitrides are chlorinated in a
vertical shaft furnace and the gaseous chlorides of zirconium and hafnium are
collected in a nickel condenser. The zirconium and hafnium chlorides are re-
duced in the Kroll process to the metals by reaction under an inert atmosphere
with magnesium. The end product, commercial grade zirconium sponge, will con-
tain about 2% hafnium suitable for non-nuclear use. Current industrial prac-
tice uses zirconium tetrachloride produced by chlorinating zircon directly
instead of the carbonitride (MF 75). In order to produce reactor grade zir-
conium, i.e., that containing about 0.3% hafnium, the commercial grade zirconium
sponge is dissolved and the hafnium is solvent extracted to hafnium thiocyanate
using methyl isobutyl ketone. The hafnium is precipitated as a hydroxide,
calcined to about 99% hafnium oxide. The resulting zirconium sponge is crushed,
compacted into consumable electrodes, and vacuum melted in an inert atmosphere
to ingot. Further product purity is achieved by applying the deBoer-vanArke
refining process. A similar procedure is applied to the hafnium solvent ex-
traction in order to obtain high purity hafnium metal. The residues generated
by the extraction processes contain graphite, coke, unreacted silicates, and
non-volatile silicates.
Wan Chang Corp. began operating the Bureau of Mines' Albany, Oregon, fa-
cility in 1955. Today it is one of the major producers of reactor grade zir-
conium and hafnium metals. Concern over the environmental and health safe-
guards at the facility grew when explosions were encountered during digging
operations near the facility's industrial waste piles. Apparently the explo-
sions were caused by rapid combustion of the zirconium in the waste piles.
At the same time the Radiation Control Section of the Oregon Department of En-
vironmental Quality (DEQ) became concerned that the large chlorinated residue
piles may be a radiological problem. A gamma radiation survey showed maximum
reading of 1200 uR/h. When the Oregon DEQ checked the radium concentration
of the piles, it found that the Ra-226 ranged from the original zircon sand
concentration of about 60 pCi/g to over 1300 pCi/g. One water sample taken
within the residue pile showed Ra-226 concentration of 45,000 pCi/L, hence
the concern of a potential ground water contamination. These radiological
parameters for the rare metals chlorinated residues can be compared with those
for uranium mill tailings.
Most uranium mills in the U.S. typically processed an average uranium ore
grade of about 0.15-0.35% uranium which would give expected radium-226 concen-
trations in the mill tailings ranging from 420-980 pCi/g. Individual tailings
samples at a given mill may have concentrations that are more or less than
these values by as much as a factor of five or so.
Due to Oregon DEQ's work at the Wah Chang facility, Oregon limited the
volume and radium content of chlorinated residue which the facility may accu-
mulate onsite before disposal in an out-of-state facility is required and man-
dated all users of zircon sand to file an application for a radioactive mate-
rials license. The criteria for release to an unrestricted area are 57 uR/h,
30 pCi/L of Ra-226 in effluent, and 0.03 WL of radon* (Wa 78). Of the twenty-
* One Working Level (1 WL) is a unit describing any concentration of
short-lived decay products of radon-222 in one liter of air which results
in the release of 1.3 x 105 MeV of potential alpha energy.
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SIX
52
a,A potential users of zircon sand, the state estimates only four will require
specific licensing. Similarly, Utah has restricted onsite accumulations to
no more than 100 tons of chlorinated residue and no more than 3 curies of
Ra-226.
Efforts to determine the extent of possible radiological problems are more
difficult for sites which have ceased rare metals processing activities for
several years either due to changes in site ownership or unfavorable economic
climate. Locating residue piles and sludge ponds, estimating amount_and_origin
of ores processed, and determining processing and waste disposal activities
must rely on historical records which are vague or nonexistent.
In the mid-19501s the responsibility for zirconium production was shifted
to private enterprise when the civilian nuclear power program was established.
In order to meet the increased demands for reactor grade zirconium, Carborundum
Corp. which operated a facility in Akron, New York, expanded its production
capacity by building a facility in Parkersburg, West Virginia. The plant's
designed capacity was 600 tons annually; it began operations in 1957. In the
mid-1960's, Amax Specialty Metals Co., Inc., became a partner and in 1967
obtained full ownership of the company. The Parkersburg site was sold in 1977
to L. B. Foster Company, a steel pipe fabrication plant. As a result of Fos-
ter's plan to expand its buildings, highly flammable waste materials were en-
countered during backhoe operations. In investigating the causes of the ex-
plosion, it was discovered that zirconium and thorium may have been buried
onsite, and that Amax Specialty Metals had not adequately terminated its li-
cense with the U.S. Nuclear Regulatory Commission (NRC) for possession of
radioactive materials. The NRC estimates that two million po.unds of zircon
ore, mainly from Nigeria, were processed at the Parkersburg plant since 1957.
A radiological survey of the site shows gamma radiation to range from a back-
ground level of 10 uR/h to 150 uR/h. Soil samples show concentrations of
thorium-232 and its decay products to range from background level of 1 pCi/g
to 10,000 pCi/g. The thorium contaminated area is limited to a few acres of
the 100 acre site. The NRC's tentative clean-up goal of 5 pCi/g above back-
ground of thorium-232 with a three to four foot overburden and deed restric-
tions on excavation was developed based on an assessment of the long-term
hazard due to thoron (radium-220). A radiological survey of Parkersburg by
a contractor to Amax Specialty Metals estimates 50,000 cubic yards of soil
may need to be removed. Some disposal alternatives being considered are
burial at a disposal facility, burial onsite in a clay lined cavity with land
use restrictions provided for the burial area, and ocean disposal. Whether
there are other locations onsite where zirconium and/or thorium are buried
may never be known since records on waste disposal and processing activities
are incomplete.
The Akron, New York, zirconium and hafnium processing facility was the
pilot plant for the Parkersburg, West Virginia, facility and presently is
owned by Amax Specialty Metals Co., Inc. Processing activities by Carborun-
dum Metals Co., Inc., began in 1953 at the Akron site to produce hafnium free
zirconium under an Atomic Energy Commission (AEC) contract.
The plant's designed processing capacity was 162 tons of zirconium an-
nually. Although the plant had a contract with the AEC to produce zirconium
metal, there does not appear to be any AEC, NRC, or MYS (an agreement state)
license for byproduct material other than for research and development pur-
poses. An industrial license with the NYS Department of Labor (DOL) for x-ray
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53
and gamma sources was in effect from 1960 through 1978.
During the summer of 1978, the EPA Region II radiation office queried the
NYS DDL and Department of Environmental Conservation about the status of the
Akron, New York, plant. The EPA was concerned that a situation similar to the
Parkersburg, West Virginia, plant may exist at the New York plant due to their
operating history. EPA was informed that Amax Specialty Metals Co., Inc., had
contracted Atcor, Inc., to perform a radiological survey of the Akron site as
the initial step in terminating its industrial license with the NYS DOL. The
June, 1978, survey showed radiation levels ranging from a background level of
7 uR/h to 1500 uR/h outside, with some building measurements up to 40 uR/h
(Le 78a). The extent of processing activities at the site is not well known
due to incomplete records. There were areas where magnesium and zirconium
residues were found but no pyrophoric incidents occurred. A single soil sample
from an area with an external radiation reading of 1500 uR/h was analyzed and
showed the soluble portion contained the radioactive material, but no further
radiochemical analyses were performed. The elemental composition of the sample
indicates it may be Nigerian ore, the principal ore processed at the Parkers-
burg plant. Surface soil samples were taken at locations with above background
gamma radiation and were spectroscopically analyzed. The range of concentra-
tions of Ra-226, Pb-214, Bi-214, Ac-212, Tl-208, and K-40 are shown in Table 2
(Le 78b). For these isotopes, the background concentrations are less than
1 pCi/g except for K-40 with concentration of 12 pCi/g. The limited results
in Table 2 indicate levels of thorium and uranium chain nuclides elevated above
expected background levels.
The monazite fraction of zircon ores typically runs about 3-10% thorium
dioxide (Th02) content with a tri-uranium octoxide (11300) content up to 0.41%.
Zircon sands of 96.7% pure zircon are currently imported and quoted on a mini-
mum basis of 65% zirconium oxide (Zi^). Hence the maximum monazite content
of incoming zircon sands should be less than 3.3%. Assuming a 10% Th02 content
in the 3.3% monazite fraction of the zircon sand, the overall Th02 content of
the zircon sand should be less than 0.33% or less than 300 pCi/g. Nigerian
sands are reported to range from 0.4 to 7% Th02- Based on this and on the as-
sumption that the monazite is some lesser fraction than 3.3% of the non-zircon
portion of the sands, then the values of 120-150 pCi/g for the thorium chain
nuclides do not seem unreasonable. Similarily, using a 0.41% t^Og content in
the monazite fraction and a maximum 3.3% monazite content, one would estimate
a maximum uranium or radium-226 concentration, assuming equilibrium, of about
38 pCi/g, which seems to be in probably fortuitous agreement with the maximum
measured values of 35 pCi/g for Ra-226. The higher K-40 values are certainly
higher than the expected background values of about 12 pCi/g. However, one
stage of the hafnium purifying process uses a potassium chloride molten mix-
ture. If this plant used this process and if some of the molten mixture were
spilled, it could account, for the higher K-40 values since the potassium
chloride probably contains about 400 pCi/g.
Between June and September, 1978, Amax Specialty Metals removed soil from
areas with high gamma radiation levels. About 25 cubic yards of soil were
shipped to the commercial low level burial site in Barnwell, South Carolina.
Two of the three lagoons or infiltration ponds were excavated to a three feet
depth and the material was disposed in a nearby hazardous chemical landfill.
The radioactivity of the excavated material is not known since no analyses
were performed prior to disposal.
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54
In September, 1978, Atcor resurveyed the site after clean-up efforts and
made recommendations for additional decontamination to reduce levels to as
low as reasonably achievable." However, it is not known to what extent these
recommendations were pursued by the site owner. It is not known whether any
attempts were made to identify the source of the slightly elevated levels of
radioactivity in the buildings.
In December, 1978, the NRC performed a survey of the Akron site (St 78).
The survey identified one area near a ridge with gamma radiation of twenty
times background. Areas near the lagoons and the tube mill building had levels
ranging from background to ten times. Analyses of the soil samples from these
three locations indicate thorium concentrations in the range 6.0-19 pCi/g with
background concentration of 1.2 pCi/g. The one air sample showed no Rn-220
daughters above background levels. Gamma radiation levels in the buildings
were within twice background, indicating little contamination after removal of
the gamma gauges.
The site could have been released for industrial use with little clean-up
necessary in order to meet the NYS DOL Industrial Code Rule 38 that no gamma
radiation levels exceed 250 uR/h at the surface and that source material in
soil be less than 0.05% by weight, which is 5,000 pCi/g of Th-232. Due to ex-
perience at rare metal processing facilities in Albany, Oregon, and Parkersburg,
West Virginia, Amax considered a more stringent clean-up program to meet the
goal of "as low as reasonably achievable." In general, the clean-up program
has been successful, although EPA Region II would have liked to have seen the
levels reduced to twice background and to 5 pCi/g for Ra-226 and Th-232. A
record of processing and disposal activities at the site would have greatly
assisted in answering questions concerning the possibility of any buried radio-
active materials.
Discussion
It appears from Figure 2 that the amount of zircon ore imported into the
United States has been increasing steadily since 1930. The imports account
for approximately 50% of the annual U.S. consumption of zircon concentrates;
the remainder is attributed to domestic production and stock piles. Australia
supplied about 60% of the imports before 1950, and over 95% after 1950. Brazil
contributed about 20% during the period 1930 to 1950. In total, the U.S. con-
sumed about 300,000 short tons of zircon before 1950 and 1.3 million short tons
thereafter.
The potential radiological problem can be likewise divided into two periods.
Prior to 1950, most of the Australian zircon was imported as a black sand mix-
ture containing zircon (40-75%), ilmenite (14-43%), rutile (7-18%), and monazite
(2-8/0) ores (MY 36). No attempts were made to separate the ores until the sand
mixture reached the processing facility. In 1948, the Commonwealth Government
declared its intent to purchase and stockpile monazite ore. As a result fu-
ture shipments of sand had most of the monazite ore separated in Australia be-
fore export.
In order to provide a conservative estimate of the radioloaical content
in the sands imported before 1950 the sand mixture is assumed to be composed
of 8% monazite ore. Assuming the Th02 content in the monazite fraction tn he
as high as 10%, then the Th02 contentZ1n the beach sands could reach 0 8? or
about 800 pci/g. Similarily? assuming the U308 content in the LnaziSe frac-
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55
tion to be as high as 1%, then the UsOg content in the beach sand could reach
0.08% or about 200 pCi/g. After 1950, the Australian ore had the monazite
fraction separated to some extent, hence the sands contain a minimum of 96.7%
zircon. Assuming the Th02 content in the monazite fraction remains 10%, then
the Th02 content in the beneficiated sands could be as high as 0.33% or about
300 pCi/g, the U30g content could reach 0.03% or about 100 pCi/g. Subsequent-
ly, the rare metals processing facilities which operated prior to 1950 may
have greater radiological problems with their chlorinated residues than those
which use ore obtained after 1950.
We would expect that any user of zircon sands would receive some monazite
in the zircon sands, since the separation of ilmenite, rutile, zircon, and
monazite, as indicated in Figure 1, is often incomplete. Hence some small
fraction of monazite, containing thorium and uranium will be present in any
industry which extracts titanium, chromium, tantalum, etc., from beach sands.
The monazite and hence the amount of thorium and uranium decay chain products
will vary with the sand origin and the degree of ore beneficiation. In fact,
facilities which need only the zircon, ilmenite, or rutile fraction of the
beach sands and insist on high purity ores may not have as great a radiological
problem as facilities which need only the monazite fraction or use sands with
little ore separation.
Producers of reactor grade zirconium have rigid specifications for thori-
um and uranium content and generally require high purity zircon which implies
low radioactivity content. To achieve this, the zircon is either purchased
as high purity material or further processed at the rare metal producing plant
to achieve purity. For metal production, then, the radioactivity remains in
the wastes while very little goes with the metal product. On the other hand,
the purity of zircon sands consumed at foundries is not critical, hence these
sands may have the highest radioactivity content. Manufacturers of refractory
materials, producers of milled or ground zircon, and ceramics manufacturers
will most likely have some portion of the radioactive content incorporated in-
to the products due to the manufacturing process.
In reviewing information from the annual Minerals Yearbook for 1929 through
1975, over twenty states were identified to have some facility which processed
beach sands in zirconium, hafnium, and rare earths production or used beach
sands in foundry processes. Table 3 provides a breakdown of the states accord-
ing to the type of processing or use activity. Facilities presently operating
in these areas of activity can be fairly easily identified and evaluated to de-
termine where these facilities dispose their chlorinated wastes and whether
the sands or concentrates used by the facilities have any appreciable monazite
fraction. For facilities which have ceased operating or changed their owner-
ship or their products, such an evaluation is more difficult.
Since only limited data and in some cases no data are available for radio-
activity and exposure levels associated with industries such as discussed above,
it seems apparent that considerable work needs to be done to assess the environ-
mental and health impact of such industries.
References
Co78 Costle, D., 1978, "Hazardous Waste Proposed Guidelines and Regula-
tions, and Proposal on Identification and Listing," Federal Register, Vol. 43,
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56
No. 243, December 18, 1978, pp. 58946-59028.
BUS Bliss, J. D., 1978, "Radioactivity in Selected Mineral Extration In-
dustries - A Literature Review," U.S. Environmental Protection Agency, Office
of Radiation Programs - Las Vegas Facility, November 1978, Technical Note
ORP/LVF-79-1.
Le78a Levesque, R. G., 1978, "Results of ATCOR's Gamma Scan Survey of
June 9, 1978," letter to H. Kail (AMAX Specialty Metals Corporation), June 20,
1978.
Le78b Levesque, R. G., 1978, "Results of Soil Samples - Analysis by
Teledyne Isotopes." letter to H. Kail (AMAX Specialty Metals Corporation),
November 16, 1978.
MF75 Mineral Facts and Problems, 1975, Bicentennial Edition, Bulletin
667, U.S. Department of Interior, U.S. Bureau of Mines.
MY Minerals Yearbook, Annual Publication 1929 through 1975, U.S. Depart-
ment of Interior, U.S. Bureau of Mines.
St78 Stohr, J. P., 1978, "Results of NRC Radiation Survey at AMAX in
Akron, New York," letter to F. Bradley (NYS Department of Labor), December 26,
1978.
Wa78 Wagstaff, D. G., 1978, "NORM - Problems in Oregon," paper presented
at the Region X Radiation Control Meeting, September 26, 1978.
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57
TABLE 1: Thorium And Uranium Composition In
Monazite Concentrates (Weight Percent)
Th02
Australia 7-8 1
Brazil 6-7 0.2
India 9-10 0.3
Madagascar 9 0.4
South Africa 6 0.1
United States 4-5 0.4
TABLE 2: Radioisotopic Concentrations In Soils
From The Akron, New York
Rare Metals Processing Facility (Le 78b)
Radioisotope Range of Concentrations (pd'/g Dry)
Ra-226 2.1 - 35
Pb-214 0.66 - 7.0
Bi-214 0.47 - 2.7
Ac-228 1.3 - 140
Pb-212 1.1 - 150
Tl-208 1.1 - 120
K-40 8.8 - 120
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TABLE 3: STATES WITH RARE METAL AND RARE EARTH PROCESSING ACTIVITIES
Producers of zirconium oxide, zirconium and
hafnium sponge metal, ingot, and alloy
Refractory firms using zircon in products
Producers of zirconium compounds and chemicals
Producers of zirconium oxide for other than
metal production
Milled and sold ground zircon
Producers of rare earth compounds and chemicals
Producers of high purity rare earth metals
Processed rare earth concentrates
Processed Canadian uranium mill solutions
for rare earths
NJ, MA, AL, MI, OR, WV, NY, OH, PA CA, NH
KY, NY, PA, MO, OH, WV, MI
NJ, NY, MA
NY, AL, OH, SC, NJ, WV
NJ, NY, OH, SC, DE, CA, PA
NY, PA, CA, CO
NJ, PA
IL, TN, NJ
MI
CJl
oo
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FIGURE 1: POTENTIAL BEACH SAND COMPOSITION
ILMENITE
MINOR
MINERALS
BEACH SANDS
I
TITANIUM - —
RUTILE
Ti02
•MH
Tn
Mi
ZIRCON
ZfiSiO/i
^m
ZlRC(
U.r-.
CHROMIUM
VANADIUM
COLUMBiUM
TANTALUM
CHROMIUM
I VANADIUM
COLUMBiUM
TANTALUM
MONAZITE
URANIUM
MINOR
MINERALS
— | KAYAr
~~| MAGNE
| GAR*-
— | STAURC
— [SILLIM
— | CHRO
-| Go
-------
SHORT
TONS
[ 100,000
I 80,000
60,000
40,000
20,000
n
r"" L
Hi
! Li
FIGURE 2: ZIRCONIUM CONCENTRATES PRODUCTION
IMPORTS TO U.S.
• —- U.S. CONSUMPTION
_ 160,000
- 140,000
_ 120,000
LT»
C-J
cn
CD
CD
CD
.=r
cn
LA
J3-
cn
cn
cn
o ir>
UD 1C
oo
cn
YEAR
cn
o
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SESSION B
WASTE HANDLING AND TRANSPORTATION
Session Chairperson
W. C. McArthur
Hittman Nuclear & Development Corp.
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62
DOE'S LOW-LEVEL WASTE MANAGEMENT PROGRAM
Sheldon Meyers
Office of Nuclear Waste Management, U.S. Department of Energy
Introduction
I wish to thank the Health Physics Society and its members for this
opportunity to discuss with you the DOE's less publicized low-level waste
program. I realize that the definition of radioactive waste by category or
type has proven troublesome. But since I have been preceded by several speakers
who have addressed low-level waste from differing perspectives, I will settle
for a simple definition of the term. In describing our low-level waste program,
I will be talking about solid radioactive waste which is currently being
disposed of by shallow land burial at carefully selected locations.
In May 1978, Secretary Schlesinger approved a reorganization within the
DOE which established an Office of Nuclear Waste Management which I now head--
reporting directly to the Assistant Secretary for Energy Technology. Within
that office there are three programmatic divisions. The Division of Spent
Fuel and Transportation deals with spent fuel storage and all waste-related
transportation requirements. The Division of Waste Isolation is responsible
for all efforts related to deep geologic repository disposal of defense and
commercial waste. The Division of Waste Products is responsible for technology
development for both commercial and defense wastes, and also handling and
treatment requirements leading to disposal of defense wastes. This includes
disposal of low-level wastes at DOE sites.
The management of low-level waste is a part of our defense waste program,
since the only low-level waste that DOE is directly responsible for is that
generated at DOE sites. We do not have responsibility for the commercial
radioactive waste burial grounds licensed by the Nuclear Regulatory Commission
(NRG). However, DOE R&D activities related to disposal of its own low-level
waste are also applicable to disposal at the licensed sites.
Before describing our low-level waste program, I would like to say a few
words about the past low-level waste situation. Past reports and Congressional
hearings have indicated that performance of low-level waste disposal operations
"is not uniformly good" and that radionuclide migration in ground water has
occurred at some sites. While releases into the environment have thus far not
been a public health hazard, they have raised questions concerning the adequacy
of the concept of shallow land disposal of low-level radioactive waste. In
view of these questions, and because of the predicted future need for this
disposal concept, DOE has underway a program to develop an improved technology
for shallow land burial, including improved environmental monitoring needs at
DOE disposal sites. Also, since some wastes now categorized as low level may
not be acceptable for shallow land burial in the future, DOE plans an expanded
program for investigating alternative disposal methods.
As a further preface, I wish to call attention to a recent Government
initiative which will impact our low-level waste program. With an earlier DOE
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63
task force report as a starting point, the President on March 13, 1978, estab-
lished an Interagency Review Group (IRG) on nuclear waste management with
representatives from 14 Government entities. The IRG was given the mandate to
undertake a comprehensive review of nuclear waste management in its broadest
sense, and to make policy recommendations to the President, as another step
towards the establishment of a National waste management policy.
Key IRG Low-Level Waste Findings
To review briefly, the IRG had as its objective the formulation of recom-
mendations for dealing comprehensively with the Nation's nuclear waste. They
have summarized the state-of-knowledge for managing the various waste types
including low-level waste. The results of their deliberations on low-level
waste can be summarized as follows.
As to technical findings, they consider that:
o Technologies exist for management and disposal of low-level waste.
Existing practice must be improved considerably. Siting must be
improved and research and development accelerated, including R&D on
alternative disposal methods.
In terms of policy, they recommend that:
o DOE assume responsibility for developing a National plan for low-level
waste, with input and involvement of other Federal agencies, states,
industry, and the public. The plan is to include establishment of
an adequate number of regionally located disposal sites.
o States be provided the option to retain management control of existing
commercial low-level land burial sites or to transfer them to the
Federal Government.
If public laws are modified to extend Federal management over commercially
operated disposal facilities, a program will be developed to implement such
legislation.
Following its review of public comment on the draft report, the IRG
reconvened to prepare a Decision Paper on Nuclear Waste Management, in fulfill-
ment of its directive from the President. This paper is scheduled to be
prepared for the President early this month (February).
We have begun implementation of some of these recommendations but, as you
can appreciate, overall implementation of the IRG's recommendations awaits
Presidential review.
The DOE Low-Level Waste Program
I plan to discuss today our current approach to the development of tech-
nology for an overall low-level waste program rather than operational detail--
other relatively recent symposia presentations and various reports have provided
considerable detail on operation of DOE's burial sites and quantities of
low-level waste being disposed of by shallow land burial.
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64
One major recent innovation affects responsibilities for management of
our program and manpower resources devoted to structuring and implementing the
program. In a decentralization move, assignments of broad responsibility for
technology development in specific waste areas have been made to DOE Field
Offices. Our Idaho Operations Office has this lead management responsibility
for low-level waste, with their on-site contractor, EG&G, in a support role.
Idaho has selected as an associate lead center the Oak Ridge Operations Office,
supported by the Oak Ridge National Laboratory. Under this arrangement, Idaho
has the responsibility for developing a detailed low-level waste management
program and for coordinating implementation of the program among all DOE
sites.
This decentralization effort has been underway for only a few months.
Formal planning documents generated in the next few months will be given broad
distribution for review and commend as one mechanism for greater involvement
of interested parties outside DOE.
Perhaps the best way to discuss program content is to discuss briefly
each of 10 major program elements. These elements are the top level of a
detailed work breakdown structure to be followed in the development of a
National low-level waste program.
1. Program Management: preparation of detailed program and technical
management plans; establishment and coordination of Review Committees
to include non-DOE personnel; establishment of system-wide quality
control and quality assurance procedures; coordinate program inter-
action among all participants and with other DOE waste management
programs.
2. Systems Analysis: program data acquisition, development of a con-
solidated data base and application of uniform methodologies; infor-
mation and program analysis; information exchange and dissemination
between participants and with other interested agencies and groups.
3. Criteria and Standards Development: DOE interface for participation
with others in formulation of generic waste disposal/management
criteria and standards; coordinate development of DOE criteria and
standards; reconcile differences between DOE and generic criteria
and standards.
4. Public and Regulatory Interface: formulate and implement programs
which will include federal, state and local agencies, industrial and
professional groups, and the public in the DOE low-level waste
management program; establish appropriate interfaces with and between
regulatory agencies, and coordinate environmental studies/activities
required by NEPA regulations.
5. Technology Development: coordination, development, refinement,
and/or implementation of environmental monitoring methods and pro-
grams, waste handling and treatment techniques, and shallow land
burial technology.
6. Waste Generation Reduction: characterize wastes at sources of
generation; forecast types and amounts of wastes generated; assess
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65
means of reducing waste generation at existing facilities; establish
waste reduction goals; coordinate improvements in waste reduction
methodologies; coordinate implementation of waste segregation and
volume-reduction policies.
7- Current System Operations: maintain cognizance of ongoing routine
operation--direct management and implementation of site-specific
operations remains the responsibility of individual DOE field offices.
8. Future System Development and Acquisition: locate and develop new
waste processing and disposal facilities and sites as required to
provide a nationwide system for low-level waste management.
9. Commercial: establish interfaces with commercial waste generators
and commerical waste burial sites as required to exercise any DOE
responsibility for handling and disposing of commercial low-level
wastes.
10. Alternatives to Shallow Land Burial: provide new emphasis on study/
research efforts on likely alternatives to shallow land burial; for
example, intermediate depth disposal, deep disposal in either existing
or new structures, and disposal in mined caverns.
A major document for public review now in preparation will delineate a
National Strategy for low-level waste. It will contain considerably more
detail on planned efforts under the program elements discussed above. Since
milestones for program accomplishments will be contained in this document, it
will require reconciling differences in Federal agency timing for the major
milestones reflecting current agency commitments. These milestones are primarily
aimed at establishing criteria and standards required for the Federal agencies
to fulfill their responsibilities. Any duplication of efforts between Federal
agencies will also become more apparent, and resources can be redirected for
better efficiency.
Another ongoing effort is development by DOE of a contingency plan for
acceptance of commercially generated low-level wastes in the event licensed
disposal capacity becomes inadequate. This plan was requested by the NRC. It
is still in the formative stages, but will address possible changes in commercial
disposal capacity and a variety of options for DOE acceptance of the wastes.
I do want to stress that, at this point, DOE considers this only as a planning
exercise which it is prudent to have available. Hopefully, circumstances
requiring its implementation will not occur.
Conclusion
In summary, low-level solid wastes are currently disposed of by shallow
land burial Continued reliance on this disposal method depends on development
of an improved technology base, generation of acceptable and comprehensive
criteria and development of stablilization techniques for sites no longer
needed to ensure that disposal areas will remain safe over the long term with
minimal reliance on continuing maintenance and surveillance. We envision
large potential increases in low-level waste quantities, particularly rubble
and soil from decommissioning actions and other high volume wastes containing
very low concentrations of radioactivity. This, plus the possibility that
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66
certain "low-level" wastes may not be acceptable for shallow land burial, have
led to an expanded DOE program to develop and assess alternative disposal
technologies.
Successful development and implementation of the planned DOE program will
to a large degree depend on assistance from organizations you represent here
at this meeting—the fact that a major meeting would focus exclusively on
low-level waste management highlights the importance of this subject. One of
our major aims is greatly increased direct involvement of interested individuals
such as yourselves. Several of the DOE and DOE contractor personnel directly
available in our low-level waste program will be available throughout this
week's meeting. I hope you will find time to discuss the DOE program with
them.
I would like to conclude by stating that we are making every reasonable
effort to assure that DOE generated low-level waste is managed in a safe and
environmentally acceptable manner. Any additional low-level waste assigned to
our custody will be treated in the same manner.
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67
A NUCLEAR POWER PLANT APPROACH TO TEMPORARY
ON-SITE STORAGE OF RADIOACTIVE WASTE
M. R. Buring
Metropolitan Edison Company
Reading, Pennsylvania 19605
E. E. Gutwein
Girbert/Commonvealth
Reading, Pennsylvania 1960S
Abstract
Designed storage space for solidified radioactive vaste and spent filters
at nuclear power plants has long been an area of concern. Increased personnel
exposures arise when waste cannot "be shipped for offsite disposal on a timely
basis. In addition, significant cost savings can "be realized in both transpor-
tation and burial costs if the short-lived isotopes in the solidified waste are
permitted to decay. In order to increase flexibility in plant operations a
large storage area for radioactive waste was designed. The design utilized
existing structures as much as possible. A large concrete roof over a heat ex-
changer vault was connected by a dock to the radwaste processing area. The
maximum anticipated radiation levels from the solidified waste were utilized
in designing the shield wall around the storage area. Salient features of the
storage area include four spent filter storage wells, a dual track overhead
crane for moving 50 ft3 liners and an enclosed truck pad to facilitate loading.
The 6000 ft^ storage area can conveniently handle 99 50 ft3 liners on 5 ft x
5 ft pallets or 396 55 gal drums on similar 5 ft x 5 ft pallets. This is approx-
imately one half of one unit's (PWR) average annual output of solidified evapor-
ator bottoms. The spent filter storage wells will permit approximately two years
decay with an estimated annual savings in transportation and burial charges of
$12,000. Provisions for maintaining radiation exposures ALARA were included in
the design of the facility. It was felt that conservatively estimated savings
in radiation exposures and shipping and burial costs justified the project.
Introduction
Storage of solidified radwaste is frequently a problem at operating nuclear
power plants. Since Three Mile Island Nuclear Station (TMINS) Unit 1 went com-
mercial in 197^, several radwaste handling and storage problems have been identi-
fied. In 19T6 plans were initiated on a Radwaste Storage area to alleviate sev-
eral of these problems. Any proposed modifications were to satisfy the following
criteria:
1. Storage space for normal volumes of solidified or compacted waste
between processing and shipping.
2. Storage space for spent filters to allow decay for reduced shipping
and burial cost.
3. Mechanized movement of processed radwaste to reduce personnel exposures.
k. Storage of processed waste in an area where personnel access is not
required to reduce personnel exposures.
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68
5- Increased speed and efficiency in loading of radwaste shipments
to reduce personnel exposures.
6. Storage space for new empty radwaste containers.
Existing Conditions
The volume of waste processed and shipped to burial sites has varied over
the years. In 1975, the first complete year of commercial operation, the vol-
ume of waste solidified for shipment was l6,l80 ft3. In 1976, the volume was
decreased to 12,260 ft3 and in 1977 to 8,000 ft3 due to an incident with a
leaky liner which curtailed shipments for several months. The 1978 volume was
13,100 ft3. When Unit 2 reaches full commercial operation, (currently expected
in January 1979), it is anticipated that the total annual volume will be 30,000
The radwaste is currently stored in two locations: in a room for which
volume reduction of radwaste is planned and below the Spent Fuel Storage Area,
one level below the radwaste processing facility. Neither of these locations
has proven to be satisfactory. The proposed volume reduction room is located
between Units 1 and 2 and is in a main traffic pattern between the units. This
area, therefore, is unsatisfactory for a storage area due to radiation exposure
considerations. Storage area on another elevation is not acceptable in that it
requires a large expenditure of manhours and manrems due to the excessive handl-
ing of the drums. Furthermore, these areas do not provide the necessary space
for flexibility in storage to allow adequate time for decay and variations in
shipping schedules.
Over the past years, further difficulties have arisen regarding the ship-
ping of solidified radwaste. TMINS Unit 1 has utilized the Maxey Flats burial
site in Morehead, Kentucky, which is now closed. More recently, the Sheffield,
Illinois, facility, also in the Midwest, has closed. TMINS is now forced to
ship waste to Barnwell, South Carolina. This facility has recently imposed a
maximum volume per month limit. In addition, burial charges are proportional
to radiation levels. Furthermore, with Unit 2 going commercial in 1979, the
radwaste problems will be compounded. These problems and others have necessi-
tated that Met-Ed provide additional radwaste storage area for both Unit 1 and
Unit 2.
Design Features
Initial Criteria
Having identified the existing conditions several general design features
or guidelines could be identified. The guidelines are as follows:
1. Locate storage area as close to and on the same elevation as the
existing processing facility.
2. Provide storage area for spent filters.
3. Provide sufficient storage areas to be able to store waste six months
at current or anticipated production rates for the two units.
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69
U. Provide truck loading capabilities.
5- Provide adequate storage for empty 50 ft3 liners.
6. Satisfy radiation dose limits surrounding the storage area.
T- Maintain radiation doses to plant personnel ALARA.
^Using the above criteria and utilizing existing facilities as much as
possible, a storage area was designed. An area outside the Auxiliary Building,
a heat exchanger vault roof, was identified as "satisfying" many of the criteria
listed above. This large flat area could be relatively easily connected to an
existing dock. After considering several other alternatives (see Evaluation of
Alternatives below), it was decided that this area over the heat exchangers could
best be modified to meet the radwaste requirements.
The design was divided into two phases. Phase 1 involved modifications
deemed necessary for better utilization of the existing facilities. Phase 2 was
directly associated with the new radwaste storage facility. Phase 2 consisted
of the following:
1. A covered dock from existing auxiliary building door to the heat exchange
vault roof.
2. Construction of recessed spent filter storage wells in this dock.
3- An overhead crane for filter cask and shield handling.
k. A hydraulic dock board for matching of truck height with dock during
loading operations.
5. A covered truck pad for transport vehicles.
6, Enclosure of dock for all-weather waste handling.
7- Shield wall between the storage area and the access road.
Radiological Criteria
(l) Spent Filter Wells. TMINS Unit 1 generates an annual average of 60 small
and 10 large spent filters with a total of about 50 curies. The isotopic break-
down of the activity on the filters was assumed to be the same isotopic distribu-
tion as for the effluent discharges (see Table 1 below).
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Table 1
Nuclide Composition of Effluent
Releases for 1976 and 1977
70
Isotopes
Co- 5 8
Co-60
1-131
Cs-137
All Others
Totals
1976
of Total Initial Curies
50.0
312.0
71.3
5-3y
8.a
2.o6y
30. ly
N/A
Curies After
2 Year Decay
O.U9
0.02
1.73
0.0
2.68
10.5
0.0
15-5
1977
Isotopes % of Total Initial Curies
K-UO 0.217
Mn-5H 1.08
Co-58 23.8
Co-60 2. ill
Cs-131* 30.9
Cs-137 ^0.0
1-135 1.86
All Others <1.0
Totals
100$
l.3E+9y
312.Od
71.3d
5.3y
2.06y
30.ly
6.6hr
Curies After
2 Year Decay
0.1
0.10
.0098
0.82
7.88
19.1
0.0
50.0
28.0
It is currently planned to set regular 50 ft3 liners into the spent filter
storage wells so that spent filters can be dropped dry from the small spent
filter transport cask into the liner. Each 50 ft3 liner will accommodate a
year's supply of spent filters. At the end of the first year, the second well
will be used, etc. After two year's decay, the first year's accumulation will
be shipped for disposal. The two year decay time (700 days) (10 half-lives for
the predominant isotope Cobalt 58) will reduce the estimated initial activity of
50 curies by an estimated UU to 70$. The longer half-lived cesiums then become
the predominant isotopes; however, their lower energy radiations are easier to
shield. The shielding for the storage wells was sized assuming each liner had
50 curies of activity on spent filters with the above-mentioned isotopic distri-
bution. The plugs for the wells were designed so that radiation levels on the
dock will be less than 5 mr/hr.
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71
(2) Shield Wall Considerations. The shield wall separating the new
radwaste storage area from a vehicle access road was sized utilizing actual
radiation levels from 50 ft3 liners of waste. Waste liners with contact read-
ings of 1 R/hr and 5 R/hr were analyzed at specific locations in the proposed
storage area. The activity in the drums was assumed to be Co-60. The shield-
ing evaluation indicated that an 18 in. concrete shield wall would be required
to satisfy the radiation zone criteria outside the storage area.
(3) Other. Numerous other areas of the design were influenced by radiolog-
ical conditions. All surfaces which potentially could be contaminated are speci-
fied to be painted with an epoxy paint which will facilitate decontamination.
Radiation monitors were placed in the storage area with a remote readout to in-
form operations personnel of the radiological conditions. Loading waste has been
simplified by providing a dock board to allow driving a fork lift directly onto
the transport trailer. The dual track crane over the spent filter storage wells
can also be utilized for truck loading. The crane has a rated 15T capacity and,
therefore, can handle a 50 ft3 liner inside of a 3-inch lead shield. Since the
storage area did not lend itself to any partitioning due to aircraft impact cri-
teria imposed On the concrete walls of the existing building, radwaste containers
will be stored in the area segregating the waste in accordance with the radiation
levels. Higher level waste will be placed in the far end of the facility to mini-
mize the exposures to personnel who enter the area with additional waste or who
are in the process of loading waste for shipment to a burial site.
Evaluation of Alternatives
Conveyors
Consideration was given to a conveyor system for storage of the waste on the
heat exchanger vault roof. The idea was to have seven parallel tracks. Waste
would be placed on the conveyor at one end and shipped from the other. Due to
the high cost of this system, potential maintenance problems, and reduced storage
area this idea was dropped.
Open (Uncovered) Storage Site
Stored waste exposed to the elements can potentially contaminate run-off
water. The roof over the proposed radwaste storage area was justified by assum-
ing that all run-off would have to be monitored and possibly processed through
the radwaste tystem. Calculations indicated that processing of normal annual
rainfall collected on the heat exchanger vault roof as contaminated liquid waste
would require about hOO hours per year of evaporator operation. The normal per
gallon cost for processing contaminated liquid is $.25/gal. Solidification and
disposal costs are approximately $3-60/gal. The estimated 200,000 gallons of
rainwater/year more than justify the cost of the roof at $130,000.
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72
Remote Storage (Avay from Plant Proper)
In addition to the increased expense, the major objection to this
option was that a storage location away from the processing area would
not be as convenient as one next to the processing area. Furthermore,
a remote storage area would necessitate longer periods of handling by
the operators in order to get to the storage area and, therefore, would
likely increase radiation doses.
Overhead Remote Operation Crane System
An overhead crane indexing system would have been the ultimate in
a radwaste handling facility. However, the cost of such a system made
its serious consideration short-lived. In addition to the higher cost,
some difficulty would have been experienced in supporting such a system
in light of the rigid aircraft impact criteria imposed on outside walls
housing safety-related equipment.
Costs
Manrems
A total of 30.211* manrems were expended in 1977 in Radwaste Proces-
sing Operations and Maintenance. Twenty-two of these were absorbed by
Maintenance Personnel, and the remainder by Operations, Supervisory, and
Radiation Safety. Utility workers handling radwaste are included in the
maintenance personnel exposure. 1978 saw a(n) crease to manrems.
Annual manrems savings by mechanized handling are estimated at 50$.
Shipping and Burial Costs
Savings in shipping and burial costs assuming a decay time of 210 days
(3 half-lives of CO-58 a predominant isotope) and assuming 1978 shipping
and burial costs:
The cost difference between shipping a shielded truck of 50 ft3 liners
and an unshielded truck is $1,000/shipment. Met-Ed currently ships approx-
imately 15 such shielded shipments/year and Unit 2 is expected to double
this. The dividing line between shielded and unshielded shipments is approx-
imately 200 mR/hr.
Provision of 210 day decay time could reduce radiation levels from
1 R/hr (requiring a shielded truck) to 0.125 R/hr, with savings of $1 OOO/
shipment.
The Barnwell South Carolina burial ground currently imposes a $1.1*0
per cubic foot radiation surcharge for each container which exceeds 0 200
R/hr over those at less than 0.200 R/hr. A three half-life storage for
decay could reduce 90% of all TMI radwaste to <0.200 R/hr. Annual radwaste
volumes are currently estimated at 15,000 ft3/year/unit and, therefore
savings of $21,000 per year/unit or $1.2 million over the plant life could
be realized, again assuming 1978 shipping and burial costs.
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73
These figures do not include the liner surcharge which is currently
$1.50/ft3.
Having to ship filters without the benefit of decay also imposes a
financial penalty. The radiation surcharge currently imposed at the
Barnwell, South Carolina burial ground per ft3 is $13.18 in the 20-UO
R/hr range, and $U.6T in the 1-5 R/hr range. This difference of $8.50/ft3
for 100 ft3/yr is a direct savings of $850. One trip per year for filter
disposal will cost approximately $2,500 with decay. Without decay, three
trips per year will be required due to radiation levels for direct savings
of $5,000/yr plus an estimated $850/yr radiation surcharge savings. Manrem
savings estimates due to remote handling and handling of decayed rather
than fresh filters are not available at this time, but are assured.
Manhours
TMI currently expends approximately 6,000 manhours per year in rad-
waste handling operations. Mechanization by use of the forklift and use
of the new storage area, versus the present temporary storage area on the
lower level, could reduce this manpower requirement by 20$. Currently
50 ft3 liners are transported from the operation level to the next lower
level by three persons in the elevator for temporary storage. These liners
must then be returned from the lower level by reversed process for loading
on the shipping vehicle. Labor savings for forklift movement are estimated
at approximately $15,000/year based on $12/manhour or $600,000 over the life
of the plant. Addition of the Unit 2 waste volumes would make lower level
storage of radwaste an intolerable situation.
Estimated Cost of the Radwaste Storage Facility
Cost Estimate
Engineering Sub Total $122,000
Construction
Phase I la,000
Phase II
Covered Dock 2^2,000
Filter Storage Wells 20,000
Monorail 53,000
Dock Board 9,000
Dock and Storage Area Roofing 130,000
Shield Wall 27,000
Sub Total $522,000
Grand Total $6itU,000
Summary
This utility feels that the savings in manrems, shipping and burial
costs, and manhours justify the expenditure of this sum to convert a
currently unused space to a radwaste storage area.
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74
RADIOACTIVE SLUDGE TREATMENT BY THE
CENTRIFUGAL DEHYDRATION METHOD
S. L. Hwang and C. M. Tsai
Health Physics Division
Institute of Nuclear Energy Research
Atomic Energy Council of Executive Yuan
P.O. Box 3-10
Lung-Tan, Taiwan 325
Republic of China
ABSTRACT
The radioactive sludge produced after chemical treatment
of radioactive liquid waste usually contains 0.3-3% of solid
by weight. The gross 3 activity of sludge produced at the
Institute of Nuclear Energy Research (INER) varies 10~1-10~3
nCi/g, and the use of centrifugal method for dehydration is
justified. The effluent from the decanter goes through the
precipitation and concentration stage and then the supernatant
is pumped so that a stable concentration of liquid waste can
be maintained during separation stage. The decanter used at
INER for treating low level radioactive waste is a Westfalia
SDB230 decanter which is of solid-bowl type. The liquid waste
from a 40MW(th) Taiwan Research Reactor (TRR) and its associated
hot laboratories is described in this report. The treatment
capacity is about 2m^/h and the characteristics of the separator
and the associated systems are discussed. As a result of im-
provement the solid in concentrated sludge is about 20% by
weight, and the production rate is about 240-250 kg/h. Fur-
thermore, no water is added to the concentrated sludge. The
solidification can be done in a closed type sludge-cement pre-
mixing system, and the homogeneous concrete can be obtained.
INTRODUCTION
The radioactive liquid waste produces radioactive radio-
active sludge after chemical treatment. The radioactive sludge
so treated contains solid generally at a few percents. The
radioactive sludge produced at INER has 3-6 wt% of solid content
depending on the period of gravity thickening. It is not econo-
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75
mical to treat such great amount of sludge by storage or solidi-
fication. Therefore, it is necessary that dehydration be firstly
made to minimize the volume of the radioactive sludge prior to
solidification. The current treatments of dehydration can be
classified as:
(1) Gravity,
(2) Air drying,
(3) Gravity filtration,
(4) Vacuum filtration,
(5) Pressure filtration,
(6) Centrifugal dehydration.
The merits and shortcomings of treatment mentioned above are
presented in Table 1 where the volume reduction ratio is defined
as:
Radio of volume reduction= Volume of feeding sludge
Volume of concentrated sludge
Table 1. The Comparisons of Different Dehydration Treatments
for Radioactive Sludge
No.
Treatment
Merits
Shortcomings
1 Gravity thickening simple operation, low volume reduction
low cost ratio
2 Air drying
simple operation,
low cost
3 Gravity filtration simple operation,
low cost
4 Vacuum filtration
more adoptive to
various sludges
Pressure filtration high filtration
speed
6 Centrifugal
dehydration
low contamination,
high volume
reduction ratio
easy to become con-
taminated, vast
space required
laborious to remove
concentrated sludge
filter frequently
replaced, compli-
cated and expensive
facilities
operation compli-
cated, easy to
become contaminated
not applicable to
particulate sludge
-1 -2
The sludge produced at INER has approximately 10 - 10
Ci/g of gross 6 activity suitable to centrifugal dehydration.
The effluent from such dehydration has less solid content. As
long as..it meets the application of gravity thickening, the
sludge is setting and the clarified liquid can be pumped out.
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76
In this way the solid content of the sludge may be applied to
batch cycle treatment in a steady condition.
The separator used at INER is called the solid bowl type
decanter. The inner diameter of the separators bowl is 22.5 cm
while that of the surface of revolution liquid is 13.5 cm. The
revolution speed of the bowl is 2850 rpm and its G force is
approximately 820 (Gr76). Ten years ago, the G force of the
separator was about 800-2000 and has increased to 3000-5000
in the recent years.
EXPERIMENTAL
The separator of the solid bowl type decanter is often used
in a BWR type nuclear power plant. The type of decanter used
at INER is a continuous centrifuge with a horizontal solid shell
bowl. A conveyor screw is enclosed for the separation of solids
from suspensions. By changing the regulating ring dam, altering
the difference in speed of the bowl shell, and using the parti-
cular type of suspension, the maximum clarification and dryness
can be obtained regardless of the specific gravity of solids.
The working principle of the bowl is as follows:
The bowl operates in two zones, i.e., separating zone and
drying zone. In order to see that the separating function meets
the requirements, it is necessary to make an appropriate selec-
tion between separating zone and drying zone. In other words,
the following two principles can be followed to decide the
proper inner diameter of revolution.
1. The solid output is required to be as dry as possible.
The degree of purity of the clarified liquid is of secondary
improtance. In this case, it is necessary to work with a long
drying zone. This is done by selecting the largest inside
diameter regulating ring dam of the bowl.
2. The effluent is required to be of maximum purity. The
degree of dryness of the separated solid mater is of secondary
importance. In contrast to principle 1, it is necessary to work
with a long separating zone, and the smallest inside diameter
regulating ring dam should be used.
In the majority of cases, however, it is necessary to com-
promise. The most satisfactory regulating ring dam and speed
difference for the suspension to be processed are found by
making several tests.
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77
RESULTS AND DISCUSSION
The results of treatment under various operation condition
will be described separately as below.
1. The feeding rate remains constant. By changing the
regulating ring dam, the results are shown in Table 2 where n
indicates the difference in speed between the bowl shell and
the conveyor screw and 0 is the diameter of regulating ring dam.
Table 2. The Efficiency of Concentration as a Function of
Ring Diameter
Feeding
rate
T
(nr/h)
2
2
2
2
n
(rpm)
18
18
18
18
Regulating
ring $
(mm)
160
145
135
127.5
Solid
content of
feeding
(wt%)
6.70
6.70
6.07
6.07
Solid
content of
effluent
(wt%)
6.70
5.48
5.67
—
Solid
content of
concentrated
sludge
(wt%)
_
20.63
23.83
-
Concentrated
products
(kg/h)
0
4.8
31.8
0
2. The diameter of the regulating ring dam is 135 mm and
the feeding rate is 2m^/h. The results of treatment under
various operation conditions are shown is Table 3.
Table 3. The Efficiency of Concentration as a Function of n
Feeding
rate
o
(mj/h)
2
2
2
2
n
(rpm)
6
12
18
30
Regulating
ring 0
(mm)
135
135
135
135
Solid
content of
feeding
(wt%)
5.60
5.16
6.07
6.35
Solid
content of
effluent
(wt%)
5.14
4.76
5.67
5.60
Solid
content of
concentrated
sludge
(wt%)
16.25
16.68
23.83
18.12
concentrated
products
(kg/h)
18.0
25.1
31.8
11.8
3. The diameter of the regulating ring is 135 mm and n is
18. The results of treatment under various feeding rates are
shown in Table 4.
Table 4. The Efficiency of Concentration as a Function of
Feeding Rate.
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78
Feeding
rate
o
(nr/h)
1
2
3
4
5
n
(rpm)
18
18
18
18
18
Regulating
ring 0
(mm)
135
135
135
135
135
Solid
content of
feeding
(wt%)
5.16
5.16
5.16
6.48
6.48
Solid
content of
effluent
(wt%)
4.79
4.76
4.29
5.91
6.21
Solid
content of
concentrated
sludge
(wt%)
23.78
21.47
19.77
16.63
14.38
concentrated
products
(kg/h)
26.2
25.1
16.6
13.0
7.7
Based on the data given in Tables 2-4, the improvement
of the dehydration system should be made.
1. The feeding and effluent receiving tank.
It is known from the above operation tests that
the concentration sludge in the dehydrated effluent
remains about 90% of the feeding. As shown in Fig.1,
the effluent cannot flow directly into the storage
tank of radioactive liquid waste. As shown in Fig.2,
after improvement the operation system consists of
two tanks for feeding and effluent receiving and one
receiving tank of sludge. The dehydrated effluent,
as a result of the gravity thickening in the settling
tank (Ko76), can remain the solid content of the sludge
of feeding about 5-6 wt% and raise the operation effi-
ciency of the decanter.
2. The use of flocculant agent(Ec70).
In order to increase the products of the concen-
trated sludge, the flocculant agent was used to enlarge
the particle size of the solid in the sludge so as to
accelerate the setting speed of the solid. The-dosage
of the commercial flocculant agent is 40-60 g/m .
When the flocculant agent is used, the production rates
of the concentrated sludge are increased to 240-250
kg/h.
In conclusion as mentioned above, concentration of the
radioactive sludge by centrifugal dehydration-method features
that the sludge can be treated in closed transportation system,
and the concentrated sludge can further be applied to cement
solidification by means of pre-mixing cement mixar. Upon soli-
dification, the cement and the concentrated sludge can mix
homogeneous and then fill in the 53 gallon drums. The operation
is simple enough yet with less contamination and weather resis-
tant .
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79
REFERENCES
EcVO Eckenfelder W.W. and Ford D.L. 1970, "Water-Pollution
Control, Experimental Procedures for Process Design,"
Gr76 Gregorio D.D. and Shell G.L. 1976, Proceedings ASCE
102, 1087.
Ko76 Kos P., 1976, "Continuous Gravity Thickening and Sludges,"
IAWP 8th International Conference October 17-22, Sydeney,
Australia.
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FEEDING
SLUDGE
RECEIVIN6
EFFLUENT
SUXME
TANK
DECANTER
LIQUID WASTE
STORAGE TAMK
CONCENTRATED
SLUD«E
Fig.1 Before the Improvement of the Dehydration System.
oo
o
-------
FEEDING
SLUDGE
RECEIVING
I
SLUDGE
RECEIVING
TANK
EFFLUENT
FEEDING AND EFFLUENT
RECEIVING TANK
CONCENTRATED
SLUDGE TANK
Fig.2 The Improvement Operation System
00
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82
REDUCTION IN WASTE HANDLING AND TRANSPORTATION
by John E. Stewart
Werner & Pfleiderer Corporation
INTRODUCTION
The reduction in volume of low level wastes and
associated handling and transportation is especially
important to health physicists. In most cases the result is
less exposure to operating personnel and less risk of
exposure to the public during transportation and
subsequent burial or other disposal.
Credit for initially exploring the benefits of volume
reduction must go to the Europeans who had to face the
radwaste problems earlier due to their limited burial and
disposal facilities. 1 2 They performed research on
volume reduction and solidification processes using
various binders in the early 1960's. This paper not only
draws upon European experience, but also that of the
U.S. and other countries for improving the radwaste
process we know as the Volume Reduction and
Solidification (VRS) system.
The radioactivity in liquid wastes is primarily in the
solids they contain. It then follows that separating out
the solids and solidifying them without the water is not
only a logical technological advance in itself, but one
with many attendant health physics benefits.
There are several low level waste solidification
systems on the market, 3 plus two systems that reduce
volume but must interface with a solidification system,
and one that both reduces volume and solidifies at the
same time. Other systems are under development. Each
of the different systems has different health physics
problems and it is impossible to discuss them all in one
paper. Therefore, this paper will discuss only the Volume
Reduction & Solidification (VRS) system of the Werner &
Pfleiderer Corporation, which has been accepted for
installation at several U.S. nuclear power plants and has
a wealth of data accumulated as a result of over 50
system years of operating experience in Europe and
elsewhere. The WPC - VRS system Topical Report has
been evaluated and accepted by the NRC for referencing
by utilities in their SAR's."
To get a better idea of what volume reduction means to
the U.S. nuclear power industry, refer to Figure 1. This
graph appeared originally 5 with only the left-hand
ordinate and we have converted MWt to MWe on the
right-hand ordinate to enable you to quickly determine
the waste quantities per 1000 MWe. From this graph, it is
readily apparent that volume reduction is essential.
Solidification, of course, is a regulatory requirement.
OPERATOR EXPOSURE REDUCTION
An integral part of every VRS system is a shielded control
room for the operator. From this room the operator can
observe the major processing operations through lead
glass windows. Closed circuit TV systems monitor all
other important phases of the process.
To illustrate this point, refer to Figure 2, which
represents a typical approach to a VRS system design.
Note that the process control panel and the crane control
panel are both in a shielded area and that there are three
lead glass windows. One window allows the operator to
directly observe the filling chamber, a second permits
observation of the drum capping and monitoring station,
and the third looks directly into the filled drum storage
area. Although not shown, several closed circuit TV
cameras are located inside the processing area with the
viewing screen in the control room. The net result is that
the operator works in an area that is at or close to normal
background radiation levels for a nuclear power plant.
All of the important phases of the process are
interlocked to prevent operation under serious abnormal
conditions and both audio and visual alarms are
provided in the control room.
10-2
Original ORNL dwg 78-18217
10
1958
1962 1966 1970 1974
CALENDAR YEAR
1978
Figure 1. Annual solid waste volume shipped from
U.S. nuclear power plants. Reference 5,
ORNL/NUREG-43.
-------
ROAD
ROLLED
DOOR
DIST
ASPHALT COLL LUBE
FEED \ &FILT OIL v
PUMP \SYS_\CpOLER\
PUMPS
CRANE TRAVEL AREA
TRUCK
LOAD! NIG
BAY
FILLED
DRUM
STORAGE
EXTRUDER
EVAP.
CAP&
MON.
STA.
\EMPTY
1DRUM
STORAGE
I OOOOOO
CRANE CONTROL
FILLING'STATION
CONTROL PANEL
83
ASPHALT STORAGE TANK
ASPHALT GEAR FEED
PUMPS /BOX /PUMPS
Figure 2. Typical design approach to VRS layout.
EQUIPMENT SEPARATION
VRS system design follows the philosophy of
separating the hardware that contains radioactive
material from other pieces of equipment. Note in Figure 2
that the asphalt feed tank is outside of the processing
area and that the radwaste hardware is separated by
shielding walls which effectively control the radiation
levels in the various areas according to their function and
anticipated occupancy times
The functional areas and their anticipated dose rate
are given in Table I below. For location within the total
system, refer to Figure 2.
TABLE I
FUNCTIONAL
LOCATION
DOSE RATE
R/HR
OCCUPANCY
HRS/YR
EXPOSURE
MAN-REMS/YR
CONTROL ROOM
FEED TANK & PUMP AREA
EXTRUDER-EVAPORATOR
(PROCESSING AREA)
DRIVE & AUXILIARY AREA
EMPTY DRUM STORAGE
FILLED DRUM STORAGE
DRUM FILLING AREA
TRUCK LOADING BAY
0.00025
0.1
0.1*
0.01
0.00025
1.0 to 10.0
1.0 to 5.0"
0.1
500-1000
10
10
10
100
0
10**
10
0.1 25 to 0.25
1
1
0.1
0.025
0
1.0
1
*Less than 0.010 R/hr during maintenance periods
because of self-cleaning action of the rotating screws in
the extruder-evaporator.
"When maintenance is required, all filled containers will
be removed and dose rate will be 0.01 R/hr.
-------
OPERATING RELIABILITY
Note in Table I that the areas of highest dose rate have
a minimun occupancy time per year. This is directly
related to equipment reliability which determines the
frequency and amount of time that personnel must be in
a particular area. The VRS process has over 50 system
years of operation. The two oldest installations are at
Marcoule, France, and Karlsruhe, West Germany. Both
of these are nuclear research centers and both keep
records of the system operation. A third installation at
Borssele, Holland (power plant) also kept track and their
results are presented in Table II below.
TABLE II
OPERATIONAL RELIABILITY
BORSSELE
&
KARLSRUHE
7,500 HOURS WITHOUT
MAINTENANCE
MARCOULE
13,000 HOURS BEFORE DISASSEMBLY
FOR ROUTINE MAINTENANCE
It is anticipated that the normal operation of the VRS
system in the U.S. will equal 2,000 hours per year. Thus,
the reliability record established in Europe indicates that
the VRS system would operate in the U.S. for years
before maintenance personnel would be required to
enter the areas of highest radiation. The exposures listed
in Table I for the processing area and other areas where
equipment containing radioactive materials are located
is based on the maximum frequency of one annual
inspection regardless of operating time.
Figure 3. Model T-120 VRS extruder-evaporator as
used in European installations.
CONVEYING
SCREWS
MIXING AND
KNEADING ELEMENTS
Figure 4. Co-rotating screws which operate with 1 mm
of clearance to provide self-cleaning action
inside extruder-evaporator.
COMPONENT DESIGN
The major piece of hardware is the extruder-
evaporator shown in Figures 3 and 4. Basically the
extruder-evaporator consists of two co-rotating screws
installed inside of heavy metal sections called barrels.
The barrels are constructed of thick nitrided steel and act
as a shield while the wastes are being processed. In
addition, at any given time during processing there is
only about two liters of waste being mixed with two liters
of asphalt binder. Thus, the low inventory in combination
with the thick steel walls of the extruder effectively keep
the radiation levels at the surface of the extruder-
evaporator at low levels.
For example, in Figure 5, the external dose rates after
processing medium level wastes of 15 and 90 Ci per
cubic meter in the feed stream are shown for Karlsruhe6,
West Germany. While the wastes are 10 to 100 times
higher than those expected at power plants, the relation-
ship of shutdown dose rates to input activity is
illustrated. It is, therefore, reasonable to expect service
dose rates of less than 10 mR/hr when shut down for
maintenance.
EFFECT OF LESS END PRODUCT
Volume reduction and solidification in the same
single-step process provide the greatest health physics
benefits. There is a reduction in the amount of handling
on-site as well as in the area required for on-site storage
prior to shipping off-site. The number of shipments to
the burial site is reduced and along with that a significant
reduction in the accident probability, which not only
involves the health physicist but the general public as
well.
Table III compares the shipments from a VRS system
with those from a non-volume reducing cement system.
TABLE III
SHIPMENT COMPARISONS
WASTE TYPE
BWR FILTER/DEMIN.
BWR DEEP BED RESINS
PWR WITH CPS *
PWR WITHOUT CPS *
DRUMS/YR
VRS
910
545
175
415
CEMENT
5,470
3,270
1,225
2,900
DRUM
REDUCTION
4,560
2,725
1,050
2,485
SHIPMENTS/YR
VRS
29
17
7
14
CEMENT
144
87
34
78
SHIPMENT
REDUCTION
115
70
27
64
*CPS — Condensate Polishing System
-------
85
According to a recently published report
(ORNL/NUREG-43, Oct. 1978, pg. 57),5 "At this time, the
main problem areas in radwaste solidification at the
power plant appear to be drum capping, monitoring, and
decontamination, which remains largely a manual
operation. To cope with drumming problems, many
plants have replaced or modified their original radwaste
processing equipment in an effort to perform more of the
operations remotely and automatically... The indication
levels of some spent resins and filter cartridges and/or
precoat filter sludges can be as high as 100 R/hr or more,
and unless adequate shielding is provided during
storage, transfer, and packaging, there is risk of high
exposure for operating personnel." All of the above
operations are done remotely in the VRS system, thus
eliminating the risk of increased personnel exposures as
reported by the nuclear plants to ORNL. It also follows
that the fewer drums or other containers there are, the
less is the handling that is required to move the
containers from the filling area through capping,
monitoring, decontamination, into on-site storage, and
finally to the transport vehicle. The reduction in
transport requirements and the resultant benefits have
already been mentioned. At the burial site, the same
benefits that apply at the plant site are in effect. In
addition to the health physics benefits obtained due to
VRS processing, there are other operational benefits
dealing mostly with economics.
If there is a long time storage at the power plant or at
the burial site, there is an additional benefit of the VRS
system with regards to drum corrosion. The VRS system
uses asphalt as a binder, which inhibits internal drum
corrosion, thus reducing the risk of exposure from the
source.
SYSTEM DESIGN PHILOSOPHY
The VRS design philosophy is to separate all
components containing radioactive materials, to isolate
the system operator from all of the potential sources of
radiation, and to reduce to a minimum the occupancy
time of any maintenance personnel when entry into a
radiation area is required. Many of the design faults
reported by ORNL and others as an operating deficiency
in nuclear power plant radwaste systems are not present
in the VRS system.
DISTILLATE COLLECTION SYSTEM. The distillate
collection system of the VRS process warrants
discussion. Water in the feed stream is heated to above
the boiling point in the extruder-evaporator, causing it to
be driven out through the steam domes mounted on the
top of the extruder. This distillate is condensed, passed
through filters and either returned to the plant system or
stored until suitable for release to the environment. The
IKonzentrat
/CN m v
Bitumen
Asphalt
, r
...
^y
t
\
\
I!
(
i p i
n „
C^^Z
70
, 150
m m. _
10 iO
110
1
so
ISO
30
LU Lu
70
50
110
IS
LU LLJ
15
15
to
20
JJ LU
Specific Activity of Concentrate
Spez. Aktivitdt des Konzentrats: 15 Ci/m3
10
;
IS
30
IKonzentrat
r
100 120
» 4
100
to
30
4
V 1 PJ
30
100
300
700
L
, .1
300
(00
LU
150
ill LU
ISO
250
100
LU U.
SO
50
100
LU LL
60
50
Spez. Aktivitdt des Konzentrats: 90 Ci/m3
Figure 5. Operating dose rates (surface mR/hr) at Karlsruhe Nuclear Research Center where
feed stream includes medium level wastes from fuel reprocessing center.
-------
86
decontamination factor, measured from waste inlet to
distillate, is 6,000. The radiation level of the distillate
filters following filtration is only a few mR/hr. during
normal changeout.
FILLING CHAMBER. The filling chamber is a shielded
area that is virtually isolated from all other areas in the
system. As the drums are filled with the mixture of
radsalts and asphalt, some residual vapor may be
present. Directly above the drum being filled is a hood
having a negative draft. Thus, all vapors are drawn
through the hood and pass through a prefilter before
exhausting to the plant HVAC system.
MONITORING. Swipe samples related to monitoring
and possible decontamination are done remotely by the
operator. The container handling system is designed to
prevent spills, but should one occur it can be cleaned
easily in the VRS system. The asphalt binder is a slow
flowing thermoplastic and it will harden quickly outside
of the container. After it solidifies, it can be scraped up
and put in any of the drums in the chamber. Any film or
residue after scraping can be cleaned with steam which
is returned to the extruder-evaporator. With the asphalt
there is no dust or free liquid, nor any solids that have to
be mechanically chipped away, all of which complicate
the health physics problem related to cleaning up such a
spill. Also, each drum is monitored and if contaminated it
can be cleaned with high-pressure heated water in the
decontamination station.
CATALYSTS AND SOLVENTS. No chemical catalysts or
promoters are needed and steam (or heated water) is the
only solvent recommended in the VRS system. This
eliminates the risks normally associated with solvents
and chemicals considered to be volatile and undesirable
in a radwaste system.
PROCESS UPSETS
LOSS OF POWER AND HEAT. Rarely would the
operator need to enter the radiation areas with the VRS
system. Even if there is a complete loss of system power
and steam heat, the operator does not have to enter the
processing area or filling area. What would actually be
required is that the operator must shut the system
controls off, then simply wait for the power or steam
service to be restored, even if the asphalt/radsalts mix in
the exturder-evaporator hardens. When power and/or
steam is restored, the operator waits until the proper
temperature profile in the extruder-evaporator is
achieved, then resumes normal operation as if nothing
had happened. This is in direct contrast to what happens
when an upset occurs with thermosetting binders, which
often solidify in place.
CHANGE IN pH VALUE OF THE FEED. The pH value of
the VRS radwaste feed stream is recommended to be in
the 7-11 range as shown in Figure 6. However, if the pH
should drop into the acid range, or rise to high alkaline
levels, the processing will continue without disruption.
The radwaste solids that go into the container will be
immobilized normally as the thermoplastic asphalt
hardens.
OIL IN THE FEED. Although oil in the feed may prevent
solidification in other systems, thus creating health
physics risks, it will not prevent solidification in the VRS
system. Oil and asphalt are compatible, both being a
petroleum derivative. The net effect is that oil will reduce
the viscosity of the asphalt in the end product in a direct
ratio according to the amount of oil. The amount of oil
normally found in radwaste feed streams, providing it
does not exceed one or two percent of the end product
volume, will not noticeably affect the VRS end product.
CHANGES IN MIX PROPORTIONS. The VRS will
produce a solidified end product even if the proportions
of feed stream wastes to asphalt are upset. These are
normally controlled remotely from the control panel.
Should upset processing conditions become normally
uncorrectable by the operator, both visual and audio
alarms will be given and the system will shut itself down
because of the interlocks designed and built into the
system.
REAGENTS. No reagents are used in the VRS system.
Therefore, reagent age and temperature effects, which in
turn could upset solidification, are eliminated. Other
possible effects of using chemical reagents are also
eliminated.
END PRODUCT BENEFITS
GENERAL. Again quoting the ORNL/NUREG-43 report
(pgs. 58 & 59),5 "The f ree-liqu id problem took most of the
discussion time in the Solidification Workshop held in
New Orleans, January12-14, 1977. Although free liquid
may appear in all the solidification systems (with the
exception of asphalt which boils the water away), nearly
all the complaints came from users of UF resins. Water is
an end product in the polymerization of UF resins, and
the amount varies according to the proportion of urea
and formaldehyde. The age of the reagent and the
pH RANGE
14
13
11
PH
RECOMMENDED
OPERATING
RANGE
Figure 6. Recommended pH values for VRS system
operation.
-------
87
temperature also have an effect . The broomstick
method used in several plants to detect free water was
admittedly inadequate, but no one has yet found a more
reliable method... The incorporation of spent resins into
cement requires careful control of the proportions of
solid and liquid in the mix to ensure adequate
mechanical strength ... the setting of cement may be
retarded or prevented by the presence of organics such
as oils and surface tension depresants."
The selection of the above excerpts is not meant to
imply that asphalt in the VRS process is perfect, but
rather to provide a basis for comparison regarding the
most common complaints with respect to use of non-
VRS systems. None of the complaints expressed in the
above excerpts are present in the VRS system or its end
product. The health physics relationship is discussed
below.
FREE WATER. There is no end product free water with
the VRS system. Water in the feed stream is boiled off
simultaneously with the mixing of the radsalt particles in
asphalt. Thus, the use of a "broomstick", or any other
method to detect free water, is completely eliminated in
the VRS system. The net result is no exposure to the
operator who otherwise would have to be at the
container to detect and, if found, drain free-water.
HOMOGENEOUS END PRODUCT. The mechanical, no
free-water mixing process of the extruder-evaporator
assures a homogeneous end product. This means no
"hot spots" due to clumps of waste or pockets of water.
LOW GAS EVOLUTION. Asphalt resists decomposition
due to internal radiation from the contained wastes up to
109 rads integrated dose. This is above the 107 rads
integrated dose normally experienced for nuclear power
plants. However, over the time period of storage and
burial, gas (hydrogen and/or methane) is generated
within all wastes, regardless of the binder used. Cement
and otherthermosetting binders form a rigid, nonflexible
solid that may contain water inside of a hermetically
sealed container. Too great a build-up of gases or
freezing of the water inside the sealed container could
cause the container to rupture and result in health
physics problems.
Asphalt, however, has no free-water and the lid on the
drum is not hermetically sealed, but rather crimped in
place. Also, asphalt is not a rigid, inflexible material.
These factors allow any gas generated inside the
container to migrate out of the asphalt and out of the
container with no internal pressure build-up.
NON-EXPLOSIVE MIX. Asphalt is an inert material.
Therefore, it is non-explosive by itself and because of its
inherent plasticity, resists external explosive forces. In
the weight percentage mixes used in the VRS process,
the end product has a negative oxygen balance.7
REDUCED DRUM CORROSION. Asphalt is an inert
material widely used for its water-proofing and
insulating qualities. Thus, the process of internal drum
corrosion is reduced significantly, and along with it the
health physics problems associated with exposure of the
contents to the environment.
ON-SITE STORAGE. The Atomic Industrial Forum has
recently published "On-site Low-Level Radwaste
ASPHALT THERMAL DATA
600 F IGNITION POINT
_ 550 F
FLASH POINT (min.)
ASTM TEST D-92
_ 410 F STEAM INLET TEMP.
300 F OPERATING TEMP.
A
ion 9noF
l_ 190 • zoo t-
SOFTENINGTEMP-
Figure 7. Thermal data of asphalt as used in the VRS
process.
Management Alternatives," ° a report prepared by NUS
Corporation. The generation rate of waste for a 1000
MWe nuclear plant was 26,300 cubic feet as solidified
with cement, and 5,200 cubic feet when volume reduced
and solidified. This report concludes that "Volume
reduction of trash and liquids result in savings in all
major cost areas: (a) capital expenditures for storage
structures, (b) annual operating costs, (c)
decommission cost, and (d) waste disposal cost." The
present-worth cost for an on-site engineered radwaste
storage (containerized storage with VR and
solidification) is $24,604,000, as compared with
$121,230,000 for shipment of solidified radwastes to
commercial shallow land burial facilities 500 miles away
and assuming 20% annual escalation in burial rates.
While the economic attraction of saving up to
$96,626,000 is outstanding, there are other benefits of
using a VRS system with on-site storage. These include
less exposure to plant personnel and the public, more
favorable public relations, and elimination of a crisis for
commercial burial facilities. VRS processing and above
ground' storage on site is the concept now in use at
Eurochemic, Mol, Belgium.9
PROCESSING AND TRANSPORTATION
SAFETY
Safety is a paramount consideration to utilities and
radwaste system operators. Safety encompasses several
factors such as flammability, explosivesness, and
radiation effects. All of these are discussed below.
Thermal data for asphalt as used in the extruder-
evaporator process is given in Figure 7. Note that the
flammability safety margin during processing is 300 F
-------
88
Because steam is used to provide process heat, the
chance for a hot spot to develop, which is possible with
electric heat, is zero.
The National Fire Code, published bytheNational Fire
Protection Association, rates asphalt as follows.
TABLE IV
ITEM
RATING
HEALTH
FLAMMABILITY
REACTIVITY
UNUSUAL REACTIVITY
Zero (Safe and non-
hazardous)
One (Requires significant
preheating before ignition
can occur)
Zero (Safe and non-
hazardous)
None
Combined with radwastes, both in the extruder-
evaporator and in the end product, the safety
characteristics are generally enhanced. For example, the
presence of borates, sulfates and other solids raises the
ignition temperature (see Table V).
The only place where ignition and sustained burning is
of real concern is in the event of a serious accident
during transportation to the burial site. The potential
hazard of that event has been evaluated by the technical
staff of the Nuclear Regulatory Commission and
published in their acceptance of the Topical Report" on
the use of asphalt in the extruder-evaporator process. To
quote their evaluation, "On the basis of our Conservative
evaluation of the radiological consequences of a severe
accident to the proposed system, we conclude that the
radwaste volume reduction and solidification (asphalt)
system is acceptable."
Another aspect of safety is the effect of an explosive
force, which was mentioned briefly earlier in this paper.
Asphalt is not explosive either by itself or when mixed
with radsalts, nor will an external explosive force cause
any damage of any consequence. That fact has been
proven in Europe by actual testing sponsored by
Eurochemic and conducted by the Royal Military School
in Brussels, Belgium. Brittle materials such as cement,
urea formaldehyde and polyester, will shatter under
explosive forces. However, asphalt normally has
sufficient plasticity that it will not shatter, though some
deformity may occur.
NET RESULT AT THE BURIAL SITE
The reduction in radwaste volume and the
corresponding reduction in the number of containers
shipped to the burial site means that the site can
accommodate more curies of waste. The lifetime of the
burial site is increased five fold or more if all buried
wastes are processed in a VRS system.
The container with asphalt encapsulated wastes
means that the external drum corrosion is the limiting
factor for container integrity. Previously the wastes were
acidic and the drums also corroded internally.10" While
no credit is given to drum integrity at this time, the
corrosion inhibiting factor of asphalt in the VRS process
is an added safety feature. Drums buried with asphalt
encapsulated wastes may last up to 15 years.
However, in the event that the drum does corrode, the
leach resistance of the non-hydroscopic binder will
retain the radwaste in place for an increased amount of
time. Both Brookhaven National Laboratory and the
Hanford Engineering Development Laboratory '213 have
shown the benefits of leach resistance on the retention of
the various radionuclides.
TABLE V
ASPHALT
IN PROCESS
CONDITION
ADVANTAGE
WATER IN FEED
AS WATER EVAPORATES,
SOLIDS TAKE ITS
PLACE
VRS OPERATING TEMP. 300F
STEAM INLET TEMP. 410F
IGNITION TEMP. 600F
WATER ISA FIRE
EXTINGUISHER, THUS
INHIBITS IGNITION
SOLIDS GENERALLY
RAISE IGNITION
TEMPERATURE
300F SAFETY MARGIN
IS MAINTAINED
DURING PROCESSING
IN THE END PRODUCT
PRESENCE OF BORATES
SULFATES & OTHER SOLIDS
PRODUCT DISCHARGE TEMP.
IS BELOW PROCESS TEMP.
SOLIDIFIED END PRODUCT
STORED AT AMBIENT TEMP.
ASPHALT IS INERT
IGNITION TEMPERATURE
RAISED
MORE THAN 300F
SAFETY MARGIN
MORE THAN 500F
SAFETY MARGIN
SIGNIFICANTLY REDUCED
ENVIRONMENTAL IMPACT &
CONTAINER CORROSION
-------
89
THE HUMAN FACTOR
The TV camera system is used by the operator to
observe the processing area and the container handling
operation. However, one or more lead glass viewing
windows are recommended to enable the operator to
more naturally observe the operations. The window(s)
permit the operator to have a 3-D plus a near natural
color view of his operations. Hopefully this combination
will enhance his interest and lessen the chances for
error.
Many features that are an integral part of the system,
such as spray nozzles inside the tanks, flush water
connections in transfer and metering lines, steam dome
clean-out provisions, etc., are designed to meet ALARA
requirements of NRC Regulatory Guide 8.8.
SUMMARY
The use of VRS and the resultant reduction in waste
handling and transportation produces many benefits
that are of prime importance to the health physicist. The
operators still handle the same number of curies but
performs significantly fewer operations. This reduction
carries over to less exposure time and fewer man-rems.
The reduction in waste handling and transportation
using the volume reduction and solidification system
with asphalt has the following advantages.
• Removes unwanted water to reduce volume.
• Uses proven binder in a reliable process.
• Elimination of free-water inhibits drum corrosion,
permits longer on-site storage.
• Reduces transport trips and lessens accident
probability.
• End product is retrievable and/or reprocessible.
• Process and end product accepted by NRC, AMI, EPA,
and burial sites.
• Uses system/equipment isolation approach in design
to reduce man-rems.
• Economically most attractive system.
All in all the Europeans have researched, developed
and proven this system over the past two decades. We
have further improved the VRS system to meet the
stringent requirements so vital to health physicists.
At the present time no credit is given to leach
resistance, however, the industry is investigating
methods to make solidified end products better. Several
10
-4 _
10
-5
10
-6
TAP WATER
(CEMENT/SLUDGE)
DEMINERALIZED WATER
(ASPHALT/SLUDGE)
=— ~
SEA WATER
^(ASPHALT/SLUDGE)
TAP WATER
(ASPHALT/SLUDGE)
i i i i i i i i i i
5 15 30 45
TIME-DAYS
60 70
Figure 8. Leach rate test results from AERE, Harwell,
England.
establishments in Europe have extensively investigated
the leach characteristics of cement and asphalt, some of
which are shown in Figure 8 and Table V. This is one of
the reasons why asphalt encapsulated wastes were
adopted so early in Europe and are gaining in popularity
here in the U.S.
A general observation is that the leach resistance of
asphalt bound wastes is 100 times better than that of
cement in the end product.14 Whenever the leach
resistance needs to be improved, there are methods
available. Hydroscopic salts, e.g., sodium sulfate, are the
worst for leaching in either cement or asphalt. A thin
layer (5 mm) of pure asphalt on the outside of the end
product or the inside of the container will virtually
eliminate leaching. Another method of improving leach
resistance is to add small quantities of calcium
hydroxide in place of or in addition to sodium hydroxide
for pH adjustment of the radwaste feed.
-------
REFERENCES
1. W. Kluger, H. Krause, O. Nentwich, FIXING OF
RADIOACTIVE RESIDUES IN BITUMEN, August
1969, presented at "Research Coordination
Meeting on the Incorporation of Radioactive
Wastes in Bitumen," Dec. 9-13, 1968.
2. W. Hild, W. Kluger, H. Krause, BITUMINIZATION
OF RADIOACTIVE WASTES AT THE NUCLEAR
RESEARCH CENTER KARLSRUHE - EXPERI-
ENCES FROM PLANT OPERATION AND
DEVELOPMENT WORK, KFK-2328 PWA-Nr
44/76, presented at NEA Seminar on the
Bituminization of Low and Medium Level Radio-
active Wastes, Antwerp, Belgium, May 1976.
3. W.F.Holcomb, S.M. Goldberg, AVAILABLE
METHODS OF SOLIDIFICATION FOR LOW-
LEVEL RADIOACTIVE WASTES IN THE UNITED
STATES, Technical Note ORP/TAD-76-4,
Technology Assessment Division, Office of Radio-
active Programs, U.S. Environmental Protection
Agency, Washington, D.C., Dec. 1976.
4. NRC TOPICAL REPORT EVALUATION AND
PROPERTY & LIABILITY INSURER'S REVIEW,
reprinted by Werner & Pfleiderer Corporation,
1978.
5. A.H. Kibbey, H.W. Godbee, E.L. Compere, A
REVIEW OF SOLID RADIOACTIVE WASTE
PRACTICES IN LIGHT-WATER COOLED
NUCLEAR POWER PLANTS, NUREG/CR-0144,
ORNL/NUREG-43 (Revision of ORNL-4942), Oak
Ridge National Laboratory, Oct. 1978. .
6. G. Meier, W. Bahr, THE INCORPORATION OF
RADIOACTIVE WASTES INTO BITUMEN, PART
1, KFK 2104, Gesellschaft fur Kernforschung
M.B.H., Karlsruhe, West Germany, April, 1975.
7. E. Backof, W. Diepold, STUDY OF THE THERMAL
AND MECHANICAL SENSITIVITY OF BITUMEN/
OXYGEN SALT MIXTURES, Translated by Ralf
Friese, KFK-tr-450, July 1975, Gesellschaft fur
Kernforschung M.B.H., Karlsruhe, West Germany.
8. On-SITE LOW-LEVEL RADWASTE MANAGE-
MENT ALTERNATIVES, prepared for Atomic
Industrial Forum by R.A. Martineit, et.al., of NUS,
draft report dated Dec. 8, 1978.
9. Personal communication with Dr. W. Hild,
Eurochemic, Mol, Belgium.
10. P. Colombo, R.M. Nielson Jr., PROPERTIES OF
RADIOACTIVE WASTES AND WASTE
CONTAINERS, BNL-NUREG-50617, Quarterly
Progress Report, July-Sept. 1976, Published Jan.
1977, Brookhaven National laboratory.
11. P. Colombo, R.M. Nielson Jr., PROPERTIES OF
RADIOACTIVE WASTES AND WASTE
CONTAINERS, BNL-NUREG-50837, Progress
Report No. 7, Oct.-Dec. 1977, Published May 1978,
Brookhaven National Laboratory.
12. R.E. Lerch, C.R. Allen, DIVISION OF WASTE
MANAGEMENT PROGRAMS PROGRESS
REPORT, July-Dec. 1977, HEDL-TME 78-48 UC-
70, published July 1978, Hanford Engineering
Development Laboratory.
13. R.E. Lerch, DIVISION OF WASTE MANAGE-
MENT, PRODUCTION AND REPROCESSING
PROGRAMS PROGRESS REPORT, HEDL-TME
77-74 UC-70, Jan.-June 1977, published July 1977,
Hanford Engineering Development Laboratory.
14. W. Bahr, W. Hild, W. Kluger, BITUMINIZATION OF
RADIOACTIVE WASTES AT THE NUCLEAR
RESEARCH CENTER KARLSRUHE, KFK 2119,
Karlsruhe, West Germany, presented at ANS
Winter Meeting, Oct. 27-31, 1974.
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91
SOLIDIFICATION OF LOW-LEVEL RADIOACTIVE LIQUID WASTE
USING A CEMENT-SILICATE PROCESS
Rodger W. Granlund
Pennsylvania State University, Health Physics Office, University Park, Pa.
and
John F. Hayes
Delaware Custom Materiel, State College, Pa.
Abstract
Extensive use has been made of silicate and Portland cement for the
solidification of industrial waste and recently this method has been suc-
cessfully used to solidify a variety of low level radioactive wastes. The
types of wastes processed to date include fuel fabrication sludges, power
reactor waste, decontamination solution, and university laboratory waste.
The cement-silicate process produces a stable solid with a minimal increase
in volume and the chemicals are relatively inexpensive and readily avail-
able. The method is adaptable to either batch or continuous processing and
the equipment is simple. The solid has leaching characteristics similar to
or better than plain Portland cement mixtures and the leaching can be fur-
ther reduced by the use of ion-exchange additives. The cement-silicate
process has been used to solidify waste containing high levels of boric acid,
oils, and organic solvents. The experience of handling the various types
of liquid waste with a cement-silicate system is described.
Introduction
The cement-silicate solidification process involves incorporation of
waste liquids and sludges in a silicate matrix using Portland cement as a
setting agent. This paper describes the use of the Chemfix Process devel-
oped by Conner (Co74a) and used by the Chemfix Corporation to solidify many
millions of gallons of industrial waste liquids and sludges. The process
was tested with several types of nuclear waste, including fission product
reactor waste, plutonium fabrication waste, and radium contaminated feed
plant raffinates, in the period 1969-1973 but it was not used until 1975,
when the patent was licensed to Delaware Custom Materiel, Inc. Since then
the process has been used to treat over a million gallons of low-level
liquid radioactive waste generated in nuclear power plants, fuel fabrica-
tion plants, research institutions and hot cell operations. Cement-silicate
systems are planned for installation in a number of nuclear reactors now
under construction or being planned (Nu76), but there are no in-plant sys-
tems now in operation. The cement-silicate process produces a stable solid
at low cost with simple equipment and shows great promise for the treatment
o.f low-level radioactive waste.
The Cement-Silicate Process
The cement-silicate process might be better described as a silicate-
cement or simply a silicate process. According to Conner (Co74b) the basic
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92
reaction is between soluble silicates and polyvalent metal ions. The com-
pound formed is thought to be based on tetrahedrally coordinated silicon
atoms alternating with oxygen atoms in a linear chain, with the metallic
ions producing crosslinking between the oxygen side groups. The structure
is similar to the naturally occurring pyroxene minerals. The silicate is
usually added to the waste as an aqeous solution of sodium silicate, but
other alkali silicates in either liquid or dry form can be used. The poly-
valent metal ions come from the waste solution, an added setting agent, or
both. The setting agent should have a low solubility,but a large reserve
capacity of metallic ions, so that it controls the reaction rate. Portland
cement is usually chosen as the setting agent because of the ready availa-
bility and the additional solidification reactions that it provides as a
cement. Lime, gypsum, calcium carbonate and other compounds containing
aluminum, iron, magnesium, nickel, copper, chromium or maganese are also
reported to be suitable setting agents.
The gel-like structure which is rapidly formed in the reaction entrains
large amounts of water or other liquids and prevents the settling of solids
during the final setting and hardening process. Reactions such as hydration,
hydrolysis, and neutralization also occur between the setting agent and the
waste to produce the final product. The process occurs at ambient tempera-
ture and pressure and can be used in either a batch or continuous mode.
The amounts of silicate and setting agent necessary to solidify a given
volume of liquid waste vary according to the composition of the waste, the
desired reaction time, and the preferred characteristics of the solid. Neu-
tral aqeous wastes with high concentrations of metal ions are easily soli-
fied with small amounts of silicate and setting agent. Highly acid or alka-
line wastes and those containing organic materials may require 2-4 times as
much material to produce an acceptable solid. Larger amounts of silicate
and setting agent form the gel at a faster rate and produce a harder solid,
although too much of either reagent can degrade the quality of the product.
The proportions may usually be varied for a given waste to produce a mate-
rial that will gel in a minute or less for batch processing in drums or a
mixture that will flow as a viscous liquid for an hour or more for addition
to large lagoons or diked areas with continuous processing equipment. A
typical mixture for one 1. of neutral pH, aqeous waste would include 80 ml.
of sodium silicate solution(density 1.4 g/ml.) and 240 g of Portland cement
and would produce 1160 ml. of solid. About 165 1. of this waste could be
solidified in the ordinary 200 1. steel drum and still leave about 10 cm of
free space above the solid. The weight of the solid would be about 223 kg.
The solid which is formed in the cement-silicate process varies from a
moist, clay-like material to a hard, dry solid similar in appearance to
concrete. The density varies from about 1.2 to 1.4 g/ml. The large amount
of water contained in the solid is not all chemically bound. In open air
the solid will dry out and lose water, with some shrinkage, but in a sealed
container the solid does not change. When buried without a container it
comes to some equilibrium moisture content with the surrounding soil. The
moisture loss is not an important factor for sealed containers of most
waste, but could be a problem with waste containing tritiated water.
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93
The solid is similar to concrete in that it is not very corrosive to
steel. A drum of solidified water was opened after being stored at room
temperature for 20 months. The solid still had the same appearance and
hardness as when the drum was sealed. The drum, which had a gross weight
of 675 pounds, was dropped from a height of 122 cm onto both ends to deter-
mine if it would still meet the requirements for a 7A container. There were
no visible openings in the drum after the test and there was no loss of con-
tents. The drum was then cut away from the solid and examined. The interior
of the drum was corrosion free, except for minor rusting at several points
where moisture had probably condensed on the lid immediately after solid-
ification.
The increase in volume of the solid, as compared to the waste liquid,
is a very important consideration in any solidification process. As the
weight and volume of additives to the waste are decreased, the cost for
containers, transportation, and burial is reduced and less burial space is
required. In the example illustrated above for water the solid has a volume
increase of 16% and a weight increase of 35% over the liquid waste. Using
only Portland cement to solidify the water (2 kg/1.) the volume increase
would be 66% and the weight increase 300%.
Leaching Studies
One of the important characteristics of a solidified waste material is
the susceptibility to loss of the incorporated radioactive material in
ground leachate. A standard method has been proposed by the IAEA (He71)
for leach testing solidified radioactive waste. However, this method re-
quires a long curing time and periodic sampling over a leach time of many
months. It is a good test for the intercomparison of standard samples, but
it is not practical for testing a large number of samples to determine the
optimum reagent proportions and operating conditions for solidifying a given
type of waste.
The following sensitive, yet relatively rapid, leach test was used to
compare various wastes solidified with the cement-silicate process. A
25-100 ml. sample of the waste was solidified and stored in a sealed con-
tainer for several days. The solid was then chopped or ground into small
pieces with maximum dimensions of 1-3 mm. A volume of 20-25 ml of the mate-
rial was weighed and then lightly packed into the barrel of a 30 ml syringe
(inside diameter 21.4mm) between glass wool plugs. The leaching solution,
which was introduced into the syringe barrel through a tube in a one-hole
stopper, passed through the sample and out through a hypodermic needle
fastened to the tipe of the syringe. The leaching solution was fed from an
aspirator bottle so that the sample was immersed at all times. The flow
rate was limited to about 1 ml/min. by crimping the hypodermic needle. Some
of the samples had a clay-like consistency and the maximum flow rate achiev-
able was less than 1 ml/min, but the slower rate did not appear to effect
the amount of material leached from the sample. Unless otherwise indicated
the leaching solution was deionized water.
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94
Approximately one 1. of leachate was collected from each sample. Tests
with a 137Cs tracer showed that essentially all of the readily leached mate-
rial would be removed by this volume of leachate. The leach rate of more
tightly bound materials was found to be a relatively constant value after
collecfilon of less than one 1. of leachate. The findings agree with pre-
vious studies by Connor (Co74b) on the leaching of polyvalent metal ions
from solidified industrial wastes.
Table 1 shows the results of leach tests conducted on samples of several
fuel fabrication plant sludges solidified with the cement-silicate process.
The uranium analyses were done using the delayed neutron activation analysis
technique or the etched fission track method. The other constituents were
determined by conventional water analysis techniques at commercial labora-
tories. The concentration of uranium and other metal ions in the leachates
was quite low. Some nitrate ion is leached from the solid, as expected,
because it is not chemically bound. The low value for chemical oxygen de-
mand (C.O.D.) in the leachate shows that the oils and other organic materials
are not readily leached from the cement-silicate solid. This result is some-
what surprising, but was also reported by Conner (Co74b) for industrial
wastes.
Table 1
Concentration in ppm.
Waste Description Analysis Waste Leachate
Fuel fabrication plant U 9.4 lxlO~
sludge, 34% O.D.S. Zr 12,000 <5
F 340 2.4
NO ^ as N 13,660 95
Fabrication plant sludge 235U 78 0.005
contaminated with C.O.D. 36,480 95
organic solvents, 21% O.D.S. Fe 1,640 < 0.01
Zn 158 < 0.01
Fuel fabrication plant U 7.1 < 0.003
sludge, 32% O.D.S. Ni 51 < .03
Pb 23 .04
Unlike polyvalent metal ions, monovalent ions are not chemically bound
in the silicate reaction and are readily leached from the solid. This has
been demonstrated using the leaching test described above with a 137Cs
tracer. In an effort to improve the cesium retention a number of additives
to the cement-silicate process were tested. The addition of shale to the
mixture greatly decreased the amount of cesium leached from the solid. Ex-
cellent results were obtained with Conasauga shale from Oak Ridge, Tennes-
see and from unidentified shale samples obtained in the vicinity of Port
Matilda, Huntington, Johnsonburg, and Bradford in Pennsylvania. Tests were
also made with sodium bentonite (Wyoming or western bentonite), calcium
bentonite (southern bentonite), and three different illite clays. The shales
and clays were ground fine enough to pass a U.S. .f40 sieve and added to the
waste mixture with the setting agent. The samples were made by solidifying
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95
a 5% N32S04 solution tagged with 137Cs, using the equivalent of 400g of
Portland cement, 200 ml. of sodium silicate solution, and 160 g of shale
or clay per 1. of solution. Samples were also prepared using only Portland
cement for solidification in the proportion of 2 kg per 1. of solution. The
relative leachability is expressed as the specific leach fraction, which is
the concentration of Cs in the leachate (yCi/g) divided by the concentra-
tion of 1;"Cs in the solid (yCi/g).
The results of the leach tests with the shale and clay additives are
shown in figure 1. The Conasauga shale additive produced a marked reduc-
tion in cesium leaching in both the cement-silicate and the cement only
solids. However, it was much more effective in the cement-silicate process
than with cement only. Both bentonite clays produced about the same reduc-
tion, but the illite clays showed a wide variation. The Beavers Bend, Okla-
homa sample showed the greatest reduction, the Goose Lake, Illinois sample
the least reduction, and the Fithian illite from Illinois was intermediate.
Samples in which ground glass replaced the shale or clay, to determine the
effect of the bulk and surface area of the additive, did not show any reduc-
tion in cesium leaching. Tests were also made with the Conasauga shale
additive in solids produced with plain water and with a simulated waste, W-7,
used in some leach studies of grouts at the Oak Ridge National Laboratory
(Mo76). The W-7 waste is highly alkaline and contains large concentrations
of sodium, nitrate, carbonate, and sulfate ions. The specific leach frac-
tion for the solid made with water was 2.6x10"^ and for the solid from the
W-7 solution 1.9x10"->, or almost the same as the average value of 1.5x10-5
obtained with the 5% Na2S04 solution. The amount of cement and silicate
used for the water solidification was about half that for the 5% Na2S04 and
for the W-7 solution about 5% more than for the 5% Na2SO^ solution.
The solid produced by the cement-silicate process has been suggested as
a liner and covering for sanitary landfills to remove metal ions and other
materials from any leachate passing through it (C674c). This technique
could also be used at radioactive waste burial grounds. A sample of cement-
silicate solid prepared using water and shale additive was used to filter
137Cs tagged water and 137Cs tagged leachate from a cement-silicate solid
without the shale. In both cases essentially all (99.98%) of the 137Cs was
removed by the cement-silicate filter with the shale additive. Thus the
presence of waste solidified by the cement-silicate process could help pre-
vent the release of materials leached from other wastes.
Field Experience
The cement-silicate process has been used to solidify many types of nu-
clear waste. One application involved about 330,000^1. of evaporator con-
centrates containing fission products in the 1-10 mC.i/1.range from decon-
tamination and hot cell operations. The mixing was done in concrete dispos-
al containers of up to 6000 1. capacity. Processing was completed in less
than one week, using field-assembled equipment and working outdoors. Even
though working conditions were adverse and the equipment less than optimum
the project was completed without significant radiation exposure or contam-
ination problems. A number of the concrete containers were not water tight
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and liquid started to seep through them after filling. However, the seepage
stopped almost immediately after the waste was solidified. The open top
containers were allowed to stand uncovered for many months before moving to
the disposal site. During this period there was some drying and shrinking
of the solid allowing precipitation to get between the container wall and
the solid. Some of this contaminated liquid leaked through the walls and
caused minor contamination.
Continuous processing at rates of 700-1100 l./min. has been used in the
treatment of sludges contaminated with very low concentrations of enriched
uranium. The treated sludges were pumped to a diked area for solidification
and storage on-site. The solid was firm enough to be walked upon in about
twenty four hours and was later covered with earth to prevent infiltration
of rain water. Such sludges are usually processed so as to produce a solid
that can be handled with convential earth moving equipment. The material
can then be easily moved at some future date, if necessary.
The nuclear reactor waste that has been processed to date has been mixed
in open 200 1. steel drums using a portable mixer. The volume of waste which
can be mixed in a drum depends upon the amount of silicate and setting agent
required, but is usually 145-165 1. A large amount of the reactor waste has
contained oils from the turbine and other sources. The oil concentration is
adjusted to 30-50% of the liquid volume, using other aqeous waste when pos-
sible, and the oil is emulsified with 4-6 1. of detergent prior to solidifi-
cation. Boric acid waste is neutralized with lime or sodium hydroxide before
solidification. Lime is preferred because it also acts as a setting agent.
Low concentrations of acid do not require pretreatment because of the excess
alkalinity in the cement. Dewatered or slurried reactor resins are easily
solidified with the cement-silicate process. The small amount of solidifi-
cation chemicals used results in a volume increase of 10% or less. However,
a relatively powerful mixer is required to adequately mix the resins with
the solidification agents. Most of the reactor waste has been material which
could not be processed with the station waste treatment system. Processing
has been done in whatever space available, even outdoors, using portable
equipment. A two man crew can solidify about 80 drums of waste per day.
Batch processing in 200 1. drums has also been used to solidify a large
quantity of process waste at a high-enrichment uranium fabrication facility.
This waste contained a large proportion of organic solvents but was success-
fully solidified using high ratios of silicate and setting agents. Research
laboratory waste containing 20% or more of liquid scintillation fluid and
various other organic solvents has been routinely solidified. A volume of
150-170 1. of waste is usually processed in each drum. The waste collected
from the various laboratories is combined in a 200 1. drum and test mixes
are tried on 25 ml samples. Because of the varying composition of wastes
received from the research laboratories, each drum is tested to find the op-
timum proportions of silicate and setting agent. The processing is easily
done by one or two persons using simple, inexpensive equipment and the direct
cost is less than that of previous methods using only absorbents, which are
no longer acceptable at licensed landfills.
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When processing waste that is difficult to solidify, it is not uncommon
to have a small volume of liquid that is not incorporated in the solid. If
this occurs it is easily observed, because the liquid collects at the top
of the container. The solid formed under the liquid is of good quality and
can be used. Small amounts of excess liquid (1-2% of the waste volume) are
usually taken up by the solid as the hydration and setting of the cement
progresses. Larger volumes may be decanted to another container for further
processing or dry cement can be added to the container to incorporate the
liquid residue.
Equipment Requirements
Mobile processing units on semi-trailer chassis have been used by the
Chemfix Corporation to solidify industrial waste in a continuous mode at
rates of about 1100 l./min. The units contain storage tanks for silicate
solution and cement plus a hopper with an agitator for mixing the liquid
waste with cement. The waste is pumped to the van, mixed with cement in the
hopper then pumped to a receiving lagoon. The silicate is injected into the
suction of the discharge pump so that it is mixed with the waste just before
discharge. These high flow rate units have been used to solidify large
volumes of sludge contaminated with uranium, but the units are not designed
to handle waste with a high concentration of radioactive material. There
are no provisions for shielding or complete containment of the waste stream.
However, the cement-silicate process involves only mixing and transfer of
materials at ambient temperature and pressure. Conventional off-the-shelf
hardware could be used to build a continuous process system for use in a
confined shielded area.
Batch processing is readily accomplished within a 200 1. drum. The drum
is filled with the liquid waste and mixed while the cement is added. Sili-
cate solution is then added quickly and the mixer is operated for a few sec-
onds more, before removing it from the drum. A 1/3-3/4 hp. mixer is suffi-
cient for liquid waste, but heavy sludges or resin beads require a mixer of
2-3 hp. Air-powered, gear-reduction mixer motors are ideal for this purpose
because of the light weight, variable speed, and the ability to be stalled
without damage. A small air mixer mounted on the lid of a 200 1. drum along
with a funnel for cement and a feed pipe for silicate solution addition has
operated very well for processing research laboratory waste. A large amount
of waste has also been processed in open drums, adding the cement and sili-
cate by hand and mixing with a clamp-on drum mixer.
A solidification kit for processing waste in a 200 1. drum has been de-
veloped by Delaware Custom Materiel, Inc. The kit consists of a drum con-
taining a disposable mixer blade, with the shaft held by bearings welded to
the inside of the lid and bottom of the drum. The upper end of the shaft
is accessible through a bung in the lid for turning with an external motor.
The cement can be added to the drum before it is capped. The liquid waste
and silicate are added through the bungs in the lid. This technique reduces
the possibility of spreading radioactive contamination during the filling and
mixing operation. An air-driven motor is clamped to the drum lid to turn the
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mixer.
Batch processing can be done in disposal containers much larger or
smaller than the 200 1. drum. In a previously mentioned project the drum
mixer was scaled up and used to mix single batches in 3800-6600 1. concrete
tanks. At the other end of the scale, prepared kits containing sufficient
silicate and cement to solidify 500 ml. of waste are being used by the radi-
oisotope users at a university. The materials are packaged in a one 1, steel
can, which is the mixing vessel and the disposal container. The waste is
added to the cement in the can and diluted to 500 ml. Then the silicate is
added and the can is capped and shaken for a few seconds. The waste is sol-
idified in less than a minute and discarded with the other solid radioactive
waste. This procedure is used only for small volumes of waste containing
relatively large amounts of radioactive material. The more hazardous liquid
waste is immediately fixed in solid form. This keeps the concentration in
the large volumes of low-activity liquid waste at the level where handling
it is not difficult.
Summary
The cement-silicate process involves the interaction of soluble sili-
cates and a setting agent with liquid waste to produce a solid material
suitable for disposal. The silicate reaction chemically binds polyvalent
metal ions so that the ions are not readily leached from the solid and shale
additives can be used to reduce the leachability of radioactive cesium.
The chemicals used in the process are inexpensive and readily available.
The volume increase of the waste in the solidification process is small re-
sulting in further savings of container, transportation, and burial costs,
as compared to other solidification processes. The reaction rate and the
chemical and physical properties of the solid can be controlled to allow
processing by continuous or batch methods and to suit the disposal site or
container.
The process has been used to solidify many types of low-level radioac-
tive waste including those which contain oils and other organic materials.
Processing equipment is simple and can be built with readily available
hardware, making the process suitable for low-volume and high-volume appli-
cations.
References
Co74a Conner, J.R., 1974, "Method of Making Waste Non-Polluting and Dispos-
able", U.S. Patent No. 3, 837, 872.
Co74b Conner,J.R., 1974 "Ultimate Disposal of Liquid Wastes by Chemical
Fixation", Proceedings of the 29th Annual Purdue Industrial Waste Con-
ference, Purdue University, West Lafayette, Indiana.
Co74c Conner, J. R., Polosky.R,J., "Method of Improving the Quality of
Leachate from Sanitary Landfills", U.S. Patent Nos. 3, 841, 102.
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Cu75 Curtiss, D.H., Heacock, H.W., 1975, "Radwaste Disposal by Incor-
poration in a Matrix", U.S. Patent No. 3, 988, 258.
He 71 Hespe, E.D., "Leach Testing of Immobolized Radioactive Waste Solids",
Atomic Energy Review, £, 195.
Mo 76 Moore, J.G., Godbee, H.W., Kibbey, A.H., Joy, D.S., 1975, "Devel-
opment of Cementious Grouts for the Incorporation of Radioactive Wastes.
Part 1: Leach Studies", ORNL-4962.
Nu76 Nucleonics Week, 17, No. 28, p. 8, July 8, 1976.
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FIGURE 1
7 LEACH TESTS FOR
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WITH CEMENT AND CEMENT SILICA?E
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COLLECTION AND DISPOSAL OF LOW LEVEL WASTE
AT AN EDUCATIONAL INSTITUTION
Dale L. Andrews, J. R. Gilchrist, and H. W. Berk,
Radiation Safety Office, University of Virginia,
Charlottesville, Virginia 22903
Abstract: Low level radioactive wastes are generated by a number of differ-
ent laboratories and departments at the University of Virginia. Radioactive
materials are utilized in a variety of research applications including
medical and basic sciences, as well as for diagnostic and therapeutic uses
at the University Hospital. Radioisotopes are purchased from commercial
sources and are produced locally for use in research and medical diagnosis
and treatment by the University of Virginia Reactor. In 1974, the University
Radiation Safety Committee adopted rules for discharging radioisotopes to the
environment which are more restrictive than the Nuclear Regulatory Commission
regulations. The committee's philosophy is that no radioactive substances
should be discharged to the environment which can be reasonably avoided, in-
cluding those used in medical diagnosis and therapy. This policy has caused
a significant increase in the accumulation of low-level radioactive wastes.
The volume of low-level wastes at the University has increased from about
1.5 M in 1969 to over 68 M in 1977. Disposal costs have increased pro-
portionately. Currently the University employs a full-time technician to
collect and package radioactive wastes under the supervision of the health
physics staff of the Radiation Safety Office. In 1976, the Radioactive
Waste Management Facility (RWMF) was completed. This facility houses the
Radiation Safety Office staff and has modern facilities for collecting and
packaging all types of radioactive wastes. The facility is being used to
limit the total cost of radioactive waste disposal, while fulfilling the
objectives of the Radiation Safety Committee. Methods used to limit waste
disposal volumes and costs are compaction, storage and decay of short half-
life isotopes, solidification of liquid wastes, and education and training of
radioactive material users throughout the University in reducing waste volume.
Introduction
The University of Virginia holds a Nuclear Regulatory Commission (NRC),
Type "A" Broad By-Product Material License, an NRC "Institutional" Reactor
Operations License, and a Virginia State License for ionizing radiation
producing equipment and radioactive material not regulated by the NRC. The
Radiation Safety Committee has administrative control over all uses of radi-
ation and radioactive materials at the University including medical therapy
and diagnosis. The Radiation Safety Committee is so structured that the
Radioactive Drug Research Committee, required by Food and Drug Administration
regulations also functions as the Medical Isotope Subcommittee, an advisory
body on medical uses of radiation in diagnosis and therapy.
The University Radiation Safety Officer is a member of the general
faculty and is appointed by the Chairman of The Radiation Safety Committee.
The Radiation Safety Officer directs the operational unit involved in radio-
active waste management, and implements the rules and regulations promulgated
by the Radiation Safety Committee for the safe use of all lonizxtig radiation
sources at the University.
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The Radiation Safety Office staff consists of professional health
physicists, technicians and clerical personnel for the routine operation
of the Radioactive Waste Management Facility and the University's radiation
safety program.
Facilities Generating Radioactive Wastes
The University is arbitrarily divided into three areas for waste manage-
ment purposes. These are the reactor facility, which houses a 2 Megawatt
swimming pool reactor, an 80 watt training reactor and experimental and
research laboratories; the academic departments, which include radioactive
materials and radiation use in the Van de Graff accelerator, the Physics
Department and the Departments of Biology, Chemistry, Materials Science and
Environmental Science; and, the Medical Center, including a medical school,
basic science research laboratories, clinical research laboratories and the
hospital which has active nuclear medicine and radiation therapy departments.
Of the above, the medical center generates by far the greatest quantity
of radioactive waste. Approximately 90% of all radioactive wastes at the
University originate within the medical center.
Development of Radioactive Waste Management
Prior to 1972, the Reactor Facility staff collected and disposed of their
own radioactive wastes and those generated by the academic departments, while
the Medical Center maintained a completely separate radioactive waste collect-
ion, storage and disposal operation. My mid-1972, the Radiation Safety
Office was established as a service department and a University-wide program
of waste collection, storage and disposal was begun. Initially, the co-
ordination was only administrative and the dual collection and storage op-
erations continued to function under the broad supervision of the Radiation
Safety Office. When a new Medical Education Building was put into service
in late 1972, a third collection and storage operation was organized to
accomodate more than 60 laboratories using radioactive materials in that
building.
In 1974, the Radiation Safety Committee revised the rules and regulat-
ions concerning discharge of radioactive materials. The revision stated that
no radioactive materials would be discharged to the sanitary sewer system if
it could be reasonably avoided. This regulation had a significant impact
on the quantity of radioactive wastes collected ( figure 1). By the end of
1974, it was apparent that the waste collection and storage operation was
costly and inefficient with the increased volumes of waste being handled,
due to both the increasing use of radioactive materials and to prohibition
of low level radioactive waste discharge to the sewer system. A sub-
committee of the Radiation Safety Committee was appointed to study the waste
disposal operation and make recommendations for improving the cost effective-
ness of waste handling.
Based on the recommendations of this Subcommittee, planning was begun
in late 1974 for a centralized radioactive waste management facility. Funds
were obtained through the University Research Policy Council and the facility
-------
Figure 1.
t/»
Of.
UJ
£
80
70
o
o
y so
3
V
Z 40
LU
| -
20
10
RADIOACTIVE WASTE VOLUME SHIPPED
FOR DISPOSAL BY CALENDAR YEAR
69 70 71 72 73 74 75 76 77 78
CALENDAR YEAR
-------
104
was completed in June 1976. A technician was added to the Radiation Safety
Office staff to assume primary responsibility for radioactive waste manage-
ment under the direction of the health physics staff. An additional secre-
tary was also added to help staff the new facility. The Radiation Safety
Office staff currently consists of three health physicists, one technician
and one full-time and one half-time secretary plus a paid student position
which is normally filled by a graduate student in the Nuclear Engineering
Department.
Radioactive Waste Management Facility
The 2500 square foot waste management building contains a small
laboratory used for environmental and bioassay sampling, a large waste
handling room, office space for two health physicists and clerical staff, and
a small emergency decontamination facility consisting of a shower, washer and
dryer and clothing supply lockers,, The shower, washing machine and floor
drains in the waste handling room and decontamination room are connected to
a hold up tank which is designed to hold contaminated water until it can
be sampled. A recirculation pump and demineralizer system is provided to re-
move radioactivity from the decontamination water prior to discharge to the
sewer system. An outside concrete pad and storage cabinet provide a storage
area for flammable liquid.
The large waste handling room contains a compactor to compact solid
wastes in 55 gallon barrels to obtain a more cost-effective package. The
compaction ratio for most waste collected is 4 to 1. A barrel handling
crane is provided so that one man can easily manipulate compacted drums in the
storage area. The room also has a fume hood which is used for opening pack-
ages of radioactive materials as they arrive at the University. All packages
containing radioactive materials entering or leaving the University are pro-
cessed through the waste facility.
The floor of the waste handling room is covered with an epoxy resin for
easy decontamination. An area monitoring system with local meters and remote
indication at the secretary's deck provide constant monitoring of the radia-
tion levels within the waste room.
Waste Packaging and Shipment Procedures
Radioactive wastes are picked up from the users' laboratory on request
and are brought to the waste facility for packaging. Each radioactive mat-
erial user is supplied with both solid and liquid waste containers. These
containers are not currently standardized and users are permitted to use
containers other than those provided if they are approved by the Radiation
Safety Officer. The collected wastes are segreated according to half-life
( Table 1) and are then packaged for shipment or set aside for decay. Short
half-life isotopes are held for approximately 10 half-lives and then monitor-
ed. If the radioactivity has decreased to background levels, the waste is
disposed of as routine non-radioactive waste material. Occasionally, when
waste volumes become prohibitively large, some short half-life material is,
by necessity, shipped for disposal.
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105
Typical Radioisotopes in Collected Wastes
TABLE 1
Long Half^Life* Short Half-Life*
3H - 12.2 years 24Na _ 15 hours
JC - 5600 years 32P _ 15 d
35S - 87.2 days 51Cr _ 27.7 d
45Ca- 163 days 99Mo_99Tc _ 66 hours
^Mn- 312 days 127xe- 36.4 days
Fe- 2.7 years 131j _ 3 days
57Co- 271 days 133Xe_ 5 days
Co- 5.2 years i^Ce- 32.2 days
'^Se- 120 days 201T1- 73 hours
125r~ 3° years
I- 60 days * Classified for Disposal Purposes
Solid wastes are placed directly into 55 gallon drums and compacted.
Liquid wastes are poured into 30 gallon drums filled with vermiculite, which
are then packed in 55 gallon overpack drums with the additional space filled
with vermiculite. Animals and biological tissues are packed in 30 gallon
drums and stored in a large walk-in freezer until shipment. Liquid scin-
tillation vials are left intact and are packed in 55 gallon drums with
alternate layers of vermiculite. All containers are currently being sup-
plied by Teledyne Isotopes of Westwood, New Jersey. The containers are
prepainted with the appropriate labeling and meet all regulations for ship-
ment. Of course, detailed records are kept of all waste collection and
disposal.
The use of radioactive materials continues to expand within the University
and the cost of waste disposal continues to rise ( figure 2). The current
method of waste collection and disposal seems to be the best approach to the
problem within the restrictions placed on radioactive material use by the
licensing agencies and the University's Radiation Safety Committee.
Objectives of the Waste Management Program
The objectives of the current waste management program are to provide
the most cost-effective and efficient radioactive waste disposal while
maintaining a good relationship between the University and the surrounding
community. By restricting the discharge of radioactive materials to a
greater extent than allowed by the NRC or State regulations, the University
is demonstrating its' commitment to the safe use of radioactive materials
within the community.
These objectives are being met by insuring that all radioactive wastes
are properly handled and contained, and by a continuing education and
information program for radioactive materials users, employees and staff of
the University, and members of the general public.
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Figure
PERCENT COST INCREASE PER UNIT VOLUME
FOR RADIOACTIVE WASTE DISPOSAL BY
CALENDAR YEAR
150
2
o
C
at
u
o
u
in
u
100
50
I I
69 70 71 72 73 74 75 76 77 78
CALENDAR YEAR
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107
COLLECTION AND HANDLING OF RADIOACTIVE WASTES
FROM A LARGE UNIVERSITY
Jamieson G. Shotts, David L. Spate, and Philip K. Lee. Health Physics Services,
University of Missouri, Columbia, Missouri
Abstract
Various types, levels, and amounts of radioactive wastes are generated by
the University of Missouri-Columbia research, teaching, and clinical programs.
Health Physics Services is responsible for collecting these wastes from the
more than 200 laboratories spread over the sprawling campus and disposing of
the wastes in an economical and safe manner. Most of the radioactive wastes
are laboratory paper trash and liquids containing low levels of activity but
some high level wastes and contaminated large animal carcasses also require
disposal. The wastes are stored at one of four interim storage locations
until ultimately disposed of by incineration, sewer release, local burial,
shipment for commercial waste disposal or decay. The economics, safety, and
handling aspects of the various disposal methods must be considered.
Discussion
Health Physics Services at the University of Missouri-Columbia (UMC) is
responsible for the disposal of all radioactive wastes produced by the Columbia
Campus. The radioactive wastes are generated by the teaching, research, and
clinical efforts of the basic sciences departments, the School of Medicine, the
University Hospital, the College of Veterinary Medicine, and the College of
Agriculture located on the sprawling campus. The waste handling responsibilities
of the group also extend to the UMC medical research laboratories in Saint
Louis which is 125 miles from the main campus.
The radioactive wastes generally are the normal solid and liquid radio-
active wastes from a basic science laboratory such as paper, gloves,
scintillation fluids, vials, and other disposables. In addition, wastes also
range from technetium milk from molybdeum cows to radioactive milk and manure
from dairy cows and even at times to the cows that produce it.
The types of waste generated by such diverse sources as those found on
the Columbia Campus require a flexible system of waste handling. The system
must be able to cope with wastes ranging from the low-level trace activities
in liquid scintillation fluids to multicurie activities produced from high
specific activity labeling procedures. In addition, the waste handling system
must be capable of contending with waste containers ranging from one milliliter
shipping vials through bags of trash and bottles of liquids to pony and cattle
carcasses.
The system must be able to absorb the expanding volumes of wastes resulting
from increasing numbers of authorized users and their radioisotope laboratories
(table 1). Growth of the radioisotope program of the campus is indicated by the
22% increase in the number of authorized radioisotope users over the past five
year period and even more directly by the 38% increase in the number of radio-
isotope laboratories. As would be assumed the number of waste pickups and
waste volumes correlate with the increasing number of authorized users (table 2).
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108
The annual number of waste pickups has increased about 43% over the past five
years while the number of containers or items collected increased by about 30%.
These increasing volumes are more dramatically indicated in the disposal
figures (table 3). Solid waste volume handled by the waste disposal system
increased by 104% over the last five years while liquid waste disposals
increased even more showing a 143% increase.
The increasing volumes of waste entering and leaving the radioactive waste
disposal system are indications of several subtle influences. The function of
a university is to train and educate students in the modern techniques needed
for the sophisticated research now prevalent. This training produces young
faculty members and research investigators trained in the use of radioactive
materials and in the methods of applying them to advance research interests.
These younger faculty members tend to rely more on research methods utilizing
radioactive materials than did their predecessors. Coupled with the research-
ers increasing use of radioactive materials is their ability, aided by the
improved instrumentation, to use the radioactive tracers more efficiently.
For example, we have noticed that investigators performing iodination of
proteins for radioimmunoassay procedures require decreasing activities of
iodine as their techniques improve. lodinations which initially took four or
five millicuries to complete may only require one-half to one millicurie to
provide the stock material for several months of assays.
The functions of the waste handling system can be separated into two
areas of responsibility: 1) the authorized user and 2) Health Physics Services.
The Authorized User
The authorized user, as the starting point in the waste flow process, is
the most critical in the waste disposal program. Only the user can predict
with any degree of certainity how the radioactive material he uses fractionates
among the types of waste that are generated. Solid waste is almost impossible
to assay and liquid waste is difficult to assay with any degree of confidence,
particularly for low energy beta emitters. Health Physics Services does not
have the time available to expend in assaying all the waste collected. It is
a requirement that the user be responsbile for recording the radionuclides and
activities in the wastes before the waste is removed from the laboratory.
Radioisotope laboratories are supplied with either of two sizes of plastic-
lined fiberboard containers lined with polyethylene bags for solid wastes and
with one gallon polyethylene jugs for liquid collections (table 4). Other
types of containers such as paint cans, small polyethylene containers and
steel drums are supplied as needed for specific handling, containment, or
volume problems. The user disposing of scintillation vials is instructed to
repackage them in their original containers for disposal, however, in some
cases the vials are collected in the fiberboard or steel drums.
It is the policy that Health Physics collects and processes all radioactive
waste. Sewer discharges from laboratories are discouraged except for some
specific instances involving large volumes of low-level wastes that are
difficult to collect and transfer. For washing of contaminated glassware,
the laboratory personnel are instructed to rinse the container into the liquid
waste jug until the remaining activity approaches background levels. The
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109
glassware then can be washed with no recorded activity to the sewer. These
procedures have reduced the sewer discharge for the Columbia Campus users
to less than ten millicuries per year excluding, of course, those radioactive
materials administered to humans for diagnostic or therapeutic procedures.
Animal carcasses containing radioactive material are held for disposal by
freezing or refrigeration. The carcasses are stored in freezers at the
authorized users' locations or transferred to Health Physics for storage in
the freezers in the waste storage facilities. The stored carcasses are wrapped
in plastic and labeled according to date, isotope, activity, authorized user,
or identification number.
Health Physics Services
Waste pickups are performed upon user request. The requests from the users
are made by telephone or by interoffice mail correspondence. The form of the
waste as to solid, liquid, or animal, isotopes and activites are identified at
the time of the request. Health Physics personnel can then arrive at the
laboratory with appropriate replacement containers. Health Physics has two
vehicles which are used for the radioactive waste collection, a van and a
station wagon. Pickups are usually made within one day and often within a few
hours of the request. At the laboratory isotopes and activities in each
containers are confirmed, the bag liners in the solid waste containers are
taped closed, and the waste transferred to the vehicle with a two wheel cart.
Upon reaching the radioactive waste storage building, waste information is
recorded on radioactive waste record cards (figure 1) and a sequential
identification number assigned to each container. The container is then tagged
with identification number, isotopes, activities, and date. Waste containers
are segregated according to form and half life in the radioactive waste storage
building.
The radioactive waste storage building is located next to the Health
Physics Services building in the Research Park (figure 2). The building
is a 728 square foot brick veneered concrete block building on a concrete slab
foundation. A flammable storage room occupies one corner. Scintillation vials
and other liquids are stored on steel shelving in this room. Room air is
exhausted at a rate of six air changes per hour by an explosion proof fan.
Two 20 cubic feet chest type freezers for small animal carcass storage are in
the large room. The remainder of the building can be used for storage of
solid wastes, empty containers, and usually one of the vehicles. The building
provides storage for about six months of waste accumulation at our current
collection rate.
Additional waste holding facilities are located at the Medical Center
where a 60 square feet storage room and two 20 cubic feet chest freezers are
available and at the University waste disposal site about five miles southwest
of the campus where a storage building with a waste compactor is available.
Ater the collected solid waste approaches a volume of about 200-500 cubic
feet of low-level materials, a burial is made at the local radioactive land
burial site. A backhoe is scheduled to dig an appropriately size burial trench,
which is generally twelve feet deep, two to four feet wide, and ten to fifteen
feet long. Burials are performed within the requirements of 10 CFR 20.304.
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no
Containers are checked for isotopes and activities and loaded for transport to
the waste disposal site. Identifying tags are removed from the containers as
the liners are emptied from the containers into the trench. Upon return to
the office the tags are cross checked with the waste disposal cards and
activities summed by isotope to assure compliance with 10 CFR 20.304. Just
before the trench is backfilled the frozen animal carcasses are transferred from
the freezers to the site, identifying tags removed, and the carcasses placed
in the trench. The burial is given an identifying number- located on the
burial site map, containers are recorded as buried on the waste disposal cards,
and the burial report completed.
An incineration amendment to the University NRG license allows open pit
incinerations. We have restricted individual incinerations to less than 100
gallons of liquid and less than one millicurie per liter specific activities.
No more than five incinerations per week and less than 52 per year are permitted,,
however, the frequency of incinerations has been much lower than this allowable
rate. Records for the incineration are prepared and maintained in a manner
similar to those for the shallow land burials.
Wastes with activities too large for disposal by local burial or incineration
are held for shipment to a commercial disposal firm. Isotopes with half lives
of less than 100 days are routinely held for decay to levels acceptable for
shallow land burial or incineration. This policy has resulted in an accumu-
lation of stored iodine-125 waste due to the recent increase in iodinations
for radioimmunoassays.
Safety margins inherent in the system are: No container is assigned an
activity of less than one microcurie, burials and incinerations tend to average
from 40 to 60% of the allowable 100%, and isotope decay is estimated somewhat
conservatively. Also all radioactive waste collected from laboratories is
considered to be radioactive and accordingly disposed. No radioactive waste is
disposed as normal waste even though the external radiation measurements may be
at background levels.
The waste disposal program is a major part of the health physics program
at the University of Missouri-Columbia. It is estimated that about one-fourth
of the effort of the six person group is devoted to waste disposal. Also
included are the costs for containers, local burial excavations, commercial
disposal charges and one-half of the operational costs of the two vehicles.
The waste disposal system is able to move radioactive wastes from a laboratory
or use area to an ultimate disposal in a safe economical manner while producing
a minimal environmental insult. It is felt that our simple and flexible waste
disposal system can cope with the various wastes and volumes produced by a
major research and teaching institution.
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Ill
TABLE 1
Authorized Users and Radioisotope Rooms by Fiscal Year
1973-74 1974-75 1975-76 1976-77 1977-78
Authorized Users 123 127 137 143 150
Radioisotope Rooms 167 174 215 218 230
TABLE 2
Radioactive Waste Totals by Fiscal Year
1973-74
282
231
736
No. of
Pickups
Containers
Solid
Gallons
Liquid
Scintillation
Vials (Gallons)*
Animals (Bags) -
Activity
*100 filled vials - 0.25 gallon
1974-75 1975-76 1976-77 1977-78
309
443
694
354
288
711
(299)
56
376
309
852
(355)
105
12,681
402
298
973
(333)
40
44,966
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112
TABLE 3
Radioactive Waste Disposed by Fiscal Year
1973-74 1974-75 1975-76 1976-77 1977-78
Local Burial
Burials 4 4 5 10 8
Volume (Ft3) 707 678 1028 1311 1510
Activity (mCi) 1046 1080 559 176 208
Incineration
Incinerations 10 9 11 12 14
Volume (gallons) 500 539 755 925 1213.5
Activity (mCi) 95 95 116 94 130
Sewer Disposal (mCi) 1.000 0.25 0.25 6.400 3.000
Commercial
Shipments 11211
Volume (Ft3) 56.8 58 43 15.5 45
Volume (Gallons) 45
Activity (mCi) 5,354 1,550 3,797 230 15,434
Cost 200.00 370.00 155.00 0 0
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113
TABLE 4
Item Source Cost
Plastic jugs University Storeroom $ 0.40 each
(1 gallon)
Plastic bags Home Plastics, Inc.
(6 mil) Des Moines, Iowa 0.42 each
Fiberboard drums Continental Can Co. 28 gal. 13.50
Inside plastic Overland, Missouri 18 gal. 7.50
Silk screened
-------
RADIOACTIVE WASTE RECORD
FIOURE 1
UNIVERSITY OF MISSOURI
in
CAMPUS Columbia
Pick up
No
1753
175A
1755
1756
1757
1758
1759
1760
Authorized
User
Johnson, H.
Johnson, H.
Johnson, H.
Johnson, D.
Johnson, H.
Johnson, H.
Johnson, H.
Pickett, E.
Isotope &
Form (S or L)
H-3 S
H-3 Animal
C-14 S
H-3 Animal
H-3 L 1 gal
H-3 L 1 gal
H-3 S
Se-75
Quantity
(mCi)
0.200
0.050
5.000
0.110
0.010
0.010
0.100
0.010
Disposition (Date)
Storage
8/12/78
8/12/78
8/12/78
8/14/78
8/15/78
8/15/78
8/18/78
8/18/78
Incinerate
8/19/78
8/19/78
Burial
8/22/78
8/22/78
8/22/78
8/22/78
8/22/78
Shipment
o
CAUTION
RADIOACTIVE
MATERIAL
ISOTOPE
AMOUNT
PATl"
DO NOT REMOVE THIS TAO
WITHOUT AUTHORIZATION OP
09-874 Pilule* in U 5 •
f Aflinelflltt Inc. 0 tafto Placa. N.r
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115
FIGURE 2
WINDOW
DOOR
SHOWER
o
FLAMMABLE
STORAGE
ROOM
STORAGE
BINS
OVERHEAD
DOOR
nn
SINK
DOOR
HEALTH PHYSICS
STORAGE BUILDJN6.
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116
AN ALTERNATIVE TO STOP THE PROLIFERATION OF
LOW-LEVEL TRU WASTE CONTAINERS
J. Bruce Peterson
Mound Facility*
Miamisburg, Ohio
Abstract
Of the many different container configurations now being utilized for
interim storage of low-level TRU waste materials none is readily acceptable
for direct shipment and isolation in the Waste Isolation Pilot Plant (WIPP).
The proliferation of these waste containers is a direct result of efforts by
the waste generators to package their unique TRU wastes into containers that
meet DOE Manual Chapter 0511 twenty-year retrievability requirements under
the differing environmental conditions of onsite storage. TRU wastes that
continue to be packaged in non-standard containers will require repackaging
to meet shipping regulations and WIPP acceptance criteria before storage in
terminal isolation. Specifications for a standard container were developed.
A prototype container has been built.
Background
As presently specified in DOE Manual Chapter 0511-044d(4), solid trans-
uranic waste packaging and storage conditions shall be such that the packages
can be readily retrieved in an intact, contamination-free condition for 20
yr.
The retrievable storage site for defense transuranic wastes at the
Idaho National engineering Laboratory (INFL) has been accepting waste since
November, 1970, and has stored this waste in an area designated the Trans-
uranic Storage Area (TSA). The packaging and storage conditions for the
waste stored at the TSA meet the requirements that the containers be readily
retrievable in an intact, contamination-free condition for 20 yr.
Current DOE Division of Waste Management plans are to continue using the
retrievable storage areas until the New Mexico Waste Isolation Pilot Plant
(WIPP) facility attains full operational status in FY-1988. According to
projections, WIPP will begin receiving transuranic wastes in FY-1983. This
waste will be stored so that it can be monitored to evaluate the behavior of
the waste types under the storage conditions. Projections indicate that the
Pilot Plant phase will continue for 3 to 5 yr, after which, with retrieval
demonstrated and experimentation successfully completed, the pilot plant will
be converted to an operational repository for permanent disposal of wastes.
*Mound Facility is operated by Monsanto Research Corporation for the U. S.
Department of Energy under Contract No. DE-AC04-76-DP00053.
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117
During 1977, Monsanto Research Corporation (Mound Facility) participated in
a study leading to the establishment of conceptual design criteria for defense
transuranic waste packaging for Interim Storage and Terminal Isolation. A
contractor questionnaire was used to gather pertinent data. Site visits were
made to formulate an integrated contractor consensus; a packaging meeting was
held to examine, discuss, and integrate packaging philosophies; and data col-
lected from these activities and from Task Force meetings were consolidated
to provide input to the conceptual design criteria.
Development of Design Criteria
An analysis of the information exchanges with the contractors dictate
that both a drum configuration and a box geometry (preferably a modular con-
cept) are needed. This analysis and mutual packaging consensus are based on
the following contractor requirements and waste generation history:
1. Present material handling systems
2. Current and future waste processing systems
3. Present material assay systems
4. Available modes of transportation
5. 71% of the low-level TRU waste generated in 1976 was packaged in
box geometry
In addition, it was found that the cost of any new packaging system is
extremely important to the contractors. This cost conservation is not only
based on future generation of low-level TRU wastes at the contractor sites,
but also strongly influenced by known and planned decontamination and decom-
missioning projects where substantial increases in low-level TRU wastes are
projected.
It was also concluded that the packaging acceptance criteria should be
consistent for both DOE and commercial TRU wastes, since both types of gen-
erators produce essentially the same types of waste.
From these criteria, a set of conceptual design specifications was
assembled for the waste container which is defined as a box or drum, includ-
ing any associated liner and/or shielding material, that immediately sur-
rounds (and is considered to be an integral, disposable part of) the waste
material. These major considerations were also incorporated: public and
generator safety, waste forms to be packaged, requirements for interim stor-
age, transportation from interim storage to terminal storage, and cost effec-
tiveness.
Structural Design
The structural design of all low-level TRU waste containers must meet the
minimum requirements of a Type A package as outlined in 49 CFR 173.398b.
Low-level TRU waste is any solid waste material, other than high-level waste,
which is contaminated with long-lived alpha emitters to the extent that,
under the provisions of DOE Manual Chapter 0511, it is not suitable for sur-
face burial, but which exhibits sufficiently low radiation levels (£500 mrem/
hr) that it is amenable to handling by "contact" methods. This minimum struc-
tural design requirement shall be required for all TRU waste packages to
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118
assure safety to personnel during handling, loading, and unloading operations.
During shipment, the Type A containers may be placed inside a reusable Type
"B" overpack. The Type "B" container must meet more rigorous structural de-
sign requirements and tests than Type A containers to provide for maximum
safety during shipment. Cost effective packaging and transportation of TRU
waste materials will require the single use Type A packages to be relatively
inexpensive but capable of meeting the requirements of contamination control
from the time the containers are filled until they are backfilled inside the
WIPP facility.
Design Life (Decomposition)
The design life of all TRU low-level waste containers for contamination-
free retrieval shall be 10 yr minimum when stored in a noncorrosive atmos-
phere (pH 7-8), 60% relative humidity, and 100°F. The design life parameters
may suggest a change in DOE Manual 0511 from 20-yr intact contamination-free
retrievability to a 10-yr intact contamination-free retrievability concept.
Life of the shipping container will start from the time the container is
manufactured until backfilled in the WIPP. The 10-yr life is based upon the
forecast that the WIPP will be fully operational for TRU waste containers in
1988. Life cycle of the container will include manufacturing, delivery,
storage, transmittal into the WIPP, analysis, and backfilling. This life
cycle should be approximately 5 yr; however, it could approach 10 yr because
the backlog of interim stored wastes will be in direct competition with
freshly packaged waste for isolation space in the Isolation Facility. All
filled waste containers must be protected from environmental conditions that
could significantly reduce the design life of the waste containers to less
than 10 yr.
Materials of Construction
Materials of construction shall be based on design life and structural
design requirements. Ferrous and nonferrous metals, plastics, reinforced
plastics, fiberboard, corrugated fibers, wood, and concrete have been con-
sidered for container materials. All these materials can meet the require-
ments for hazardous materials transportation and are acceptable in the WIPP
in limited quantities. Therefore, the choice of materials, or combinations
thereof, can be made from the above group. However, choice will be influ-
enced by the waste form, container design, economics, and, most important,
final WIPP TRU Waste Acceptance Criteria.
Maximum Weight of Container and Contents
The weight of a single container filled to 98% capacity is limited to
25,000 Ib (11,400 kg) based on a contents density of 125 lb/ft3 (2000 kg/m3).
This design weight is based on the 25,000 Ib (11,400 kg) maximum capacity of
the WIPP low-level hoist cage.
The container family should be modular, having a shape which will pro-
vide maximum packing efficiency in storage. The cylindrical container has a
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119
packing efficiency of -0.69 and the void space will be 31 ft3 for every 100
ft3 of waste in terminal isolation. The cylindrical container, up to 8 ft3,
is readily mass produced and available in metal, plastic, and fiberboard.
However, because of the underground location for isolation of TRU wastes,
emphasis must be placed on container shape with higher packing efficiencies
for the waste materials.
Dimensions
Waste container dimensions should be based on criteria to provide flexi-
bility in mode of transportation.
Handling Appurtenances
All low-level TRU waste containers must be provided with cleats, offsets,
or chimes which permit handling by fork lift.
Security Seal
The outside of each waste container must incorporate a feature such as
a seal that is not readily breakable and that, while intact, will be evidence
that the package has not been illicitly opened.
Cost
Current low-level waste packages which can meet the requirements of DOE
Manual Chapter 0511, WIPP, and DOT Type A have costs ranging from $3.57/ft3
(4x4x7 ft fiberglass reinforced polyester resin box) to $18.19/ft3 (DOT 17H,
55-gal, stainless steel drum) for the packaging materials. Cost per cubic
foot of storage volume for the standardized container family should be to-
ward the lower end of this range to be cost effective.
Commercial Construction
Continuing efforts to develop an Acceptable TRU Waste Container System
were enhanced by the results of a survey of container manufacturers completed
in January 1978. The purpose of the survey was to determine if any commer-
cially available containers were applicable to the shipment and storage of
low-level TRU waste. The survey provided an overview of current packaging
technology and availability. A marketing information center supplied a mail-
ing list of 4191 National Manufacturers in seven major container categories
who were currently engaged in the manufacture of containers used for packag-
ing and shipment of various industrial commodities. The manufacturers were
contacted by mail and invited to submit technical information on: container
types; size, shape, internal volume; weight; closures; DOT certification (if
applicable); performance data; unit and quantity cost.
One container manufacturer, Lanson Industries, submitted a proposal to
work with Mound Facility to develop a standardized waste container system
which could meet the above criteria.
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Two container sizes were designed by Lanson. A basic 2x2x3 ft (12 ft3)
rectangular container which will overpack a 17C 55-gal drum and a 4x4x7 ft
rectangular container to overpack the FRP plywood boxes were designed. These
container dimensions are slightly larger than stated to provide the overpack
feature. They are sized such that the small containers will fit inside the
larger container with minimum void volume. Both container designs are top-
loading; however, the large container is fitted with handling appurtenances
which will accommodate rotation onto its side for loading large equipment or
boxes with a forklift.
These containers will be manufactured from Corten A Type 4 weathering
steel which has five to eight times the corrosion resistance of low carbon
steel and, if painted, will hold paint two to three times longer than low
carbon steel.
The 4x4x7 ft container designs included a container of 11 gauge rein-
forced metal and a 3/8 in. thick container designed to withstand the % atmos-
phere reduced pressure test without reinforcement. All the containers have
clamped, gasketed closures with a feature for welding.
On January 10-12, 1979, the qualification testing of the top loading,
4x4x7 ft overpack prototype was completed.
The prototype passed the following tests with only minor damage and no
loss of contents:
1. Reduced pressure - 7.5 psig
2. Vacuum - 7.5 psig
3. Puncture test - 13-lb pin at 40 in.
4. Drop test - 48 in. with 14,700-lb load
5. Compression test - 25,000 Ib for 24 hr, also 85,000-lb
structural engineering analysis on
compression
The design features, quality of workmanship, testing results, and com-
petitive costs contribute to the fact that this Mound/Lanson prototype is
far superior to any of the contact handled TRU waste containers currently in
use. However, for this container to be competitively priced ($700-800 each)
at approximately $6/ft3, it must be manufactured in large quantities on a
semi-automatic production line. The concept of central procurement would
allow the procuring contractor to purchase in large lots and distribute in
small quantities, as needed. The central procurement concept has been
around for a while and might be one for GSA to consider.
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VOLUME REDUCTION OF LOW-LEVEL RADIOACTIVE WASTE
WITH A HAMMERMILL
by
Winborn D. Gregory
University of Rochester
Health Physics, Box RB&B
601 Elmwood Avenue
Rochester, NY 14642
ABSTRACT
A Jacobson Model J-3 Hammermill was recently installed at the University
of Rochester to process low-level radioactive wastes from hospital and re-
search laboratories. The hammermill will handle both glass and plastic vials
of all types. The waste is poured into a hopper located on the top of the
seven foot high assembly. The material is gravity fed into the hammermill
which is driven by a 15-horsepower motor at 3600 rpm. The waste is pounded
by the rotating hammers into a metal screen perforated with one inch holes.
The ground-up product is discharged from the bottom of the unit into a 55-
gallon shipping drum. A volume reduction of 4s1 has been achieved on an
equal mixture of glass and plastic vials.
DISUCSSION
A hammermill is an industrial type machine usually associated with farm
products. A Model J-3 Hammermill was ordered from the Jacobson Machine
Works, Inc., 2445 Nevada Avenue North, Minneapolis, MN 55427. The basic unit
is shown in the mid portion of the picture. The hammermill was ordered with
the custom made base, motor mount, and inlet hopper as shown. The total as-
sembly weighs 1300 pounds, and costs $4,068.00 including transportation.
The hammermill was installed in an existing basement room which is used
for receipt of all radioactive waste in the Medical Center. Installation was
accomplished by positioning the unit in location and bolting on the inlet
motor. A 6 inch vent pipe was installed from the hammermill fan to a nearby
isotope hood. A drum lid was adapted to fit the 12x12 inch discharge outlet.
A pallet truck is used to position an empty drum and elevate it to make a
tight seal with the adapted drum lid.
After gaining some operating experience, two problems were evident.
First, the inlet hopper was only 10 inches wide. A galvanized steel ex-
tension was made that was bolted in place. This extended the hopper to 22
inches wide by 36 inches long making it possible to empty the entire contents
of a 13-gallon waste collection container with no spillage. A slide gate was
also fabricated as a part of the hopper. The gate may be adjusted to limit
the flow of material into the mill so that an overload will not occur. A
controlled flow is important for processing plastic since the motor will
stall or fuses will blow if too much is fed in at once.
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INJ
FIGURE 1. HAMMERMILL
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The second problem was the dust in the discharge air from the fan. An
attached fan draws air down the hopper through the mill and the waste barrel,
and finally up from the barrel through the fan and out the discharge vent
pipe. This airflow is necessary for cooling the hammer operation, and for
pulling trash through the screen. The problem is that much of the finer dust
and paper is not deposited into the drum, but is carried out the discharge
vent. Several steps were taken to solve this problem. A small water supply
was added to the inlet hopper so that a trickle of water would enter the mill
to dampen the dust. The air flow was reduced by installing a damper in the
vent pipe, and a hole was cut in the duct between the drum and the fan. This
latter step allows some make-up air to enter the fan. A gate was installed
so that the make-up air could be regulated.
Finally, a filter box was constructed to trap dust in the discharge. It
consists of both a fine mesh screen and a furnace filter. A clean-out door
is provided so that the filter can be changed and the box cleaned out with a
small vacuum cleaner. This is usually necessary after filling two to three
55-gallon drums.
There are two items which require periodic inspection and replacement.
They are the screen and the hammers. The screen is a curved piece of steel
plate perforated with one inch holes Other hole sizes are available, but
this size gave the best results for general purpose use. The screen lasted
five months, and its replacement cost is only $20. Access to the screen is
gained by the removal of two hand nuts and a hinged cover.
The other item requiring inspection is the hammers. There are four rows
of hard faced steel hammers. These rotate around the center of the axis at
3600 rpm, and just barely clear the screen. This pounds the waste through
the holes in the screen. The hammers must be rotated and reversed so that
all four edges are used. Properly rotated, it is estimated that a set of
hammers will last about a year depending on the amount of use the mill re-
ceives. A replacement set is about $60. It takes about an hour for two
people to rotate or change a set of hammers.
In actual operation it is necessary to wear ear and eye protection, lab
coat and gloves. Odor masks are also worn (3M Company) both for comfort
around pungent waste, and for protection from Iodine-125 vapor. Glass vials
feed themselves by gravity, but the lighter plastic vials need assistance by
pushing open the hinged safety gates located in the inlet hopper. These
gates prevent the flyback of material after it enters the mill.
In use, segregated quantities of glass and plastic are fed into the
mill. For glass the discharge is a finely ground product. Plastic comes
out in chunks no larger than one inch in diameter as determined by the screen
size. An overall volume reduction of 4:1 is achieved when equal quantities
of glass and plastic are processed. The resulting gross weight of a 55-
gallon drum filled in this manner is about 400 pounds. For plastic alone
the volume reduction is 2:1 with a gross weight of 265 pounds for a full
drum. It requires about one hour for one person to fill a 55-gallon drum
using the mill.
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This is about the same amount of time required to handle waste using a
compactor. An institutional type compactor had been used to reduce the
volume of glass waste. The compactor was a Slugger-S model made by Inter-
national Dynetics Corporation. It is of the screw drive type, and crushes
waste directly into 55-gallon drums. A volume reduction of 3:1 was usually
achieved when compacting glass alone. Plastic will not compact.
This compactor is still in use to reduce the volume of paper wastes and
other solids that have large quantities of radioactivity associated with them.
In order to determine how the hammermill effected the overall waste
volume, it is necessary to review the waste collections from previous years.
Not including drums filled with animal carcasses, an average of 18 drums per
month were filled in both 1976 and 1977. In 1978 through the end of August
this had increased to an average of 20 drums per month. During those years
the large compactor was in use. From September of 1978, when the hammermill
was fully utilized, through January 1979, an average of only 14 drums per
month were filled. This is a reduction of six drums per month. At a cost
of $60/drum, this gives an annual savings of approximately $4320. The pur-
chase cost of the mill with electrical hook-up, alterations and maintenance
items is about $5000. This makes the payback period just over one year.
It is interesting to note that during the five months operating experi-
ence, six of the fourteen drums per month are still filled with compacted
waste, two are filled with mini-vials and Nalge filmware containing liquids
to which vermiculite has been added, and the remaining six per month are
filled by the hammermill. It is expected that this ratio of 1:1 between
compacted material and milled material will change since a new policy has
been instituted of further segregating burnable waste into plastics only and
paper/gloves only. This segregation is done at the user level and should
provide additional waste that can be processed in the mill. The filtration
system will also be improved so that higher activities of waste that are now
compacted can be safely run through the mill without contaminating the labo-
ratory and the environment. This should provide greater volume reduction in
the overall waste picture, and increased savings.
The experience gained using this hammermill indicates that a hammermill
is more effective in volume reduction than a compactor. The hammermill also
lends itself to many adaptations which might include automatic feeding via a
conveyor or hopper system. It could also be used to pulverize filled scin-
tillation vials if a method were devised to extract the liquid and carry off
the solid materials via an auger.
ACKNOWLEDGEMENT
The assistance of Robert J. Williams of the R. E. Williams Co., Inc. of
Buffalo, NY is gratefully acknowledged. As the local representative of the
Jacobson Machine Works, his operating experience and ideas were invaluable.
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SESSION C
WASTE DISPOSAL OPERATIONS AND ALTERNATIVES
Session Chairperson
A. A. Moghissi
U.S. Environmental Protection Agency
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REGULATORY CONSIDERATIONS FOR RADIOACTIVE WASTE DISPOSAL
Teledyne Isotopes
D.R. Fuhrman
S.A. Black
J.P. Pasinosky
Abstract
Packaging, transportation, and burial of radioactive wastes involves a com-
plex, sometimes nebulous set of regulatory conditions, restrictions, and ex-
ceptions for proper handling. These regulations affect the shipper of radio-
active wastes directly as in the case of the Department of Transportation (DOT),
Nuclear Regulatory Commission (NRC), state highway permits or bridge and tunnel
authorities, or indirectly by imposing restrictions on the burial site of which
the shipper must be aware.
Due to the complexity of the regulations involving the various levels and
types of radioactive wastes, this paper concentrates on "low level" (i.e., Type A
and LSA quantities) wastes. When considering the shipment of Type A or LSA waste,
the shipper must first consider the packaging, marking, labeling, etc. require-
ments as specified by the DOT and compatible with the respective burial site re-
strictions. The appropriate shipping papers must be completed certifying the
packaging, marking, and labeling in accordance with DOT regulations.
In transporting radioactive wastes, the DOT imposes requirements and restric-
tions on the carrier. Driver's logs and examination certificates must be in the
driver's possession. In addition, requirements for the safe operation of vehi-
cles carrying hazardous materials are imposed by the DOT. In addition to the
DOT regulations, certain states, localities, highway authorities or bridge and
tunnel authorities impose restrictions or require permit authorization prior to
transporting radioactive wastes through their jurisdiction. In many states, the
respective environmental agencies have recently imposed permit requirements for
transport of any hazardous materials.
The regulatory process governing radioactive waste disposal has become a
maze of requirements, conditions, restrictions, and exemptions which are often
redundant and ambiguous, which raises the costs to commercial waste disposal
companies, carriers, and ultimately the generator.
Introduction
The question of which and what regulations are applicable must have gone
through the mind of any person involved in disposing radioactive wastes. This
length paper could not attempt to identify and describe all the regulatory pro-
cesses for each type of waste, but it is hoped that it will serve as a guide to
direct the generators of radioactive wastes to the correct sources or at least
to initiate the right questions which will lead them to the correct sources.
The first part of the paper will give an overview of the regulations af-
fecting all phases of radioactive waste disposal, i.e. generator to carrier to
disposal site. In this section the applicable regulatory agencies at various
stages and a list of some of the regulatory considerations each group must con-
tend with, will be identified.
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The second part will describe, in greater detail, the regulatory consid-
erations for the disposal of low specific activity (LSA) material. The dis-
cussion will include which regulations apply to LSA material shipped by various
methods, i.e. packaged and unpacked material in exclusive use vehicles, and
materials shipped in non-exclusive vehicles.
Regulatory Overview
Regulations pertaining to the disposal of radioactive wastes are diverse
and complex, requiring careful review of the appropriate regulations according
to the particular wastes in question. Due to large potential liabilities from
infractions or accidents, generators of radioactive wastes need to be cognizant
of all aspects of disposal and be assured that all parties involved in the trans-
portation and disposal of their wastes are reputable firms with extensive es-
perience and liability coverage.
Generally, radioactive wastes generated are disposed by two basic methods:
(1) The generator acts as the shipper and coordinates all transportation and
burial. In this case, the generator would perform all the necessary documenta-
tion, marking, labeling, packaging, and monitoring needed to comply with the
regulations affecting the shipment. (2) The generator contracts a disposal ser-
vice to act as a consultant and to remove and transfer the waste to an author-
ized burial site. In this case, the disposal service may supply the necessary
packaging, labeling, shipping documents, and monitoring to insure that the gen-
erator will adequately comply with all the regualtions governing transportation
and disposal. It is important that careful consideration is given in selecting
the dispossl service because the liability, at least in part, may ultimately
reside with the generator.
In 1966, the Department of Transportation Act was passed giving the DOT
regulatory responsibilities for the safe transport of radioactive materials by
all modes of transportation in interstate or international shipments except for
postal shipments. Postal shipments are the jurisdiction of the U.S. Postal
Service, in 39CFR.
Economic considerations for transportation of radioactive materials are
under the jurisdiction of the Interstate Commerce Commission (ICC) for land ship-
ments and Civil Aeronautics Board (CAB) for air shipments through the issuance
of operating authorizations and controlling tariff rates.
The Nuclear Regulatory Commission (NRC) regulates the possession and use,
including transport, of byproduct, source and special nuclear materials through
the licensing of these materials. The transfer of fissile material or quantities
exceeding Type A to a carrier is subject to the conditions and requirements set
forth in 10CFR71.
Various states have entered into agreement with the NRC which entitles them
to acquire regulatory authority for the possession and use of byproduct, source
and special nuclear material. The exception is for critical quantities of SNM.
These states, called "Agreement States", have developed regulations for the safe
use of radioactive materials including intrastate shipments.
In addition to federal regulations pertaining to the possession, use, .and
transportation of byproduct, source and special nuclear materials, states have
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the authority to regulate radioactive materials which do not fall into one of
these three categories (e.g. radium-226 and cobalt-57). These Materials are
usually subject to the conditions set forth by radiological health groups in
each respective state.
Recently a number of states have adopted regulations, through state environ-
mental protection groups, to control the transportation of certain types of
hazardous wastes, including non-byproduct material. These agencies have developed
a manifest system to account for the waste at any stage during the transfer from
the generator to the disposal site known as the "cradle to graveyard" system.
International shipments are subject to additional requirements as set forth
by the DOT in 49CFR. Each shipment must be identified by its contents, mode of
transport, and destination to determine which international authorities have
jurisdiction over such shipments.
Regulatory Considerations of the Shipper
In considering the shipment of radioactive wastes, it is necessary to eval-
uate the proper regulatory requirements which are primarily subject to the type
of isotopes, quantity, and form. The DOT details requirements for transportation
based on the type of material in question(e.g. limited quantity, LSA, Type A,
Type B, etc.) Once the material is properly identified and the quantity and the
form determined, appropriate packaging can be selected. Generally, DOT specifi-
cation packaging is required except in cases of limited quantities and low spe-
cific activity (LSA) materials shipped under certain conditions for transport.
DOT specification packaging must meet certain test criteria to qualify for a
Type A, Type B, or Large Quantity package.
In the case of Type A packaging the regulations now provide for a pure per-
formance based DOT, Spec. 7A, Type A general package. A shipper must assure that
his package will compare to tests and design specifications for the 7A package.
The construction must be adequate to prevent the loss or dispersal of its contents
and to maintain its radiation shielding properities if the package is subjected
to "normal" conditions of transport.
Type B packaging, in addition to general packaging requirements and perform-
ance standards for "normal" conditions of transport, must also meet certain test
conditions incident in an accident with a limited loss of shielding integrity and
essentially no loss of containment.
Each package offered for shipment must be marked and labeled in accordance
with requirements in 49CFR except for certain shipments of limited quantity or
LSA material. Two labels must be placed on opposite sides of the package in
accordance with the requirements in 49CFR.400. This symbol was recommended by the
International Commission of Radiation Protection (ICRP) in 1956 and adopted by the
American National Standards Institute (ANSI) as the standard radiation symbol.
The type of label is based on the external radiation levels at contact and
three feet from the package, as defined in 49CFR173.399. The radiation level at
three feet is termed the Transport Index, which is directly transcribed on "Radio-
active Yellow II" and "Radioactive Yellow III" labels.
Contamination control, as described in 49CFR173.397, is required for any pack-
age offered for transportation and for the vehicle after being used for "exclu-
sive use" shipments of radioactive materials.
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Shipping documents, required with each shipment of radioactive material,
must include shipping papers which identify the material and other required
information. In addition, the shipper's certification is required as amended
April 15, 1976 and must be instituted by July 1, 1979.
Certain "special" conditions require added precautions and packaging spec-
ifications as in the case of pyrophoric material and liquids. The specific
requirements pertaining to these shipments must be carefully considered in ac-
cordance with 49CFR.
Fissile radioactive materials require additional labeling with Fissile
Class I, II, or III labels. Special packaging and shipping requirements are im-
posed on these shipments to ensure against nuclear criticality. Packaging
specifications, labeling, and special procedures are detailed in 49CFR173.396 of
the DOT regulations and 10CFR71 of the NRC Regulations.
Placarding of the transport vehicle is normally the responsibility of the
carrier for loads containing packages bearing a "Radioactive Yellow III" label.
The exception is for "full-load" LSA shipments which require placarding of the
vehicle and is under the responsibility of the shipper. For shipments with pack-
ages in excess of 200 mr/hr at contact and 10 mr/hr at three feet the shipper
also has the responsibility to provide to the carrier specific instructions, in-
cluded with the shipping papers, which detail the radiation levels at three feet
from the package, external surface of the vehicle, two meters from the vehicle,
and in any area in the vehicle normally occupied by personnel.
In addition to the DOT, the NRC imposes specific requirements in 10CFR71
which control the packaging and transportation of fissile material and quantities
greater than Type A. Specific requirements must be carefully reviewed to deter-
mine the applicability of these standards. Shipments of less than Type A quan-
tities, shipments for medical use, various quantities of fissile material, and
certain Type B shipments, are exempt from the provisions of this part.
In considering the destination for burial, the shipper must be aware of re-
quirements and conditions imposed by the burial site. Each burial site has a
list of conditions which are readily available on request. Particular packaging
specifications, in addition to DOT requirements, may be imposed on the shipper
for certain types of materials such as bulk liquids, scintillation vials, or
animal carcasses. Restrictions on total activity and transuranic elements are
also a consideration. Careful determination of the type and method of packaging
and proper documentation of the waste will avoid a possible rejection of the
waste at the burial site.
Regulatory Considerations of the Carrier
When a disposal service is used, many of the above considerations are per-
formed by the particular company. The disposal service acts as an intermediate
agent supplying all the necessary packaging, labels, and shipping documents to
meet the necessary regulatory requirements. The disposal service may transport
the waste directly to the burial site or store the waste until aggregate quan-
tities from multiple pickups total a sufficient amount to economically send to
a burial site. Regardless of which method is employed, additional regulations
are imposed on the disposal service or the common carrier.
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During transport of the waste DOT regulations affect every aspect of the
transportation. Basic requirements are imposed on a hazardous waste driver such
as written examinations, medical examinations, traffic convictions and accidents,
driving test, previous employment records, and an annual review of the driving
record. When operating a vehicle, pre-inspection of the vehicle is required,
driver's logs must be maintained, and certain restrictions for inclement weather
are imposed.
Added restrictions have also been imposed on the transporters of hazardous
wastes when traveling on certain roads, bridges, tunnels, or through certain
communities. A definite routing plan must be determined before leaving a shipper'
facility. Information regarding these restrictions may usually be obtained
through bridge and tunnel authorities, the DOT, or other local agencies with
authority over transportation on roadways.
If the disposal service uses intermediate storage, regulations imposed by
local agencies may also be a consideration in addition to state and federal agen-
cies. The disposal service is required to obtain a license for possession and
transfer to a disposal site from the NRC and state in "non-Agreement States", or
from the state in "Agreement States". Local fire or health codes may impose con-
ditions on the storage facility such as the use of automatic fire systems.
Other regulatory agencies may be involved with certain shipments depending
on the transportation mode and destination of the shipment. Agencies involved may
include the International Air Transport Association (LATA) or International Atomic
Energy Agency (IAEA) for international shipments, and the U.S. Postal Service,
U.S. Coast Guard, Federal Aviation Administration, or Interstate Commerce Com-
mission for domestic shipments.
Regulatory Considerations of the Disposal Site
Today, regulations governing disposal site operations are very complex and
the ruling authority is not always well defined. The regulatory processes have
become subject to many external forces such as environmental and public activist
groups.
Generally, regulatory responsibility is maintained by the NRC for byproduct,
source, and special nuclear materials for activities conducted in "non-Agreement
States". In "Agreement States" regulatory responsibilities, except for special
nuclear material, are subject to state control. Stringent requirements are im-
posed on the burial sites to minimize mishaps during burial operations and pre-
vent releases to the environment. Details of trench configuration and parameters
are well defined and incorporated in the operating licenses. The trenches must
be filled and backfilled to certain specifications. To evaluate the containment,
extensive environmental monitoring is required at numerous locations around the
trenches and control zone perimeters. Burial sites also have restrictions on
transuranic isotopes and possession limits. After a trench is completed, back-
filling techniques are employed to insure proper drainage. A marker, called a
tombstone, is placed on one end of the trench and is inscribed with the total
activity, volume, and date of completion of burial operations.
Details Pertaining to a Shipment of LSA Material
In the following analysis, a shipment of low specific activity (LSA)
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material will be evaluated according to various modes of transportation. LSA
material is defined in 49CFR173.389 to include the following:
1. Uranium or thorium ores and physical or chemical concentrates of these
ores;
2. Unirradiated natural or depleted uranium or unirradiated natural thorium;
3. Tritium oxide in aqueous solutions provided the concentration does not
exceed 5 millicuries per milliliter;
4. Material in which the activity is essentially uniformly distributed and
in which the estimated average concentration per gram of contents does
not exceed:
(i) 0.0001 millicuries of Group I radionuclides; or
(ii) 0.005 millicuries of Group II radionuclides; or
(iii) 0.3 millicuries of Groups III or IV radionuclides.
5. Objects of nonradioactive material externally contaminated with radio-
active material, provided that the radioactive material is not readily
dispersible and the surface contamination when averaged over an area
of 1 square meter, does not exceed 220,000 dpm per square centimeter
of Group I radionuclides or 2,200,000 dpm per square centimeter of other
radionuclides.
Once the radioactive material has been determined to meet the criteria for
LSA, packaging requirements are subject to the method of transportation (i.e.
exclusive use vs. non-exclusive use; and closed transport vehicle vs. any vehicle).
For packaged shipments of LSA radioactive material transported in vehicles
assigned as exclusive use, an exemption from specification packaging, marking,
and labeling is provided if the shipment meets various additional conditions:
1. Materials must be packaged in strong, tight packages so that there will
be no leakage of radioactive material under conditions normally incident
to transportation.
2. Packages must not have any significant removable surface contamination
as defined in 173.397.
3. External radiation levels must comply with 173.393.
4. Shipments must be loaded by the consignor and unloaded by the consignee
from the transport vehicle in which originally loaded.
5. There must be no loose radioactive material in the vehicle.
6. Shipments must be braced to prevent leakage or a shift in the load under
conditions normally incident to transportation.
7. Except for shipments of unconcentrated uranium or thorium ores, the
transport vehicle must be placarded in accordance with 172.500.
8. The outside of each outside package must be marked "Radioactive-LSA".
9. Specific instructions for maintenance of exclusive use shipments must
be provided by the shipper to the carrier.
Unpackaged shipments of LSA materials must be transported in closed trans-
port vehicles and comply with the following conditions in addition to packaged
LSA shipments.
1. Authorized materials are limited to the following:
(i) Uranium or thorium ores and physical or chemical concentrates of
those ores.
(ii) Uranium metal or natural thorium metal, or alloys of these mate-
rials; or
(iii) Materials of low radioactive concentrations, if the average
estimated radioactive concentration does not exceed 0.001 milli-
curies per gram and the contribution from Group I material does
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not exceed one percent of the total radioactivity.
(iv) Objects of nonradioactive material externally contaminated with
radioactive material, if the radioactive material is not readily
dispersible and the surface contamination, when averaged over one
square meter, does not exceed 220,000 dpm per square centimeter of
the other radionuclides.
2. Bulk liquids must be transported in certain specification type containers
specified in 49CFR.173.
LSA shipments transported in non-exclusive use vehicles must be packaged in
accordance with the requirements for Type A shipments and marked and labeled in
accordance with 172.300 and 172.400.
Summary
The complexity involved with properly classifying radioactive wastes then
deciding which set of regulations apply with all the conditions, requirements, ex-
ceptions and exemptions, lend itself to frequent infractions of the regulations
for the average person who deals with occasional shipments. Instititutions gen-
erating large volumes of waste can afford to staff persons who will become fami-
liar with the regulations and keep abreast of amendments. The smaller generators
are subject to limited knowledge of the regulations pertaining to their shipments,
or to rely on a disposal service.
Regardless of the types of wastes or quantities that are being disposed, a
thorough review of the applicable regulations must be performed to ensure full
compliance and a safe transfer.
Bibliography
(1) Adam, J.A. and Rogers, V.L., NUREG-0456, "A Classification System for Radio-
active Waste Disposal - What Waste Goes Where?", FBDU-224-10, June 1978.
National Technical Information Service, Springfield, Virgina, 22161.
(2) Edling, Don A., Monsanto Research Corporation, "Certification of Packaging:
Compliance with DOT Specification 7A Packaging Requirements", October 1976.
National Technical Information Service, Springfield, Virginia, 22161.
(3) "A Review of the Department of Transportation (DOT) Regulations for Trans-
portation of Radioactive Materials", Ocoober 1977, U.S. Department of Trans-
portation, Material Transportation Bureau, Office of Hazardous Materials
Operations, Washington, D.C., 20590.
(4) NUREG-0383, Volumes I & II, "Directory of Certificates of Compliance for
Radioactive materials Packages", December 1977- National Technical Inform-
ation Service, Springfield, Virginia, 22161.
(5) Proceedings of the Fifth International Symposium, "Packaging and Transpor-
tation of Radioactive Materials", Volume II, May 1978, Las Vegas, Nevada.
(6) 10CFR, Nuclear Regulatory Commission, Part 71.
(7) 39CFR, U.S. Postal Service.
(8) 40CFR, Department of Environmental Protection.
(9) 49CFR, Department of Transportation.
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OPERATIONAL EXPERIENCE AT CHEM-NUCLEAR'S
BARNWELL FACILITY
by
David G. Ebenhack
Chem-Nuclear Systems, Inc.
Abstract
Chem-Nuclear's low level burial ground located in
Barnwell, S. C. with primary licensing by the State of
South Carolina has seen a substantial increase in volume of
waste and presently serves the majority of fuel and non-
fuel cycle radwaste generators in the country. The waste,
upon receipt, is monitored and disposed of in one of our
engineered trenches. The packaging requirements, trench
design and surface management minimize the possibility of
release through the water pathway. The Health Physics
practices and the environmental programs evaluate and monitor
the personnel and population exposure pathways.
Discussion
Chem-Nuclear Systems, Inc. (CNSI) is a nuclear
service orinetated corporation providing expertise, service
and assistance in a number of areas for both fuel cycle and
non-fuel cycle customers. One of Chem-Nuclear1s major operations
and the one for which we are the most widely known is the operation
of a low-level radioactive waste burial site.
Chem-Nuclear1s facilities are located outside of
Barnwell, South Carolina adjacent to DOE's Savannah River
Plant and Allied General's Reprocessing Plant. Within the
approximate 300 acre tract operated by Chem-Nuclear is the
burial site itself as well as transportation and maintenance
facilities, support facilities for the mobile solidification,
resin, technical and decontamination services offered by CNSI
and administrative facilities.
Primary licensing is through South Carolina's Department
of Health and Environmental Control, Bureau of Radiological
Health. All radiologically controlled activities, burial
packaging and burial site design, construction, monitoring and
perpetual care considerations are regulated by the State.
Transportation and cask compliance and regulatory activities
are shared by the State, DOT and NRC. Chem-Nuclear is also
licensed by the Nuclear Regulatory Commission to possess
quantities of special nuclear material greater than can be
authorized by the State of South Carolina under the Agreement
State Program.
-------
134
Burial activities have increased significantly during
the last several years. The closing of the Sheffield, Illinois
site left the Barnwell site the only remaining commercial low-
level burial site east of Beatty, Nevada. The volume in cubic
feet of waste buried each year since the initial licensing
and operation in 1971, is depicted on the slide.
In 1978 we buried a total of 2,225,049 cubic feet of
waste with a collective activity of 652,061 curies. The State
of South Carolina has recently placed a ceiling on the annual
volume of our burial operations which averages 200,000 cubic
feet per month. With an average volume of 185,421 cubic feet
per month in 1978, the importance of volume reduction at the
point of origin becomes obvious.
Approximately 225 customers shipped waste to our site
in 1978. Six of these customers, however, service a large
number of additional customers, thus, bringing the total number
of organizations utilizing the Barnwell site to an estimated
five hundred (500) . The nuclear industry which makes up the
fuel cycle accounts for approximately 75% of the waste by volume.
Medical, academic, industrial and research facilities, catagorized
as non-fuel cycle, generate the remaining 25%. The percentage
(by volume) of waste received from each state is shown here.
Regionally, the largest volume comes from the Northeast with
47.8% of the total, followed by the Southeast with 32.5%, the
North Central with 18.2% and the Mid-West with 1.4%. We
received no waste in 1978 from the far West.
Waste shipped to the site arrives in a very large
variety of package shapes, sizes, weights; cask and trailer
types with radiation and contamination levels ranging from
background to -*-' 20,000 R/hr and 500,000 dpm/100 cm^ respective-
ly. Special handling and offloading procedures have been
developed and are continuously being improved upon to maintain
and reduce personnel exposure as low as reasonably achievable.
Contamination levels on the site are kept at background levels
to further assure radiological control. The average monthly
absorbed dose to the different catagories of site personnel
shows that we have maintained exposures to about half or
less of the established standards. We expect to be able to
reduce this further through refinement of the handling and
offloading techniques. One recent change in this area is the
establishment of a set of site criteria. These criteria
predominately specify special packaging and loading configuration
which when fully impelmented, should reduce exposure and increase
the site safety and efficiency.
All shipments, prior to acceptance on the site, are
surveyed for contamination and radiation levels. The shipping
papers are checked for accuracy and completeness, and the
packaging is checked for compliance with our site criteria, S. C.
and NRC license requirements and DOT specifications. The
State makes spot inspection of shipments to monitor our per-
formance as well as the shipper's compliance with the DOT and
license requirements.
-------
135
After shipments are received on site, they are
directed to one of the trenches for offloading. Fuel cycle
and non-fuel cycle waste is segregated into different trenches.
After offloading, the truck is taken to the exit gate where
it, or any other vehicle or piece of equipment which has been
in the controlled area, is monitored for contamination prior
to exit. If radiation or contamination levels on sole use
vehicles are detected above 0.5 mr/hr or 2200 dpm/100 cm2,
then release is not permitted until sufficiently decontaminated.
Unrestricted use vehicles are decontaminated to essentially
background. Decontamination is accomplished by wiping,
vacuuming, sandblasting and removal - disposal or a combination
of these. The site Health Pnysics staff performs the site
survey and decontamination activities and monitors all off-
loading evolutions.
The engineered trenches consist of two basic designs -
the regular and the slit trench. The slit trenches are
20 feet deep, 3 feet wide and 250 to 500 feet long. The slit
trench design is utilized for high exposure rate - mostly
component waste. We have two active slit trenches - one
utilized when offloading horizontal casks and the other is
a specially-engineered, TV-monitored pit for offloading
vertical cask.
The regular trenches are nominally twenty-two feet deep
with dimensions varying from 50 to 100 feet wide and 500 to
1000 feet long. There is a nominal 1% slope to the side
where a french drain system is installed with a nominal slope
of 0.3% end to end. Sampling points are installed every 100
feet along the french drain with a sump installed every 500
feet. Prior to initial waste disposed in a trench, 2-3
feet of pervious sand is backfilled in the bottom of the
trench to allow any liquid to transgress to french drain
system and to allow several feet of safety margin between the
first detection of water and the actual waste level, allowing
time to take steps to prevent the emersion of the waste.
All trench design, details and construction activities
must be approved and inspected by the State.
Remote hook-up and release techniques are utilized
routinely in the offloading evolutions as a means of minimizing
personnel exposure. Another special technique utilized
is what we call "Toner tubes". These are tube structures
placed in the waste material during filling operations which
are utilized at a later date for the offloading of high exposure
rate drums. The restricted opening provides substantial
shielding as does the slit trench design. The design shown
in the picture has recently been modified to prevent the
entry of rain and surface water as well as provide more positive
contamination control. The filled tube is capped with a
shield drum of concrete, the guide structure is removed and
the opening filled and capped the same as all trench areas.
-------
136
Since migration via the hydrological pathway is so
important, trenches are constructed/ filled and maintained in
a manner which retards the entry of surface water into or
through the trench. The original sand layer on top of the site
is removed and the walls of the trenches are made of compacted
clay. After the waste is placed in the trench, we fill the
void spaces with sand backfill and use a 10,000 pound vibrating
compactor to accelerate the settling process. A minimum of
two (2) feet of compacted clay is then placed over the trench
to act as a moisture barrier. Our license dictates a minimal
addition of three (3) feet of earth cover over the clay. In
actual practice, 5 to 10 feet of additional cover is provided.
Then, with the addition of topsoil and fertilizer, the area
is contoured and seeded to enhance the surface runoff while
minimizing erosion. License requirements specifying specially
approved solidification media for fuel cycle liquids and
specially designated packaging requirements for non-fuel
cycle liquids further reduces the potential for migration from
the trenches.
There are two major inhibitors to more efficient
operation. Since operations on the site are almost exclusively
conducted outdoors, the weather can have a significant impact.
For example, last week we had a day of freezing cold weather,
coupled with an ice storm which shut down operations by noon.
The next day, it had warmed up some, the power was on and the
ice was gone but what remained was at least 12 inches of
very slippery, gooee red clay. The red clay has fair ion-
exchange properties - a good location for a burial ground, but
really hard to work in when muddy. Just plain rain muddies
up the site and lightning storms shut down the cranes. During
the summer, the heat, insects and dust also leave something
to be desired.
The other major inhibitor to efficient orderly
operations at the site is, unfortunately, the customers.
There are an average of 480 shipments arriving at the site
each month. During the last four weeks, there were sixty-four
waste shipment discrepancies noted: Ten against the site
criteria, sixteen related to the paperwork, eleven dealing
with equipment, eighteen related to improper packaging and nine
for contamination levels. Waste not solidified or packaged
properly, shipping papers which are either not legible or
incomplete, and contamination problems all hold up and add
costs to the disposal of your waste.
The environmental monitoring program consists of soil
and vegetation samples, environmental TLD's, air samples and
well water samples. The major route of migration off site
would be via the water migration path. For this reason, there
is an extensive well monitoring program. In addition to the
sumps and sample points at the bottom of the trenches, there
are also monitoring wells scattered throughout the active burial
area, around the perimeter of the site and off site. The
initial monitoring well at a specific location extends to
-------
137
the water table and is core sampled while drilling. The
core samples provide not only analytical data but also shows
the depth of any additional saturated sand layers that may run
through the clay bed. Monitor wells are installed at the same
location for each saturated layer. All wells are sealed,
grouted and capped to preclude any entry of contamination from
the surface. Prior to each sampling, the wells are pumped
down and allowed to recharge to provide a more accurate
indication of the present water condition. To date, only
very limited migration has been detected. Very close to one
trench, samples have shown organics and tritium contamination
only.
The burial area (approximately 235 acres) is owned by
the State of South Carolina and leased to Chem-Nuclear for
operation. The State's extensive environmental monitoring
program in the area will be maintained into the future
through the perpetual care fund. This fund was established
at the time of initial licensing, is controlled by the State
and is based upon the volume of waste disposed.
All areas - trench and well designs and locations;
solidification and packaging methods, environmental monitoring
and modeling, the perpetual care fund, offloading, safety and
health physics procedures, special techniques for volume
reduction and for increased efficiency in trench space
utilization are in a continuous state of review and improvement
by both CNSJ and State personnel.
I would like to extend an invitation to you to visit
our facilities in Barnwell if you are ever in the area. If anyone
has any questions, I will be glad to try to answer them.
-------
138
Based on 10.5 month, 1978:
Average Exposure
No. Avg. ( mrem/month)
H.P. Techs. 4 198
Office, Mgm., Supv.
(Includes Dispatcher,
Janitor, Warehouseman) 9 26
Offloaders 15 174
Truck Drivers 8 21
Equipment Operators 6 176
Maintenance
Personnel 6 6
(Mechanics, Welder
etc)
-------
139
CUBIC FEET/BY STATE/6 MONTHS JULY, 1978 THRU DECEMBER, 1978
STATE
Alabama
Arkansas
Arizona
Colorado
California
Connecticut
Delaware
Florida
Georgia
Hawaii
Illinois
Indiana
Iowa
Kentucky
Louisiana
Maryland
Massachusetts
Maine
Michigan
Missouri
Minnesota
Mississippi
Montana
New Jersey
New York
Nevada
North Carolina
Nebraska
North Dakota
New Hampshire
Oregon
Ohio
Oklahoma
Pennsylvania
Puerto Rico
Rhode Island
South Carolina
South Dakota
Tennessee
Texas
Utah
Virginia
Vermont
Washington
Washington, D.C.
West Virginia
Utah
Wisconsin
VOLUME
57,121.04
1,830.12
9.0
200.95
516.8
34,829.16
249.0
42,932.75
14,775.25
2.0
90,020.11
4.0
21,214.75
3,990.56
732.3
38,619.02
80,375.25
18,876.24
31,119.28
2,761.57
7,681.57
805.5
21.0
60,585.72
165,504.6
4.0
75,462.28
9,738.8
36.9
3,307.7
82.5
42,739.61
32.38
115,947.06
448.35
1,314.9
79,517.57
.6
46,862.47
5,469.63
11.9
33,431.76
5,683.6
88.2
1,947.89
1,576.13
40.0
3,381.98
5.2
.2
.0
.0
.0
3.2
.0
3.9
1.3
.0
8.2
.0
1.9
.4
.1
3.5
7.3
1.7
2.8
.3
.7
.1
.0
5.5
15.0
.0
6.8
.9
.0
.3
.0
3.9
.0
10.5
.0
.1
7.2
.0
4.3
.5
.0
3.0
.5
.0
.2
.1
.0
.3
TOTAL
1,101,903.75
99.9%
-------
1971 1972 1973 1974 1975 1976 1977
VOLUME BURIED - BARN WELL
1978
1979 £
-------
141
MANAGEMENT AND SURVEILLANCE OF A UNIVERSITY RADIOACTIVE WASTE BURIAL SITE
Philip K. Lee, Jamieson G. Shotts, and David L. Spate. Health Physics Services,
University of Missouri, Columbia, Missouri
Abstract
A radioactive waste burial site is operated at the University of Missouri
under the conditions of 10 CFR 20.304. Over 90% of the radioactive wastes
generated by the laboratories and clinics,- Exclusive of the Research Reactor
Facility, are economically disposed at this site. During the seven year oper-
ation about 200 cubic meters of low-level wastes containing about l.SxlO11
becquerels have been buried in 3.6 meter deep and 0.6 meter wide trenches. A
radiation surveillance program confirms that radiation levels in the vicinity
of the burial site are well within acceptable limits.
Discussion
All licensees of the Nuclear Regulatory Commission are permitted to dis-
pose of small activities of radioactive wastes by local land burial. General
provisions contained in Section 20.304 of Title 10 of the Code of Federal
Regulations (10 CFR 20.304) stipulate allowable burial conditions such as
nuclide activities, burial depths, and burial frequencies for the local land
burials. Recently it has been announced that the provisions for generally
permitted land disposals are under discussion and may be modified or withdrawn.
Hopefully any changes in the regulations will not completely eliminate the
opportunity for institutions to operate a local radioactive waste site for the
efficient disposal of low-level solid wastes.
The University of Missouri has operated a radioactive burial site within
the conditions of 10 CFR 20.304 for the past seven years. This designated
burial site is a nine-tenth acre fenced plot on a University research farm
located about five miles from the main portion of the Columbia Campus. A metal
building at the site is used for temporary waste storage and for processing
radioactive wastes scheduled for transfer to commercial firms for off-site
disposal; This building is also used to a limited extent for the temporary
storage of radioactive wastes and there is a freezer in the building for storing
contaminated animal carcasses that are to be buried.
The soil at the site consists of a water permeable clay layer to a depth
of about three meters. Below the surface soil layer is about one meter of
water impermeable clay and then about a one-half meter layer of weathered lime-
stone and clay over the limestone bedrock. Burial trenches which are about
four meters deep may penetrate into the impermeable clay region but will not
break through it. Surface water drains to the north across the site into a
natural drainage ravine near the farm boundary about 200 meters from the north
end of the burial site.
Most of the radioactive wastes generated at the University research and
clinical laboratories consist of slightly contaminated paper trash, plastic
vials, animal bedding, and animal carcasses. This high volume of low activity
waste is well suited for shallow land burial disposal. The wastes are collected
from the campus locations and accumulated in the waste storage areas on campus
-------
142
or in the storage building at the waste site. Burials are scheduled with
respect to weather conditions and the amounts of accumulated wastes.
Burial trenches are dug to a depth of about four meters by a University
operated backhoe. The width and length of an individual burial trench is
dependent upon the volume and types of wastes to be buried. Usually the
trenches are about sixty centimeters wide and four meters long; however,
trenches double this normal width are convenient for burial of large animal
carcasses. The location and size of a particular burial trench is recorded
on the map of the burial site as shown in figure 1 and referenced by a number
code giving the calendar year and burial number for that year, such that the
number 78-6 would indicate the sixth burial at the site in 1978.
Each item placed in the burial trench, whether a bag of laboratory trash,
animal carcasses, or container of animal bedding, is identified according to
the contaminating nuclide and activity. Records for each burial indicate
the activities of all radioisotopes, total volume of uncompacted wastes,
numbers of items buried and the fraction of an allowable burial in reference
to the allowable activity limits defined in 10 CFR 20.304. Other waste records
can be used, if required, to trace an item buried to its source laboratory.
Following burial the wastes are covered with at least 1.2 meter of dirt and
the burial located and coded on the site map.
There have been 40 burials at the waste site as of the end of 1978. These
burials represent over 95% of the solid radioactive wastes generated at the
Columbia Campus over the past seven years. The present waste site should
accommodate the University for the next fifteen years; however, the area can
be expanded if needed.
A total of 1.4X1011 Bq (3.87xl03 mCi) have been disposed by land burial
in the 40 burials over the past seven years. The uncompacted volume of the
disposed wastes is estimated to be 199 m3 (7027 ft3). Each burial is distinc-
tively different but an average burial would be 3.58x109 Bq (99.9 mCi) with
a volume of 4.96 m3 (175 ft3) and would represent 47.4% of an allowable burial
as defined in 10 CFR 20.304. A summary of the annual burial totals is given
in table 1.
The volume of waste buried each year has been fairly consistant ranging
from about 11 m3 to 39 m3 for the calendar year as shown in figure 2. However,
there has been a decrease in the activities from about 4.8x10 Bq being
disposed in 1973 to only 4xl09 to 9xl09 Bq being disposed in each of the past
three years as shown in figure 3. One of the reasons for the decrease in
disposed activity is that labeling procedures requiring relatively large
amounts of tritium are not being performed by campus users to the extent of
several years ago.
The residual activity of previously buried wastes is continuously being
reduced by radioactive decay. A computer program calculates this decay
reduction and indicates the radioisotope inventories in the site. Residual
activities have reached a plateau over the past two years due to decreased
levels of activities per burial and the decay of the previously disposed
radionuclides. The variation in residual activity over the past seven years
is indicated by figure 4. At the end of 1978 the total activity of the buried
wastes was calculated as being l.llxlQ11 Bq (3.0 Ci). Of the total activity
-------
143
tritium accounted for 97.4% and carbon-14 accounted for 2.49%. Other radio-
nuclides which were minor contributors to the residual activity include
scandium-46 at 0.04%, iodine-125 at 0.03%, calcium-45 at 0.02% and strontium-85
at 0.02%. Twenty-three other radionuclides are identified by the disposal
records as being represented at very low activities in the burial site. The
total activities of the different radioisotopes in the burial site at the end
of 1978 are listed in table 2. ;
Radiation surveys and environmental samplings have been conducted in the
waste site area to confirm that radiation levels are maintained within accept-
able limits. After the backfill of each burial, the surface exposure rate is
measured with a portable survey meter. There was only one measurement above
normal background recorded during a post burial survey. This elevated reading
was due to some items that had not been properly covered by the backfill
operation. The materials were collected and were properly disposed in a later
burial.
Surveys are also taken monthly at fifteen identified locations around the
site boundary and in the waste storage building. Elevated exposure rates in
and around the waste storage building are produced by radioactive wastes stored
in the building. These wastes are primarily from the Research Reactor Facility
and are being stored and processed before transfer to a commercial waste
disposal vendor. Exposure rates within the building range up to 40 mR/h while
the exposure rates at the site boundary nearest to the building have been as
high as 0.15 mR/h when the building is full of stored wastes. Other survey
points beyond the influence of the radioactive material in the storage building
are at normal background levels of less than 0.05 mR/h. Thermoluminescent
dosimeters positioned at the monitoring locations for extended time periods
confirm the exposure rate measurements.
Soil and grass samples were obtained in 1978 at nine of the monitoring
locations at the site and analyzed for radioactivity. No activity was observed
above 1 Bq/g of gamma activity or above 1 Bq/g of beta activity on any of the
samples. Soil samples from depths ranging to three meters were also taken at
the middle of the north boundary of the waste site. These soil samples, taken
in 1977, were also below 1 Bq/g of beta activity. The environmental sample
measurements confirm that the soil and vegetation have not been contaminated
with radionuclides from the burial site.
Water samples taken from the surface of the waste site and in the drain-
age area below the site showed neither gamma nor beta activity above back-
ground levels nor any tritium activity above the 400 Bq/Z detection level.
Ground water samples from three meter deep sample holes at the north edge of
the site did show slight positive tritium activity in 1977 when heavy rains
probably flushed the tritium from the burial pit near the sample point.
Concentrations of 10" Bq/fc of tritium were observed in the ground water from
the site at that time.
Operating costs of the waste site are minimal. The major expense is the
digging and backfilling charges from the University service furnishing the
backhoe equipment. These charges average about $40 per burial which are
small in comparison to the approximate $800 expense of having the wastes
removed by a commercial disposal firm. Personnel time and efforts for a local
-------
144
burial would be approximately the same as preparing the waste for shipment.
This time is estimated to be about 1.5 man days per burial. Packaging expenses
would be greater for a commercial shipment because of the cost of the containers.
Containment or packaging costs for local burial are minimal because certified
shipping containers are not required.
The local shallow land burial site has proven to be an effective means for
disposal of almost all of the low-level solid wastes generated by the University
campus. It has been demonstrated that a low-level waste disposal site properly
maintained and operated within the conditions of 10 CFR 20.304 is a safe,
efficient, and economical disposal method for large quantities of low-level
radioactive wastes. It is recommended that continued use of this type of
burial sites be allowed for institutions willing to accept the responsibility
for proper custodial maintenance.
-------
145
TABLE 1
Summary of Radioactive Burials
University of Missouri
Year
(Calendar)
1972
1973
1974
1975
1976
1977
1978
Totals
Burials
(Number)
1
6
3
6
9
8
7
40
Volume
(Cubic Meters)
11.5
30.6
13.4
31.2
35.2
39.1
37.5
198.5
Activity
(Becquerels)
1.23X109
4.84xl010
3.37X1010
4. 08x10 10
4.76X109
5.65X109
8.84X109
1.43X1011
-------
146
TABLE 2
Residual Activities at Burial
(December 29, 1978)
Site
Radioisotope
Hydrogen-3
Carbon-14
Sodium-22
Phosphorus-32
Sulphur-35
Chlorine-36
Calcium-45
Scandium-46
Chromium-51
Magnesium-54
Iron-55
Iron-59
Cobalt-57
Cobalt-60
Zinc-65
Selenium-75
Arsenic-76
Strontium-85
Rubidium-86
Molybdenum-99
Technetium-99m
Cadmium-109
Iodine-125
Iodine-131
Cerium-144
Terbium-160
Mercury-203
Lead-210
Unknown Beta Emitters
Activity (becquerels)
11
1.08x10
2.77X109
4.51X105
l.OlxlO7
9.51X106
6.99X106
2.64xl07
4.96xl07
7.20xl06
1.33x10*
5.40X105
3.96xl02
2.82X106
4.44xl05
2.96X105
2.36X10
2.02X105
3.06X107
2.07X106
7.40X106
7.10X105
l.llxlO5
Totals
l.llxlO11
-------
UNIVERSITY OF MISSOURI
RADIOACTIVE WSTE BURIAL GROUNDS
SINCLAIR FARM
HEALTH PHYSICS SERVICES
MV1SED I-IO-TTM. DRAFTSMAN &u>^
i-ao-nfe* SCALE r-io'
-------
FIGURE 2
148
|
o
15
3
UJ
h-
CO
o
LU
2
200
190
180
170
160
ISO
140
'30
120
HO
100
80
60
50
40
30
20
10
0
VOLUMES OF BURIED
RADIOACTIVE WASTES
CURRENT
YEAR'S
PREVIOUS ACCUMULATION
(corned forward)
1972 1973 1974 1975 1976 1977 1978
CALENDAR YEAR
-------
FIGURE 3
149
0)
3
cr
o
0)
CO
LU
CO
I
U.
O
1.5x10
II
1.4x10
II
1.3x10
II
II
1.2x10
l.lxlO11
I.OxlO11
.10
9x10
8x10
10
7x10
10
6x10
10
5x10
10
> 4x10
10
3x10
10
2x10
10
IxlO
10
0_
ACTIVITY OF BURIED WASTES
CURRENT
YEAR'S
Y/Ss
PREVIOUS ACTIVITY
(carried forward)
1972 1973 1974 1975 1976 1977 1978
CALENDAR YEAR
-------
150
FIGURE 4
0>
O
0>
6x10
5x10
^ 4xlO'v .
(A
UJ
CC
3xlOlu.
2x10 .
IxlO
10
RESIDUAL RADIOACTIVITY
OF BURIED WASTES
1972 1973 1974 1975 1976 1977 1978
CALENDAR YEAR
-------
151
A PRELIMINARY IMPACT ASSESSMENT
OF INSTITUTIONAL RADIOACTIVE WASTE DISPOSAL
R. Andersen
University of Colorado
T. Beck, L. Cooley, and M. McCampbell
University of Maryland at Baltimore
Abstract
A significant fraction of the low level radioactive wastes
which are buried in the commercial, shallow land burial sites in the
U.S. originate from non-fuel cycle sources. The primary radwaste pro-
ducers in this category include large medical and academic institu-
tions. This paper considers, in a preliminary manner, some of the im-
pacts of disposing of institutional radwastes via the same methods and
systems as are used to dispose of fuel cycle radwastes.
Nuclide content and activity concentrations of institutional and
reactor radwastes differ greatly. The varied physical and chemical
forms of institutional radwastes may not be compatible for burial with
reactor radwastes. Animal carcasses and other biological materials
may adversely affect containment and transport parameters in the bur-
ial trenches. The time and expense involved in the commercial burial
of institutional radwastes may not be commensurate with the actual
radiological content of the materials. Finally, this method of dis-
posal may not represent an optimum use of limited burial space.
Introduction
In 1975, 12% of the low level radioactive wastes which were bur-
ied at shallow land burial sites in the U.S. were institutional rad-
wastes (An78). The total percentage of low level radwastes which were
buried from all non-fuel cycle sources was 39% (Ho79). This paper
will discuss the impacts of institutional radwastes only, but many of
the points that are made could equally apply to all non-fuel cycle
radwastes (but with 3.25 times the quantitative significance).
Method of Identifying Impacts
Impacts are the result or consequence of utilizing a material i-
tem within a particular process (e.g. driving an automobile, mining
coal, or generating electricity with a nuclear power plant). In this
case, the impacts that are considered are the result of utilizing the
material, institutional radwaste, within the process of disposal by
shallow land burial at commercial sites. It is convenient to break
down the process and the material into distinct elements, and then to
identify specific impacts within a matrix of these elements. This is
shown in Figure 1.
The disposal process is broken down into three elemnts: Collect-
ion, transport and burial. Collection includes picking up the rad-
wastes at a use point within the institution (e.g. university lab),
processing and packaging the material at a central collection point
(e.g. health physics lab), and storing the material for shipment.
the extent of these processes varies according to the type of insti-
tut:i~v- -~'' ' ~~ ~s J-l"~ -~J- -nuclide use program.
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152
Transport includes either direct shipment of the radwastes from
the institution to the burial site, or, shipment to an intermediate
facility (i.e. commercial radwaste disposal company warehouse) for
consolidation into a larger shipment to the burial site. Burial in-
cludes offloading the material into the burial trench, backfilling
over the material, and the residence of the material at the site.
The institutional radwaste material in defined with three para-
meters: volume, form and radioactivity. The volume refers to the
volume of the disposal package, and not necessarily the volume of the
contents, unless otherwise specified.
Form refers to the physical or chemical form of the waste. This
includes scintillation vials, packaged as glass or plastic vials con-
taining organic scintillation fluids; biological materials, includ-
ing animal carcasses, excreta, tissue cultures, blood samples, etc.;
solid wastes, including paper, gloves, labware, etc.; and solidified
or adsorbed liquids, which may include aqueous solutions or organic
fluids. Radioactivity refers to the radionuclide and activity con-
tent of the waste.
Once the specific impacts have been identified, they are grouped
together into more general impact categories for a discussion of
their significance and possible alternatives. A schematic example is
shown in Figure 2.
Impacts of Institutional Radwaste Disposal
In Figure 3, the specific impacts have been grouped together
into three general impact categories: A discussion will be made of
each category. Where possible, quantitative data will be cited or
estimated for specific impacts in order to discuss their significance.
Resources impacts include direct and indirect costs, materials
consumed in the disposal process, and limited or non-renewable re-
sources used in the disposal process.
The acquisition and establishment of facilities and equipment
for processing, packaging and storing radwastes pose unique problems
for institutions. These items are considered as impacts because the
largest part of their expense is met through public funds (either in
the form of overhead in research money allocations, or as allocated
budget items from state or federal funds) or in medical fees (in the
case of hospitals). These funding resources are becoming limited
through inflation and leglislated restrictions upon taxes, state and
federal budgets and medical costs.
Comprehensive data on moneys spent on facilities which are spec-
ifically intended for radwaste management within institutions is not
available. However, costs for radwaste processing facilities built
in 1975-77 at several major universities ranged from $95,000 to
$165,000 (An77).
A lack of adequate facilities for compacting solid radioactive
wastes, processing liquids, and consolidating radwastes into large
shipments, particularly at smaller institutions, yields secondary
impacts as the result of increased volume, more unprocessed liquids
or organics at the burial site, and increased transport costs (which
are typically a function of the quantitv of ' '
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153
Personnel time expended in the collection phase of radwaste dis-
>sal has impacts similar to those of facilities and equipment. Funds
:e derived from the same sources. A lack of adequate staff results
i the same secondary impacts discussed above. It should be noted
lat often, the unavoidable demands of radwaste disposal, in terms
: time and money, often restrict other, "less necessary", radiation
ifety functions. "Less necessary"in this case is defined by the res-
jctive radiation safety officer.
During 1977 interviews with 15 institutional radiation safety
Eficers (An77), data was obtained regarding the amount of personnel
Lme spent on the collection phase of radwaste disposal (at the in-
titution). Dividing the time spent by the total volume shipped by
lat respective institution in the preceeding-,twelve months yielded
i average time allotment of 0.6 person-hours/ft3 of waste shipped.
scause all of these programs were at major institutions, an extra-
Dlation to the time expended by the entire institutional population
annot appropriately be made.
Containers used in institutional radioactive waste disposal are
ppically 55 gallon steel drums(An78). Thirty gallon steel drums,
Lberboard or wood boxes, or steel pails are also used. In order to
zmceptualize the number of containers that are consumed in the in-
titutional radwaste disposal process, it is convenient to translate
le total volume of radioactive wastes shipped for burial in 1977
274,433 ft3 - Be79) into an equivalent number of 55 gallon drums.
tiis yields 37,388 drums, or, a quantity which if placed in a single
ine, would extend for more than fourteen miles.
The cost of disposal containers varies according to the type of
Dntainer, whether it is new or reconditioned, and the quantity pur-
tised at one time (which is often a function of storage facilities) .
rices for 55 gallon drums ranges from $15 to $25 each. Using the
rum equivalency figure, above, yields a container cost of $560,000
3 $935,000 for 1977.
Shipping radioactive wastes consumes fuel, both for a local
adwaste pickup and for a cross country trip to a burial site. Be-
ause there are now only three sites available, some institutions
hip radwastes as far as 1,500 miles. Shipments may be made as
art of a combined load (with non-radioactive materials) via a
Deal shipper, or may be made in truckload quantities (76 to 154
5 gallon drums) by the institution or a local radwaste disposal
ampany. Quantification of the amount of fuel expended in this
tiase of disposal cannot be made with the available data.
The final resources impact to be considered is use of burial
pace. In view of increased public opposition to the establishment
E burial sites within their locality (e.g. as in New Mexico), and
ie increased time and capital requirements for establishing a site,
irial space must be considered a limited resource. Space which is
tilized for institutional radwastes cannot be used for the disposal
E other radwastes (e.g. reactor wastes).
In view of the low activity_concentrations of institutional rad-
istes, consisting primarily of JH, l*C and -LZDI, it is not clear that
ie disposal of these was*-** in a burial site represents an optimum
3e nf ^^^.^^^^^.x.ity concentration for institutional
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154
radwastes shipped for burial in 1975 was less than 0.03 Ci/m (An78) .
By comparison, the average activity concentration of reactor low level
radwastes shipped for burial in the same period was 1.49 Ci/mJ, con-
sisting primarily of mixed activation and fission products (Ep77).
The impact which is the least quantifiable at this point, but
may be the most significant, is that of environment. Although much
research is underway to assess qualitatively and quantitatively the
effects of organic solvents, liquids, and biological materials upon
radwaste containment and transport parameters in the burial trench
(e.g. Co— and Ep—), the full significance of introducing these
materials into the sites may not be known for some time.
What is known is that institutional radwastes are a source of
liquids, organics, and biological materials in the trenches. In
1977, approximately 486,134 liters of scintillation fluids, in the
form of scintillation vials, 31,834 ft3 of adsorbed and solidified
liquids, and 27,718 ft3 of biological wastes were shipped to the
burial sites (Be79). It should be noted that the burial facility
at Barnwell, SC, is currently burying non-fuel cycle radwaste in
seperate trenches from fuel cycle radwaste (Oa79).
Health and safety impacts include personnel exposures to rad-
iation, chemicals and bio-materials and the potential for accidents,
including fire, expolosion or traffic accidents. Radiation exposure
may be the least significant of these. Typical annual exposures
of personnel handling radwastes at 15 institutions were less than
100 mRem/yr (An77). Although such a small sample cannot be extra-
polated with great confidence upon the total institutional popula-
tion, such small exposures do seem consistent with the low activity
concentrations of beta and low energy gamma emitters in the waste.
The noxious and toxic properties of handling toluene are well
documented (He73). A lack of adequate facilities to provide per-
sonnel protection from organic solvent vapors or noxious fumes from
biological wastes often results minimal handling and processing of
these materials at the institution. This is reflected in drums of
unabsorbed/unsolidified liquids which occasionally reach the burial
site (An79 and Oa79).
Fire, explosion and traffic accidents also occur in the dispos-
al of institutional radioactive wastes. Two examples from first-
hand experience as an employee of a commercial radwaste disposal
company will typify these impacts. A fire in a radwaste storage
area at a research institution caused several thousand dollars of
damage to the facilities. Fire personnel attributed the cause of
the fire to spontaneous ignition of toluene vapors from the thirty
drums of scintillation vials stored in the area.
In a second instance, a tractor trailer, carrying a load of
institutional radwaste to the burial site, wrecked in the Washing-
ton, D.C. area. Although the release of materials within the trail-
er was minimal, and no radioactive contamination outside of the
trailer was noted, a national newspaper headlined the event the
following day as "Atom Truck Crashes on Capitol Beltway".
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155
Alternatives
A comprehensive assesment of alternatives is beyond the scope
of this paper. However, three alternative do warrant a brief dis-
cussion.
In addition to its impact at the burial site, the disposal of
nominally contaminated scintillation fluid represents an opportun-
ity cost. Recycle of scintillation fluid would reduce the impact
of the material at the site, as well as reduce the demands for pro-
duction of the scintillation fluids. Industrial scale recycling
of the fluid is not currently nted. However.- small scale recycle
through evaporation-distillation has been used at at least two
universities (Lu77 and Ha77). No quantitative data on the effic-
iency of this method is available.
A less sophisticated recycle procedure was noted at several
universities (An77), and could easily be applied in many situa-
tions where scintllation counting is used. This involves a re-
use of scintillation vials (and fluid) which contain activity
levels below a predetermined limit (dependent upon the application
for which the recycled vials are intended. Where the samples con-
tain non-miscible materials (e.g. contamination survey wipes), a
recycle of 50% of the vials is common.
Incineration of waste scintillation fluid is another alter-
native. The potential impacts of the organics in the trenches and
a reduction in volume of waste buried would result. If the incin-
eration facilities were on-site or within the region, transport-
ation impacts would be reduced. Additionally, packaging the mat-
erial for incineration, rather than containment, would allow the
use of less expensive, or recyclable shipping containers.
Estimated activity concentrations in waste scintillation fluid
are 7xlO~3 Ci/ml of 3H and 0.24 Ci/ml of 14C (Gr77). Applying
these activity concentrations to the entire volume of scintillation
fluid (in vials) shipped in 1977 (486,134 liters! results in a
total activity of 3.4 Ci of 3H and 0.24 Ci of 14C. These levels
of activity would seem to make incineration a reasonable alter-
native, at least from a radiological point of view.
Local or regional burial sites for non-fuel cycle radwastes
is another alternative. Some institutions already practice this by
on-site burial, or by burying radwastes in a local landfill (An78).
Regional waste facilities would reduce or eliminate the transport
and burial impacts. They could include an incineration facility
to reduce some of the collection impacts.
Conclusion
The current direct cost of institutional radwaste disposal is
$10.90/ft3 (Be79). This means a total 1977 expenditure of nearly
$3 million for containers, transport and burial. Assuming that in-
stitutional radwastes still only comprise 30% of all non-fuel cycle
radwastes, as they did in 1975 (An78), then the total direct costs
for non-fuel cycle radwaste disposal would exceed $10 million. This
does not include personnel, facility and equipment costs, much of
whic - - -—-—.".-.-,•",- -ublic sector.
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156
Assuming that only 35% - 40% of the total monies spent on non-
fuel cycle radwaste disposal are for burial, this yields a total
burial expenditure of no more than $4 million spent in burial costs
in 1977 by non-fuel cycle sources. Regional waste disposal concepts,
whether burial, incineration, or some combination of the two, would
further reduce these monies to about $800,000/site, assuming only
five regional sites. Considering only institutional radwaste sources
would reduce this to $250,000/site. Whichever figure is used, the
question is, is this revenue source sufficient incentive for in-
dustry to do the planning, evaluation and capital outlay necessary
for the establishment of such sites?
Another alternative would be to reduce bural site impacts by
restricting, through regulation, the allowable forms of radwaste
for burial (e.g. solidification of scintillation fluids). This
would make impacts within the institution, in the form of additional
personnel and facilities requirements, more acute. More likely,
such waste processing as would be required would be taken over by
intermediate radwaste disposal companies, which would greatly in-
flate radwaste disposal costs.
It is clear that more consideration of this problem-is needed.
That is why this paper is only-"A"Preliminary Assessment ..."
References
An77 Andersen R., unpublished data from interviews of 15 in-
stitutional radiation safety officers during the study
referenced in An78.
An78 Andersen, R., Beck, T., Cooley, L., Straus, C., Institu-
tional Radioactive Wastes, NUREG/CR0028, University of
Maryland at Baltimore, March 1978.
Be79 Beck, T., Cooley, L., McCampbell, M., Andersen, R., In-
stitutional Radioactive Waste-1977, University of Mary-
land at Baltimore, (forthcoming).
Co— Colombo, P., Weiss, A., Francis, A., Evaluation of Isotope
Migration; Land Burial Water Chemistry of Commercially Op-
erated Low Level Radioactive Waste Disposal Sites, Quarterly
Reports, Brookhaven National Labs, Upton, NY, (ongoing).
Ep— U.S.EPA, Office of Radiation Programs, Radiological Mea-
surement at the Maxey Flats Radioactive Waste Burial Site
11974 to 1975, USEPA-520/5-76/020 (ongoing study)."
Ha77 Harwood, G., University of Southern California, personal
communication, Spring 1977.
He73 U.S.Dept of H.E.W., Criteria for a Recommended Standard
. . . Occupational Exposure to Toluene, USDHEW, HSM 73-
11023, 1973.'
Ho79 Holcomb, W., A Summary of Shallow Land Burial of Radio-
active Wastes at Commercial Sites Between 1962 and 197iL
with Projections, Nuclear Safety, 19:1. Jan-Feb 1978.
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157
Gr77 Granlund, R. , Incineration of Waste Scintillation Fluid",
presented at "Management of Low Level Radioactive Waste
Symposium", Atlanta, GA, May 23-27,1977.
Lu77 Lundberg, R. , San Diego State University, personal commun-
ication, March 1977.
Oa77 Oakley, H., Chem Nuclear Systems, Barnwell, SC, personal
communication, 1979.
RADWASTE PARAMETERS
DISPOSAL
PROCESSES
Collection
Transport
Burial
Volume
Facilities
Personnel
Containers
Fuel
Accidents
Bur. Space
Form
Chem Exp.
Fire/Expl.
Fire/Expl.
Bur. Trench
Activity
Rad . Exp
Integrity
Figure 1. Matrix of Radwaste Parameters and Disposal Processes
with Specific Impacts
PARAMETER
(Volume)
4
PROCESS
(Burial)
w
H
w
o
PH o
(0
W ft
u e
D O
Q U
SPECIFIC IMPACT
(Burial Space)
GENERAL IMPACT
(Resource)
SIGNIFICANCE
(12% of Volume
Buried in 1975) u
w C
en 0
W -H
U -P
O (0
« M
CM (U
C
W-H
o u
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158
RESOURCES ENVIRONMENT HEALTH & SAFETY
Personnel Burial Trench Chem. Exposure
Facilities Integrity Fire/Expl.
Containers Accidents
Fuel
Burial Space
Figure 3. Impacts of Institutional Radioactive Waste Disposal
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159
THE STATUS OF LOW-LEVEL RADIOACTIVE WASTE DISPOSAL-
HOW TO PLAN A DISASTER!
Wilson C. McArthur
Hittman Nuclear and Development Corporation, Columbia, Maryland
Abstract
The nuclear industry is faced with serious problems in the transportation
and burial of low-level radioactive wastes. Soaring burial costs, state regu-
lations regarding transportation routes, and lack of direction from regulatory
agencies are problems that must quickly be resolved.
In order to gain control of this situation four major steps must be
taken. First, states must accept their fair share of responsibility in the
"waste" problem. Regulatory agencies must recognize the seriousness of the
problem and develop a schedule for action. The nuclear industry must assert
itself in a positive manner regarding the safety of nuclear power, and the
low-level waste burial ground situation must improve.
Introduction
The present status of the low-level waste scene is quite complicated and
appears to be headed in the direction of more confusion.
Table I outlines the current nuclear reactor status given by the Atomic
Industry Forum (AIF) as of November 29, 1978. There are a total of 203 reactors
which have operating licenses, construction permits, or are on order. These
203 reactors total 197,918 MWe. In addition, Commonwealth Edison Company
(CECO) recently placed an order for two 1,150 MWe nuclear units from Westinghouse.
However, at the same time, Public Service Electric and Gas Company cancelled
all four of its floating platform nuclear plants. Projections had previously
indicated that only two domestic nuclear reactors would be ordered during
1979; those two were the CECO plants which have already been ordered. There-
fore, the prospects for nuclear steam supply system vendors during 1979 appear
to be bleak indeed.
In a recent AIF study, it was indicated that a normal PWR produced 40,000
ft3 of solid low-level waste per year. A BWR produces 55,000 ft3 per year.
Considering the number of plants now having operating licenses, this would
result in a range of 2.8 to 4.0 x 106 ft3 per year to be shipped to licensed
burial grounds. As a comparison, the amount of low-level solid waste produced
by a PWR in a 40-year life of a plant is 1.6 x 106 ft3. An average-size PWR
containment vessel has approximately five million cubic feet of space. The
low-level solid waste produced by this plant would fill approximately 32
percent of the containment in 40 years. One may ask why such a comparison?
The response is that a recent on-site storage study considered a decommis-
sioning alternative of placing the low-level solid waste in "fixed" burial
facilities, i.e., the containment, after retrieval from on-site burial facili-
ties. This simple arithmetic is provided so that low-level waste from an
operating nuclear power plant can be gained.
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160
The attitude in our industry today is very similar to that of being in
"limbo." For example, the regulatory agencies appear to be moving slowly in
making decisions regarding low-level waste. One only has to look at the
status of high level waste to know that there is chaos; and low-level waste is
following suit. It is my feeling that low-level waste could be the "tail
wagging the dog." As an example of the lack of recognition of the magnitude
of the problem, one DOT official was recently asked, and it was reported in
Nucleonics Week, why there was not an official government position regarding
transportation of radioactive waste. His response was, "I personally have not
seen anything to indicate an emergency. I doubt that the DOT will jump into
the fray to make a precipitous decision." Further, in Nucleonics Week a
utility executive recently stated that, "The nuclear option is dead, there is
no question about it. We at Consumers Power continue to plan nuclear plants
but each year we wipe them off the planning board—they are too expensive."
It is obvious that there is confusion and lack of direction in regard to
low-level waste.
Burial Grounds
What are the problem areas facing this industry? First and foremost,
there must be a burial site for low-level waste if we are to bury the waste.
Those that are familiar with this industry know that the low-level waste
burial ground history has been quite confusing. Table II shows the history of
low-level waste burial sites. For example, in 1975, there were six operating
burial sites. Currently, only three of these sites are operating; two west of
the Continental Divide and one in the eastern part of the United States. The
Barnwell site located in South Carolina recently considered placing a limitation
of 135,000 ft3/month on shipments into the burial site. Although this limitation
has not as yet been imposed there is always the possibility that the State of
South Carolina will become more concerned about taking most of the waste east
of the Mississippi. Also, the State of South Carolina is beginning to scrutinize
what comes into the site. For example, recently, the State placed a hold on
receiving oily waste shipments. The State is currently evaluating the various
methods of solidifying oily waste and feel that until such criteria is set,
the site can no longer receive this type of waste. The State of South Carolina
is also beginning to look at organics contained in liquid scintillation vials.
There is always the impending threat of placing a hold on burial of urea-formal-
dehyde solidified low-level waste. In addition, Chem Nuclear Systems, Inc.
(CNSI), recently suspended efforts to open a burial site in New Mexico due to
licensing problems. Nuclear Engineering Company (NECO) has run into continual
red tape in attempting to license additional space for Sheffield. It was
reported in Nucleonics Week that one NECO official said, "...it's paralysis by
analysis." There are two reasons—safety and financial—why Maxey Flats will
probably remain closed. First, Kentucky will insist that state and federal
studies, now under way, "prove that it is 100 percent safe to bury low-level
waste at the site before operations resume. Second, the state would have to
charge taxes high enough to discourage private industry from operating the
site in order to maintain a profit and ensure long-term care.
However, there is some attempt to open a low-level waste burial site in
Lyons, Kansas at the site of the old high-level waste salt mines. The opening
of other low-level waste burial sites would be of benefit to the nuclear
industry.
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161
Figure I shows the increase in burial costs, another problem faced by the
industry.- During 1978, burial costs at the Barnwell, South Carolina site
increased approximately 85 percent for waste in the 0-200 mRr/hr range. This
is a significant price increase. On December 28, 1978, users of the Barnwell
site were notified of additional price increases, primarily for higher level
waste, various handling and surcharges. At this point, the cost of burial is
becoming an important factor in the budgets for operating nuclear power plants.
Transportation
A second and perhaps even more significant problem, is that of trans-
portation. The routes to get to the burial sites are being threatened and
costs are increasing. Restrictive statutes and ordinances have been adopted
in over 50 states and localities. For example, Clergy and Laity Concerned
(CALC), a group of northwest nuclear opponents, hope to choke off transportation
routes to Hanford. Their aim is the Hanford facility itself, but the impact
will be to stop the receipt of radioactive waste at the low-level burial
grounds.
Figure II shows the increase in the cost of transportation. As one can
see, the increase from 1973 to 1978 has been approximately 30 percent. As a
comparison, Table III gives a simple analysis of the impact of the cost of
transportation of low-level radioactive waste for one particular nuclear plant
located in the midwest. This analysis is for a shipping cask that contains
approximately 170 ft3 of waste. Before the Sheffield, Illinois facility was
closed, the plant was paying about $400 for tranporting the waste to the
burial site. When Sheffield closed the cost for transportation resulted in a
four-fold increase for shipments to Barnwell. If the Barnwell facility were
to close, the cost for transportation would be about an eight-fold increase.
This analysis does not include the increased probability of an accident on the
highways. This cost increase does not parallel the cost increase received
from the burial grounds. It is my feeling that the transportation problem
could become the major problem facing the industry due to the fact that many
states are considering the banning of routes for transporting any type of
radioactive waste through their localities and states.
Equipment Costs
Due to significant efforts, and changes in design, one radwaste hauling
vendor has been able to control or practically maintain the costs for some of
his shipping casks over a five-year period. This, of course, is not the case
with most pieces of equipment. The escalation rate has been averaging approxi-
mately eight percent per year.
Economics
Table IV is a recent analysis of how different costs have risen since
1967 and since one year ago this January. The point is that transportation
*The top curve is the price increases for western burial sites operated by
NECO. The cottom curve is the price increases for the Barnwell site operated
by CNSI.
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162
and equipment costs are falling within the ranges shown in Table IV. The
recent burial price increases have indicated a potential "run-away" situation.
With increases of approximately 85 percent for the 0-200 mR/hr range waste
occurring during 1978, one must keep a cautious eye on the future if he is to
budget this portion of operating costs adequately.
Solutions
What must be done to effectively get the nuclear industry out of the
dilemma that we now face? I believe that there are basically four major steps
that must be taken.
1. First of all, states must accept the fact that they are involved in
the production of radioactive waste if nuclear plants are located in
their states. Most states produce medical and institutional radio-
active waste and all states produce some toxic chemical wastes. An
official of the State of South Carolina recently related that the
state was doing its share by accepting low-level radioactive waste
and it was time that someone else handled the other types of waste,
such as toxic chemical waste. I am suggesting that a serious look
be taken by the states and that their involvement in production of
radioactive and toxic chemical wastes be reviewed. The states must
somehow become involved; either by providing a low-level waste
burial ground, a medical/institutional radioactive waste burial
ground, a toxic waste burial ground, or by entering some cooperative
venture. A great deal of cooperation would be required by the
states, but it is a simple fact that all of the states make a con-
tribution to the waste produced and they should share in the handling
and burial of these wastes.
2. The second major step is that the regulatory agencies must not only
recognize the peril of not solving the low-level waste problem, but
that they must develop a schedule to seek answers to these problems
before the problem becomes even more serious. For example, the DOE
is considering a contingency plan to take low-level waste at its
burial sites in emergency situations. However, to my knowledge, no
real plan has been placed in effect and if Barnwell were to close
for some particular reason, the availability of casks to ship low-level
waste to Beatty and Hanford would create a real problem for the
nuclear industry.
3. The third major step is that the nuclear industry must go on the
"warpath." There is enough evidence to show the economic advantages
and safety of nuclear power that a more positive attitude must be
taken. Is nuclear power really cheaper? The New England power
plants have been producing power for 1.293 cents/kwh over the past
two years compared with 2.662 cents for oil-fired capacity. Table V
gives recent cost estimates from Ebasco Services, Inc. indicating
their estimates of nuclear versus fossil over a 10-year period. The
nuclear plants are currently estimated to cost approximately 1,648/kw
vs. 1,226/kw for fossil plants. Most people recognize that the
nuclear power question is now a political rather than a technical or
environmental one. We must direct our warpath towards the political
arena and become as positive and outspoken as the anti-nuclear
proponents.
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163
4. The fourth major step is to improve the burial ground situation.
There are basically two alternatives as far as I can see. First, we
can continue with commercially-operated sites and, second we could
proceed with government-controlled burial sites. At this time, I am
not sure which is the best direction to take. It is obvious that
there could be some advantage now to having government-controlled
burial grounds because the ability to open government burial sites
would be easier. However, there are disadvantages that must be
considered. This question must be studied and a decision made prior
to a decision being forced upon the industry.
Conclusion
The bottom line is that unless some steps are taken, an already serious
problem could become a disaster. There are power plants that have such limited
storage capacity that if Barnwell were to close down, and if radwaste disposal
vendors could not provide adequate shipping casks for hauling the waste to the
West Coast, these plants would be faced with closing down due to radwaste
disposal problems. It is about time that we take this situation in hand
before it becomes a much more serious problem. We have an opportunity to do
just this if we will step forward and speak out, talk to our Congressmen and
write letters to the appropriate authorities making these decisions.
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164
TABLE I
NUCLEAR REACTOR STATUS REPORT
72 Reactors with operating licenses 52,273 MWe
90 Reactors with construction permits 98,968 MWe
37 Reactors on order 42,565 MWe
0 Letters of intent/options 0
203 Total 197,918 MWe
TABLE II
HISTORY OF LOW-LEVEL WASTE BURIAL SITES
Name
West Valley
Maxey Flats
Sheffield
Barnwell
Hanford
Beatty
Location
New York
Kentucky
Illinois
South Carolina
Washington
Nevada
Operating
No
No
No
Yes
Yes
Yes
Comments
Operations suspended
in March, 1975
Operations suspended
in December, 1977
Operations suspended
in April, 1978
135,000 ft3 per month
limitation imposed
Chem Nuclear
SWECo
New Mexico
Kansas
No
No
Licensing effort
suspended
Licensing application
in
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165
TABLE III
TRANSPORTATION COSTS FROM A MIDWEST PLANT
Plant to Sheffield, Illinois
Plant to Barnwell, South Carolina
Plant to Beatty, Nevada
Approximate Cost
$ 400
1,700
3,400
Item
TABLE IV
HOW DIFFERENT COSTS HAVE RISEN
Percent since
1967
Overall living costs
Food, including meals out
Housing and operations
Transportation
Clothing and upkeep
Medical care
Personal care
Entertainment
101
112
110
90
63
125
86
79
Percent since
Year Ago
8.8
11.3
9.8
6.6
4.1
8.5
6.7
4.7
Year
1969
1978
TABLE V
COST ESTIMATES*
Nuclear**
226/kw
1,648/kw
Fossil
183/kw
1,266/kw
*Ebasco Services, Inc.
**Regulatory changes account for 78 percent
of soaring nuclear plant costs.
-------
5 PER
CU. FT.
S.OOr
.00
3-00
2.00
i .OCr
FIGURE I
BURIAL PRICE INCREASES
X • DATE ON WHICH PRICE
INCREASE ANNOUNCED
NEW YORK
SITE CLOSED
MARCH II, 1975
51-'0 March It
C-200 mr/hr 1975
1975
ILLINOIS
SITE CLOSED
APRIL 9, 197S
.63
nr-lR/hr
S2.50 "1
SI. 80
S
•30
Dec. 12
$1.
f
H-1
S'-75 April
1976
KENTUCKY
SITE EFFECTIVELY CLOSED
JUNE 19, 1976
$2.58
KENTUCKY
SITE OFFICIALLY CLOSED
DECEMBER 1977
S2.95
$2.10
Dec. 6,
1977
$4.45
3.65
Apr!I 2
1978
JULY I.JAN. 1, APRIL 1.
197«. 1975 1975
JAN. 15,
1976
MAY 15,
1976
JAN. 1.
1977
JAM. 1.
1978
JUNE 1.
1978
EFFECTIVE DATE OF INCREASE
-------
in
o
o
o
00
•c
o
o
X
u
COST OF TRANSPORTATION
1973
1974
1975
YEAR
1976
1977
197S
CT»
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168
SHALLOW LAND BURIAL--WHY OR WHY NOT?
Warren T. Thompson, Civil Engineering, The University of Texas at Austin
(presently employed by Union Carbide Corp., Nuclear Division, Oak Ridge,
Tennessee)
Joe 0. Ledbetter, Civil Engineering, The University of Texas at Austin
Gerard A. Rohlich, Civil Engineering, The University of Texas at Austin
Abstract
This paper summarizes a master's thesis on the state-of-the-art for
shallow land burial of solid low-level radioactive wastes. The coverage of
the thesis, which is condensed for this paper, ranges from site selection to
problem case histories. Inherent in such coverage is the assessment of risk,
the discussion of operational and management problems and the real signi-
ficance of off-site migration—this topic will be discussed in light of the
stands taken that the migration is a serious problem and that it is not.
Emphasis is on the engineering parameters of importance in site selection, and
what pretreatment, if any, is needed.
Introduction
This paper will present considerations which should be included in making
the political decision to bury or not to bury. Such a decision should be made
with a background of what has been done, what problems have occurred, what
alternatives exist, and what risks are entailed.
History of Shallow Land Burial
Shallow land burial (SLB) of low-level radioactive wastes has been practiced
in the United States since the advent of the atomic age in the 1940s. SLB
competed with disposal at sea until the 1960s when the latter practice was
abandoned (Len67). Starting in I960, the Atomic Energy Commission (AEC) set
up criteria for licensing commercial land burial sites (USC76). These criteria
were:
1. "A written commitment from a responsible state official that the
state would assume control over the burial site in event of default
or abandonment of the site by the commercial operator."
2. "The geological and hydrological characteristics of the site must be
such that containment of the waste materials is assured in a manner
that will not endanger public health and safety."
3. "The waste must be in solid form prior to burial."
4. "Establishment of an environmental monitoring program."
5. "The packages in which the wastes are transported to the burial site
meet the NRC and DOT standards for packaging and transportation."
(Note: NRC is the Nuclear Regulatory Commission and DOT is the
Department of Transportation.)
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169
6. "As part of the licensing process, any future burial grounds licensed
by the NRC would require a review under the provisions of the National
Environmental Policy Act (NEPA)."
There have been 6 commercial sites licensed since September 1962; they
are: Beatty, Nevada; Maxey Flats, Kentucky; Richland, Washington; and Sheffield,
Illinois (Nuclear Engineering Co., Inc.); West Valley, New York (Nuclear Fuel
Services, Inc.); and Barnwell, South Carolina (Chem-Nuclear Systems, Inc.)
The Barnwell site is now receiving about 80% of all the wastes being buried
commercially.
West Valley and Maxey Flats are closed due to operational problems and
Sheffield has used up its authorized space. In addition to the commercial
sites, there are 5 major Department of Energy facilities which have shallow
land burial. They are Oak Ridge National Laboratory, Oak Ridge, Tennessee;
Los Alamos Scientific Laboratory, Los Alamos, New Mexico; Idaho National
Engineering Laboratory; Idaho; Hanford, Richland, Washington; and Savannah
River Plant, South Carolina.
The restrictions on wastes accepted at commercial sites include packaging
of the wastes, solidification of liquids and sludges, gases only in containers
(if at all), limits on transuranics, limits on fissile material, and limits on
chemical toxicity.
Problems With Shallow Land Burial
Problems arising at low-level solid radioactive waste burial grounds may
be site-specific, dependent on the regional hydrologic, geologic, and climatic
conditions, or they may be common to all sites. The major problem has been
the migration of radioactivity away from the burial sites in concentrations
that have caused political concerns although knowledgeable radiological health
professionals have stated that the observed levels of contamination posed no
threat to the safety and health of the public (Me76b). Migrations have occurred
at 4 sites: 2 commercial, Maxey Flats and West Valley; and 2 Federal facilities,
Oak Ridge National Laboratory and Idaho National Engineering Laboratory (USC76;
Me76b).
The migration problems have probably been a result of the emphasis on
economics and convenience in the selection and operation of the disposal
sites. The criteria utilized for site selection included factors such as
isolation of the site from population centers and water supplies, depth of
burial to minimize radiation at surface, limits on form and type of waste, and
operational procedures. The only reference to geologic or hydrologic con-
siderations was the specification of the depth of the groundwater table (USC76).
The migration of radioactivity at Maxey Flats has been attributed pri-
marily to the accumulation of water in the trenches, both those which had been
completed and those still receiving waste, and the subsequent hydrological
transport of the radioactivity from the site. In the period from 1972 to
1975, the State of Kentucky and the licensee were able to correct most of the
problems (Co76).
A problem of diversion of contaminated equipment that was to be buried
has received wide publicity in connection with the Beatty site. The State of
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170
Nevada found that tools, clocks, and other contaminated equipment had been
removed from the site by employees. The license was suspended until the
situation was corrected (USC76).
Other problems have related primarily to poor management procedures in
that maps and records do not exist; they were never prepared or they burned.
There have been some problems with fires and explosions. There is no record
of any problems of migration or effects from the possible emanation of gases
from burial sites. Shrinkage of the buried wastes resulted in accumulation of
surface water in the burial trenches at West Valley (Mo68).
Mobilization of the Waste
Solid radioactive waste disposal in terrestrial environments is subject
to infiltration and leaching by ground and surface waters. The initial result
of water contacting the waste would most likely be formation of leachate
having the approximate characteristics of ordinary commercial landfill leachate
plus the presence of radioactivity.
Because of the low density of the waste, oxygen is available for aerobic
decomposition during the early stages of organic breakdown; this decomposition
yields carbon dioxide, water, and nitrate. As the oxygen supply diminishes,
anaerobic decomposition produces methane, carbon dioxide, water, organic
acids, nitrogen, ammonia, and sulfides of iron, manganese, and hydrogen.
Carbon dioxide production, as an example, could dissolve strontium in the
bicarbonate form (Gia77; Ma65).
Other characteristics of landfill leachate include high chemical oxygen
demand (COD, up to many 1000s of mg/1), volatile acids, and a pH in the range
of 3.7 to 8.5. Constituents which release or mobilize contaminants from the
waste include acetic, propionic, isobutyric, and valeric acids. In brief,
landfill leachate contains agents capable of solubilizing cobalt, strontium,
cesium, and other radionuclides, including some of those generally considered
relatively insoluble such as plutonium (Me76a,b).
Alternatives for Waste Disposal
Treating and/or packaging of low-level radioactive waste prior to burial
can serve several purposes, including (1) volume reduction; (2) the reduction
of the mobility of the radioactivity; (3) enhancement of the safety associated
with the handling and disposal; and (4) recovery of plutonium and other trans-
uranics (Gil77). The most burdensome problem facing SLB is not the radio-
activity of the waste but the large quantity of waste (NRC76). This problem
is likely to get worse because of opposition to the establishment of new sites
(NY76).
Several methods are available for the pretreatment of waste before disposal.
These include sorting, incineration with or without pneumatic classification
beforehand, baling and compaction with or without prior shredding, and acid
digestion (Gil77; NRC76). Wastes may be hand sorted at its source or at the
disposal site into combustible/noncombustible glass, metals, hazardous chemicals/
radioactive, and short/long lived radionuclides (Gil77; NRC76).
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171
Incineration is perhaps the ultimate volume reduction technique short of
chemical separation and concentration. Volume reductions of 70 to 90 percent
can be achieved (Mc70). Other advantages of incineration include the stabili-
zation of the organic waste, a product with better understood properties for
better leachate prediction and/or prevention, and the end of the threats of
long-burning, subterranean fires at the sites (NRC76). Incineration has
received widespread acceptance in Europe, but in the Unitd States, the attitude
exists that the extra volume reduction often does not justify the additional
cost (Mc70). Because baling and compaction offer relatively low cost volume
reductions to 75 percent while requiring little segregation or sorting, they
have been more attractive than incineration in the United States (Mc70; St75).
Acid digestion is still in the pilot plant research stage (Gil77).
In addition to pretreatment schemes, there are probably viable alterna-
tives of engineering the containment of the wastes regardless of whether any
pretreatment has been applied. These alternatives include the catchment and
treatment of the leachate (especially the early flows), the interruption of
the water access to the burial trenches and/or the wastes and the sorption of
the radionuclides from the migrating water.
The catchment and treatment of leachate can be accomplished by preleaching
the wastes in containers or pits, by putting the trenches on an impervious bed
of clay or plastic, by pumping the waters in or beneath the trenches, or by
leaching the pretreated waste, especially the ashes from incineration. The
interruption of the water access to the wastes can be accomplished by cutoff
structures or by encapsulation of the wastes. The sorption of the radionuclides
from the migrating water cannot be relied on to provide total containment of
all the waste radionuclides; however, sorption may give enough holdup to
reduce the migration rate to acceptable values.
Current research at several locations throughout the United States,
including the University of Texas at Austin, is aimed at providing satisfactory
engineering designs for shallow land burial of low level radioactive wastes.
Risk and Risk Assessment
The design of a waste disposal site is based on either complete containment
of the waste, thereby allowing for zero planned discharge to the environment
and subsequently low risk to the environment, or the design may be based on a
finite amount of radioactive contamination leakage. The leakage would have
maximum limits based on the existing or proposed risk criteria developed by
the regulatory agency. The level of risk allowed will tend to dictate the
disposal practice for low-level radioactive waste.
Two of the major problems with risk assessment of a radioactive waste
burial site are: first, scientist and health physicists are unable to determine
the exact dose which would prove to be innocuous by both the genetic and
somatic definitions (NAS72; Led65). This problem centers around the how safe
is safe enough" cliche. Secondly, the design period of the burial sites can
only be verified by historical data, accelerated life testing or models based
upon the experience of similar systems. This is a major problem with radioactive
waste burial sites due to the long hazardous lifetimes of the radioactive
waste (USE77; USE78a,b).
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172
The range of possible risk associated with the disposal of low-level
radioactive waste is related to the amount of money spent on ensuring contain-
ment of the radioactivity. The range of risk could be extreme. High risk
could be the result of very little research and maintenance of a disposal
site, and low risk could be the result of encapsulation and/or solar disposal.
Outside of the possibility of a moratorium being placed on all waste production,
the risk from future waste cannot be entirely eliminated.
Conclusion
Why or why not utilize SLB as a method for disposing of low-level radio-
active waste must be decided in the near future. Maxey Flats, Sheffield, and
West Valley, three of the four eastern disposal sites, have already quit
receiving waste, resulting in 80% of all low-level radioactive waste being
sent to Barnwell.
The reasons why SLB should be utilized include economics and simplicity
of site operations. Yet, these two advantages have also been major sources of
problems at existing sites. Economics, being the culprit, in that dollars
were saved by not practicing thorough site selection techniques, and by inadequate
engineering design of the burial sites. Simplicity of site operations has
resulted in improper disposal practices and lax operational management programs.
The past operational practices and problems, together with results from
ongoing research should provide enough data to allow for the proper design and
operation of disposal sites so as to ensure safe containment of the radio-
activity. Utilization of pretreatment and volume reduction techniques together
with proper site management should allow for optimal use of areas allocated
for disposal operations, and result in minimum risks to public health and the
environment.
References
Co76 Comptroller General of the United States, 1976, "Report to the Congress:
Improvements needed in the Land Disposal of Radioactive Waste - A
Problem of Centuries," U. S. General Accounting Office, Washington,
D. C.
Gia77 Giardina, P. A., DeBonis, M. F., and Eng, J., 1977, "Summary Report on
the Low-Level Radioactive Waste Burial Site, West Valley, New York,
(1963-1975)," U. S. EPA.
Gil77 Gilmore, W. R., 1977, Radioactive Waste Disposal Low and High Level,
Noyes Data Corporation; Park Ridge, New Jersey.
Led65 Ledbetter, J. 0., 1965, "Environmental Hazard of Radioactive Waste,"
Journal of the Sanitary Engineering Division, ASCE, Vol. 91, No. SA1,
pp. 59-66.
Lem67 Lenneman, W. L., 1967, "United States Atomic Energy Commission Interim
Radioactive Waste Burial Program," Proc. of a_ Symposium on the Disposal
of_ Radioactive Waste into the Ground, Int. Atomic Energy~Agency,
Vienna, pp. 261-300.
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Ma65 Mawson, C. A., 1965, Management of Radioactive Wastes, D. Van Nostrand
Company, Inc., New York, pp. 110-124.
Mc70 McLain, S., and Hungerford, L. B., 1970, "Low-Level Radioactive Wastes,"
Talk Presented to the Class on Engineering and the Environment, The
Department of Nuclear Engineering, Purdue University.
Me76a Meyer, G. L., 1976, "Preliminary Data on the Occurrence of Transuranium
Nuclides in the Environment at the Radioactive Waste Burial Site
Maxey Flats, Kentucky," U. S. EPA Report No. EPA-520/3-75-021.
Me76b Meyer, G. L. , 1976, "Recent Experience with Land Burial of Solid Low-Level
Radioactive Wastes," Presented at the Int. Atomic Energy Agency
Symposium on Mgt. of Radioactive Waste from the Nuclear Fuel Cycle,
Vienna, Austria.
M068 Morton, R. J. , 1968, "Land Burial of Solid Radioactive Wastes: Study
of Commercial Operations and Facilities," U. S. AEC Report No.
WASH-1143.
NAS72 National Academy of Sciences, 1972, "The Effects on Populations of
Exposure of Low Levels of Ionizing Radiation, Advisory Committee on
the Biological Effects of Ionizing Radiation," (Beir Report), Washington,
D. C.
NRC76 National Research Council, 1976, "The Shallow Land Burial of Low-Level
Radioactively Contaminated Solid Waste," National Academy of Sciences,
Committee on Natural Resources, Committee on Radioactive Waste Mgt.,
Panel on Land Burial.
NY76 New York State Department of Environmental Conservation, 1976, "Draft,
Solid Waste Management Facility Content Guidelines for Plans and
Specifications."
St75 Stone, R. , 1975, "Evaluation of Solid Waste Baling and Balefills,"
National Technical Information Service Report No. PB-247-185.
USC76 U. S. Congress, 1976, "Hearings Held on Low Level Radioactive Waste
Disposal."
USE77 U. S. Environmental Protection Agency, 1977, "Rationale for Establish-
ing Risk Acceptability Levels for Radioactive Waste Criteria," Office
of Radiation Programs, U. S. EPA, Washington, D. C.
USE78a U S. Environmental Protection Agency, 1978, "What Control Measures
Should Be Undertaken for Radioactive Waste?" EPA Public Forum on
Environmental Protection Criteria for Radioactive Waste, Denver,
Colorado.
USE78b U. S. Environmental Protection Agency, 1978, "Considerations of Environ-
mental Protection Criteria for Radioactive Waste-Background Report,"
Waste Environmental Standards, Program Office of Radiation Programs,
U. S. EPA, Washington, D. C.
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WASTE MANAGEMENT OF URANIUM MINING
AND MILLING OPERATIONS
(Nancy P. Kirner, State of Washington, Olympia, WA 98504; A. Alan
Moghissi, U.S. Environmental Protection Agency, Washington D.C.
20460); Pamela A. Blackburn, Western Nuclear, Inc., Wellpinit, WA
99040)
ABSTRACT
Uranium mining and milling operations associated with waste practices
are somewhat different from other nuclear operations in that no new radio-
active material is generated.
The methods and procedures of uranium mining and milling operations
using the acid-leach solvent extraction method are described using the
Sherwood Project of Western Nuclear, Inc., as an example.
DISCUSSION
Uranium mining and milling operations remove, pulverize, and disperse
radioactivity which is naturally deposited in the Earth's crust. No new
radioactivity is created through these operations. The uranium and its
daughters, including radon a gasseous radionuclide, are merely converted
to a more dispersible form. It is the large volume of this low concen-
tration waste coupled with its readily dispersible chemical and/or physical
form which makes the safe disposal of uranium tailings a unique challenge.
Most currently extracted ores contain uranium concentrations of
from 0.05% to 1% (IA76), with the more recently constructed U.S. mills
utilizing the lower concentrations. This presentation will attempt to
describe the various sources of waste from Uranium mining and milling.
Western Nuclear, Inc., Sherwood Project will be an example of how a
typical open pit mine and acid leach mill manages its waste.
SHERWOOD - GENERAL DESCRIPTION
Each site has particular characteristics which affect how it manages
its waste. The Sherwood Project is located on the Spokane Indian Reservation,
approximately 33 miles north, northwest from Spokane, Washington. Its loca-
tion high atop a bluff overlooking Lake Roosevelt affords it one of the most
picturesque settings for a uranium milling operation in the U.S.. The top-
ography is generally hilly. Annual rainfall at the site is estimated to be
20 inches (50 cm) per year, slightly more than Spokane, The average temperature
is 44.7°F (9.3"c). The immediate vicinity of the mill is sparcely populated
with approximately one person (0.9) per square mile (0.5 person/km ). The
predominant industries are ranching and multi-use forest with farming and
recreation minimally represented. The only groundwater on the site consists
of low volume springs used previously for livestock waterings (US76). Lake
Roosevelt is the major source of water in the near vicinity of the mill.
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MINE WASTE SOURCES AND CONTROLS
At Sherwood's open pit mine, waste rock is the largest volume mine
waste, and is anticipated to be 88% of the volume of the mined rock. (F178)
The ore is routinely sorted using its radiation emission properties.
A portable GM probe is used while the ore is still in the ground. A second
GM probe tower is employed to measure the average ore content of each
truck load of excavated material. From these two determinations, the
material is divided into three piles: waste rock (or overburden), protore
(or ore that is not yet economic to mill), and ore destined for the milling
operation. Part of the overburden has already been used as a source of
crushed basalt for construction materials for the mill access road and the
horizontal drain blanket at the base of the tailings dam. Most of the
overburden, however, is transferred from one area to another where it is
known there is no underlying ore body.
Low grade ore or protore is another source of mine waste and the
definition of this ore varies from one mill to another. For the Sherwood
Project, protore contains less than 0.06% and more than 0.035% uranium
(Fi76). This protore is piled on 2 pads to await changes in the economics
of the uranium industry. Ore, that with 0.06% and higher uranium content,
is stockpiled on one pad near the crusher building to await processing in
the mill.
Erosion during ore storage is controlled by two means (US76). The
ore pads themselves are sloped towards the mill so as to impound any
surface runoff. There will also be some terracing on the slopes towards
Lake Roosevelt so as to alleviate erosion. Currently all ore dumps are
sloped into the mine site to allow any runoff to be reabsorbed on-site.
Because of the low concentration of uranium in this ore body, the
Sherwood Project is not expected to generate a large volume of mining
equipment which cannot be decontaminated to acceptable levels for release
to uncontrolled areas. Any unrecoverable contaminated material will be
put into the tailings retention system. No such waste has been generated
to date.
Most of the liquid mine effluent is due to surface runoff of the
seasonal rains. Very little water is used for drilling. The Sherwood
Project has encountered little groundwater in the mine area, and that water
which has been found is very localized ground water (Fi76).
Wind erosion tends to be naturally controlled. The stockpiled
material breaks down easily; however, its high moisture content tends to
control dusting naturally. The Sherwood Project has a system for water
spraying the mine areas to control the dust but, as yet, it has not been
necessary to use the system for control of wind erosion.
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176
The stockpiled ore next enters the crushing circuit. Once crushed,
the fine ore storage building protects the majority of the crushed ore
from erosion by wind and rain. Some minimal erosion can occur as the
building has been erected on pedestals for ease of maintenance. The
expected lifetime of the ore body is 10.6 years (US76). After that time,
erosion of the mine area is planned to be controlled by limited restora-
tion of the topography with overburden, sandy surface soil, and seeding
with native vegetation.
Potential sources of airborne contaminants from the open pit mine
are primarily the operation of excavation equipment and trucking on
unsurfaced roads. In comparison, blasting is a small source of dust (FI78),
Non-sanitary waste water from the mill has been used to control dusting
in the mine area.
A limited airborne monitoring program had been operating for two
years prior to the beginning of mining operations. A more elaborate
monitoring program begun in April, 1977 has seen no appreciable increase
in activity since mining operations commenced (Me78).
MILL WASTE SOURCES
A short discussion of the mill process will aid understanding of
waste sources. The Sherwood mill employs an acid leach process using a
combination of Sulfuric Acid and Sodium chlorate at a pH of about one to
dissolve the uranium from the crude ore (US76). This slurry is then
treated with polymeric fluculent to separate the uranium bearing solution
from the barren waste by the process of counter current decantation. The
dissolved uranium then enters the solvent extraction process which uses
high flash point kerosene as the carrier for the active extractant - a
tertiary amine. Following extraction into the solvent, followed by an
ammonium sulfate strip of the solvent, the concentrated uranium in an
aqueous phase is neutralized by anhydrous ammonia, yielding the yellow-
cake product [ammonium diuranate (NH,)2 U?OJ (Mi78). The remaining
water content of the yellowcake is removed through a thickener process,
centrifuge, washing the sulfates out, and finally evaporative heating in
the roaster. The dried yellowcake is then transferred to a hopper to
await packaging in 55 gallon drums.
This typical milling process, therefore, generates the following
wastes:
unreacted, acidic, barren ore
flocculated and barren precipitates
waste organic solution
neutralized, barren ammonium sulfate solution
aqueous wastes from the thickeners and centrifuge
contaminated runoff
dusts from the crushing circuit
sulfuric acid vapors
dusts from dryers and packaging areas
radon emmissions from radium bearing wastes
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177
Non-routine wastes also result from:
emergency spills or overflows
decommissioning of plant
long term emissions and erosions from tailings area
Chemical constituents other than Uranium of the ore body which need to
be considered include:
Molybdenum
Vanadium
Zinc
Calcium
As with most acid leach mills (IA76), salts are a major threat
to the ground water. Fortunately, there is minimal ground water at the
Sherwood Project and the sorptive capacities of the soil are natural
barriers to control migration of the salts (US76).
WASTE HANDLING SYSTEM
Tailings area. The most prominent feature of the Sherwood Project
and the feature which will handle the major share of all wastes generated
is the tailings retention system. The mill tailings are deposited in an
above ground dammed structure. This tailings area is located in a valley
which drains eventually to Lake Roosevelt (the backwaters of the dammed
Spokane River). The 163 acre (63 hectares) dammed tailings area (WN76)
is surrounded by a diversion ditch designed to divert the projected 100
year flood which was postulated to follow the day when the projected 50
year flood occurred. This tailings retention area is created by a 63 foot
(19 m) high, 2400 foot (740 m) long dike which is designed to contain
approximately two years of tailings (US76). The dike will be increased
in stages to an eventual height of 108 feet (32 m) to contain the tailings
from the projected 10.6 years of mill operations.
The entire tailings area is, or will be, lined with a 30 mil (1.7 mm)
polyester reinforced hypalon liner. Hypalon is a synthetic rubber which is
reasonably resistant to weathering, ozone, and sunlight. It has high resista-
bility to acids and alkalis. It is also moderately resistant to various
organic solvents and biological degredation. It is usually supplied in the
unvulcanized form and can be seamed by heat sealing or solvent welding (St78).
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178
In 1974-1975 when this waste retention system was first proposed,
storage of the tailings in the mined out portion of the mine area was
considered, but was rejected because of the increased hazard to any
ground water sources in the area and to Lake Roosevelt. The use of the
above ground lined tailings system was preferred at that time because
it essentially removed the aqueous wastes from the environment (US76).
A small portion of the hypalon liner was strength tested in place using
heavy construction equipment. Aside from minor punctures from pointed
rocks, no tears or major flaws were noted.
The synthetic liner on the waste ponds is designed to contain all
liquid effluents from the mill, thus facilitating the re-use of water and
storage of rain water. The liner should also prevent migration of this
contaminated water through the underlying silty sands. Thus the large
amount of salts normally generated through the acid leach process should
remain adequately sequestered in the lined tailings structuref
The tailings retention system receives the neutralized tailings slurry.
The previously acid tailings are neutralized with calcium oxide. The addi-
tion of the calcium oxide is beneficial because the chemical similarities
between calcium and radium reduces mobility of the radium in the tailings.
Also, the majority of the heavy metals are also settled out during this
process. Since water is at a premium at the Sherwood mill, a pontoon
mounted pump is floated in the tailings pond to pump the neutralized water
back into the mill process following treatment with Barium chloride for
radium removal in the aggitated barium chloride treatment tank.
This removal is accomplished in the primary and secondary precipitation
ponds. These ponds comprise a total area of 3/4 acre (3000 m ) (US76)
and are lined with the same reinforced hypalon liner material used for the
main tailings pond (WN76). The supernatent which is drawn from the ponds is
routed back to the mill process. Any excess water which cannot be routed
back to the mill is allowed to overflow into the unlined seepage/evaporation
pond.
2
An industrial waste pond of approximately 1/8 acre (500 m ) has also
been constructed and lined with the reinforced hypalon (WN76). This pond
has, so far, been used only to contain and evaporate the backwash water
generated during cleaning of the filter media from the drinking water
treatment plant. It has the capability, however, of containing wastes
from any of the processes in the plant, but particularly those from the
solvent extraction process, truck maintenance bays, and laboratory areas.
OTHER WASTES
One major feature of the Sherwood mill is that it has reclaimed a
major fraction of its waste streams. Since the mill is processing low
grade ore, average 0.089% uranium oxide (Fi78), and since the site has
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179
no substantive on-site water supply and since the costs of pumping water
from Lake Roosevelt are high, it is economically critical to recycle.
Thus waste streams from the acid leach process, the barren ammonium sulfate
(counter-current decantation) circuit, and any waste organic solution are
fed back into the mill process. Aside from the economic considerations,
recycling has the obvious benefit of minimizing water pollution.
The mill has been designed so that no liquid can escape the process
area during normal operation as a result of any foreseeable tank or pipe
failure (WN76). The floor grates, building walls, sumps and berms were
designed so that there will be adequate volume to contain spills within
the buildings or process areas. For example, the mill building will contain
the contents of two leaching tanks which burst simultaneously.
Contaminated runoff is unlikely to occur from the mill area since
the tailings area is sloped to contain any such runoff which is generated
within the tailings area. In addition, the walls of the diversion ditches,
much of which is comprised of granite, are expected to divert any surface
runoff from intruding into the tailings area (US76).
Airborne wastes are controlled by the use of a wet scrubber atop the
crusher and bag houses above the crushed ore storage buildings and conveyor
transfer points. Dust control may be enhanced in dry weather by the use of
water sprays. The operation utilizes a wet grinding system, thus eliminat-
ing the need for dust collection equipment (WN76).
Acid vapors in the leaching circuit, if generated, are collected
through a common header and vented through demisters to trap acid and
chlorine gas. Any water collected is then routed back to the process.
A forced air ventilation system is used to collect and vent to the outside
organic vapors generated in the solvent extraction portion of the process.
The precipitation and dewatering of the yellowcake is essentially a wet
process, and thus the airborne effluent control at this stage (WN76) is
relatively simple.
The roaster drying area is equipped with a wet venturi scrubber to
collect greater than 99% of the uranium particles contained in this process
effluent. The packaging area is equipped with a micro pulsair bag house
collector (WN76).
Radon emissions from radium bearing wastes are of most concern in the
milling process prior to the transfer of tailings to the tailings pond.
During tne acid leaching process, radon is controlled by ventillation of
the mill building itself; the Counter Current Decantation process is out-
side, therefore, no emission control system is utilized in this wet process
(WN76). Currently, the fine ore storage bin is the largest radon source
at the mill site (Me78, Mi78).
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180
Radon emanations from the operating tailings pond are generally uncon-
trolled, using only fresh air dilution. Calculations have been made to
estimate the population doses from Rn emissions per year to the popula-
tion of a 50-mile radius. This dose is estimated to be 4.79 man-rem to the
lung (US76).
RECLAMATION OF THE TAILINGS AREA
A tailings area stabilization plan has been proposed which includes
provisions for both stabilization and for long term maintenance. Estimates
of the amount of cover (overburden) needed to reduce radon emissions to
approximately twice the background emanation rate were made and were calcu-
lated to be approximately 13 feet (4 m) (WN76, Cu76, Kr63).
One essential feature of the stabilization plan is the grading of the
tailings and its 13 feet of cover to blend with the surrounding area so as
to minimize erosion. The operators of the Sherwood Project have secured a
surety bond from the Bureau of Indian Affairs in the amount of $6 million
(April, 1978) to cover mill decommissioning and tailings stabilization costs.
A separate bond was also secured from the Bureau of Indian Affairs in the
amount of $176,000 to cover maintenance and monitoring of the reclaimed and
stabilized area for a period of 50 years. Provision was made to re-evaluate
the adequacy of both bonds periodically. This will probably be done at
the time of radioactive materials license renewals.
SUMMARY
Each uranium extraction method, acid leach, basic leach, heap leach,
in-situ leach, has its own characteristic waste problems (IA76). These
problems may be allevaited or enlarged by the characteristics of a partic-
ular location. For instance, in-situ leaching by its very nature has solved
the problem of what to do with large volumes of solid waste which is so
typical of acid leach mills. However, in-situ leaching has perhaps a more
environmentally significant problem with the large volume of liquid waste
which is generated which may threaten the ground water (US78a). In Texas
(US78a), such waste has been allowed to be discharged into more or less
sequestered underground formations. This may or may not be an acceptable
method of disposal elsewhere, depending on the characteristics of the
particular site.
The Sherwood Project endeavored to employ state-of-the-art technology
to solve their typical acid leach waste problems. The unique features of
Sherwood's system are the hypalon liner used in the tailings retention system
and their water conservation techniques. Both of these approaches have
resulted from the site specific characteristics of minimal ground water, the
height of the mill from Lake Roosevelt, and its remote location from natural
tailings system liners, such as bentonite clay.
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181
REFERENCES
Cu76 Culot, M.V., Olson, HJ.G., and Schaiger, K.J., "Effective Diffusion
Coefficient of Radon in Concrete Theory and Method for Field Measurements",
Health Physics, 30,263.
Fi78 Filler, R., Personal Communication, Western Nuclear, Inc.
Kr63 Kraner, H.W., Schroeder, G.L., Evans, R.D., "Measurements of the
Effect of Atmospheric Variables on Radon-222 Flux and Soil-Gas Concen-
trations," Proceedings of the First International Symposium on Natural
Radiation Environment, Adams and Lowder, editors, Rice University, Houston,
Texas, 210.
IA76 IAEA Safety Series 44, "Management of Wastes from the Mining and
Milling of Uranium and Thorium Ores", International Atomic Energy Agency.
Me78 Meenach, G.T., Environmental Monitoring Program Semi-Annual Report,
Sherwood Project, Western Nuclear, Inc.
Mi78 Miyoshi, K., Personal Communication, Western Nuclear, Inc.
St78 Stewart, W.S., "State of the Art Study of Land Impoundment Techniques",
EPA-600/2-78-196.
US76 US Department of the Interior, Bureau of Indian Affairs, Final
Environmental Statement, Sherwood Uranium Project, Spokane Indian
Reservation.
US78a US Nuclear Regulatory Commission, "Groundwater Elements of in situ
Leach Mining of Uranium," prepared under contract by Geraghty and Miller,
Inc., NUREG/CR-0311.
US78b US Nuclear Regulatory Commission, "Final Environmental Statement,
Irigaray Uranium Solution Mining Project", NUREG-0481.
WN76 Western Nuclear, Inc., Application for Radioactive Materials License,
June, 1976, with subsequent revisions.
DISCLAIMER
This paper was prepared while one of the authors (A.A.M.) was at Georgia
Institute of Technology, Atlanta, Georgia. Mention of commercial products
in this paper does not constitute an endorsement by the State of Washington,
Georgia Institute of Technology, U.S. Environmental Protection Agency or
Western Nuclear, Inc.
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182
MANAGEMENT OF LOW-LEVEL NATURAL RADIOACTIVITY WASTES OF
PHOSPHATE MINING AND PROCESSING*
C. E. Roessler, Z. A. Smith and W. E. Bolch, Department of Environmental
Engineering Sciences, and J. A. Wethington, Jr., Department of Nuclear
Engineering Sciences, University of Florida, Gainesville, FL 32611
Abstract - Redistribution of the uranium-series radionuclides associated
with phosphate deposits produces what are, in effect, low-level radio-
active wastes. Proposed management of these materials involved
(1) restricting by-product uses to applications not causing significant
exposures to man; and (2) returning other material to the land with
simulation of the natural radioactivity depth profile by covering high
activity overburden, tailings, clays, gypsum and other radioactivity-
bearing wastes with a lower activity overburden layer. Radon flux
models are being developed to aid design of materials placement.
Introduction
Increasing attention is being directed to the potential radiation exposure
from naturally-occurring radionuclides at concentrations only an order of
magnitude above the average values at the earth's surface. This is leading to
a new dimension in radioactive waste management - large quantities of relatively
low activity materials, involvement of industries other than the traditional
nuclear industries and impingement on established commerce.
One of these natural radioactivity sources is the phosphate industry.
Naturally-occurring uranium series radioactivity associated with phosphate
deposits of marine origin is redistributed in mining, beneficiation, chemical
processing, distribution and use of products and by-products and waste
management (Gu75, Ro78).
One of the more significant exposure routes to man from this radioactivity
source is that of inhalation of airborne radon progeny in structures that (a) are
built in a closely-coupled fashion over lands containing elevated near-surface
concentrations of 226Ra or (b) incorporate radium-bearing building materials
(US75). Consequently, this paper will concentrate on the fate of radium and on
measures that limit radon exhalation.
Historically, there has been little concern and no regulatory constraints
on the potential radioactive wastes from the phosphate industry. Recently,
however, recommendations have appeared calling for licensing, restriction of the
use of certain by-products, and increased monitoring and regulation to control
natural radioactivity from the phosphate and other industries (Boo77, Boh78).
Then in late 1978, proposed hazardous waste regulations of the U.S. Environmental
Protection Agency designated overburden and slimes from phosphate surface
mining, gypsum from wet-process phosphoric acid production and slag from
elemental phosphorus production as hazardous solid wastes (US78).
^Supported in part by the Florida Phosphate Council and by the University of
Florida Engineering and Industrial Experiment Station.
scope of this paper will be limited to the Florida phosphate industry.
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183
The Radioactivity Source
In the typical undisturbed, pre-mining profile, 226Ra concentrations are
on the order of 0.5 pCi/g at the surface, increase gradually with depth through
the 1 to 20 m overburden and reach values on the order of 40 pCi/g near and in
the ore or "matrix" (Ro78).* In mixed overburden that has been cast aside to
expose the matrix, concentrations are typically an order of magnitude higher
than in the original surface soil but local concentrations may be found ranging
up to that of the matrix.
For the purpose of this paper, phosphate operations can be conveniently
divided into three categories - mining, wet-process chemical operations and
the thermal process. Mining includes overburden removal, excavation and
transport of the exposed matrix and beneficiation of the matrix to separate the
phosphate rock product from sand and clay. In wet-process chemical operations,
phosphate rock is processed to produce phosphoric acid and phosphate products
such as fertilizer and animal feed ingredients. In the thermal process,
phosphate rock is reduced in an electric furnace to produce elemental phosphorus.
The partitioning of radioactivity in phosphate mining and chemical operation has
been described previously (Gu75, Ro77, Ro78). The flow of materials and
radioactivity in mining and beneficiation, phosphoric acid production and
electric furnace operations are summarized in Figures 1-3. The 226Ra concen-
trations of various fractions are presented in Table 1, column 2. Included in
the table but not identified in a figure is "debris" which consists of higher
radioactivity spoils on lands mined prior to the adoption of the flotation
process (i.e. prior to the 1940's). At that time, only the coarse, pebble
product fraction was recovered and all the balance of the input matrix was
returned to the land. Also included in the table is an entry for the sediments
and scales (gypsum and other impurities) that occur in phosphoric acid reactors,
filters, piping and tanks; these contain the highest 226Ra concentrations
observed in the industry.
Rough estimates were made of the production of these wastes annually by a
nominal facility, annually by the entire Central Florida industry and cumula-
tively to date. Table 2 presents volumes, quantities and radium activities
associated with each of the major waste types and the areas committed to
overburden spoils, clay settling ponds, tailings dumps and gypsum piles. While
the radionuclide quantities are significant, most occur as tremendous volumes
of low concentration materials.
As summarized in Table 3, most of these "waste" materials have an onsite
use, have entered into commerce or have potential commercial use. While uses
do represent potential routes for radiation exposure to the public, they also
factor into the costs associated with management alternatives, and certain uses
may even present a positive benefit to society. All of these facets must be
considered and balanced in developing a reasonable posture concerning large
volume materials of low activity concentrations.
*Radioactivity concentrations given in this paper are current values for
Central Florida. In the North Florida phosphate mining region, surface soil
concentrations are comparable but those in matrix and various products are 25
to 50% of those in Central Florida.
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184
General Considerations for Waste Management
In view of the large quantities of relatively low activity materials and
considerations such as the need for overburden and tailings as replacement for
mined volumes, the potentially recoverable phosphate and uranium resources of
dewatered clays, and the superiority of furnace slag to other materials
locally available as aggregates, it is neither practical nor prudent to ship
these materials to off-site burial grounds as is done with small quantities of
"conventional" low level radioactive waste from the nuclear industry and
nuclear research. Rather the solution lies in first developing criteria or
standards for population exposure from technologically enhanced natural
radiation, next designating limits for unrestricted use of materials and land
and then pursuing one or both of two options:
1) restricting land use and commercial utilization of materials according
to radioactivity parameters, and/or
2) employing in situ management of materials in mining, reclamation and
disposal where necessary to meet criteria for the desired land use.
Criteria for Materials and Land Use
Establishment of an indoor airborne radon progeny standard is a prerequisite
for regulating materials for radioactivity in order to limit exposure by this
route. Values ranging from 0.005 WL to 0.05 WL above background have been
proposed; hopefully a standard will soon be established (St71, US77, HRS78).
Most recently, the hazardous waste regulations propose a maximum addition to
background of 0.03 WL (US78).
With an airborne radon progeny standard, models and empirical observations
can be used to derive standards in terms of measureable lands and materials
parameters. The indoor airborne radon progeny concentration is a function of a
number of variables including the radon exhalation rate or flux at the inputing
surface. Flux, in turn, is a function of the radon generating capacity, thick-
ness and radon transport characteristics of the underlying radium-bearing
material and of transmission characteristics of any cover (such as a covering
over a wall or a floor over fill and soil). Thus for both building materials
and under-floor soil and fill, the radon-generating capacity of the materials
is a fundamental parameter. This radon generating capacity can be expressed
as the "emanating radium concentration", Cf, , which can be defined:
C® = C_ -E,
Ra Ra '
where C is the materials radium concentration which is also the
radon production rate per unit mass, and
E is the "emanating power" or "emanating fraction", i.e., the
fraction of the produced radon that is released to void space.
Radioactivity limits for building materials, land and fill, adopted for
the purpose of limiting exposure by the airborne radon route, should be based
on Cj|a or jointly on its two components, CRa and E. Values of E and Cj|a for
various phosphate waste and by-product materials are listed in Table 1,
columns 3 and 4.
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185
The relationship between average indoor radon progeny concentrations in
existing structures and measureable land radiological parameters has been
quite variable in field studies in Central Florida. In University of Florida
studies, the soil radioactivity concentration predicting an indoor radon progeny
concentration of 0.03 WL in conventionally-constructed slab-on-grade structures
was on the order of 6 to 19 pCi/g for various data-fitting models. Based on
studies in Central Florida, the proposed EPA hazardous waste regulations
specify a bRa concentration limit of 5 pCi/g for land and building materials
If the nominal emanating fraction of typical Florida soils is taken to be
about 0.20, the 5 pCi/g standard corresponds to a C| value of about 1 pCi/g.
If this concept is extended to other materials, absofute radium concentrations
corresponding to this emanating radium value can be calculated as indicated in
Table 1, column 5.
Some of the Table 1 materials meet the 1 pCi/g emanating radium criterion
and can be used as construction materials or as lands or fill for residential
construction without incorporating any special measures for limiting radon
progeny exposure. On the other hand, the radiological characteristics of land
reclamation materials are highly variable and individual cases would have to
be evaluated. Under the proposed criteria, use as building material would be
contraindicated for some of the materials such as by-product gypsum and some
parcels of land would require special restrictions on the type of construction.
The status of slag will be uncertain until the wide disparity in reported
emanating fraction values can be resolved.
Materials Management
Since some of the materials do not meet the proposed criteria for
unrestricted use, some form of management in place would be indicated. This
consists of either mixing higher and lower radioactivity materials to reduce
the concentration or selective placement in mining, disposal, reclamation and
construction site preparation to more nearly simulate the natural radioactivity
concentration profile with the higher activity materials under a low radio-
activity cover.
Mathematical modeling of radon flux can be used as an aid in planning
materials management schemes to be tested in the field. The expected flux
at the surface of an emanating layer can be calculated from the finite single
layer model, equation A-l in Table 4. For most of the materials of concern
here, the thickness approaches an essentially infinite value after several
meters and the maximum flux can be calculated from the simpler infinite single
layer model, equation A-2 in the table. If the values of all the other
parameters can be anticipated, this equation reduces to:
J = K C_ ,
o> Ra
where K is a constant for the specified material and conditions.
Values of K for some conditions expected in Florida have been computed as
shown in Table 1, column 6.
For the indoor radon progeny inhalation exposure pathway, limiting
conditions are based on the projected indoor radon progeny levels in future
-------
186
structures built over the site of concern. In the field studies, the soil
surface radon flux corresponding to 0.034 WL (0.03 WL above an average background
of 0.004 WL) for existing structures of conventional slab-on-grade construction
ranged from 4 to 19 pCi/m2-s, depending upon the model used to fit the data.
A conservative approach would limit flux to 4 or 5 pCi/m2-s.
Covering a lower layer of more active, radon source materials with an
upper, attenuating layer of available lower activity material can be represented
by the bi-layer model, equation B in Table 4. For the bi-layer situation, any
required reduction from the flux predicted for the uncovered case can be expressed
as an "apparent transmission" (acutally the net effect of the attenuation of
radon originating in the lower layer and the contribution of radon from the
cover material itself):
T = J-/J , or, in the maximum case, T = J^/J^.
The number of combinations expressing the range of expected conditions can be
reduced by presenting the transmission as a function of the ratio of the soil
radium concentrations in the two layers and constructing families of attenuation
curves for pairs of materials. Figure 4 is a set of curves for high activity
overburden covered by low activity upper layer overburden.
As an example, consider a thick layer of 11.7 pCi/g high activity over-
burden at 10% moisture. The infinite single layer model predicts a flux of
7.4 pCi/m2-s. Using a design objective of 4 pCi/m2>s this requires a layering
with T = 0.54. Using 2.3 pCi/g overburden as cover, the activity ratio is
0.2; from the curve the required cover thickness is 0.7 m. This is an attainable
condition. Similar types of analysis can be performed for placement of materials
produced in mining or for covering chemical plant waste and by-products.
The models suggest that placing the higher activity materials under
reasonable thicknesses of available low activity cover (upper layer overburden)
will achieve radon fluxes acceptable for unrestricted use. This can be
achieved directly in mining if it is feasible to separate the overburden and
place the higher activity lower layer (with its leached zone and adventitious
matrix) and other waste materials in the bottom of mine cuts with a covering
of upper layer overburden. Incidently, such selective placement also reduces
possible radionuclide uptake by plants and minimizes the potential for
contamination of near-surface ground water and run-off.
The modeling calculations are being extended to various concentrations of
other materials and simulations of materials placement in test columns are
currently in progress. The ultimate test of these methods will be the extent
to which they are verified in future field operations.
Summary and Conclusions
The naturally occurring uranium series radioactivity appearing in phosphate
mining spoils and the waste and by-products of ..eneficiation and chemical plant
operations is for the most part contained in large volumes of relatively low
radioactivity concentration. The solution to management of "materials and lands
with radioactivity levels that do not meet unrestricted use criteria consists
of 1) restricting land and materials use and/or 2) selective in situ placement
of higher radioactivity materials in a manner that meets criteria for the
desired use.
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187
Inhalation exposure to airborne radon progeny originiating in building
materials and in lands used for construction purposes constitutes one of the
more significant exposure pathways and this paper concentrated on limitation
of exposure by this route. The definitive approach to this problem awaits an
adopted indoor airborne radon progeny standard. The development of criteria
for materials and lands is further complicated by the observed high variability
in the relationship between airborne radon progeny concentrations and
radiological characteristics of lands.
The emanating radium concentration is a more meaningful single parameter
for evaluating materials than the absolute radium concentration. A method for
estimating cover thickness requirements was presented.
Concepts have been formulated and models are showing directions to proceed.
However, empirical observations show large variability in the various parameters
needed to predict radiation exposure and design mitigative measures. Much
work still needs to be done to measure basic parameters of the materials in
question, fine tune mathematical models and perform testing and validations.
A complete approach involves evaluation of other nuclides of the uranium
series and other exposure pathways including suspension of airborne particulates,
uptake by crops from the root zone of the soil with subsequent tranmission through
the agricultural food chain and transfer to surface and ground water.
Acknowledgements
The work of Mr. Bruce Butler in performing emanating fraction measurements
is greatly appreciated.
References
Bo78 Bolch W. E. , Desai N. and Roessler C. E., 1978, "Modeling Radon Flux",
Natural Radioactivity Studies - Radioactivity of Lands and Associated
Structures, Final Report Volume Two, Technical Reports, University of
Florida, College of Engineering (Gainesville, FL) 98-134.
Boh78 Bohlinger, L. H., 1978, "Natural Radioactivity Contamination Problems"
in 9th Annual National Conference on Radiation Control, June 19-23, 1977,
HEW Publication (FDA) 78-8054, 168~-171.
Boo77 Booth, G. F. , 1977, "The Need for Radiation Controls in Phosphate and
Related Industries", Health Physics JJ2, 285.
Fi78 Fitzgerald J. E. Jr. and Sensintaffar E. L., 1978, "Radiation Exposure
From Construction Materials Utilizing By-product Gypsum from Phosphate
Mining", Radioactivity in Consumer Products (edited by Moghissi A. A.,
Paras P., Carter M. W. and Barker R. F.) NUREG/CP-0001, 351-368.
Gu75 Guimond R. J. and Windham S. T., 1975, "Radioactivity Distribution
in Phosphate Products, By-products, Effluents, and Wastes", U. S.
Environmental Protection Agency Technical Note ORP/CSD-75-3.
HRS78 Florida Department of Health and Rehabilitative Services, Radiological
Health Services, 1978, "Study of Radon Daughter Concentrations in
Structures in Polk and Hillsborough Counties", IV-9.
Li78 Lindeken C. L. and Coles D. G., 197S, "The Radium-226 Content of
Agricultural Gypsums" in Radioactivity in Consumer Products (edited by
Moghissi A. A., Paras P., Carter M. W. and Barker R. F.) NUREG/CP-0001,
369-375.
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188
Ro77 Roessler C. E., Smith Z. A., Bolch W. E. and Prince R. J., 1977,
"Uranium and Radium-226 in Florida Phosphate Materials", University of
Florida (Gainesville, FL).
Ro78 Roessler C. E., Kautz R., Bolch W. E. and Wethington J. A. Jr., 1978,
"The Effect of Mining and Land Reclamation on the Radiological Character-
istics of the Terrestrial Environment of Florida's Phosphate Regions",
in Proceedings of the Symposium The Natural Radiation Environment III,
April 23-38, 1978, in press.
St71 Steinfield J. L., 1971, "Recommendations of Actions for Radiation
Exposure Levels in Dwellings Constructed on or with Uranium Mill Tailings",
recommendations from the Surgeon General of the U. S. Public Health
Service to the State of Colorado, reprinted in "Preliminary Findings -
Radon Daughter Levels in Structures Constructed on Reclaimed Florida
Phosphate Land", U. S. Environmental Protection Agency, Technical Note
ORP/CSD-75-4.
US75 U. S. Environmental Protection Agency. Office of Radiation Programs,
1975, "Preliminary Findings - Radon Daughter Levels in Structures
Constructed on Reclaimed Florida Phosphate Land", Technical Note
ORP/CSD-75-4.
US77 U. S. Environmental Protection Agency, Office of Radiation Programs,
1977, "Draft Radiation Protection for Florida Phosphate Lands - Proposed
Recommendat ions".
US78 U. S. Environmental Protection Agency, 1978, "Hazardous Waste Proposed
Guidelines and Regulations and Proposal on Identification and Listing",
Federal Register. 43, (243) 58946-59028.
-------
TABLE 1. Some Radon-generating Properties of Florida Phosphate-related Materials
Numerical Values are for Central Florida
Material
226Ra
(CRa> .
~/-i,' /„ a)
Emanating
Fraction , N
fv\ b>
Emanating C for
Radium (cf )
Ka r>e - 1 «/-,• /„
J./CRa
Overburden
Range through profile
Leached zone
Mixed overburden lands
Clays
Sand Tailings
Debris Lands
Gypsum
Misc. Sediment, Scale
Slag
pCi/g
A. MINING AND BENEFICIATION
0.2-46 — .
9 (2-46) 0.16 '
5 (1-35) 0.16
28 (14-52) 0.20
5 (2-12) 0.10
10 (3-32) 0.12
B. WET PROCESS PHOSPHORIC
32 (21-65)
(80-380)
0.12
1.4 (0.2-7.4)
0.8 (0.2-5.6)
5.6 (2.8-10)
0.5 (0.2-1.2)
1.2 (0.4-2.8)
ACID PRODUCTION
3.8 (2.5-7.8)
6.3
6.3
5.0
10.0
8.3
8.3
0.63
0.63
0.24
0.45
0.73
C. THERMAL PROCESS
69 (45-101)
0.05
(0.0002-0.004)
e)
3.5 (2.2-5.1)
(0.01-0.4)
20.0
(200-400)
a) For comparison, some ZZbRa concentrations in products are:
Mining and beneficiation: matrix (ore), 38 (18-84); pebble (product), 57 (45-97); rock concentrate
(product), 35(24-50) ;
Wet process phosphoric acid and fertilizer production: phosphoric acid, <1 pCi/g; ammoniated phosphates,
4 (1-12); triple superphosphates, 20 (15-32).
b) Standard deviations of the distributions of mean emanation power for multiple samples of the various
materials were: overburden and debris 50-60%; clays and tailings 13-15%; gypsum 30%; and slag 60%,
expressed as percent of the category mean.
c) Clays calculated at 33% moisture, all others calculated at 10% moisture.
d) Assumed to be similar to mixed overburden.
e) Analysis by second laboratory; discrepancy unexplained.
— indicates not measured or not calculated.
oo
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190
TABLE 2. Estimated Inventory of Florida Phosphate Solid Wastes and By-products*
Material
(Form, 226Ra)
Parameter
Facility Florida
Annual Annual
Florida
Cumulative
A. MINING AND BENEFICIATION
Overburden:
(spoils, 5 pCi/g)
Clays ;+
(settling ponds,
28 pCi/g)
Sand Tailings:
(spoils, 5 pCi/g)
B.
Gypsum;
(gypsum piles,
5pCi/g)
Sediments & Scales:
(80-380 pCi/g)
Volume, m3
Mass, tons (metric)
Area, hectares
226Ra activity. Ci
Volume, m3
Mass, tons (metric)
Area, hectares
226Ra activity, Ci
Volume, m3
Mass, tons (metric)
Area, hectares
226Ra activity, Ci
WET PROCESS PHOSPHORIC
Volume, m3
Mass, tons (metric)
Area, hectares
226Ra activity, Ci
9.9 x 106 1
1.4 x 107 1
80
70
2.0 x 106 2
2.8 x 106 3
50
80
2.3 x 106 3
4.2 x 106 5
30
20
ACID PRODUCTION
.3 x 108
.9 x 108
1000
940
.5 x 107
.6 x 107
660
1000
.0 x 107
.5 x 107
350
270
2.3 x 106 2.1 x 107
2.5 x 106 2.3 x 107
10 100
80 730
No Quantity Estimate
2.5 x 109
3.6 x 109
26,000
18,000
2.8 x 108
4.0 x 108
17,000
11,000
1.1 x 109
6.0 x 108
9000
3000
1.3 x 108
1.4 x 108
730
4400
C. THERMAL PROCESS
Slag:
(slag pit, 60 pCi/g)
Volume , m3
Mass, tons (metric)
226Ra activity, Ci
1.1 x 105 2
1.4 x 105 2
10
.1 x 105
.7 x 105
17
6.7 x 106
8.6 x 106
550
*Quantities and areas were back-calculated from published mining and chemical
plant production data and represent only rough estimates.
Clays volume and mass reported on dry weight basis. Settled clays actually
contains 15% solids.
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191
TABLE 3. Current and Potential Values and Uses of Phosphate Industry Wastes
A. MINING AND BENEFICIATION
Overburden - Replacement for volume removed in mining; land reclaimed for
development, agriculture, recreation.
Clays - Replacement volume (30% of removed matrix); potential phosphate and
uranium resource if recovery technology developed; potential soil conditioner
to improve fertility, exchange capacity, moisture retention; potential
commercial uses.
Sand Tailings - Replacement volume (40% of removed matrix); fill materials.
B. WET PROCESS PHOSPHORIC ACID PRODUCTION
Gypsum - Soil conditioner and Ca source (Li78); building materials use outside
U.S. (Fi78); potential chemical raw material.
Sediments and Scales - No known use or values.
C. THERMAL PROCESS
Slag - Crushed and to commercial use as aggregate; or as light weight
aggregate (expanded form).
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192
Table 4. General Models for Soil Surface Radon Flux*
Definitions
Description
Radon flux at the soil-air
interface
Radium concentration in the soil
media (grams, dry weight)
Bulk density, volume per unit
of dry weight
Emanating power, amount of radon
Symbol
J
CRa
P
E
Units
pCi/m2-s
pCi/g
g/cm3
pCi/pCi
released to void space per
unit of radium (or radon)
Decay constant for radon-222
Porosity
Media diffusion coefficient
D = 5.0 x 10"2 e"0-1 for values
of D from 5 x 10"2 to 5 x 10"3
P
D
cm2/s
Depth of media
cm
A. Single Layer Models
1. Finite Single Layer
Jd = 10 • CRapE J— • tanh
2. Infinite Single Layer - special case, very thick active layer (d •* °°)
Jc= =
CRa P E 'T
B. Bilayer Model
J2 = -104m2D2(2A2eW
where
- 2)e
2w-y
ml =
kl =
y =
k2 = CRa2P2E2/P2
m2D2
4 m D, tanh z
. ,
i
* From Bolch et al.(Bo78).
-------
193
MINING
SITE
Overburden
To overburden
spoil piles
Matrix
Weight: 100%
226 Ra: 38pCi/g
| Benef iciation
I Plant
Product
WASHER
Waste
Material
\ 22
Clays
Weight: 30%
226 Ra: 26pCi/g
(Flotation \
^ J
Pebble
Weight: 10%
226Ra: 57pCi/g
FLOTATION
Product
Waste |
Material \
V 2
Sand Tailings
Weight: 40%
226 Ra: 5pCi/g
Rock Concentrate
Weight: 20%
226 Ra: 37pCi/g
ROCK STORAGE
AND LOADING
To clay
settling
To sand
disposal
To dryer,
chemical plant,
or customer
Fig. 1. Simplified Flow Diagram for Phosphate Mining and Benef iciation.
Radium Concentrations are for Central Florida.
-------
194
Phosphate Rock
226 Ra: 37 pCi/g
238 U: 32pCi/g
Sulfuric Acid
[A. Phosphoric \
Acid \
Plant \
ACIDULATION
'
'
FILTRATION
Product
f Phosphoric Acid \
I 226 Ra: <1pCi/g
V 23B U: 30pCi/g >
By- („* Gypsum A
product V 238 U; <, pCJ/g 7
j
|
t
To gypsum
pile
To further processing,
fertilizer production, and/or
market
Fig. 2. Simplified Flow Chart for Wet Process Phosphoric Acid Production. Radionuclide
Concentrations are for Central Florida.
f Phosphate Rock
( 226 Ra: 37pCi/g J
V 238 U: 32pCi/g )
Coke
226 Ra: 1 pCi/g
238 U: 2pCi/g
ELECTRIC
FURNACE
By-product
By-product
f Slag
( 226 Ra: 64pCi/g
V 238 U: 63pCi/g
Ferrophosphorus \
226 Ra: 2pCi/g )
"" U: 41pCi/g 7
238
Fig. 3. Simplified Flow Diagram for Thermal Process Elemental Phosphorus
Production. Radionuclide Concentrations are for Central Florida.
-------
195
CB.1 -11.7PCI/Q
J = 7.4 pCi/m2
2 3(1) 4 5
UPPER LAYER DEPTH, M(m)
Fig. 4. Typical Design Curve for Attenuation of Radon Flux.
Low Activity Overburden over High Activity Overburden.
-------
196
ACTIVITY MEASUREMENTS AT A WASTE VOLUME REDUCTION FACILITY
J. Richardson, D.A. Lee; Bruce Nuclear Power Development,
Ontario Hydro, Tiverton, Ontario, NOG 2GO
Abstract
The monitoring program for Ontario Hydro's radioactive waste
management site will be described, several aspects of which will
be discussed in detail. The program at this facility includes
categorization, volume reduction processing, and storage of solid
radioactive wastes from nuclear generating stations of the CANDU
type. At the present time, two types of volume reduction process
are in operation - incineration and compaction. Following catego-
rization and processing, wastes are stored in in-ground concrete
trenches or tile-holes, or in above-ground quadricells.
The monitoring program is divided into three areas: public
safety, worker safety, and structural integrity. Development
projects with respect to the monitoring program have been under-
taken to achieve activity accounting for the total waste manage-
ment program. In particular, a field measurement for the radio-
activity content of radioactive ash containers and compacted
waste drums.
General
The radioactive waste management site currently operated by
Ontario Hydro is located at the Bruce Nuclear Power Development
on the east shoreline of Lake Huron about 140 miles north-west of
Toronto. The site consists of waste processing facilities,namely
an incinerator and compactor, and various types of storage struc-
tures for the different categories of solid radioactive wastes.
The actual site layout was developed in the following stages:
. Stage 1 - concrete trench and tile hole storage
structures.
. Stage 2 - waste volume reduction facility.
. Stage 3 - concrete trench and tile hole storage
structures.
. Stage A - concrete quadricell storage structures
for bulk resin storage.
The waste volume reduction facility or W.V.R.F. includes a
starved-air type radioactive waste incinerator (Trecan, Canada)
and a compactor (Stock Equip. Co.) for processing the large volu-
me, low specific activity wastes prior to storage.
The various concrete storage structures have the following
major design criteria:
. minimum operational life of 50 years.
. radioactive contents retrievable.
. isolated from ground water.
. adequate shielding.
-------
Solid Radioactive
Was tes from
Nuclear Generating
Sta tIons
100% - volume and activity
Non-
processible
was tes
12%- volume
98.4% - activity
Processib
wastes
88%- volume
1. 6%- ac tivity
Combus t ibles
- Incineration
63% - volume
0.05%- activity
Non-combus-
tibles (PVC,
metals)
- Compaction
- trenches /Storage
- trenches
FIG 1 - RADIOACTIVE WASTE MANAGEMENT PROCESS
25% - volume
1.55%- activity
-------
198
The radioactive wastes currently processed and stored at
this site originate from the nuclear generating stations at
B.N.P.D. and Pickering.
Waste categories
The solid radioactive wastes originating from the nuclear
generating stations are classified into the following categories
prior to any volume reduction processes:
Solid Waste
Category
Type 1
Type 2
Type 3
Beta-Gamma
(Nominal Ci
< 0.1
0.1 to
> 100
Activity
/m3 <" )
100
For the purpose of segregating the waste packages, direct
gamma dose rate measurements are used with Type 1 being equiva-
lent to a contact dose rate of less than 200 m rad/h, Type 2 less
than 15 rad/h and Type 3 in excess of 15 rad/h.
Type 1 wastes are typically the general dry garbage with low
levels of radioactive contamination such as waste paper, used
protective clothing, metal and plastic scrap materials. The type
2 and 3 radioactive wastes are usually associated with the reac-
tor system and are typically discarded filters, ion-exchange co-
lumns, bulk ion-exchange resin and reactor components. In gene-
ral, the Type 1 and 2 level wastes are stored in the concrete
trenches and the Type 3 in the concrete tile holes and quadri-
cells.
Waste management process
The overall radioactive waste management process for the
site is illustrated in Fig. 1. The Type 2 and 3 wastes are nor-
mally regarded as non-processible and transferred directly to
the storage structures. This accounts for about 12% of the
wastes by volume and greater than 98% of the wastes by activity
content. The majority of the Type 1 wastes are processed in the
W.V.R.F. prior to storage.
(1) Nominal curie -
that quantity of beta gamma emitting
radioactive material which emits
3.7 x 10
second .
i o
photons of 0.8 MeV per
-------
EFFLUENT TO
ATMOSPHERE
RELIEF '
BATCH LOADING DOOR
FOR RADIOACTIVE WASTE
BAGHOUSE
FILTER
1600F - 1800F
AFTERBURNER
PRIMARY
INCINERATOR
CHAMBER
1000F
HEAT
EXCHANGER
IGNITION BURNERS
A
PROPANE
FUEL
INDUCED
DRAFT FAN
CONTAINER FOR
RADIOACTIVE ASH
HEAT
EXCHANGER
AIR BLOWER
UNDERFIR
SECONDAR
AIR BLOWER
CONTAINER FOR
RADIOACTIVE ASH
4a
RADIOACTIVITY
MONITOR
VALVES
FOR
GRAVITY
CLEANOUT
LIQUID
WASTE
OIL PUMP
SHELL COOLING
& AFTERBURNER
SECONDARY
AIR BLOWER
•AFTERBURNERS
IGNITION BURNER;;
PRIMARY COMBUSTION
AIR BLOWER
TEMPERATURE INDICATING
CONTROLLER
TR) TEMPERATURE RECORDER
PRESSURE INDICATING
CONTROLLER
TEMPERATURE INDICATOR
UD
FIGURE 2 INCINERATOR SCHEMATIC
-------
200
These wastes are further segregated prior to processing
into combustible wastes for the incinerator and non-combustible
wastes, such as P.V.C. materials and metals, for the compactor.
The combustible wastes account for about 63% of the total by
volume and less than 0.05% by activity content. The non-combus-
tible or compacted wastes account for 25% of the total by volume
and about 1.55% by activity content.
The current solid waste volume processed at the site is
about 3000 m3 per year with an estimated total activity content
of about 500 Ci. The incinerator provides an overall volume
reduction of about 25 and the compactor about 2.
Monitoring program - general
The monitoring program for the storage site and waste volume
reduction facility (W.V.R.F.) has been designed to monitor per-
formance in the areas of public safety, worker safety, and
structural integrity of storage facilities. The associated
activities are broken down as follows:
Public safety - radioactivity of liquid effluent
- radioactivity of airborne effluent
- radiation fields at fence line
- radioactivity in ground water
Worker Safety - hazard identification
- working environment surveys
- internal and external dosimetry
- worker knowledge
Structural Integrity - record of stored activity
- radioactivity in ground water
Activity measurements - radioactive incineration
Figure 2 is a schematic of the radioactive waste incinerator
Routine radioactivity measurements have been made on samples of
the incinerator ash, heat exchanger deposits, baghouse ash, and
stack monitor filters. The stack monitor, supplied by Radeco,
provides continuous on-line monitoring for operational control.
The particulate filter, charcoal cartridge, and tritium bubbler
samples are analyzed in the site Health Physics laboratory for
compliance monitoring purposes.
The filters and ash samples are analyzed with a computer -
based gamma spectrometry system. The stack bubbler samples are
analyzed for tritium using an automatic liquid scintillation
counter.
The incinerator is loaded with 20 to 22 m3 of low level
radioactive waste in 0.05 m3 plastic bags. The waste averages
about 1% of the activity limit defined for the type 1 waste
category (excluding tritium).
-------
201
TABLE 1; LOW LEVEL RAP-WASTE INCINERATION - "TYPICAL"
PARTICULATE ACTIVITY
(yCi/burn) (%)
Incinerator Ash
Baghouse Ash
Stack Effluent
19000
2000
75
90
9.6
0.4
1-131
(yCi/burn)
50
H-3
(uCi/burn)
10'
TABLE 2: RADIONUCLIDE DISTRIBUTION FOR INCINERATION
RADIONUCLIDE
Ce-144
Sb-124
Cr-51
Ru-103
Cs-134
Ru-106
Cs-137
Zr-95
Nb-95
Zn-65
Co-60
La-140
Others
INCINERATION
ASH (%)
15.1
N.D.
4.5
3.0
1.8
5.2
5.6
9.1
16.0
5.0
14.8
15.2
4.7
BAGHOUSE
ASH (%)
< 1
10
N.D.
< 1
10
7
49
< 1
< 1
20
3
N.D.
1
STACK EFFLUENT
(%)
1.7
N.D.
N.D.
4.3
11.7
< 1
54.8
< 1
N.D.
25.1
< 1
N.D.
2.4
(Gd-153, Ce-141, Sn-113,
Sb-125, Mn-54, Fe-59, Ba-140)
100%
100%
100%
-------
202
The process reduces the waste to about 0.5 m 3 of ash of which
90% remains in the incinerator primary chamber and 10% is trapped
in the baghouse filters.
The distribution of the waste activity following a typical
burn is shown in table 1. 100% of the radioiodine and tritium
is driven off in the stack emissions. About 0.4% of the total
particulate activity is observed in the stack emissions. As in
the case of the radioiodine and tritium, the final distribution
of radionuclides in the particulate activity depends on the
physical and chemical properties of the elements involved (see
table 2). For instance, cesium, which has relatively low melting
point (29°C) and boiling point (670°C), constitutes a large
fraction of the activity found in the ash in the baghouse (ope-
rates at 200°C) and in the effluent gases. In the other extreme,
chromium has a melting point of 1900°C and a boiling point of
2200°C and is not detectable in the baghouse ash or in the efflu-
ent gases, but remains in the primary chamber.
Field measurements
In order to facilitate the routine measurement and account-
ing of curie content of processed wastes under field conditions,
relationships have been determined for field measurements using
gamma dose-rate survey meters.
Ash from the incinerator is discharged into 2.5 m3 steel
cubic containers prior to storage. Equation (1) is the relation-
ship determined for this type of container.
Curie content (yCi/m3) = 10^ x mrem/h (@ 1 ft)
— (1)
Non-combustibles are compacted into 0.2 m3 drums for which
the relationship in equation (2) was determined.
Curie content (yCi/m3) = 2 x mrem/h (contact)
— (2)
Future programs
No attempt has been made to date to analyze for Carbon-14
or alpha-emitters. This work is currently in the development
stage .
-------
203
A COMPARISON OF ALTERNATIVES FOR
LOW-LEVEL RADIOACTIVE WASTE DISPOSAL
By
Paul J. Macbeth
Ford, Bacon & Davis Utah, Inc.
ABSTRACT
A comparative analysis of alternatives for disposal of low-
level radioactive wastes has been performed for the U.S. NRC.
A systematic evaluation of all possible disposal mechanisms
identifying options most viable for further analysis is presented.
Generic reference disposal facility concepts for each viable al-
ternative are evaluated to provide a consistent, meaningful com-
parison based on technological, economic, and sociopolitical
factors. The results of the comparative analysis are presented
in a convenient matrix format to facilitate intercomparisons and
to promote understanding of the complexities of the tradeoffs
involved in selecting waste disposal options.
The concepts judged to be the most viable alternatives
to the current practice in this country of disposal of low-level
wastes by shallow land burial include improvements to shallow
land burial, ocean disposal, intermediate depth burial (10-15 m
deep), disposal in natural or mined cavities, and disposal in ex-
posed or covered structures. Representative waste disposal
facility concepts for each of these alternatives were analyzed
as the basis of the evaluation, using reference waste volumes and
facility lifetimes.
I. INTRODUCTION
This paper describes an evaluation of alternative methods
for the disposal of low-level radioactive wastes performed for
the U.S. Nuclear Regulatory Commission (NRC) by Ford, Bacon &
Davis Utah, Inc. (FBDU). Alternative methods for waste disposal
must be evaluated to assure that safer or more effective tech-
niques are not overlooked.
II. SCREENING OF ALTERNATIVES
A comprehensive review of all possible methods which have
been identified or proposed for low-level radioactive waste dis-
posal was performed, based on a systematic methodology for iden-
tifying disposal options ensuring that no viable choices have
been overlooked. This first study objective included systemat-
ically identifying, cataloging and describing possible low-level
waste disposal alternatives.
-------
204
The range of all possible low-level radioactive waste
disposal alternatives was divided into categories to provide a
systematic means for identifying all options. The categories
were subdivided to arrive at specific disposal methods, shown
in Figure 1. After analysis and review, those alternatives
warranting further evaluation were selected. The selected alter-
natives are the basis for this report, and include the base case
of typical shallow land burial, improvements to present practices,
deeper burial, disposal in mined cavities, disposal in engineered
structures, and disposal in the ocean.
To assure completeness of the initial listing and adequacy
of the selection of viable alternatives, a panel of technically
competent individuals of recognized waste management expertise
was consulted for review and guidance. A formal report of the
results from this phase of the study has been published.1
The most viable alternatives selected are compared with
current solid low-level waste disposal by shallow land burial
using a rigorous and detailed analysis. The results of this
effort are presented in a convenient matrix format to facilitate
their use in decisions pertaining to selection of viable alter-
natives in national low-level waste management programs.
For the generic alternatives selected for further evalua-
tion, several additional factors require specification to allow
a meaningful comparative analysis. These factors include the
location, size, and type of disposal facility designed for each
alternative method. For this study generic Eastern U.S. and
Western U.S. locations and possible ocean disposal sites were
assumed to obtain transportation factors, a volume of waste to
be accommodated was given, and the disposal facilities were con-
ceptually designed to reasonably accommodate and contain the
wastes.
Based on the reference disposal facilities for each alter-
native method studied, technical, sociopolitical and economic
factors were evaluated as the basis of a comparative analysis.
Values for the parameters required for performing the evaluations
are specified; but it should be understood that the performance
of any actual waste disposal facility will depend on the condi-
tions that exist at the real site, which may vary from those
assumed for this study. By changing some of the site- and
facility- specific factors, some of the conclusions of this
comparative analysis could be changed. However, a uniform,
consistent approach has been taken for all alternatives evaluated
in this report, which provides a rational basis for waste manage-
ment decisions and allows appropriate evaluation of the tradeoffs
involved in disposal option selection.
III. EVALUATIONS
Reference disposal facilities for each alternative have
been selected as a base case in performing comparative analyses
-------
205
of the clifferent types of disposal options. These reference
facilities are all based on disposal capacity for a constant
volume of waste having a given radioactivity inventory. The
conceptual designs of the facilities represent an estimation
of the types of design criteria that may be required in the
future for waste disposal sites. The results from disposal
of waste in the reference facility provide a uniform basis for
comparison of the alternatives with both costs and effects being
appropriate indices for the comparison. Other approaches for
the comparative analysis coudl be selected. For instance,
either the costs for construction or the resultant effects from
waste disposal at a reference facility could be held constant,
the designs varied, and the comparison based on non-fixed vari-
ables. The approach taken in this study, however, provides a
consistent basis for comparing alternatives, and is appropriate
for the preconceptual design 'stage of development of the refer-
ence disposal facilities.
The volume of waste to be disposed of was assumed to be
630,000 m3, which roughly corresponds to the expected output of
1,000 typical light-water reactors for one year (1,000 Reference
Reactor Years (RRY) of low-level waste).2 This volume of waste
would correspond to roughly 800,000 megawatt-years of electricity
production (MW(e)-yr). Wastes from non-fuel cycle sources will
also be accommodated in the reference facilities. The generic
inventory has been adjusted to account for wastes from both sources,
The generic reference facilities were assumed to handle this volume
of waste in a twenty-year operating period.
The environmental effects are subdivided into non-radiologi-
cal and radiological impacts. The non-radiological effects in-
clude impacts on construction and waste management workers.
Radiological impacts include direct radiation exposures to workers
and the public along the transportation route and in the area of
the disposal facility.
The exposure pathways calculated for the various alterna-
tives may not necessarily be an exhaustive listing of all the
possible mechanisms for human exposure at each site. However,
they do provide a consistent basis for comparison of the alter-
natives and are representative of the most important types of
impact that would be expected from implementation of waste dis-
posal operations. Obviously, by changing the conceptual design,
the radiological impacts would be changed, as would be the asso-
ciated costs. Prior to implementation of any of the alternatives,
it is anticipated that formal cost-benefit analyses and trade-
offs would be performed to optimize the results from the selected
option. The analyses reported in this study are useful, however,
in providing perspective and guidance for selection among the
various alternatives, and are presented for that purpose.
Institutional control over disposal sites is assumed to be
maintained for 150 years after operations cease. Any future site
reclamation efforts would occur after that time period.
-------
206
The assessment of sociopolitical implications is somewhat
subjective. However, available published research and informa-
tion on the topic3'4 has beer used for guidance. Additionally,
many of the social acceptance issues depend on adequate demon-
stration that the technological problems have been appropriately
solved. Assuming that the disposal alternatives meet the minimum
constraints of being technically sound, the sociopolitical issues
hinge mainly on requirements for governmental agreement and con-
trol, as is the case with ocean disposal in international waters.
These issues are considered in the weighting factors used in the
comparative analysis.
IV. COMPARATIVE ANALYSES
After completion of the technical, sociopolitical and cost
evaluations for each concept, the major factors relating to each
of these areas are quantified. Some of the items important to
the comparison of alternatives, however, can be quantified only
by subjectively ranking one concept against another. Other items
(such as cost, for example) are quantified during the course of
the evaluation. Care must therefore be exercised to assure that
the different alternatives are uniformly assessed.
Once the important evaluation factors and parameters have
been quantified for each of the disposal alternatives, the
factors and the alternatives are jointly displayed in a conven-
ient matrix format. One additional factor in the comparison is
an estimate of the relative importance (weighting) of the evalua-
tion factors and parameters used in the comparison. For instance,
an estimate of how heavily costs should be considered in relation
to sociopolitical issues allows comparison of the different con-
cepts.
The weighting factors are somewhat subjective. The weight-
ing factors used in this report were determined from a survey of
knowledgeable persons serving as advisors on this study. The
use of these weighting factors allows quantitative comparison of
the alternatives, but does not mean that others may not be more
appropriate for different circumstances.
The comparative analysis is useful in demonstrating that
the selection of best or optimum alternatives for low-level
radioactive waste disposal involves complex tradeoffs among
several factors. It also shows that there is more than one
method of safely handling low-level radioactive wastes. However,
going from the generic generalized concepts studied in this pro-
ject to specific designs at real sites will lead to important
differences in the values of the evaluation parameters. This
comparative analysis should, therefore, be used with care for
guidance in selecting optimum choices, and not be directly
applied to specific, actual disposal operations.
The radiological effects from waste disposal are summarized
in Table 1. Increased transportation exposures to the western
site account for the differences between eastern and western
-------
207
locations. Improving the base case would reduce exposures by
about 20%. Disposal in mined cavities would eliminate about
90% of the potential doses. Disposal in structures would in-
crease potential exposures by about 20% because of greater
availability of the waste to future reclamation activities.
Ocean disposal would eliminate all but about 20% of the potential
exposures from the base case. In general, it can be noted that
reclamation events contribute the most significant portion of
the potential doses. Elimination of the possibility of recla-
mation after only 150 years beyond disposal would effectively
reduce the consequences of waste disposal activities.
Non-radiological effects are based on accident statistics
from comparable industries,5 and are summarized in Table 2.
Crew sizes are estimated, and transportation risks are calcu-
lated based on shipping distances.6 Because of the larger
construction crew sizes, the structural disposal concepts pro-
vide approximately twice as large an impact as the base case.
Cost estimates for the preconceptual design facilities are
presented in Table 3. Capital and operating costs are differen-
tiated, and profit, escallation and financing charges included.
Monitoring and surveillance activities for 150 years have been
included in the operating budgets.
Table 4 contains the overall comparative analysis of all
alternatives. Weighting factors are included for the various
evaluation parameters. Selection of other weighting factors
could change some of the conclusions of this analysis. However,
a consistent approach has been taken and some insight into the
tradeoffs involved obtained.
REFERENCES
P.J.Macbeth, et al, "Screening of Alternative Methods for
the Disposal of Low-Level Radioactive Wastes" NUREG/CR-
0308, October, 1978.
Alternatives for Managing Wastes from Reactors and Post
Fission Operations in the LWR Fuel Cycle, ERDA-76-43,May 1976.
F.Peret, "Radioactive Waste Storage and Disposal: Methodol-
ogies for Impact Assessment" Ph.D. Dissertation, University
of California, Berkley, 1975.
J.W.Bartlett, et al, "Advanced Methods for Management and
Disposal of Radioactive Wastes" BNWL-1978, March,1976.
"Accident Facts-1977 Edition" National Safety Council,
Chicago, 111./ 1977.
"Environmental Survey of Transportation of Radioactive Mat-
erials to and from Nuclear Power Plants" WASH-1238, Dec. 1972,
-------
TABLE 1
SUMMARY OF RADIOLOGICAL EFFECT FOR ALTERNATIVES
Reclamation Events
Alternative
Shallow-Land Burial-Eastern Site
Shallow-Land Burial-Western Site
Improved Burial-Eastern Site
Improved Burial-Western Site
Deeper Burial-Eastern Site
Deeper Burial-Western Site
Abandoned Mine-Eastern Site
Abandoned Mine-Western Site
New Horizontal Shaft Mine-Eastern Site
New Horizontal Shaft Mine-Western Site
New Vertical Shaft Mine-Eastern Site
New Vertical Shaft Mine-Western Site
Above Grade Structure-Eastern Site
Above Grade Structure-Western Site
Buried Structure-Eastern Site
Buried Structure-Western Site
Direct Ocean Dumping
Ocean Projectile Disposal
Short Term Events
Total
Hell Hater Single Cununulative
Inhalation/Direct Gamma/Food Transportation/Consumption/Container Accidents Effect
110 340
110
99
99
0
0
0
0
0
0
0
0
220
220
220
220
0
0
340
310
310
0
0
0
0
O
0
0
0
670
670
670
670
0
0
625
625
560
560
0
0
0
0
0
0
0
0
625
625
625
625
1
0
9.5
33
9.5
33
9.5
33
14
38
14
38
14
38
9.5
33
9.5
33
38
38
80
80
1.5
1.5
1.5
1.5
0
0
0
0
0
0
1.5
1.5
1.5
1.5
O
0
200
200
150
150
150
150
100
100
100
100
100
100
100
100
100
100
200
200
1364
1388
1130
1154
161
184
114
138
114
138
114
138
1626
1650
1626
1650
239
238
Normalize
Effect
1.0
1.0
0.8
0.8
0.1
0.1
0.1
0.1
0.1
0.1
0.1
0.1
1.2
1.2
1.2
1.2
0.2
0.2
r\s
O
CO
-------
TABLE 2
SUMMARY OF NON-RADIOLOGICAL EFFECTS FOR ALTERNATIVES
Transportation
Alternatives
Shallow-Land Burial-Eastern Site
Shallow-Land Burial-Western Site
Improved Burial-Eastern Site
Improved Burial-Western Site
Deeper Burial-Eastern Site
Deeper Burial-Western Site
Abandoned Mine-Eastern Site
Abandoned Mine-Western Site
New Horizontal Shaft Mine-Eastern Site
New Horizontal Shaft Mine-Western Site
New Vertical Shaft Mine-Eastern Site
New Vertical Shaft Mine-Western Site
Above Grade Structure-Eastern Site
Above Grade Structure-Western Site
Buried Structure-Eastern Site
Buried Structure-Western Site
Direct Ocean Dumping
Ocean Projectile Disposal
Total Train
Car Miles
528,000
1,848,000
528,000
1,848,000
528,000
1,848,000
792,000
2,112,000
792,000
2,112,000
792,000
2,112,000
528,000
1,848,000
528,000
1,848,000
2,112,000
2,112,000
Total
Injuries
0.20
0.70
0.20
0.70
0.20
0.70
0.30
0.8O
0.30
0.80
O.30
0.80
0.20
0.70
0.20
0.70
0.80
0.80
Total
Fatalities
0.002
0.01
0.002
0.01
0.002
0.01
0.003
0.01
0.003
0.01
0.003
0.01
0.002
0.01
0.002
0.01
0.01
0.01
Construction
Crew Size
(My)
20
17
22
20
31
28
12
12
54
54
62
62
330
326
363
352
NA
NA
Total
Total
Injuries Fatalities
0.61
0.52
0.67
0.61
0.95
0.85
0.63
0.63
2.84
2.84
3.26
3.26
10.06
9.94
11.07
10.73
NA
NA
0.01
0.01
0.01
0.01
0.01
0.01
0.01
0.01
0.06
0.06
0.07
0.07
0.12
0.12
0.13
0.12
NA
NA
Crew Size
(My)
265
265
265
265
265
265
265
265
265
265
265
265
265
265
265
265
300
300
Operations
Total
Injuries
8.06
8.06
8.06
8.06
8.06
8.06
8.06
8.06
8.06
8.06
8.06
8.06
8.06
8.06
8.06
8.06
9.15
9.15
Total
Fatalities
0.09
0.09
0.09
0.09
0.09
0.09
0.09
0.09
0.09
0.09
0.09
0.09
0.09
0.09
0.09
0.09
0.11
0.11
Cummulative
Effects
9.9
10
10
10
10
10
10
10
12
13
13
13
20,
20.
21.
21.
11.
11.
.4
.0
.5
.2
.7
.0
.6
.7
.3
.3
.8
.4
.9
.6
.7
,2
2
r\i
o
vo
-------
TABLE 3
COST ESTIMATE SUMMARY FOR ALTERNATIVES ($MILLIONS)
1.
2.
3.
4.
5.
6.
7.
8.
9.
10.
11.
2.
3.
4.
5.
6.
7.
8,
ALTERNATIVE
Shallow-Land Burial-Eastern
Site
Shallow-Land Burial-Western
Site
Improved Burial-Eastern Site
Improved Burial-Western Site
Deeper Burial-Eastern Site
Deeper Burial-Western Site
Abandoned Mine-Eastern Site
Abandoned Mine-Western Site
New Horizontal Shaft Mine
-Eastern Site
New Horizontal Shaft Mine
-Western Site
New Vertical Shaft Mine
-Eastern Site
New Verticle Shaft Mine
-Western Site
Above Grade Structure
-Eastern Site
Above Grade Structure
-Western Site
Buried Structure
-Eastern Site
Buried Structure
-Western Site
Direct Ocean Dumping
Ocean Projectile Disposal
Capital
Costs
12.26
9.42
13.81
10.98
18.70
15.87
7.06
7.06
29.57
29.57
33.73
33.73
178.33
175.60
192.13
189.40
3.65
3.65
Operating
Costs
23.65
23.65
23.65
23.65
23.75
23.75
24.47
24.47
24.47
24.47
24.47
24.47
24.63
24.63
24.63
24.63
73.94
484.67
Summary
Contln-
gency
10.77
9.92
11.24
10.39
12.74
11.89
9.46
9.46
16.21
16.21
17.46
17.46
60.89
60.07
65.03
64.21
23.28
146.50
Financing
Escallation
& Profit
35.22
29.44
38.29
32.58
48.18
42.48
22.55
22.55
70.40
70.40
78.78
78.78
121.31
115.81
128.52
123.02
40.04
221.58
Total
Facility
Costs
81.90
72.44
86.99
77.60
103.37
93.99
63.54
63.54
140.65
140.65
154.44
154.44
385.16
376.11
410.31
401.26
140.91
856.40
Transpor-
tation
Costs
67.74
237.10
67.74
237.10
67.74
237.10
101.60
270.96
101.60
270.96
101.60
270.96
67.74
237.10
67.74
237.10
101. SO
101.60
Total Total Unit,
Costs Costs ($/m )
149.64
309.54
154.73
314.70
171.11
331.09
165.14
334.50
242.25
411.61
256.04
425.40
452.90
613.21
478.05
638.36
242.51
958.00
238
491
246
499
272
525
262
531
384
653
406
675
719
973
759
1013
385
1521
Normal-
ization
Costs
1
2
1
2
1
2
1
2
1
2
1
2
3
4
3
4
1
6
.0
.1
.0
.1
.1
.2
.1
.2
.6
.7
.7
.8
.0
.1
.2
.3
.6
.4
-------
TABLE 4
WEIGHTED COMPARATIVE ANALYSIS FOR ALTERNATIVES
uompatiDility/Site/Saf eguards/Env./Availability/Inst . /Public /Individual /Industry Weighted
with Waste/Selection/ /Effects/of Technique/Control/Accept . /Costs /Costs /Coinparisti
Weighting Factor
Alternative
Shallow-Land Burial-Eastern Site
Shallow- Land Burial-Western Site
Improved Burial-Eastern Site
Improved Burial-Western Site
Deeper Burial-Eastern Site
Deeper Burial-Western Site
Abandoned Mine-Eastern Site
Abandoned Mine-Western Site
New Horizontal Shaft Mine-Eastern Site
New Horizontal Shaft Mine-Western Site
New Vertical Shaft Mine-Eastern Site
New Vertical Shaft Mine-Western Site
Above Grade Structure-Eastern Site
Above Grade Structure-Western Site
Buried Structure-Eastern Site
Buried Structure-Western Site
Direct Ocean Dumping
Ocean Projectile Disposal
0
0.
0.
0.
0.
0.
0.
0 .
0.
0.
0.
0.
0.
0.
0.
0.
0.
0.
0.
.075
075
075
075
075
075
075
075
075
075
075
075
075
075
075
075
075
075
075
0.117
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0
.117
.117
.117
.117
.140
. 129
.176
. 164
. 164
. 152
.152
. 140
.105
.094
.105
.094
.140
. 140
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0.065
.065
.065
.065
.065
.059
. 059
.052
.052
.052
.052
.052
.052
. 078
.078
.072
.072
.033
.033
0. Ill
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0
.111
.111
.100
. Ill
.056
.067
.056
. 067
.078
.078
.078
.089
.178
.178
. 189
. 189
.078
.078
0
0.
0.
0.
0.
0.
0.
0.
0.
0.
0.
0.
0.
0.
0.
0.
0.
0.
0.
.095
095
095
095
095
105
105
114
114
124
124
124
124
105
105
105
105
095
124
0.108
0.
0.
0.
0.
0.
0.
0.
0.
0.
0.
0.
0.
0.
0.
0.
0.
0.
0.
108
108
108
108
119
119
130
130
130
130
130
130
130
130
119
119
130
130
0.161
0.161
0.161
0.144
0.144
0. 128
0. 128
0.128
0.128
0. 112
0.112
0.112
0.112
0. 176
0. 176
0. 144
0. 144
0. 224
0.192
0.143
0.143
0. 300
0. 143
0.300
0. 157
0.315
0.157
0.315
0. 229
0.386
0.243
0.400
0. 429
0.586
0. 458
0.615
0.229
0.915
0
0
0
0
•0
0
0
0
0
0
0
0
0.
0
0.
0
0
0
0.125
. 125
.263
. 125
.263
. 138
.275 '
.138
.275
. 200
.338
.213
.350
375
.513
400
.538
.200
.800
1.00
1. 00
1.29
0. 97
1.28
0.98
1.27
1.03
1.32
1. 16
1.45
1.18
1.47
1.65
1.94
1.67
1.95
1.20
2.49
IN3
-------
212
CATEGORIZATION OF ALTERNATIVES FOR DISPOSAL
OF LOW-LEVEL RADIOACTIVE WASTES
EXTRATERRESTRIAL
DISPOSAL
HIGH EARTH ORBIT
SOLAR & PLANET
IMPACT
SOLAR ORBITS
SOLAR ESCAPE
ORBITING
ACCELERATOR
2. ATMOSPHERIC DISPOSAL
DIRECT
VIA JUPITER SWINGBY
• SINGLE BURN BEYOND EARTH ESCAPE
• CIRCULAR SOLAR ORBIT
• VENUS OR MARS SWINGBY
• DIRECT
• VIA JUPITER SWINGBY
yNOER LINED COUNTRY :
"i ME'THOO USED IN PAST OR
IN USE TODAY 1
NON-UNDERLINED COUNTRY:
{METHOD IS UNDER REVIEW 1
• EJECTION FROM EARTH ORBIT
UPPER ATMOSPHERE
LOWER ATMOSPHERE
• ROCKETS
• AIRCRAFT
• BALLOONS
USA. NEA, IAEA
USA, NEA, IAEA
USA, NEA, IAEA
USA
y§*. £i*!ADA, FOR, UK,
Ei*!£li NETHERLANDS
ysOLisLfiiuSTAlA."
NEA
• ALLCOUNTRIES WITH NUCLEAR FACILITIES USE STACKS: HOWEVER.
NOT EVERY COUNTRY HAS PUBLISHED REFERENCES TO STATE THIS.
3. DISPOSAL TO WATE RS i/ii BELGIUM, FRANCE,
1
^j OCEAN DUMPING J
V /
>j DISCHARGES TO f
•
4. CRUSTAL DISPOSAL
N^
MIXING WITH SOIL f
V
"*"! RELEASE TO AQUIFERS ^»
LAND DISPOSAL f*»
1 I jb'
• LAKES ••••
• RIVERS ••••
• SEAS ••••
• COASTAL WATERS ••••
f
••• THIS DISPOSAL METHOD IS PRACTICED BY /
ALL COUNTRIES GENERALLY. WHICH / /
HAVE OPERATING REACTORS. IAEA. NEA f f
• TAILINGS PONDS
• SOIL CRIBS
• DIRECT MIXING
• LEACHING PONDS
'//
• PERCHED WATER ZONES ^ 1 I
• REGIONAL WATERTABLES lit
• STABLE CONNATE WATER ZONES | f \
• SHALLOW BURIAL W \
• DEEP BURIAL f
• NATURAL CAVITIES OR CAVERNS ^T ^^^
• MINED CAVITIES
^
• EXPLODED CAVITIES p^ USA
FORMATIONS P"l • DRILLED HOLES K J
X0| • ROCK MELTING t^. USA V
^
k
X DISPOSAL IN ICE _.
SHEETS
SUBMANTLE DISPOSAL }f*
• HYOROFRACTURE p^ ^fL
• SEABED DISPOSAL BY EMPLACEMENT ^^^
• SHALLOW PITS
• DRILLED HOLES
• CAVERNS
• MELTDOWN CONCEPT
• FREE FLOW CONCEPT
• INJECTION INTO SHALLOW MAGMA
CHAMBERS
^^
V N
\
ITALY. NETHERLANDS,
SWITZERLAND. JAPAN.
DENMARK. INDIA, IAEA,
USA, UK, JAPAN
USA, USSR^CSSR. GDR,
IAEA, NEA
USA, CANADA, FRANCE,
USSR, FOR, GOR. JAPAN.
DENMARK, NORWAY, UK
.CSSR, IAEA, NEA
SWITZERLAND. SWEDEN,
AUSTRIA, INDIA, CSSR,
USA, CANADA, FRANCE.
NETHERLANDS, UK, NEA,
ROMANIA, SPAIN, USSR,
DENMARK, POLAND. GOR.
IAEA, FOR JAPAN, BELGIUM
• EVAPORITESISALT/
ANHYDRITESI
• CRYSTALLINE AND
• SHALE AND CLAY
• CLASTICS
• DEEP CONFINED STRATA
• SALT DOMES
• SALT BEDS
• CRYSTALLINE IGNEOUS
ROCK
USA, CANADA. FRANCE,
NETHERLANDS. SWEDEN,
N CSSR, OOR. IAEA. NEA,
USSR
• HIGH SEDIMENTATION
AREAS
• DEEP STABLE SEABED
• SUBOUCTION ZONES,
DEEP TRENCHES
USA, CANADA, FRANCI,
AUSTRALIA, AUSTRIA,
ITALY, JAPAN, FOR. UK.
IAEA, NEA
USA, IAEA, NEA
5.
STRUCTURAL CONCEPTS
ABOVE GRADE
BELOW GRADE
• CONCRETE STRUCTURES
• METAL STRUCTURES
• TANKS
• OTHER
• CONCRETE STRUCTURES
• METAL STRUCTURES
• TANKS
• OTHER
USA, SJ.LGm.M. CANAUA,
• UK. FRANCE" FOR,! INDIA.
lAEA.TlEA
USA. BELGIUM, CANADA,
• UK. FRANCE^ FGR. INDIA.
6. DISPOSAL BY CONVERSION
V _
N_
TRANSMUTATION |T
• FISSION REACTORS
• FUSION REACTORS
• ACCELERATORS
• THERMONUCLEAR DEVICES
FIGURE 1.
USA. JAPAN. BELGIUM,
, ITALY, KiR, IAEA, NEA,
AUSTRALIA, FRANCE
-------
213
SESSION D
REGULATORY ASPECTS
Session Chairperson
C. R. Price
State of Virginia
-------
214
DISPOSAL CLASSIFICATION OF LOW-LEVEL RADIOACTIVE WASTE
J.A. Adam
Nuclear Regulatory Commission
I will be presenting a system for classifying radioactive waste that
has been developed for the Nuclear Regulatory Commission (NRC) by Ford,
Bacon & Davis Utah under the lead of Dr. Vern Rogers. It is a system for
classifying waste based on the minimum requirements for safe disposal. It
was first given public attention in NUREG-0456, "A Classification System
for Radioactive Waste Disposal - What Waste Goes Where?", June 1978. That
document reported a study in progress and did not represent the fully devel-
oped system. We have made considerable progress since the publication of
NUREG-0456 and today I will present the complete classification system.
I would like to emphasize that the system which I am going to present
is not being proposed by NRC at this time. The system will be one of
several alternatives that will be considered during the development of a
waste classification regulation.
Early in the development of the classification system, it was decided
that the system should be based on the requirements for safe disposal. We
recognized that there are varying degrees of toxicity of radioactive wastes
and various degrees of confinement that can be achieved by disposal.
Three generic disposal options were considered:
- disposal as ordinary (non-radioactive) waste,
- containment in a licensed disposal facility, and
- isolation (i.e., deep geologic repository).
For the containment option, variations based on the accessibility of
the waste to man, periods of administrative control or restricted land use,
and presence or absence of groundwater were also considered.
The steps in developing the complete system are:
- develop study guidelines (what is safe?),
- determine the pathways to man,
- formulate the classes to be considered,
- determine disposal concentration guides, and finally
- classify the waste from the major waste stream or sources.
The first step, defining safe disposal, was accomplished by formulat-
ing a set of study guidelines. The study guidelines dealt with exposures
to few and many individuals, "as low as reasonably achievable" (ALARA),
positive net benefit, and periods of administrative control or restricted
land use.
Two of the study guidelines turned out to be dominant; restrictions on
the exposure of the maximally exposed individuals and the period of admin-
istrative control. We used 500 mrem/year, whole body or organ, as the ex-
-------
215
posure limitation for the critical individuals and 150 years as the period
of administrative control.
Because the analyses that were used in developing the classification
system are consequence rather than risk analyses, the use of a 500 mrem per
year exposure restriction for the maximally exposed individuals is con-
sidered to be conservative. We chose to use consequence rather than risk
analyses because of the difficulties in assigning probabilities to the numer-
ous events that can take place over the many years involved (risk equals
probability times consequence). This is particularly true when dealing with
generic disposal rather than waste and site specific disposal.
As an analogy, consider the risk from the head-on collision of two cars.
The risk is the consequence of a collision times the probability of a colli-
sion. For any given speed the risk is greater on a curvy, two-lane road
than on a divided interstate highway. Consequence alone is not an adequate
measure of risk. However, if there is an identifiable speed at which head-
on collisions would not result in any serious injury, then we could assert
that driving at that speed is safe without any statement of risk. Likewise,
if we can show that the consequence from waste disposal will not result in
any serious injury, we can assert that such disposal is safe without state-
ments of risk. In this context, 500 mrem per year to the maximally exposed
individuals is used to define safe disposal and not to estimate the degree
of risk. Further, just as the proper design of roads reduces the risk of
driving, we envision that reasonable caution by future man and other waste
management criteria such as careful site selection will reduce the risks
far below the levels of the postulated consequences.
More simply, we can insure that wastes will be safely disposed of by
insuring that they are sent to disposal facilities capable of their safe
disposal. This can be done by classifying the wastes according to minimum
requirements for safe disposal as determined by consequence analysis. Other
criteria and standards for other functions, such as waste preparation and
site selection, can be used to reduce the risks from disposal to as low as
reasonably achievable. This conservative approach was adopted because of
the difficulties in specifying detailed site characteristics and probabili-
ties for a non-site specific waste classification system.
To identify an appropriate period of restricted land use a series of
calculations were made. These calculations were for the change in potential
exposures with various periods of decay for typical mixtures of isotopes.
The results of the various calculations were fairly consistent. During the
first hundred years or so, the very short lived isotopes control the level
of potential exposures. After a few hundreds of years the long-lived iso-
topes, in particular carbon-14, control. From about one hundred years to
a few hundreds of years cesium-137 controls. Typically, the decay between
one hundred and one hundred and fifty years resulted in a factor of 2.5 re-
duction in potential exposures. The decay between one hundred and fifty
years and two hundred years resulted in a reduction of 1.5. We chose one
hundred and fifty years as an appropriate period for restricted land use.
It should be noted that the restricted land use we postulate is passive in
nature as compared to active administrative control. Restricted land use
does not mean complete restriction. Use of the land, such as for use as
an airfield, may be permissible or even desirable.
-------
216
The pathways that were investigated were of two types; onsite reclaimer
events and offsite migration. The onsite reclaimer events include:
- a worker or reclaimer digging into the waste and inhaling resus-
pended waste,
- consumption of food grown onsite,
- consumption of well water obtained from on or near the site, and
- direct exposure of workers or reclaimers.
Routine and accidental airborne releases during disposal operations were
also considered. Because such airborne release are not directly related
to the type of disposal but rather the care taken in preparing and handling
the waste, airborne releases are not included in the classification system.
These generic pathways are used as surrogates for families of more
specific pathway scenarios. For example, a reclaimer may be digging on a
site which has been released from restricted use to dig a basement, emplace
fence posts, lay sewers, or any number of reasons, including unforeseeable
future activities of man.
The offsite migration events include:
- ground water migration of the waste to surface waters (rivers) and
subsequent consumption; and
- surface erosion of the waste to surface waters.
Three conclusions regarding offsite migration were reached. First, off-
site migrations were seldom controlling pathways. Second, groundwater
migration was not affected by periods of administrative control or restrict-
ed land use. And third, offsite migration rates and consequences are
strongly dependent on site specific parameters and could not be adequately
addressed in generic analyses. Because offsite migration is seldom controll-
ing and is highly site specific, it is not included in the waste classifica-
tion system. However, the burden of substantiating that offsite migration
would not be controlling at a given site remains with the classification
study.
There are several observations regarding the onsite reclaimer events.
The consequences of the reclaimer (but not worker) inhalation, direct ex-
posure and food consumption events can be reduced through the use of a
period of restricted land use, or eliminated if the wastes are buried
sufficiently deep. Because the waste concentrations in well water can be
close to maximum near the site, but outside of a restricted area, a period
of administrative control of restricted land use may not reduce the con-
sequences of well water consumption. Siting of disposal facilities in dry
regions can eliminate the well water event for as long as the site remains
dry. However, it can not be asserted, at least generically, that a site
will remain dry indefinitely. Having observed that most of the decay of
typical waste will occur in the first few hundred years it was concluded
that assuming a dry site would remain dry for at least a few hundred years
(150 years) was appropriate.
The classification system postulates five classes of terrestrial dis-
posal. (This does not mean that NRC will consider all five classes when
developing their waste classification reeulat-lnnfi.x
-------
217
CLASS A:
Isolation or the best that can be reasonably accomplished. Because
isolation is intended to be the best that can be achieved, no limitations
are placed on Class A waste by the classification system.
CLASS B:
Man does not have ready access to the waste (i.e., deep burial) and
the site is intially dry. The well water pathway is assumed to exist after
one hundred and fifty years.
CLASS C;
Man does not have ready access to the waste (i.e., deep burial) but
there is ground water present. The well water pathway is assumed to exist
prior to any decay of the waste.
CLASS D:
Man does have access to the waste (i.e., shallow burial) after a period
of administrative control or restricted land use (150 years). The well water
pathway exists prior to decay of the waste. The reclaimer digging into the
waste, food consumption, and direct exposure pathways are assumed to exist
after 150 years of waste decay.
CLASS E:
Man does have access to the waste from the time of disposal (i.e.,
sanitary landfill). The assumed pathways are; worker inhalation of resus-
pended waste, food and well water consumption, and direct exposure.
For class D and E, variations could have been postulated based on the
absence of well water for a dry site. It was concluded that these varia-
tions would not be significant for most wastes because of the dominance of
the inhalation and food pathways. However, for some isotopes, such as
tritium, disposal in a dry site would offer a significant advantage.
The table shows examples of calculated allowable average concentrations
for the various classes of waste. The controlling pathway is also noted.
CLASS AVERAGE ALLOWABLE CONCENTRATIONS (Ci/m3)
H-3 Sr-90 1-129 Pu-239
A -
B 430,OOO1 381 0.3 90
C 941 2.41 0.31 901
D 941 0.022 0.31 O.I3
E 0.052 0.000232 0.141 0.00033
1 2 3
well water consumption food consumption dust inhalation
-------
218
Comparison of the allowable concentration for tritium (H-3) among the
5 classes illustrates the importance of decay and the choice of disposal
options for short lived isotopes. The comparison also shows that when man
has ready access to the waste and there is no decay, food rather than well
water consumption can be the controlling pathway for tritium.
The allowable concentration for 1-129 in class E waste is lower than
for the other classes even though well water is assumed to be the controll-
ing pathway for all the classes. This is the result of not allowing for
dilution of the waste by surrounding soil for class E waste. Class E
assumes uncontrolled (unlicensed) disposal without qualification as to
dilution of the waste, depth of burial or thickness of cover. These may
be unduly conservative assumptions.
If the development of the classification system were to stop with the
calculation of average allowable concentrations the system would be un-
workable for at least two reasons. First, for many important isotopes, the
allowable levels are not readily detected. Second, besides allowable con-
centrations, there needs to be consideration of surface contamination,
activation of structural materials, hot spots, and other physical properties
of the waste which affects disposal. Many of these considerations were
discussed in NUREG-0456.
To make the classification system workable, we are proposing to classify
radioactive waste according to the source of generation. Classification
by source would involve an analysis of existing and newly acquired informa-
tion on the isotopic mix, statistical variances, activity levels, and
physical properties of the waste from the major waste streams. As an example,
the results of such an analysis could be: the waste from the clean-up of
the primary cooling system of a PWR (except certain resins) are Class D,
medical waste (with a few exceptions) are Class D, and the waste from the
cleanup of the secondary cooling system of the PWR under normal operating
conditions are Class E. For classification by source to be successful the
methodology must be workable for both the waste generator and the disposal
facility operator and at the same time must provide a means to insure safe
disposal.
The radioactive waste disposal classification system which has been
described is being proposed to NRC as a practical first step for assuring
the safe disposal of radioactive waste. This system and alternative classifi-
cation systems, along with the development of criteria and standards for
waste disposal and other regulatory requirements, will be considered by NRC
in the development of the waste classification regulation.
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A HISTORICAL REVIEW OF FEDERAL/STATE ROLES IN REGULATING
COMMERCIAL LOW LEVEL RADIOACTIVE WASTE BURIAL GROUNDS
G. Wayne Kerr
U. S. Nuclear Regulatory Commission
Washington, D. C. 20555
Abstract
The subject of disposal of low level radioactive waste has generated
considerable interest among State and Federal agencies, State and Federal
legislatures, the public and the industry in the past three to four years.
The background and regulatory processes which have been followed since
1961 in regulating the low level waste burial grounds provide a useful point
of reference for consideration of the currently evolving changes in this
sector of the nuclear industry. The background is discussed in this paper
as well as the type of activities conducted at the burial sites and the
processes followed in licensing and regulating the existing commercial
burial sites. The paper also discusses the possible future roles of
Federal and State governments in regulating such sites.
Introduction
Almost every sector of the nuclear industry generates high interest
amongst the public and the industry and in government agencies from time to
time. In the Washington area almost every day some article related to nuclear
matters appears in the daily papers. Subjects range from nuclear power plants,
transportation of radioactive materials, high and low level waste, uranium
mills to smoke detectors. The fact that this entire symposium is devoted to
low level radioactive waste management is indicative of the high level of
interest in this subject. The subject of waste disposal, particularly high
level, but also low level to a significant extent is one of the most intract-
able subjects which the technical and regulatory experts must deal with.
Some would argue, however, that the subject is not so much a question
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regarding the technical aspects as it is the political and institutional
aspects. It is also sometimes stated that it is not a lack of technical
knowledge but a lack of decisions which makes the subject intractable.
The high level of public interest in this subject began to surface
in late 1974 when the State of Kentucky issued a report on the results of
a special six-month environmental study at the Maxey Flats waste burial ground.
The report noted the contribution of radioactivity to the surrounding environs
resulting from the operations at the site but concluded that it did not
create a public health hazard. This was followed by an NRC report on the
same site and a January 1976 report by EPA on the same subject. During early
1976, three other events took place which highlighted the attention the
Federal Government was giving to this subject. First, the GAO issued a
report on January 12, 1976 on land disposal of radioactive waste. Second,
the Conservation, Energy and Natural Resources Subcommittee of the House
Committee on Government Operations held hearings on low level waste early in
1976. During FY 1977 authorization hearings for NRC, the subject of regulation
of low level sites was discussed in Congress. Congressional interest in this
subject has remained high through 1978 including various proposals for organi-
zational and institutional frameworks for addressing the subject at the
Federal level. One of the most recent publications is the "Report to the
President by the Interagency Review Group on Nuclear Waste Management"
(DOE, 78).
Why Are There Commercial Low Level Sites?
Prior to 1960, disposal in designated areas of the Atlantic and Pacific
Oceans was the conventional method for the disposal of wastes generated by
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commercial users. In June I960, the Atomic Energy Commission placed a mora-
torium on issuance of new licenses for sea disposal. The AEC established an
interim program in I960 whereby radioactive wastes generated by licensed users
were accepted for burial at the Commission's Oak Ridge National Laboratory
in Tennessee and at the National Reactor Testing Station in Idaho though
existing licenses for sea disposal were permitted to remain in effect.
In September 1962 the first commercial land burial facility located near
Beatty, Nevada was licensed and became available for use by the private
sector. Other commercial burial grounds were established in Kentucky (1962),
New York (1963), Washington (1965), Illinois (1967), and South Carolina
(1971). In a press release dated May 28, 1963, the Atomic Energy Commission
stated that the Commission's burial grounds would no longer be available to
licensees for waste material shipped on or after August 12, 1963. It further
stated that "AEC's withdrawal is in line with its policy to foster industrial
participation in the atomic energy program." Some of these burial grounds
were presumably established because it was felt they might serve to attract
other nuclear related operations to the area but this has apparently not been
the case.
What is a Low Level Site?
"Low level radioactive waste" has never been formally defined. It is
generally considered to be any waste other than high level wastes which are
defined in Appendix F of 10 CFR Part 50 of the Commission's regulations.
Therefore, low level wastes cover such things as minimally contaminated
papers, metal, boxes, etc. and even "suspect" waste, but in fact can also
include megacuries of radioactive material such as 3H> 60Co, 90gr and others.
Nevertheless, it is generally known that the six sites noted previously
constitute the commercial low level burial grounds in the United States.
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The sites vary in licensed acreage from 20 to 260 acres and are located
in both dry areas such as Nevada and Washington to wet areas for the other
four sites. Table 1 shows the volume of wastes and the amount of activity
that have been buried at each since their initial operation. The physical
operations conducted at a burial site are rather straightforward and do
not involve complex machinery, mechanical control systems or highly skilled
staff. Operations carried out are similar to a sanitary land fill operation
with a greater degree of care exerted during trench excavation, trench
filling and trench capping. Of course, certain specialized operating proced-
ures must be observed during these operations. However, operations in
general are rather modest when compared with the operation of a fuel fabrica-
tion plant, uranium mill or a nuclear power plant.
How Are the Sites Licensed and Regulated?
A regulatory requirement from the very beginning was that the land for a
disposal site be owned by a State or the Federal Government. All are located
on State-owned land except the Washington site which is on Federally-owned
land. Regulatory jurisdiction for each of the sites was determined by the
type of materials to be buried or handled at the site. If the site is
located in an Agreement State* the regulatory authority is exercised by
the State unless the quantity of special nuclear material to be handled
prior to burial exceeds those quantities specified in 10 CFR 150.11. If
quantities exceed that, a license for handling special nuclear material
has to be obtained from the NRC.
* A State which has entered into an agreement with the AEC (now NRC)
pursuant to Section 274 of the Atomic Energy Act of 1954, as amended,
whereby the Commission relinquishes and the State assumes, certain
regulatory authority over the use of source material, byproduct material,
and small quantities of special nuclear material.
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In non-Agreement States, burial grounds were and continue to be,
regulated by AEG (NRC). It should be noted, however, that the burial
grounds also receive non-agreement material and accelerator produced
radioactive material which is not subject to control by NRC. Such
material is regulated by the Agreement States and the non-Agreement State
of Illinois.
The procedures for licensing sites included a submission of information
by the applicant on the geological and hydrological characteristics of the
site and the usual information related to qualifications of the applicant,
operating procedures, and the applicant's radiation safety program. Although
the information on geology and hydrology obtained on the sites at the time
they were initially licensed would be considered modest by today's standards,
the information obtained was evaluated by appropriate technical specialists
in the State and Federal agencies. It should be noted that the National
Environmental Policy Act (NEPA) process for Federal agencies was not in
effect at the time any sites were originally licensed by AEC.
During the early 1960's it was felt that the disposal of waste in the
ground took care of the matter. These sentiments were no doubt expressed
with a degree of certainty that they did not deserve. Notwithstanding this,
the States involved in all six sites, through various mechanisms, established
environmental monitoring programs in the vicinity of the site. Environ-
mental monitoring programs by nature are designed to provide verification
that the "system" was performing as expected or if it was not performing as
expected one could take such action as necessary to correct it before
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the problem became serious. I have noted earlier in this paper that one of
the initiating events for the publicity surrounding this subject was a 1974
environmental monitoring report issued by Kentucky. I believe it worth
noting that the Kentucky environmental monitoring program did precisely what
it was supposed to do, that is, detect a possible problem at a low threshold.
Another aspect of the regulation of these sites is that a,ll operations are
subject to periodic inspection by the appropriate regulatory agency.
In each case, the States (as site owner or lessor in the case of
Washington) established disposal fees which were intended to provide a per-
petual care fund when the sites were closed. The funds were not expected to be
used for corrective action since major problems in site performance were not
expected. It is obvious that the funds accumulated to date are insufficient
for major corrective action and may be even insufficient for long term main-
tenance. There are no uniform national standards that can be used at present
regarding the establishment of such funds. At the May 1973 meeting of the
National Conference of Radiation Control Program Directors, a task force on
radioactive waste management presented a report and certain recommendations.
Among other things, it recommended a national and/or regional coordinated
approach to the question of establishing disposal sites as may be required,
recognizing that proliferation of such sites may not be in the public interest.
It also noted that information is required in the areas of criteria and
requirements regarding perpetual care of land burial grounds and the legal and
financial implications of the State's perpetual care responsibilities.
Some States have expressed the parochial view that they might establish
a burial ground just to be used for wastes generated within their own borders.
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If this view prevails there could be more than 50 burial grounds in the United
States. I think most of us here today would agree that would be highly
undesirable. In this regard the Supreme Court ruled in June 1978 that a New
Jersey law prohibiting importation of wastes from neighboring States (this
case involved non-radiological wastes) violates the Constitution because a
State is not free to impede interstate commerce simply because that commerce
is in valueless wastes. This should be considered by those who might suggest
a burial ground just to handle the wastes generated within its borders. One
should also question the economic viability of such an operation, at least in
many States. The concept of regional sites to address the problem of adequate
distribution of capacity should be pursued vigorously. This obviously will
require a high degree of cooperation among several States if it is to be
effective.
Thoughts for the Future
A fair question to ask is "Is low level waste burial a viable option?"
I leave this to those closer to the subject technically, but would observe
that there are still three low level sites operating today. Certainly the
technical and regulatory climate is different today than it was in 1962 when
the first site was licensed. The most notable factors are NEPA requirements,
more detailed information required to evaluate site suitability, and the high
level of public interest and involvement. I would judge that the use of solid
land burial as a disposal method will continue although with various modifica-
tions such as new techniques, waste classification, and volume reduction. I
believe that volume reduction is an area where the industry, both waste gener-
ators and waste collection and disposal firms, should take the initiative.
One never knows when an action taken in some sector of the nuclear industry
affects another. For example, the NRC recently received a communication from
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the State of South Carolina indicating that rather large volumes (as much as
50,000 ft per month) of waste in the form of spent fuel racks are being
received for burial at the Barnwell, South Carolina facility due to reracking
to accommodate the increasing quantities of spent fuel stored at nuclear power
plants. It would seem that some effort at the waste generating sites to
reduce the volume of these wastes would be desirable.
What does the future hold for Federal/State roles in this area of
regulation of low level waste disposal? Although the NRC's task force report
on this subject (NRC77) recommended NRC reassert regulatory authority over
the burial grounds, the Commission did not adopt this as formal policy.
It felt there were a number of unresolved questions on the matter and there
was no compelling need to make a final decision at that time (December
1977) since the States were adequately protecting the public health and
safety. The Commission also believed it was more urgent to proceed with other
elements of the low level waste program including development of regulations,
standards, guides and a study of alternatives to shallow land burial. The
report of the task force for the review of nuclear waste management (DOE, 78a)
commonly referred to as the "Deutch Report" noted various possibilities for
addressing the question of regulatory control of the burial sites. At the
July 1978 meeting of the National Governor's Association (NGA) they stated
that "The Governors believe that long-term program plans for low-level radio-
active waste which continue to permit private operation and 'agreement-state'
regulation of low-level waste burial grounds on a cooperative basis with
Federal authorities, wherever this is both preferred and practicable, should
be finalized as expeditiously as possible." The subject was also considered
by the Interagency Review Group established by President Carter to develop
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recommendations for the management of nuclear waste (DOE, 78). The IRG
recommended the concept of State "consultation and concurrence" for high-
level waste facility siting plans and that States have the option to retain
management control of existing commercial low level waste sites or to transfer
such control to the Federal Government. They further recommended that the
Department of Energy assume responsibility for developing and coordinating the
national plan for low level waste. All of this leads to the conclusion that
the concern over low level waste burial grounds is attributable to the insti-
tutional questions to an extent at least as great, if not greater than the
technical questions. I previously mentioned the regulatory jurisdiction from
the early days of this program. I think it is well recognized at the Federal
level that the States will have a significant role to play in the regulation
of low level waste burial. I am equally certain the States realize that the
Federal Government has a significant role to play no matter where the jurisdic-
tion lies. It should be noted that Section 14(b) of the NRC Authorization Act
for FY 1979 directs the NRC to prepare a report on means for improving opportu-
nities for State participation for siting, licensing and developing nuclear
waste storage or disposal facilities. The NRC Working Paper "Means For
Improving State Participation in the Federal Nuclear Waste Management Programs"
(NUREG-0513) was published on December 20, 1978 (NRC, 78 ). Although it is
heavily oriented to the high level waste program, it notes that a key item of
interest to the States in the low level waste area is the question of Federal
control of the sites. However, other issues such as waste classification,
chemical toxicity of wastes and volume reduction are also of high interest
to the States.
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It certainly seems that a prime need is to get a more formalized
regulatory framework in place so that this sector of the industry can be
regulated on a more consistent, uniform and effective basis. It is vitally
important that the regulatory framework be one which can be implemented by
either the States or the Federal Government. This would help assure that any
regulatory or technical uncertainties in low-level waste management are
minimized.
A closing word about the role of government. From time to time we have
heard various people proclaim there should be less regulation. In recent
months there have been a number of steps taken to deregulate the airlines
and there have been some actions taken to reduce the extent of OSHA regula-
tions.
Heclo (1977) in his book "A Government of Strangers" stated:
"Americans have long expressed impatience with red tape and
Washington bureaucrats, but few of the heavy demands they
make on the federal government can be satisfied without some
form of organized bureaucratic activity. Public opinion
polls show, for example, that the overwhelming majority of
Americans agree that the federal government should control
inflation; avoid depression; assure international peace;
regulate (but not run) private business; and see to it that
the poor are taken care of, the hungry fed, and every person
assured a minimum standard of living. But a comparably large
majority also agree that the federal government is so big and
bureaucratic that it should return more taxes to subnational
governments and count mainly on the states to decide what
programs should be started and continued."
The issue of low level radioactive waste disposal might be included in
this list.
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REFERENCES
1. DOE/TID-28817 (Draft), 1978, "Report to the President by
the Interagency Review Group on Nuclear Waste Management."
2. DOE/ER-0004/D, 1978a, "Report of Task Force for Review of
Nuclear Waste Management."
3. Heclo, H., 1977, "A Government of Strangers" p. 113
(Washington, D. C., The Brookings Institution.)
4. NGA, 1978, Subcommittee on Nuclear Power Report to Natural
Resources and Environmental Management Committee, p. 2.
5. NRC, NUREG-0217, 1977, p. 3, "Task Force Report on Review
of the Federal/State Program for Regulation of Commercial
Low-Level Radioactive Waste Burial Grounds."
6. NRC, NUREG-0513, 1978, p. 15, "Means For Improving State
Participation in the Federal Nuclear Waste Management Programs."
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Table 1. Volume of wastes and amount of activity buried at low level sites;
cumulative through 1977.
Location
Totals
Year Licensed Volume buried X10 ft
BPM (Ci)*
SM (Ibs)
SNM (kg)
17.91
4,307,108
3,042,318
1480.80
Pu (kgm)
Beatty, NV
Maxey Flats, KY
Hanford, WA
Sheffield, IL
West Valley, NY
Barnwell, SC
1962
1962
1965
1967
1963
1971
2.16
4.77
.58
3.0
2.36
5.04
156,059
2,400,691
434,358
58,000
700,000
558,000
142,976
532,988
14,354
594,000
1,026,000
732,000
191.46
431.8
93.74
54.8
56.0
653.0
14.3**
63.76
30.63***
13.4
4.0
0
122.09
* BPM, byproduct material; SM, source material; SNM, special nuclear material, including plutonium.
** Cumulative through 1976.
*** 1970 through 1977 only.
ro
co
o
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REGULATORY ASPECTS OF LOW LEVEL WASTE DISPOSAL
Heyward G. Shealy
S. C. Department of Health and Environmental Control
Columbia, S. C.
Abstract
Regulating a low-level waste disposal site reveals many interesting
facets of how this nation's low-level waste is being managed. Incidents
and occurrences that happen to radioactive waste prior to arriving
at a burial site and after it is received at the burial site will be
described. Regulating the site involves numerous disciplines of
Engineering, Chemistry, Hydrology, Geology, Health Physics, and others.
Regulatory involvement of these elements will be reviewed. Classification
of low-level waste as presently being implemented at the Chem-Nuclear,
Barnwell, South Carolina, site will be discussed.
Discussion
Before commencing the substance of my presentation, I would like
to briefly discuss the State's role in the regulation of ionizing
radiation. The State of South Carolina became involved in the regula-
tion of ionizing radiation in 1967 when the South Carolina Legislature
enacted the Atomic Energy and Radiation Control Act. On September 15,
1969, the State became an "Agreement State" pursuant to Section 274
of the Atomic Energy Act of 1954 as amended, thereby assuming regulatory
responsibilities for certain radioactive materials. Such responsibility
includes the licensing of low-level waste burial facilities; and, on
April 13, 1971, one such license was issued to Chem-Nuclear Systems, Inc.,
authorizing the use of approximately 250 acres of property in Barnwell
County, near the Savannah River Plant property, as a low-level radioactive
waste burial facility. This information provides you with some under-
standing of the regulatory atmosphere within our State.
The Barnwell, S. C., Chem-Nuclear site is one of the three remaining
low-level commercial burial sites in the U. S. that is presently receiving
commercial low-level waste for land disposal. The other two sites are
located on the west coast. The volume of low-level radioactive waste
disposed of at the Barnwell site has increased significantly over the
past several years. For example, in 1975 the total volume of waste
buried was 638,137 cu. ft.; whereas, in 1978, this volume had increased
to 2,173,955 cu. ft. Much of this volume increase was due to the Maxey
Flats and West Valley shutdown. The recent termination of the Sheffield,
Illinois operation has further complicated the disposal of commercial
low-level radioactive waste.
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As a result of closing the Sheffield site, the State of South
Carolina deemed it necessary to impose a monthly restriction on the
volume of waste that could be disposed of at the Barnwell site. This
volume restriction was imposed on the operator of the site because of
concerns over using available burial space much faster than was originally
planned. Approximately one-half of the waste buried at the Barnwell
site during 1978 originated outside the region of the U. S. that the site
was intended to serve. The volume restriction has not been the first
such action imposed on Chem-Nuclear's operation. The State of S. C.
initially restricted burial of plutonium or any other transuranic
elements. During 1976, Chem-Nuclear was not allowed to ship or receive
additional bulk shipments of contaminated liquids from nuclear power
reactor sites. Solidification at the point of origin was required rather
than solidify the liquids at the burial site for disposal. Also, we
have recently imposed a restriction on the site operator not to receive
or bury radioactive waste containing more than 1% oil by volume.
The Chem-Nuclear site property is owned by the State of South
Carolina, whereupon, the State leases the site for Chem-Nuclear to
operate. Chem-Nuclear's operation is licensed by both the State of South
Carolina and the U. S. Nuclear Regulatory Commission. Special nuclear
material disposal authorization is licensed by NRC in quantities greater
than is authorized under State jurisdiction. The joint licensing of the
site operation has proven to be a satisfactory arrangement. Close
communication between the NRC Office of State Programs and Region II
Compliance Office has demonstrated a close Federal-State partership in
this respect.
During 1978, 12 incidents occured with waste being shipped to the
Chem-Nuclear site that required investigation. These include contamina-
tion of personnel, breach of package integrity, liquid waste shipments
contaminated vehicles, and freight, and vehicle accidents. A total of 35
man-days was expended due to investigation and follow-up action for pro-
tection of the public health and safety. Many of these investigations
could have been avoided had the waste been prepared and packaged properly
for shipment and disposal.
Routine monitoring (spot checks) of packages received at the
Chem-Nuclear site indicates to us that not enough attention is being
given to the management of low-level waste at the point of origin.
During a routine inspection of a shipment of institutional waste from
one of our major educational institutions, the following was found.
The shipment of 74 - 55 gallon steel drums contained glass jugs, coke
bottles, wine bottles, etc. filled with liquid chemicals (organics)
containing tracer amounts of radioactivity. In addition to violating
Chem-Nuclear's license, the shipper was also in violation of DOT
regulations.
Solidified waste that arrives at the burial site is also routinely
inspected by personnel from the Bureau of Radiological Health. It is
evident from our observations that quality assurance programs are
inadequate with respect to preparation of liquid waste for shipment and
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ultimate disposal. Solidification medias and the management of organics
has been of major concern at the Chem-Nuclear burial site. Standards in
these areas and areas I have not touched on are needed. We need these
standards now.
An important aspect of perpetual maintenance of a low-level burial
facility is a program of daily site upkeep and continuous environmental
monitoring. The former is important to prevent the occurrence of problem
situations and the latter to detect potential problems in their infancy.
Chem-Nuclear Systems, Inc. presently undertakes all site maintenance
being reviewed periodically by Health Physicists from the Bureau of
Radiological Health. Both the State and Chem-Nuclear conduct extensive
on and off site environmental monitoring programs.
In regulating the burial site operation, the staff of our Bureau of
Radiological Health reviews and approves changes in site operation,
inspects each trench prior to use for burial, physically inspects the
site routinely for erosion control and trench conditions, and audits the
entire burial operation. Based on our experience at the Chem-Nuclear
site, we believe that properly engineered trenches and management of
surface water is an important factor in the management of a low-level
radioactive waste burial facility.
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BURIAL OF SMALL QUANTITIES
OF RADIONUCLIDES WITHOUT PRIOR
NRC APPROVAL
John W. N. Hickey
U.S. Nuclear Regulatory Commission
Washington, D.C. 20555
Abstract
The U.S. Nuclear Regulatory Commission has proposed to delete a regulation
(10 CFR 20.304) which allows licensees to bury specified small quantities of
radionuclides anywhere without notification or approval of NRC, subject to cer-
tain conditions. In developing this proposed deletion, the NRC staff contacted
licensees, state officials, inspectors, and others to obtain information on
disposal practices. The Commission and its staff tentatively concluded that
burials pursuant to 10 CFR § 20.304 should stop, and that the impact of the
proposed deletion would be small. If review of public comments does not
change this evaluation, 10 CFR § 20.304 will be deleted, and burials will have
to be specifically approved in advance by NRC.
Background
U.S. Nuclear Regulatory Commission (NRC or Commission) regulations currently
allow licensees to bury small quantities of radionuclides anywhere without notifi-
cation or specific approval of NRC. Specifically, 10 CFR § 20.304 provides that
no licensee shall dispose of licensed material by burial in soil unless:
(a) The total quantity of material buried at any one location and time
does not exceed 1,000 times the amount specified in Appendix C of
10 CFR Part 20,
(b) Burial is at least four feet deep,
(c) Burials are at least six feet apart, and
(d) Not more than 12 burials are made per year. Records of the burials
must be maintained by the licensee as provided in § 20.401.
Burials which do not comply with § 20.304 must receive specific Commission
approval as provided by 10 CFR § 20.302.
Requests for reevaluation of 10 CFR § 20.304
Section 20.304 has not been amended since its adoption in 1957. Over the
last few years, NRC has received several requests for reevaluation of this regu-
lation. For example, the National Conference of Radiation Control Program
Directors and Officials from the majority of the States have requested such a
reevaluation.
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The reason given for the need to reevaluate § 20.304 is that people
could be overexposed to radiation by inadvertently disturbing burials, since
licenses may be terminated, burial records lost, land sold, etc.
The NRC agreed that § 20.304 should be reviewed, and the radiological risks
and potential impacts of revising § 20.304 were assessed as described below.
Radiological risk associated with § 20.304
Section 20.304 allows burial of quantities up to 1,000 times the amounts
listed in Appendix C of 10 CFR Part 20. The Appendix C values are approximately
the lesser of two values: (1) the amount that a standard man would inhale when
exposed for one year to the highest concentration allowed for unrestricted
areas, or (2) the amount which produces 1 milliroentgen per hour of gamma
radiation exposure at 10 centimeters.
There is a remote possibility that radionuclides buried pursuant to § 20.304
could deliver large doses to individuals disturbing a burial site. However, an
individual would have to dig up the buried material, and remain near the site long
enough to inhale a large fraction of the material or be exposed to a large direct
radiation dose.
The NRC staff concluded that the risk associated with burials pursuant to
§ 20.304 is small. However, it was recognized that amendments to § 20.304 could
improve public health protection by improving data and controls over burials of
even small quantities of radionuclides. For the NRC staff to decide whether such
amendments were justified, it was necessary to assess the potential impacts on
licensees.
Potential impacts of amendments to § 20.304
Amendment or deletion of § 20.304 would not prohibit burials. Rather, licensees
wishing to continue burials previously conducted pursuant to § 20.304 would have
to apply to NRC for approval in advance, or send waste to a commercial burial
ground. The staff estimated that obtaining prior approval would require a few
man-days extra effort on the part of each licensee.
Because licensees are not required to inform NRC of burials made pursuant to
§ 20.304, it is difficult for the staff to estimate how many licensees might be
affected by an amendment to the regulation. However, the staff made a rough esti-
mate by contacting NRC inspectors and State officials. This survey covered 25
states, over 5,000 NRC licensees, and over 5,000 licensees under the NRC Agreement
States. (There are over 16,000 NRC and Agreement State licensees in the U.S.)
A summary of the information collected is shown in Table I. It should be
emphasized that these are rough estimates; exact data could only be obtained by
the time-consuming process of contacting thousands of licensees for review of
records. Also, it should be noted that some Agreement States have already
prohibited burials which NRC would permit pursuant to § 20.304.
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Based on this survey, the NRC staff estimated that less than 100 licensees
are using § 20.304 to perform burials. Therefore, it was concluded that the
impact of amendments to § 20.304 would be slight. The staff then considered the
alternatives described below.
TABLE I
NRC AND AGREEMENT STATE LICENSEES PERFORMING § 20.304 BURIALS
Approx. Number
States Surveyed Total Material Licensees Performing Burials
NRC Region I (11 states) 2300 (non-agreement) less than 5%
NRC Region III (8 states) 2900 (non-agreement) less than 25
California 1700 5
Texas 1310 10-15
New York 1000 none
Florida 700 none
South Carolina 151 none
Oregon 170 none
Arkansas 248 1
New Hampshire 59 2
Kansas 207 3
Notes:
I. There are 25 Agreement States regulating over 9000 materials
licensees. NRC regulates about 7000 materials licensees.
2. Licensee totals for individual states are Agreement State
licensees only.
3. There have been no unlicensed burials in New York, Florida,
South Carolina, and Oregon for several years.
4. Source: NRC inspectors and Agreement State officials.
Alternatives for action on § 20.304
No Action - The "no action" alternative was rejected because it was not
responsive to the expressed concerns of the public.
Notification - The staff considered amendments to § 20.304 which would
merely require notification of NRC and the Agreement States after the burials
were performed. This was rejected because burials could still be conducted
without prior regulatory review for suitability of location, proper marking, etc.
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Additional restrictions - The staff considered additional restrictions such
as (1) confinement of burials to restricted areas, (2) removal of buried materials
prior to termination of licenses, or (3) additional limitations on quantities and
types of radionuclides to be buried. The staff concluded that generic decisions
on these issues were less desirable than case-by-case decisions.
Deletion - The staff chose deletion of § 20.304 as the best alternative
because it is responsive to the concerns of the public and would have little
impact on licensees.
Conclusion
The NRC staff recommended deletion of § 20.304 to the Commission, and the
deletion has been approved as a proposed amendment and published for public comment
in the Federal Register (43 FR 56677, December 4, 1978). Copies of the Federal
Register notice were mailed to all NRC licensees and Agreement State officials.
The public comment period ended on February 2, 1979. However, I would encourage
attendees who have additional comments to submit them within the next few weeks,
and I will see that they are considered.
Although our review of the public comments is not complete, we have already
identified the following issues which will have to be addressed:
1. The public comments contain data identifying licensees using § 20.304
and assessing the potential impact of deletion of § 20.304. This data
will have to be reviewed so that the impact analysis of the proposed
regulation can be updated and improved.
2. If § 20.304 is deleted, some licensees using § 20.304 may apply to NRC
for approval to continue burials. The staff will need to provide
guidance to licensees on how these applications will be handled.
3. If NRC deletes § 20.304, appropriate actions on the part of the Agree-
ment States will have to be determined.
4. Some licenses contain conditions which allow burials in compliance
with § 20.304; that is, specific NRC approval has been obtained for
the burials. The impact of deletion of § 20.304 on these licenses,
if any, needs to be determined.
After our review of the public comments is complete, we will make a recommenda-
tion to the Commission for final action on § 20.304. This process usually takes
several months. I should point out that none of the alternatives described pre-
viously have been ruled out. After public comments have been considered, any
alternative could be chosen.
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WASTE MANAGEMENT PRACTICES IN DECOMMISSIONING NUCLEAR FACILITIES*
H. W. Dickson
Health and Safety Research Division
Oak Ridge National Laboratory
Oak Ridge, Tennessee 37830
ABSTRACT
Several thousand sites exist in the United States where nuclear activ-
ities have been conducted over the past 30 to 40 years. Questions regarding
potential public health hazards due to residual radioactivity and radiation
fields at abandoned and inactive sites have prompted careful ongoing review
of these sites by federal agencies including the Department of Energy (DOE)
and the Nuclear Regulatory Commission (NRC). In some instances, these
reviews are serving to point out poor low-level waste management practices
of the past. Many of the sites in question lack adequate documentation
on the radiological conditions at the time of release for unrestricted use
or were released without appropriate restrictions. Recent investigations
have identified residual contamination and radiation levels on some
sites which exceed present-day standards and guidelines. The NRC, DOE,
and Environmental Protection Agency are all involved in developing
decontamination and decommissioning (D&D) procedures and guidelines
which will assure that nuclear facilities are decommissioned in a manner
that will be acceptable to the nuclear industry, various regulatory
agencies, other stakeholders, and the general public.
INTRODUCTION
Decontamination and decommissioning (D&D) of nuclear facilities is
playing an increasingly greater role in demonstrating the credibility of
Research sponsored by the Division of Operational and Environmental Compliance,
U.S. Department of Energy under contract W-7405-eng-26 with Union Carbide
Corporation.
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the nuclear industry. Several thousand sites exist in the United
States where nuclear activities have been conducted over the past 30 to
40 years. Many hundreds of these sites either have been abandoned or
have become inactive. A few have been totally decommissioned and re-
leased for unrestricted use. Recently, questions regarding potential
public health hazards due to residual radioactivity and radiation fields
at the abandoned and inactive sites have prompted careful review of
these sites by federal agencies including the Department of Energy (DOE)
and the Nuclear Regulatory Commission (NRC).
The DOE is responsible for the radioactivity in facilities it owns
or controls. Also, DOE has assumed the responsibility for abandoned or
inactive sites which were under the control of its predecessors, the
Manhattan Engineer District (MED), the Atomic Energy Commission (AEC),
and the Energy Research and Development Administration (ERDA). Of
immediate concern to DOE are 22 inactive uranium mill sites in the
western part of the United States. In addition, DOE has the respon-
sibility for reviewing more than one hundred excess MED and AEC sites
that played a role in the early development of the atomic energy pro-
gram.
Decommissioning criteria applied to NRC licensees prior to 1965
were not as stringent as present guidelines (Di78). Documentation of
the final radiological status of the properties involved may be in-
adequate. As a consequence, NRC has initiated a systematic program to
review all of its docket files of licenses terminated prior to 1965. In
addition, formal radiological surveys are being conducted at a few
selected sites with a known potential for residual contamination.
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While individual states have not as yet undertaken extensive review
programs, it is well-known that a few problem areas exist. For example, a
former nuclear facility in Tennessee and another in New York (DOE78), both
with significant levels of residual contamination, have become inactive
and essentially abandoned. The site in Tennessee was partially decon-
taminated at federal expense by the Oak Ridge National Laboratory (ORNL),
and at least a portion of the New York site may become a ward of the
state for cleanup.
OBSERVATIONS
Some nuclear sites have been decontaminated successfully and decommis-
sioned. Former AEC reactors, including the Piqua, Elk River, and BONUS
reactors, have been decommissioned. In one of the most ambitious decommis-
sioning actions ever undertaken, the Elk River Reactor in Minnesota was
completely dismantled and removed from the site. The NRC also has decom-
missioned a large number of formerly licensed sites with documentation
verifying that the sites met the established decommissioning criteria (Di78).
However, it has been pointed out that a number of sites have either been
abandoned or allowed to become inactive without adequate documentation
of radiological conditions at the time of release or without imposing
appropriate restrictions. As a consequence, recent investigations
(Ha77, Di77, Le78b, Pe78) have identified residual contamination and
radiation levels on these sites which exceed applicable standards
and guidelines (ANSI78, Di78).
Of the 22 inactive uranium mill sites, 16 are accessible to the
general public, seven have had no significant stabilization against erosion,
and 16 show evidence of off-site contamination (Go76). While some of
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these locations are remote, at least four of the sites are within a 16-
km radius of a population exceeding 10,000 persons (Go76). Many of
these sites have existed without active surveillance for 10 to 15 years.
From these observations, it is apparent that some early waste management
practices were less than adequate by today's standards even though waste
management was judged to be adequate at many of these sites under the then
existing standards.
In the case of the excess MED/AEC sites, properties with significant
levels of residual contamination and/or radiation levels have been identified
in or near major metropolitan areas (Di77, Le78a, Le78b). While the
total quantities of residual radioactive materials may be less than
those quantities at inactive uranium mills, there are more people who
potentially could be exposed. Most of these sites were contaminated in
the 1940's and 1950's and have been inactive for 10, 20, or even 30
years.
The NRC also has discovered previously licensed sites that have been
decommissioned without adequate verification of the radiological status (Pe78)
It is difficult to assess the possible extent of this problem. The NRC esti-
mates that as many as 8,000 source material and special nuclear material
licenses have been terminated over the years prior to 1965. Again, because
waste management practices in the past were not as thorough as present
practices, many of these sites could not be decommissioned using present-
day decommissioning criteria without substantial decontamination.
FACTORS CONTRIBUTING TO RESIDUAL CONTAMINATION PROBLEMS
It should be pointed out that many of the current radiological prob-
lem sites had their beginning long before the days of the AEC and other
regulatory authorities. For example, Lindsay and Company began operation
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in West Chicago in 1931 (Fr78). Another example is the former Vitro Rare
Metals Plant in Canonsburg, Pennsylvania, which was used as early as 1911
for the commercial extraction of radium from carnotite ore (Le78a).
Consequently, at least a portion of the present problems can be blamed
on a total lack of regulation. Although advisory groups had been formed
as early as 1929, no regulatory authority existed until the Atomic
Energy Act of 1946 when Congress established the AEC. The AEC and its
successors had no regulatory authority over naturally occurring, non-
ppc
source material (e.g., Ra) until Congress passed the Uranium Mill
Tailings Radiation Control Act of 1978 (PL 95-604) which defines tailings
as "byproduct material," thus, giving NRC authority over such materials.
Users or handlers of large quantities of radioactive materials
(e.g., uranium mills) have tended to use large scale industrial pro-
cessing techniques which have a few percent loss and/or spillage. As a
consequence, the facility involved became generally contaminated with
low-level radioactive waste. Much of the feed material contained "natural"
radioactivity which was considered rather innocuous. Efforts to prevent the
spread of materials which had been extracted recently from the earth received
little attention. Even to this day, several uranium mill tailings piles
have had no deliberate surface stabilization (Go76) to prevent erosion
or security measures to prevent casual access by the public.
In some cases, the large user would contract for waste disposal via
conventional industrial means. As a result, radioactive waste has been
placed in muncipal or industrial landfills or other such accessible
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locations. The examples of this are numerous and include Middlesex, New
Jersey, and Burrell Township, Pennsylvania. These specific examples are
covered in greater detail by Goldsmith (Go79).
In other cases, the large users possessed a property which was unused,
and perhaps unusable, for other purposes which became the collecting place for
nuclear waste. Although the site was not strictly considered a waste burial
site, radioactive material accumulated there over the years awaiting ultimate
.disposal. Specific cases are represented by the Kerr McGee site (Fr78)
(old Lindsay Light and Chemical Company) in West Chicago, Illinois,
Canonsburg, Pennsylvania, (Le78a), and the Haist property (Le78b) in
Tonawanda, New York, which was used by Union Carbide under a lease
arrangement with property owners and MED.
Many licensees who used small quantities of radioactivity took
advantage of the on-site burial provisions of 10 CFR 20. While this
provides expedient removal of radioactive waste from sight, the problem
of ultimate disposal was simply deferred to license termination. It is
uncertain as to whether many sites can be decommissioned and released
for unrestricted use when substantial quantities of radioactive materials
are known to be buried on the site, even if the material is below licensable
concentrations (e.g., ores containing by weight 0.05% or more of uranium).
The pressures of commercial competition and governmental regulation
caused the. termination of many nuclear activities. In the case of uranium
mills, antiquated equipment and a low profit margin caused by a depression
in the price of uranium were responsible for the premature closing of several
mills. Some firms with marginal operations tended to short-cut on waste
management procedures to maintain a favorable economic picture. Such
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was the case with the American Nuclear Corporation in Oak Ridge, Tennessee.
Government (AEC or state) inspections were too infrequent to detect
items of noncompliance on a timely basis. As a result, a facility could
experience significant degradation in general housekeeping in the period
between inspections, which in some cases might be as long as several
years.
In a few cases, sites have been virtually abandoned. One can
speculate that the reasons for this abandonment range from ignorance of
decommissioning requirements to the more likely case of financial insolven-
cy. Since it has not been regulatory practice (NRC78) to require decom-
missioning funding arrangements (e.g., posting of bond) in advance of
decommissioning for small users, the licensee frequently does not have
the financial resources to cope with the cleanup and decontamination
required to be able to obtain consent for unrestricted release.
There have been numerous cases where radioactive waste materials
have been misused. The removal of tailings and their subsequent use as
fill around homes, schools, and other buildings in Grand Junction,
Colorado is one noteworthy case. In fact, many of the inactive uranium
mill tailings sites are accessible to the general public (Go76). Conse-
quently, the tailings materials easily could be misused at these sites.
The misapplication of radioactive materials extends to other source
material as confirmed by a review of NRC records (Cr78). Another example
concerns the unauthorized removal of contaminated tools and equipment
from the commercial burial site of the Nuclear Engineering Company at
Beatty, Nevada.
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Along similar lines, radioactive material has been transferred to
clean sites without specific application, probably in ignorance of the
radiological hazards involved. Examples of this include the spread of
contamination from the Kerr McGee site in West Chicago to at least 75
other locations in in the Chicago area (Fr78) and the relocation of a
major portion of the radioactive residues from the Haist property (Le78b)
to the nearby Seaway Industrial Park in Tonawanda, New York (Le78c).
Numerous small areas of radioactive contamination can also be found in
residental areas of Canonsburg, Pennsylvania, presumably spread there
from the early operations at the Vitro site (Le78a).
Another problem has been the lack of a comprehensive, internally
consistent set of decommissioning criteria and numerical guidelines.
Many contamination limit proposals have been adopted for use at specific
sites, apparently with marginal scientific justification. The Grand
Junction Remedial Action Criteria (CFR76) were written specifically to
resolve the dilemma at Grand Junction but may have applicability to
other sites contaminated with radium. The Environmental Protection
Agency (EPA) is the federal agency responsible for providing federal
guidance on radiation exposure related to the release of contaminated
property. As an example, EPA is considering interim recommendations for
radiation levels at new structures located on Florida phosphate lands
(FR76). While these fragmentary guidelines are of value for specific
applications, a master set of decommissioning criteria with general
applicability does not exist. Surely it is not practical for the nuclear
industry to develop a new and different set of criteria for each D&D
action.
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CORRECTIVE ACTION
While waste management practices involved in the decommissioning of
nuclear facilities in the past have resulted in unacceptably high levels of
residual contamination at many sites, a number of steps have been taken
recently to correct this situation. Several federal agencies are actively
*•
pursuing programs to correct past D&D deficiencies and to provide improved
D&D processes in the future.
For many of the sites formerly utilized by MED and AEC, available
records before the recent resurveys were not adequate to identify the
radiological condition at the time government controls were relinquished
(Cr78). Records for some formerly licensed sites are similarly lacking
in pertinent radiological information (Cr78). Both DOE and NRC have
programs to determine the adequacy of documentation and to make new
surveys if warranted. The DOE program is known as the Formerly Utilized
Sites-Remedial Action Program.
In addition to the review of terminated licenses which is being
conducted by the NRC, the whole decommissioning policy of that agency is
being reevaluated (NRC78). The NRC has sponsored considerable research
to determine the technology, safety, and costs associated with decommis-
sioning reactors (Sm78) and fuel reprocessing plants (Sc77). The DOE
was instrumental in the passage of Public Law 95-604, Uranium Mill
Tailings Radiation Control Act of 1978. This law provides the legal
basis for remedial action at the inactive mill tailings sites and at the
former Vitro Rare Metals Plant in Canonsburg, Pennsylvania. The EPA
continues to work on the development of appropriate criteria and guidelines
(FR77, FR78); however, a comprehensive set of decommissioning criteria
is in an embryonic state.
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Interest has been shown by a number of technical societies such as the
American Nuclear Society and the Health Physics Society, especially with
respect to their standards committees. Other interest groups such as the
Atomic Industrial Forum (Ro78) and the American National Standards
Institute (ANSI78) continue to make contributions in the D&D field.
While all of the problems related to waste management in D&D activities
have not been solved, it is encouraging to see so much interest and effort.
One concern is that all of this effort is not well coordinated. An inter-
agency task force much like the one organized in Canada (AECB77) to
provide D&D criteria might be the answer to a more efficient production
of the much needed guidance in this country. For example, DOE could
take a leading role in this activity since it is encumbent upon DOE to
implement D&D at a large number of facilities including the excess
MED/AEC sites, inactive uranium mill sites, and 300 to 400 excess
contractor facilities. In all likelihood, Congress would have to act to
set up the machinery for such a broad scope effort. Other participants
in this undertaking should include, but not be limited to, NRC, EPA, and
state regulatory agencies.
SUMMARY
A number of factors have contributed to the marginal waste management
practices observed in decommissioning of nuclear facilities. Some of the
more pertinent factors include:
1. lack of regulation—particularly with respect to naturally
occurring radioactivity;
2. poor control measures on large scale industrial processes;
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3. radioactive waste disposal by conventional methods such as
dumps, landfills, and on-site burial;
4. misapplication of waste products containing radioactive
material;
5. short-cutting of waste management procedures to increase the
profit margin;
6. abandonment of sites;
7. lack of continuing surveillance over inactive sites; and
8. lack of a comprehensive set of D&D criteria.
As a consequence, recent investigations have revealed residual contamination
and radiation levels on some sites which exceed present-day standards and
guidelines. Efforts by major federal agencies including DOE, NRC, and EPA
are serving to correct these deficiencies. An interagency task force could
be the most expedient approach to arrive at the D&D guidance which is
urgently needed by the nuclear industry.
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REFERENCES
AECB77 Atomic Energy Control Board, 1977. "Criteria For Radioactive
Clean-up in Canada," Information Bulletin 77-2.
ANSI78 American National Standards Institute, 1978, Control of Radioactive
Surface Contamination on Materials. Equipment and Facilities to be Rel-
1 eased for Uncontrolled Use. ANSI-N13.12.
Cr78 Crow W. T., 1978, "Problems at Inactive or Abandoned Fuel Cycle
Facility Sites," Ninth Annual National Conference on Radiation Control,
HEW Publication (FDA) 78-8054, p. 271.
CFR76 Code of Federal Regulations, 1976, Title 10, Part 712, Grand
Junction Remedial Action Criteria.
Di77 Dickson H. W., Leggett R. W., Haywood F- F., Goldsmith W. A.,
Cottrell W. D., and Fox W. F., 1977, Radiological Survey of the Middlesex
Sampling Plant. Middlesex, New Jersey, DOE/EV-0005/1.
Di78 Dickson H. W., 1978, Standards and Guidelines Pertinent to the
Development of Decommissioning Criteria for Sites Contaminated with
Radioactive Material, ORNL/OEPA-4.
DOE78 U.S. Department of Energy, 1978, Western New York Nuclear Service
Center Study Final Report for Public Comment. TID 28905-1.
FR76 Federal Register, 1976, "Interim Recommendations for Radiation
Levels at New Structures on Florida Phosphate Lands," Vol. 41, p. 26066.
FR77 Federal Register 1977, "Transuranic Elements," November 3, 1977.
FR78 Federal Register 1978, "Criteria for Radioactive Waste,"
Vol. 43, pp. 53262-8.
Fr78 Frigerio N. A., Larson T. J., and Stowe R. S., 1978, Thorium
Residuals in West Chicago. Illinois. NUREG/CR-0413, ANL/ES-67.
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Go76 Goldsmith W. A., 1976, "Radiological Aspects of Inactive Uranium
Mill Sites: An Overview," Nucl. Saf. 17(6). 722.
Go79 Goldsmith W. A., Crawford D. J., Haywood F. F., and Leggett R. W.,
"Previous Management Practices for Naturally Occurring Radionuclides
Wastes: Current Radiological Status," Proceedings of Low-Level Radio-
active Waste Management Symposium of the Health Physics Society
Williamsburg, VA, February 11-15. 1979.
Ha77 Haywood F. F., Goldsmith W. A., Perdue P. T., Fox W. F., and
Shinpaugh W. H., 1977, Assessment of Radiological Impact of the Inactive
Uranium Mill Tailings Pile at Salt Lake City. Utah. ORNL/TM-5251.
Le78a Leggett R. W., Haywood F. F., Barton C. J., Cottrell W. D., Perdue
P. T., Ryan M. T., Burden J. E., Stone D. R., Hamilton R. E., Anderson
D. L., Doane R. W., Ellis B. S., Fox W. F., Johnson W. M., and
Shinpaugh W. H., 1978, Radiological Survey of the Former VITRO Rare Metals
Plant, Canonsburg. Pennsylvania. DOE/EV-0005/3.
Le78b Leggett R. W., Cottrell W. D., Dickson H. W., Golden C. A., Fox W. F.
and Anderson D. L., 1978, Radiological Survey of the Former Haist Property
DOE/EV-0005/4.
Le78c Leggett R. W., Cottrell W. D., and Dickson H. W., 1978, Radiological
Survey of the Seaway Industrial Park, Tonawanda, New York. DOE/EV-0005/6.
NRC78 U.S. Nuclear Regulatory Commission, 1978, Plan for Reevaluation of
NRC Policy on Decommissioning of Nuclear Facilities. NUREG-0436.
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Pe78 Perdue P. T., Leggett R. W., and Haywood F- F., 1978, "A Technique
for Evaluating Airborne Concentrations of Radon Isotopes," Proceedings
of the Third Natural Radiation in the Environment Symposium. Houston.
Texas. April 23-28. 1978.
Ro78 Roger W. A., Staton S. S., Frendberg R. L.t and Morton H. W., 1978,
De Minimus Concentrations of Radionuclides in Solid Waste, AIF/NESP-016.
Sc77 Schneider K. J. and Jenkins C. E., 1977, Technology, Safety and
Costs of Decommissioning a Reference Nuclear Fuel Reprocessing Plant,
NUREG-0278.
Sm78 Smith R. I., Konzek G. J., and Kennedy W. E., 1978, Technology.
Safety and Costs of Decommissioning a Reference Pressurized Water
Reactor Power Station. NUREG/CR-0130.
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Decommissioning Standards - the Radioactive Waste Impact
by
J. L. Russell and W. N. Crofford
Office of Radiation Programs
U.S. Environmental Protection Agency
Introductions and Conclusions
Several considerations are important in establishing standards
for decommissioning nuclear facilities, sites and materials. This
review includes discussions of some of these considerations and
attempts to evaluate their relative importance. Items covered include
the form of the standards, timing for decommissioning, occupational
radiation protection, costs and financial provisions, and low-level
radioactive waste.
Decommissioning appears more closely related to radiation
protection than to waste management, although it is often carried under
waste management programs or activities. Basically, decommissioning is
the removal of radioactive contamination from facilities, sites and
materials so that they can be returned to unrestricted use or other
actions designed to minimize radiation exposure of the public. It is
the removed material that is the waste and, as such, it must be managed
and disposed of in an environmentally safe manner. It is important to
make this distinction even though, for programmatic purposes,
decommissioning may be carried under waste management activities.
It was concluded that the waste disposal problem from
decommissioning activities is significant in that it may produce volumes
comparable to volumes produced during the total operating life of a
reactor. However, this volume does not appear to place an inordinate
demand on shallow land burial capacity. It appears that the greater
problems will be associated with occupational exposures and costs, both
of which are sensitive to the timing of decommissioning actions.
Other areas which are not addressed in this paper but which may
become important in considerations of decommissioning standards
include: differences between existing facilities and planned
facilities regarding decommissioning, concepts related to dedicated
sites and facilities, issues of the relative costs and benefits of
various methods such as removal/disposal or fixation in place, and the
period for which impact assessments must be conducted. All of these
topics must be examined during the development of decommissioning
standards but their possible influence on the standard itself is
currently not clear.
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REVIEW OF EXISTING STANDARDS. REGULATIONS AND CRITERIA
The most difficult problem in establishing radiation protection
standards for decommissioning is the choice of the rationale or basis
of the standard. Once the rationale is chosen, the approach to setting
the standard (or the form of the standard) is defined and the standard
effort can proceed in a straightforward fashion. For example, if the
existing FRC guidance is to provide the basis for decommissioning
standards, the principal task becomes the ALAP approach, the familiar
cost-effectiveness assessment, and the standard is established at a
level where further expenditures provide little additional health
protection. Other rationale are available, however, and must be
examined in light of identifying which rationale would potentially
provide the greatest assurance of maximizing public health and
environmental protection.
As a initial step in identifying rationale which may be suitable
for decommissioning, a thorough review is planned of existing
standards, regulations and criteria which deal with areas such as
exposure rate limits in uncontrolled areas, decontamination criteria
and standards, and others. The objective of the review is more than a
listing of applicable standards, however it is the rationale for the
standards which requires identification. This identification will
require a thorough search of the history of the applicable standards
and will include such items as:
the reason the standard was established,
the contribution and opinion of expert advisory groups to the
development of the standard,
the use or interpretation made of existing authoritative guidance,
such as that of the Federal Radiation Council, and
the input by the public and governmental agencies during the
public comment period, both written and oral.
This information can be useful in determining the basis for
decommissioning standards in that it will make available the logic
previously followed by those who addressed the decommissioning issue,
or one closely related. Too often the thinking performed in the past
is neglected because it is believed new or unique issues are being
confronted. While there is no great assurance that past thinking will
provide a sound rationale for the current effort, it is believed it
will offer sufficient insight to be helpful.
Previous efforts that have been chosen for examination thus far
include:
-10 CFR Part 20, Paragraph 105, U.S. NRC
-EPA's proposed Federal Radiation Protection Guidance on Dose
Limits to Persons Exposed to Transuranium Elements in the
Environments
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-the Surgeon General's Recommendations to the State of Colorado
for Radiation Protection for Individuals Living in Residence
Contaminated with Uranium Mill Tailings
-Regulatory Guide 1.86, Termination of Operating Licenses for
Nuclear Reactors, U.S. NRC.
-ANSI Standard N328, Control of Radioactive Surface Contamination
on Materials, Equipment and Facilities to be Released for
Uncontrolled Use
TIMING
At the end of its operating life, a nuclear power plant itself or
other nuclear facilities become a form of radioactive waste. As such,
these facilities and sites must be restored to some level of
radioactive exposure that can be found acceptable (or not unacceptable)
for their unrestricted use in the future. An important part of the
decommissioning qustion is the time at which facilities and sites must
meet dose level requirements. This is especially pertinent to the
commercial reactors because at the time of reactor retirement radiation
levels will be quite high, primarily due to activitation products.
Since most of the activation products have relatively short lifetimes,
less than 5 years half-lives, there are potentially significant
benefits to be gained by delaying decommissioning until there is
appreciable decay of the activation products. However, delay of
decommissioning is undesireable from the view of both the possibility
of institutional failures and the responsibilities of the generation
which received the benefits to reduce the impact (or cost) to future
generations. Thus, the challenge is to select the timing for
decommissioning to optimize the benefits to be obtained through
radioactive decay and to minimize the remoteness of the generation
conducting the decommissioning from the generation receiving the
benefits.
An important consideration here is the ultimate use to which the
site or facility is to be put. Obviously, a contaminated laboratory
that is planned for future use must be decommissioned (or
decontaminated) to a level that will protect laboratory workers during
this future use. However, the case is not as clear-cut for a nuclear
power and centralized station power is unclear. A nuclear power plant
site represents a large investment for the owner-utility in terms of
transmission lines, switch yards and site improvements, especially
cooling water provisions. This results in a strong incentive for
continued use of the site for centralized power, nuclear or not. Thus,
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central power sites may eventually become mixtures of nuclear and
fossile fuel generation. In addition, utilities may choose to recycle
certain structures or other parts of a site, such as containment or
cooling water structures, if nuclear remains a viable option. These
considerations serve to greatly complicate the timing question for
decommissioning. However, timing requirements must be established to
assure public health and environmental protection for future
generations. Thus, such uncertainties should be considered to the
extent possible at this time with the concept that future activities
may require a reevaluation of the timing requirements.
As an initial estimate of what the timing for decommissioning
should be, it is suggested that EPA's proposed Radiation Protection
Criteria (EP) for the Disposal of Radioactive Waste be used. Criterion
No. 2 of this proposal states, "Controls which are based on
institutional functions should not be relied upon for longer than 100
years to assure continued isolation from the biosphere." While this
criterion addresses radioactive waste disposal rather than
decommissioning, these two issues share a common theme in the timing
area, that of the acceptance of continuing reliance on institutions for
health and environmental protection. Applied here the result is the
scenario:
Nuclear power plant licensing and construction 10 years
Nuclear power plant operations 40 years
Maximum decommissioning time 50 years
Total 100 years
This scenario provides a miximum 50 year period for decommissioning
while meeting the intent and spirit of the criterion. A 50 year period
following reactor shutdown also provides a most substantial reduction
in radiation exposure levels from activation products.
OCCUPATIONAL RADIATION PROTECTION
The major health related issue in determining timing requirements
for decommissioning is occupational radiation exposure. In most cases
and especially in the reactor situation, appreciable reductions in
occupational exposures incurred during decommissioning can be realized
by postponing decommissioning activities until short-lived
radionuclides have significantly decayed. This is based on the
assumption, or even possibly the fact, that most of the residual
radioactivity at a reactor will be due to activation products which are
predominantly short-lived materials. The situation at nuclear
facilites other than reactors is not as clear because the contaminating
materials will have longer lived characteristics.
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Decommissioning evaluations for reactors normally follow a
scenario which includes removal and offsite shipment of all spent fuel,
packaging and offsite shipment of the low-level radioactive waste, and
mothballing the reactor for a period of time. This time period is
largely controlled by the decay rate of cobalt-60 a significant
activation product in terms of production and external exposure with a
half life about 5 years. Longer lived activation products are also
present, such as Ni-59 and Nb-9^ but not in the same magnitude as
cobalt and other short-lived radionuclides. Since cobalt-60 is a
predominant radiation source and has longest half life of the
shorter-lived materials, occupational exposure reductions to be gained
by delayng decommissioning can be directly related to the half life of
cobalt-60. For example, a 20 year delay would produce a reduction of 16
and a 50 year delay the maximum as discussed above results in a factor
of 1,000 reduction. Therefore, it appears highly that a maximum timing
limit of 50 years for reactors will provide an optimum point for
occupational exposures balanced with economic and institutional
considerations.
Facilities providing the fuel for and managing the waste from
reactors, along with most of the defense and research oriented
facilities pose an additional problem in decommissioning since the
contamination is a result of longer-lived materials. First, the front
end of the fuel cycle is dominated by uranium and its daughters
thorium-230 and radium-226 which lead to the radon-222 decay chain.
Because of the long-lived nature of these products, there is nothing to
be gained through a delay and decommissioning activities should be
conducted immediately upon plant closings. As for the tail end of the
fuel cycle, regardless of decisions concerning recycle, facilities most
likely will be highly contaminated with longer-lived materials and
little will be gained through delays in decommissioning such
facilities. A similar situation can be expected at most of the defense
related and research oriented facilities.
DECOMMISSIONING COSTS AND FINANCIAL PROVISIONS
Current estimates indicate decommissioning costs may be great and,
what is worse yet, the current estimates have not been substantiated by
actual proactice. Thus, no evidence exists to prove that the costs may
not be significantly greater than current estimates. There is no way
to avoid some costs since prudent public health policies dictate that
nuclear facilities and sites be returned to a condition of
insignificant radiation threat to the public. A logical policy
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regarding such costs would be that the beneficiaries of the activity be
responsible for the decommissioning of the activity-related sites and
facilities. In addition since the decommissioning costs appear great,
it seems logical that financial provisions should be made before the
activity is initiated, or at the very least during the early stages of
the activity, to assure the existence of sufficient funding when
decommissioning takes place.
The diverse nature of nuclear facilities will lead to numerous
different solutions to the funding problem. For instance, nuclear
power plants, as regulated utilities, have a relatively secure economic
future, especially when compared to such facilities as uranium mills.
The situation for reactors appears relatively simple, because the
electric utility that can afford to build nuclear power plants is
presumed to have a continuing role. If the utility could demonstrate
that the decommissioning cost was a small fraction of the power
production cost, say less than 5% to 10$, this cost could be included
in the operating costs and regulated by the controlling Public Utility
Commission (PUC). Perhaps PUCs could even drop the decommissioning cost
from the operating costs, if sufficient emphasis was placed on the
relatively small fraction the decommissioning cost would be and on a
commitment by the utility that decommissioning would occur and costs
would be borne by the utility. The major obstacle to this approach is
the lack of firm cost data for decommissioning. It should be
recognized that any funding requirements would require approval of the
regulating PUC.
The situatiuon is not as simple for other nuclear facilites,
however. First the question of financial responsibility and a
continuing role must be addressed since most of these facilities are
owned by companies in a competitive market with no oversight provided
for the common good, in contrast to utility regulations by PUCs. A
case in point here is the company which owns the NFS facility in West
Valley, New York. According to a recent study the company can
terminate its' lease for the site at the end of 1980 with the question
as to their financial responsibility unclear, but apparently with
little likelihood that the company will be liable. The cost for
decommissioning this site is estimated at 536 (DE) million, or much
greater than the initial cost of the plant, constructed during the
early and middle 60's. This illustrates the second point to be made in
decommissioning non-reactor sites and facilities - the cost of
decommissioning may be far greater than the initial capital
investment. In addition to the West Valley situation, the abandoned
uranium mill tailings sites in the western part of the U.S. further
amplify this financial problem.
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The only logical conclusion that can be drawn here is that
arrangements must be made before licensing such activities to assure
both responsibility for decommissioning and adequate funding.
LOW-LEVEL RADIOACTIVE WASTE FROM DECOMMISSIONING
Estimated volumes of low-level radioactive waste from a reference
PWR and from a generic fuel reprocessing plant were taken from the work
performed by Battelle. (SraXSc). The volume of waste from the PWR was
17,924 cubic meters, all of which would be suitable for disposal by
methods currently practiced. Volumes of waste from the reprocessing
plant were separated into TRU and non-TRU contaminated. The TRU-
contaminated waste estimate was U,600 cubic meters and was further
divided into high, intermediate, and low-level categories. The non-TRU
contaminated waste estimate was 3,100 cubic meters presumably all
suitable for disposal by current practice.
To gain insight into what impact this waste volume will have on
waste disposal capacity, a review of waste generated annually by
reactors is made. Phillips (Ph) summarized this data through 197*5 as:
Average High Low
(All in cubic meters)
BWR 1,000 to 2,000 1,780 178
PWR 200 to 500 810 10
The average represents an assessment and projection of what volumes
could be expected. It was also noted the volumes increased as the
reactors aged, at least for the first few years of operation.
A comparison of annual volumes versus decommissioning volumes
indicates that over its operating life, a reactor is expected to
generate anywhere from about the same volume to as much as four times
the volume that will be generated during decommissioning. This assumes
that volumes from decommissioning BWR's will be about the same as for
PWR's. Thus, while decommissioning will result in a significant
increase in demand for waste disposal capacity, it appears that such
demand will not be inordinately excessive. The non-TRU waste from
reprocessing plant decommissioning is a factor of four lower and
represents an insignificant additional burden. TRU waste from
reprocessing represents a much greater problem in that suitable
disposal methods do not currently exist.
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A second approach at assessing the waste volume problem in
decommissioning is to determine the trench volume that would be
required if the waste is disposed of by shallow land burial methods.
The estimated PWR waste volume would fill a trench of dimensions 9m
wide by 9m deep by 220m long or 15m wide by 9m deep by 130m long.
Given the spacing required between trenches, the decommissioning volume
thus represents a land use commitment of about one acre per reactor.
This would not appear to be an inordinate commitment of land,
especially when compared to fossil fuel power demands. Thus, it is
concluded that the low-level waste volume from decommissioning is not a
serious problem nor a limiting consideration.
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References
(Ph) J.W. Phillips and G.A. Gaul, "An Analysis of Low-Level Solid
Radioactive Wastes from LWR's. Through 1975," ORP-TAD-772-2,
Nov. 1977.
(Sm) R. T. Smith, et. al., "Technology, Safety and Costs of Decom-
missioning a Reference Pressurized Water Reactor Power Station"
NUREG/CR-0130, June, 1978
(So) K. J. Schneider and C.E. Jenkins, "Technology, Safety and Costs of
Decommissioning a Reference Nuclear Fuel Reprocessing Plant,"
NUREG-0278, Oct. 1977.
(De) U.S. Department of Energy, "Western New York Nuclear Service
Center Study - Final Report for Public, TID-28905-1, Nov. 1978
(EP) U.S. Environmental Protection Agency, "Environmental Protection
Criteria for Radioactive Waste," U3 F.R. 53262, Nov. 1978.
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MODELING AND ENVIRONMENTAL ASSESSMENT OF LAND DISPOSAL
METHODS FOR LOW-LEVEL RADIOACTIVE WASTES
G. Lewis Meyer,
Stephen T. Bard,
Cheng Y. Hung
and
James Neiheisel
Office of Radiation Programs
U.S. Environmental Protection Agency
Washington, D.C. 20460
ABSTRACT
The Environmental Protection Agency (EPA) is evaluating the potential
environmental impact and risk to man from disposing of low-level radioactive
waste (LLW) in the ground. A generic model is being developed to simulate
the interaction of a variety of LLW waste types and land disposal methods
and operations with the natural characteristics of the site and their
potential impact on the environment. Shallow land burial will be used as a
base case. This model will be an important tool in EPA's work to develop
environmental standards for LLW. This paper presents the general
objectives, information requirements, components, and pathways considered in
the environmental assessment model and some considerations in developing the
model.
INTRODUCTION
Since 1945, low-level radioactive wastes (LLW) have been disposed of
principally by shallow land burial and, until 1975, it was believed that
there would be "zero" activity released from the burial
sites. (1»2) Studies by the Environmental Protection Agency (EPA) and
Department of Energy (DOE) at several burial grounds have shown, however,
that radionuclides can escape from the burial trenches to the uncontrolled
environment. (3,4,5) in 1975, it was determined that, "the activity
detected in the environment around the site does not create a public health
hazard at this time".(3,5)
In 1979, we still can not realistically estimate what the environmental
impact of LLW disposal will be in 50, 100, or 1,000 years, if temporary
remedial actions at the burial site are halted. Nor can we advise others
whether present and proposed disposal methods are environmentally acceptable.
The Environmental Protection Agency has issued draft environmental
protection criteria for storage and disposal of all forms of radioactive
waste. (6) EPA is now preparing environmental radiation protection
standards for high-level radioactive waste (HLW) and spent fuel management
(7) and will prepare a radioactive waste standards rationale document by
mid-1979. (8)
Presuming the recommendations of the Interagency Review Group to the
an* annnovedL EPA will develop a proposed standard for LLW by the
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The environmental standard for LLW should comply with the environmental
protection criteria. For compliance, (1) the health risk from disposing of
LLW should not be unacceptable, (2) future generations should not be
subjected to risks greater than present generations, and (3) the health risk
should be further lowered as much as reasonable, taking into account
economic and social factors. If the environmental criteria are met, it
follows that the needs of NEPA will also be satisfied.
Benefit-cost analyses, estimated health risks, and comparison of
alternative methods will all be considered during the development of the LLW
standard. The focus of this paper, however, is on development of an
environmental assessment model to simulate the health risks from disposing
of LLW by a shallow land burial (SLB) method.
GENERAL APPROACH
EPA's Criteria for Radioactive Wastes (6) recommend that the
environmental standard should be based on predetermined models and should
examine the projected effectiveness of alternative methods of control.
Since a site specific model will not satisfy all these requirements, a
generic environmental assessment model is required to estimate the health
risk from LLW disposal methods to support development of an environmental
standard for disposal of LLW. The proposed generic model will be a dynamic
simulation model which will include release transport, pathways and dose and
health effect submodels. The release and transport models will consider a
range of geological, hydrological, meteorological, and waste properties and
characteristics which may exist at present and potentially acceptable future
sites. We are currently developing the model and collecting relevant data.
(Details of these procedures follow.)
A. Data Base
There is already an extensive data base on the transport of
radionuclides in the environment. The data required for our generic
environmental assessment model are summarized as follows:
1. Waste form and inventory
2 Hydrological and meterological data of the site
3. Hydrogeological and geochemical characteristics of the disposal
site and waste
4. Hydrological, demographic, and land-use distribution of site
5. Transport of non-volatile radionuclides by surface and ground water
and wind
6. Transport of gaseous radionuclides through trench cover and their
consequent transport by wind
7. Hydrogeological characteristics of trench cover
8. Macro-porosity of buried waste
9. Environmental impact assessment of leachate evaporator
10. Summary of available and current environmental study results
11. Summary of current engineering and operations practises for
disposal and their associated costs.
12. Potential biological pathway
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In some cases, information needed for these categories is available
from previous studies. Other data which characterize simple physical and
chemical processes can be obtained from simplified laboratory models or
experiments. However, some data can only be obtained from field
observations at existing burial sites. The data available will be reviewed
and where the information is not sufficient, supplemental data will be
obtained from ongoing and proposed field studies by DOE, NRC and EPA at a
number of existing shallow land burial sites. At the present time, it
appears that additional information is needed on: gaseous and surface
transport of radionuclides, trench cover characteristics, biological
pathways, waste forms, and current engineering practices and costs.
The parameters required for the generic environmental assessment model
will be obtained from the results of the statistical analysis of the site
specific and general data described above.
B. Description of Model
The environmental assessment model consists of four submodels: a
release model; a primary transport model; an environmental transport model;
and a dose and health effects model. Figure 1 shows the general flow of
radionuclide transport simulation.
CONTAINMENT AT SITE ["*
1
\JteLEASE SUBMODEL /
' "
1
ANNUAL INFILTRATION
ANALYSIS
PREPARATION MODEL
\ PRIMARY TRANSPORT /
\ SUBMODEL /
1
GROUNDWATER SU
RECEPTOR
1®
RFACE WATER
RECEPTOR
®
AIR L
RECEPTER
@
1
T
AND SURFACE
RECEPTOR
4
\ ENVIRON TRANSPORT /
\ SUBMODEL /
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The release model simulates the release of the radionuclides from the
waste by a driving force, due to either natural processes or human
activities. Water is considered to be the most important driving force to
release radionuclides from the waste. The predominant release modes are
expected to be leaching, erosion, and gas generation. The radionuclides
released are then available for transport to environmental receptors through
the primary transport model. The environmental receptors considered include
groundwater, surface water, air, and land surface. The environmental
transport model simulates the transport of radionuclides among the
receptors. The resultant accumulation of radionuclides in the environmental
receptors are then available for biological uptake by human beings through
food chains, drinking water, inhalation, and direct irradiation pathways.
The results of the pathway simulations then serve as input to the biological
pathway model. The main environmental assessment model then integrates the
total health effects for the total radionuclide uptake into humans through
all critical pathways, both for individuals and for population.
C. Model Development and Analysis
The analyses will be accomplished by a system model which includes a
main model and two independent preparation models. The main model is
designed to accept inputs from the direct interpretation of collected data
and from the simulation results obtained from the preparation models and to
simulate the health effects therefrom.
The two preparation models are designed to simulate respectively (1)
the synthesized hydrological conditions for a normalized driving force
(infiltration) analysis and (2) the total number of health effects per unit
intake of each radionuclide from each environmental receptor- The
relationship between the main program and the preparation models is also
shown in Figure 1.
The environmental assessment analysis will be conducted for SLB and for
alternative land disposal systems. The scenarios described in the following
section will be used as input to the established system model to simulate
the health effects resulting from a disposal system implemented at a
postulated generic site as a function of time. Use of these health effects
in benefit-cost analyses are beyond the scope of this paper and are,
therefore, not covered.
VIABLE POTENTIAL DISPOSAL SYSTEMS
A. Shallow Land Burial - The Base Case
Shallow land burial (SLB) is the method currently used for disposing of
most LLW in the United States and, it is believed, SLB will continue to be
used for disposal of LLW in the future. Since there are extensive data and
practical experience available with SLB, it will be used in EPA's
environmental assessment and standards development studies as a base case
from which to compare other disposal alternatives.
In its most basic form, disposal of LLW by SLB consists of excavation
of a large open-cut trench in the ground, placing wastes therein, and
covering the wastes with the earth excavated thereform. As such, SLB can be
treated and analyzed as a disposal system in which (1) the form of the
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waste, (2) the hydrogeology, meterology, and environment of the site, and
(3) special emplacement and protective engineering practices all perform and
interact together to provide varying degrees of radionuclide retention. The
degree of retention (i.e., performance) by the disposal system can be
improved by changing one or more components of the system. Ideally, the SLB
site, or system, would retain all the radionuclides until they have decayed
to innocuous levels.
At present, there appear to be three viable variations of the SLB
method: conventional SLB; improved SLB; and intermediate-depth SLB. Each
method has its own unique engineering practices which result in its own
associated costs; it may also, have different retention capabilities. A
brief description of the three SLB methods follows.
Conventional SLB; This is the most basic and commonly used disposal method
to date. Largely untreated wastes are placed in the trench and are covered
with about one meter of earth. It is the least expensive disposal method
and, in most cases, offers the least protection to the biosphere. In humid
climates, precipitation can penetrate the earthen trench cover, leach
radionuclides from the waste, and potentially release radionuclides to the
environment.
Improved SLB; In this method, one or more components of the disposal system
are modified to improve the performance (retention) of the system, or site.
For example, converting a soluble waste into a solid with low leachability
would retard the leaching of radionuclides from the wastes and greatly
reduce the potential release of radionuclides to the environment. In
another example the wastes could be covered with 2-3 meters of earth instead
of one meter to reduce infiltration of precipitation. This method,
naturally, is more expensive than conventional SLB but, in return, offers
better retention of radionuclides and consequent environmental protection.
Intermediate-Depth SLB; In this method, wastes would be placed in trenches
at least 10 meters below land surface and the trench would be backfilled to
slightly above ground level. Other improvements in waste form and in site
engineering such as suggested for improved SLB could be made, also.
However, the main intent of this method is to reduce infiltration of
precipitation and accessability of the wastes from human intrusion. This
method is, also, more expensive than the other two SLB methods but, in
return, it offers much more environmental protection. Shallow ground-water
tables such as found in the eastern U.S. and the availability of geologic
formations in which the open-cut trenches will remain open to depths of 18
meters or more without collapsing will limit the application this method.
B. Pathways and Scenarios for Shallow Land Disposal
In applying a generic approach to our environmental assessment
modelling to SLB, it appears that three basic scenarios will cover the most
sensitive parameters for radionuclide retention. These are: (1) an arid
zone site for which no particular regard is given to the permeability of the
disposal medium (because of a lack driving force or water); (2) a humid zone
site with low-permeability disposal medium (which would fill like a
"bathtub" in the event the trench cap leaked); and (3) a humid zone site
with a moderately permeable disposal medium (which would allow water
infiltratine through the trench cap to leak out the bottom like a "sieve").
. will be developed by altering the input parameters.
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It is anticipated that all of the components identified earlier as
making up the generic land "disposal system" will be present for SLB and all
of the pathways involved in radionuclide transport will be considered in
each of these three scenarios. The water pathway would certainly be less
important for conventional SLB and improved SLB in the arid climates,
whereas, all of the pathways would be important for these methods in the two
humid climate scenarios. In intermediate-depth SLB (10-l8m deep), the
ground-water pathways would be considered most important.
A summary table of the transport and biological pathways to man which
will be considered is given in Table 1.
TABLE 1. TRANSPORT AND BIOLOGICAL PATHWAYS TO BE CONSIDERED
TRANSPORT
PATHWAY
GROUNDWATER
TRANSPORT
LEACHATE
OVERFLOW
WASTE EROSION
GASSEOUS
EMISSION
RESUSPENSION
DESORPTION
DIRECT
EXPOSURE
HUMAN
INTRUSION
TRANSPORT
MEDIUM
GROUNDWATER
SURFACEWATER
SURFACEWATER
ATMOSPHERE
ATMOSPHERE
SURFACE AND/
OR GROUNDWATER
...
ATMOSPHERE
REVEISING
RECEPTORS
GROUNDWATER &
SURFACEWATER
SURFACEWATER
SURFACEWATER
AIR
AIR
SURFACE AND/
OR GROUNDWATER
...
AIR
BIOLOGICAL
PATHWAYS
DIRECT & INDIRECT
INGESTION
DIRECT & INDIRECT
INGESTION
DIRECT & INDIRECT
INGESTION
INHALATION
INHALATION
DIRECT & INDIRECT
INGESTION
DIRECT EXPOSURE
INHALATION
INDIVIDUAL
DOSE
ASSESSMENT
YES
YES
YES
YES
YES
YES
YES
YES
POPULATION
DOSE
ASSESSMENT
YES
YES
YES
YES
YES
YES
NO
NO
C. Viable Alternative Disposal Methods
As noted earlier, SLB is being analyzed as the base case for the
disposal of LLW. However, some wastes may not be suited for disposal by the
SLB method by virtue their specific activity, half-life, or chemical or
physical character, or cost of preparation. In such cases, alternative
disposal methods may be required or more cost-effective. An environmental
assessment of these alternatives will be required to simulate and compare
the alternatives.
On the basis of present information, much of which is of a preliminary
nature, the following seven disposal methods appear to be viable and
suitable for the disposal of one or more major types of LLW; deep geological
disposal; deep well injection; hydrofracturing; engineered surface storage;
ocean disposal-on-sea-floor; and ocean disposal-beneath-sea-floor. Brief
descriptions of these potential alternatives follow.
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Deep Geological Disposal; LLW can be emplaced in mined, solution, or
explosion cavaties in geologic formations more than 100 meters deep(8).
This method has already been studied in some detail for the disposal of
non-radioactive hazardous wastes, transuranic wastes and HLW. It is
suitable for solids or solidified liquids and for a wide range of specific
activities and half-lifes which require a higher degree of containment.
Deep-Well Injection; The disposal of liquid wastes by injection into deep
geological formations has been used widely for more than 40 years (more than
200 hazardous waste wells, 40,000 brine injection wells, and several LLW
wells) (9,10). Liquids are injected through specially constructed wells
under controlled pressure into permeable formations or saline aquifers at
considerable depths. This method appears to be a viable alternative to
certain liquid LLW presently being solidified for disposal in shallow land
burial sites.
Hydrofracture; Hydrofracturing is widely used in the petroleum industry to
stimulate the recovery of oil. This technique has been adapted for LLW
disposal at the Oak Ridge National Laboratory(ll). in a demonstration at
ORNL: (a) fractures were opened along planes essentially parallel to the
bedding in the Conasauga Shale by hydraulic pressure applied through an
injection well at selected zones; (b) pumping a propping agent of sand and a
gelling agent under pressure into the open fractures; and (c) then breaking
the gel by injection of a special compound to open fractures. A grout
slurry, containing cement, radioactive waste, and clay was then injected
into the open fractures and solidified as grout sheets parallel to bedding
in the impervious shale. Limitations of this method include separation of
the liquid radioactive waste from the grout during emplacement and
determining whether the grout is emplaced horizonally.
Engineered Surface Structures; Engineered surface structures, such as those
described by Morisawa et al (12)> Would be capable of containing as much
LLW as would be produced in the next 20 years in Japan. France already
uses another form of surface storage for disposal of certain LLW (13).
The Atomic Industrial Forum has also recently completed extensive studies on
the on-site storage of LLW in engineered structures at power stations (I1*)
and has found it to be a feasible but expensive management method. This
method would normally be limited to very short-lived radionuclides unless
removal was anticipated.
Ocean Disposal-On-Seafloor and Ocean Disposal-Beneath-Seafloor; LLW could
be deposited (a) directly on the ocean floor or (b) in open-cut trenches in
the sediment of the ocean floor and then covered. The seafloor sediments
have relatively high cation exchange capacities and provide radionuclide
retention capabilities as an additional barrier to the migration of
radionuclides. Seafloor disposal is an alternative disposal method in which
additional advantages may exist in the presence of a water barrier and more
remote isolation between the waste and the biosphere than exists in SLB.
It can be seen from an examination of the gross characteristics of the
various alternatives that the wastes are disposed of into deeper geologic
formations at depths greater than 100 meters except for the engineered
surface structure and ocean disposal methods.
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Those disposal methods in which wastes are disposed of at depths
greater than 100 meters include: deep geological disposal, deep well
injection, and hydrofracturing. It is believed that the environmental
assessment model developed for disposal of high-level wastes in a deep
geological formation would, to some extent, be useful and adaptable to
simulate the impact of LLW disposal by these alternatives. With the
exception of operational spillage, the near-surface pathways such as surface
water, atmosphere, soil, biological, and erosion, would be significantily
less important. The ground water and intruder pathways would assume primary
importance.
Environmental assessment of ocean disposal will require a considerably
different model.
CONSIDERATIONS IN IMPLEMENTING THE MODEL
Development of an environmental assessment model to support a LLW
standard will not be simple. There are a number of factors such as timing,
philosophy, national affairs, and the state of technology development which
will affect the depth and accuracy of analysis. These are discussed briefly
below.
Time Constraint; EPA must develop a proposed environmental standard for LLW
by the end of FY 1982. To be useful, the environmental assessment modeling
and analysis must be completed by early FY 1982.
Data Available: In some cases, the data required for simulating important
processes are not readily available (i.e., leaching of wastes, radionuclide
transport through unsaturated soil, etc.). In other cases, it will be
necessary to use best estimates of combined field and laboratory data.
Unusual Period of Analysis; The model will attempt to analyze
hydrogeological and radiological processes which cover 100's and 1,000's of
years. The scientific and engineering professions involved are not
experienced in making predictions over such long periods of time.
Complexity of the Model Limited; The model must be simplified to some
degree because (1) it must analyze both the fast and the slow responses of
very complex hydrogeological, ecological, and radiological processes and (2)
the costs of computer time would be prohibitive if a sophisticated model is
used to make an environmental analysis of a disposal method such as SLB when
many possible variations in parameters and consequent changes to performance
must be considered.
Caution Required in Using Output; It must be clearly realized that the
model may produce estimates which may be in error by orders of magnitude
because of the long periods of time involved, the many unknown parameters,
and uncertainty in the reliability of certain data. Therefore, output from
the model must be used with caution and good judgment.
Benefits and Necessity of Model; Although the environmental assessment
model will have limitations, it's use is both benefical and necessary. It
forces us to organize our thoughts to consider the total disposal system and
it helps us to identify our real data needs. Also, modeling is, to our
knowledge, the best tool to compare and estimate the impact of a large
complex system such as a SLB facility.
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In summary, it is believed that development of an environmental
assessment model is necessary; that the model must be relatively simple; and
that we must use caution and common sense to successfully apply it.
REFERENCES
1. U.S. Atomic Energy Agency, Draft Generic Environmental Statement Mixed
Oxide Fuel (GESMO), WASH-1337, (1974).
2. U.S. Atomic Energy Agency, Proposed Final Environmental Statement for
Liquid Metal Fast Breeder Reactor Program, WASH-1535, (1975).
3. Meyer, G.L., Preliminary Data on the Occurrence of Transuranium
Nuclides in the Environmental at the Radioactive Waste Burial Site
Maxey Flats, Kentucky, Environmental Protection Agency Report
USEPA-520/3-75-021, (1976).
4. Duguid, J.O., Status Report on Radioactivity Movement from Burial
Grounds in Melton and Bethel Valleys, Oak Ridge National Laboratory
Report, ORNL-5017, (1975).
5. Giardina, P.A., DeBonis, M.F., Eng, J., and Meyer, G.L., Summary Report
on Low-Level Radioactive Waste Burial Site, West Valley, New York
(1963-1975), Environmental Protection Agency Report, EPA-902/4-77-0010,
(1977).
6. U.S. Environmental Protection Agency, Federal Register Notice, Nov. 15,
PART IX, Criteria for Radioactive Wastes, (1978).
7. Arthur D. Little, Inc., Technical Support for Radiation Standards for
High-Level Radioactive Waste Management, Sub Task 2., Environmental
Protection Agency Draft Report, EPA 68-01-1470, (1977).
8. Deutch, J.M., Draft Report to the President by the Interagency Review
Group on Nuclear Waste Management, U.S. Department of Energy Report,
TID-28817, (1978).
9. Reeder, L.R., Cobbs, J.H., Field J.W., Finley, W.D., Vokurka, S.C., and
Rolfe,B.N., Review and Assessment of Deep-Well Injection of Hazardous
Waste, Vol I-III, Environmental Protection Agency Report,
EPA-600/2-77-029a, (1977).
10. Energy Research and Development Administration, (ERDA), Waste
Management Operations, Idaho National Engineering Laboratory, Final
EIS, ERDA Report 1536, p. 111-63 (1977).
11. Sun, J.R., Hydraulic Fracturing as a Tool for Disposal of Wastes in
Shale, Underground Waste Management and Artificial Recharge, Vol. 1.,
P219-270, (1973).
12. Morisawa, s., Inove, Y., Wadachi, Y., and Kato, K., Radiological Safety
Assessment for Low-Level Radioactive Solid Waste Storage Facility,
Preliminary Risk Evaluation by Reliability Techniques, Health Physics,
Vol. 35, (1978).
13. Bardet, C., Experience de Sept Annees de Stockage de Dechets
Radioactifs Solides de Faible et Moyenne Activite en Surface ou en
Transchees Betonnees (IAEA-SM-207/39), Management of Radioactive Wastes
from the Nuclear Fuel Cycle, Vol. II, p. 351-357, (1976).
14. Atomic Industrial Forum Inc. (AIF), Draft Report On-Site Low Level
Radwaste Management Alternatives, NUS Corp. TM, (1979).
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270
THE USE OF A RISK LIMIT AS AN ENVIRONMENTAL SAFETY STANDARD
FOR RADIOACTIVE WASTE DISPOSAL SITES*
A. E. Desrosiers
Battelle, Pacific Northwest Laboratories
P.O. Box 999
Richland, WA 99352
Edwin Njoku
University of Kansas
Lawrence, KS 66044
Abstract
This paper demonstrates how the recommendations of ICRP Publication 26 may
be applied to setting environmental radiation standards for radioactive waste
disposal sites. Traditionally, such standards prescribe the allowable radiation
dose to the maximally exposed offsite individual. Dose limits are usually es-
tablished for the total body and for individual organs. In this paper, the risk
factors recommended by ICRP for individual organs and the doses to those organs
are combined to calculate the total risk per unit of ingested radioactivity. The
allowable ingestion of radioactivity is then calculated from ICRP's individual
risk limit. When these data are compared to normally derived ingestion limits,
significant differences appear whenever a relatively insensitive organ receives
the majority of the dose. Maximum allowable concentrations of radionuclides in
water are derived from the ingestion limits and the concept of an effective
water consumption rate. Using risk assessments in environmental standards for
waste disposal sites would allow 1) a rapid and conservative (cautious) assess-
ment of the potential health impact of a waste disposal facility, and 2) a
simpler evaluation of the impact of ingesting several radionuclides even if each
radionuclide affects different human organs.
Introduction
The International Commission on Radiological Protection (ICRP; ICRP 77) and
the U.S. Environmental Protection Agency (EPA; EPA 78) have recently proposed the
incorporation of risk assessments into the process of setting environmental health
and safety standards. In this context, risk is a combined measure of the proba-
bility and severity of health effects which may result from doses of radiation.
At low doses and dose rates, a linear, zero-intercept regression of observed
incidence rates of health effects versus dose is the basis for calculations of
probabilities. Although the observed health effects occur only at high doses
and dose rates, the probabilities are assumed to be accurate at low doses and
dose rates. The resulting model is not, strictly speaking, a predictor of en-
vironmental health impact, but rather an analytical aid for the comparison of
alternative regulatory strategies. The term health effect normally refers to
a mortality which results from neoplastic disease or to a mortality or severe
morbidity arising from genetic defects. The risk factor is defined to be the
*Work under Contract EY-76-C-06-1830 with the Department of Energy
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271
number of health effects per unit of radiation dose and per person-year at
risk.
ICRP Publication 26, (ICRP 77) recommends an upper limit of 10~5 to 10~6
health effects per person-year at risk for the hazard associated with non-
occupational doses of radiation. The risk limit applies to the sum of risks
from annual external exposures and 50 yr internal dose commitments; the risk
which results from nonuniform exposures is the sum of the risks to each sensi-
tive organ or tissue. This system of risk assessment provides an appropriate
framework for calculating the potential effects of radionuclide ingestion. The
determination of acceptable levels for this risk, however, requires a political
value judgment rather than a scientific evaluation of facts. The former is be-
yond the scope of this paper.
The EPA (EPA 78) acknowledges that acceptable levels of risk are not as
dependent upon statistical considerations as upon society's perceptions of the
benefits and risks of radiation exposures. The proposed radioactive waste
criteria hold that assessments of risk are key elements in the selection of
waste management alternatives. EPA has not proposed numerical standards for
environmental and public health risks associated with radioactive waste disposal.
This paper demonstrates how secondary standards for environmental radio-
activity in the vicinity of waste disposal facilities might include analyses of
risk, discusses practical difficulties encountered in implementing ICRP's method
of risk assessment, and proposes a simple scheme for deriving operating limits
useful in environmental assessments of radioactive waste disposal facilities.
Using an arbitrary nonoccupational dose limit for radioactive waste disposal
facilities, the risk factor for each radionuclide may be converted to a maximum
allowable intake (secondary standard), which is in turn proportional to a maxi-
mum concentration in ground or surface waters (operating limit). These operating
limits may be used to evaluate sites and disposal methods or to compare radio-
active waste management alternatives.
Due to the state of knowledge concerning several key parameters, the con-
tents of this paper should be considered as an example of a methodology rather
than a recommendation for standards or limits.
Method
We selected 12 radionuclides (3H, 51Cr, 51+Mn, 59Fe, 57Co, 59Co, 60Co, 65Zn,
131I, 13LfCs, 137Cs, 239Pu) for consideration on the basis of their significance
in low-level and high-level waste or due to inhomogeneities of internal dose
distribution that enhanced the value of this demonstration. A compilation of
radionuclides which are significant for waste management might also include
14C, 90Sr, "Tc, 129I, 226Ra, 230Th, 232Th, 237Np, 240Pu, 2^Am, and 2"3Am.
However, our purpose is not to perform impact assessments of specific disposal
facilities but to suggest a method of integrating risk assessments into the
derivation of secondary standards and operating limits and to discuss practical
aspects of implementing ICRP's recommendations (ICRP 77). For similar reasons,
we only treat ingestion of radionuclides. We calculated the average 50 yr dose
equivalent commitment (hereafter called "dose") per pCi of ingested radioactivity
to 11 organs (Table 1) of significance in risk analysis (ICRP 77). Each dose
calculation consisted of two parts: the dose to a target organ from radioactivit
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272
in that organ (D t) and the dose to a target organ contributed by penetrating
radiations whose source is a different organ (Dg^.t)• The former calculations
generally employed the method of ICRP Publication 2 (ICRP 57), the latter are
based on Monte Carlo calculations (MIRD 75). All bone marrow calculations,
however, are based upon Monte Carlo calculations (MIRD 75). The quality factor
for 3H beta particles in these calculations is 1.7; this prevents a potential
conflict in comparing these results to models used by NRG (NRC 77) . Some para-
meters for biological uptake and removal rate in organs were not available to
us. In these cases we assumed an organ or tissue uptake proportional to the
uptake by the soft tissues of standard man (ICRP 75). Biological removal rates
that were unknown were conservatively estimated to yield high ratios of dose
per pCi ingested.
Monte Carlo calculations of doses to an organ due to penetrating radiation
in another organ were calculated from the relationship
D = D (S /S )
S"*t S^S S"^t S~*"S
where D and D ^ are the doses from radioactivity which is external or internal
to the organ and S and S are the appropriate ratios of absorbed dose per
unit cumulated activity. The S factors have been tabulated for many radionuclides
and combinations of source and target organs (MIRD 75). Since S is defined in
units of dose per unit cumulated activity, the ratio S ^.t/S ^ is simply the
ratio of dose to the target organ from a different source organ divided by the
dose to the source organ from radioactivity in the source organ. Multiplying
this ratio by D ^ gives the dose to the target organ from radioactivity in the
source organ.
Since we only considered ingested radioactivity, and since the lower large
intestine (LLI) has the greatest cumulated activity, it was the only organ treated
as a source of external exposure to other organs. The upper large intestine,(ULI)
and the small intestine (SI) may also have appreciable cumulative activities.
These organs were not considered as sources of penetrating radiation for other
organs because we achieved the purpose of our demonstration - to show that D
may be significant - using only dose contributions from the LLI. Additional8^1"
doses from the ULI and SI would not qualitatively change our results. For rela-
tively insoluble radionuclides, this method will approximate the correct external
organ dose since a large fraction of the cumulated activity will occur in the LLI.
For soluble radionuclides, the cumulated activity is divided among all target
organs and the doses from radioactivity within the organ will be considerably
larger than doses from penetrating radiation which is emitted in other organs.
Hence in the case of soluble radionuclides, the ratio D /D will be low and
the inclusion of the effects of cumulated radioactivity*£n the ULI and SI would
not significantly alter the results.
Since the ratio ss^.t^B^B will obviously vary according to the mean path or
range of the radiation produced by each nuclear transition, we corrected this
ratio whenever the error would otherwise exceed 10%.
While these approximations are adequate for the present demonstration the
process of standard setting or impact evaluation would require more rigorous
derivations of the dose conversion constants.
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273
The organ specific risk parameters (Table 1) are proportionality factors
which relate the sensitivity of specific organs for radiation induced stochastic
effects to the total risk that results from whole body exposures (ICRP 77). The
risk factors specific to the gonads and breast are averages for both sexes; cancel
cure rates are included in the calculation of risk to breast and thyroid. The
product of the organ specific risk factor and the organ specific dose factor may
be summed for all 11 organs to yield the total stochastic risk per pCi ingested.
The details of this calculation are presented in the case of 58Co (Table 1). The
effective whole body dose equivalent is the ratio of the risk/yCi ingested divided
by the risk/mrem under conditions of uniform irradiation. In the case of insolubJ
58Co, this is 2.6 mrem/yCi (Table 2). That is, ingesting one yCi of insoluble 58C
is equivalent in terms of risk to a uniform whole body dose of 2.6 mrem.
Risk assessments will therefore be systematically incorporated into standards
and criteria for waste disposal if the primary standards are given in units of ef-
fective whole body dose equivalent. ICRP's recommendation of a maximum nonoccupa-
tional risk due to radiation exposures (.10~5 to 10"6 yr"1) is equivalent to an
effective annual whole body dose of 7 to 70 mrem. Assuming for the sake of these
calculations that 7 mrem/yr is the primary standard, the maximum allowable intake
of 58Co is 2.7 yCi/yr (Table 2). The use of 7 mrem/yr in this paper does not con-
stitute advocacy of such a standard, but merely the adoption of a convenient benct
mark.
Operating Limits
For a given radionuclide, the maximum allowable intake for nonoccupationally
exposed individuals is inversely proportional to the dose per yCi ingested and
proportional to the maximum allowable dose. This intake, expressed in yCi/yr,
may then be related to an operating limit, or maximum concentration in ground
or surface waters. Assuming we are concerned with intrusion of radionuclides
into potable or irrigation water and subsequent ingestion by humans (Figure 1),
the yearly radionuclide intake from any pathway will be proportional to the
concentration of radioactivity in the source of contaminated water. Hence the
maximum allowable concentration in water of radionuclide i (W.) is
7 (mrem/yr)
W
R± (mrem/yCi) s f± U (i/yr)
where fjj is the ratio of the concentration of radionuclide i in the material whic
is ingested in pathway j to the concentration of radionuclide i in the drinking 01
irrigation water (yd/kg ingested per yCi/fc of water); R is the effective whole
body dose equivalent per ingested yCi of radionuclide i and U.s is the rate at whic
material j is consumed (kg/yr). The term Ij f-^j Uj may be viewed as an effective
rate of water consumption; fij is then a weighting factor specific to each combin-
ation of radionuclide and pathway.
Assuming that a human population has a single source of drinking and irriga-
tion water, that all vegetables and fruit are cultivated by irrigation, that
drinking water is unfiltered, and that meat and milk are produced from herds
raised on irrigated pastures, we selected values of f±j and U-s that are appropri-
ate for an average adult (NRC 77, Ba 76, Ei 73, Table 3). For 137Cs and 239Pu,
this model yields the following relationships between allowable radionuclide
concentrations in water and the effective whole body dose per unit ingestion to
an average member of the nonoccupational population.
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274
W (137Cs) = 4.1 x 10-^ R-l (yCi/£)
US
W (239Pu) = 2.5 x 10~3 ^ (yCi/£)
Results
Table 2 lists the effective whole body dose equivalent per unit ingestion
for the 12 radionuclides considered here and compares the secondary standards
derived according to limits on effective whole body dose with secondary stan-
dards calculated according to the dose commitment to critical organs. The
ratio of the latter to the former indicates the degree to which uniform limits
for whole body and critical organ doses result in unequal assessment of risk.
The limiting concentrations of 137Cs and 239Pu in water supplies are
137Cs: 1 x 10-5 yCi/£
239Pu: 1.8 x 10-5 yCi/£
under the assumptions of our scenario.
Discussion
The risk to each organ which was calculated in the case of 58Co (Table 1)
indicates that the critical organ does not necessarily bear the majority of the
stochastic risk. For 58Co, the total risk is 3.7 x 10~7 yCi"1 and the risk to the
LLI is 1.5 x 10~7 yCi"1. In particular, the risk to the gonads is approximately
0.8 x 10~7 yCi"1, in spite of the fact that 58Co is not readily absorbed by the
small intestine- The reason: irradiation of the gonads by 58CO .in the LLI pro-
duces approximately 80% of the total dose equivalent which is committed to the
gonads by ingestion of 58Co. This phenomenon also occurs when other relatively
insoluble radionuclides which emit penetrating radiation are ingested.
The preliminary nature of these results must be emphasized. We did not
calculate external organ dose to the gonads from 58Co in the SI or ULI. More-
over, we acknowledge that refinements in uptake, distribution, and "S" factors,
the ratios of dose commitment to cumulated activity in an organ (MIRD 75), may
change these results. Nevertheless, we feel that these data indicate a more
thorough understanding of external or "between-organ" doses may be necessary
before a complete risk assessment is possible in the case of environmental
impact assessments associated with radioactive waste disposal. These results
also indicate that placing equal limits on whole body doses and organ dose
commitments will not produce standards which equalize the risks from each
radionuclide. For 60Co, the allowable intake calculated from considerations
of the risk to specific organs is 6 times greater than the allowable intake
calculated according to the critical organ concept.
Although accurate internal dosimetry calculations are crucial to the environ-
mental impact appraisals, the final results cannot be more accurate than the
accuracy of the intake data. Table 3 indicates that the consumption of fish is
the most important route for ingestion of 137Cs. However, the bioaccumulation
factor and the yearly ingestion rate are, in general, poorly known. Moreover,
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275
organ doses and hence risk, may vary significantly according to age. Usage
factors may vary according to geographical or socioeconomic factors. Agri-
cultural methods may affect the values of f. ..
The calculation of an effective water consumption rate and the weighting
factors, f.., allows a rapid assessment of the relative importance of the
ingestion pathways and provides a focal point in the environmental pathway
analysis where an operating limit may be set. Although the primary limits
will be based on dose to humans, the derived concentration limit for water
provides a convenient design basis for radiological engineers charged with
selecting the operating characteristics of a repository or shallow burial
site. The present MFCs for continuous occupational exposure to water do not
serve this purpose adequately. If only human drinking water were to be con-
sidered in the case of 137Cs, the term £4 f..U. would be equal to 370 £/yr.
By including the other ingestion pathways, lnejsum is 17,000 £/yr, a 47-fold
increase. Doubtless these results would vary widely from site to site. The
precise chemical and physical form of radionuclides which are leached or dis-
solved from waste and enter ground waters may also cause significant changes
in the estimated values of the parameters in these environmental pathways.
Nevertheless, the methodology demonstrated in this report serves as a basis
for performing environmental risk assessments and setting site-specific
environmental limits for waste repository designs.
Summary
Risk assessments related to waste disposal require relatively thorough
analyses of the environmental behavior of relatively few radionuclides.
Present models of the impact of ingested environmental radioactivity provide
calculations of internal organ doses. These dose assessments may be inade-
quate for thorough assessments of risk, in part, because penetrating radiation
emitted in one organ is not considered in doses to other organs; age and uptake
rate may also be factors. The number of organs and tissues included in the
present models may also need to be increased.
Environmental standards based on effective whole body doses and critical
organ dose commitments do not result in equivalent risks for all radionuclides.
The concept of an effective water consumption rate simplifies the process
of deriving operating limits based upon dose limits and risk assessments. These
operating limits may serve as design guidelines for radioactive waste repository
engineers or they may serve to assist regulators in distinguishing between ac-
ceptable and nonacceptable repository designs.
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276
References
Ba 76 Baker, D. A., Hoenes, G. R. and Soldat, J. K., "Food - An Interactive
Code to Calculate Internal Radiation Doses from Contaminated Food
Products", BNWL-SA-5523, Battelle, Pacific Northwest Laboratories,
Richland, WA, 1976.
Ei 73 Eisenbud, M., Environmental Radioactivity, 2nd ed., Academic Press,
New York, 1973.
EPA 78 "Criteria for Radioactive Waste", USEPA, Federal Register, 15
November 1978.
ICRP 59 Report of ICRP Committee II on Permissible Dose for Internal
Radiation, ICRP Publication 2, Pergamon Press, Oxford, 1959.
ICRP 75 Report of the Task Group on Reference Man, ICRP Publication 23,
Pergamon Press, Oxford, 1975.
ICRP 77 Recommendations of the ICRP. ICRP Publication 26, Pergamon Press,
Oxford, 1977.
MIRD 75 S, Absorbed Dose per Unit Cumulated Activity for Selected Radio-
nuclides and Organs, MIRD Pamphlet No. 11, Society of Nuclear
Medicine, New York, 1975.
NRC 77 Calculation of Annual Dose to Man from Routine Releases of Reactor
Effluents. Regulatory Guide 1.109, Rev. 1, USNRC, Washington, D.C.,
1977.
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TABLE 1
277
Risk Parameters
Organ
Gonads
Bone
Red Marrow
Breast
Lung
Thyroid
Lower Large
Intestine
Weighting
Factor (W )
x —
0.25
0.03
0.12
0.15
0.12
0.03
0.06
Risk x 10-9
(mrem"1)
35*
4.2
17
21
17
4.2
8.4
Risk x ID'8
from 58Co in-
gestion (yCi"1)
7.8
4.5
1.4
2.3
1.7
0.17
15.
Stomach, Small
Intestine
Upper Large
Intestine, Liver 0.24
TOTAL 1.00
34
140
8.4
37
*Risk of genetic effect in first two generations
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278
TABLE 2
Total Risk Coefficients and Allowable Intakes
(1)
Radio-
nuclide
H-3
Cr-51
Mn-54
Fe-59
Co-57
Co-58
Co-60
Zn-65
1-131
Cs-134
Cs-137
Pu-239
(2)
Effective Whole
Body Dose Equiva-
lent per Unit
Ingestion (mrea/yCi)
0.10
0.19
2.6
5.7
0.61
2.6
7.0
6.7
71.
40.
40.
140.
(3)
Standard: 7
mrem/yr Effec-
tive Whole Body
Dose Equivalent
69.
37-
2.8
1.3
11.
2.7
1.0
1.1
0.10
0.18
0.18
0.05
Allowable Intake (pCi)
(4)
Standard; 7 mrem/yr
Whole Body Dose Equiva-
lent or Critical Organ
Dose Equivalent Commitment*
69.
10.
0.50
0.21
1.6
0.48
0.17
1.0
0.01
0.06
0.10
0.008
Ratio
(3) (4)
1
4
6
6
7
5
6
1
10
3
2
6
*I-131 intake calculated at 21mrem/yr
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279
FIGURE1
ENVIRONMENTAL PATHWAYS RELEVANT TO RADIOACTIVE WASTE DISPOSAL
PATHWAY
POTABLE WATER
FISH
GARDEN CROPS
MEAT
MILK
f|,U/kgl
137Cs
1
2000
11
4
3
239
"VPu
1
4
12
0.9
<0.01
U.(kg/yr
370
7
190
95
110
ASSUMES 0.017 tlm hr OF IRRIGATION WATER-9 MONTH
GROWING SEASON; 100 yr ACCUMULATION OF 1?7Cs;
1000 yr ACCUMULATION OF 239Pu
TABLE 3
VALUES OF fj. AND U. FOR AN AVERAGE ADULT
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280
RADON — AN ENVIRONMENTAL POLLUTANT?
William A. Mills
Criteria & Standards Division
Office of Radiation Programs
U.S. Environmental Protection Agency
Abstract
Radiological concerns with the disposal and use of mining
and milling residues have heightened to the point that Federal
agencies are asking or being asked to formulate new regulations
for controlling radon daughters from a variety of sources —
radioactivity previously considered to be part of our natural
environment. Based on information derived from epidemiologic
studies of underground miners, particularly uranium miners, the
health impact on the general public is being projected.
Depending on the assumptions made, these projections vary
widely. Because of these variations in health risks, decisions
on control measures have even wider implications on economic
and social considerations. Thus the question: Is radon an
environmental pollutant? While not fully answering the
question, recognizing the uncertainties in assessing and con-
trolling radon daughters can put the question in better
perspective.
Discussion
Sources of radon have become very much a part of the radiation protection
scene. Up until a few years ago, most health physicists probably thought of
exposures to radon and its daughters only in terms of exposures to under-
ground uranium miners and the episode of Grand Junction with the permitted use
of uranium mill tailings in occupied structures. Today, however, such a
peripheral view is not possible and we must view this natural occurring —
human enhanced — potential exposure source with more direct vision.
Evidence of this current direction is recorded in the passage of the
Uranium Mill Tailings Radiation Control Act of 1978, the lengthy deliberations
of the Nuclear Regulatory Commission regarding the inclusion of guidelines for
radon daughter exposures in its "S-3" table for the uranium fuel cycle, and
EPA's proposed actions under the Resource Conservation and Recovery Act
related to "hazardous wastes." Radon daughter exposures play a predominant
role in the decisions to be made in implementing regulations for these and
other responsibilities. Yet, in reality, our knowledge of the biological
effects associated with radon, of general environmental levels, of sources,
of proper instrumentation and measurement procedures, and of the appropriate
dosimetry to be used, is among the most lacking in all of health physics.
The keystone to health physics has been dosimetry — this allows us to do mar-
velous things with units and facilitates understanding problems on a common
ground. However, for radon and its daughters, this common ground is so
complex that many of us now prefer to go directly from how we describe
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281
exposures, working levels, to risk of lung cancer, bypassing the rad/rem
calculations.
I titled this brief discussion "Radon — An Environmental Pollutant?",
leaving it as a question so that each of you, after some thought, can provide
your answers. What I intend to do now is to give you a brief on the problem
so that your thoughts on an answer will be more than a "knee-jerk" response.
In my opinion, too often the response has been that radon exposures are part
of natural background and why attempt to do anything about it — after all,
we evolved in a natural radiation environment. I trust that you will at
least give more thought than this reaction to the question.
On a risk basis, and I will quantitate this later, radon daughters
represent the highest health risk of all radionuclides present in the atmos-
phere. The so-called natural background ambient outdoor levels of radon
range from 40-1000 pCi/m3, according to NCRP Publication No. 45 (NCRP75),
with an average of about 100-200 pCi/m3. The outside ambient background is
about 0.001 WL, but highly variable. However, since most of us spend about
3/4 of our time indoors and the concentration can build up, the exposure
levels we experience are greater than 0.001 WL and average about 0.004 WL
(Ge78). The exposure level variations in residential structures can be con-
siderable, perhaps even by as much as a factor of +10. Actually we have very
limited information on "natural background levels" of radon and radon
daughters, making any decisions using background as a reference point a very
tenuous one at best.
But the determination of whether or not radon is an environmental
problem cannot be answered using solely the range in natural background
levels. The basis for this determination has to lie in how we project the
risks from a given exposure to radon and its daughters. For this deter-
mination we must turn to a limited health effects base which exists in
epidemiologic studies of miners, primarily uranium miners; studies which have
been made in several countries. The most often quoted studies are of U.S.,
Swedish, Czechoslovakian, and Canadian miners.
While any of these studies can be criticized for one reason or another,
it is rather accepted that the risk of lung cancer induction is a function of
cumulative exposure (i.e., CWLM) and that linearity of this relationship
cannot be totally rejected. It can even be generally agreed that, regardless
of whether one uses the absolute model for risk estimates or the relative
risk model, the risk estimates are within factors of two to three when such
estimates are limited to miners. The absolute risk takes the difference
between the risk in the irradiated population and the risk in the nonirradi-
ated population, e.g., number of excess cases per 10" person years per WLM.
Relative risk takes the ratio of risk between the risk in the irradiated
population and the risk in the nonirradiated population, e.g., percent
increase in cases per WLM.
The more serious problem arises when we take the observations for these
populations and have to extrapolate to exposures of the general public. Here
we run into a multitude of complexing factors of differences — differences
in sex, age distribution, breathing rates, smoking habits, in degree of radio-
nuclide equilibria, other environmental factors, etc. These factors do not
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282
act in one direction, e.g., slower breathing rates decrease the exposure, and
younger ages have longer periods at risk; smoking goes both ways, increasing
the promotion of lung cancer but perhaps decreasing the dose because of
thickening of the bronchial epithelium.
Thus, in the end we are cautious on making such extrapolations, but they
must be made because, regardless of how uncertain our health risk may be, the
magnitude of the risk is large enough that the public health implications
cannot be ignored. Drs. Ellett and Nelson of our program, in an analysis to
appear in connection with our efforts to evaluate potential health impact of
radon levels in Florida, have estimated this risk of increased fatal lung
cancer to be between 1-5% per WLM on a relative risk model. This corresponds
to lifetime risks of about 10~3/WLM. Using an absolute risk model, an upper
bound value of about 10~VWLM has been proposed by an ad hoc group of the
Nuclear Energy Agency. I will not attempt here to evaluate the pros and cons
of these estimates because in the end the differences derived are not terribly
important to the regulatory process and, in my opinion, differences in risk
of a factor of 10 are not major determinates for this problem.
Assuming a mortality rate, i.e., lung cancer deaths, of 3% per WLM, it
can be estimated that the risk of lifetime exposure at 0.02 WL is about 1 in
50. More properly stated, this risk is about 2000 excess cancer deaths in a
population of 100,000 persons exposed. The 0.02 WL value is given because
this is a proposed recommendation being considered for the phosphate situation
in Florida. Taking the 0.004 WL value given earlier for indoor background,
we have about 400 cancer deaths attributable to radon daughters, in a total
of about 2900 (^42/yr for 70 yrs), or about 10% of the total lung cancer
deaths. If we take the NEA's absolute model derived value, we can account for
about 1% of the "natural" incidence of lung cancer. In either case, the
apparent contribution of radon daughters exposure to the current number of
lung cancers is not trival. For perspective, consider that for a continuous
lifetime dose equivalent rate of 25 mrem/yr, whole body external exposure,
the risk estimate is 20-100 cancer deaths per 100,000 persons exposed.
While I recognize that in many instances emanations from a given source
cannot be discriminated from the noise of background surrounding that source,
a true public health protection role requires that we give this some attention
and act prudently, especially for those who reside on or near the source and
for which the practice of "as low as reasonably achievable" has obvious public
health meaning. It is also true that, for such sources, we must put ourselves
in a preventative frame of mind and not be in a corrective mode. Reasonable
preventative measures are socially more acceptable and less costly than man-
dated corrective measures; and as an aside, a preventative frame of mind helps
our creditability. In my own mind, I have answered the question — YES!
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283
References
Ge 78 George, A.C. and Breslin, A.J., The Distribution of Ambient Radon
and Radon Daughters in Residential Buildings in the New Jersey-New
York Area, presented at Natural Radiation Environment III,
Houston, TX, 1978 (in press).
NCRP 75 "Natural Background Radiation in the United States," NCRP Report
No. 45. Recommendations of the National Council on Radiation
Protection and Measurements, Washington, DC (November 15, 1975).
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PROPOSED FEDERAL RADIATION GUIDANCE
FOR RADIOACTIVE WASTES
J. E. Martin, H. J. Pettengill, and S. Lichtman
Office of Radiation Programs
U.S. Environmental Protection Agency
Abstract; The Administrator of EPA, in his role of providing Federal
radiation guidance, will soon forward radiation protection guidance
for radioactive wastes to the President for approval. The recom-
mended guidance addresses: (1) the types of materials to be
categorized as radioactive wastes and considered for control; (2)
the use of institutional functions, engineered controls, and natural
barriers to isolate wastes; (3) the potential health risks associ-
ated with wastes, and the factors involved in determining them; (U)
factors which determine the unacceptability of various levels of
risk; and (5) other considerations such as retrievability, site
locations, and communication to succeeding generations to assure
continued isolation. When approved by the President, the
recommendations will guide Federal agencies in making policy,
programmatic, operational, and standard-setting decisions in
-providing radiation protection for radioactive wastes.
Introduction
Federal radiation protection guidance is being developed to establish
the basic radiation protection principles which Federal agencies should
apply in the formulation of policies, plans, programs, standards, and
other decisions involving the storage and disposal of all forms of radio-
active wastes. The guidance would not apply to non-Federal sectors of our
society except through regulatory or other actions by an affected Federal
agency.
The recommended guides were developed in accordance with Executive
Order 10831 and Public Law 86-373 (U.S.C. 2021(h)), which charge the
Administrator to "...advise the President with respect to radiation
matters, directly or indirectly affecting health, including guidance to
all Federal agencies in the formulation of radiation standards and in the
establishment and execution of programs of cooperation with States." The
Department of Energy will provide and operate disposal facilities for
several types of radioactive wastes. The Nuclear Regulatory Commission is
preparing specific regulations for licensing such facilities in accordance
with environmental protection standards which the Environmental Protection
Agency will establish for the various types of radioactive wastes Each
of these Federal activities will be guided by the recommended guidance.
The Agency received considerable assistance from the public in
developing this guidance through participation in two Public Workshops and
a Public Forum. The bases for the guidance are being developed in a
Federal Guidance Report No. 10 which reflects the comments received from
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the Public Forum participants and from public review of the proposed
criteria, which were published for comment in the Federal Register on
November 15, 1978.
Radioactive Waste Materials
The National Environmental Policy Act of 1969 states that each
generation has a responsibility "as trustee of the environment for
succeeding generations." Some radioactive waste materials are both
long-lived and highly toxic, and thus may pose substantial health risks to
both present and future generations.
The primary factor in deciding that a material is a "waste" is the
material's residual value. If it has no designated net value as a
resource or a product or is impractical to control as such, the material
should be considered a waste. Because these recommended guides are
addressed to the treatment of radioactive wastes, as opposed to other
radioactive materials with current or possible future uses, materials
which may have such uses would not be considered wastes subject to the
guidance. Depending on the circumstances, the judgment of the material's
remaining value will be made by the person or organization possessing or
controlling it, or by a government agency on behalf of the society as a
whole. Once a material has been designated as a waste, whether or not it
should be disposed of under this guidance depends on its radiological
hazard and the capability and costs of controls.
The Agency stated in its proposed criteria that the requirement to
keep radiation exposure as low as practicable precluded the establishment
of a general "de minimus" level below which waste materials that are also
radioactive would not be considered radioactive waste. Considerable
public comment was received on this statement with the bulk of opinion
being that such a level was required, and the application of the balance
of the criteria as proposed would result in cie facto levels of jie
minimus. This guidance is designed, as were the proposed criteria, to
allow for various radioactive materials to be excluded for justifiable
causes from additional control because of their radioactivity content, but
these exclusions are to be determined on the basis of the particular
circumstances involved using the principles established in the guidance.
It is proper for Federal agencies to designate the materials that would
and would not be subject to types of controls; in the Agency's judgment
this approach should be used rather than attempting to define an
arbitrary, general level of de minimus which may be overly restrictive for
some situations or non-conservative for others.
Three broad classes of materials should be considered radioactive
wastes for disposal purposes. The first class consists of those materials
produced artifically from nuclear reactions or by fabrication from
naturally radioactive materials into concentrated sources. These
materials are directly controllable and their circumstances are readily
identified and described.
A second class of substances contains diffuse radioactive materials
of natural origin. However, these materials would be subject to control
only if they could result in exposures greater than those which would have
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occurred, through any pathway, prior to the disposal of the material. The
rationale underlying this distinction is that requiring the material to be
treated as radioactive waste implies that its potential radiological risk
is a decisive characteristic in deciding its control. Yet if failure to
treat the material as radioactive waste would not increase exposures over
pre-existing levels, requiring disposal of the material as a radioactive
waste seems unreasonable. If such an improvement is desirable, it should
be for reasons other than that the material is a waste containing some
natural radioactivity.
The final class of wastes which should be considered radioactive
wastes for purposes of disposal under this guidance includes any retained
waste materials which are prohibited by government regulatory action from
unrestricted discharge into the general environment due to their radio-
activity. While this category of substances will obviously contain many
wastes covered by the two previous classes, certain materials fall into
neither category. Regardless, the Agency believes the restriction on
discharge to the general environment is in itself sufficient basis for
requiring any such material to be considered for disposal in accordance
with this guidance. Examples of such materials could include those
removed from effluent streams at nuclear reactors or from the processing
of ores which contain uranium or thorium.
Recommended Guide Number One designates broad types of materials in
these three classes that should be regarded as wastes whose radioactivity
content requires consideration in their disposal. Future study and
information may lead to the designation of other materials in accordance
with the principles discussed above. Usual methods of control of a
material solely as waste may be used if they provide a degree of control
consistent with this guidance.
Control of Radioactive Waste
Federal control of any radioactive waste should attempt to meet a
fundamental goal of complete isolation over the period it would represent
unacceptable risks to humans. Controls for radioactive wastes are of
three general types: engineered barriers, natural barriers, and institu-
tional mechanisms. Engineered barriers such as containers or structures
generally can be considered only as interim measures for containment,
despite the fact that some structures have survived intact through the
ages. Stable geologic media with high retention characteristics are an
example of natural barriers. Institutional controls are those which
depend on some social order to prevent humans from coming in contact with
wastes, such as controlling site boundaries, guarding a structure, land
use policies, record keeping, monitoring, etc.
It generally is accepted that long-term isolation should depend on
stable natural barriers. Institutional mechanisms can be very effective
and are essential in the early stages of management of any waste, but for
practical reasons they can be relied on for only limited periods! The
appropriate time period for relying on institutional controls was dis-
cussed extensively during the developmental stages of this guidance since
the issue is a matter of judgment. A maximum time period of 100 years was
recommended for such controls to be depended upon with any degree of
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assurance. Public comments recommended shorter times, longer times, and
substitution of general goal statements; however, the Agency has not been
persuaded that a different time period would constitute a better
recommendation for reliance on institutional controls.
For wastes that represent unacceptable risks for longer than 100
years, Federal decision makers should, therefore, not approve radiation
protection systems that are based primarily on restricting customary uses
of land and of ground or surface waters for longer than 100 years. This
does not mean that institutional controls are required for 100 years, or
that they must stop at that point if society can still maintain them; only
that people making the initial disposal decision should not plan on their
use to maintain protection beyond 100 years.
Risk Assessment
The risk a waste poses over time is the pivotal part of any
environmental and public health protection policy for radioactive wastes,
and is the key consideration in deciding whether and how to store or
dispose of the wastes. The term risk expresses a general concept encom-
passing both the probability of occurrence of adverse effects and their
severity. Both aspects are basic to decisions on the acceptability of
controls for radioactive wastes; however, there is no generally accepted
methodology for determining the risk that society would accept in a given
set of circumstances.
Risk determinations rest on a number of factors, especially the total
amount of waste material at a particular location, its persistence due to
form and concentration, its potential to enter the biosphere and produce
adverse effects on individuals and populations, the effectiveness of
various controls, and the inherent uncertainties of many of the parameters.
Determinations of risk are most useful in making public policy
decisions if they estimate effects on health; however, projections of
population size, land use, and human factors become very uncertain beyond
a few hundred years. Because of the persistence of many wastes, however,
it is recommended that health effects estimates be performed for at least
1000 years to consider both acute exposure risks and chronic exposure
risks represented in many types of wastes. Some very significant waste
materials persist well beyond 1000 years; thus, it is important that these
be considered in choosing the best control alternative. Such comparisons
can be made with health effects estimates based on very general
assumptions beyond 1000 years; they can also be reasonably made by other
parameters such as curies released, calculated doses to assumed
individuals, etc. Public comments divided over this recommended degree of
emphasis before and after 1000 years, principally, we believe, over
perceptions that resulting risk estimates would be over- or under-
conservative. The intent was, rather, that risk estimates be sufficient
to recognize the potential level of risk and its persistence and yet be of
a reasonable form for responsible decision-making. Reducing the rigor
required for estimates beyond 1000 years was an attempt to consider both
circumstances.
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Examination of risks for waste disposal systems in terms of the
consequences of exposure and their probability of occurrence will result
in values greater than zero which vary with the types of waste, the
controls available, and the kinds and severity of events that can be
postulated to disrupt isolation. Consequences may range over chronic and
acute exposures of a few individuals or large groups, with probabilities
ranging from very remote to near certainty over the long term for some
waste types and controls.
Risk Acceptability Considerations
All societies experience risks and have developed patterns of
acceptance of various types of risks. For high-technology societies these
patterns are difficult to discern, and they often change in relatively
short time frames because such societies are continually creating new
sources of risks. As a starting point, it appears that any present or
future risks due to radioactive wastes would not be acceptable to society
if reasonable controls that are available and economically feasible have
not been used. Also, because of the responsibility each generation has to
succeeding ones, a key social consideration is that at a minimum the
current generation should not pose larger risks to a future generation
than it would be willing to accept for itself. This does not mean that
the risk has to be the same in future generations, only that it would not
be unacceptable to the current generation if imposed on it — even though
the projected impact on the first few generations may be well below levels
determined to be acceptable.
In general, risks are more readily acceptable to society if the
consequences are common ones that are not highly dreaded, and are well
understood situations which can be related to other common risks. On the
other hand, generally beneficial circumstances which have high potential
for harm appear to be acceptable to people if occurrence of the harm is
virtually impossible; that is, the probability is very low. These inter-
relationships between probability of occurrence and consequence are not
linear functions, and thus an unacceptable risk value can not be chosen as
a product of the two. Since a broad range of circumstances is possible
for wastes, a continuum of probability/consequence is possible; thus, it
is only possible to give general guidelines for deciding when various
levels of risk would be unacceptable. These are rooted in the basic
concepts discussed above: risks due to events likely to occur will be
acceptable only if adverse consequences are low and are of a common type;
events with high adverse consequence potential must be virtually certain
not to occur.
High consequence events resulting from waste disposal can be compared
to those associated with productive, shorter term technologies such as
dams, dikes, and large stores of toxic or explosive chemicals. High
consequence events for radioactive wastes should be considered unaccept-
able unless their probability is only a small fraction of that deemed
acceptable for shorter term productive technologies.
Exposure situations which are likely to occur for certain kinds of
wastes would be unacceptable unless predicted effects would be of ques-
tionable significance, even on a statistical basis. This means that if
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acute exposure is projected as a likely result of waste disposal, it
should affect only a few individuals randomly. If large groups are likely
to be chronically exposed, the projected risks should be small and no
greater than comparable risks that society has already willingly accepted.
Each Federal agency that is involved in establishing policy for
radioactive waste disposal will be responsible for estimating the
consequences of likely radiation exposure circumstances for wastes, and
for determining whether they are unacceptable based on an examination of
their consistency with other established social values that have evolved
for comparable circumstances.
Supplementary Protection Goals
A number of other subjects such as provisions for retrievability,
choosing site locations, and passive communication to future generations
may provide positive aspects for control of radiological hazards; however,
their application may, in some instances, undermine the goal of providing
permanent isolation for wastes. For example, it is difficult to maintain
retrievability or conduct a monitoring program without compromising the
ability to provide isolation. It is not appropriate to depend upon
methods and techniques such as these or other similar ones for long-term
control; nonetheless, when such methods could be reasonably applied to
enhance overall protection from wastes, it is prudent to use them.
Provisions for retrievability of radioactive waste is desirable, if
isolation is not compromised, in order to provide a mechanism for correct-
ing losses in protection due to unforeseen circumstances. Wastes may also
become future resources which has led some to argue that retrievability
should be required for this possibility; however, it appears that such
materials ought not be disposed of in the first place.
It is desirable to select sites if they are reasonably available
where the action of natural forces over time such as erosion, sedimen-
tation, and crystallization can be projected to improve, rather than
reduce, isolation of the wastes over the time they may represent
potentially unacceptable risks if they were to enter the biosphere. Such
sites may not be practicably available, but if they exist among available
alternatives they should be considered for use. Similarly, if geological
media are used, they should reduce the effect of potential adverse
interaction of the wastes with water to the greatest extent possible.
Public comment on the proposed criteria discussed the desirability of
choosing waste disposal locations away from valuable natural resources to
reduce the likelihood of inadvertant destruction of the isolation
capability of the site. Federal agencies should consider this aspect of
siting when it is practicable to include it, all other features being
equal.
In many disposal situations, the residual risk will mainly be attrib-
utable to potential failure mechanisms involving eventual intrusion by
humans. Passive methods of communicating the hazard, such as markers
which call attention to the waste, may sometimes be judged to provide a
net reduction of risk. Other passive methods, such as creating records
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describing the waste, or setting aside of the land title to the disposal
site, have value in reducing the likelihood of intrusion for some limited
time.
An example of a circumstance where land title transfer is reasonable
is a current site that has been in use for some time where optimal
environmental isolation is no longer a practicable alternative, such as an
abandoned mill tailings site, a nuclear test facility site, etc. In these
cases, Federal ownership of the land beyond the normal period of institu-
tional control would be reasonable to minimize potential intrusion. Such
decisions should be made on a case-by-case basis and provision for
specifically treating such exceptions should be addressed in standards and
regulations which are promulgated for these types of wastes.
Recommendations
Recommended guides have been developed which address each of the
considerations presented above. It is intended that upon adoption, each
of the following recommended guides would be applied, unless a specific
circumstance is excepted in a guide, by Federal agencies in providing
environmental and public health protection for radioactive wastes:
Recommended Guide No. _1_: Radioactive materials should be considered
by Federal agencies for control as radioactive wastes if they have no
designated net product or resource value and: 1) are human-produced by
nuclear reactions, fabricated from naturally radioactive materials into
discrete sources, or as a result of regulatory activities are prohibited
from uncontrolled discharge to the environment; or 2) contain diffuse
naturally-occurring radioactive materials that, if disposed into the
biosphere, would increase exposure to humans above that which would occur
normally in pathways due to the pre-existing natural state of the area.
Examples * of radioactive waste materials that should be subject to
environmental protection requirements are:
a. All radioactive materials associated with the operation and
decommissioning of nuclear reactors for commercial, military,
research, or other purposes and the supporting fuel cycles, including
spent fuel if discarded, fuel reprocessing wastes, and radionuclides
removed from process streams or effluents.
b. Artificially produced radioisotopes, including concentrated
radium sources, for medical, industrial, and research use and waste
materials contaminated with them.
c. The naturally-radioactive residues of mining, milling, and
processing of uranium and phosphate ores.
«
The materials listed should be subject to this Federal radiation
protection guidance even though some such materials may not upon
examination require any control above that they would receive as ordinarv
wastes; Federal agencies may also designate other radioactive materials
for consideration if they are found to satisfy this guide.
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Recommended Guide No. 2\ Radioactive wastes should be controlled to
meet a fundamental goal of complete isolation over the period they repre-
sent unacceptable risk. Controls which are based on institutional
functions should not be relied upon for longer than 100 years to provide
such isolation; radioactive wastes that represent unacceptable risk in
excess of 100 years should be controlled by engineered and natural
barriers.
Recommended Guide No. 3_: Radiation protection requirements for
radioactive wastes should be based primarily on an assessment of risk to
individuals and populations; such assessments should examine at least the
following factors:
a. The amount and concentration of radioactive waste in a location
and its physical, chemical, and radiological properties;
b. the projected effectiveness of proposed alternative methods of
control;
c. the potential adverse health effects on human individuals and
populations for a reasonable range of future population sizes and
distributions, and of uses of land, air, water, and mineral
resources for one thousand years, or any shorter period of hazard
persistence;
d. estimates of environmental effects using general parameters or of
health effects based on generalized assumptions for as long as
the wastes pose a hazard to humans when such estimates could
influence the choice of a control option;
e. the probabilities of releases of radioactive materials to the
general environment due to failures of natural or engineered
barriers, loss of institutional controls, or intrusion; and
f. the uncertainties in the risk assessments and the models used for
determining them.
Recommended Guide No. ^: Any risks due to radioactive waste storage
or disposal activities should be deemed unacceptable unless it has been
justified that the further reduction in risk that could be achieved by
more complete isolation is impracticable on the basis of technical,
economic, and social considerations; in addition, any method of control
should be considered unacceptable if:
a. radiation risks to a future generation are greater than those
acceptable to the current generation;
b. probable events could result in adverse consequences greater than
those of a comparable nature generally accepted by society; or
c. the probabilities of highly adverse consequences are more than a
small fraction of the probabilities of high consequence events
associated with productive technologies which are accepted by
society.
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Recommended Guide No. 5: Certain additional procedures and
techniques should also be applied to waste disposal systems which
otherwise satisfy these guides if their use provides a reasonable net
improvement in environmental and public health protection. Among these
are:
a. procedures or techniques designed to enhance the retrievability
of the waste;
b. selection of waste disposal locations which take advantage of
natural processes to enhance isolation over time and minimize
unintentional disruption due to resource recovery; and
c. passive methods of communicating to future people the potential
hazards which could result from an accidental or intentional
disturbance of disposed radioactive wastes.
Implementation and Follow-Up
It is expected that each Federal agency will use the recommended
Federal radiation protection guides as a basis for developing detailed
radiation protection requirements for radioactive wastes which are
consistent with its particular responsibilities. The Agency, in
cooperation with other Federal agencies, will follow the implementation of
these recommended guides, and will promote the coordination necessary to
achieve an effective Federal radiation protection program for radioactive
wastes. Periodically, the Agency will interpret and expand upon each of
the recommendations as required to assure effective implementation by
Federal agencies in accordance with new and changing information.
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SESSION E
ENVIRONMENTAL - PUBLIC HEALTH ASPECTS I
Session Chairperson
S. V. Kaye
Oak Ridge National Laboratory
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PREVIOUS MANAGEMENT PRACTICES FOR NATURALLY OCCURRING
RADIONUCLIDE WASTES: CURRENT RADIOLOGICAL STATUS
W. A. Goldsmith, D. J. Crawford,
F. F. Haywood, and R. W. Leggett
Health and Safety Research Division
Oak Ridge National Laboratory*
Oak Ridge, Tennessee 37830
Abstract
Many installations used during the early days of the United States'
atomic energy program have been released in recent years for unrestricted
private uses. These installations include lands and buildings used for
the storage of radioactive wastes resulting from refining and processing
of uranium and thorium. Waste management practices at these sites in
the 1940's and 1950's were not conducted with today's emphasis on as-
low-as-reasonably-achievable (ALARA) principles. Consequently, many of
these older waste storage areas are contaminated with naturally occurring
radionuclides in concentrations which are orders of magnitude greater
than those found ordinarily in the earth's crust. Current and potential
elevated human exposures at fifteen of these sites are due primarily to
radon daughters and external-gamma radiation. A wide variety of exposure
conditions may be found at these sites — ranging from slightly above
background to more than thirty times the guidelines recommended for the
public. Remedial actions are contemplated for a number of these sites
where contamination levels or radiation exposures exceed current guidelines.
Introduction
Early in this nation's development of atomic energy, extensive
efforts were made to utilize all available resources for a program to
demonstrate controlled nuclear fission. Initially, this program was
administered by the Department of the Army under the Manhattan Engineer
District (MED). Although conducted in wartime secrecy, the MED program
encompassed a wide variety of materials research and development activities
as well as various commercial source material handling operations.
Contracts for needed services were entered and terminated as required.
Administration of the MED program was taken over by the Atomic
Energy Commission (AEC) after the conclusion of World War II. Although
the initial MED/AEC contractor facilities processed uranium, increased
interest in the thorium fuel cycle resulted in a corresponding increase
in the number of facilities involved in thorium processing. Services
provided by private commercial firms under MED/AEC contracts covered a
*0perated by Union Carbide Corporation under contract W-7405-ene-26
with the U. S. Department of Energy.
By acceptance of this article, the
publisher or recipient acknowledges
the U.S. Government's right to
retain a nonexclusive, royalty-free
license in and to any copyright
covering the article.
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wide variety of activities such as ore transport and storage; dissolution
and leaching of ores; production of mill concentrate (yellow-cake);
refining of mill concentrate; conversion of this refined product to other
compounds and/or metal; smelting, rolling, extrusion, cutting, and
packaging uranium and thorium metal products for distribution to other
institutions such as national laboratories; and the recovery of uranium
from scrap and salvaged material.
Disposal of radioactive residues frequently consisted of shallow-
land burial on-site or at some federally owned or leased land in the
vicinity of the site. At the termination of contract operations,
efforts were made to decontaminate buildings, land, and equipment to
levels consistent with guidelines which existed at that time. However,
no consideration was given to cleanup in accordance with ALARA objectives.
In many cases, present records are insufficient to document the
radiological condition of these sites, most of which are now in the
public domain. The overall program to determine the current status of
these sites has been described elsewhere (e.g., Ha78, DOE78). This
paper will present results found at three sites involved with waste
disposal during MED/AEC activities; these results represent the range of
findings obtained during the current resurvey program.
Early Solid Waste Management
The handling and processing of ores and ore concentrates produced
large volumes of low-level solid residues at MED/AEC contractor facilities.
Furthermore, solid wastes were generated when equipment and building
surfaces were decontaminated or discarded. These wastes, particularly
the residues, contained most of the radioactive material present in the
original ore or ore concentrate. Consequently, these wastes are the
source of present day radiation exposures at MED/AEC sites.
The highest priority items at MED/AEC sites, particularly during
World War II, were the development of processing technology and the
production of source material. Typically, a portion of the contractor's
land would be dedicated to surface storage or shallow burial of process
residues. If this proved to be impractical, a contract for these purposes
would be established with a nearby property owner. Residues were generally
regarded as wastes when the source material content was no longer recover-
able by the contractor's processes. Thus, qualititative and quantitative
aspects of the radionuclide inventory of residues were quite dependent on
the processing history of the material.
Previous waste management practices are intimately associated with
the present radiological status of fifteen former MED/AEC sites surveyed
to date by the Oak Ridge National Laboratory (ORNL). The contamination
of these sites by radionuclides such as 230Th, 232Th, 227Ac, 228Ra,
225Ra, and 210Pb indicates that problems could be associated with
inhalation and ingestion of these long-lived materials. On-site measure-
ments indicate that external gamma radiation exposures and, where
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structures are present, indoor radon daughter concentrations constitute
the principal radiation exposure problems at these formerly utilized
sites (e.g., Le78a).
Radiological survey results indicate that present exposures range
from those which cannot be distinguished from background to more than
thirty times the guideline values recommended for the general public
(Cr78, Le78a). These results are due to a wide range both in exposure
rates and in site occupancy. Radiological survey results from three
sites will be used to demonstrate the relationship between previous
solid waste management practices and present radiation exposures at
formerly utilized sites.
Former Vitro Rare Metals Plant, Canonsburg, Pennsylvania
The 8-ha (18-acre) site at Canonsburg, Pennsylvania, was used for
the commercial extraction of 226Ra from 1911 to 1922. From 1930 to
1942, radium and uranium salts were extracted for commercial purposes.
From 1942 to 1957, uranium was recovered from ores, concentrates, and
scrap materials under MED/AEC contracts. The site remained under the
control of various AEC licenses until 1966. Since 1967, the property
has been developed by the present owner. Various light industries
currently occupy the twelve buildings in what is now an industrial park.
Approximately 125 persons are employed at this site. None of these
employees is performing any work which is related to the nuclear industry.
The site is divided into three separate parcels, designated A, B,
and C as shown in Fig. 1. Extraction of radium began on the western side
of parcel A. Residues were dumped on parcels A and B. As the plant
expanded toward the east of parcel A, new buildings were constructed
over the residues. Liquid slurry wastes were impounded in a swampy area
located in parcel C; this area was later filled with site residues and
covered with uncontaminated dirt.
Results of the radiological survey (Le78a) indicate that a layer of
contaminated soil can be found within 1 m of the surface at almost any
point on the site. Apparently, all buildings on the site are built over
or directly adjacent to contaminated soils. A "typical" boring on
parcel A would indicate contamination to a depth of almost 3 nr the
average 226Ra content of the core would be about 200 pCi/g. In parcel
B, about the same results could be expected except for an area in the
center which has up to 2 m of fill material over the contamination.
A 150-cm layer of highly contaminated muck (up to 17,000 pCi/g of 226Ra)
would be encountered within 1 m of the surface of parcel C. The ratio
of 226Ra activity to 238U and 227Ac activities varied widely from sample
to sample because of the wide variations in processes and in feed
material used to generate the residues.
A summary of radiation exposures being received by the industrial
park employees is given in Table 1. These elevated exposures can be
attributed directly to the contaminated residues which cover and underlie
practically the entire site. Variations in external gamma radiation
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levels can be correlated fairly well with variations in subsurface
contamination (Le78a). Radon and radon daughter concentrations in
buildings can be attributed to 226Ra contamination in surface and subsurface
soils, in floor and former process drains, and on interior surfaces of
the buildings. Daughters of 219Rn, attributable to 227Ac contamination
in surface soils, were also detected inside structures (Le78a).
The ranges of average exposures inside structures shown in Table 1
represent the lowest and highest average obtained in any building on
site. Each of these individual building averages represents numerous
individual measurements. However, the range for airborne 230Th represents
the range observed in individual spot samples; these should not be
construed as average values.
A summary of current exposure guidelines for an individual in the
general public is given in Table 2. These guidelines, except for radon
daughters in commercial structures, assume that the exposure is
continuous. Comparisons between Table 2 and Table 1 show that employees
at the Canonsburg site are exposed to average radon and radon daughter
concentrations which are far in excess of the guidelines. In fact, only
a portion of one of the twelve buildings had average radon and radon
daughter concentrations below the guidelines. Remedial measures are
obviously required to reduce on-site exposures to radon and its daughters.
Results of spot samples and the presence of alpha contamination on
practically all building surfaces indicate that occupants of some of the
buildings on this site may be exposed to average concentrations of long-
lived airborne radionuclides (particularly 230Th) which exceed guideline
values.
Exposures at the Canonsburg site are far higher than those associated
with the other MED/AEC sites surveyed by ORNL. The Canonsburg site is
the only MED/AEC site which was specifically included in the "Uranium
Mill Tailings Radiation Control Act of 1978" (H.R. 13650, 95th Congress).
Under the provisions of this act, the Secretary of Energy is authorized
to conduct remedial measures at this site.
Pennsylvania Railroad Landfill Site, Burrell Township, Pennsylvania
The Pennsylvania Railroad Landfill Site is located approximately
one mile southeast of Blairsville, Pennsylvania, in Burrell Township.
This property, owned by the Properties Division of the Penn Central
Transportation Company, lies between the Conemaugh River and the mainline
tracks of ConRail (see Fig. 2) and consists of approximately 25 ha
(60 acres).
During the period of October 1956 through January 1957, an estimated
10,500 metric tons of radioactive material were shipped by rail from the
uranium processing plant in Canonsburg, Pennsylvania, and were dumped on
the site. The material contained approximately 5000 metric tons (dry
weight) of waste residues containing an average 0.097% U^OQ by weight.
The uranium (approximately 5 metric tons of UsOg) was classified as
"unrecoverable material-measured." The wet weight of the residues was
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298
estimated to be 9000 metric tons, and since the material was shipped
wet, it appears that approximately 1500 metric tons of possibly nonradio-
active materials were mixed with the radioactive materials during loading.
The waste residues were generated under an AEC contract at the Canonsburg
plant.
The area where the residues were dumped was previously the river
bed of the Conemaugh River, which had been diverted approximately 150 m
to the south several years earlier. Apparently, this site was chosen,
in part, because the dumping site and scattering technique used in
unloading from the railroad cars would cause the material to be widely
dispersed and intermixed with large volumes of nonradioactive wastes.
Furthermore, it was thought that the material would be confined in a
large chasm approximately 10 m deep. There were, and still are, no
public thoroughfare passes through the site or in its immediate vicinity,
other than the railroad. The nearest dwelling is approximately 150 m
from the site. There are probably a few persons, for example, railroad
workers or hunters, who may occasionally be on the site.
Results of the radiological survey (Le78b) indicated more than 75%
of the residues lie at least 3 m beneath the surface. It was estimated
on the basis of historic records that about 1.5 Ci 238U was dumped
on this site. Auger and core hole analyses performed during the recent
survey could account for approximately 1.3 Ci (Le78b). Most of this
activity was in an area of less than 2 ha (4 acres). Thus, it appears
that most of the dumped residue has been accounted for by the recent
survey.
Areas of surface contamination appear to coincide to a large extent
with those of subsurface contamination shown in Fig. 2. Beta-gamma dose
rates as high as 5.4 mrad/hr were measured at 1 cm above the surface;
however, most values were below 0.1 mrad/hr. The average 226Ra content
of numerous surface soil samples taken from the shaded area of Fig. 2
was 10 pCi/g; 238U average was 3.9 pCi/g. Background concentrations in
the Burrell Township core soils are 1.9 and 0.9 pCi/g, respectively.
A summary of current exposure conditions at the Burrell Township
site is given in Table 3. The average outdoor 222Rn concentration on
the site was 0.52 pCi/A. However, instantaneous measurements as high as
9.7 pCi/£ were observed. Radon daughter concentrations measured on the
site were reasonably typical of outdoor radon daughter measurements in
that area of Pennsylvania. Average background gamma radiation in the
Burrell Township area was found to be 8 yR/hr. Thus, the general average
of external gamma radiation on site is slightly above background.
Furthermore, the concentrations of radionuclides in all water samples
were below the concentration guide for water in 10 CFR 20.
In summary, the Pennsylvania Railroad Landfill Site is contaminated
by about 4 Ci of 22&Ra and 1.5 Ci of 238n spread over &n a^^f about
2 ha (4 acres). Although most of the contamination is presently a
few meters below the surface, the random dumping of materials has
resulted in some areas of significant surface contamination Radiation
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299
exposures on this site are slightly above regional background. Occupancy
of the site is very infrequent. Scenarios can be offered, such as
building structures over contaminated soils on the site, which could
result in overexposures to site occupants.
This site is much more typical of the MED/AEC sites surveyed by
ORNL than is the Canonsburg site. The Burrell Township site has a
relatively small amount of measurable contamination. However, some type
of remedial action, consistent with ALARA principles, could be taken to
reduce potential radiation exposures to potential site occupants.
Middlesex Landfill Site, Middlesex, New Jersey
In 1948, about 6000 m3 of dirt contaminated with pitchblende ores
were brought to the 10-ha (23-acre) landfill site from the former Middlesex
Sampling Plant. In 1960, elevated gamma radiation levels were detected
on this site by civil defense monitors during a local civil defense
exercise. A radiological survey of the site was made at that time by
the AEC, and it was found that external gamma radiation levels over an
area of approximately 2000 m2 were 20 to 50 times background levels for
the surrounding area. The AEC subsequently removed approximately 600 m3
of the contaminated material nearest the surface and covered the area
with about 1 m of uncontaminated dirt. This action reportedly lowered
the external gamma radiation levels to no more than 50 yR/hr.
In 1974, a second survey of the site was performed to reevaluate the
radiological conditions. During the time between the 1960 and 1974 surveys,
an approximately 2-ha parcel of the landfill site (originally owned by
the Borough of Middlesex) had been sold to the Middlesex Presbyterian
Church, and a church had been constructed on the parcel. During weekdays,
part of the church building and grounds is used as a day care center for
local children. The church and the Middlesex Municipal Building are
located on the western edge of the site. It appears from discussions
with local people that both the church and the Middlesex Municipal
Building were constructed on "non-fill" or solid ground. The landfill
site is surrounded by residences which approach to within 0.5 km of the
south and west and to Bound Brook on the eastern and northern edges.
Results of the 1974 AEC survey indicate that the remaining contamination
on the property was in an area (see Fig. 3) bounded by the baseline and
by the lines designated as 300R, 2+0, and 6+0 (Cr78).
Radiological survey results showed that average surface contamination
of 226Ra and 238U in soil were indistinguishable from background activity
of about 1 pCi/g of each. A few subsurface samples contained detectable
activity caused by small pieces of material, presumed to be pitchblende
ore. These contaminated samples were found in the general area referred
to in the previous paragraph, generally at depths of less than 4 m.
A summary of present radiation exposures is shown in Table 4.
Background gamma radiation measurements in this area ranged from 5 to
10 yR/hr, with an average of 8 yR/hr. Hence, the average external gamma
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300
radiation at this site is within the range of area background. The
maximum external gamma radiation was associated with surface contamination
in an area of about 50 m2. Construction of a building on this contaminated
area, particularly a building with a basement, could lead to elevated
human exposures. Furthermore, the underground contamination poses the
potential for producing elevated exposures if future activity at the
site were to uncover pieces of uranium ore at or near the surface.
This site represents the radiological condition at those MED/AEC
sites where present radiation exposures cannot be distinguished from
background over almost the entire site. Minor remedial measures at
these sites are expected to involve minimal expenditures.
Summary and conclusions
The three sites summarized in this paper represent the range of
results found in ORNL surveys of waste storage areas at former MED/AEC
sites. Most of the sites would be typified by the Burrell Township
site—current radiation exposures averaged over the site are slightly,
but demonstrably, greater than background; small portions of the site
contain highly contaminated material in close proximity to the surface
of the ground. Exposures to 222Rn and its daughters in structures built
on these contaminated areas would probably exceed guidelines. Exposures
to external gamma radiation will probably exceed guidelines at several
points within the contaminated areas. However, the Burrell Township
site does not demonstrate appreciable radionuclide migration due to
surface runoff. This is an appreciable problem at three other MED/AEC
sites. Remedial actions are required at sites such as Burrell Township.
The Middlesex Landfill Site represents the lower end of the
spectrum of present exposures found at MED/AEC sites. Exposures can be
differentiated from background only at small portions of these sites.
Minor remedial action would be required at these sites.
The Canonsburg site represents the upper end of the exposure
spectrum found at MED/AEC sites surveyed by« ORNL. Practically all of
the present exposures at Canonsburg can be traced to previous waste
management practices at that site. Extensive remedial measures are
required to reduce radon and daughter concentrations in buildings as
soon as possible at Canonsburg.
Waste management practices employed during the time that MED/AEC
contracted activities were performed have a major bearing on the extent
of current radiation exposures at these sites. The magnitude of these
exposures is directly related to the amount of contamination still
present at a site. Hence, sites with the best planned waste management
practices generally have the lowest present day exposures.
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301
REFERENCES
Cr78 Crawford D. J., Cottrell W. D., Wagner E. B., Shinpaugh W. H.,
Leggett R. W., Haywood F. F., Doane R. W., and Christian D. J.
1978, "Radiological Survey of the Middlesex Landfill Site, Middlesex,
New Jersey," Oak Ridge National Laboratory, to be published as a
Department of Energy report.
DOE78 U. S. Department of Energy, Office of Public Affairs 1978, "DOE
Updates List of Former Nuclear Sites Included in Radiological
Survey Program," Press Release R-78-226, June 29.
Ha78 Haywood F. F. 1978, "In Search of Yesteryear," Presented at the
23rd Annual Health Physics Society meeting, Minneapolis, Minnesota,
June 18-23, 1978.
Le78a Leggett R. W., Haywood F. F., Barton C. J., Cottrell W. D.,
Perdue P. T., Ryan M. T., Burden J. E., Stone D. K., Hamilton R. E.,
Anderson D. L., Doane R. W., Ellis B. S., Fox W. F., Johnson W. M.,
and Shinpaugh W. H. 1978, "Radiological Survey of the Former Vitro
Rare Metals Plant, Canonsburg, Pennsylvania," Oak Ridge National
Laboratory, DOE/EV-0005/3.
Le78b Leggett R. W., Crawford D. J., Ellis B. S., Haywood F. F., Wagner E. B.,
Loy E. T., Cottrell W. D., Anderson D. L., and Shinpaugh W. H. 1978,
"Radiological Survey of the Pennsylvania Railroad Landfill Site,
Burrell Township, Pennsylvania," Oak Ridge National Laboratory,
to be published as a Department of Energy report.
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302
Table 1. Summary of current on-site exposures at the
Canonsburg, Pennsylvania, site
Exposure source
Range of average
values observed
Maximum value
observed
Radon in air inside
structures
Radon daughters in air
inside structures
External gamma radiation
inside structures
Airborne 230Th inside
structures
Radon in air outside
structures
External gamma radiation
outside structures
2.6 to 107 pCi/A average 227 pCi/£
daytime concentration
0.01 to 0.43 WL* average 0.5 WL+
daytime concentrations
20 to 80 yR/hr averaged 310 yR/hr
over buildings
0.003 to 0.2 pCi/m3 0.2 pCi/m3
range of spot samples
2.5 to 17 pCi/Jl average 69 pCi/£
24-hr concentration
110 to 210 yR/hr averaged 1600 yR/hr
over parcels A, B, and C
*The WL (working level) is defined as any combination of short-
lived radon progeny per liter of air which will result in the ultimate
emission of 1.3 x 105 MeV of alpha energy by decay to 210Pb.
"^Measured during good ventilation conditions. Under poor
ventilation conditions (cold weather) maximum is estimated to be 1.9 WL.
Table 2. Summary of current guidelines for exposure to
a member of the general public
Exposure source
Guideline value
Documentation
Radon in air
Radon daughters in air
commercial structure
residential structure
Airborne 230Th (insoluble)
External gamma radiation
(whole body)
3 pCi/£
0.03 WL
0.03 WL
0.01 WL
0.08 pCi/m3
500 mrem/yr
10 CFR 20
10 CFR 20
10 CFR 712
10 CFR 712
10 CFR 20
10 CFR 20
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303
Table 3. Summary of current on-site exposures at the
Burrell Township, Pennsylvania, site*
Exposure source
Radon in air
Radon daughters in air
External gamma radiation (at 1 m)
Beta-gamma radiation (at 1 cm)
Average values
observed
0.52 pCi/£
0.0009 WL
11 uR/hr
<0.1 mrad/hr
Maximum values
observed
9.7 pCi/A
0.001 WL
630 yR/hr
5.4 mrad/hr
*Includes component of exposure due to background.
Table 4. Summary of current on-site exposures at the
Middlesex, New Jersey, landfill site
Exposure source
Average values observed* Maximum observed value
Radon in air
External gamma radiation
0.04 pCi/£
5 pR/hr
Calculated to be 0.01
pCi/£ above background
32 uR/hr
*Includes component of exposure due to background.
-------
U-800
L-TOO
ota 1+0 z+o 'tlffe^—4+o s+o 6+0 7+0-— e-K> >9
L-4OO-,
12+0
13+0
^
*<*
t?
L. _^
X
rl
GEOR6E STREf
KNN, GENUAL RAILROAD
, I I I I I
T 1
i
1
~~~-1
i —
"^
N
^"^i^Jt-4-a-i
LINE
sc*u
IOOFI
00
o
Fig. 1. Layout of the present setting of the site at Canonsburg, Pennsylvania.
-------
ORNL-DWG 78-7737
900R
I I I I 1 I I I
--T-+-4-+-J-4-4-4-4
_^__h_4_4_+_4-4
.<^f - H ---I--
200 L
300 L
-1 -2
4OOL
O 50 100
I I I
200
SCALE(FEET)
Fig. 2. Layout of landfill site at Burrell Township, Pennsylvania. Shaded
areas are those where subsurface contamination has been found.
CO
o
en
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306
ORNL DWG 78-20630
200' L
100' L
100* R
ZOO' R
600' R
Fig. 1
Fig. 3. Present setting of the Middlesex, New Jersey, landfill
site.
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307
Radiological Impact of Uranium Tailings and Alternatives for Their Management
M. H. Momeni, W. E. Kisieleski, S. Tyler, A. Zielen
Y. Yuan and C. J. Roberts
Division of Environmental Impact Studies
Argonne National Laboratory, Argonne, Illinois 60439
ABSTRACT
The radiological hazards associated with uranium tailings arises from
inhalation of airborne particulates and radon daughters, ingestion of food
grown in contaminated ground, and from external exposures to pollutants in the
vicinity. Uncontrolled tailings piles are mobile sources of fugitive dust that
may produce a practically uncleanable adjacent environment. A practical pro-
cedure for managing solid tailings is addition of surface moisture, mechanical
and gravitational separation of slimes, and storage of slimes below solution
tailings. Presently practical alternatives for tailings management are vari-
ations of two basic methods—surface and below-ground disposal. Protocol for
tailing management should be based on both reduction of exposure pathways and
containment throughout the predictable future. Isolation of tailings by
natural materials such as clay lenses and combinations of overburden, top soil,
vegetation and rip-rap may provide both minimization of exposure and stability.
Experimental measurement of radon flux over two inactive tailings, acid and
carbonate leached tailings resulted in average specific flux values of
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308
tailings pond beaches and eroded tailings piles. These releases result in
inhalation of particulates and radon progeny, ingestion of food produced from
contaminated ground and water, and external exposure to beta-gamma radiation.
Generally, liquid-borne mill wastes (tailings) are impounded in tailings
ponds. Acid tailings are about 50% solids by weight. In many older mills the
peripheral dikes are composed of the solid tailings themselves. These mills
generally separate coarse sands in covering the exterior surface of the dam
and discharge the finer particles (slimes) into the interior surface of the
dam. Because of their mobility, the dusts and sands from these tailings
contaminate the land and surface waters collected from the watersheds. In the
following sections a selected alternative for tailings management to mitigate
the potential releases are analyzed and a selection of the experimental data
are reported.
FIELD INVESTIGATION
Experimental Procedures
The field investigations were made in support of the Uranium Milling
Operation Generic Environmental Impact Statement for United States Nuclear
Regulatory Commission. A major part of the field study has been conducted
since June, 1977, at the Anaconda Uranium Mill, Bluewater, New Mexico, in
cooperation with William E. Gray, Director of Environmental Affairs, Uranium
Mining and Refining. Between 1955 and 1978 this mill extracted uranium using
an acid leaching process with a throughput of 3500 tonnes per day of about
0.25% uranium ore. At present, the mill processes 5400 tonnes of ore per day.
The tailings are pumped to a retention area of about 8 x 105 m2. Before
reconstruction of the present dikes, tailings overflow were collected in catch
basins. These deposits are about 125 cm in depth. Between 1953 and 1956
Anaconda also operated a carbonate leach process. The tailings from the
carbonate process were separated from those of the acid leach process. Figure
1 shows the location of these tailings areas; they are designated ACID (inactive
acid tailings), ALKO (inactive carbonate tailings) and MAIN (presently active
tailings). These inactive tailings were covered by an average of about 85 cm
of clay during the fall of 1977.
Radon concentration in the air was continuously measured at three stations
(#102, #103 and #104) on the Anaconda mill site, (Figure 1) and at two stations
(Elks and Bride) located about 25 km east and west of the mill. Design and
calibration of the continuous radon and working level (CRWM) monitors were
previously reported (Mo78, Mo79c). Radon concentration was also measured based
on integrating air sampling techniques (Si69).
Radon flux, i.e. the amount of radon-222 that is transported across a
unit area of the surface per unit time, was measured using the accumulation
method, charcoal cannister, and ANL continuous radon flux monitor (Mo79c). In
this report, flux measured only by the accumulation method is reported.
Accumulation method for measurement of flux has been used by Kraner et al
(1964), Wilkening (1972), Bernhardt (1975) and Clements (1978).
The accumulation of radon, Q, (pCi) in a collecting device placed directly
over a surface area, A, is related to the radon flux, $, for the collection
period, At:
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309
Q = A • $ • At (1)
assuming (a) that At « radon-222 half-life of 3.92 days, (b) that back
diffusion of radon is low, and (c) that the accumulator does not disturb the
exhalation process. In general, criteria (a) and (b) can be met for a At of
less than one hour, but satisfaction of the criteria (c) is not yet proven.
Radon flux is dependent on wind velocity over the surface and the accumulator
will limit the surface wind over the area.
The accumulated radon, Q, was measured by determining the accumulator
radon concentration, Xg = Q/V, where V is the accumulator volume of approx-
imately 100 1. Unless $ is very large, i.e., xg >> Xa> the concentration of
radon in air x trapped within the accumulator must be measured. In this
study two techniques for measuring Q were utilized. In the first technique,
the radon concentration in air, shortly before placing of the device on the
surface, and the radon concentration in the accumulator were measured. A
collection time, At, of 10 and 20 minutes over tailings and soil, respectively,
were used. In the second technique, air in the accumulator was sampled at 10,
20 and 30 minutes after placing the accumulator on the gound. The initial
concentration of radon in air, Xa» was estimated from extrapolation of the
radon concentration in the accumulator to zero collection time. The flux was
calculated from:
(Xe - X,) ' h
$ = —§ §- (2)
At
where, h and A are the height of the accumulator (65 cm) and the area of the
collector surface (0.16 m2), respectively (h = V/A). Statistical comparison
of flux values showed that the two methods gave similar results.
Samples of radon in the ambient air and from inside the collector were
obtained by means of an in-line filter and desicant using evaculated 0.4-
and 1.4- liter ZnS scintillation cells (Eberline), respectively. After an
elapsed time of about 4 hours, the radon concentrations were measured using
an Eberline SAC-R5 alpha scintillation counter. Each cell was cross calibra-
ted using 100-cm3 Lucas cells (Johnston Laboratories) and an NBS radium solu-
tion (Lucas 1957, 1977).
Flux from the uncovered surface of the two inactive tailings (ALKO and
ACID, Fig. 1) was measured between August, 1977, and October, 1978. On each
pile region an area of about 1000 m2 was selected for study. Within this
area each tailings region was cored at several locations to a depth of approx-
imately 1.5 m, and samples were collected at approximately 15 cm intervals.
Selected samples were analyzed for Ra-226 concentration and particle size
distribution.
Within each of the tailings region, test areas were randomly selected.
Radon flux at the selected test areas was measured over several weeks. One
area was selected for measurement of flux attenuation through native soil
cover (ACIDS and ALK05). Native soil near Anaconda is mostly weathered fine
grained silty sandstone and mudstone with patchy presence of evaporities and
limestone. The regional soil is composed of montmorillonite clay, kaolinite,
quartz and unidentified amorphous material (Fu77). The test plots were
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310
covered initially with soil to a depth of about 7.5 cm and compacted. Radon
flux was measured after a waiting period of several days. Presently, each
test tailings plot had been covered to a depth of about 225 cm.
Experimental Results
Average radon-222 background was 0.3 pCi/1, ranging from 0.2 to 2.5 pCi/1.
Radon concentrations were measured using Continuous Radon Monitor (CRM) and
integrated air sampling method (SI69). Net radon concentration (i.e. radon
attributable to the mill operation) measured at stations #102, #103 and #104
were, respectively, 0.4, 1.2 and 0.4 pCi/1. The average of radon concentrations
measured at six locations on the MAIN tailings was 8.6 pCi/1 (6 to 15 pCi/1 in
range).
Radon flux is influenced by atmospheric pressure (C174), wind speed (Kr64),
stability, and climatic factors such as soil moisture and temperature. Our
flux measurements using Continuous Radon Flux Monitor (CRFM) will be reported
elsewhere (Mo79d). In order to minimize diurnal effects, radon concentration
in air on ACID and ALKO was measured between 10 a.m. and 1 p.m.
Normalized flux, i.e. the ratio (covered tailings/control tailings) was
assumed to be independent of meteorological variables. Therefore, radon flux
at both covered tailings (ACIDS and ALK05) and control tailing sites (ACID2 and
ALK02) were measured at the same time, The average of 49 measurements of radon
concentration in air during the measurements of radon flux on each area was
2.7 pCi/1 (0.4 to 12.2 pCi/1 in range) on ACID and 4.4 pCi/1 (0.1 to 12.7 pCi/1
in range) on ALKO.
Radon flux, $, measured at stations #100, #101 and #105 (Fig. 1) was
between 0.4 and 1.5 (pCi/m2*sec). Ra-226 concentration of soil was measured
using high-resolution gamma spectroscopy (Ge detector). Radium concentration
within the 10 cm of soil surface was in the range of 0.7 to 2.5 pCi Ra-226/g.
Additional measurements of radon flux and radium concentration in the soil
distant from the mill site are in progress.
The average radium concentration in cores through the ACID tailing was
616.6 pCi Ra-226/g with a range of 220 to 1800 pCi Ra-226/g. The average
radium concentration in ALKO was 601.4 pCi Ra-226/g with a range of 70 to 1000
pCi Ra-226/g. Radium concentration did not indicate a distribution pattern with
depth. Less than 50% of tailings by weight were 200 ym and 125 ym in diameter,
respectively for ACID and ALKO. In comparison these sizes are smaller than
those from the MAIN tailings (= 400 ym). Average core moisture in both test
tailings was about 10%. The moisture in the MAIN tailings beach is from a dry
0.2% to a complete saturation.
The average of 49 radon flux measurements (Table 1) made over each control
tailings (ACID2, ALK02) between August, 1977 and October, 1978 was 4 15 6 and
174.2 (pCi Rn-222/m2-sec). The average flux for ACID and ALKO was 376 8 and
190.0 (pCi Rn-222/mz-sec), respectively. The range of flux was between 59.7 to
1103 (pCi Rn-222/m^-sec) for ACID and 45.1 to 762 (pCi Rn-222/m2-sec) for ALKO
(Table 1).
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311
Figures 2 and 3 show the measured radon flux through the soil cover. The
particle size of the soil covering was 50% (by weight) less than 300 ym in
diameter and with a moisture content of less than 5% during deposition. Soil
moisture was higher at the point of contact with the tailings but in general
it remained at about 5% near the surface.
Estimation of Diffusion Coefficients
Radon flux from the surface of a single, homogeneous layer of material is
a function of radium concentration, R, emanating power of the tailings, e,
(fraction of the radon generated which is released into the diffusion space),
and bulk diffusion coefficient, Dt, given by:
tanh
(
x 104 (3)
where p is the tailings density, A is the radioactive decay constant of radon,
pt is the porosity of the tailings, and z is the depth of the tailings deposit.
Table 1 gives the average measured flux and diffusion coefficient Dfc calcu-
lated by using the following parameters: e = 0.25, Pfc = 0.3, X = 2.1 x 1Q~6 sec""1
z = 125 cm, p = 1.6 g/cm3 and R = 617 and 601 pCi Ra-226/g for ACID and ALKO,
respectively. Emanation rates from domestic uranium ores vary only slightly
with ore moisture between 10% and 80% of saturation. Emanating power for
domestic uranium ores is between 0.01 - 0.9 with an average of about 0.25
(Au75, Me74). For uranium tailings e = 0.23 (Cu73). In these measurements
average specific flux, $ (pCi Rn-222/m2«sec) / (pCi Ra-226/g) is 0.61 and 0.67
for ACID and 0.32 and 0.29 for ALKO. The lower specific flux and diffusion
coefficient at the ALKO pile may be partially due to the smaller tailings size
relative to the ACID pile. Based on these parameters, a 100-cm tailing depth
is effectively an infinite depth (i.e. more than 95% of flux from infinite
depth) for radon diffusion and the values given in Table 1 are effectively
<(>„. Specific flux <|> measured (Si69) for loose sediments rich in clay is
= 0.37, for sandy soils <|> = 0.18, and for an average of clays and heavy
loams <|> = 0.28 (pCi Rn-222/m2'sec) / (pCi Ra-226/g). The specific flux for
abandoned Vitro tailings has been reported to be 1.6 (pCi Rn-222/m2-sec) /
(pCi Ra-226/g), (Sc74). In our radiological analysis of the Bear Creek Project
(NRC77) an average specific flux of a, = 1 (pCi Rn-222/m2-sec) / (pCi Ra-226/g)
for tailings containing both dry and wet beaches was utilized.
The radon flux from tailings which escapes from the surface of soil cover,
$c, is dependent on the same parameters given for the Equation (3). The
emanation factor e measured for various soil conditions is reported to be
0.14 to 0.29 (Si69). In this study a value of e = 0.25 for soil cover was
used. Radon flux through a soil cover is given by:
$ «) = $ (£ = 0) f(O (4)
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312
where £ is the depth of the soil cover over tailings which release flux
= 0) = $t. The attenuation function f (5) is:
«o
where k = rc , h =/ rt , and T = Tanh (h • z)
-1
(5)
The above equation may be further simplified by substituting 6 =^ — ]-
into the Equation 5 which then reduces to:
f (?) = 2e~k5 (1 + 6kT) + (1 - 6kT) e"2k5 (6)
and assuming Pt = PC. Equation (6) was fitted to the measured flux using the
Marquardt (Ma63) method. Figures 2 and 3 as a function of soil cover 5 give
the average measured flux and predicted flux using the Equation 6. The radon
diffusion coefficients through the soil cover Dc is 3.69 x 10~3 and
3.60 x io~3 (cm2/sec) for the data shown in Figures 2 and 3. The number of
radon measurements through the soil cover was 47 for each of the tailings;
they were spread over 387 days.
Laboratory measurements of radon diffusion coefficients for soil covers
obtained from the Powder River Basin, Shirley Basin, and Ambrosia Lake, under
a porosity of 0.43 to 0.60; a moisture content of 1% to 11%; and a compaction
of 65% to 89% gave values between 3.9 x io~3 cm2/sec and 3.2 x 1Q~2 cm2/sec
(Ro79). The effect of cover moisture is a reduction of the diffusion coef-
ficient. Diffusion coefficients for soil obtained from Powder River Basin,
similar to the soils from the Bear Creek Project, are 1.8 x 10~2, 1.6 x IQ"4
and 2.1 x 1Q~5 cma/sec, respectively, at 5%, 17% and 30% moisture (Ro79).
Since the soil cover contains Ra-226, the contribution to radon flux from
thick soil covers should be considered. A composite soil sample had an average
of 1.5 pCi Ra-226/g. Figures 2 and 3 show flux after application of this
correction using a finite difference multilayer simulation code. Range of
radon flux for background is also shown in Figures 2 and 3. The depth of
cover needed to achieve a reduction of radon flux to twice the average back-
ground (0.75 pCi Rn-222/m2-sec) is about 450 cm based on the flux corrected
for the cover Ra-226 activity. But the depth of cover is only 400 cm using
only the predicted flux, i.e. that not corrected for the Rn-226 in the soil.
TAILING MANAGEMENT
A major objective of tailings management during both milling-operation
and after decommissioning is the mitigation of radiation exposure by short-
term impoundment and long-term containment. This objective may be accomp-
lished by: (a) reduction or elimination of airborne radioactive releases
(b) reduction or elimination of contamination of surface and groundwater,'
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313
(c) insuring long-term stability and isolation of the tailings (Ma78, Sc77).
In addition, in order to minimize the potential hazards associated with mill
wastes, a mill site which is as remote as possible from both human population
and local streams should be selected.
Practical alternatives for the management of tailings (NRC77) under condi-
tions which satisfy the "as low as reasonably achievable" criteria, are
variations of two methods, surface and below grade disposal. In each case
disposal areas are designed to minimize both vertical and lateral seepage of
tailings solution. An example of surface storage of tailings is the Bear Creek
Project (NRC77) which was built using technologies available at the time. This
mill has utilized the natural valley contour to impound the tailings. Seepage
through the tailings dam foundation was controlled by excavating a cut-off
trench to the top of the bedrock beneath the center of the dam and backfilling
with impervious material. The dam itself contains a central impervious core.
The surface of the tailings retention area (about 6 * 105 m2) was covered with
compacted clay to a minimum depth of 1 meter. The retention area thus created
is capable of storing about 6 x 1Q6 tonnes of solids.
Tailings can be disposed below grade in areas such as open pit mines or
specially prepared pits. This type of disposal, when available and conditions
permit, allows a deep cover to be applied without subjecting its surface to
excessive erosive forces. An example of such disposal is the Morton Ranch
facility (NRC78).
Active Tailings; Particulates
The tailings beach area (the surface which is not covered by tailings
solution) may comprise from 10% to 90% of the retention area depending on
the annual evaporation rate, rate of seepage and the recycling of the tailings
solution. The quantity of tailings material released from the beach as
fugitive dust is dependent on wind velocity, surface, moisture and the physical
protection of the surface. Upon aging fine tailings tend to conglomerate on
larger grains (Ca65). In this condition and otherwise undisturbed, they are
relatively resistant to erosion by wind.
Based on the physical characteristics given in Table 2 and meteoro-
logical parameters, typical of the southwestern United States, listed in
Table 3 the rate at which tailings particulates are transported by wind was
estimated using the UDAD code (Mo79). The specific activity for particles
less than or equal to 35 ym in diameter was assumed to be 62.5 pCi/g for
U-238 and U-234, and 1250 pCi/g for Th-230, Ra-226, Pb-210 and Po-210. This
is equivalent to assuming an average specific activity for the uranium ore
of 500 pCi/g and a specific activity for the small tailings particles that
is 2.5 times greater than the average. Figure 4 gives the total Ra-226
concentration of airborne radioactivity as a function of distance from the
geometric center of the tailing beach (0.3 km2 in area) in the direction of
maximum transport (northeast). These concentrations include resuspension of
Ra-226 from the ground after accumulation from 20 years of mill operation.
Because the concentrations of Th-230, Pb-210, and Po-210 in the tailings are
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314
about the same as for Ra-226, the concentrations given in Figure 4 are
equally applicable to these radionuclides, except for Pb-210. As shown in
Figure 5, the concentration of lead in the air begins to increase beyond about
1 km from the pile because of the contribution from decay of radon daughter
products in the air. For example, at 10 km in the base case the total Pb-210
is 3. x IQ-4 pCi/m3 in comparison with about 1.5 * 10~6 pCi/m3 released
directly from the tailings surface. Since the specific activity of uranium
in the tailings was assumed to be about 5% of that for Ra-226, the resulting
airborne concentration of uranium is 5% of the values given in Figure 4.
The amount of dust released and, therefore, the concentration of air-
borne particulates is dependent on the moisture content of the tailings
surface. As Figure 4 indicates, an increase in surface moisture (not
necessarily the moisture content of the bulk tailing) from a dry, 0.2%
moisture in the base case (a typical condition for a dry beach in a south-
western climate) to a moist (10%) and wet (30%) condition results in a marked
decrease in airborne concentration. At 1 km from geometric center of the
tailings beach, the predicted average concentration of Ra-226 in air is
about 9 x HP4, 3 x 1Q~6 and 1.3 x 1Q~6 pCi Ra-226/m3, respectively, for the
base case, moist and wet tailings surface. Concentration of radium in air
for 1% and 5% moisture is 2.04 x 10~5 and 5.35 x 10~6 (pCi Ra-226/m3).
respectively. This results in a reduction by factors of about 4, 17, 30 and
70 times in airborne Ra-226 concentration with an increase in moisture from
0.2% to 1%, 5%, 10% and 30%, respectively. The 30% moisture is about the
saturation condition of the tailings. The concentration predicted for 10%
and 30% moistures are conservative and for 30% moisture the rate of tailings
transported may be practically zero. The predicted concentrations for
moistures exceeding 10% are only an order of magnitude estimates.
The accumulation of ground contamination as a result of the deposition
of airborne radioactivity during the 20 years of mill operation also was
estimated using the UDAD code and the data from Tables 2 and 3. The results
of the calculation of Ra-226 surface soil concentration as a function of
distance from the pile are shown in Figure 6 by the lower of the two curves
for each case. The dependence of the total fallout on the surface moisture
content of the beaches is clear from this figure. This dependence is not
surprising since the deposition rate is directly proportional to the concen-
tration of the radionuclide in the air, and the influence of surface moisture
on the air concentration was shown in Figures 4 and 5. Even for the base case
with dry beaches, the buildup of radium contamination on the ground, as pre-
dicted by the UDAD code and presented in Figure 6, is not large compared to
background levels if the fallout is assumed to mix through the upper layer of
soil. The average concentration of Ra-226 in soil is about 1.5 pCi/g, so a
soil layer to plow depth of 25 cm contains 6 x 1Q5 pCi/m2. Figure 6 shows
that at 1 km from the center of the pile, the Ra-226 concentration in the plow
layer would be increased about 0.3% by deposition of tailings particulates.
In addition to the vertical flux of dust particles which accounts for
the ground concentrations calculated by the UDAD code, a horizontal flux of
relatively large particles, similar to desert sand, can contribute signifi-
cantly to soil contamination under conditions of high wind. With winds in
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315
excess of 19 knots tailings particles as large as 1000 ym in diameter are
transported. A total horizontal flux exceeding 5 kg/m.hr has been measured
(mo79). Fortunately, winds of sufficient velocity to generate such a large
horizontal flux are infrequent, but during a severe wind storm the Ra-226
concentration near a tailings pile could be increased by more than an order
of magnitude in a short period. In Figure 6 the cross-hatched area between
the two curves in each case is a qualitative estimate of surface contamination
from the horizontal flux of particles.
Dose commitments to bone and lung from inhalation of Th-230, Ra-226,
Pb-210 at a selected distance and in the direction of maximum transport of
tailings from the dry beaches (the base case) were computed using UDAD code
and are listed in Table 4. Dose commitments for the moist and wet surface
conditions are about the same ratio as the 226Ra concentrations in air for
base case/moist/wet, viz. 1/30/70. External gamma exposures from the
ground contamination due to fallout (not including horizontal flux) of the
airborne tailings and from airborne radionuclides are given in Table 5.
The dose commitment to any organ from inhalation plus external exposures
from tailings particulates under base case conditions is less than the 25
mrem/year standard published by EPA (40CFR190), assuming a controlled boundary
to exclude the general population at 1 km from the tailings center. In
these analyses, the contribution of ingestion exposure was not included.
The ingestion pathway may contribute equal to the combined dose commitment
from inhalation and external gamma exposure.
Active Tailings; Radon
Radon flux, $, from the surface of the tailings was assumed to be 500
pCi 222Rn/m2.sec (dry tailings), 50 pCi Rn-222/m2.sec (moist) and 10 pCi
Rn-222/m2.sec (wet). These correspond to specific fluxs of $ = 1, 0.1 and
0.02 (pCi Rn-222/m2.sec) / (pCi Ra-226/g). The diffusion coefficient of
radon is exponentially dependent on the moisture content. An increase in
moisture from a base case dry condition to saturation results in a decrease
of about 10-lt in the diffusion coefficient and a decrease by about 10~2 in
specific flux (Ro79).
Figure 7 gives radon concentrations in the direction of maximum tailings
transport for the three surface moisture conditions. Concentrations of
radon-222 at 0.1 km (Fig. 7), downwind from the geometric center of the tail-
ings beaches are 10.3 pCi Rn-222/1 and at 1 km only 0.92 pCi Rn-222/1. At 10-
km distance radon concentration decreases to 0.03 pCi/1. Corresponding working
levels are estimated to be 7.4 x 10" 3 at the center of the tailings, decreasing
to 2.7 x 10~3 at 1 km and to 2.5 x 10"1* at 10 km. Measurements at several
inactive tailings piles have yielded average airborne radon concentrations,
three feet above the surface and directly over the tailings, ranging from 3.5
to 19 pCi/1 (3.5 x 103 - 1.9 x lO4 pCi/m3) (US-DH69). Our average of measure-
ments at the six locations on the MAIN tailings was 8.6 pCi/1 (6 to 15 pCi/1
in range). Radon concentration predicted for the base case is 10.3 pCi/1 on
the tailings (Fig. 7). A detailed comparison of theoretical predictions and
experimental data has been previously reported (Mo78).
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316
Dose commitments from inhalation of Rn-222 and its progeny are given
in Table 6, for the base case, moist and wet tailings surfaces, assuming
an average dose conversion factor of 0.625 (mrem/year) / (pCi Rn-222/m )
for combined exposure to inside and outside radon concentrations. The
dose from radon progeny exceeds by far the dose from the long-lived par-
ticulates. A comparison of the dose commitments given in Table 6 indicates
that even under wet conditions and at a distance of 1 km, a dose commitment
of 11.5 mrem/year to the bronchial epithelium is expected. But, at present,
the only radiation limit for exposure of the general public is a maximum
permissible radon concentration (MPC) of 3 pCi/1. Figure 8 gives the iso-
pleths for the safety factor (SF) defined as follows:
SF = 1 ' concentrat:i-on of Rn-222 in air |
I maximum permissible concentrationJ
At 2 km northeast, Figure 8 shows that the safety factor is about SF = -3,
corresponding to a radon concentration of 1/1000 of the maximum permissible
concentration. Under the base case condition a boundary 0.6 km from the
tailings would satisfy the criteria that SF = 0, while with moist beaches
the radon concentration right on the surface of tailings is less than the
3 pCi/1 MPC.
Mill Tailings Decommissioning
The protocol for decommissioning uranium mill tailings should be based
both on minimization of exposure from all pathways and on insuring the
containment of the tailings throughout a predictable future. Tailings
decommissioning protocol should consider the transition period from active
tailings storage to completion of drying prior to isolation of the tail-
ings. Because the rate of release of radioactivity from dry tailings is
relatively high, the environmental impact produced during the interim
drying-out period, if it is extended for a long time, could exceed the
impacts of the operating phase when the tailings were kept moist. "Pro-
gressive decommissioning" of tailings pile by covering with overburden
material more or less continuously in parallel with drying operations is
both technically practical and financially feasible.
Viable decommissioning protocols under present technological and com-
petitive conditions are isolation of tailings within a soil lens of clay
material and burying beneath overburden from the mining operation and a
combination of vegetation and rip-rap if required to reduce erosion.
Examples of decommissioning plans are given in Chapter 10 of the Bear
Creek Project environmental statement (NRC77). The suggested decommission-
ing protocol requires that the tailings solids be placed in a clay lined
impoundment and covered with 25 cm of compacted montmorillonite clay thereby
enclosing the tailing in an artificial clay lens. Further, the clay cap
is to be covered with an additional 180 cm of overburden and top soil.
Placement of the clay cap, compacted to engineering specifications of'the
Bear Creek model, requires initial covering of the tailings with overburden
to support machinery. Thus, the total cover may exceed 300 cm at many loca-
tions. The minimum thickness specified by the Bear Creek model, will reduce
the radon flux and gamma radiation to twice the normal background level
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317
Soil utilized for ACID and ALKO (Figs. 2 and 3) contains silt and
sand in addition to clay and not compacted to Bear Creek specifications.
Even so, 400 cm of this soil would reduce the radon flux to the background
level. This depth of cover is explicitly for the measured conditions
(Table 1) and may be less or more for other tailings depending on the
radon flux from bare tailings and radium-226 activity of the soil cover.
In most of the southwestern United States, the rate of surface denuda-
tion is less than 15 cm per 1000 year, which is greater than the denudation
rate for other parts of the continental United States. At this rate a
soil cap of 300 cm protected from erosion by surface water run-off with
both rip-rap and vegetation will last for about 20,000 years. Even so,
the cover of these tailings will need occasional reconstruction to remedy
the local effects of water erosion. In perspective, this period is long
in comparison to predictable geological conditions and concern for the in-
tegrity of small scale features seems misplaced. For example, less than
10,000 years ago glaciers covered most the north-middle-western states and
was responsible for creating lakes and changing the direction of rivers
such as the Ohio River.
An alternative for tailings management (NRC78) after decommissioning
is below-grade containment, such as planned for Morton Ranch. Below-grade
tailings disposal, if it is hydrologically feasible, allows deeper burial
of the tailings. Subsurface disposal of tailings results in a configuration
which is stable and less prone to surface erosion than above-grade disposal.
The choice of either model, Bear Creek or Morton Ranch, is dependent on
the site-specific hydrological conditions. Choice of overburden contain-
ing substantial Ra-226 activity (> 5 pCi/g) for the top cover of tailings
may itself generate a substantial radon flux. Adoption of either models
for decommissioning of tailings results in mitigation of future exposures
from inhalation, ingestion, and external irradiation.
Alternative methods for disposal of tailings have been discussed in
Chapter 10 of the Bear Creek Project environment statement (NRC77) including
removal of radioactivity during processing of the ore. At present, more
advanced alternative disposal methods such as this are not technologically
and financially feasible because the large volume of low level activity.
CONCLUSION
For some of the older mills, the dose commitment from inhalation of
airborne tailings particulates, ingestion of contaminated food and external
exposure to individuals living nearby may exceed the 25 mrem/year limit.
Also, because of fallout of airborne tailings particulates and horizontal
transport of tailings, sand, a large area of uncontrolled land in the imme-
diate vicinity may be contaminated to an extent which is unacceptable.
Since expansion of the land controlled by the mill is often not practical
or possible, consideration of alternative procedures for management of
uranium tailings is necessary both to achieve compliance with the 25 mrem/year
standard and to protect the environment, the mobility of the tailings sands
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318
and fugitive dust should be reduced. A suggested approach is to minimize
dried beach tailings surfaces and/or maintain a wet surface condition.
The release rates from tailings completely submersed under solution is
negligible, but for some mills, submersion is not possible because of
the impact of the large surface head area on the groundwater quality.
Fig. 9 depicts a suggested alternative procedure for maximizing evapora-
tion and reducing the release rates of both radon and particulates. The
multiple discharge of mill effluents minimizes the formation of large,
dry deltas and maintains almost the entire surface of the tailings in a
wet to moist condition. Recycling of solution over the surface in some
cases may not result in a sufficient rate of volume reduction and may re-
quire the addition of auxilliary evaporation ponds as depicted in Fig. 9.
In other cases, since essentially the entire surface of the main pond is
utilized for evaporation, a pond of smaller area but greater depth may
become practical. This in turn would reduce the cost of final stabiliza-
tion of the tailings.
Gravitational settling encourages deposition of the larger sand par-
ticles of lower specific activity near the discharge points resulting in
relatively greater deposition of finer particles (slimes) with higher
specific activity at deeper depths below the tailings solution. This
tends to reduce the radioactivity in the more exposed and drier tailings,
and therefore, decreases the dose commitments which result from transport
of this material. Mechanical separation of sands and slimes prior to dis-
charge is suggested as a useful alternative procedure to augment this
natural, differential settling process. Concentration of the slimes in
the liquid discharge also reduces solution seepage by closing the channels
for leakage.
Above ground storage of tailings after decommissioning in a configura-
tion similar to that depicted in Fig. 9 will require periodic maintenance
to insure that they remain isolated from the environment. For this reason,
permanent disposal above grade is not recommended if any practical alter-
natives exist. Below ground level storage will provide more protection
from erosion forces and reduce maintenance and, therefore, is the preferred
method. In light of higher uranium market prices, reprocessing of some of
the older tailings may provide an incentive for reconstruction of tailings
retention areas in accord with the regulatory objectives listed by Martin
and Miller (Ma78). Application of a minimum of 4 m of cover as a mixture
of overburden, clay, soil and rip-rap over the tailings and protection of
the slopes at dyke boundaries from erosion may provide an acceptable above
ground alternative for storage of tailings if local conditions prevent
disposal below grade. Stability of this configuration is dependent on the
slope of the dykes, reduction of erosion and control of water run-off.
The radiation dose commitment from the isolated tailings should be compar-
able to that of natural background. Decommissioning of the tailings should
progress in parallel with drying of the ponds rather than being deferred
until the beaches are completely dry. Deposition of overburden on the
wet tailings beaches using presently available technology is both practical
and financially feasible. Any protocol for decommissioning of tailings
should include both minimization of exposures and assurance of predictable
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319
long-term stability of containment. Management of uranium tailings at all
new mills so that the hazard from radon and particulates is negligible
during active milling and after decommissioning is feasible and practical.
Further studies are needed, however, to provide reliable data for the
analysis of reclamation plans in order to insure that they are cost effec-
tive.
References
Au75 Austain S.R., 1975, "A Laboratory Study of Radon Emanation from Domestic Uranium Ores,"
in: Radon in Uranium Mining, p. 151-169, International Atomic Energy Agency, IAEA-PL-565/8.
Ba41 Bagnold R.A., 1941, "The Physics of Blown Sand and Desert Dunes," (London: Methuen and Co.).
Be64 Belley Pierre-Yves, 1964, "Sand Movement by Wind," Technical Memorandum, No.l, January,
U.S. Army Corps of Engineers.
Ca65 Capes C.E. and Danckwerts P.V., 1965, "Granule Formulation by the Agglomeration of Damp
Powders, Part 1: The Mechanism of Granule Growth," Trans. Instn. Chem. Engrs. 43, p. T116-T-124.
Ch45 Chepil W.S., 1945, "Dynamics of Wind Erosion, I: Nature of Movement of Soil by Wind,"
Soil Sci. 60, p. 305-320.
Ch45 Chepil W.S., 1945, "Initiation of Soil Movement," Soil Sci. 60, p. 397-411.
Ch39 Chepil W.S. and Milne R.A., 1939, "Comparative Study of Soil Drifting in the Field and in
Wind Tunnel," Sci. Agri. 19, p. 249-257.
Ch41 Chepil W.S. and Milne R.A., 1941, "Wind Erosion of Soil in Relation to Roughness of Surface,"
Soil Sci. 52, p. 417-431.
C178 Clements W.E., Barr S., and Marple M.L., 1978, "Uranium Mill Tailings Piles as Sources of
Atmospheric Radon," in: Natural Radiation Environment III (Edited by Adams J.S., Lowder W.M.,
and Gesell T.F.).
C174 Clements W.E. and Wilkening M.H., 1974, "Atmospheric Pressure Effects on Rn-222 Transport
Across the Earth-Air Interface," J. of Geo. Phys. Res. 79, p. 5025-5029.
Cu73 Culot M.V.J., Olson H.G., and Schiager K.J., 1973, "Radon Progeny Control in Buildings,"
COO-2273-1, Colorado State University.
Do75 Douglas R.L. and Hans J.M., 1975, "Gamma Radiation Surveys at Inactive Uranium Mill Sites,"
Technology note ORP/LV-75-5, U.S. Environmental Protection Agency, Las Vegas, Nevada.
Fu77 Fugro, Inc., 1977, "Pond Lining Study," Anaconda Milling Facilities near Grants, New Mexico.
Kr64 Kramer H.W., Schroeder G.L. and Evans R.D., 1964, "Measurement of the Effects of Atmos-
pheric Variables on Radon-222 Flux and Soil Gas Concentrations," in: Natural Radiation
Environment. p. 191-215 (Edited by Adams J.A.S. and Lowder W.M.) (The University of Chicago
Press, Chicago).
Ma63 Marquardt D.W., 1963, "An Algorithm for Least-Squares Estimation of Nonlinear Parameters,"
Journal of Soc. Indust. Appl. Math. 11, p. 431-441.
Ma78 Martin J.B. and Miller H.J., 1978, "Generic Environmental Impact Statement on U.S. Uranium
Milling Industry," in: Seminar on Management, Stabilization and Environmental Impact of
Uranium Mill Tailings. Nuclear Energy Agency (OECD), July 24th-28th.
Me74 Megumi K. and Mamuro T., 1974, "Emanation and Exhalation of Radon and Thoron Gases from
Soil Particles," J, Geophys. Res. 79. p. 3357-3360.
Mo79a Momeni M.H., Guill P., Kisieleski W., and Rayno D., "Radioactive Composition and Physical
Characteristics of Tailings," Argonne National Laboratory, Argonne, Illinois (to be published).
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320
Mo78 Momeni M.H., Kisieleski W.E., Yuan Y., and Roberts C.J., 1978, "Radiological and Environ-
mental Studies at Uranium Mills: A Comparison of Theoretical and Experimental Data," in:
Management, Stabilization and Environmental Impact of Uranium Mill Tailings, Proceedings of
the NEA Seminar, Organization for Economic Cooperation and Development (OECD-AEN).
Mo79b Momeni M.H., Yuan Y. and Zielen A.J., "Uranium Dispersion and Dosimetry Code, UDAD,"
Argonne National Laboratory, Argonne, Illinois (in press).
Mo79c Momeni M.H., Miranda J., Kisieleski W., and Kretz N., 1979, "Continuous Measurement of
Rn-222 Flux, Concentration and Working Level," Health Physics Society Annual Meeting,
Philadelphia, PA, July 8-13.
NRC77 Nuclear Regulatory Commission, 1977, "Bear Creek Project Final Environmental Statement,"
Appendix K, NUREG-0129.
NRC78 Nuclear Regulatory Commission, 1978, "Morton Ranch Project, Environmental Impact State-
ment," NUREG-0439.
Ro79 Rogers V., Overmeyer R.F., Jensen C.M. and Canon E., 1979, "Characterization of Uranium
Tailings Cover Materials for Radon Flux Reduction," prepared for Argonne National Laboratory
by Ford, Bacon & Davis, Utah, Inc. FBDU218-1.
Sc77 Scarano R.A., Martin J.B., and Magno P.J., 1977, "Current Uranium Mill Licensing Issues,"
presented at Atomic Industrial Forum, Inc. Fuel Cycle Conference, Kansas City, Missouri.
Sc78 Scarano R.A., Linehan J., 1978, "Current Nuclear Regulatory Commission Licensing Review
Progress: Uranium Mill Tailings Management," in: Management, Stabilization and Environ-
mental Impact of Uranium Mill Tailings. Proceedings of the NEA Seminar, Organization for
Economic Cooperation and Development (OECD-AEN).
Sc74 Schiager, K.T., 1974, "Analysis of Radiation Exposures on or New Uranium Mill Tailings
Piles," in: Rad. Data and Reports 14. p. 411.
Si69 Sill C., 1969, "An Integrating Air Sampler for Determination of Rn-222," Health Physics 16,
p. 371-377.
Si69 Sisigina T.I., 1969, "Assessment of Radon Emanation from the Surface of Extensive Terri-
tories," in: Nuclear Meteorology, a proceeding of the All. Union Conference on Nuclear
Meteorology, Obninsk (Edited by Makhon'ko E.P. and Malaphov S.G.) (translated UDC 551.510.71).
USDH69 U.S. Department of Health, 1969, Colorado State Health Agency and Utah State Health
Agency, Supt. Docs., U.S. Government Printing Office (Washington, D.C.: AEC) .
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TABLE 1. Flux <|>t(z) (pCi Rn-222/m2-sec) , Specific Flux
(pCi Rn-222/m2-sec)/(pCi Ra-226/g), Bulk Diffusion
Coefficient Dt (cm2/sec) at Two Tailings Designated
ACID and ALKO for Number of Measurements N
Tailing Area N <|>t(z)
ACID
ACID.
ACID.
ALKO
ALKO.
ALKO.
2
5
2
5
55
49
6
55
49
6
376.
415.
169.
190.
174.
175.
8
6
0
0
2
1
0.
0.
0.
0.
61
67
*
32
29
*
2
3
4
5
4
5
.4
.0
.5
.7
.8
.0
X
X
X
X
X
X
10
10
10
10
10
10
-3
-3
-k
-1+
-it
-4
*ACID.5 and ALKO.5 are covered with soil and Ra-226 concentration has not been
determined yet.
TABLE 2. Physical Characteristics of Tailings Solids
Particle
Diameter (ym)
Deposition
Velocity (m/sec)
Settling
Velocity (m/sec)
Fractional
Distribution
0
1
5
35
>100
.1
.0
.0
.0
.
Radon release rate:
Density
Tailing
of tailings:
area : *
1.0
1.0
1.0
8.8
7.2
4700
2.4
3 x
x 10
x 10
x 10
x 10
x 10
-2
-2
—2
-2
-1
7
7
1.
8.
7
.2
.2
80
82
.2
x
X
X
X
X
10
10
10
10
10
-7
-5
-3
-2
-1
0.
0.
0.
0.
0.
02
03
25
50
20
Ci/year
g/cm3
ID'1
km2
*This area was subdivided into 16 areas, each with an area of 1.88 x 10~2 km2,
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322
TABLE 3. Annual Relative Frequency of Occurrence (Sum of All Stability
Classes) Metset Bluewater, New Mexico
Wind Speed, knots
Wind Direction
N
NNE
NE
ENE
E
ESE
SE
SSE
S
SSW
SW
WSW
W
WNW
NW
NNW
Column totals:
0-3
0.01690
0.01390
0.02130
0.02100
0.02060
0.02780
0.03320
0.02740
0.02910
0.03010
0.04950
0.03150
0.02860
0.02850
0.02740
0.01540
0.42220
4-6
0.01560
0.01260
0.01990
0.02050
0.02080
0.02710
0.03430
0.02640
0.02970
0.02910
0.04670
0.02790
0.04740
0.02770
0.02880
0.01410
0.40860
7-10
0.00510
0.00390
0.00630
0.00620
0.00620
0.00880
0.00910
0.00730
0.00310
0.00760
0.01280
0.00710
0.00610
0.00610
0.00630
0.00430
0.11130
11-16
0.00150
0.00100
0.00220
0.00220
0.00220
0.00310
0.00250
0.00300
0.00310
0.00310
0.00490
0.00280
0.00270
0.00260
0.00260
0.00140
0.04090
17-21
0.00060
0.00040
0.00070
0.00050
0.00050
0.00070
0.00050
0.00070
0.00060
0.00060
0.00070
0.00030
0.00080
0.00070
0.00070
0.00040
0.00940
Over 21
0.00040
0.00030
0.00060
0.00040
0.00050
0.00040
0.00040
0.00050
0.00060
0.00060
0.00040
0.00020
0.00050
0.00060
0.00060
0.00040
0.00740
Row To
0.040!
0.032:
0.0511
0.050!
0.050!
0.067!
0.080(
0.065:
0.071;
0.071]
0.115C
0.069J
0.066]
0.0662
0.0664
0.036C
0.9997
TABLE 4. Total Inhalation Dose Commitment (mrem/year) from
Particulates at the Direction of Maximum
Dispersion and Base Case Condition
Distance (km)
0.3
1.0
5.0
10.0
50.0
80.0
Whole-body
2.26E-01
1.33E-02
2.61E-03
2.46E-03
2.18E-03
2.06E-03
Nasopharyngeal
4.32E+01
2 . 31E+00
1.37E-01
4.70E-02
1.08E-02
9.13E-03
Tracheobronchial
9.17E-03
5.56E-04
9.11E-05
7.99E-03
6.86E-05
6.48E-05
Pulmonary
1.86E+00
1.24E-01
3.30E-02
3.29E-02
2.99E-02
2.83E-02
Bone
6.07E-*
3.64E-
7.82E-
7 . 51E-
6.73E-
6.37E-
-------
323
TABLE 5. Total External* Dose Commitment (mrem/year) from
Ground Contamination and Airborne Radionuclides
Dose
(mrem/year)
Whole Body
Ground
Airborne
Ovaries
Ground
Airborne
Testes
Ground
Airborne
Bone Marrow
Ground
Airborne
Bone
Ground
Airborne
Distance (km)
0.1
9.26E+00
6.58E-01
6.85E+00
3.64E-01
7.95E+00
6.87E-01
9.85E+00
7.85E-01
1.07E+01
8.38E-01
1.0
2.74E-01
1.05E-00
2.03E-01
7.11E-01
2.36E-01
9.67E-01
2.92E-01
1.16E+00
3.18E-01
1.24E+00
5.0
1.23E-02
5.00E-01
9.04E-03
3.69E-01
1.06E-02
4.27E-01
1.33E-02
5.25E-01
1.45E-02
5.64E-01
10
3.65E-03
2.35E-01
2 . 60E-03
1.74E-01
3.13E-03
1.99E-01
4.13E-03
2.46E-01
4.51E-03
2.64E-01
50
6.09E-04
3.53E-02
3.49E-04
2.63E-02
5.17E-04
2.99E-02
8.90E-04
3.69E-02
9.74E-04
3.96E-02
80
5.05E-04
1.97E-02
2.78E-04
1.46E-02
4.28E-04
1.66E-02
7.67E-04
2.05E-02
8.40E-04
2.21E-02
*Direct gamma and beta radiation from the tailings and contamination from
creeping sand tailings are not included in these calculations. On tailings
pile direct radiation dose to whole body is 1.0 x 101* to 1.5 * 101* mrem/year.
But direct radiation at 0.5 km from tailings is small relative to background.
TABLE 6. Dose Commitments (mrem/year) from Inhalation of
Rn-222 and Radon Daughters to Bronchial Epithelium
under Conditions of Dry Tailings, Moist and Wet
Beaches as a Function of Distance from the Tailings
and in the Direction of Maximum Dispersion
Distance (km)
0.1
1.0
5.0
10.0
20.0
50.0
80.0
Dry
6.41E+03
5.74E+02
4.49E+01
1.73E+01
7.13E+01
2.33E+00
1 . 30E+00
Moist
6.41E+02
5.74E+01
4.49E+00
1.73E+00
7.13E-01
2.33E-01
1.39E-01
Wet
1.28E+02
1.15E+01
8.98E-01
3.46E-01
1.43E-01
4.66E-02
2.60E-02
-------
\
1A
•B
P
14.0
1.0 2.0 3.0 4.0 5.0 6.0 7.0 8.0 9.0 10.0 11.0 12.0 13.0
488 METERS PER DIVISION
1. Experimental stations at Anaconda Uranium Mill, Bluewater,
New Mexico. Areas designated by letters are A (Station
#101, #102), B (Station #103), C (Station #104), D (Sta-
tion #105), and E (Station #100). Inactive Tailings are
ACID and ALKO. The tailings retention area still active
is designated as MAIN.
-------
325
• CORRECTED FOR 226Ro IN SOIL COVER
* AVERAGE MEASURED FLUX
PREDICTED FLUX
2.
300 400 500
DEPTH OF SOIL COVER, cm
Measured flux over ACID tailings (see Figure 1), predicted flux,
and flux corrected for the radioactivity content of soil cover,
as a function of soil cover thickness.
1000
700
500
ALKO TAILINGS
• CORRECTED FOR 226Ra IN SOIL COVER
* AVERAGE MEASURED FLUX
PREDICTED FLUX
100
200 300 400 500
DEPTH OF SOIL COVER, cm
600
700
3. Measured flux over ALKO tailings (see Figure 1), predicted flux,
and flux corrected for the radioactivity content of soil cover,
as a function of soil cover thickness.
-------
TAILING MANAGEMENT
TAILING MANAGEMENT
a
a
o 2
8 •
0
oc.
-------
TAILING MANAGEMENT
TAILING MANAGEMENT
7
20th y«or of operation
at maximum dl*p*r«lon
LEGEND
a • Bos* ca«*
O m Mo let b«och««
A - U«t b«ach««
-\
• I
• I*
Dl«tonc* CkrtO
20th y*ar of operation
at maximum di*p«r«ion
LEGEND
O • BOB* ea««
O • Mol*t b*och«*
A - W«t b«och«*
-I
!•
te
te
Dl»tanc«
2»-UAN-70
Ra-226 contamination of ground as a result of 20 years of mill
operation as a function of distance from the geometric center
of beach tailings and in the direction of maximum tailings
transport, predicted for three surface moisture conditions:
base case (dry), moist and wet beaches. The cross-hatched area
represents the estimated surface activity from horizontal tail-
ings flux. The symbols superimposed on the curves do not repre-
sent experimental data.
Concentration of Rn-222 in air as a function of distance from the
geometric center of beach tailings predicted for the direction of
maximum transport and for three surface moisture conditions (dry),
moist, wet). The sumbols superimposed on the curves do not repre-
sent experimental data.
Co
r\>
-------
Dry Beach: Lung | |
Log (concentration of radon/MPC)
8. Safety factor isopleth, i.e., distances from
the geometric center of the beach tailings
having equal concentration of Rn-222 in air
normalized to maximum permissible concentra-
tion (MFC) of radon (3 pCi/1). A safety
factor of -3 indicates that the concentra-
tion of radon-222 in air is only 1/1000 of
the MFC.
9. Sketch of a suggested design for minimizing
fugitive dust and radon-222 emission from
the surface of a tailings pond. Features
include multiple ports for discharging mill
effluents, recirculation of the solution
over the tailings, and excess-solution
evaporation ponds with plastic membranes
and clay liners to minimize leakage. The
retention area design is based on the Bear
Creek model, with a clay-cored dike keyed
to an impervious base.
CO
PO
OD
-------
329
DOSE MODELLING iVOK RADQN-fRCDUUlNG ACTIVITIES IN HMVILY POPULATED AREAS
Lee Bettenhausen* and Veronica Burrows
U. S. JUivironmental Protection Agency
Philadelphia, Pennsylvania
Abstract
The Area Source Radiological Emission Analysis Code (AREAC) was
used to estimate population and individual exposures resulting from radon
effusion sources in two heavily populated areas. One situation results
from disposal of radium and uranium processing wastes at a plant site
over the past sixty years. The other situation occurs as a result of
mining operations.
The mining operation results in radon contamination of the air near
a metropolitan area. The source could affect 234, 386 persons living
within a ten-mile radius. Results of the AREAC model calculations for
radon diffusion and subsequent population exposure using meteorology
and demography for the region are presented.
The waste disposal situation was modelled using measured concen-
trations of radium in soil and relating these concentrations to radon
emanation rates and consequent area source strengths for radon. This
provided input to the diffusion calculations employing regional meteor-
ology and demography. The resulting radon concentrations at specific
receptor sites and at sector centroids gave population exposure estimates.
The disposal site is located in an area of many small population centers
on the fringe of a metropolitan area. 47, 351 persons live within ten
miles.
Introduction
This paper discusses use of a computer code to calculate the
radiological impact of airborne radon in two situations involving
relatively large populations. The results presented herein are
theoretical ones from the computational model. The purpose of the
work was to develop a tool to evaluate the effectiveness of proposed
control measures in reducing population exposure. Another use of the
computational model would be to guide a field measurement program and
to provide a rational method for interpolation and extrapolation of a
few well-chosen field measurements in the area of the radon source. No
experimental radiation exposure data was obtained in the course of this
work.
The computational tool used was AREAC (Area Source Radiological
Emissions Analysis Code), developed by the U. S. Environmental Protect-
ion Agency's Office of Radiation Programs (ML76). The code model was
applied to a radium and uranium waste disposal site and to a radon-
emitting mining operation using regional meteorological and demographic
data. The paper briefly describes the computational methods used and
presents results in terms of exposure at particular locations and to
area populations as the consequence of radon gas emission in the two
situations.
"presently with U. S. Nuclear Regulatory Commission, King of Prussia, Pa.
-------
330
The AREAC code applies the well-known Gaussian diffusion equation
to an emission source of finite area. The source geometry can be either
rectangular or circular, For meteorological input, the code uses the
joint probability distribution for wind velocity and atmospheric
stability (S168), along with an algorithm for atmospheric dispersion
(TuTl) to generate dispersion coefficients from each of the specified
source points to, first, up to 6 specific receptor sites and, second,
to the 16 azimuthal sectors and up to 12 radial segments for population
around the source. The dispersion coefficients multiplied by the area
source strength and specified dose conversion factor to compute annual
radiation exposure to individuals at the specific receptor sites and to
the population distribution specified by angular and radial segment
about the source.
A Mining Operation Emitting Radon
The modelling method and the results obtained can be better under-
stood when applied to a relatively simple situation. In this situation,
a mining operation emits approximately 6 microCuries per second of radon
into the atmosphere as radon-contaminated ventilation air from an
exhaust fan house, essentially a point source. The emitter is located
between two metropolitan areas. Total 1970 census population within
a ten-mile radius of the site was 234, 386. Two available sets of
meteorological data were used in the calculations, since no specific
data was obtainable nor directly applicable to the emission site.
One data set was the ten-year average for the regional airport, about
8 miles distant; the other data set was from a nuclear power plant site
26 miles away. Neither set was truly appropriate for the emission site
because of differing topography. The two sets of meteorological data
also differ considerably. Nevertheless, computed individual and
population doses are comparable, demonstrating that the modelling is
not very sensitive to meteorology.
A dose conversion factor of 4.0\x 10r2 (mrem/yr)/(Curie/m3) (EP77)
is used throughout this work. Resultant population exposures for the two
meteorological regimes are presented graphically in Figures 1. and 2.
Total population dose within the ten-mile radius and maximum individual
dose vathin one mile are 293 person-rem/yr and 313 mrem/yr and 475
person-rem/yr and 336 mrem/yr for the cases of airport meteorological
data and nuclear plant site meteorological data, respectively. This
theoretical model employed a point source of emission directly into the
atmosphere with straightforward diffusion to obtain the results presented.
The next situation is more complex and demonstrates the area source
feature of the AREAC code.
Waste Disposal Involving Radium and Uranium
The second situation modelled was the theoretical radiation exposure
resulting from radon emitted by an old site contaminated by radioactive
wastes from radium and uranium mineral processing. A detailed radio-
logical survey (D078J includes measurements of soil contamination levels
of 226Ra on the site and its environs and as a function of soil depth
for selected bore samples. Schiager's method (Sc74) for estimating radon
-------
331
flux into the atmosphere from radium concentration in soil was used:
4>ffn (pCi/m -sec) = 1.6 CRa (pCi/g).
Schiager tested this relationship against measurements for the Salt
Lake City Vitro uranium mill tailings pile with reasonable agreement -
6.8 pCi/1 calculated radon concentration compared with a measured value
of 10 pCi/1. For this work, the Schiager relationship was tested
against 1975 field measurements for the Climax pile (Du77) with the
following result: Calculated radon concentration at pile edge, 22 pCi/l$
measured values, SE corner 30, NE corner 22, center 1$, SW corner 38 and
NW corner 26 pCi/1. The agreement was considered satisfactory.
Two source terms were computed. One was an areal average of radium
concentration through the upper 6 feet of soil over the entire site. The
other source used only the hot spot or relatively small area of high
radium concentration near the surface in one place on the site. As in
the first situation, two different sets of meteorological data were
used. One set was from the regional airport 17 miles away. The second
set was from a nuclear power plant site 30 miles away. Again, neither
set was really appropriate because of the differing topography.
Results for the two differing sets of meteorological data generally
agreed within a factor of 2 for individual and population doses which
could theoretically result from radon emanation at the site to the
47, 351 persons residing in 1970 within a ten-mile radius. Figure 3
depicts population doses using a circular hot spot of radon emanation
corresponding to a soil concentration of 760 pCi/g 226Ra and airport
meteorological data. Figure 4 shows the theoretical population dose
estimates for the situation with 88 pCi/g over a rectangular site, again
using airport meteorological data. A summary of population doses is
given in Table 1. The model was also used to obtain estimated
individual exposures at specific locations close to the site. A summary
of the results of these calculations is given in Table 2.
Conclusion
A computational modelling method for estimating individual exposure
and population dose has been described. The method and its application
to two different sources of radon emission into the environment were
presented. The model provides a tool to evaluate various radon emission
control alternatives. The computational results obtained should be
compared with field measurements and the model appropriately adjusted
if necessary to utilize the method to obtain valid population dose
estimates for these particular radiological situations and others like
them.
References
Du?7 Duncan, D., G. Boysen, L. Grossman and G.Franz, Outdoor Radon Study
U974-1975), Technical Note ORP/LV-77-1, U. S. Environmental Protection
Agency, Las Vegas, Nevada
-------
332
References (Continued)
D078 1978, Formerly Utilized MED/AEG Sites Remedial Action Program,
Report DOE/EV-OQO5/3, U. S. Department of Energy, Washington, D.C.
EP77 Radiological Quality of the Environment in the United States,
1977, Report EPA 520/1-77-009, U. S. Environmental Protection Agency,
Washington, D. G. 204.60
Mi76 Michlewicz, D., 1976, Area Source Radiological Emission Analysis
Code, Report EPA 520/1-76-017, U. S. Environmental Protection Agency,
Washington, D.C. 20460
Sc74 Schiager, K., 1974, "Analysis of Radiation Exposures on or Near
Uranium Mill- Tailings Piles", Radiation Data and Reports. 15_, 411
S168 Slade, D., 1968, Meteorology and Atomic Energy, U. S. Atomic Energy
Commission Report TID 24190, Oak Ridge, Tennessee
Tu71 Turner, D., 1971, Workbook of Atmospheric Dispersion Estimates,
U. S. Jiiivironmental Protection Agency Publication AP-26, Research
triangle Bark, North Carolina, 27711
TABLE 1
SUMMARY OF POPULATION DOSES NEAR DISPOSAL SITE
hot spot source averaged source
" " Met 2
263
Population doses are person-rem/yeai-
TABLE 2
distance, m.
and direction
125, SSW
261, N
283, ENE
291, NE
335, E
753, WSW
hot spot
Met 1
x
1.70
1.14
1.58
0.46
0.16
source
Met 2
x
0.64
0.76
0.70
0.54
0.05
averaged
Met 1
x
x
X
X
X
JbiiiiL.
source
Met 2
x
X
0.72
0.50
0.48
0.05
Radiation doses are in rem/year
x indicates no calculation performed
Met 1 is meteorological data from nuclear plant site
Met 2 is meteorological data from repinnal_ air port
-------
333
FIGURE 1.
0 POPULATION DOSE
MINING SITE
airport met. data
underlined figures are
radii in statute miles
figures in sectors
are person-rem/year
-------
334
FIGURE 2.
POPULATION DOSE
MINING SITE
n-site met. data
underlined figures are
radii in statute miles
figures in sectors
are person-rein/year
-------
N
0
JL-
335
FIGURE 3.
POPULATION DOSE
DISPOSAL SITE
hot spot source
0
0
-A_-
0
•_3_-
0
0
0
12
18
263
0
0
0
0
0
0
0
0
0
O1
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0
underlined figures are
radii in statute miles
figures in sectors
are person-rem/year
-------
10
336
FIGURE 4.
POPULATION DOSE
DISPOSAL SITE
average source
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0 underlined figures are
radii in statute miles
figures in sectors
are person-rein/year
-------
337
RA-226 CONCENTRATIONS IN THE HYDROGRAPHIC BASINS
NEAR URANIUM MINING AND MILLING IN BRAZIL
A.S. Paschoa, G.B. Baptista, E.G. Montenegro,
A.C. Miranda, and G.M. Sigaud
Pontificia Universidade Catolica do Rio de Janeiro, Depto. de Fisica, Rua
Marques de Sao Vicente 225, Z.C. 19, Rio de Janeiro, R.J. 22453, Brasil
"Abstract". A monitoring survey of the 226Ra concentrations in river wa-
ters in the vicinity of the mining area and future milling facilities in
the Pogos de Caldas region began in January 1977. The objective of the
monitoring survey is to establish a baseline to allow future comparisons
between the 226Ra concentrations in waters of the hydrographic basins of
the Pogos de Caldas plateau before and after the beginning of full scale
commercial operations. Open pit mining started in July 1977 in the urani-
um deposits of Campo do Cercado, but the main uranium body has not been
reached yet. Seasonal variations in riverflow are apparently accompanied
by little variations in the 226Ra concentrations in river waters. A crude
calculational dosimetric model is in the process of being developed to
estimate annual dose equivalent to an individual from 22°Ra via drinking
water and irrigation patterns as a first step to calculate the collective
dose equivalent commitment to the population of the Pogos de Caldas pla-
teau and surroundings.
"Introduction"
The brazilian region of Pogos de Caldas is a remarkable example of
non-explosive volcanic intrusive. The radiogeology of the Pogos de Caldas
region has been described by Roser et al. (Ro64), and later supplemented
by other authors (Cu76)(Ad77). The Pogos de Caldas plateau was formed
when the central part of the alkaline plug, 35 kilometers diameter,thrust
its way 400 meters above the adjacent substratum. Subsequently the cen-
tral part of the plug, was lowered by chemical weathering and erosion ,
and further mineralization occurred (Cu76). Uranium deposits were discov-
ered in this region by Frahya in 1948 (Fr50).
Most uranium ores from the Pogos de Caldas deposits allow to recover
about 0..2% U308(An74). There are two main uranium deposits in the Pogos
de Caldas plateau, namely, Campo do Cercado, and Campo do Agostinho; to-
talizing over 10000 metric tons U308 (An74). In the latter mining opera-
tions did not started yet. Although mining operations in the Campo do Ce_r
cado began in July 1977, the main uranium body has not been reached yet.
Five hundred metric tons U308 are expected to be extracted annualy from
Campo do Cercado through open pit mining when the operations will reach
the commercial level. This projected annual production will enable PWR
reactors to generate about 2.9 x103 Mw(e)/year (Pa77). However, in or-
der to extract annually 500 metric tons U308, a total amount of 1.40xio6
metric tons of gross mineral ores will be mined at 18% efficiency (Fe78).
This total amount of ore, after being processed, will be converted into
1.05 xio6 metric tons of sterile ores (i.e., < 0.01% U308) plus 3.5 x io5
metric tons of sludge. The sterile ores will contain about 30 g 226Ra ,
while the 1.4x102g*26Ra, initially in secular equilibrium with the 238U
-------
338
content of the 500 metric tons V^OQ extracted per year, will be contained
in the sludge mixed in a volume of the order of 10 m liquid plus solid
phases. Therefore, the concentration of 226Ra in the sludge is expected
to be at the yCi/S, level.
Figure 1 shows a map of the Pocos de Caldas plateau with the loca-
tion of Campo do Cercado indicated by a circle. Conventional leaching
processes (Wo58) are being tested to extract uranium from the ores of
Campo do Cercado. Milling facilities will be located in the same geograph
ical region where the mining operations of Campo do Cercado are taking
place.
The future tailings from the mining and milling operations in the
Pocos de Caldas region are expected to be kept in a way that 226Ra leach-
ing can be reduced as much as possible. However, the stabilization proce-
dures to be adopted and the actual percentage of 226Ra which will leach
from the tailings are still unknown.
Although it is well recognized today that the radioactive contamina-
tion of water utilized in uranium mining and milling should be avoided ,
and coal mining may be more damaging for a local environment than uranium
mining (NP77), detriment may result from the uranium mining and milling
operations in POC.OS de Caldas in the case of significant 226Ra contamina-
tion of waters of the hydrographic basin of the region. A recent study
carried out under the auspices of the American Physical Society concluded
that "for regional and local population exposure, radionuclides in urani-
um mill tailings are potentially at least as important as the actinide
chain elements in high-level waste; the relative accessibility of mill
tailings contrast with the isolation proposed for other actinide-contain-
ing wastes" (AP78). The rationale behind this statement becomes transpar-
ent through the fact that the long half-life of 226Ra (1.602x 1Q3 years )
imposes the necessity of considerable effort to find a solution to pre-
vent any amount of 226Ra,large enough to result in detriment,from leach-
ing from the tailings of commercial uranium milling operations. To the
best knowledge of the authors, there is not yet any long term solution
for segregating, efficiently, uranium mill tailings from the surrounding
environment, therefore, monitoring the environment is still the most ef-
fective way to detect faulty procedures or fails in the waste management
of the tailings from uranium mining and milling operations. The local
water contamination by 226Ra resulting from uranium operations can only
be assessed with any degree of confiability if the baseline of natural
radioactivity in the surrounding region is previously established for
future comparisons. A number of investigators have published reports of
the impact on the local environment from operations in several parts of
the world involving extraction and treatment of uranium for nuclear in-
dustry (Ru61)(Ts63)(Ha68)(Ha70)(Iy70)(Ki71)(Ka75)(Pr76)(Ea77)(Ko78), but
the only report so far dealing with baseline studies prior to the begin-
ning of uranium operations at any particular site is the one related to
the work now being developed in Australia (Br78).
The monitoring survey of the 226Ra concentrations in the waters of
the main hydrographic basins of the Pogos de Caldas plateau started in
January 1977, six months before the beginning «* «-^« ..-„-,•, ,•„,•„„ „
-------
339
tions in the region.Other investigators are undertaking studies on sedi-
mentation of 226Ra and on 222Rn emanation from soils and atmospheric dif
fusion. ~~
A simulation model to calculate the internal alpha dose to an indi-
vidual from 226Ra intake via the pathways of drinking water and irriga-
tion paterns is in the process of being developed based upon scenarios
consistent with today's state of art. This model will constitute the ba-
sis to calculate the collective dose commitment to the population of the
Po£os de Caldas plateau and surroundings.
"Sampling arid Experimental Procedures"
The water drainage of the POC.OS de Caldas plateau occurs by three
natural streams. The Rio das Antas crosses the plateau in the general di^
rection south-to-north, as can be seen in the map shown in Figure 1, and
drains about seventy percent of the waters which fall on the plateau
The Rio das Antas will be the natural recipient of liquid wastes from
the mining operations in the Campo do Cercado and Campo do Agostinho, as
well as from the milling operations. There is a plan to use water from
Rio das Antas for the water supply system of the city of Pogos de Caldas
(PD70), located near the northern edge of the plateau. Two small dams ,
Saturnine de Brito and Bortolan, receive waters from the Rio das Antas,
which is also tributary of the Rio Lambari. The latter together with Rio
Pardo are the main contributors to the water volume of Graminha dam
which appears in Figure 1. The Graminha dam covers thirty square kilome-
ters and is near two cities, Palmeiral on the eastern shore and Caconde
located five kilometers from the dam in the northwestern direction. The
Rio Verde flows into the Rio Pardo forty kilometers downstream from its
source, which is located ten kilometers from the center of Campo do Cer-
cado in the southeastern direction, after draining about twenty percent
of the waters that fall on the POC.OS de Caldas plateau by crossing the
border of the volcanic intrusive through the eastern edge. The Rio Verde
will also receive part of the liquid wastes from the mining operations
in the Campo do Cercado. This river passes near two cities, Pocinhos in-
side the border of the plateau,and Caldas outside. A brook called Ribei-
rao da Prata, which runs across the city of Sguas da Prata outside the
western border of the volcanic intrusive, plus other smaller natural
streams around the plateau,drain under ten percent of the water system.
Water samples are taken monthly from 28 sites of collection which
are shown in Figure 1, and represent approximately the part of the hy-
drographic basins most likely to be affected by the uranium mining and
milling operations. Water samples are collected in one liter plastic bo_t
ties and transported to Rio de Janeiro about five hundred kilometers
apart from the collection sites without previous filtration and acidifi-
cation. Once in the laboratory in Rio de Janeiro, the water samples are
filtered through 0.45um membrane filter and are subsequently acidified
to a pH 2 or below with hydrochloric acid. Comparisons between samples
filtered and acidified in the site of collection and then transported
and those transported without these two preliminary procedures have
shown that the adsorption rates in the first few weeks after collection
differ very little. Less than five percent of the 226Ra presented ini-
tially in solution is lost onto the walls of the recipients in which
-------
340
the samples are transported without previous filtration and acidifica-
tion. On one hand filtration through a 0.45ym membrane filter is a time
consuming procedure and cannot be easily performed in the collecting si-
tes of the POC.OS de Caldas region. On the other hand, acidification
without previous filtration might prove misleading for determining the
226Ra concentrations of the water samples, because if the suspended so~
lids would contain significant amounts of 226Ra, the acid solution could
cause leaching,thereby resulting in wrong values for the 226Ra concentra_
tions in the water.
A simpler version of the well known de-emanation method (Ha67) is
used to determine the 226Ra concentration in the water samples. After
filtration and acidification, a 120 ml aliquot is taken from each water
sample for 222Rn ingrowth into a 150 ml volume capacity bubbler. Co-pre-
cipitation with barium is not used because the levels of 226Ra concentra^
tion of interest are high enough to be detected by direct de-emanation
of 222Rn, which is at a known state of equilibrium with 226Ra in solu-
tion, from the bubbler into a scintillation flask. Light pulses from the
scintillation flask are counted in a photomultiplier coupled with high
voltage supply, sealer and timer, after equilibrium between 222Rn and
short-lived daughters is established. The experimental lower limit of
detection with these simple procedures is about 0.20 pCi226Ra/&. The de-
tection system has been inter-calibrated with other somewhat similar sys_
terns used around the world through the International Atomic Energy Agen-
cy (IAEA) Coordinated Programme on Studies on the Source, Distribution ,
Movement and Deposition of Radium in Inland Waterways and Aquifers.
"Results and Comments"
Results of the analyses of the 226Ra concentrations in the waters
of Pogos de Caldas region are shown in Table 1 for the wet and dry sea-
sons, represented by the months of January and July respectively. In
Table 1 the letters A, M, L, and R denote collection sites associated
with the water system of the Rio das Antas, letter V denotes collection
sites in the Rio Verde and some small tributaries, F-l represents a
collection site in the southern base of Morro do Ferro, and collection
sites denoted P-l, P-2, and P-3 are located, respectively, in the
Ribeirao da Prata in the city of Sguas da Prata, in the source of
mineral water called Fonte do Villela also in the city of Sguas da Pra-
ta, and in the Ribeirao do Quartel which flows directly into the Ribei-
rao da Prata.
The 226Ra concentrations in the wet season are, in general,slightly
higher than those recorded in the dry season, though the variations are
within the range of the experimental errors. However, for waters collec-
ted in the sites denoted by R-3 and P-3 the results for the dry season
are higher than those for the wet season. Tentative explanations for the
results, apparently anomalous, of R-3 and P-3 may be as follows: (i) R-3
is a collection site located in the Rio das Antas after the points of
release of the sewage system of the city of Pocos de Caldas, so 226Ra
may be dumped into the river with contaminated sewage: and (ii) P-3 is
a collection site located in the cSrrego do Quartel, in the town of Cas-
cata, where in the dry season the water is usually clear, but in July
1977 there was resuspension of sediments bec»uss sf «-^~ „-„„«._„„,..; _*
-------
341
a small barrage upstream, so the 226Ra might have been leached out of the
suspensed sediments increasing the 22&Ra concentration in the water. An
other anomalous result appears to be the high 226Ra concentration (i.e. ,
3.9 pCi226Ra/£) of the water sample collected in January 1977 in the site
A-8, followed by non-detected 226Ra concentrations (i.e., < 0.20
pCi226Ra/£) in the months of July 1977 and January 1978. However, no plaiu
sible explanation has been found so far for this fact.
The water samples collected in the site P-2 present the highest
226Ra concentrations in all months shown in Table 1. Such water samples
are from a source of natural mineral water (Fonte do Villela). An earlier
analysis of this mineral water has resulted also in a 226Ra concentration
about 30 pCi226Ra/£ (Ha74). The water samples collected from the Ribeirao
da Prata at the site P-l, 30 meters apart from the Fonte do Villela, have
non-detected concentrations (i.e., < 0,20 pCi226Ra/£), which indicates
that the river water is not affected by the 226Ra source for the Fonte do
Villela nearby.
The water sample collected in the site M-l in January 1977 presented
the 22&Ra concentration of 26 pCi226Ra/&. This site of collection has been
destroyed by mining excavations in the center of Campo do Cercado.
The geographical configuration of the Pogos de Caldas plateau is
such that two hills, Morro do Cercado, and four kilometers north of this,
Morro do Ferro, constitute the main elevations that divide the waters
that fall in the region. The flowrate of Rio das Antas are shown in the
graph of Figure 2a. Data on the rain precipitations in the Pocos de Cal-
das region are summarized in the graph of Figure 2b; where the dots re-
present the monthly average of pluviometric records for the year 1977 ,
and the vertical bars indicate the range of monthly average values recor_
ded between 1961 and 1968. Observing the graph of Figure 2b one can con-
clude that in general the dry season is characterized by rain precipita-
tions lower than lOOmra/month between May and September, with a minimum in
July. The maximum rain precipitation is usually reached in January, char-
acterizing the peak of the wet season. The flowrate of Rio das Antas fol-
lows the pattern of rain precipitation in the Pbgos de Caldas plateau ,
and as can be seen from Figure 2a, the flowrate of this river in the wet
season is about four times higher than in the dry season.
Figure 2c shows the monthly 226Ra concentrations, for the year 1977,
of the water samples collected in the sites M-4, A-l, and V-2, located
respectively in the mining area, Rio das Antas, and Rio Verde. The 226Ra
concentrations in waters collected in the sites M-4, A-l, and V-2 had dif_
ferent distributions throughout the year 1977. According to Figure 2c ,
water samples collected in A-l appear to have had slightly higher concen-
tration in the wet season than in the dry season. Of course, further data
are still necessary to confirm such observation. However, this can be
explained by the following reasoning: on the one hand, the volumetric in-
crease of water in rivers during the wet season dilutes the 226Ra concen-
o o c
tration in the river water; on the other hand, the amount of "bRa in so-
lution in river water can be enhanced by the resuspension of 226Ra
adsorbed in the sediments. In the rivers of the Pogos de Caldas region ,
the latter effect seems to predominate. Variations of 26Ra concentration
,1s waters as a function of time of the year have already been observed
-------
342
elsewhere (Pa78), although under different conditions.
The water samples collected in M-4, which is located at a point only
2,5 km apart from the center of the mining operations of Campo do Cerca-
do in a small tributary of Rio das Antas, presented, in 1977, 26Ra con-
centrations higher than those water samples collected either in A-l or
V-2. Furthermore, between July 1977 and December 1977 the 226Ra concen-
trations of the waters collected in the site M-4 increased from (2.8±0.5)
pCi226Ra/£ up to (6.0±1.0) pCi226Ra/;i while before July 1977 the 226Ra
concentrations never reached 2.8 pCi22°Ra/£ level. This remarkable in-
crease in the 226Ra concentration in the waters collected in the site
M-4 indicates that the uranium mining operations are releasing 226Ra to
the small stream which flows into the Rio das Antas after crossing the
Campo do Cercado in the southwestern direction.
The 226Ra concentrations in the waters collected in the site V-2
were not detected (i.e., < 0.20 pCi226Ra/£) until September 1977, when
started increasing steadily up to (1.3±0.4) pCi226Ra/£ in January 1978.
The collection site V-2 is located in an indirect affluent of the Rio
Verde, and is 2.5 km distant, in the northeastern direction, from the
center of Campo do Cercado. The site V-2 is near an area where prelimin£
ry work for future mining operations started around August 1977.
"Dosimetry"
A calculational dosimetric model is in the process of being devel-
oped to estimate the annual internal dose equivalent from 226Ra intake
via the pathways of drinking water and irrigation patterns, as the basis
to calculate the collective dose equivalent commitment to the population
of. the POQOS de Caldas plateau and surroundings. The dosimetry is essen-
tially based upon exponential models recommended by the International
Commission on Radiological Protection, ICRP (IC59), and used successful-
ly in the computer program HERMES (F171) and in the report WASH-1258
(So73).
The ICRP hypotheses and recommended values (IC59)(IC74) were used
for calculating the annual dose equivalent for the whole body and selec-
ted organs of an adult individual. Accordingly, the usage factor for wa-
ter ingestion is 438£/year (IC74). The human intake of the main food pro
ducts of the region are the following: coffee - 23 kg/year; potato - 90"
kg/year;corn, tomato, bean, and rice - 115 kg/year each (IC74). Further
details of the dosimetric model and of the irrigation characteristics of
the Pogos de Caldas region will be given elsewhere.
Figure 3 shows a series of graphs of the annual dose equivalent, in
mrem*/year, from 226Ra via the pathways of drinking water and irrigation
patterns, as a function of the 226Ra concentration in the water. The
upper line of each graphic bar in Figure 3 indicates an annual dose equi
valent calculated under the assumption that 222Rn is totaly retained in
the organ, while the lower line of each bar indicates the contrary as-
sumption, that is total 222Rn escape from the organ. Experimental data
* 1 mrem = 10 rem = 10 sievert (= 10~5l/kg
-------
343
on 222Rn escape from human organs are not generally available in the open
literature, with the exception of the 0.6 fraction of 222Rn that escapes
from bone (Ro58).
Observing the graphs of Figure 3, one can say that when the 226Ra
concentration in the drinking water is 3 pCi226Ra/£, the annual dose equi
valent to the whole body is approximately 10 mrem/year, while to the bone"
the annual dose equivalent is about 15 mrem/year. However, if the 226Ra
concentration in the water to be used to irrigate plantations of coffee ,
potato, corn, tomato,bean, and rice is 30 pCi226Ra/Jl, the annual dose
equivalent to an adult individual, consuming these food products at the
same rate as the reference man (IC74), can reach about 700 mrem/year to
the whole body and over 1 rem/year to the bone. Although the maximum per-
missible concentration for 226Ra in drinking water is generally lower
than that for waters with potential to be used for irrigation purposes ,
the use of water with the same concentration may result in lower dose
when used for drinking than when used for irrigation of food plantations.
This somewhat striking observation should not take anyone by surprise ,
since in 1967, Eisenbud based on studies made in several parts of the
world had already noticed that "the principal daily intake of radium is
from food rather than from water" (Ei67).
"Concluding Remarks"
Evidently, further studies are needed to support any conclusion that
one may decide to draw at this stage of an assessment like this, which
constitutes only a preliminary step towards a long term project. However,
several tentative conclusions and general observations deserve to be re-
gistered here for future references.
(i) The 226Ra concentration levels in the river waters of the Pocos
de Caldas region are in general under 1.0 pCi226Ra/£, with the exception
of the waters of few small streams crossing the area where the uranium
deposits are now being explored, even though the 226Ra concentrations in
these waters are still lower than 30 pCi226Ra/£.
(ii) The 226Ra concentrations in river waters of the Pocos de Caldas
plateau may be higher in the wet season than in the dry season, but the
differences are slight, so more refined studies are to be untertaken to
allow a definitive conclusion concerning this matter.
(iii) A model to calculate the collective dose equivalent to the po-
pulation of the Pocos de Caldas region from the 226Ra existing naturally
in solution in the river waters will allow to estimate the dose enhance-
ment from any increase in the 226Ra concentrations in the usable river
waters. Appropriate data are now being gathered to calculate, according
the ICRP recommendations, the collective dose equivalent from 226Ra to
the population of the Pocos de Caldas region.
(iv) If 226Ra, from the future tailings of the uranium mining and
milling operations in the Pocos de Caldas plateau, would be released to
the environment in amounts such that the 226Ra concentrations in the wa-
ters used to irrigate the typical food plantations of the region would
reach 30 pCi226Ra/£, one could expect an annual dose equivalent as high
-------
344
as 700 mrem/year to the whole body (and over 1 rem/year to the hone) of
an adult individual eating those foods at the same rate as the reference
man.
(v) The highest annual dose equivalent via the pathway of drinking
water under the present conditions of the POC.OS de Caldas region would be
around 95 mrem/year to the whole body (150 mrem/year to the bone) of an
adult individual who would drink daily watej^from the Fonte do Villela ,
with 226Ra concentration just under 30 pCi Ra/&. Notwithstanding, the
Fonte do Villela constitutes a touristic attraction and its water, accbrji
ing to one anonymous tourist , is good not only for those who are
thirsty, but also is "good for health"!
"Acknowledgements"
This work has been performed with the support of the International
Atomic Energy Agency (IAEA), Comissao Nacional de Energia Nuclear (CNEN),
Financiadora de Empreendimentos e Pesquisas (FINEP) e Conselho Nacional
de Desenvolvimento Cientifico e Tecnologico (CNPq).
"References"
Ad77 J.A.S. Adams. 1977. The Geological Origins of Radioactive Anomalies,.
Iri Proceedings of the International Symposium on Areas of High Natural
Radioactivity. June 16-20, 1975. Pocos de Caldas, Minas Gerais ( Thomas
L. Cullen and Eduardo Penna Franca, editors), 5.
An74 J.R. Andrade Ramos and M.O. Fraenkel. 1975. Main Uranium Occurrences
in Brazil. In Proceedings of a Symposium on Formation of Uranium Ore De
posits. May 6-10, 1974. Athens. International Atomic Energy Agency, ~~
STI/PUB/374.
AP78 American Physical Society (APS). 1978. Report to the APS by the
Study Group on Nuclear Fuel Cycles and Waste Management. Reviews of
Modern Physics, 5£ PII, S71.
Br78 A.J. Brownscombe, D.R. Davy, M.S. Giles, and A.R. Williams. 1978.
Three Baseline Studies in the Environment of the Uranium Deposit at
Yeerlirrie, Western Australia. Australian Atomic Energy Commission
AAEC/E442, ISBN 0 642 59650 6.
Cu76 T.L. Cullen and A.S. Paschoa. 1976. Morro do Ferro, an Invitation to
Radioecology. In Proceedings of the Health Physics Society Tenth Mid^-
year Topical Symposium on Natural Radioactivity in Man's Environment.
October 11-13, 1976. Saratoga Springs, New York, 301.
Ea77 G.G. Eadie and R.F. Kaufmann. 1977. Radiological Evaluation of the
Effects of Uranium Mining and Milling Operations on Selected Ground Wa-
ter Supplies in the Grants Mineral Belt, New Mexico. Health Physics
_32_, 231. '
Ei67 M. Eisenbud. 1967. Radionuclides in the Environment. In Proceedings
of a Symposium on Diagnosis and Treatment of Deposited Ralionuclides
Mav 15-17. 1967. Richland. Washington. 1.
May 15-17, 1967. Richland, Washington,
-------
345
Fe78 P.A.M. Ferreira. 1978. personal communication.
F171 J.L. Fletcher and W.L. Dotson (compilers). 1971. A Digital Computer
Code for Estimating Regional Radiological Effects from the Nuclear Pow-
er Industry, USAEC, Report HELD-TME 71 - 168, Hanford Engineering Deve-
lopment Laboratory.
Fr50 R. Frahya. 1950. Relatorio da Diretoria - 1948 - Divisao de Fomento
da Produgao Mineral, Boletim N° 87 (Alberto I. Erichsen, editor) 112.
Ha74 P.L. Hainberger, I.R. de Oliveira Paiva, H.A. Salles Andrade, G.
Zundel, and T.L. Cullen. 1974. Radioactivity in Brazilian Mineral Wa-
ters. Radiological Health and Data Reports, 15, 483.
Ha67 J.H. Harley. 1967. Manual of Standard Procedures, Health and Safety
Laboratory, N.Y.O. 4700 (2nd. edition), E.Ra 01-01, 06-06.
Ha68 B. Havlik, J. Grafova, and A. Nycova. 1968. Radium-226. Liberation
from Uranium Ore Processing Mill Waste Solids and Uranium Rocks into
Surface Streams - I. Health Physics, 14, 417 - II, 423.
Ha70 B. Havlik. 1970. Radioactive Pollution of Rivers of Czechoslovakia.
Health Physics, 19, 617.
IC59 International Commission on Radiological Protection (ICRP). 1959.
Publication 2, Pergamin Press, 1960; Health Physics, .3, 1, 1960.
IC74 International Commission on Radiological Protection (ICRP). 1974.
Publication 23, Report of the Task Group on Reference Man, Committee II,
Pergamon Press, Oxford.
IC77 International Commission on Radiological Protection (ICRP). 1977.
Publication 26, Pergamon Press, Oxford.
Iy70 M.A.R. lyengar and P- Markose. 1970. Monitoring of the Aquatic Envi-
ronment in the Neighborhood of Uranium Mill at Saduguda, Bihar. In Pro-
ceedings of a National Symposium on Radiation Physics. November 24-27,
1970. Trombay.
Ka75 R.F. Kaufmann, G.G. Eadie, and C.R. Russel. 1975. Ground Water Quali_
ty Impacts of Uranium Mining and Milling in the Grants Mineral Belt.
New Mexico. U.S.E.P.A. 906/9-75-002.
Ki71 P. Kirchman, A. Lafontaine, G. Cantillon, R. Boulenger. 1971. Trans-
fert dans la Chaine Alimentaire et 1'Homme, du Radium-226 Provenant
d'Effluent Industriels Diverse dans les Cours d'Eau. In Proceedings of
the International Symposium on Radioecology Applied to the Protection
of Man and His Environment. September 7.10, 1971. Rome.
Ko78 I. Kobal, J. Kristan, M. Skofljanec, S. Jerancic, and M. Ancik. 197&
Radioactivity of Spring and Surface Waters in the Region of the Uranium
Ore Deposit at Zirovski Vrh. Journal of Radioanalytical Chemistry, 44,
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346
NP77 Nuclear Power Issues and Choices. 1977. Report of the Nuclear Energy
Policy Study Group. Sponsored by Ford Foundation. (This report is
widely known as the Mitre-Report).
Pa77 A.S. Paschoa and G.B. Baptista. 1977. Environmental Impact of a Nu-
clear Industry at an Early Stage of Development: Peculiar Aspects. In
Applications of Environmental Impact Analysis to the Nuclear Power In-
dustry. Proceedings of a Regional Seminar Organized by the Internatio-
nal Atomic Energy Agency. August 29 - September 2, 1977. Buenos Aires.
IAEA - 212, 269.
Pa78 A.C. Paul, V.S. Londhe, and K.C. Pillai. 1978. Radium-228 and
Radium-226 Levels in a River Environment and Its Modification by Human
Activities. Ir\_ The Natural Radiation Environment III (To be published
by the U.S. Department of Energy).
PD70 Piano de Desenvolvimento Integrado de Pogos de Caldas, 1970-1971 (Iri
tegrated Development Plan for the City of Pocos de Caldas, 1970-1971).
1970. Prefeitura Municipal de Pocos de Caldas.
Pr76 J. Pradel and P. Zettwoog. 1976. La Radioprotection dans L1extraction
et le Traitment de 1'Uranium et du Thorium, Septembre 9-11, 1974.
Bourdeau, International Labour Office ed., Geneve, 249.
Ro64 F.X. Roser, G. Kegel, and T.L. Cullen. 1964. Radiogeology of Some
High-Background Areas of Brazil, In the Natural Radiation Environment
(John A.S. Adams and Wayne Lowder, editors), 865.
Ro58. R.E. Rowland, J. Jowsey, and J.H. Marshall. 1958. Radon Escape from
Bone Mineral. Radiation Research, ^, 298.
Ru61 D.E. Rushing. 1961. Determination of Distribution Coefficient and
Water Leaching of Radium in Tailing Fractions from Alkaline Circuit
Uranium Ore Processing Mills. U.S.A.E.G. Ds No. WIN-125.
So73 J.K. Soldat, D.B. Shipler, D.A. Baker, D.H. Denham, and N.M.
Robinson. 1973. A Computational Model for Calculating Doses from Radio-
nuclides in the Environment. In U.S.A.E.G. Final Environmental State-
ment - ALAP - LWR Effluents, Vol. 2, Report 1258.
Ts63 E.G. Tsivoglou. 1963. Environmental Monitoring in the Vicinity of
Uranium Mills. In Proceedings of the International Atomic Energy Agen-
cy Symposium on Radiological Health and Safety in Mining of Nuclear
Materials. August 26.31, 1963. Vienna, 231.
Wo58 R.J. Woody and D.R. George, 1958. Acid Leaching of Uranium Ores. In
Uranium Ore Processing (John W. Clegg and Denis D. Foley, editors). —
Addison - Wesley Publishing Company, Inc., 115.
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347
Table 1. Ra-226 concentrations -in waters of the Pogos de Caldas region.
Collection
site*
A-l
A-2
A-3
A-4
A-5
A-6
A1- 6**
A- 7
A-8
A-9
A-10
L-l
L-2
R-l
R-2
R-3
M-l***
M-2
M-3
M-4
V-l
V-2
V-3
V-4
P-l
P-2
P-3
F-l
Jan 77
wet season
1.8± 0.7 t
0.6± 0.2
0.3±0.2
0.4±0.2
0.4 + 0.2
0.3±0.2
0.6±0.3
0.5±0.2
3.9±0.7
1.0±0.3
ND tt
ND
ND
0.6±0.3
ND
ND
26.0±6.0
1.3±0.4
5.0± 1.0
2.6±0.5
ND
ND
ND
ND
ND
29.4±6.0
ND
ND
Jul 77
dry season
pCi^oRa/*
1.2± 0.7
ND tt
ND
ND
ND
ND
ND
ND
ND
1.4±0.4
ND
ND
ND
ND
ND
0.8±0.3
-
1.110.4
4.1± 0.7
2.8±0.5
ND
ND
ND
ND
ND
28.0±6.0
0.6±0.3
ND
Jul 78
wet season
1.6± 0.7
0.3± 0.2
0.3± 0.2
0.2± 0.2
0.3± 0.2
0.3± 0.2
0.3± 0.2
0.4± 0.2
ND
1.7± 0.4
ND
ND
ND
ND
ND
ND
-
1.2±0.4
6.0± 1.0
5.0± 1.0
ND
1.3± 0.5
ND
ND
ND
26.0±6.0
ND
ND
* Geographical locations of collection sites are shown in Figure L
** A'-6 is a well twenty meters apart from collection site A-6 in
Antas river .
*** M-l was a water source annihilated by the mining excavations.
t la due to statistical counting and calibration uncertainties.
tt ND = not detected, i.e. < 0.20 pCi 226Ra/£.
-------
348
CACOXDE
PAIMEIRA1
«*i,
A,/ ^ ''NCOS "\
« \
ilOAS V\
'5kr
Figure 1. Sample sites for water collection in the POC.OS de Caldas region,
and the idain rivers of the hydrographic'basins. Map is based upon Institu-
te Brasileiro de Geografia e Estatistica (IBGE) charts.
-------
a)
ANTAS RIVER
349
50 -
LU 40
»-
< 30
S 20
O
u. |0
* X
x b)
-
X x .
* X
* X x X
1 1 1 1 1 1 1 1 1 1 1 1 1
J FMAMJ JASONDJ
6.0
5.0
o, 40
o
cc
if)
CM 3.0
CM
O
0.
2.0
1.0
0.0
o E
C) E
i
o o Q
o o t—
I—
o M-4 "
LU
•A-l £
o A V - 2 2
o —
°' or
o o o
o
- O _J
• ' X
9 ^""
• • • 2
•*••*** * °
A 2
A
DETECTION LIMIT * _ _
I I I I I. 1 1 1 I I I 1 1
s
POCOS DE CALDAS
1
300
e
e •
-250
200 '!
o
-150
e
100
(
o
•50
1 1
' 1
1 1 I I 1 1 Y I 1 1 I 1
JFMAMJJASONDJ (1978)
YEAR'- 1977
JFMAMJ JASOND
Figure 2. a) Flow rate of Rio das Antas; b) data on rain precipitation in
PaoQS de Caldas: and c) Ra-226 concentrations in water samples collected
in ^hc mtning area fM-4), Rio das Antas (A-l), and Rio Verde (V-2).
-------
350
concentration ( pCi Ra/l)
Figure 3. Annual dose equivalent from 226Ra via the pathways of drinking
water (bars with hachures) and irrigation patterns (dark bars) . Upper
lines of the bars indicate annual dose equivalent with no 222Rn escape ,.
while lower lines indicate total 222Rn escape.
-------
351
EVALUATION OF THE ENVIRONMENTAL DOSE COMMITMENT
DUE TO RADIUM-CONTAMINATED SOIL
Joyce Feldman, Radiation Branch, U. S. EPA Region II
Jeanette Eng, New Jersey Department of Environmental Protection
Paul A. Giardina, Radiation Branch, U. S. EPA Region II
Abstract
The Middlesex Sampling Plant located in Middlesex, NJ was a uranium ore
sampling plant operating during the 1940s and 1950s. A radiological problem was
identified during a routine program to resurvey selected former MED/AEC sites
which are no longer under government control. The survey, when conducted by the
U. S. Department of Energy (DOE), indicated that the Middlesex facility had a
radium and radon problem on-site as well as off-site, where some of the
contaminated soil was used as landfill. The old sampling plant is presently
being used as a Marine Corps Reserve Training Center. Subsequent, more detailed
studies have identified possible solutions to the contamination problem. The U.
S. Environmental Protection Agency (EPA) is examining cleanup options based on a
cost/benefit analysis utilizing the environmental dose commitment concept
rather than an annual dose calculation. The practice of using dose to local
populations as a basis for impact assessment can lead to a large underestimate
of the total potential impact from the continuous environmental release of
radon.
Introduction
Activities performed in connection with the Manhattan Engineer District
(MED) efforts to develop the first atomic bombs and with postwar research
sponsored by the U. S. Atomic Energy Commission (AEC) to develop nuclear energy
led to the contamination of numerous sites in this country. The U. S.
Department of Energy (DOE) has conducted a program to determine the present
radiological status of those sites formerly used by MED/AEC and their
contractors. Based on radiological monitoring most of the sites have been found
to present no radiological hazard to present occupants. In such cases, the
sites are approved for unrestricted use. At some of the sites, however,
measurements of radioactive contamination levels have indicated a need for some
form of remedial action before the site can be released for unrestricted use.
One of the locations surveyed, the Middlesex Sampling Plant in Middlesex,
NJ, was used during its MED/AEC period for sampling, weighing, assaying and
storing of uranium and thorium ores (FB78a). One of the major ores stored there
was Belgian Congo uranium ore, which had a concentration of about 60% uranium.
In equilibrium in the ore were radium and its decay products (FB78b). These
elements formed the major constituents of the radioactive contamination found at
the Sampling Plant site. The levels measured both on-site and off-site were
found to exceed background measurements nearby by factors of up to 2000
(FB78b).
As a result of these findings the DOE has presented several possible
remedial actions for the on-site and the off-site contaminated areas. The DOE
presentation was based on levels of radioactive contaminants as measured on-
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site. A measurement of gamma radiation levels or of alpha contamination would
be a realistic indication of a potential radiological hazard attributable to
fixed contamination.
Such a study of concentrations may tend to misrepresent the off-site hazard
when one or more of the radioactive constituents is a gaseous product, capable
of diffusing during its lifetime and moving off-site. With the diffusion of
radon gas, the maximally exposed individual is not a valid reflection of maximum
population dose. There is a greater distribution of the gas over a large area,
leading to a severe underestimate of the population dose. Consideration
should be given to evaluation of the total health impact of any suggested
actions. Each option will involve exposure of some segment of the general
population to radioactive materials. These will take the form of gases (radon
emitted from the tailings) or particulates (dust laden with loosened radioactive
soil from earth moving operations).
The short-term effects of operations have been addressed in the Engineering
Evaluation and the Environmental Analysis, as have the financial costs.
However, long-term effects of exposure to gases and particulates which have not
been contained either prior to or during cleanup should be considered. Further,
materials remaining after cleanup operations have been completed will lead to a
very low level, but long term, dose to the general population. There should be a
comparison of cost of the cleanup options and the anticipated health effects
from materials remaining. This would give a basis for appraisal of each
remedial action alternative.
Proposed Options
The Middlesex Sampling Plant and locations nearby affected by its
operations require cleanup efforts of some type. In all cases, remedial actions
would involve removal of buildings on site, excavation of soils and emplaced
utilities such as the contaminated drainage system, asphalt coverings, lawns,
and sidewalks. Follow up actions would include restoration of land surfaces to
their original grade. The remedial actions being examined by DOE are based on
levels of contamination which may be allowed to remain (FB78b). They include
the following:
1. Interim storage on site : Areas of especially high contamination would
be removed, packaged and stored on-site for the period ending with
transfer to a permanent storage area. This would be used in addition
to alternative actions and is designed to limit exposures to the
public by packaging of contaminated portions of buildings and soils
immediately. Additional clean up could then be implemented at a later
time.
2. Long-term storage on-site : Packaged contaminated materials and soils
from both on-site and off-site locations would be stored on the
Sampling Plant site indefinitely. Such a program would require that
the site not be open to public access for a prolonged period.
Applicable criteria for such a long-term storage program would also
need to be considered.
3. Removal of materials in interim storage and direct removal of site
soils and buildings to a disposal site : This would follow
implementation of alternative 1.
4. Direct removal of all contamination to a disposal site : Such a
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program would require more time to implement, since arrangements for
final disposition would need to be made before removal of contaminated
materials could be initiated.
The DOE has presented these options in an engineering report. Evaluations
of the options have been presented in a second report which presents an
environmental analysis of the effects of all options. The effects discussed are
examined in light of gamma radiation levels and radon concentrations now
existing at each location. According to the DOE report potential health effects
from radon daughter inhalation by occupants of the site buildings would result
in one radiation associated lung cancer to building occupants every 70 years.
External gamma exposures would yield the equivalent of one cancer incidence
every 6,000 years. Projections of health effects have been made based on
exposures to maximally exposed individuals off-site as well as to regular
occupants of on-site facilities.
Environmental Dose Commitment
The assessments suggested here for deriving health effects projections
depart in two significant respects from practice common in the past for
assessing the significance of radiation exposures. The first of these is the
use of the concept of environmental dose commitments (EDC). This concept
considers the totality of doses to all populations over the lifetime of the
radionuclide in the biosphere and not just to a maximally exposed individual
located near a facility (US73).
The second departure from past practice is to evaluate potential health
effects rather than to minimize radiation dose as the end point. In retrospect
it is perhaps obvious that the focus for determination of the appropriate level
for any guidance should be its public health impact, but in the past
minimization of dose has served as a useful surrogate for the health impact
because of uncertainties about the magnitude and shape of the relationship
between dose and effect.
Under the concept of environmental dose commitment the health impact
analysis thus considers the total impact of releases of radioactive materials to
the environment by including radiation doses committed to local, regional,
national, and worldwide populations, as well as doses committed due to the long-
term persistence of some of these materials in the environment following their
release. The analysis would serve to identify the effluents from the site
(gaseous as well as wind-borne solids and soluble or insoluble contaminants
carried off by water erosion) which represent the major components of risk to
the population. This would lead to a better-defined view of the need to control
long-lived contaminants as well as a recognition of the excess control measures
for any short-lived radioactive materials.
To make a determination of the degree of-control which can reasonably be
required by any applicable standards, an analysis of cost-effectiveness of risk
reduction must be performed. This will be discussed below.
For the Middlesex Sampling Plant Site calculations of the EDC may be
performed with very little additional data. Population estimates for the area
for the period of interest exist now. This would probably be 100 years, the
period recommended by the EPA for institutional control of a waste repository.
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The dose rates have been measured on-site. Off-site estimates of radon may be
made knowing the source term and using existing wind rose data. From this
information, a dose per year from each radionuclide of interest (basically
radium and radon isotopes) may be found. Determining these values for the
period of interest will yield the EDC. Based on the EDC in person-rem, health
effects due to each deposited radionuclide may then be considered.
The DOE report has addressed radiological impacts in terms of potential
health effects to occupants of the Sampling Plant buildings as well as to
nearest off-site residents for radioactive contamination levels existing now.
Consideration has also been given to contamination problems due to vegetation
uptake factors for foods grown in contaminated soil off-site (FB78b).
Environmental impacts due to remedial action implementation have been examined.
As a short-term problem many of these impacts, such as releases of radioactive
airborne particulates, would have their greatest potential effects on workers
involved in the cleanup operations, rather than on the general population. In
addition to air quality effects there would be radiological effects on local
stream water quality. These effects have been discussed to some extent in the
engineering report (FB78a) and should be considered in the decision-making
process, although contribution their may be minimal.
Cost-Effectiveness Studies
The basic thrust of the health effects calculations described above would
be to establish the need for and extent of remedial actions. The Engineering
Evaluation presents three cleanup level options which should also be examined in
this manner. Each of the alternative remedial action proposals mentioned above
has been addressed with three cleanup levels for radium contamination of the
soil: 10 pCi/g above background, 5 pCi/g above background, and background.
Costs of these cleanup options have been listed in each case for each location
(FB78a). It is at this point that a health effects analysis would prove most
helpful in evaluating the cleanup options from a cost-benefit standpoint. By
applying an EDC calculation to the cleanup levels presented, a risk assessment
could be made which would allow decision making from a cost-effectiveness
approach.
A total population health effects determination would be made for each of
the suggested cleanup levels, applied to the 100-year period mentioned above.
Such a determination would then permit a systematic evaluation of the cleanup
level options in light of cost of implementation vs. reduction of health
effects.
In addition to a study of health effects at each of the suggested cleanup
levels, a second determination of health effects due to the storage options
should also be made. The Sampling Plant site has been proposed for both interim
storage (alternative 1) and long-term storage (alternative 2). While either of
these options would be able to reduce dose to the public by removal of sources
from public access to containment, there would still be some dose involved,
first in the physical cleanup process and later at a much reduced level from the
storage area. These options should similarly be examined to view the relative
speed with which they could be implemented, thereby considering total doses
involved.
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Recommendation
A health effects analysis should be utilized to assess the suggested
remedial actions for cleanup of the radiological contamination in and around the
Middlesex Sampling Plant site. Such an assessment would include an appraisal of
cost of each option compared to reduction of EDC. This would, in turn, give a
risk reduction for all population exposed to the contamination from the site
rather than a measure of dose reduction to the nearest off-site location. The
use of the EDC will allow a decision-maker to proceed with greater knowledge of
long-term risk and the cost associated with all options.
References
FB78a Ford, Bacon & Davis Utah, Inc., "Engineering Evaluation of the Former
Middlesex Sampling Plant and Associated Properties Middlesex, NJ,"
Draft Report, August, 1978 UC 230-001.
FB78b Ford, Bacon & Davis Utah, Inc., "Environmental Analysis of the Former
Middlesex Sampling Plant and Associated Properties Middlesex, NJ,"
Preliminary Draft, September 1978 FBDU 230-005.
MA74 Martin, J. A. Jr., C. B. Nelson, and P- A. Cuny, " A Computer Code for
Calculating Doses, and Ground Depositions Due to Atmospheric
Emissions of Radionuclides, U. S. Environmental Protection Agency
EPA 520/1-74-004, May 1974.
US73 U. S. Environmental Protection Agency, "Environmental Radiation Dose
Commitment: An Application to the Nuclear Power Industry," EPA-
520/4-73-002.
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ASSESSMENT OF RADON PROGENY INHALATION EXPOSURE
FROM LOW-LEVEL WASTES OF PHOSPHATE MINING IN FLORIDA*
Darrell R. Fisher
Pacific Northwest Laboratory
P. 0. Box 999
Richland, WA 99352
Charles E. Roessler
Department of Environmental
Engineering Sciences
University of Florida
Gainesville, FL 32611
Abstract
The redistribution of naturally-occurring uranium series radionuclides
as a result of phosphate mining, processing, product use, and waste disposal
presents several potential radiation pathways to man. Of particular impor-
tance is exposure to radon-222 progeny in structures built on reclaimed lands
in Florida.
We analyzed indoor radon daughter sampling data from Polk County, and
categorized the data by land and structure type. We determined the average
population-weighted concentration in about 4,400 homes to be about 0.009
working level (WL) in addition to a background of 0.003 WL. We also deter-
mined that the average annual cumulative indoor exposure on reclaimed land
was approximately 0.02 working level months (WLM). A relatively small num-
ber of houses on high-activity overburden accounted for 38% of the total
population exposure.
We are proposing a generally applicable model to relate lung cancer risk
to the average annual exposure, the risk coefficient, the expected lung can-
cer mortality from all other causes, the duration of the exposure and the num-
ber of years for observation of the effects. Health risk estimates were per-
formed for present levels and population size, and also for several scenarios
anticipating new growth and construction — with and without imposed standards
to limit indoor radon progeny levels. The model suggests that for an equilib-
rium condition, about one additional case of radiogenic lung cancer every two
years in the Polk County population might be expected.
Introduction
Greater than average terrestrial concentrations of uranium and radium are
associated with the phosphate rock matrix in Florida. The natural soil radium
content of undisturbed (unmined) land increases with depth from the surface
into the phosphate ore layer (Ro78) . When the phosphate rock is mined, the
overburden is removed by draglines and laid to one side. The matrix is then
removed and slurried with water to a washer/beneficiation plant. Sand
* Research performed at the University of Florida/Gainesville, under contract
with the Florida Phosphate Council.
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357
tailings (waste by-products) are pumped back to previously mined areas or
other disposal sites for use in land reclamation.
Prior to the development of the flotation process during the 1940's, the
fine (<1 mm dia.) fraction of phosphate mixed with sand was not recovered with
the pebble fraction, and was instead returned to the land as "debris." Later,
during land restoration operations, the debris piles were redistributed as new
land surfaces or fill.
Some unmined lands in the region with near-surface phosphate deposits
(little, if any, overburden) and elevated radium concentrations have been
identified. These have been termed by some as "mineralized" land (HRS78).
In addition, unmined lands with enhanced natural radioactivity due to fill or
other cover materials exist, and are referred to as "radioactive fill" lands.
Often the distinction between "fill" and "mineralized" lands is less than
adequate.
Thus, the redistribution of naturally-occurring uranium series radio-
nuclides following phosphate mining, processing, product use, waste disposal,
and land reclamation has increased the concentrations of near-surface radio-
activity. Although several possible pathways to man may result from the
technologically-enhanced radiation, the evaluation of radon progeny exposures
in structures on these lands (primarily in Polk County) is particularly
important.
The possibility of increased radiation-induced lung cancer from radon
daughter inhalation prompted the Environmental Protection Agency (EPA) to con-
sider the imposition of radiation standards specific to Florida homes and
phosphate-related lands, and to designate overburden, slimes, and tailings
from surface phosphate mining as "hazardous wastes" by reason of their radio-
activity content (FR78).
The objectives of this study were 1) to assess the distribution of popu-
lation exposures to indoor radon daughter concentrations, and 2) to estimate
the lung cancer risks to residents of Polk County, whose homes are built on
either reclaimed phosphate mine lands or unmined parcels with elevated soil
radium levels.
Methods
Population exposures. We analyzed the indoor radon progeny data that
were available for Polk County. Measurements were performed by the EPA (EPA75),
the State of Florida Department of Health and Rehabilitative Services (HRS;
HRS78), and the University of Florida College of Engineering (UF;UF78). We
combined the UF quarterly sampling data for 25 homes with the additional data
from a larger sampling of structures by the HRS to identify radiological char-
acteristics according to land and structure type. The data for mobile homes
resulted primarily from HRS measurements. Only a small amount of real data
was available from the UF concerning the "fill" land category and from the
HRS concerning "mineralized" lands. Since the radon progeny concentrations
reported for houses on these two land categories were similar, they were
combined into a single exposure analysis category.
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358
We then employed population statistics to estimate the distribution of
population exposures in Polk County. Less than 5% of the residences in Polk
County are situated on reclaimed land (HRS78). From the sampling data we
estimated that an additional 1% of all residential acreage (unmined) in the
County could be classified as "mineralized" near-surface deposits or fill
materials with elevated natural radioactivity.
Annual cumulative population exposures were estimated for each category
from multiplication of the mean indoor radon progeny concentration (WL, abovi
background) by an occupancy factor and a breathing rate correction factor
(for continuous indoor exposure)*
Risk analysis. The BEIR Report (NAS72) suggested that the mechanism of
radiogenic lung cancer is dependent upon metaplastic perturbations (lesions,
inflammations, etc.) in the bronchial epithelial tissues by nonradioactive
irritants. Studies on underground miners tend to support the hypothesis thai
a synergistic relationship exists between exposure to inhaled radon daughter
activity and other carcinogens or lung irritants (Ar76, He76), and that the
combined effect is multiplicative rather than additive (Do77) . It is there-
fore reasonable to presume that a given amount of radiation exposure to the
lungs will be more harmful to a subject whose expectation of cancer is alrea<
high, than to one whose expectation is lower, i.e., that a given amount of
lung exposure will not produce equivalent effects in all people.
We have proposed the following risk model to assist in the assessment o:
biological effects (lung cancer) from the inhalation of radon daughters. In
simplest terms, the risk model states that the additional cases of lung can-
cer will be proportional to the product of the natural incidence and the cumi
lative exposure. The constant of proportionality is the risk coefficient.
Lifetime risk per unit radiation (or cumulative exposure) was proposed as th<
best statistic for use in predicting lung cancers amoung populations exposed
to radon daughters in air (Ar78). The present approach involves considera-
tion of several time factors, including the latent period for lung cancer
induction, the duration of the exposure, and the follow-up (observation)
period. A linear, zero intercept dose response curve was assumed (Co78).
The expected number of lung cancer cases during a specified period of
time among an unexposed population group may be represented by the symbol Nz
The absolute risk (Nacjd) is the additional lung cancer mortality due to the
inhalation of radon decay products (above and beyond normal background leveL
(Ja73a). The total number of lung cancer observed (Ntot) in a population
exposed to the radiation hazard is therefore
Ntot = Nz + Nadd- (a)
The percent increase in risk relative to the natural risk is
Ir = 100 Nadd/Nz. (fe)
36 WLM/WL-yr. To estimate cumulative WLM, the assumption was made that a
third of one's total weekly inhalation is breathed occupationally Thus
3-12 = 36 "working-month equivalents" are breathed each year. We'also '
assumed an average occupancy factor of 0.7.
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359
Specifically, the expected number of "naturally occurring" cases of lung
cancer (Nz) is
Nz = P0 Z t2 (10) (c)
where Po is the population size, Z is the age-adjusted lung cancer mortality
rate (cases/106 person-yr), and t2 is the time frame or observation period (yr)
The additional risk is a function of the exposure level, the risk co-
efficient and the lung cancer mortality rate from all other causes. In a
previous report, Jacobi (Ja73a) expressed this relationship as
N'add(E) - arel • E • N'erw(E), (d)
where N'add(E) referred to the additional lung cancer mortality as a result of
a cumulative working level exposure (E), arej_ was the relative risk coefficient
or inverse doubling dose, and N'erw(E) represented the expected number of lung
cancer deaths. Jacobi 's equation can be restated more explicitly using new
symbols as
Nadd = K Wa tj N2 (e)
= K Wa t: P0 Z t2 (10~6), (f)
where K is the risk coefficient constant (WLM"1), Wa represents the average
annual exposure (WLM/yr) above background, and tj is the duration of the expo-
sure (yr). Wa is the product of the average annual indoor radon progeny con-
centration, the breathing rate correction factor, and the occupancy factor.
An appropriate expected lifetime should normally be used for the value of t2
the observation period. The product of Wa tj is the equivalent of the cumu-
lative working level month (CWLM) exposure unit frequently encountered in
some of the literature.
The risk model must be modified to account for a changing population size
if the analysis is performed over a lengthy period of time (Fi78). If Po is
the original population size and i is the annual rate of change, then the num-
ber of expected (natural) lung cancer mortalities in an expanding (or decreasing)
population for t2 years is thus
t2-l
Nz = P0 Z(10-5) 2-* (1 + i)n. (8)
n=0
and the anticipated number of additional lung cancers due to the radiation
component is
t-1
Nadd a K Wa Z P0(lCTe) \t^2 +
n=l
(t1 - n) (t2 - n) (• . (h)
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360
The total number of cancer mortalities from all causes combined (Ntot) is the
grand sum of equations (g) plus (h) . We assumed that the annual rate of
change, whether positive or negative, remains constant for t2 number of years,
and also that the values Nz and Nacjd are much less than the value Po (Fi78).
The primary source for dose response data for human lung cancer risk
following inhalation exposure to radon and radon daughters is the epidemiology
of underground miners. A review (Fi78) of absolute risk factors for under-
ground miners of various locations around the world resulted in the value
5 ± 4 (cases/106 person-yr-WLM) as the best estimate of the risk. Given this
risk per unit exposure and the natural incidence of lung cancer in a popula-
tion, the risk coefficient can be determined. The lung cancer incidence
rate (Z) for all males in Florida is 449 cases/106 person-yr (HEW74) . Assum-
ing that Florida males have essentially the same "natural" risk as the base
populations of the underground miners,* the risk co.efficient is
K = Nadd (WLM')/^ (i)
_ 5 (cases/ 10 6 person-yr-WLM) m ^^-i
449 (cases/106 person-yr) U'UJ"L WliM '
We therefore estimated the doubling dose (reciprocal of K) to be about 90 WLM,
which compares with other reported values of 100 WLM (Au76), 110 WLM (Ja73b),
and a value of 60 WLM used previously by the EPA (Mi77) .
The natural age-adjusted incidence (Z) by county and state in the U.S.
is available. The annual lung cancer incidence in Florida for members of the
general population (males and females) is about 244 cases/106 person-yr (HEW74),
The individual natural risk is strongly a function of age, sex, and smoking.
To tentatively delineate the impact and significance of radiation-induced
lung cancer in Polk County, four population exposure scenarios were developed.
For each of the scenarios a constant age distribution with time was assumed.
We accounted for an assumed 10-yr lung cancer latent period (Au76) by setting
an upper limit on the "effective" exposure period (t^ of t2 - 10 years.
Furthermore, we assumed the risk coefficient K to apply equally well to female
adults and children in the general population. Considering the many assump-
tions required and the limitations on the available data for radiation carcino-
genesis, the reader's attention should be directed towards relative differences
in calculated risk rather than on absolute numbers generated.
* Actually the natural or expected incidence of lung cancer among the male
population from which the fluorspar miners, and the American uranium and
metal miners were taken is somewhat lower: 416 cases/106 person-yr (derived
from NAS72). However, since a disproportionate number of miners were
smokers, their expected incidence rate could have been higher than this
Therefore, the Florida male cancer incidence was chosen for the present'
analysis of a Florida population. The incidence for all males in the U S
is 380 cases/106 person-yr (HEW74).
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Results
We found the average indoor radon progeny "background" concentration in
Florida homes to be about 0.003 WL (Fi78). Indoor levels were confirmed to
be significantly higher in structures on reclaimed land. A summary of indoor
radon progeny levels in Polk County homes by land and structure type is pre-
sented in Table 1. Since the radiological data were found to be log-normally
distributed, we reported the geometric mean rather than the arithmetic average.
Table 1.
Classifications of Polk County Structures and Their Indoor
Radon Progeny Levels; A Composite of UF and HRS Data
Exposure Category
Radon Progeny
Concentration, (WL)
Structure Type
Slab-on-grade
Land Type
Crawl Space Homes
and Mobile Homes
Undisturbed
Higher-activity
overburden and debris
Lower activity
overburden
Tailings
Unmined, near-surface
radioactive deposits
and fill
Undisturbed
Reclaimed
Mean*
(Range)
0.003 (0.001-0.010)
0.043 (0.019-0.140)
0.008 (0.004-0.018)
0.008 (0.002-0.038)
0.019 (0.003-0.045)
0.003 (0.001-0.010)
0.006 (0.001-0.014)
*geometric mean
(a)
(b)
(c)
(d)
(a) combination of UF "debris" lands and the higher-activity population
in the HRS "overburden" category.
(b) combination of the lower-activity population in the HRS "overburden"
category, and the UF type "overburden".
(c) HRS "mineralized" and UF "radioactive fill" land types in this cate-
gory are grouped together for convenience, and remain to be further
defined.
(d) all reclaimed or otherwise altered land types.
Approximate population, structure type, and population exposure (to
levels in excess of background) distributions are given in Table 2. We found
the population-weighted radon progeny mean concentration in about 4,400 homes
on disturbed lands to be about 0.009 WL above background. The corresponding
"excess" population exposure was thus estimated to be about 117 WL-persons
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362
above background. The average annual cumulative exposure (Wa) was found to
be 0.009 WL • 36 WLM/WL-yr • 0.7 = 0.02 WLM for a typical individual resident
Table 2. Estimated
Population Exposure from Elevated
Radon Progeny Concentrations
Land Category
A. RECLAIMED LAND
1 . S 1 ab-on-g rade :
High-activity overburden
and debris lands
Low-activity overburden
Tailings
2. Crawl space and mobile
homes: all reclaimed
Total:
B. RADIOACTIVE DEPOSITS,
AND FILL
1. Slab-on-grade
2. Crawl space and mobile
homes
Total:
C. UNDISTURBED LANDS
All structure types
Grand Total:
Excess Population
Estimated Exposure
Residences Persons WL-Persons %
322 1128 45.1 38
843 2951 14.8 13
995 3482 17.4 15
1440 3603 10.8 9
3600 11164 88.1 75
473 1656 26.5 23
316 790 2.4 2
789 2446 28.9 25
74468 262390
78857 276000 117.0 100
From Table 2 it can be seen that about 38% of the additional pupulation
exposure appears to be attributable to a relatively small fraction of resi-
dences on the higher-activity overburden or debris lands where the indoor
radon progeny concentrations were found to be distributed geometrically
around a value of 0.04 WL above background. Since contributions to the total
population exposure from the occupancy of public buildings and other non-
residential structures were small, they were not included in the exposure
analysis.
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363
Having established probable population size and exposure distribution
parameters, it was possible to estimate health risk implications, A summary
of population risks to lung cancer according to equations (g) and (h) for four
hypothetical 70-year scenarios is shown in Table 3. It can be seen that at
current indoor radon progeny levels, population size, age distribution and
number of dwellings, about 34 additional lung cancer deaths might be attri-
buted to increased levels of indoor airborne radioactivity. If distributed
evenly with time, the excess cancers might occur at the rate of one every
two years. For Polk County as a whole, the 34 added cases over a 70-year
period would represent an increase of 0.7% (too small to be detected statis-
tically). However, the additional cases represent an increase of approxima-
tely 14.5% in expected lung cancer mortality among a base population of
13,610 persons at risk on reclaimed land. In particular, for residents of
debris reclaimed land, 13 additional lung cancer deaths are estimated during
the 70-year period according to the first scenario, which corresponds to a
67% increase over their natural risk.
The possible effect of restrictions on indoor radon progeny levels can
be seen when scenarios three and for are compared. The result of decreased
indoor activity might prevent about 12 cases, or about one very six years.
Table 3. Theoretical Population Lung Cancer Risk Scenarios
for a 70-Year Exposure Period for Current and
Restricted Indoor Working Level Concentrations(a'
Population Lung Cancer Deaths
Scenario Level (avg.) Growth rate/year ^ Nadd Ntot
1 present 0% 232 34 266
(0.0086 WL)
2 present 1.5% first 30 332 47 379
(0.0086 WL) years, and 0%
next 40 years
3 present 1% 334 42 376
(0.0086 WL)
4 restricted(b) 1% 334 30 364
(0.0064 WL average
for existing struc-
tures and 0.0053 WL
for new construction)
(a) K = 0.011 WLM"1, Wa = 0.217 WLM/WL-yr, tx = 60 yr, t2 = 70 yr,
Z = 244 cases/106 person-yr, and Po = 13,610 persons.
(b) Limits of 0.02 WL for existing structures, and 0.01 WL for new structures.
Lifetime theoretical risks to individual residents (Po = 1) were deter-
mined (Fi78); these varied from 0.00001 to 0.219 depending on the indoor level,
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364
residency period, and smoking pattern. Ntot for the "average" individual
(Wa =0.02 WLM) was found to be 0.0175 (Ir = 2.4%) for 10-yr residency and
0.0195 (Ir = 14%) for lifetime residency on reclaimed land. For the average
resident on debris land (Wa =1.01 WLM), Ntot was found to be 0.019 (Ir = 11%)
for 10-yr residency and 0.029 (Ir = 67%) for lifetime residency.
Discussion
In general, most indoor radon progeny levels are low, and the population
health hazard is rather small. However, levels in some existing homes are un-
acceptably high, and these should be reduced. Debris category reclaimed lands
may not be suitable as future home sites, unless proper construction methods
are incorporated to limit indoor radon progeny levels.
The exercise of predicting health effects into the future is complicated
by uncertainties in utilization of reclaimed lands and population trends. It
is also difficult to predict future indoor radon progeny concentrations, since
the technology is available to reduce current levels.
It is quite possible that the risk coefficient for underground miners,
which we used in the risk analysis, is biased by other lung irritants charac-
teristic of mining atmospheres. Therefore, better human risk data for non-
miners is needed. For extension to the general public, the present risk
coefficient likely produces a cautious overestimate rather than a nearest
approximation of the biological effects from long-term low-level inhalation
exposure to radon and radon daughters.
References
Ar76 Archer V.E., Gillam D.J., and Wagoner J.K., 1976, "Respiratory Disease
Mortality Among Uranium Miners," Annals of the New York Academy of
Science 271: 280-293.
Ar78 Archer V.E., Radford E.P., and Axelson 0., "Radon Daughter Cancer in
Man: Factors in Exposure-response Relationships," presented at the
Health Physics Society 22nd Annual Meeting, Minneapolis, MN,
June 19-23, 1978.
Au76 Auxier J.A., 1976, "Respiratory Exposures in Buildings due to Radon
Progeny," Health Phys. 31: 119-125.
Co78 Cohen A.F., and Cohen B.L., 1978, "Tests of the Linearity Assumptions
in the Dose-effect Relationship for Radiation-induced Cancer," pre-
sented at the Health Physics Society 22nd Annual Meeting, Minneapolis,
June 19-23, 1978.
Do77 Douglas B., 1977, Occupational Health and Safety Newsletter 7(11): 5.
EPA75 U.S. Environmental Protection Agency, 1975, Preliminary Findings
Radon Daughter Levels in Structures Constructed on Reclaimed Florida
Phosphate Land, Technical Note ORP/CSD-74-4. ~ ~~
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Fi78 Fisher D.R., 1978, Risk Evaluation and Dosimetry for Indoor Radon Pro-
geny on Reclaimed Florida Phosphate Lands, Ph.D. Dissertation,
University of Florida.
FR78 Federal Register, 1978, "Hazardous Wastes," j43_(243): 58945-59028.
He76 Hewitt D., 1976, "Appendix C: Radiogenic Lung Cancer in Ontario
Uranium Miners 1955-74," in Report of the Royal Commission on the
Health and Safety of Workers in Mines, Province of Ontario, Toronto,
p 319-329.
HEW74 U.S. Department of Health, Education and Welfare, 1974, U.S. Cancer
Mortality by County: 1950-1969. National Cancer Institute, Publication
No. (NIH) 74-615 (Washington, DC, U.S. Government Printing Office).
HRS78 Department of Health and Rehabilitative Services, Radiological Health
Services, State of Florida, Study of Radon Daughter Concentrations
in Structures in Polk and Hillsborough Counties, January 1978.
Ja73a Jacobi W., 1973, "Lung Cancer Risk by Inhalation of Rn-222 Decay
Products," BNWL-TR-126 (English translation from Biophysik 10(2);
103-114) .
Ja73b Jacobi W., 1973, "Relation Between Cumulative Exposure to Radon Dau-
ghters, Lung Dose, and Cancer Risk," in Noble Gases Symposium, Stanley
R.E., and Moghissi A.A., eds. (CONF-730915), p 492-500.
Mi77 Mills W.A., Guimond R.J., and Windham S.T., 1977, "Radiation Exposures
in the Florida Phosphate Industry," Fourth International Radiation
Protection Association congress, April 24-30, 1977, Paris.
NAS72 National Academy of Sciences, Report of the Advisory Committee on
the Biological Effects of Ionizing Radiations, 1972, The Effects on
Populations of Exposure to Low Levels of Ionizing Radiation, National
Research Council.
Ro78 Roessler C.E., Kautz R., Bolch W.E., and Wethington J.A., 1978, "The
Effects of Mining and Land Reclamation on the Radiological Character-
istics of the Terrestrial Environment of Florida's Phosphate Regions,"
Trans. Third International Symposium on the Natural Radiation
Environment (NRE III), Houston, Texas, April 23-28, 1978.
UF78 University of Florida College of Engineering, 1978. Radioactivity of
Lands and Associated Structures, Final Report to the Florida Phosphate
Council (Volumes I, II, III, IV).
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RECOMMENDATIONS FOR REMEDIAL
ACTION AND DECOMMISSIONING
OF A RADIOACTIVE WASTE
BURIAL SITE
Paul A. Giardina, Radiation Branch, U. S. EPA Region II
Jeanette Eng, New Jersey Department of Environmental Protection
Joyce Feldman, Radiation Branch, U. S. EPA Region II
Abstract
For the past year, the Nuclear Fuel Services, Inc. (NFS) site located in
West Valley, NY has been the subject of state and federal efforts to determine
decontamination and decommissioning options. In 1978, the U. S. Environmental
Protection Agency (EPA) issued Criteria for Radioactive Wastes for storage and
disposal of all forms of radioactive wastes. Under an Atomic Energy Commission
(AEC) license, NFS operated the only commercial fuel reprocessing facility in
the United States. As a result of the reprocessing activities, the site
contains liquid high-level radioactive waste, buried cladding hulls and
defective fuel elements, a spent fuel storage pool, and a low-level burial
ground. Low-level radioactive material contained therein also comes from
sources other than NFS's operations. The site received a license from the State
of New York to perform low-level burial operations and radioactive material was
buried until 1975. Studies of the low-level burial area show radioactive gases
have leaked through the trench caps and the caps are more permeable than the
surrounding soil allowing water infiltration into the trenches. Active site
maintenance is used to prevent trench water overflow through the trench caps.
Other remedial actions have been described for the site and are undergoing
implementation. The West Valley site will be examined to determine the extent
of remedial action and decommissioning activities which may be necessary based
on the proposed EPA environmental criteria for radioactive waste.
Introduction
The State of New York authorized the establishment of a commercially
operated, low-level radioactive waste disposal area on part of a larger site
containing a facility for the reprocessing of nuclear fuel. A site in
Cattaraugus County approximately 30 miles southwest of Buffalo, New York known
as West Valley was licensed as a disposal site and in November 1963 the first
radioactive material was buried. Shallow trenches were dug to hold the waste
and an earthen cap (cover) was placed over each trench after it was filled with
the waste.
In the mid 1960s several burial trenches in the northern portion of the
site (trenches 1-7) began to fill with water shortly after they had been
covered. This posed a serious potential problem as the water could carry buried
radionuclides out of the trenches and into the environment. To eliminate or
reduce the water accumulation, burial procedures were changed for trenches in
the southern portion (trenches 8-13) of the site. The new procedures were
required by the State in 1968 and stipulated new trench capping methods were
used to prevent surface water from entering the trenches.
By 1974 three of the trenches in the north area had developed high levels
of water. Water levels in the south trenches, where modified capping procedures
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had been used, remained low. To date, no significant water accumulation has
occurred in the south trenches. In March 1975, water in one trench (trench 4) in
the north area seeped through the trench cap. The contaminated surface runoff
from this seepage was detected by the NYSDEC surveillance program and confirmed
during an onsite inspection. A similar seepage was noted shortly thereafter
along the west side of the cap on trench 5 (Ne75). Based on this occurrence, NFS
closed the burial site and has not reopened it.
During the summer of 1978 remedial action was taken to prevent further
water infiltration into the north burial trenches. The essence of this remedial
action was to place additional earth over the cover of these trenches. Eight
feet of dirt was placed over the trenches and compaction of the cover material
was done using bulldozers. To date eight feet of dirt cover now exists over all
burial trenches in the north and south portion of the site.
During the period from October 1963 through March 1975 more than 2,000,000
cubic feet of low-level radioactive waste were buried in the West Valley
trenches. Kilocurie quantities of strontium-90, tritium, and cobalt-60 have
been emplaced in the trenches along with such isotopes as radium-226, plutonium-
238, plutonium-239, uranium-233, uranium-235, thorium-232, and americium-241
(Ke73)
Several reports and papers (Gi77a, 6i77b, En77, USDOE78) have been
published detailing conditions at the low-level burial site at the West Valley
site. These works identify environmental pathways by which radioactivity buried
in the low-level trenches has been observed leaving the site. One such pathway
involves the formation of gases, radioactive in nature, through chemical
interactions in the burial trenches caused by reactions between water which has
infiltrated into the trenches and the waste material, some of which is organic.
Another pathway identified involves radioactively contaminated water
percolating through burial trench caps and running off to adjacent streams.
This pathway causes still another pathway involving the discharge of radioactive
liquid from the trenches into a nearby stream. These releases occur in a
controlled fashion after treatment has been performed on the radioactive
material. Finally, lateral migration of small quantities of tritium through the
strata adjacent to the trenches has also been observed, but to a very small
extent.
In November 1978 EPA published proposed Criteria for Radioactive Wastes in
the Federal Register (Co78). These criteria (a) define radioactive wastes,
indicate which types of wastes should be controlled, and give examples of where
these wastes originate; (b) state the goal of radioactive waste control and
define limitations on institutional and other controls over certain time
periods; (c) discuss the factors to be considered in assessing risk to the
general public and the general environment; (d) discuss the factors which would
result in unacceptable risk for different methods of disposal; (e) require that
the selection, design, and operation of a disposal site must enhance isolation
of radioactive wastes; and (f) discuss the appropriateness of retrievability of
waste and communication of waste disposal locations to future generations.
Analysis
By reviewing the conditions at the site and comparing the status of the
site to the proposed EPA criteria, determinations can be made as to potential
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remedial actions and decommissioning actions needed to assure that the low-level
site poses no unnecessary risk to public health and safety and to the
environment and that all low-level radioactive materials are ultimately
disposed of in an environmentally acceptable manner.
Proposed Criterion No. 1
Radioactive materials should be considered radioactive wastes requiring
control for environmental and public health protection if they have no
designated product or resource value and (a) are human-produced by nuclear
fission or activation, fabricated from naturally radioactive materials into
discrete sources, or as a result of regulatory activities are prohibited from
uncontrolled discharge to the environment; or (b) contain diffuse naturally-
occurring radioactive materials that, if disposed into the biosphere, would
increase exposure to humans above that which would occur normally in pathways
due to the preexisting natural state of the area. Examples of radioactive waste
materials that should be subject to environmental protection requirements are:
All radioactive materials associated with the operation and
decommissioning of nuclear reactors for commercial, military, research, or
other purposes and the supporting fuel cycles, including spent fuel if
discarded, fuel reprocessing wastes, and radionuclides removed from process
streams or effluents.
Artifically produced radioisotopes, including discrete radium sources,for
medical, industrial, and research use and waste materials contaminated with
them.
The naturally-radioactive residues of mining, milling, and processing of
uranium and phosphate ores.
This criterion addresses the issue of which materials should be considered
as radioactive waste.
In general radioactive material buried at the low-level site would be
considered radioactive waste based on this criterion.
Proposed Criterion No. 2
The fundamental goal for controlling any type of radioactive waste should
be complete isolation over its hazardous lifetime. Controls which are based on
institutional functions should not be relied upon for longer than 100 years to
provide such isolation; radioactive wastes with a hazardous lifetime longer than
100 years should be controlled by as many engineered and natural barriers as are
necessary.
This criterion addresses the issue of control of radioactive waste.
At the present time remedial action appears necessary to assure that this
criterion is attained. Complete isolation of the radioactivity contained in the
low-level burial area had not been accomplished before the summer of 1978.
Whether the remedial action taken during 1978 will be sufficient to meet this
criterion is as of yet unclear. Further actions such as using liners for the
trenches may or may not be necessary to isolate the waste for long periods of
time. However, it has not been positively demonstrated that liners would
achieve successful isolation either.
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Based on an inventory of material buried at the site (Ke73) one can
calculate the approximate hazardous lifetime of the radioactivity buried at the
site. If it is assumed that Ra-226, Am-241, and C-14 are the three isotopes
which will remain hazardous for the longest period of time, in 1,000 years 3.64
Curies, 3.85 Curies and 392 Curies of Ra-226, Am-241 and C-14 would remain
respectively. After 10,000 years 0.07 Curies and 133 Curies of Ra-226 and C-14
would remain. All the other isotopes buried on site should have activities
below one Curie by 1,000 years. These estimates neglect any radioactivity that
has left the site.
From this it can be seen that potentially hazardous levels of certain
nuclides will remain in the low-level burial site well after the 100 years
specified in criterion 2 as an upper limit for reliance on control by
institutional functions. Based on this it would seem clear that the remedial
actions taken to date be reviewed so as to assure that after 100 years the
material buried will remain isolated from man and the environment for a
sufficient period of time to allow certain long-lived radionuclides such as Ra-
226, Am-241, and C-14 to decay away to innocuous levels. If the current
remedial actions which have or will be implemented at the site cannot meet this
isolation and control criterion, further actions should be prescribed.
Proposed Criterion No. 3
Radiation protection requirements for radioactive wastes should be based
primarily on an assessment of risk to individuals and populations; such
assessments should be based on predetermined models and should examine at least
the following factors:
a. The amount and concentration of radioactive waste in a location and
its physical, chemical, and radiological properties;
b. The projected effectiveness of alternative methods of control;
c. The potential adverse health effects on individuals and populations
for a reasonable range of future population sizes and distributions,
and of uses of land, air, water, and mineral resources for 1,000
years, or any shorter period of hazard persistence;
d. Estimates of environmental effects using general parameters or of
health effects based on generalized assumptions for as long as the
wastes pose a hazard to humans, when such estimates could influence
the choice of a control option;
e. The probabilities of releases of radioactive materials to the general
environment due to failures of natural or engineered barriers, loss of
institutional controls, or intrusion; and
f. The uncertainties in the risk assessments and the models used for
determining them.
This criterion addresses the issue of risk assessment.
To meet this criterion several actions should be undertaken. These are as
follows:
a. Using existing inventories and data from ongoing studies, the
physical and chemical properties of the waste should be reviewed to
determine any unique characteristics which would lead to future
problems in isolating the waste.
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b. Alternative methods of control should be reviewed in the future as new
methods become available. Existing methods which have not been
considered, if any, should also be reviewed.
c. The potential adverse health effects on individuals and populations
for reasonable future population sizes and distributions should be
estimated for a period of up to 1,000 years. This should probably be
in the form of a risk assessment and should address future land, air,
water, and mineral resource uses.
d. Should the material buried at the site be shown to have a hazard
potential greater than 1,000 years, an effort should be made to define
this potential in terms of environmental and health effects for the
hazardous lifetime of the waste.
e. The probabilities of release of the radioactive material buried at the
site to the general environment due to failures of natural or
engineered barriers, loss of institutional controls, or intrusion
should be quantified.
f. The uncertainties in the analyses used to determine matters discussed
in "a" through "e" above should be addressed.
Should the results of any of these undertakings listed in "a" through "f"
show unnecessary risks, further remedial actions above what is currently being
done will be necessary.
Proposed Criterion No. 4
Any risks due to radioactive waste management or disposal activities should
be deemed unacceptable unless it has been justified that the further reduction
in risk that could be achieved by more complete isolation is impracticable on
the basis of technical and social considerations; in addition, risks associated
with any given method of control should be considered unacceptable if:
a. Risks to a future generation are greater than those acceptable to the
current generation;
b. Probable events could result in adverse consequences greater than
those of a comparable nature generally accepted by society; or
c. The probabilities of highly adverse consequences are more than a small
fraction of the probabilities of high consequence events associated
with productive technologies which are accepted by society.
This criterion addresses the issue of unacceptable risk.
To fulfill this criterion three determinations should be made:
a. Risks to future generations will be no greater than those acceptable
by this generation. This does not mean risks must be equal to or less
than those imposed on current generations. They could be greater as
long as the current generation would accept the risk.
b. Probable events will not create adverse consequences greater than
those of a comparable nature generally accepted by society.
c. Probabilities of highly adverse consequences are no more than a small
fraction of the probabilities of high consequence events which are
associated with productive technologies and which are accepted by
society.
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Should it be impossible to make reasonably assured affirmative
determinations on these points it will not be possible to make a positive
determination on the acceptability of the risk at the West Valley site and
further remedial attention will be needed to asssure this criterion is attained
and the risk is acceptable.
Proposed Criterion No. 5
Locations for radioactive waste disposal should be chosen so as to avoid
adverse environmental and human health impacts and, wherever practicable, to
enhance isolation over time.
This criterion addresses location and waste isolation.
Since the site currently exists, it is questionable whether a criterion
involving siting is applicable. If it is found that the site for reasons of
geology, hydrology, and meteorology cannot avoid adverse environmental and
human health impacts and that removal of the waste and disposal in another place
is less risky, then this criterion might apply and might cause evaluation of
this alternative (the authors note that there does not seem to be any
substantial evidence supporting this premise at this time). However, this
criterion should be closely scrutinized if the site is considered for future
disposal.
Proposed Criterion No. 6
Certain additional procedures and techniques should also be applied to
waste disposal systems which otherwise satisfy these criteria if use of these
additional procedures and techniques provide a net improvement in environmental
and public health protection. Among these are:
a. Procedures or techniques designed to enhance the retrievability of
the waste; and
b. Passive methods of communicating to future people the potential
hazards which could result from an accidental or intentional
disturbance of disposed radioactive wastes.
This criterion addresses supplementary protection goals.
Two areas of remedial action are evident from this criterion. One involves
developing the best possible monitoring system around the site with the goal of
detecting potential problems before they become hazardous. The other involves
designing better passive means of communicating the potential hazards to future
generations. The idea of retrievability of waste does not seem feasible since
the material is already buried. However, should the site be considered for use
in the future it would be wise to consider upgrading waste containerization to
allow retrievability for as long a period of time as is reasonably possible.
The past history of the West Valley low-level radioactive waste burial site
reveals that radioactivity has been and can be released to the environment.
Remedial actions have been undertaken to alleviate this problem. Work must now
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be done to assess the hazard potential of the radioactive material over time, to
assure that the risk attributable to the site is adequately assessed, to assure
the acceptability of the risk, and to assure that the controls at the site are
adequate. This work may show that more remedial attention is needed at the site
in the form of engineered barriers and safeguards to meet the proposed EPA
criteria for radioactive waste. Upgrading monitoring and passive
communications safeguards also seems desireable.
References
Gi77a Giardina, P. A., De Bonis, M. F., Eng, J., Meyer,G.L., "Summary
Report on the Low-Level Radioactive Waste Burial Site, West Valley, New York
(1963-1975), EPA-902/ 4-77-010, February, 1977.
Gi77b Giardina, P. A., "Preliminary Pathway Observations of Radionuclide
Movement to the Environment from Low-Level Radioactive Waste Disposal Site in a
Humid Climate," Presented at the Twenty-Second Annual Meeting of the Health
Physics Society, July 3-8, 1977.
USDOE78 U. S. Department of Energy, Western New York Nuclear Service
Center Study Volume 2, TID-28905-2, December, 1978.
En77 Eng, J. Giardina, P. A., "Investigation of Gas Formation on a Low-
Level Radioactive Waste Disposal Site," Proceedings of the Fifth National
Conference, Energy and the Environment American Institute of Chemical
Engineers, November 1977.
Co78 Costle, D. M., "Criteria for Radioactive Wastes," Federal Register.
Vol. 43, No. 221, November 15, 1978, pp. 53262-68.
Ne75 New York State Department of Environmental Conservation,
"Radioactivity in Air, Milk, and Water for Jan-Mar 1975," Environmental
Radiation Bulletin, November, 1975.
Ke73 Kelleher, W., and Michael, E., "Low-Level Radioactive Waste Burial
Site Inventory for the West Valley Site, Cattaraugus County, New York," 1973.
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ACCEPTABLE RESIDUAL RADIOACTIVE CONTAMINATION LEVELS
FOR SITES OF DECOMMISSIONED NUCLEAR FACILITIES
W. E. Kennedy, Jr., R. B. McPherson and E. C. Watson
Pacific Northwest Laboratory
Richland, Washington 99352
Abstract
Examination of existing guidelines and regulations has led to the conclu-
sion that there is need for a general method to derive residual contamination
levels that should be acceptable to the public in return for use of any site
of decommissioned nuclear facilities. The method used for this study is to
determine these acceptable environmental levels based on a maximum annual dose
to an individual from the residual contamination via all environmental path-
ways. A maximum annual dose criterion of one mrem is selected for purposes
of illustrating the method. Acceptable residual radioactive contamination
levels for the sites of three current-design nuclear facilities are presented.t
The reference facilities considered are: a Fuel Reprocessing Plant (FRP), a
Pressurized Water Reactor (PWR), and a small Mixed Oxide Fuel Fabrication
Plant (MOX). Using maximum annual dose as a basis permits an accurate account-
ing of the impact of radionuclide mixtures at each unique nuclear site. The
results presented consider all probable radiation exposure pathways contribut-
ing significantly to the maximum annual dose to an individual from chronic
exposure to the residual radioactive contamination on the site. For chronic
radiation exposure, the year in which the maximum annual dose occurs depends
upon the chemical and physical characteristics of the residual radionuclides,
the body organ of reference, and the radiation exposure pathway. The accept-
able residual radioactive contamination levels are based on the calculated
maximum annual dose resulting from the radionuclide mixture accumulated from
a calculated annual release during the facility operating lifetime. The
results of this study show acceptable radioactive contamination levels at
plant shutdown in units of microcuries per square meter of 5.0 x 10~^ for
the FRP, 1.2 x 10~2 for the PWR, and 1.3 x 10~2 for the MOX.
Introduction
Our examination of existing guidelines and regulations has led us to the
conclusion that there is a need for a general method to derive acceptable
radioactive contamination levels that can be applied for the release of any
decommissioned nuclear site for public use (Sc77, Sm78, Je79) . Some guidance
currently exists defining the levels of radioactive surface contamination
which are acceptable to the U. S. Nuclear Regulatory Commission for the termi-
nation of operating licenses (NRC74, AEC70). Other suggested guidance is
directed toward specific types of nuclear facilities, or accident situations
involving radioactivity (CFR49-76, ERDA75, He74, Gu64, Ha75, ANSI78).
None of these guidelines is sufficiently flexible to accommodate the
various radionuclide mixtures or site specific features unique to each nuclear
t Based on work prepared for the Division of Engineering Standards, Office
of Standards Development of the U. S. Nuclear Regulatory Commission.
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facility. We believe that the methodology used to calculate acceptable residual
radioactive contamination levels should be based on a more general concept; a
concept that is capable of accommodating these unique radionuclide mixtures
and site specific features. One general concept for determining acceptable
radioactive contamination levels is to compare established annual dose limits
with calculated annual doses to members of the public.
There are currently no unique regulations or specific guides on acceptable
maximum annual dose to individuals living on or near a decommissioned site.
Guidance that could be interpreted as annual dose limit recommendations speci-
fically for the cases of interest here include:
• Recommendations of the International Committee on Radiation
Protection (ICRP), Publication 9 (ICRP66)
• Surgeon General's Guidelines (DREW) (PHS71)
• Appendix I of 10 CFR 50, Guides for Design Objectives for Light-Water-
Cooled Nuclear Power Reactors (NRG) (CFR10-76)
• Proposed Federal Guidance for the Environmental Limits of Transuranium
Elements (EPA) (EPA77)
• 40 CFR 90 Environmental Radiation Protection Requirements for Normal
Operations of Activities in the Uranium Fuel Cycle (NRC) (CFR40-77)
The purpose of this paper is to describe our methodology for determining
acceptable residual radioactive contamination levels based on the concept of
limiting the annual dose to members of the public. It is not our purpose to
recommend or even propose dose limits for the exposure of members of the public
to residual radioactive contamination left at decommissioned nuclear sites.
Thus, example acceptable levels of residual radioactive contamination are
calculated only to demonstrate the annual dose-based methodology for an assumed
annual dose of one millirem. The reference sites we considered are a Fuel Re-
processing Plant (FRP), Pressurized Water Reactor (PWR), and a small Mixed-
Oxide Fuel Fabrication Plant (MOX).
The following terminology is used in developing the annual dose-based
methodology:
• Disposition Criteria The acceptable radioactive contamination levels
for public use of decommissioned nuclear sites, based on a maximum
annual dose limit.
• Organs of Reference Radiation doses are calculated for specific organs
of the human body. In this study these organs of reference are the
thyroid glands, lungs, total body, and bone.
- Exposure Pathways The radiation exposure pathways represent ways by
which people are exposed to radiation. Exposure pathways of concern
in calculating the dose to members of the public located on a decom-
missioned nuclear site are inhalation of radionuclides, external exposure •
from radioactive surface contamination, and the ingestion of food pro-
ducts containing radionuclides.
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• Decay Periods The continually changing mixture of the radionuclide
inventories results in annual doses that are time-dependent due to
radioactive decay. This dependence is demonstrated by calculating the
doses at the time of the facility shutdown, and at 10, 30, and 100
years after shutdown.
• Annual Dose The annual dose is the radiation dose equivalent calculated
during any year following continuous exposure. It is the sum of the
doses received during the year of interest from all exposure pathways
including the dose resulting in that same year from the intake of radio-
nuclides during previous years. The highest value calculated for any
year following the unrestricted public release of the site is referred
to as the maximum annual dose.
• Maximum Exposed Individual The maximum exposed individual is assumed to
reside at the location of the highest airborne radionuclide concentration,
and maximized exposure pathway parameters are used.
• Unrestricted Use The unrestricted use of the decommissioned site means
that the potential exposure to members of the public from residual radio-
active contamination levels will not exceed the maximum annual dose limit
as may be established by Federal and state regulatory agencies. Decom-
missioning a site will, in general, result in the unrestricted public
use of land areas that the public had been denied use of during the
nuclear facility operational life.
Maximum Annual Radiation Dose
The annual dose from radiation exposures originating from sources external
to the human body tends to decrease with time after shutdown due to radiative
decay of the residual radioactivity. Annual doses to organs of reference re-
sulting from chronically ingested or inhaled radionuclides, however, tend to
increase with time after shutdown until a maximum value is reached. The annual
dose from internal sources then tends to decrease with time due to both radio-
active decay of the residual radioactivity and biological elimination of radio-
nuclides deposited in the organ of reference. For continuous exposure to a
radioactively decaying source, the year in which this maximum annual dose occurs
depends on the chemical and physical characteristics of the radionuclides, the
organ of reference and the environmental pathways considered.
A fundamental relationship for the calculation of radiation dose to man
from environmental pathways for any radionuclide is given as follows:
R. = C. U D. (1)
ipr ip p ipr
where:
R. • the radiation dose equivalent or committed dose
equivalent from nuclide i via pathway p to organ r;
• the concentration of nuclide i in the media of pathway p;
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376
U • the exposure rate or intake rate associated with
pathway p; and
D. -a radiation dose equivalent or committed dose equiva-
lpr lent factor: a factor for a given nuclide i, pathway p
and organ r that converts a specified concentration of
the radionuclide and the intake rate of that radio-
nuclide to the radiation dose equivalent or committed
dose equivalent.
Specific equations tailored to each exposure pathway are derived from
Equation 1. A more complete discussion of the radiation dose model, the varia-
tion of annual dose with time after shutdown, and the parameters used can be
found in the literature (Sc77, Sm78, Je79). The principal difference among
pathways is the manner in which the radionuclide concentrations in air, soil,
or food products are calculated as an integral part of the computerized models
used in this study (So74, Ba76). They are functions of such parameters as the
radionuclide release rates, resuspension rates, deposition rates, root uptake
parameters, and atmospheric dispersion.
Disposition Criteria Methodology
The methodology we used to determine disposition criteria based on annual
dose is shown in Figure 1. The three steps in this methodology are:
• Compute the Maximum Annual Doses The maximum annual doses to the organs
of reference resulting from the radioactive contamination present at
the radioactive decay times of interest are calculated using the radia-
tion dose methodology discussed previously.
• Compute the Contamination Levels The residual contamination levels,
or disposition criteria, expressed in units of microcuries per square
meter (yCi/m2) are calculated for the organs of reference using a normal-
ized annual dose value of 1 mrem.
• Determine the Maximum Acceptable Contamination Level The maximum accept-
able contamination level at the assumed maximum annual dose is determined
by selecting the most restrictive calculated organ dose derived from all
exposure pathways. This value is dependent on the composition of the
radionuclide inventory.
Example Calculations
Radioactive contamination is expected to be present on the site as a
result of effluents released during normal operations over the anticipated
plant life. The concentration of deposited radionuclides was estimated using
an NRC computer model (Sa76). Continuous annual releases at a constant level
were assumed over the plant life. We assumed the effective release height
to be 100 meters for the FRP and 10 meters for the PWR and MOX. Deposition
values reported are for the area within the site boundary (the area within
a radius of 1,000 meters from the point of release). Offsite contamination
levels are expected to be lower.
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The resulting surface contamination compositions for the reference FRP,
PWR and MOX are shown in Tables 1, 2 and 3 for various times after plant shut-
down. It should be noted that the contamination levels defined for the site
by Tables 1 through 3 are probably higher than might be encountered at real
facilities. This is primarily because no credit was taken for weathering
effects on the radioactive contamination either during the facility operating
life or during the time after shutdown. For specific sites, comprehensive
measurements will be necessary at shutdown to characterize the quantity and
mixture of the deposited radioactive contamination.
Maximum annual doses calculated for the site radionuclide inventories
(Step 1 in Figure 1) are listed in the literature (Sc77, Sm78, Je79). Accept-
able contamination levels for a dose of 1 mrem/yr are calculated (Step 2 of
Figure 1), and disposition criteria based on the most restrictive annual dose
to any organ of reference (Step 3 of Figure 1) are determined. These values
are listed in Table 4 in units of yCi/m2. The dominant radionuclide contrib-
utor to the organ doses are listed in Table 4 to help illustrate the dependence
of the calculated doses on the assumed radionuclide mixture.
For the FRP, the disposition criteria are controlled at all times after
shutdown by the dose to the thyroid gland because of the dominance of 129j
in the radionuclide mixture. For the PWR, because of 90Sr in the radionuclide
mixture, the deposition criteria are controlled by the dose to bone at all
times after shutdown. The primary pathway for both 129i and 90sr deposition
in humans is the consumption of leafy vegetables grown on the site.
The change in limiting organ dose from lungs to bone between 0 and 10
years after shutdown for the MOX reflects the impact of the time-dependent
resuspension factor used in our analysis. We assumed that the quantity of
material available for resuspension decreased with time after ground deposition
using the Anspaugh model (An75) until a constant value is reached around 20
years after deposition. With more material resuspended at short times after
shutdown, the dose to the lungs is limiting. As the amount of resuspended
material decreases, the translocation of material from the lung and GI-tract
to the bone controls the annual dose.
Summary and Conclusions
The methodology that we have presented in this paper can be used to calcu-
late defensible acceptable residual contamination levels that are directly
relatable to risk assessment with the proviso that an annual dose limit will
be established. Our methodology is shown to be flexible enough to permit
variations in the radionuclide mixtures and site specific data required in
the calculation of the annual doses. The maximum annual dose should be used
for comparison to an annual dose limit, and not a 50-year committed dose
equivalent or the first year dose. The first year dose is not conservative
when internal exposure is the dominant pathway, and it is not appropriate to
compare a 50-year committed dose to an annual dose limit.
-------
378
References
AEC70 U.S. Atomic Energy Commission, 1970, Guidelines for Decontamination of
Facilities and Equipment Prior to Release for Unrestricted Use or Termina-
tion of Licenses for By-Product, Source or Special Nuclear Material.
AN75 Anspaugh, L. R., Shinn, J. H., and Phelps, P- L., 1975, Resuspension and
Redistribution of Plutonium in Soils, UCRL-76419, 14-18.
ANSI78 ANSI Standard N328, 1978, Control of Radioactive Surface Contamination
on Materials, Equipment and Facilities to be Released for Uncontrolled Use,
published for ANSI national trial and use.
Ba76 Baker, D. A. Hoenes, G. R., and Soldat, J. K., 1976, "FOOD - An Inter-
active Code to Calculate Internal Radiation Doses from Contamination Food
Products," Environmental Model-Ing and Simulation, Proceedings of a Con-
ference held in Cincinnati, OH (April 20-22, 1976) EPA, Washington, DC,
204-208.
CFR10-76 U.S. Code of Federal Regulations, Title 10, Part 50, Appendix I,
1976, "Licensing of Production and Utilization Facilities," Superintendent
of Documents, GPO, Washington, DC 20402.
CFR40-77 U.S. Code of Federal Regulations, Title 40, Part 190, 1977, "Environ-
mental Radiation Protection Standards for Nuclear Power Operations," Superin-
tendent of Documents, GPO, Washington, DC 20402.
CFR49-76 U.S. Code of Federal Regulations, Title 49, Part 173, 1976, "Trans-
portation," Superintendent of Documents, GPO, Washington, DC 20402.
EPA77 U.S. Environmental Protection Agency, 1977, Proposed Guidance on Dose
Limits for Persons Exposed to Transuranium Elements in the General Environ-
ment, EPA 520/4-77-016.
ERDA75 U.S. Energy Research and Development Administration, 1975, "Prevention
Control and Abatement of Air and Water Pollution," U.S. ERDA Manual,
Chapter 0510.
Gu64 Guthrie, C. E. and Nichols, J. P., 1964, Theoretical Possibilities and
Consequences of Major Accidents in 233U - 239pu Fuel Fabrication and Radio-
isotope Processing Plants, ORNL-3441, Oak Ridge National Laboratory, Oak
Ridge, TN 37830.
>
Ha75 Hazle, A. J. and Crist, B. L., 1975, Colorado's Plutonium-Soil Standard,
Colorado Department of Health, Occupational and Radiological Health Division,
Denver, CO..
He74 Healy, J. W., 1974, A Proposed Interim Standard for Plutonium in Soils,
LA-5483-MS, Los Alamos Scientific Laboratory, Los Alamos, NM.
ICRP66 International Commission on Radiological Protection, 1966, "Recommendations
of the International Commission on Radiological Protection," ICRP Publication .
9, Pergamon Press, London.
-------
379
Je79 Jenkins, C. E., Murphy, E. S. and Schneider, K. J., Technology, Safety
and Costs of Decommissioning a Reference Small Mixed-Oxide Fuel Fabrication
Plant, NUREG/CR-0129, U.S. Nuclear Regulatory Commission Report by Pacific
Northwest Laboratory, Richland, WA 99352.
NRC74 U.S. Nuclear Regulatory Commission, 1974, Termination of Operating
Licenses for Nuclear Reactors, Regulatory Guide 1.86.
NRC76 U.S. Nuclear Regulatory Commission, 1976, Final Generic Environmental
Statement on the Use of Recycled Plutonium in Mixed-Oxide fuel in Light-
Water-Cooled Reactors, NUREG-0002, Vol. 3.
PHS71 Surgeon General, U.S. Public Health Service, Surgeon General's Guide-
lines, 1971, "Use of Uranium Mill Tailings for Constructive Purposes."
Hearings before the Subcommittee on Raw Materials of the Joint Committee
on Atomic Energy, October 28 and 29, 1971, 52-54.
Sa76 Sagendorf, J. F. and Goll, J. T., 1976, XOQDOQ - Program for the Meteoro-
logical Evaluation of Routine Effluent Releases at Nuclear Power Stations,
Draft NRC Report.
Sc77 Schneider, K. J. and Jenkins, C. E., Study Coordinators, 1977, Technology,
Safety and Cost of Decommissioning a Reference Nuclear Fuel Reprocessing
Plant, NUREG-0278, Report of U.S. Nuclear Regulatory Commission, by Battelle
Pacific Northwest Laboratory, Richland, WA 99352.
Sm78 Smith, R. I., Konzek, G. J. and Kennedy, W. E., Jr., Study Coordinators,
1978, Technology, Safety and Costs of Decommissioning a Reference Pressurized
Water Reactor Station, NUREG/CR-0130, Report of U.S. Nuclear Regulatory Com-
mission by Battelle Pacific Northwest Laboratory, Richland, WA 99352.
So74 Soldat, J. K., Robinson, N. M. and Baker, D. A., 1974, Models and Computer
Codes for Evaluating Environmental Radiation Doses, BNWL-1754, USAEC Report,
Battelle, Pacific Northwest Laboratories, Richland, WA 99352.
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380
Table 1. Estimated Maximum Radioactivity Deposited
on the FRP Site Over a 30-Year1 Operating Lifetime
Deposited Radioactivity (yCi/m2)
Selected Times After Shutdown
Radionuclide
89Sr
90Sr
90Y
91Y
95Zr
95Nb
103Ru
106Ru
110A
125Sb
127Te
129Te
129j
131Z
13l+Cs
137Cs
141Ce
lkkCe
147Pm
15ltEu
155Eu
231+U
235U
236U
238U
238pu
239pu
2t|°Pu
241pu
2<42pu
21tlAm
21t3Am
242Cm
241tCm
Shutdown
1.5E-5
1.4E-3
1.4E-3
2.8E-5
5 . 5E-5
5.5E-5
4.9E-5
2 . 4E-3
2.0E-6
2.7E-5
2.1E-6
2.5E-7
6.4E-3
1.6E-5
5.5E-4
2.0E-3
5.1E-6
7.4E-4
3.1E-4
6.8E-5
3.7E-5
2.0E-7
4.4E-9
7.4E-8
8.0E-8
1.6E-4
2. OE-5
3.1E-5
2.7E-3
8.6E-8
l.OE-4
4.6E-7
9 . 4E-6
3.6E-5
10 Years
— (a)
1.1E-3
1.1E-3
__
_._
— _
__
2.6E-6
8.8E-11
2.1E-6
1.7E-16
6.4E-3
—
1.9E-5
1.6E-3
—
9.9E-8
2.2E-5
3. OE-5
8.7E-6
2.0E-7
4.4E-9
7.4E-8
8.0E-8
1.5E-4
2. OE-5
3.1E-5
1.3E-3
8.6E-8
1.5E-4
4.6E-7
1.7E-12
2.4E-5
30 Years
__
6.9E-4
6.9E-4
—
__
__
2.9E-12
__
1.3E-8
__
__
6.4E-3
__
2.3E-8
9.9E-4
—
1.8E-15
1.1E-7
6.0E-6
4.8E-7
2.0E-7
4.4E-9
7.4E-8
8.0E-8
1.3E-4
2 . OE-5
3 . 1E-5
3.0E-4
8.6E-8
1.9E-4
4.6E-7
__
1.1E-5
at
100 Years
__
1.3E-4
1.3E-4
—
__
__
__
__
__
6.4E-3
__
— —
2.0E-4
__
l.OE-15
3.1E-8
1.9E-11
2.0E-7
4.4E-9
7.4E-8
8.0E-8
7.3E-5
2. OE-5
3. OE-5
1.9E-6
8.6E-8
1.9E-4
4.6E-7
7.4E-7
(a) Dash indicates deposition is less than 10~15
-------
Table 2. Estimated Maximum Radioactivity Deposited
on the PWR Site Over the 40-Year Plant Lifetime
from GESMO (NRC76)Study Annual Releases(a)
381
Deposited Radioactivity (yCi/m2)
at Selected Times After Shutdown
Radionuclide
5"Mn
58Co
6°Co
59Fe
89Sr
90Sr
90Y
131X
133];
134Cs
137Cs
Shutdown
4.2E-4
1.5E-4
5.3E-3
6.3E-6
7.5E-6
9.0E-4
9.0E-4
2 . 8E-5
2.8E-6
5.2E-4
9.3E-3
10 Years
9 . 1E-8
—
1.4E-3
__
—
7 . OE-4
7 . OE-4
—
__
1.8E-5
7.4E-3
30 Years
4.3E-15
__
9.8E-5
__
__
4.2E-4
4.2E-4
— _
__
2.1E-8
4.7E-3
100 Years
__(b)
__
8.7E-9
__
__
7 . 1E-5
7 . 1E-5
__
__
—
9.3E-4
(a) Normalized to the reference PWR power rating of 1175 MWe and plant
capacity factor of 0.75.
(b) A dash indicated values less than 10~15 yCi/m2.
-------
Table 3. Estimated Maximum Radioactivity Deposited
on the MOX Site Over a 10-Year Operating Lifetime
382
Deposited Radioactivity (pCi/m2)
at Selected Times After Shutdown
Radionuclide
234Th
233pa
234u
23 5n
23 GU
237u
238u
237Np
238pu
239pu
240pu
2<+lpu
242pu
2"lAm
Shutdown
1.5E-4
2.4E-6
3.0E-4
7 . OE-6
6.2E-7
4.5E-3
1.5E-4
2.4E-6
8.8E-1
7 . 2E+0
3 . 5E+0
1.8E+2
8.0E-4
1.9E+0
10 Years
__(a)
2.4E-6
3.0E-4
7. OE-6
6.2E-7
—
1.5E-4
2.4E-6
8.1E-1
7 . 2E+0
3 . 5E+0
1.1E+2
8.0E-4
4.1E+0
30 Years
2.4E-6
3.0E-4
7 . OE-6
6.2E-7
—
1.5E-4
2.4E-6
7.0E-1
7 . 2E+0
3 . 5E+0
4.2E+1
8.0E-4
6.3E+0
100 Years
__
2 . 4E-6
3.0E-4
7. OE-6
6.2E-7
—
1.5E-4
2 . 4E-6
4.0E-1
7 . 2E+0
3 . 5E+0
1 . 5E+0
8.0E-4
6 . 8E+0
(a) A dash indicates values less than 10 7 pCi/m2.
-------
383
Table 4. Disposition Criteria on the Reference
FRP, PWR3 and MOX Sites Corresponding to a
Maximum Annual Dose of 1 mrem per
FRP
PWR
MOX
Class Y
Time After
Shutdown
Years
0
10
30
100
0
10
30
100
0
10
30
100
Limiting
Organ
Thyroid
Thyroid
Thyroid
Thyroid
Bone
Bone
Bone
Bone
Lungs
Bone
Bone
Bone
Dominant
Radionuclide
129j
129X
129];.
129]-
9°Sr
90Sr
90Sr
9°Sr
239pu
239pu _ 2«tlAm
239Pu _ 2klM
239Pu - 241^
Disposition
Criteria
UCi/m2
0.0050
0.0038
0.0029
0.0020
0.012
0.0096
0.0091
0.0098
0.013
0.49
0.23
0.077
(a) Includes doses from inhalation,external exposure, and ingestion of
food products.
-------
• CONTAMINATION CHARACTERISTICS
'LAND AND FACILITY CONDITIONS
'POSTULATED SCENARIOS FOR
RELEASE
STEP 1: COMPUTE THE MAX I MUM ANNUAL
DOSES FOR THE REFERENCE
RADIONUCLIDE INVENTORIES
ASSUMED ANNUAL DOSE
LIMIT OF 1 mrem/yr
STEP 2: COMPUTE CONTAMINATION LEVELS
THAT PRODUCE THE ANNUAL DOSE
VALUES OF 1 mrem/yr;
REPORT INUNITSOFpCi/m2
STEP 3: DETERMINE THE MAXIMUM
ACCEPTABLE CONTAMINATION
LEVEL AT 1 mrem/yr,
USiNG THE MOST RESTRICTIVE
COMBINATION OF EXPOSURE
PATHWAYS AND ORGAN DOSES
(DISPOSITION CRITERIA)
Fig. 1. Disposition Criteria Methodology
-------
385
DOSIMETRIC AND RISK/BENEFIT IMPLICATIONS OF Am-241
IN SMOKE DETECTORS DISPOSED OF IN NORMAL WASTES
McDonald E. Wrenn and Norman Cohen
Institute of Environmental Medicine
New York University Medical Center
550 First Avenue
New York, New York 10016
Abstract
The risks associated with 2'*1Am in residential-type smoke detectors
have been assessed (using a linear dose-response model) by evaluating
collective dose commitments (70 years) to the public delivered during normal
detector use and disposal. The risk of radiation induced harm results mainly
from external exposure, with only a small contribution from waste management
practices. A reference case of 100 million detectors distributed in residences
at a rate of 10 million per year leads to an external dose of 340 man-rems per
year to the U.S. population and a cancer risk of 0.07 cases per year. Collective
background is 20 million man-rems per year. Lives saved by detectors from fires
are estimated at 2400 cases per year, for a benefit (lives saved) per risk
(potential cancers induced) ratio of 34,000.
Objective
The analysis reported here is based partly on the results of a study (Wr79)
conducted for the lonization Smoke Detector Bureau of the National Electrical
Manufacturers' Association to assess the potential radiation risks associated
with home-type ionization smoke detectors containing 21tlAm, and put these risks
in perspective relative to the benefits to be gained; similar evaluations have
been reported elsewhere (Nu78; Or77; Na77a).
This evaluation addresses only single station (home-type) ionization
smoke detectors; a reference case has been chosen for analysis in which
10 million detectors, having a mean effective life in the home of ten years, are
produced and distributed per year. At equilibrium there would be 100 million
detectors deployed, each with two microcuries of activity, less than that dis-
tributed in the past but representative of current output.
Approach
The approach taken was to assess the potential radiation exposure to
individuals and populations associated with the deployment of these detectors.
The collective and individual risks were then estimated by assuming a linear
dose-response relationship and multiplying the dose by an effects coefficient
established from human and animal information. For this purpose, total cancer
induction was taken as the measure of risk.
The benefit to risk ratio is expressed as the potential number of lives
saved from fires versus the estimated number of lives at risk as a result of
radiation exposure. In addition, the individual and collective doses associated
with the deployment of the detectors can be compared to natural background as a
means of putting the exposures in perspective.
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386
Finally, a comprehensive literature review was conducted and pertinent cal-
culations were adopted from other analyses, with suitable modifications to the
reference U.S. case chosen. In addition, detailed consideration was given to
specific situations not analyzed previously.
Detectors, 21tlAm Source Characteristics, and Integrity Tests
Detectors consist of a "tamper-proof" chamber housing the ionization source,
electronics to sense a change in voltage when smoke enters the sensitive volume,
and warning or signaling alarms,.
The active portion (21tlAm source) is manufactured by three different
manufacturers by similar but not identical processes. The following is a
brief description of the general process:
The majority of Ionization Chamber Smoke Detectors (ICSD) utilize the
oxide form of the alpha-emitting radionuclide 21tlAm (21tlAm02). Americium oxide
is uniformly mixed with gold, formed into a briquette and sintered at above
800°C. The briquette is then mounted between a backing of silver and a front
cover of gold or goId/palladium alloy and sealed by hot forging. The composite
material thus formed is cold rolled to give the desired activity loading which
ranges from 0.01 to 2.5 uCi/mm2. An additional corrosion resistant material,
either rhodium or gold, is commonly electrodeposited on the top surface. Total
foil activity ranged from 0.5 to 130 yCi in the past with modern designs utiliz-
ing 1 to 2 yCi 21* Am. The sources are commonly fixed onto metallic holders
usually stainless steel or plated brass by soldering or crimping them to the
holder wall.
Tests of detector and source integrity have been made by various organiza-
tions including the source and detector manufacturers, independent testing labor-
atories, and government-related laboratories. Tests include acid leaching,
abrasion, corrosive atmospheres, and high temperatures simulating fires (Ba76b;
Ea77; Gr76; Ha75,78; Hi76; Ni69; Sw69).
Removable fraction (i.e., by wiping plus immersion in water) ranged be-
tween 10~3 to 10~6. In most fire tests the airborne fraction is normally one
to two orders of magnitude less than the mechanically removable fraction; the
latter averaged 10~ in the reports surveyed. We have used 10~2 for our analysis
for conservatism and to take into account the possible effect of high temperature
incineration.
The number of detectors in use has increased greatly in the last few
years while at the same time the average amount of 21tlAm per detector has de-
creased. In 1970, for 53,000 smoke detectors (not just home-type) containing
21tlAm, the average activity per detector was 79 yCi with a total of about 4.7
Ci cumulative in all detectors. By 1975, the average activity of 21|1Am per
detector had dropped to 16 yd, but the total Ci used, 10.8, increased because
the number of detectors increased to 388,000. By 1977, detector sales had
increased to 7.3 million (this is based on a calendar year from July 1 of
the previous year through June 30 of the reference year) with an average
activity of 5.7 yCi, and a total amount of 21tlAm of 42 Ci that year (Nu78)
-------
387
Thus to mid-1977, nationally, there has been about 105 Ci distributed in
detectors. Industry figures show that in calendar year 1978, slightly more
than 11 million ionization detectors were sold, with an average activity of
about 4 yCi.
For 1979 the expected activity per detector is less than 2 yCi (Ha78).
Thus although there has been an enormous expansion in production, the total
activity distributed in detectors has increased less rapidly than the number
of units because advances in technology has allowed the use of less activity
per unit. Sales of 10 to 11 million units are expected in 1979.
Radiological Characteristics of 2ltlAm
Americium is an actinide element of atomic number 95 first identified
late in 1944. Americium-241, the isotope used in smoke detectors, has a
radiological half-life of 433 years and emits alpha particles and gamma and
X-ray photons with abundances as follows: E =5.48 MeV; Y! = 59.5 MeV, 36%;
Y2 = 0.026 MeV, 2.6%; and Neptunium L-X raysawith an average energy of 18 keV
emitted with an abundance of 37.6%.
Dosimetry and Risk Evaluation
Based on the results of animal experiments and some metabolic informa-
tion obtained in vivo in man, americium is retained primarily in three organs,
lung, after inhalation, from which it is transported primarily to liver and
skeleton (Du73; In59; Wr72).
The model and metabolic parameters of ICRP-19 (In72) were used in the
calculations of inhaled dose. In calculating the ingestion dose, the
ICRP-2 (1959) model was used for which a GI absorption of lO"1* and other
metabolic parameters taken from ICRP-19. Calculations of the inhalation dose
were made with the DACRIN Code (Ba75; Ba76a) similar to those first reported
by Strom and Watson (St75) but modified to use ICRP-19 rather than ICRP-2
parameters (Wa78).
Estimates of cancer risk were compiled (Bai76; Be72; Ene77; Ma72; Ma76a; Me7
Ne75) from several publications to derive a risk of l.ZxlO'Vyd ingested. Per
unit activity, inhalation carries four orders of magnitude greater risk than
ingestion. Risk and dose estimates are shown in Table 1. Americium-241 oxide
probably behaves intermediately between Class W and Y. The risk estimate for
inhalation of Class W is 2.7xlO~2. For external exposure, we used the BEIR
absolute linear risk coefficient, 2x10"Vrem (Be72).
The 59 keV gamma ray gives a dose rate of 9 nrads per hour at 1 meter
from a detector containing 1 microcurie.
Assessment of Exposure and Risk to House Occupants for Normal ICSD Use -
External Dose
A. Individual Dose
The most important component of dose turns out to be external dose from
the detectors.
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388
The potential for external exposures by smoke detectors has been eval-
uated by the NEA-OECD (Or77), which developed a methodology to make an assess-
ment of individual and collective doses from the detectors. This methodology
was applied by Johnson (Jo78) to the United States population and will also be
adopted in the present evaluation. In addition, Graham (GrTS^has calculated
and measured dose rates from different detectors containing 2Am. The excess
dose noted by Graham above 9 nrads per hour per microcurie of Am at 1 meter
results from inclusion of the X r-ys in the external dose estimates.
Although the detector housing normally provides some shielding from the
X rays, they are not completely eliminated. The X rays have not been included
in the dose estimates here however because they are not sufficiently penetrating
to be considered as "whole body" radiation, and their application in "total body"
risk estimates would be inappropriate.
The average annual dose, according to the formula developed by OECD and
for persons sleeping approximately 6 feet from a detector, 8 hours per day,
would be 14 microrads, or roughly 80 minutes of natural background. The dose
varies as the inverse square from the detector and at one meter would be four
times larger. Little information exists on distances people actually sleep
from detectors on the average, so this might overestimate or underestimate the
dose.
B. Collective Dose Estimate
The collective dose has been evaluated by assuming that 100 million
detectors containing 2 yCi each are distributed in 50 million homes, with 90%
of the units being positioned in hallways and the other 10% in bedrooms. Again
assuming an average distance of 2 meters, the collective dose is estimated as
340 man-rems. The assumption is made that 3 people are exposed for 1 hour per
day in the halls, and 2 people for 8 hours per day in the bedroom.
If all the detectors were installed in halls this estimate would be 232
man-rems and if all were installed in bedrooms, 1872 man-rems, assuming an
occupancy of 2.95 people/structure unit.
Summary of Calculation of Average Annual Collective (Dc) and Individual
(Pi) Dose
and
Dc = (DihVh + DibPbV I
where
r is the specific gamma ray constant for 21tlAm = 8.7 nrad
per hour per yCi at 1 meter;
A = Am activity of detector, 2 yCi;
n = number of Installed detectors (taken as 2 times the number
of dwelling units);
t = hours in year, 8760
-------
389
d = average distance of detectors from people, taken as
2 m;
F, - fraction of detectors in hallways = 0.9;
F, = fraction of detectors in bedrooms = 0.1;
P^ = three people in the hall;
P, = two people in the bedroom;
0^ = occupancy factor in bedroom taken as 8 hours per day (2 m
from detector);
0^ = occupancy factor in halls (2 m from detector) taken as 1
hour/day;
Oh = occupancy factor in halls (2 m from detector) taken as 1 hour/
day;
D.,_.= individual dose rate in hall; and
D., = individual dose rate in bedroom.
For these values, D. becomes 14.2 yrads per year, with 12.6 from the bedroom
(D, ) and 1.6 (D, ) from the hall. The collective dose D is 340 man-rads.
b h c
Exposure and Risk for Waste Disposal
The dose associated with disposal is evaluated by considering both
incineration and burial of detectors. Since home-type detectors have no
requirement for special disposal, it is prudent to assume that all of them
will be disposed of in normal refuse.
A. Individual and Population Dose from Incineration
The analysis adopted here is based on the NEA-OECD (Or77) analysis of
this problem. Assumptions used are:
5
Population feeding one disposal route = 1.5x10
Number of private homes (2 ICSD's/home) = 5xl05
Average activity in single station = 2 yCi ** Am
Total amount of refuse = 5xl08 kg/yr
Removal efficiency from stack =0.9 (90%)
For the reference case of 10 million units per year, this would involve 100
incinerators.
With well-designed sources less than 1% of the activity is likely to
escape or become airborne. If it is assumed that the stack is 50 m high, and
that removal of 90% of the particulates from the effluent takes place (a
likely circumstance since all modern incinerators will have particulate re-
moval devices, such as electrostatic precipitators or scrubbers, generally
with efficiencies exceeding 90%), then a maximum downwind concentration
averaged over a year is calculated (Br64), as ^ 10"11 yCi/m3. Assuming a?
breathing rate of 20 m3 per day, leads to an annual inhalation of 1.5xlO~7
pCi. This gives a dose commitment to lung and bone of 100 and 80 yrems
respectively, roughly 1/5000 normal annual background alpha doses to
these organs.
The estimation of collective dose to the U.S. population from incinera-
tion is given in Table 2.
-------
390
It is assumed that 10 million units are sold per year each having an
activity of 2 yCi. In order to assess the amount that might be inhaled from
eventual incineration of these devices, one needs to know the fraction of
the total units that might be incinerated, the fraction of the activity that
would be released and the fraction of the activity that would return from the
environment back to man. All of these can be estimated to give an order
of magnitude of the collective risk. From studies of fallout it is known that
cumulative human retention of plutonium inhaled from weapons testing debris
has been about 3xlO~8 (Ri75). This refers to debris that has descended
rather uniformly over the U.S. In the case of a large number of point sources,
such as incinerators, the distribution would not be uniform throughout the U.S.
but rather would be concentrated in areas which have a greater population
density where the incinerators are located. For purposes of this analysis
it is assumed that this will introduce a 30 fold factor greater collective
inhalation by people. It is also assumed that 10% of the units produced
eventually are incinerated, and that the fraction of material released from an
incinerated unit is one part in a hundred. This leads to an estimate of 2xlO~3
yCi collective inhalation with a concurrent risk estimate of 2xlO~ cancers per
year from the operation of the whole industry from incineration.
Clearly, modern incinerators will have particle filtration systems designed
to minimize the total mass loading to the environment which will reduce the
emission below the amount chosen for this analysis. From this calculation it
is clear that one is dealing with a very small risk, referenced to a U.S. pop-
ulation size of 200 million. It is possible that the estimate of the cumula-
tive activity retained may be too low or too high, but in any event is probably
not in error by two orders of magnitude which would still leave this risk
estimate a very small one.
One may ask whether or not there would be a significant addition to the
dose from resuspension of material which deposits on the ground. Several
studies suggest that the dose from resuspension of insoluble actinides would
at most be comparable to that received on direct inhalation (Wa74). This
might, therefore, increase the risk estimate by a factor of two.
B. Activity in Incinerator Slag or Ash
The concentration of 21tlAm in incinerator ash or slag may be evaluated
as follows: in the U.S. (Table 3) about 20% of solid waste is left as ash
after incineration. It is assumed that all of the 21tlAm will remain with
the ash. Estimates of solid waste generation per capita of 2.7 kg/day have
been made for 1980 (En77).
Assuming two detectors in each of 50 million homes, with a mean life of
ten years each, leads to the disposal of 10 million units per year. At
present, some 8 to 13% of solid waste is incinerated, the rest poing to
landfill.
Future incineration could increase or decrease. Accordingly, it is
assumed that 10% are incinerated.
For total incineration of all solid waste the average concentration in '
ash or slag is obtained by dividing the total 241Am discarded in detectors
by the residual ash weight from one year's worth of incinerated solid waste, or:
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391
(lOxlO6 detectors/y)(2xl06 pCi/detector)
(50xl06 houses)(2.7 kg/d, person)(2.95 persons/house) x
(365 d/y)(0.21 g ash/g solid waste)(103 g/kg)
0.66 pCi/g
For 10% of waste being incinerated, an average concentration would then
be 0.07 pCi/g of waste slag which is small compared to the normal alpha
actinide content in soil, about 5 pCi/g. However, there is no available data
on alpha emitters in solid waste incinerator residue with which to compare
this expectation.
The NEA-OECD study concluded that the majority of the activity would remain
with the slag, and an analysis of the various uses to which slag might be put
has shown these routes of exposure to be unimportant.
C. Disposal: Burial or Sanitary Landfill
The risk from inhalation to an individual from resuspension in the
vicinity of an uncovered landfill can be estimated using the approach developed
by Johnson (Jo78). It is assumed that the 21|1Am is uniformly distributed and
resuspends from soil as does uranium, a natural actinide. Johnson concludes that
an average air concentration of S.AxlO""1 pCi/m3 results from average uranium
soil content of 0.93 pCi/g.
The concentrations of residue in landfill can be evaluated from the cal-
culations of 21tlAm in slag from incineration. In landfill it is assumed that
the waste is mixed with an equivalent of soil, and that the fivefold reduction
in mass associated with incineration does not occur. Thus, this average con-
centration would be:
0.7 PCi/g
10
which would lead to a local air content due to resuspension of:
(§ifi) x 5.4x10-" = 4.0xlO-6 pCi/m3.
This substantially overestimates the expected concentration in air since
the resuspension value is valid only for a much larger area source than a
single or multiple landfill. Thus, for 70 years, inhalation of 20 m /day
would lend to an inhalation of 2 pCi, associated with a lifetime risk on the
order of magnitude of 10~7.
The collective risk has not been evaluated for this route but is clearly
limited and likely to be less than by other pathways.
D. Dose from Ingestion
A relatively simple calculation shows that the potential risk from inges-
tion is less than the risk evaluated for inhalation. Bennett (Be78) has
shown that 21flAm from weapons testing, which is widely distributed in the
-------
392
environment, is very poorly returned to man via diet. The cumulative
deposition on the U.S. is now about 5000 Ci, and the current annual average
dietary intake about 0.4 pCl. If no change in the fraction in diet occurs
over the next 70 years, then lifetime intake would be 0.4x70 - 27 pCi. In
200 million people, this represents an integrated lifetime intake of 5.4x10
yCi, or 10~6 of that distributed in the environment.
Assuming that 10% of the 21tlAm in smoke detectors leaches or otherwise
becomes available to move about the environment, the collective ingestion
risk for disposal of the reference 10 million units (integrated 70 years into
the future) would be:
)(107 detectors) (0.1)(10-6) 1.5xlO~5 (risk/yd) = 3xlO
'
~5
(
Metector
An alternative approach may be taken by calculating the 21tlAm content
on food grown directly on old landfill converted to agricultural uses. This
approach has been used by the Nuclear Regulatory Commission (Nu78) using a
concentration in waste of 0.05 pCi/g of waste. The concentration of 0.07
pCi/g derived here is comparable.
Sindfe 21+1Am is discriminated against by most animals, uptake by vegeta-
tion represents the most significant likely route.
If it is assumed that the waste is comparable to soil, that the Am
is evenly dispersed, that all the 21tlAm in detectors becomes available, that
plant uptake is 10"% and that 10 g daily of a person's diet comes from plants
grown in waste containing such detector origin contamination, then:
10-" x 10 -f- x 365
Q
x 70 y = 2xlO~6 yCi on a per capita basis,,
If one million people had such diets then the collective intake would be
equal to that derived from the assumption of uniform distribution.
Summary of Disposal Intakes and Risks
for 10 Million Units/Year
Collective
Incineration (inhalation)
Landfill (ingestion/food)
Activity (yCi)
2x10" 3
2
Risk
2.4x10""
3xlO~ 5
,The collective estimates of potential harm (risk) are summarized
in Table 4.
Risk/Benefit
In 1977, 9,950 civilian fire deaths were estimated for the U.S., of
which 8,600 occurred in structure fires, and 7,800 in residential structures..
Accordingly, "the United States fire death problem is heavily concentrated
in residential fires." It is likely, therefore, that the ICSD's will be use-
ful in reducing this toll of 7,800, which will be used for the analysis here.
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393
There undoubtedly will be a reduction of nonfatal injuries to civilians
(33,400 in 1977) and possibly also to firefighters (106,100 in 1977) if fires
are detected earlier and response accordingly can be more rapid.
Property loss from fires in structures was 5.2 billion dollars, and
if the proportion of property loss from residential fires to all structure
fires is the same as the relative loss of life, then the value of residential
fire property loss would have been about 4.7 billion dollars (De78).
Approximately 9,950 persons lose their lives in fire each year in the
U.S. (46.4. fire deaths/million persons). This is almost two times higher
than any other technically advanced country reporting, with Canada ranking
second with 29.2 fire deaths/million persons. Two-thirds or approximately
8,600 persons/year, lose their lives from fires occurring in buildings, 9
out of 10 of which occur in private homes. Since about 75% of these fires
occur during the night, people are usually asleep and are not aware of the
developing fire until it is too late to save themselves. From these data,
one could reasonable conclude that if the dwelling had been equipped with
some type of early warning fire detection device, many of these fatalities
could have been avoided (Br77, De78).
A number of studies have been conducted which allow a rough estimate of
the effectiveness of smoke detectors in reducing the fire death rate.
Halpin, et al. (Ha77) have assessed the potential impact of fire pro-
tection systems by studying the outcome and histories of 73 fatal fires in
the Maryland/Washington, D.C. area between June, 1976 and 1977- The method
chosen was to make a thorough investigation of all fires in which fatalities
occurred and then to examine the data to make a professional judgment as to
the likely efficacy of various fire protection systems. The fires were
primarily residential in classification.
There were 114 fatalities in these 73 fires, and about 2.6 million dollars
in property losses. The conclusions were that 100 of the 114 fatalities could
have been saved and 114 of the 123 injuries prevented, if fire protection
systems had been installed. Property losses would have been about 73% lower
if detector systems had been operating.
A 90% potential effectiveness of smoke detectors could be inferred as a
result of this analysis. However, this study showed, as did others, that the
human factor reduced the effectiveness below that theoretically attainable.
People do not always realize that their lives are in danger when warned of a
fire, and do not always behave accordingly, i.e., in a manner to insure their
own survival.
The mortality rate was greatest among the young (less than age 9) and
elderly (greater than 60), being 20 and 24% respectively of the total. This
is consistent with the distribution of deaths in national fire statistics (Na77b),
Thus, life shortening is particularly high among the young involved in fires.
In Canada, McGuire and Ruscoe (1962) concluded that ICSD's may save up to 45%
of average adults, and 35% of children and the infirm who die in fires.
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394
Although neither this nor other studies provide a highly accurate
basis for judging the effectiveness of smoke detectors, the potential
effectiveness is nevertheless up to 90% and the practical effectiveness,
although lower, is still substantial. Thus, the use of smoke detectors
should be coupled with an educational program for proper response in case
of an alarm.
For the purpose of this report an effectiveness of 45% will be assumed,
consistent with the data reported by Halpin, et^ al. (Ha77) and the figure
used in the NEA-OECD study (Or77).
Ten million detectors installed in 5 million homes, with about 2.95
persons/household, would serve a population of 14.75 million. For a home
fire rate of 7,800/year, and an effectiveness of 0.45, and a U.S. population
of 215 million, we get lives saved as
7,800 x 15 x 0.45 = 240 lives/year
If these are introduced over a ten-year period, then the number of lives
saved per year will be cumulative, or about 2,400 per year.
We may now compare this with our estimates of 7xlO~2 potential indirect
effects per year, or a benefit to risk ratio of 34,000 lives saved for every
potential life shortened. The former are identifiable early deaths averted and
the latter are non-observable effects which result from a calculation assuming
a linear relationship between radiation dose and biological effects at levels
well below natural background.
The major contribution to risk comes not from internal exposure but from
external exposure. The risk could also be expressed as a fraction of natural
background. For individuals the dose to a maximally-exposed individual was
shown to be less than 1% of the natural background exposure. On a collective
basis the natural background delivers about 20 million man-rems to the U.S.
population per year, whereas doses from 100 million installed ICSD's would
collectively be about 400 man-rems, or one part in 250,000.
Thus, the dose from ICSD's is trivial relative to natural background,
the variation in natural background, and from most other human activities
involving alteration of natural radiation exposure.
The collective disposal and external risks are shown in Table 4. The
risks from external exposure (7xlO~2) appear to be two orders of magnitude
greater than the risks from internal exposure (2xlO~1*); waste disposal
practices do not contribute significantly to the risk. The benefits, by
any reasonable measure, appear to be four to five orders of magnitude greater
than the risks.
Translated into human terms the potential for saving lives far exceeds
the risk associated with the widespread deployment of smoke detectors con-
taining 21>1Am.
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395
Acknowledgment s
We wish to thank those who assisted with this report: E.G. Watson for
the computer dose calculations and Paul Linsalata for summaries of parts of
the literature.
This research was partially supported by the Department of Energy,
Contract No. EY-76-S-02-3382 and is part of center programs supported by
Grant No. ES 00260, from the National Institute of Environmental Health
Sciences, and Grant No. CA 13343, from the National Cancer Institute.
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Calculating Organ Dose from Acute or Chronic Radionuclide Inhalation:
Modification for Gastrointestinal Tract Dose," BNWL-B-389, Supplement.
Bai76 Bair, W.J. and Thomas, J.M. , 1976, "Prediction of the Health Effects of
Inhaled Transuranium Elements from Experimental Animal Data," In: Trans-
uranium Nuclides in the Environment, IAEA, Vienna, pp. 569-585.
Ba76a Battelle Northwest Laboratories, 1976, "DACRIN: A Computer Program for
Calculating Organ Dose from Acute or Chronic Radionuclide Inhalation," BNWL-
B-389, Revised from 1974 Report.
Ba76b Battelle Memorial Institute, 1976, "Final Report on Radioactivity Loss
at Elevated Temperatures from lonization Fire Detectors," Columbus, Ohio.
Be72 BEIR Report, Report of the Advisory Committee on the Biological Effects
of Ionizing Radiations, Division of Medical Sciences, 1972, "The Effects on
Populations of Exposure to Low Levels of Ionizing Radiation," National
Academy of Sciences, National Research Council, Washington, D.C.
Be78 Bennett, B.C., 1978, "Environmental Aspects of Americium," Doctoral
Dissertation, New York University Medical Center.
Br64 Bryant, P.M., 1964, "Methods of Estimation of the Dispersion of Wind-
borne Material and Data to Assist in Their Applications," AHSB(RP)R42, HMSO.
Br77 Bright, R.G., 1977, "Status and Problems of Fire Detection for Life
Study in the United States," Proceedings of a Symposium, 1975, Council Fire
Detection for Life Safety, National Academy of Sciences/Nuclear Regulatory
Commission, Washington, D.C.
Br78 Bright, R.G., 1978, "Technical Developments of Domestic Fire Detectors,
Presented at the International Fire, Security and Safety Exhibition and
Conference, April 24-28, 1978, London, National Bureau of Standards.
De78 Derry, L., 1978, "A Study of United States Fire Experience," Fire
Journal, pp. 67-77.
Du73 Durbin, P.W. , 1973, "Metabolism and Biological Effects of the Trans-
Plutonium Elements," In: Handbook of Experimental Pharmacology, Ch. 18,
Vol. XXXVI (H.C. Hodge, J.N. Stannard, and J.B. Hursh, Eds.), Springer-
Verlag.
-------
396
Ea77 EAD Metallurgical Test Results for Foil Integrity, 1977, Attachments
to Letter of November 10, 1977, Radosavljevic, to U.S. Nuclear Regulatory
Commission, and Test Results tollected with Nuclear Alpha Foil AMX-110,
Tonowanda, New York.
Ene77 Energy Research and Development Administration Report, "Health Effects
from Transuranic Element Exposures," ERDA-1545D, Draft Environmental Impact
Statement, Rocky Flats Plant Site, Vol. II, Golden, Colorado.
En77 Environmental Protection Agency, 1977, "Municipal-Scale Thermal
Processing of Solid Wastes," PB-263, p. 396.
Gr76 Greenberg, G. and Dooley, D.A., "Am-241 Foil Integrity Tests," Per-
formed for Nuclear Radiation Developments Corporation."
Gr78 Graham, C.L., 1978, "Radiation Dose Rates - Various Smoke Detectors,"
Fire Journal, p. 109.
Ha75 Hall, E.G. and Hunt, D.G., 1975, "A Summary of Testing Programme on
Alpha Foils Used in lonization Chamber Smoke Detectors," TRC Report, 378,
The Radiochemical Centre, Ltd., Amersham.
Ha77 Halpin, B.M., Dinan, J.J. , and Peters, O.J., 1977, "Assessment of the
Potential Impact of Fire Protection Systems on Actual Fire Incidents,"
Applied Physics Laboratory, Johns Hopkins University.
Ha78 Hall, E.G. and Hunt, D.E., 1978, In: Radioactivity in Consumer
Products. NUREG/CP-0001.
Hi76 Hill, M.D., Wrixon, A.D. and Wilkins, B.T., 1976, "Radiological
Protection Tests for Products Which Can Lead to Exposure of the Public to
Ionising Radiation," National Radiological Protection Board, R42, Harwell,
Didcot, Oxon. 0X11 ORA.
In59 International Commission on Radiological Protection, 1959, ICRP-2,
Report of Committee II, "Permissible Dose for Internal Radiation," Pergamon
Press.
In72 International Commission on Radiological Protection, 1972, ICRP-19,
"The Metabolism of Compounds of Plutonium and Other Actinides," Pergamon
Press.
Jo77 Johnson, J.E., 1978, Memo to lonization Smoke Detector Manufacturers,
February 28, 1978.
Jo78 Johnson, J.E., 1978, "Smoke Detectors Containing Radioactive Materials,"
In: Radioactivity in Consumer Products, NUREG/CO-0001.
Ma72 Mays, C.W. and Lloyd, R.D., 1972, "Bone Sarcoma Incidence Versus Alpha
Particle Dose," In: Radiobiology of Plutonium (B.J. Stover and W.S.S. Jee,
Eds.), J.W. Press, Salt Lake City, pp. 409-439.
Ma76a Mays, C.W. , Spiess, H., Taylor, G.N., Lloyd, R.D., Jee, W.S.S.,
McFarland, S.S., Taysam, D.H., Brammer, T.W., Brammer, D., and Hollard, T.A.,
1976, "Estimated Risk to Human Bone from 239Pu," In: Health Effects of
Plutonium and Radium (W.S.S. Jee, Ed.), Salt Lake City, pp. 343-362.
Ma76b Mays, C.W., 1976, "Estimated Risk from 239PU to Human Bone, Liver and
Lung," In: Biological and Environmental Effects of Low Level Radiation.
Vol. II, International Atomic Energy Agency, Vienna, pp. 373-384.
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397
Mc62 McGuire, J.H. and Ruscoe, B.E., 1962, "The Value of a Fire Detector in
the Home, Fire Study No. 9," Ottawa, Canada, National Research Council,
Division of Building Research.
Me75 Medical Research Council, 1975, "The Toxicity of Plutonium," Her
Majesty's Stationery Office, London.
Na77a National Council on Radiological Protection, 1977, "Radiation Exposure
from Consumer Products and Miscellaneous Sources," NCRP-56.
Na77b National Safety Council, 1977, "Accident Facts," Chicago, Illinois, U.S,
Ne75 Newcombe, H.F., 1975, "Mutation and the Amount of Human 111 Health,"
In: Radiation Research; Biomedical, Chemical and Physical Perspectives
(O.F. Nygaard, e_t al. , Eds.), Academic Press, New York, pp. 937-946.
Ni69 Niemeyer, R.G. , 1969, "Containment Integrity of 226Ra and 21|1Am Foils
Employed in Smoke Detectors," Oak Ridge National Laboratory.
Nu78 Nuclear Regulatory Commission, 1978, "An Interim Staff Analysis of
the Environmental Effects of lonization-Type Smoke Detectors."
Or77 Organization for Economic Cooperation and Development, Nuclear Energy
Agency, 1977, "Recommendations for lonization Chamber Smoke Detectors in
Implementation of Radiation Protection Standards."
Ri75 Richmond, C.R., 1975, "Current Status of Information Obrained from
Plutonium Contaminated People," Proceedings of the Fifth International
Congress of Radiation Research (O.F. Nygaard, H.I. Adler and W.K. Sinclair,
Eds.), Seattle, Washington, pp. 1248-1266.
St75 Strom, P.O. and Watson, E.G., 1975, "Calculated Doses from Inhaled
Radionuclides and Potential Risk Equivalence to Whole Body Radiation," IAEA
Symposium on Transuranium Nuclides in the Environment.
Sw69 Swiss Reactor Institute, Department of Radiation Control, 1969,
"Assessment of the Behavior of the Cerebras FES 6 Fire Detector and its
Radiation Sources During a Fire," Zurich.
Wa74 WASH-1535, 1974, "Pu Toxicity," App. IIG, Vol. 4.
Wa78 Watson, J.C., 1978, personal communication.
Wr72 Wrenn, M.E., Rosen, J.C. and Cohen, N., 1972, "In Vivo Measurement of
Am-241 in Man," In: IAEA Symposium on Assessment of Radioactive Contami-
nation in Man. Vienna.
Wr79 Wrenn, M.E. and Cohen, N., 1979, "Assessment of Risks and Benefits of
Home lonization-Type Smoke Detectors," Draft report prepared for the
lonization Smoke Detector Bureau of the National Electrical Manufacturers'
Association.
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398
TABLE 1
50-Year Committed Dose Equivalent and Risk Values
Adopted for 2l>1Am. 0.5 Micron Particles
rem/yCi
Linear
Estimate of
Risk/yCi Inhaled
or Ingested
Inhalation
lung*
liver
skeleton
Ingestion (fi
liver
skeleton
Class Y
680
500
740
= 10-")
0.9
1.4
Class W
69
1000
1600
Cancer Risk/rem
1.6x10-"
io-5
4xlO~6
TOTAL
io-5
4xlO"6
TOTAL
Class Y Class W
l.lxlO"1 l.lxlO"2
5x10- 3 ID'2
3xlO~3 0.6xlO~2
l.ZxlO-1 2.7x10-^
9xlO"6
5.6xlQ-6
1.5xlO~a
* The dose to lung is to the pulmonary region where lung cancers are most
often found in inhalation studies with actinides in experimental animals
(Bai76).
TABLE 2
Calculation of Collective Dose and Risk
from Incineration of Smoke Detectors
Basis:
One year's production (P) of 10 million units
Activity per unit (A), 2 yCi
Fraction of units incinerated (0.1) Fj
Fraction of activity released from the detector
during incineration (FR) = 10"2
Fraction of activity released from detector which
is released from the incinerator stack, Fs - 0.1
Cumulative fraction of that released and is retained
by man, by inhalation, Cm = 10"6
Ratio of activity inhaled to that retained, Fm = 10
Risk per unit activity inhaled, R = 0.1 yCi (cancer
induction)
Then, the total estimate of risk from one year's operation
until complete disposal would be:
PAFIFRFsCmFmR
which is 2x10-" cancers, associated with the collective
inhalation of 2xlO~3 yCi.
-------
TABLE 3
Composition of Solid Waste
Average for 21 U.S. Cities, 1966-1969
Weight
Category (%)
Food 18.2
Garden 7.9
Paper 43.8
Metals 9.1
Glass and Ceramics 9.0
Plastics, rubber, leather 3.0
Textiles 2.7
Wood 2.5
Dirt, ash, etc. 3.7
100.0
Composition Typical Composition
Moisture
Carbon
Oxygen
Hydrogen
Sulfur
Nitrogen
Ash
28.2
25.6
21.2
3.5
0.1
0.6
20.8
TABLE 4
Collective Disposal Risks
Lifetime Committed
Cases/year
Incineration 2xlO_*
Burial 2xlO_5
Firefighters 10 5
External person-rem
home 340
transport 2
warehousing 6
retailing 17
Total 365 7x10
-------
BANQUET ADDRESS
MANAGING LOW-LEVEL RADIOACTIVE WASTES:
BIOETHICAL CONCERNS
Margaret N. Maxey, Ph.D.
University of Detroit
Detroit, Michigan 48221
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401
INTRODUCTION
Mow that the high-level waste problem has attracted considerable public
exposure and exploitation, the mounting controversy over low-level waste
management has begun to compete for center stage. The strategy planned at
Critical Mass in 1974 continues to be successful. At that forum, Dr. Margaret Mead
exhorted her audience "to make people feel that everything they value in the
world is at stake" if the development of nuclear energy is allowed to continue.
She stated then that "Americans aren't afraid of dying suddenly but of dying
slowly." She therefore urged her listeners to concentrate on those aspects of
nuclear power that would evoke the most fear, namely, the long-range deteriorating
effects of low-level radiation. The association in the public mind between
low-level wastes and low-level radiation continues to bear fruit in the strategy
of inducing fear. However, there is reason to be convinced that in the long term,
reasonable arguments will prevail. To that end, the public must be confronted with
the social necessity of an equitable mangement of hazards having the potential for
harmful health effects and social consequences.
Sy "equitable management" I mean that policy makers should first be well informed
about the broad spectrum of both natural and ordinary hazards that may have health
effects for large segments of a population, then make comparisons of actual costs
per capita to reduce them, and only then set criteria and standards that will get
the most public health protection for the many out of a finite amount of money.
Hazard management is equitable only if it is proportional in relation to the actual
harm that can be identified and reduced by expenditures of human effort, time, and
money.
In view of ethical requirements for social justice and equity, we need to address
three major problems: the adequacy of conceptual tools for assessing biohazards;
the disagreement among scientific experts; the origins of value-conflicts underlying
expert disagreement.
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402
I. CONCEPTUAL TOOLS: How we structure hazard-management.
Contrary to a popular misconception, "hazards" have neither a bare facticity
nor an intrinsic morality predetermining how human beings should behave in relation
to them. Hazards are not baldly "there" in nature or in human transactions with it.
What people regard as hazardous in any given era reflects what they have come to
know about their environment, and what they value as essential or desirable on a scale
of real possibilities. In short, human beings structure hazards. In that sense,
hazards are human artifacts. A hazard is not by definition "toxicity of substance"
or "violence of event" or "magnitude of consequences" that can be known, classified,
and predicted. A hazard exists only when, and to the degree that, harmful exposure of
and assimilation by the human body or other valued living systems becomes a genuine,
not merely imaginable, possibility. And that possibility exists only when there is
an inability or failure to devise and maintain controlling actions or safeguards.
Because there are vast uncertainties about "how the world works," it serves
no human purpose to bewail our "legacy of risks to future generations," and then make
the fraudulent claim that the goal of hazard management must be to assure centuries
of control over toxic elements or prediction of future adverse events. I concur
with Prof. William Clark in his statement that hazard management is "the adaptive
design of hazard structure," and that the primary goal of hazard management is "to
increase our ability to tolerate error and to take productive risks."^ This
statement stands in sharp contrast to a popular yet unexamined notion -- expressed
as well as anyone by Wolf Hflfele — that "we are locked in a world of untested
hypotheses (of unimplemented trials) because we dare not let experience prove us
wrong. The costs of failure have grown too great."3 Not only does this notion
reflect the New Pessimism, the defeatism and pseudoscientific dire predictions,
pervading our cultural climate; but it also constitutes in itself the ultimate hazard •
the failure to design and maintain structures of social resiliency. It is the social
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403
ideal of resiliency that has been a major driving force behind the emergence of
highly complex and technologically advanced societies. The social ideal of resiliency
impels us to cope with the risks that are inherent and unavoidable in the human
condition by doing risk-analysis.
The hazy connection between hazards and risks has given rise to another
popular misconception. If popular literature on the subject is any indication, "risk"
is steadily acquiring the moral opprobrium reserved for other four-letter words. I
do not intend to add to that moralizing. Suffice it to say that "risk" has begun to
carry all the baggage associated with uncertain consequences of so-called "hard"
technology in a world of big, bad, centralized, corporate industrialism. What accounts
for this state of affairs? At the very least, many have adopted the uncritical
assumption that risk is a normative concept for certifying consequences to human
beings that are harmful, dangerous, or "bad." These contrast sharply with consequences
that are beneficial, pleasurable, or good, and by implication, risk-free. We have
already grown accustomed to graphs which imply this dichotomy: one axis measures
risks; the other axis measures benefits. This assumption is altogether understandable
because it reflects a basic value-conflict about the nature of risk-taking.
For some persons, risk-taking is by definition hazardous, harmful, and perhaps the
result of some demonic compulsion suppressing nobler human pursuits. For others,
the word risk stands for the opportunity to undertake what is challenging and
venturesome, innovative and fulfilling to the human spirit in its endeavor to live
"the good life." This value-conflict has developed because risk-taking is not
inherently good or bad -- neither in a psychological sense nor in a moral sense.
The fact that the concept of risk is negatively overloaded in practical usage has
no theoretical justification. In any case risk-taking is inherent in the capacity
of a social -- hence moral -- being to make conscientious, consequential decisions.
Because of a facile identification of risks with hazards, a false antithesis
has been set up between risks and benefits - as if there were a way to have one without
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404
the other. Granted that it is a common phrase already entrenched in our regulatory
lexicon, "risk-benefit analysis" is misleading. "Benefit" is a term that stands for
a known or virtually certain result or reward. Risk implies an unknown or uncertain
outcome, a mere possibility weighted on the side of probability that intended benefit
will not materialize and, instead, harm may occur. As William May suggests, it would
be far more accurate to talk about either harm-benefit analysis (so that both words
would refer to expected outcomes ) or risk-hope analysis (so that both words would
clearly signal possibilities only. The trouble with the phrase risk-benefit is
twofold: it fails to express a proper symmetry, and it tends to obscure the primacy
of benefit within the normative structure of human action.6 That is to say, the
primary motivating force of human activity is the foreseen and intended benefit
which can be gained or lost. Even in the case of activity undertaken to avoid
harm or injury, it is the intended benefit of that avoidance that is primary. In
other words, in concrete decisions, what is actually "at risk" is the possibility that
intended benefit may be gained or lost. If risk-analysis is going to reflect the
normative structure of human action, then the concept of risk must focus primarily
on the benefit acquired or foregone. When harm results, it is clearly unwanted and
unintended. Risks and benefits are inseparable, not antithetical.
I will be the first to admit that these remarks will appear to be hopelessly
ivory-towered and out-of-touch, in view of entrenched interests in doing risk-
benefit analysis as usual. But I refuse to be deterred, because they have important
implications for the common good of society.
In my view, a major problem about the growing dispute over low-level radiation
hazards and low-level waste management is the inadequacy, not of risk-analysis,
but of harm-benefit analysis. The first order of business here should be to gain some
refinement in the concept of benefit. In one of the most comprehensive and insightful
•j •
studies of risk to date, Okrent and Whipple suggest that we should make qualitative
distinctions which reflect significantly different types of benefit, namely those goods
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(1) essential to society (e.g. food, water, energy at sufficient levels,
in short basic goods);
(2) beneficial or advantageous to society (e.g. most manufacturing);and
(3) of peripheral, if any value to society (e.g. aerosal deodorants which have
readily available substitutes at similar cost and lower
likelihood of harm.)
Using these refinements, we can distinguish between corresponding levels of harm.
Just as there are basic goods, there are also basic harms that may result from being
deprived of goods essential to subsistence and material well-being. The obligation
of a society to avoid basic harms and provide access to basic goods has been formulated
in the ethical principled justice and equity. As for second-level benefits which
are advantageous because they improve the quality of life of a society, the total
outcomes of any social policy toward such improvements will have an unclear mix
of benefits and harms. In our era, one of the most difficult questions we face as a
society is flow to make comparative judgments about the moral desirabilities of various
harms and second-level benefits, especially when they are different in kind.
Automobile and airplane manufacturing afford major economic benefits to
employees, to capital investors, and to the general health of international economies.
Yet each time someone drives a car or enables an airplane to take off, the benefits
one pursues may entail the possibility (risk) of unintentionally causing the death or
serious impairment of a fellow human being. Any society must, at some point in policy
formation, deliberately decide how we ought to balance economic benefits and costs ;
against possible harm or loss of life.
According to critics of such balancing, a human life is of infinite value,and
its loss or impairment cannot be put in a class with other "negative consequences,"
much less be given a finite monetary value. To do so indicates the moral bankruptcy
of our materialistic, consumerized, decadent society. Cost/risk/benefit quantifications,
say its critics, manifest a loss of respect for the sacredness of human life.
Those who defend this conceptual tool have sometimes used simple observations,
such at "Tkere are necessary tradeoffs in any public policy decision," or "Everyone
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puts a finite, monetary value on one's life when buying life insurance, installing
safety mechanisms in a home or auto, taking hazardous jobs because they pay higher
wages." Such analogies are true enough, but not sufficient. The public must be
educated and confronted with the fact that any society or viable economy has but a
finite amount of money to spend on health protection and safety, and that the
ethical problem is to get the most protection for the most people from this finite
amount. To put a government, or an industry, or a company in a position of financial
insolvency on the grounds ttiat their financial liability for loss of life is Infinite
violates equity and justice. These principles express the obligation of a'society to provi
access to basic goods and a reasonable quality of life for its citizens as a whole.
As a conceptual tool which attempts to enhance informed consent, cost/risk/
benefit quantifications are merely one tool among many others whereby policy makers
endeavor to allocate finite amounts of money in a just and equitable manner. They
are not tools for putting some callous "dollar value" on human life or injury as a
moral judgment of individual worth, much less of using economic losses to society as
a measurement of personal expendability. We are in fact maximizing the value we as
a society place on human life when we endeavor to allocate limited amounts of money
in such a way as to reduce widespread hazards, thereby preventing as much loss of life
and protection from injury as possible.
The fact that our tools for balancing economic costs against risks to human
life are not morally or ethically objectionable does not amount to saying that they
are psychologically easy and acceptable to the general public. Far from it. The
task of public education in this matter is monumental. Furthermore, I concur with
my colleague in social ethics, George Pickering, in his observation that "we are
going to have to do more than find some level of 'acceptable risk'; we are going
to have to come to terms with the question of 'justifiable harm.1 There are, after
all, some kinds of harm which cannot be avoided; but there are other kinds of harm
which any society should not allow and against which it should adopt protective or
remedial measures to the best of its ability."8 Which is which becomes the problem.
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We must face up to the discomforting task of formulating a more enlightened
concept of, and method of informed consent to, unavoidable hence justifiable harm,
and not divert attention away from it by focusing exclusively on "acceptable risk"
criteria. In my judgment, our failure to undertake this task lies at the root of
the second problem I noted at the outset: namely, the frustrating dilemma of a
policy-maker who wishes to set safety standards on the basis of informed consent --
yet when he turns to scientists upon whom he relies for "expert testimony," he
finds that they have basic disagreements about what data should count, how it should
be interpreted, and what level of health protection is "acceptable" or "safe enough
/
to be safe."
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II. EXPERTS AND STANDARDS FOR "SAFETY": How we have institutionalized dissensus
and value conflicts.
Aaron Wildavsky has recently observed, "Experts are used to disagreeing,
Q
but they are not so used to failing to understand why they disagree."'
From my reading, research, and reflection on the problem of "expert dissensus"
(vs. consensus) especially in the matter of radiation protection, I have come to the
conclusion that at the heart of the matter lies a misconception about safety,
especially as it relates to risk estimates and risk acceptability.
A case in point is the unending controversy over whether or not there is a
threshold for radiation below which no harmful effect occurs. A threshold concept
has been generally accepted for most toxic elements. It carries the implication that
below a threshold dose any exposure is "absolutely safe." But over the twenty years
of evolution in radiation protection philosophy, the ICRP and NCRP came to adopt
a conservative assumption, namely that it would be more prudent to assume some
harmful effect from any radiation dose, however small, than to assume a threshold dose
and then discover data proving it to be false. This conservative assumption carries
the implication that there is no absolutely safe radiation dose except zero, and
every dose greater than zero entails a corresponding risk of genetic or somatic
harm. In the ensuing process of applying a linear no-threshold hypothesis to the
development of standards, regulatory institutions and some of theirexpert advisers seen
to have forgotten that their quest for radiation limits rests only on a hypothesis,
a conservative assumption, and not on a scientifically established fact.
As Dr. G. Hoyt Whipple has observed, "The data on the biological effects of radiation
can be interpreted in terms of a threshold dose, but even the vast amount of
radiobiological data cannot conclusively prove the existence, or absence, of a .
threshold."10
Given this state of affairs, the dilemma of the policy maker could be
mitigated if two factors in the controversy wer^-clarified: (1) the Caning of safe;
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and (2) the meaning of threshold.
As noted above, a profound misconception of "safety" dominates the controversy
over radiation protection. The working assumption of policy makers and regulators
has been that safety is an intrinsic, measurable, absolute property that any given
system, or product or activity can and should possess. Our society has institutionalized
and appointed the regulator to measure approximations to that elusive property. The
mandate of the regulator is to make evermore stringent regulations, presumably to
come ever closer to that property by reducing risks. However, the only risks he is
expected to monitor and minimize are a small percentage of the total spectrum of
risks tolerated by members of society as a whole. Intent on making a set of risks
publicly "acceptable," as an index of "safety," the professional regulator must
continue to propose risk-reduction without regard to economic costs or social impacts
of ever-changing regulations. Seemingly, he is "only giving the public what it
wants," namely safety. This spiral is likely to continue unless or until the public
comprehends the fact that safety is not an intrinsic property measured by approaching
zero-risk. Safety is an evolving, relational value-judgment derived from current personal
or social priorities. Risks can be scientifically measured, quantified, and predicted
in probabilistic terms. Safety, however, cannot be measured, much less pre-determined
by the presence or absence of risks. Judgments of safety are judgments about the
justifiability or unjustifiability of harm. The process of reasoning whereby ethical
safety-policy decisions are made ought to be dictated—not by risk avoidance, an
impossible ideal — but by comprehensive risk/risk assessments and cost/risk/benefit
ratios. When these comparisons make it clear that a point of diminishing returns on
allocations of money, time, and effort has been reached by comparison with other
Potential hazards in a society, then the particular product or process under scrutiny
is "safe enough." If indeed unintended and unwanted harm should occur despite carefully
wrought safety-policy decisions, then such harm can be judged "justifiable" because
unavoidable or negligible by comparison with other harms and essential benefits.
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If policy makers were more circumspect about the process of reasoning from
which they ought to derive ethical safety-policy decisions, there might also be
less ambiguity about a disputed "threshold concept."
With their increasingly sophisticated measurements in radiobiology, specialists
are capable of identifying, and extrapolating from, even minute effects of exposure
to radiation. But it is a qualitatively distinct cognitive leap to make the value
judgment that a zero-threshold for so-called "safe" radiation exposure ought to be
written into regulatory standards. Certain radiobiologists make this value judgment.
However, scientific judgments about putative effects from radiation exposure cannot
and ought not to be substituted for an ethically responsible value judgment about
"safety." For the policy maker, the practical threshold concept cannot be evaded.
There can and must be a practical threshold below which the possibility of unintended,
unwanted, and comparatively insignificant harm becomes ethically justifiable.
This justification derives from a reasoning process which concludes that such effects
are unavoidable and negligible by comparison with other greater radiation exposures --
both naturally occurring and applied by humans — and with other potential hazards
against which citizens ought to be protected first.
The near clinical paranoia about cancer which some citizens experience cannot
be avoided. Indeed it may even be exacerbated by policy makers and politicians who
are trafficking in the fearsome mystique now surrounding radiation sources. The
pathologic fear of radioactivity and radiation will in time be overcome, just as
mankind has transformed its fear of fire, steam locomotives, electricity, the
automobile and the airplane. Meanwhife, however, the fear of cancer is only a symptom
of much more pervasive fears about the fate of our human species.
The dfssensus among scientists may mean that we need to devise innovative
institutional methods for dealing fairly with their complaints without undermining
still further public confidence in expert professionals, in safety-policy decisions,
*
and in regulatory actions. But to do so with circumspection, we must recognize the
origin of basic disagreements over value judgments.
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III. UNDERLYING PHILOSOPHIES: Why we derive conflicting values.
The mounting controversy over low-level radiation and low-level wastes
has revealed basic value-conflicts. We are compelled, therefore, to probe more
deeply into the origins of these value-conflicts — namely, the philosophical and
ethical principles from which values derive their justification.
In this regard, it is fortunate that the Natural Resources Defense Council
(a self-styled public intenst group in the USA) has recently disseminated a "Report"
supplementing its highly critical "Comments on Criteria for Radioactive Waste"
proposed by the Environmental Protection Agency. I consider it a fortunate
development — not because these political tracts are likely to advance either the
public interest or public understanding of the complex issues they purport to
address -- but because it throws an illuminating spotlight on the NRDC's ethical
and philosophical assumptions.
In their critical Comments, the NRDC authors repeatedly chastize the EPA and
proposed criteria for evading what they choose to call "the fundamental mandate of EPA"
and "an uncompromisable standard" -- namely, "non-degradation of the environment."
This is their rendering for "protection of the environment." The NRDC authors commend
the EPA at one point for comparing hazards from human activity to hazards from the
"pre-existing natural state of the area." (7) As their reason for feeling that this
is an appropriate standard, the NRDC authors state that "it emphasizes the role of
a trustee as one who maintains the non-renewable environment as it was originally,
to pass on to the next trustee." This fundamental goal is a key consideration)
$ays the NRDC, "because if any degradation is allowed (in the name of 'allowable
radiation exposure1), there is no clear bound at which degradation becomes, by
anyone's standard, too much." (3-4)
In their supplementary Report, the NRDC authors make the claim that the
"public-'s attitude toward the environment is one of non-degradation."(26)
Test Ban Treaty, the demise of the Plowshare Program, and
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public concern over long-lived nuclides and tailing piles, the NRDC claims that it
speaks for the public on two contentions:
(1) "No amount of radioactive contamination is 'acceptable1. . . .
(2) "The effects of radiation on future generations are of prime import
and can not be discounted."(26)
On behalf of the public, the NRDC authors are of the opinion that — given their
assumption that longevity of hazardous lifetimes of radioactive wastes constitutes an
unfair imposition of hazards and risks upon unconsulted future generations -- the
ethical principles of equity and participation require waste disposal criteria to be
neutral to future generations. Although admitting that the ideal of a totally neutral
allocation of benefits and risks is unattainable, it nonetheless serves a purpose
useful to the NRDC polHkal strategy: it
"finds practical application in refuting the arguments that a present
commitment to nuclear power is fair becaase investments in a technological
society now via nuclear power will benefit the future as a result of an
enhanced society, more than they hurt as a result of waste hazards."(11)
The NRDC authors preface their own proposed criteria by stating that "the
least unfair way of managing intertemporal relationships is for each generation to
try-to leave the earth as it was when they arrived. As a goal, the only acceptable
distribution of hazards and benefits is the neutral allocation, where no pattern of
benefits and hazards is imposed."(28)
From their version of a theory of Justice and Equity, the NRDC authors
derive a criterion which purports to consider only the risks to future generations,
and to ignore the net benefits of using nuclear energy. The original unmined ore
bodies and their cumulative risks to future generations are to be established as the
eeasure against which cumulative risks from nuclear operations of all types (mining,
milling, fuel processing, decommissioning, waste disposal) are to be compared for
acceptability. Considerations of cost are secondary. If it should happen that our
society does not
»
"wish to bear the monetary costs of justice, then we should explicitly
acknowledge that we prefer being wealthy and evil to being poor and
righteous and not try to justify our moral vacillation with a cloud of •
cost/benefit models."(29; emphasis addsd)
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These NRDC tracts give us much to ponder about beams and motes in the eyes
of special interest politics, and in that regard, they are a helpful exercise in
edification. But as for providing the public with thoughtful and persuasive analysis
of the intellectual questions posed by human transactions with natural and man-made
radiation sources, they are an exercise in obscurantism.
By espousing as a fundamental philosophical principle, "non-degradation of the
environment," the NRDC joins with the Sierra Club in defining a "degraded environment"
as any place that human actions have affected or changed. Formulas such as these
obscure two questionable assumptions:
(1) that an untouched "natural environment" by definition manifests a
superior, if not sacred order which human interventions violate to
some degree; and secondly,
(2) that a trustee of a so-called "natural environment" can do nothing more
nor less than pass it along in its original pristine state; to do
otherwise is to be guilty of a moral wrong.
The philosophy of non-degradation has a long history, as is clear to anyone
who has read Book I of Georgius Agricola's DE RE METALLICA, published in 1556.
This sixteenth century inventory of objections to disturbing the earth cite the
following:
"The earth does not conceal and remove from our eyes those things which
are useful and necessary to mankind, but on the contrary, like a beneficent
and kindly mother she yields in large abundance from her bounty and brings
into the light of day the herbs, vegetables, grains, and fruits, and
the trees. The minerals on the other hand she buries far beneath in the
depth of the ground; therefore they should not be sought. But they are dug
out by wicked men who, as the poets say, are the products of the Iron Age.
(6-7)
. . . The strongest argument of the detractors is that the fields are
devastated by mining operations. . . the woods and groves are cut down,
for there is a need of an endless amount of wood for timbers, machines,
and the smelting of metals. And when the woods and groves are felled,
then are exterminated the beasts and birds, very many of which furnish
a pleasant and agreeable food for man. Further, when the ores are
washed, the water which has been used poisons the brooks and streams,
and either destroys the fish or drives them away. . . Thus it is said,
it is clear to all that there is greater detriment from mining than
the value of the metals which the-,mining produces. (8)
And next they raise a great outcry against other metals, as iron, than
which they say nothing more pernicious could have been brought into the
Jtfefeof man Ear it is employed in making swords, javelins, spears,
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pikes, arrows -- weapons by which men are wounded, and which cause
slaughter, robbery, and wars. ... It is claimed too, that lead
is a pestilential and noxious metal, for men are punished by means of
molten lead. (11)
They contend that, inasmuch as Nature has conceded metals far within
the depths of the earth, and because they are not necessary to human
life, they are therefore despised and repudiated by the noblest, and
should not be mined, and seeing that when brought to light they
have always proved the cause of very great evils, it follows that
mining is not useful to mankind, but on the contrary harmful and
destructive." (11-12)
The MRDC appeal to a non-degradation philosophy only demonstrates that those who
are ignorant of history are condemned to repeat it.
Furthermore, the philosophy of non-degradation uncritically assumes the
idea that a benign environment is rapidly being ruined by human beings, However,
the historical record attests that an untamed environment has repeatedly wrought
massive human degradation through catastrophic effects of famines, plagues, floods,
tornadoes, earthquakes, etc. The fundamental problem, therefore, is not to maintain
some simplistic "non-degradation" of the environment. Rather the problem is a complex
one of devising appropriate means to protect both life-sustaining and aesthetic
qualities of the biosphere, and at the same time develop technologies which provide
basic human goods as a necessary condition for maintaining a preferred environmental
quality. As a fundamental, meaningful principle for securing that environmental
protection, "non-degradation" is vacuous.
As for preoccupation with risks to future generations and their proposed
principle of neutrality, the NRDC authors seem committed to perpetuating a politically
powerful, yet no less fraudulent myth -- namely, that the hazards of radioactive
wastes foist unprecedented risks onto unconsulted future generations because the
index of their hazard to the future is measured by and equivalent to the longevity
of their radioactive half-life. This is absurd. Intellectual honesty should compel
those who know better to state as often as necessary that any risks of adverse health
»
effects from radiation sources — both to present and future generations --must be measured
only in relation to environmental
pathways which determine the degree oflikelihood of harmful exposure of and assimilation
by the human body.
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which
Perhaps time alone can domesticate the exaggerated fears/surround radiation
in general, and radioactive wastes in particular. In the meantime, our concern for
the risks bequeathed to future generations will be better expressed if we reject
outright two simple-minded notions: (a) first, that such risks have an existence
in splendid isolation from the benefits which justify them, and (b) secondly,
that such ethical principles as equity and participation require a neutral allocation
of risks and benefits to the future.
The first notion merits rejection because the legacy of any generation to an
immediate as well as remote future is not mere "risks" and "hazards," but to the
contrary, an entire social order striving to provide material well-being, institutional
stability, and creative freedom for its citizenry. Risks and promises, harms and
benefits are inseparably interdependent within any sustainable social order.
As for the second notion, the ethical principle of equity requires a society
to provide its citizens with reliable access to those basic goods which sustain
material well-being. The principle of participation requires a society to provide
institutionalized methods of consent for its citizens, who in turn are obligated to
contribute to and abide by outcomes of those methods. It is nonsense for anyone to
arrogate to themselves the wisdom either to decide for future generations what is in
their best interest in securing basic goods and protection from basic harms, or
to suppress -- under the guise of "neutrality to the future" — any method of devising
conceptual tools which might enable the present generation to deal constructively
with its uncertainties and responsibilities toward the living. Ethical responsibility
is primarily for the living who happen to be the only foundation we have to provide
for the well-being of future generations.
From a bioethical perspective, there is ample justification for policy makers
in the present generation to establish criteria and standards for health protection
by reference to naturally occurring radiation sources from which man-made applications
derived. But it is not justifiable on the basis of a pseudo-philosophy of
or trusteeship over some pre-existing "natural state."
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Those responsible in society for providing basic goods, methods of informed
consent, and an equitable management of biohazards have an ethical obligation to
derive value judgments of safety, acceptable risk, and justifiable harm from a
philosophy of congruence with a pattern of benefits and harms already established
by naturally occurring radiation sources with which human beings have lived and
evolved throughout recorded history. That is to say, the philosophy of congruence
and of logical consistency require a policy maker to form value judgments on the
relative benefits of providing protection against radiation by first taking account of
wide variations in personal exposures and population exposure from naturally occurring
background sources.
External sources of exposure include cosmic rays, together with the radionuclides
they produce, and primordial radionuclides in the earth. Internal exposure from
natural radionuclides inhaled or ingested via food and drinking water augment that
external exposure appreciably (e.g. potassium-40 adds 17 mrem.) Large segments of
the population in the United States receive natural external radiation doses varying
from 41 to 105 mrem per year simply because of geographic location. Variations in
natural exposure to thorium in monazite sands along the southeastern coast of India
range from 130 mrem to 2,800 mrem; while on the coast of Brazil, exposure ranges
from 90 to 2,800 mrem with an average of 550 mrem per year. There is no scientifically
established evidence, despite contrived attempts to prove it, that there are
basic harms to those so exposed.
Human tolerance for, indeed dependence upon, such wide variations in natural
radiation sources for several millenia demonstrate that increments from man-made
applications of those natural sources can be kept well within the range of those
variations without inflicting unjustifiable harm or deprivation of basic goods.
I am fully aware that this conclusion is contrary to what has been assumed.
by regulatory agencies when they have set excessively conservative standards in
the past. With our increased knowledge of the benefits and harms of natural
radiation, however, there is ethical justification/ffcr their graduaT7e>ision.
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It is a matter of fact that the largest increment from man-made radiation exposure
conies from medical and dental health practices, and these exposures are 10 to
100 times greater than other man-made sources which by contrast are stringently
regulated.
From the perspective of bioethics, the inequitable management of biohazards
in general, and of radiation protection in particular, has become a matter for
national embarrassment. There is clearly a category of negligible risk and negligible
harm which in practice ought to be ignored. This category conincides with the
ethical principle of justifiable harm.
An application of the philosophy of congruence and of negligible harm is
already a part of the public record in testimony submitted to the .Nuclear Regulatory
I o
Commission in the matter of Perkins Nuclear Station, North Carolina. In public
testimony, comparisons have been made between radon releases which result from
mining and milling of uranium, with radon naturally released from the earth. According
to the record, Dr. R. L. Gotchy "provided calculations out to 10,000 years of the
comparative population exposure resulting from radon emanation from the nuclear
fuel cycle compared to the naturally occurring exposures. These calculations show
that exposures due to radon releases from mining and milling are insignificant
compared to natural background radiation exposures." Although agreeing with
Dr. Gotchy's estimates based upon the data he had used, Dr. Hamilton "decried
extrapolations of health effects Into the distant future as being misleading" and
not truly meaningful, because they do not take any account of repair mechanisms.
When questioned by the NRC Board, Dr. Hamilton testified that:
. . . variations in normal living style, traveling about the country
and going indoors or outdoors result in doses that are many orders
of magnitude greater than the increase in^dose resulting from
radon-222 emanating from tailings and mining.
He concluded that low levels of radiation such as these are "completely insignificant
and without any reality."
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The NRC Board concluded its findings by stating that the best mechanism
available to characterize the significance of increments released by mining and
milling is:
to compare such releases with those associated with natural background.
The increase in background associated with Perkins is so small compared
with background, and so small in comparison with the fluctuations in
background, as to be completely undetectable. Under such a circumstance
the impact cannot be significant.
IV. CONCLUSION: Suggested Bioethical Principles for Setting Criteria and
Standards for Radiation Health Protection
In view of the above reflections, I suggest that the following principles
might better serve as guidance in the formulation of social policies for radiation
health protection:
(1) Any involuntary risks imposed by social policies for radiation
protection must be congruent with, must not be in excess of, and
may be reasonably less than, those involuntary risks imposed by the
wide variations in naturally occurring toxic elements and harmful
effects from our natural environment.
(2) There must be evidence that basic goods and essential benefits
cannot be more satisfactorily obtained through other alternative
means which entail fewer risks.
(3) The basic goods and essential benefits must be demonstrated or
virtually assured, and they should outweigh the possibility of
basic harm to human well-being.
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REFERENCES
1. Hans Jonas, "Technology and Responsibility. Reflections on the New Tasks
of Ethics," Social Research 40 (1), Spring, 1973, 31-54
2. William C. Clark, "Managing the Unknown," in Managing Technological Hazard:
Research Needs and Opportunities, ed. by R. M. Kates (University of Colorado:
Institute of Behavioral Science, 1977), pp. 111-142
3. Ibid.
4. Silvan Tomkins, "Ideological Conflicts about the Nature of Risk-Taking Behavior,"
in Risk-Taking Behavior, Richard E. Carney, ed. (Springfield, 111.: Charles
Thomas, 1971), pp. 182-192
5. William F. May, "The Right to Know and the Right to Create," Science.
Technology and Human Values (#23; April 1978), pp. 34-41
6. Alan Gewirth, Reason and Morality. (Chicago: University of Chicago Press,
1978).pp. 48 fF!By normative structure of human action, Gewirth means.that
". ... every agent implicitly makes evaluative judgments about the goodness of
his purposes and hence about the necessary goodness of the freedom and well-being
that are necessary conditions of his acting to achieve his purposes. . . ;
every agent implicitly makes a deontic judgment in which he claims'that he has
rights to freedom and well being. . . ' every agent must claim these rights for
the sufficient reason that he is a prospective agent who has purposes he wants to
fulfill, so that he logically must accept the generalization that all prospective
purposive agents have rights to freedom and well-being." (48)
7. D. Okrent and C. Whipple, Approach to Societal Risk Acceptance Criteria and Risk
Management. University of California, Los Angeles, UCLA-ENG-7746, June 1977.
8. George Pickering, "Energy and Well-Being: Whose?" Proceedings of Energy:
The Ethical Issues. Springfield, Ohio: Ohio Institute for Appropriate
Technology, 9 December 1978.
9. Aaron Wildavsky, "No Risk Is the Highest Risk of All," AMERICAN SCIENTIST
(67; Jan-Feb, 1979) pp. 32-37.
10. G. Hoyt Whipple, "Low Level Radiation: Is There a Need To Reduce the Limit?"
Public Presentation, Atomic Industrial Forum Conference on Nuclear Power:
Issues and Audiences, Houston Texas, 10-13,September 1978.
11. Dimitri Rotow, Thomas Cochran, Arthur Tamplin,"NRDC Comments on Criteria for
Radioactive Waste Proposed by Environmental Protection Agency, Federal Register.
Vol. 43, No. 221, 15 Nov. 1973." Issued 5 January 1979. ,..«..«
Thomas B. Cochran and Dimitri Rotow, "Radioactive Waste Management Criteria,1!
Prepared for U. S. Department of Energy, Contract ER-78-C-01-6596, 5 January 1979.
12. Georgius Agricola, De Re Metal!ica.(New York: Dover Publications, 1950).
13. NRC Decision, Perkins Nuclear Station. Units 1.2. and 3, "Partial Initial
Decision Environmental Consequences of the Uranium Fuel tycle, 31 July 197o,
PP. 28,664-28,672.
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SESSION F
ENVIRONMENTAL - PUBLIC HEALTH ASPECTS II
Session Chairperson
R. M. Fry
State of Kentucky
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A REVIEW OF ENVIRONMENTAL SURVEILLANCE DATA AROUND LOW-LEVEL
WASTE DISPOSAL AREAS AT OAK RIDGE NATIONAL LABORATORY
Thomas W. Oakes and Kenneth E. Shank
Environmental Surveillance and Evaluation Section
Industrial Safety and Applied Health Physics Division
Oak Ridge National Laboratory*
Oak Ridge, Tennessee 37830
ABSTRACT
White Oak Creek and Melton Branch tributary surface
streams flow through the Oak Ridge National Laboratory (ORNL)
reservation and receive treated low-level radioactive liquid
waste which originates from various Laboratory operations.
The streams receive additional low-level liquid waste generated
by seepage of radioactive materials from solid-waste burial
grounds, hydrofracture sites, and intermediate-level liquid-
waste sites. Over the years, various liquid-waste treatment
and disposal processes have been employed at ORNL; some of
these processes have included: settling basins, impoundment,
storage tanks, evaporation, ground disposal in trenches and
pits, and hydrofracture. Burial of solid radioactive waste was
initiated in the early 1940's, and there are six burial grounds
at ORNL with two currently in use. Monitoring at White Oak Dam,
the last liquid control point for the Laboratory, was started
in the late 1940's and is continuing. Presently, a network of
five environmental monitoring stations is in operation to
monitor the radionuclide content of surface waters in the White
Oak watershed. In this paper, the solid waste burial grounds
will be described in detail, and the environmental data tabulated
over the past 29 years will be presented. The various monitoring
systems used during the years will also be reviewed. The liquid
effluent discharge trends at ORNL from the radioactive waste
operations will be discussed.
INTRODUCTION
Six solid radioactive waste disposal areas (SRWDA) have been used
since the operation of Oak Ridge National Laboratory (ORNL) began in 1943.
The location of these SRWDA's are shown in Figure 1. Location of the
first three SRWDA's were selected primarily for convenience (We 76). Very
little geologic or hydrologic considerations were given to the site
selections.
Operated by Union Carbide Corporation under contract W-7405-eng-26 with
the U.S. Department of Energy.
-------
ORNL-DWO 65-12157*7*
SOLID WASTE DISPOSAL
AREA NO. 3
X
;
y
1
,SOLIO WASTE DISPOSAL
AREA NO. 2
\
PROCESS WASTE
SETTLING .BASIN
I
SEWAGE TREATMENT PLANT
SOLID WASTE DISPOSAL
r
'SOLID WASTE DISPOSAL
AREA NO. I
l.L.W. TRENCHES
(NOT IN USE)
SAMPLING STATION
APPROXIMATE BOUNDARY
OF ORNL COMPLEX
1000
I...
0
.1
1000 2000
1 , I
FEET
-------
423
As the volume of waste increased, more attention was given to site
selection. Areas underlain by Conasauga shale formation make excellent
sites for waste disposal, as the shale is easily excavated and has ion
exchange properties that inhibit the migration of water-soluble nuclides
through the soil. Melton Valley is underlain by this formation and is the
location of three of the SRWDA's that became operational since 1951. The
current operational status and land area of the solid-waste areas are
given in Table 1. Other sources of contamination on the site include
settling basins, impoundments, trenches, and pits.
Table 1. Operational status of ORNL Radioactive Solid Waste
Storage Area
SRWDA
1
2
3
4
5
6
Operating Dates
1943 - 1944
1944 - 1946
1946 - 1951
1951 - 1959
1959
1969
Status
Closed
Closed
Closed
Closed
Operating
Operating
Land Used
(acres)*
1
4
7
23
33
68
2
One acre = 4047 m .
SOURCES OF CONTAMINATION
SRWDA No. 1
SRWDA No. 1, with a total area of one acre, is located at the foot of
Haw Ridge. It is at the edge of the Laboratory complex and is about 25 ft
south of White Oak Creek (We 76). This site was selected on the basis of
its proximity to the Laboratory, and no consideration of waste leaching
into the water system was given. Waste was dumped into open trenches and
backfilled. There are no available records showing the quantity or kind
of solid waste disposed of in these areas. Very little monitoring data is
available around SRWDA No. 1 (We 76). The date of closure of this area
was 1944; it was closed because water was found in one of the trenches.
In 1946, the site was surveyed for surface contamination, and soil
samples were analyzed. The results from only two areas indicated activity
above background. Water samples from two wells and a surface seep in this
area were analyzedqfor Sr, Cs, and transuranic elements in 1975. Low
concentrations of Sr (0-4-dpm/mJl) were present in one of the wells. No
detectable quantities of Cs and transuranic elements were found (Du 75).
SRWDA No. 2
SRWDA No. 2 was operated between 1943-1946, and covered a total area
of about four acres. The site is located on the lower half of a hill near
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424
the east entrance of the Laboratory. Selection of the site was most
likely based on consideration for the reduction of personnel exposure
during the transportation of the waste (We 76). No attention was given t
environmental protection.
There are no available records documenting the quantity or kind of
solid waste disposed of in this area. It has been ascertained that solid
waste contaminated by beta or gamma activity was placed in black iron dru
and buried in the trenches. Liquid waste contaminated by plutonium was
placed in stainless steel drums and either buried in trenches or stored
without burial in a "natural ravine" eroded in the denuded slope (We 76).
The use of the SRWDA No. 2 site was later found to be incompatible
with the long-range land-use planning at the Laboratory, and the operatic
was terminated in 1946. After closure, most of the waste is said to have
been exhumed and reburied in SRWDA No. 3. The stainless steel drums con-
taining liquid plutonium waste were removed intact, but the black iron
drums containing beta-gamma solid waste had deteriorated. Thus, the sur-
rounding earth was also removed and reburied at SRWDA No. 3. The hillsid
of the SRWDA No. 2 site was then bulldozed to smooth out the irregulariti
and was seeded (We 76).
During August, 1976, 13 core samples were collected at various point
in SRWDA No. 2. Water samples were also collected from the core holes.
Activity levels in water samples were found not to be significantly diffe
from baseline samples when analyzed for H, gross-alpha, and gross-beta
activity. A representative portion of the homogenized whole core was use
for this analysis. The average uranium and plutonium concentrations were
found to be 0.47 pCi/g and 0.06 pCi/g, respectively (Oa 77). The average
radioisotope concentration for soil samples near the perimeter of the DOE
area in Oak Ridge have been found to contain 0.66 pCi/g of uranium and 0.
pCi/g of plutonium (Oa 76).
137
The average Cs concentration for the upper third and the entire
core was measured to be 0.7 and 0.3 pCi/g, respectively. Both of these
values are substantially lower than the value of 1.0 pCi/g, the average
value of samples collected in 1976 from 16 sites throughout eastern and
central Tennessee (Oa 76). It should be noted that these soil samples a
fromggores several feet long and are being compared with topsoil samples.
For Sr, the average values for the cores were ^0.57 and ^0.53 pCi/g
for the upper third and the entire core, respectively (Oa 76) .
SRWDA No. 3
SRWDA No. 3 is about 0.6 miles west of the west entrance to the Labc
complex. The site is a flat, forested area at the foot of Haw Ridge.
Waste Area No. 3 presumably was chosen because of its proximity to the
Laboratory, out-of-sight location, and because the soil could be readily
excavated (We 76). The area became operational in 1946. Alpha-contamins
wastes were dumped in unlined trenches and covered with concrete, whereas
the beta-gamma waste was covered with native soil.
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425
Samples of well water from the area were analyzed inQ1964 and indicated
small amounts of the trivalent rare earths (TRE), Sr, Sr, and H (We 76)
Well samples were also collected in 1973 and analyses indicated Sr levels
as high as 3.0 dpm/m£. Soil samples were collected and analyzed during 1978
and the results are given in Eldridge (79).
SRWDA No. 4
During 1948-50, a study (St 51) of the geology and hydrology of the
Laboratory site was conducted. Disposal of waste in the Conasauga shale
belt was recommended. SRWDA No. 4 was opened in 1951 in the closest area
to the Laboratory underlain by Conasauga shale. Trench orientation was
variable and lacked any consistent relationship to original site topography
(We 76). Auger holes one to two feet in diameter in this area were used
for the disposal of higher radioactive level waste. The site was closed in
1959 and resulted in a total disposal area of 23 acres.
A number of small seeps have developed near the rim of the terrace in
the center third of the area, and others are reported to have developed in
the central part of the site. During 1959 and 1960, sampling of wells and
streams in and near this area indicated that both ground water and surface
water were contaminated (We 76) . Eight of the sixteen wells showed beta-
gamma contamination. qWater,samples from two seeps indicated contamination
of Sr, Cs, Zr- Nb, Co, and TRE. The section of Whi^Oak (jgeek
flowing-bv SRWDA No. 4 indicated radioactive contaminants of Ru, Sr,
J Po, Pu, and TRE (We 76). In 1964, water samplesgwere collected from
six wells and one seep-.and each were found to contain ' Sr, H, TRE,
and minor amounts of Ru (We 76). Discharges of Sr from SRWDA No. 4
versus precipitation are given in Table 2 from Stueber (78).
90
Table 2. Sr discharges vs. precipitation
_ _
Precipitation Total Sr Discharge
Water Year* cm Ci
1967
1968
1969
1970
1971
1972
1973
1974
1975
1976
1977
154
114
102
122
123
120
181
175
147
124
129
2.7
2.0
2.1
1.6
1.2
2.4
1.6
5.2
3.2
5.1
2.3
September 1 - August 31
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426
Soil samples were collected in 1973 gjong J^e south s^e of SRWDA Nc
These samples contained small amounts of Co, Cs, and Sr (Du 76).
soil along White Oak Creek east of the area has been contaminated by seep
from SRWDA No. 4 and discharges of the creek. Near this site is a contan
nated flood plain area which was once flooded by an intermediate pond. A
was constructed in early 1944 to help create an intermediate retention pc
between the Laboratory and White Oak Lake. The dam was breached in late
1944, and a small pond remained until 1950 (Du 76). Results of analyses
soil and sediment are given in the section on flood plains.
SRWDA No. 5
SRWDA No. 5 opened in 1958 and consisted of two sections on the hill
east of White Oak Creek and south of Haw Ridge. This area was opened in
1958, because space in SRWDA No. 4 was approaching exhaustion. Initially,
same burial procedures were used at this site as had been used at the pre
ceding sites; that is alpha-contaminated waste was interred in the lower
of the area and capped with concrete, and the beta-gamma contaminated was
was simply covered with weathered shale. This segregation procedure was
discontinued sometime during the operational life of the site. The trenc
lengths vary from < 40 feet to > 500 feet. These trenches were oriented
parallel to the topographic slopes (We 76). Water samples from-several w
were collected in 1964. The principal contaminants found were Sr, Sr
Ru, H, and TRE. Several new wells were cored and sampled. The data
suggested that at this time, only minor movement of radioactivity had
occurred. In 1960, samples from these wells indicated that SRWDA No. 5 w
the major source of H (We 76) in White Oak Creek.
Most of the transport of radionuclides in the surface water is
monitored at Station 4 on Melton Branch. Additional data are given in th
section on monitoring stations. In 1974, thirteen small seeps were sampl
along the south §dge of the arj^e These samples contained measurable amo
of total alpbjg, Sr, H, and Sb. Eleven of the samples contained con
trations of Sr ranging from 9 x 10 to 6.1 x 10 yCi/mJl (Du 76).
SRWDA No. 6
SRWDA No. 6 is located immediately northwest of White Oak Lake. This
site is about 70 acres in size and was opened in 1969. Trenches initiall
were excavated as long as possible, but are now limited to a length about
feet. This procedure was initiated in order to reduce the collection of
water in the trenches to an acceptable level (We 76). Some monitoring
around this area has been completed. The results indicate some movement
radioactivity, but it is too early to judge the true meaning of the resu]
Floodplain Areas
There are four floodplain areas on the site that are contaminated.
For the purpose of this paper, only the floodplain that was established t
the construction of the dam in 1944 will be discussed.
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427
137
During 1944, a study of the Cs distribution in soil, roots, ground
vegetation^ overstory, litter, mammals, flies, and insects from the 2 ha
(20,000 m ) flood plain area was made^o 76). The highest concentration
found in soil was 84,400 pCi/g. All Cs concentrations below the 17-cm
depth, outside the floodplain area were equal to background (Vo 76). Con-
centjjations in the roots ranged as high as 12,500 pCi/g. The concentration
of Cs in ground vegetation ranged from 4.6 to 182 pCi/g. Other results
can be found in Voris (76).
Waste Ponds
Another source of radioactive discharges from past waste treatment
procedures are the waste ponds. Three waste ponds have been in use at ORNL
over the years of operation. An example of the activity in these ponds is
from a study in2197Z, JJTa 77) of Waste.,Eond 2. This study indicated approxi-
mately 5 Ci of ' Pu, 200 Ci of 1J/Cs and 33 Ci of Sr in the bottom
sediment. Another source of discharged waste is the Intermediate-Level
Waste Pits.
Seepage Pits and Trenches
In 1951, the construction of pits for disposal of intermediate-level
liquid was begun. The first pit was opened in 1951 and immediately closed
because of its poor location. The second pit was opened in 1952, and large
quantities of intermediate-level waste were disposed of for the first time.
Pits 3 and 4 became operational in 1955 and 1956, respectively. Trenches 5,
6, and 7 were opened between 1960-1962. These trenches were taken out of
routine service in 1965 as part of a plan to implement disposal of intermediate-
level waste by hydrofracture (Du 75).
Small amounts of Sr and Cs have been observed in-seepage from
trenches 6 and 7. The major seepage problem has been with Ru, as seen in
Table 3 and Figure 2.
Table 3. Annual discharges of Ru to the Clinch River
Year Ci Year Ci
1959
1960
1961
1962
1963
1964
520
1900
2000
1400
430
191
1965
1966
1967
1968
1969
1970
69
29
17
5
1.7
1.2
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428
10000
ORNL-DWG. 79-8861
1000
100
10
-
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\ /
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1958 1960 1962 1964
YEAR
1966
1968
Fig. 2. Annual discharges of 106RU to the Clinch River
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429
WATER SAMPLING
Low-level radioactive liquid wastes originating from ORNL operations
are discharged, after preliminary treatment, to White Oak Creek and Melton
Branch, which are small tributaries of the Clinch River. The radioactive
content of the White Oak Creek discharge is determined at White Oak Creek
Stations 1, 2, and 3 and at White Oak Dam, which is the last control point
along the stream prior to the entry of White Oak Creek into the Clinch
River. Samples are collected at several locations in the Clinch River,
beginning at a point above the entry of the wastes into the River and
ending at Center's Ferry near Kingston, Tennessee, the nearest population
center downstream. Location maps of these stations are given in Figures 1
and 3.
Station 1 - White Oak Creek Station 1 monitors the effluent from the
ORNL Process Waste Treatment Plant.
Station 2 - Station 2 is located on White Oak Creek a short distance
upstream from Station 1 and provides data on radionuclide content from
operation discharges above the Process Waste Treatment Plant.
Station 3 Water Monitoring Station 3 is located on White Oak Creek a
short distance above the confluence of White Oak Creek and Melton Branch.
This station measures the streamflow and radionuclides content from the
ORNL plant effluents; SRWDA Nos. 1, 2, 3, 4, and portions of 5; contami-
nated floodplain sediments; and other potential sources.
Station 4 Station 4 is located on Melton Branch, a short distance
above the confluence with White Oak Creek. This station measures stream-
flow and radionuclide content from SRWDA No. 5, several experimental
reactor sites, and other areas.
Station 5 - Samples of White Oak Creek effluent are collected at
White Oak Dam by a continuous proportional sampler which was designed and
constructed at ORNL. Proportional sampling is necessary to obtain a truly
representative sample, since streamflow and concentration of radioactive
materials in the stream may vary independently over a relatively wide
range in a relatively short period of time, depending upon weather and
operating conditions. Streamflow at White Oak Dam is measured by means
of a Stevens water-level recorder and stilling well in the lake pool in
conjunction with the White Oak Dam gate, which serves as a rectangular
weir through which the water flows. The weir system was rated by the U.S.
Geological Survey and found to be accurate to within 5%.
Samples are collected weekly from White Oak Dam and analyzed for
gross beta activity as a control measure and as a means of evaluating the
gross concentration of radioactivity entering the Clinch River. Portions
of the weekly samples are composited, proportional to the flow, into
monthly composite samples that are subjected to more detailed analyses by
wet chemical and gamma spectrometric techniques. The weekly samples are
-------
K 25
WATER INTAKE
CRM 14.5
WHITE OAK DAM
MONITORING
STATION
BfllOCE
MELTON HILL DAM
CRM 23.1
WATER°"SAMPUNG STATION
Fig. 3. Water monitoring locations in the Clinch River.
-P.
w
o
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431
analyzed for the transuranic alpha emitters, total strontium, and 131I,
which represent the elements in the waste stream with the highest hazard
indices.
The monthly composites are concentrated and analyzed by radiochemical
and gamma spectrometric techniques, normally for the following radionu-
clides: yUSr, 1J/Cs, X^Ce, iU6Ru, 95Zr-95Nb, 5°Co, \ trivalent rare
earths, and gross beta. Analyses for other nuclides may be performed as
the need arises. These analyses are performed to determine the percentage
distribution and concentrations of the various nuclides in the effluent
stream and to calculate the quantity of each radionuclide released to the
Clinch River. More frequent analyses are made if concentration levels in
White Oak Creek vary significantly from the experienced normal.
Calculations are made of the concentrations of radioactivity in the
Clinch River for the point of entry of the wastes, using the concentrations
measured at White Oak Dam and the dilution provided by the River. These
calculations are based on uniform mixing of the two streams within a short
distance downstream from the point of entry of the wastes. The calculated
concentration of each radionuclide in the River is compared with its respect
MFC value as specified by Chapter 0524 of the DOE Manual and the resulting
fractions are summed to arrive at the % MFC in the Clinch River.
w
The annual discharges of radionuclides to the Clinch River as measured
at White Oak Dam from 1949-1978 are given in Tables 4 and 5. The measured
MFC at White Oak Dam and the calculated MFC in the Clinch River from
1974-1978 are given in Figures 4 and 5, respectively. A comparison of MFC
at White Oak Dam and the calculated value in the River from 1974-1978 is
given in Table 6. The total amount of radionuclides discharged into the
Clinch River is given in Table 7. The amount of H released to the river
in three-year periods from 1965-1977 is given in Figure 6.
As a followup to this calculated concentration, two sampling stations
are maintained in the Clinch River below the point of entry of the wastes:
one at the ORGDP water intake, Clinch River mile (CRM) 14.5, and the other
at Center's Ferry near Kingston, Tennessee (CRM 4.5). In addition, a
sampling station is maintained in the Clinch River above the point of entry
of the waste at Melton Hill Dam (CRM 23.1) to provide background data.
ORGDP Water Intake Sampling Station - The ORGDP water sampling
station, which was designed and constructed at ORNL, collects a sample from
the Clinch River that is proportional to the flow in the River near the
water intake of the ORGDP plant water system (the first point of Clinch
River water usage downstream from the point of entry of ORNL wastes). The
samples are brought into the laboratory at weekly intervals, acidified, and
combined into quarterly composite samples for analysis. The quarterly
composite samples are concentrated by evaporation and analyzed by wet
chemical and gamma spectrometric techniques for gross activity and for
those radionuclides present in significant amounts. Average concentrations
of individual radionuclides are determined from the analytical data,
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432
Table 4. Annual discharges (Ci) of radionuclides to the Clinch River
1949-1963
Year
1949
1950
1951
1952
1953
1954
1955
1956
1957
1958
1959
1960
1961
1962
1963
137Cs
77
19
20
9.9
6.4
22
63
170
89
55
76
31
15
5.6
3.5
106D
Ru
110
23
18
15
26
11
31
29
60
42
520
1900
2000
1400
430
89Sr
NA
0.28
1.9
2.0
1.7
0.98
90Sr
150
38
29
72
130
140
93
100
83
150
60
28
22
9.4
7.8
TRE*(-Ce)
77
30
11
26
110
160
150
140
110
240
94
48
24
11
9.4
144Ce
18
NA
NA
23
6.7
24
85
59
13
30
48
27
4.2
1.2
1.5
95Zr
180
15
4.5
19
7.6
14
5.2
12
23
6.0
27
38
20
2.2
0.34
95Nb
22
42
2.2
18
3.6
9.2
5.7
15
.7.1
6.0
30
45
70
7.7
0.71
131X
77
19
18
20
2.1
3.5
7.0
3.5
1.2
8.2
0.5
5.3
3.7
0.36
0.44
60Co
NA
6.6
46
4.8
8.7
77
72
31
14
14
3H TRU
NA** 0,009 (for
0.0379
0.0734
0.0296
0.0838
0.0729
0.2491
0.2840
0.1454
0.0755
0.6770
0.1860
0.0682
0.0627
0.1660
Total rare earths minus cerium.
**No analysis performed.
Table 5. Annual discharges (Ci) of radionuclides to the Clinch River
L9'64-1978
Year
1964
1965
1966
1967
1968
1969
1970
1971
1972
1973
1974
1975
1976
1977
1978
137Cs
6.0
2.1
1.6
2.7
1.1
1.4
2.0
0.9
1.7
2.3
1.2
0.62
0.24
0.21
0.27
106D
Ru
191
69
29
17
5
1.7
1.2
0.50
0.52
0.69
0.22
0.30
0.16
0.20
0.21
89Sr
0.79
0.59
0.85
0.73
0.55
0.31
0.27
0.20
NA**
90Sr
6.6
3.4
3.0
5.1
2.8
3.1
3.9
3.4
6.05
6.7
6.0
7.2
4.5
2.7
2.0
TRE*(-Ce)
13
5.9
4.9
8.5
4.4
4.6
4.7
2.9
5.2
NA
"*C.
0.3
0.1
0.1
0.2
0.03
0.02
0.06
0.05
0.03
0.02
0.02
NA
95Zr
0.16
0.33
0.67
0.49
0.27
0.18
0.02
0,01
0.01
0.05
0.02
NA
95Nb
0.07
0.33
0.67
0.49
0.27
0.18
0.02
0.01
0.01
0.05
0.02
NA
131T
0.29
0.20
0.24
0.91
0.31
0.54
0.32
0.21
0,34
0.46
0.23
0.28
0.03
0.03
0.04
6°Co
15
12
7.4
3.1
1.2
1.0
1.0
0.8
1.3
1.1
0.60
0,50
0.86
0.44
0.36
3H
1929
1161
3090
13273
9685
12247
9473
8945
10600
15000
8633
11061
7422
6249
6292
TRU
0.0775
0.4979
0.1599
1.0335
0.0440
0.1989
0.4003
0.0456
0.0687
0.0759
0.02
0.02
0.01
0.03
0.03
Total rare earths minus cerium.
**No analysis performed.
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433
250
200
150
100
50
ORNL-DWG 79-8857
1974 1975 1976 1977 1978
YEAR
Fig. A. Percent of MFC total over White Oak Dam.
0.50
0.40
0.30
0.20
0.10
ORNL-DWG. 79-8860
1974 1975 1976 1977 1978
YEAR
Fig
. 5. Percent of MFC total for dilution calculation in the
Clinch River.
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434
Table 6. Total percent MFC for all radionuclides
1974-1978
Year
1974
1975
1976
1977
1978
White Oak
Dam
203%
234
160
112
75
D.F.
688
646
423
537
481
Calc. C.R.*
0.29%
0.36
0.38
0.21
0.16
Calculated concentrations in Clinch River using dilution
factors.
Table 7. Total amount of radionuclides discharged
into Clinch River
1949-1978
Time Interval
1949-78
1949-78
1949-78
1949-78
1949-78
1949-78
1964-78
1955-78
1949-74
1949-71
Radionuclide
90
USr
95
XT1»
QS
7
106.,
Ru
131I
137Cs
3H
6°Co
144Ce
TRE(-Ce)*
Ci
1,179
286
375
6,932
174
687
124,945
321
342
1,289
Trivalentrare earths minus cerium.
-------
435
o
36000
32000
24000
16000
8000
ORNL-DWG 79-8855
1965 1968
1971
YEAR
1974 1977
Fig. 6. Tritium releases to the Clinch River (3-yr. intervals).
0.30
o
Q.
0.20
0.10
y
ORNL-DWG. 79-8858
\
r
A
1974 1975 1976 1977 1978
YEAR
Fig. 7. Percent of MFC total at the ORGDP intake.
w
-------
436
and the percent MFC for the mixture is calculated from the concentration
values measured. T^e MFC at this point from 1974-1978 are given in
Figure 7- W
The ORGDP sampling station was established to provide data relative to
the concentrations of radioactivity in water taken from the Clinch River
for normal treatment plant usage, to provide an index of the hazard to the
ORGDP plant population as the first user of Clinch River water downstream
from the point of entry of the ORNL wastes, and to provide data for com-
parison with the concentrations in the Clinch River calculated from White
Oak Dam releases and the dilution provided by the River.
Center's Ferry Sampling Station - A "grab" sample is collected daily
at the Center's Ferry sampling station which is located on the Clinch River
at CRM 4.5. Thermal stratification exists at this location, with the cold
water of the Clinch River running under the warmer backwater of the Watts
Bar Reservoir. The sample is collected 25 ft below the surface of the
water to ensure the collection of Clinch River water for the sample. Fluc-
tuations in concentration at this location are relatively small due to the
distance downstream from the point of entry of the wastes, and dispersion
of radioactivity in the Clinch River water is complete; thus a grab sample
is considered adequate. The daily grab samples are composited, acidified,
and analyzed on a quarterly basis. The preparation of the sample and the
analyses performed are the same as those cited for the ORGDP water sampling
station.
Samples from the Center's Ferry sampling station provide data relative
to the average concentration of radioactive materials in the Clinch River
at the nearest population center (Kingston, Tennessee) downstream from the
point of entry of the wastes. The MFC from 1976-1978 is given in Figure 8.
A comparison of the calculated MFC atWORGDP and Center's Ferry is given in
Table 8. w
Table 8. Total percent MFC i Clinch River
w
Year ORGDP C.F.* Calc. C.R.**
1974
1975
1976
1977
1978
0.26%
0.23
0.15
0.10
0.11
0.21%
0.15
0.15
0.12
0.12
0.36%
0.49
0.51
0.28
0.24
Center's Ferry
**
Calculated concentrations in Clinch River
using dilution factors
-------
437
0.30
0.20
o
Q.
55?
0.10
ORNL-DWG. 79-8859
1974 1975 1976
YEAR
1977 1978
Fig. 3. Percent of MPCW total at Kingston, Tennessee.
800
ORNL-DWG. 79-8856
600
cc
o
o
a.
400
200
I
30
ORNL DISCHARGE
CRM 20.8
20
CLINCH RIVER MILE
Fig
. 9. Cesium-137 in river silt (1967).
-------
438
Melton Hill Dam Sampling Station - The Melton Hill Dam sampling station
collects a sample proportional to the flow of water through the power-
generating turbine, which represents all of the discharge from the Dam
other than a minor amount discharged in the operation of the lock. The
sampler was designed and constructed at ORNL and is located on the lower
side of the Dam, with the intake positioned at the tail race of the turbines.
The sampler is keyed to the turbine operation such that a sample is col-
lected only when the turbines are operating, even though water from the
tail race is continuously pumped through the sampling lines. Samples are
collected from the station at weekly intervals, acidified, and composited
for a three-month period in polyethylene containers. The quarterly sample
is processed and analyzed in the manner set forth (above) for the ORGDP
water sampling station.
In addition to meeting routine monitoring requirements, water monitoring
aids-in the detection of abnormalities. For example, in 1969, high levels
of Co were detected in the water at Melton Hill (GEM 23.1). River silt
was then analyzed and found to contain 16 pCi/g of Co above.Melton Hill.
Concentration below ORNL discharges was 1.2 pCi/g (CRM 19.1). Investigation
resulted in the finding of a waste tank leak at American Nuclear Corporation
located at approximately CRM 51. Other types, of samples, such as sediment
and fish, are also collected for the purpose of monitoring the discharges
from the plant site.
Fish Sampling
Several species of fish from the Clinch River are sampled. Ten fish
of each species are composited for each sample, and the-samples are analyzed
by gamma spectrometry and radiochemical techniques for the critical radio-
nuclides that contribute to dose. An estimate of man's intake of radio-
nuclides from eating the fish is made by assuming an annual rate of fish
consumption of 37 Ibs. The estimated percentage of MPI is calculated by
assuming a maximum permissible intake of fish to be comparable to daily
intake of 2.2£ of water containing the MFC of these radionuclides for a
period of one year. The highest recorded $PI in the last 15 years for any
species has been 8.1%.
Sediment Sampling
Annual average concentrations of radionuclides in the surface layers
of bottom sediments in the Clinch and Tennessee Rivers were conducted from
1951-1969. This information is relevant to the dispersal and movement of
radioactivity released from ORNL. Data varied from year to year with the
changes in the amounts of radionuclides released. The principal radio-
nuclid|0gontained1^ the bottom sediments is 7Cs with lesser amounts of
Co, Ru, and Ce being retained. Only a small amount of 1 Sr released
to the river is incorporated £g the bottom sediments. Water analyses have
shown that nearly all of the Sr is either in solution or associated
principally with suspended solids and, thus, passes through the river sys-
tem. Sediment data can indicate discharges as shown in Figure 9.
-------
439
SUMMARY
I. Facts with regard to the releases from ORNL waste practices over the
last twenty years.
A. Ruthenium Releases.
1. A large amount of ruthenium was released during 1959-64.
This reflected the seepage of Ru from the ILW waste
pit disposal area following the transfer of a large
quantity of material to the pit in 1959.
2. The Conasauga shale did not retain ruthenium as well
as other radionuclides.
B. Tritium Releases.
1. Large quantities of tritium have been released to the
Clinch River in recent years.
2. Origin of the bulk of the tritium is apparently the
burial grounds where tritium-bearing waste from another
laboratory was buried.
3. From a practical standpoint, little can be done to
inhibit or control these releases.
C. There has been a downward trend in the number of curies released
for all other radionuclides.
Ci
Radionuclides 1959 1978
6°Co 77 0.36
90Sr 60 2.0
131I 0.5 0.04
137Cs 76 0.27
II. A number of corrective measures have been taken at ORNL to reduce the
discharges. Some of these are:
A. The method of disposal of liquid waste into pits and trenches was
discontinued in 1965, and the use of hydrofracture was begun in
1966.
B. Rerouting of White Oak Creek to bypass floodplain areas.
C. The installation of the new Process Waste Treatment Plant
which becameQoperational in April, 1976, has significantly
reduced the Sr discharges to White Oak Creek from facility
operations.
-------
440
D. Paved intercepter ditches have been placed in the burial
ground area to reduce the surface runoff in that area.
E. Hydrological consideration is being reviewed in the selection
of a new burial site.
F. Small trenches are being used to reduce the amount of water
collected in them after disposal has occurred.
G. Volume reduction programs have been initiated.
Better monitoring programs are also being initiated. Other proposed
corrective measures for the reduction of releases and proposed facili-
ties for the disposal of low-level waste are being considered. These
efforts have resulted in a reduction in the discharges from the site
as has been shown in this paper.
-------
441
REFERENCES
DU 75 Duguid, J. 0., Status Report on Radioactivity Movement from
Burial Grounds in Melton and Bethel Valleys, ORNL-5017 (1975).
Du 76 Duguid, J. 0., Annual Progress Report of Burial Ground Studies
at Oak Ridge National Laboratory, Period Ending September 30,
1975, ORNL-5141 (1976).
El 79 Eldridge, J. S., Oakes, T. W., Shank, K. E., and Stueber, A. M.,
"Instrumental Methods Used in Environmental Surveillance Programs
Around a Low-Level Radioactive Burial Site: in Proceedings of
the Twelfth Midyear Topical Symposium of the Health Physics
Society, Williamsburg, Virginia, February 12-15, 1979.
Oa 76 Oakes, T. W., Shank, K. E., and Easterly, C. E., "Natural and
Man-Made Radionuclide Concentrations in Tennessee Soils" in
Proceedings of the Tenth Midyear Topical Symposium of the
Health Physics Society, Saratoga Springs, New York, October 11-
13, 1976.
Oa 77 Oakes, T. W., and Shank, K. E., Subsurface Investigation of the
Energy System Research Laboratory Site at Oak Ridge National
Laboratory, ORNL-TM-5695 (1977).
St 51 Stockdale, P. B., Geological Conditions at the Oak Ridge National
Laboratory (X-10) Area Relevant to the Disposal of Radioactive
Waste, ORO-58 (1951).
St 78 Stuaber, A. M. , Edgar, D. E. , McFadden, A. F., and Scott, T. G.,
Preliminary Investigation of Sr in White Oak Creek Between
Monitoring Stations 2 and 3, Oak Ridge National Laboratory,
ORNL/TM-6510 (1978).
We 76 Webster, D. A., A Review of Hydrologic and Geologic Conditions
Related to the Radioactive Solid-Waste Burial Grounds at Oak
Ridge National Laboratory, Oak Ridge, Tennessee, Open File
Report 76-727, U.S. Department of the Interior, Geological
Survey, 1976.
Vo 76 Voris, P. V., and Dahlman, R. C., Floodplain Data: Ecosystem
Characteristics and Cs Concentrations in Biota and Soil,
ORNL/TM-5526 (1976) .
-------
442
RHO-SA-99
EVALUATION OF A DECOMMISSIONED RADWASTE POND
D. Paine, K. R. Price+, and R. M. Mitchell
Rockwell Hanford Operations, Richland, Washington
Abstract - An eight hectare radwaste pond (216-S-17) which received cool-
ing water effluent from a nuclear fuel reprocessing plant at Hanford from 1951
to 1954 was contaminated due to unplanned releases. Subsequently, it was de-
commissioned by covering the area with 45-60 cm of backfill. Soil erosion and
nuisance contamination, in the form of tumbleweeds (Salsola kali), occurred
between 1954 and 1972 due to the lack of a specific revegetation program.
Siberian Wheatgrass (Agropyron sibericum) and Cereal Rye (Secale cereale) were
planted in 1972 and allowed to grow under natural conditioTi?! A" routine
evaluation of the site disclosed the presence of contaminated Siberian
Wheatgrass plants. This report describes the results of a radioecological
study of the site in 1978. Nondestructive methods developed for in situ
evaluation to determine radionuclide inventories and transport parameters for
biotic and abiotic compartments are presented concomitant with standard
procedures. Results indicate that Siberian Wheatgrass is a suitable perennial
for revegetation of low-level waste disposal sites in an arid environment.
INTRODUCTION
In October 1951, 216-S-17 Pond became operational by receiving process-
vessel cooling water and steam condensate from the Redox nuclear fuel repro-
cessing plant (Fig. 1). The original pond site was formed by creating an
earthen,dike approximately one meter high on the north and west sides of the
polygonal area indicated in Figure 2. The designated use area was approxi-
mately 23 hectares, but photographs indicate that only 8 hectares were
inundated at any time. The pond averaged about 0.3 meters deep with a
maximum depth of 0.6 meters.
S-17 Pond received effluent until April 1954 when it was deactivated due
to unplanned releases of fission products and consequently radionuclide concen-
trations in the pond sediments "exceeded prescribed limits" . A deter-
mination was then made to plug the pipeline near the outlet and the pond was
allowed to dry up. The contaminated pond area was reportedly covered with
45-60 cm of backfill to prevent the spread of the major contaminants, i.e.;
137Cs, 90Sr, and 238y. Due to extensive wind erosion near the inlet at
the northeast corner (Fig. 2), additional soil was added to a depth of
approximately 105 cm above the contaminated sediment layer. In succeeding
years, wind erosion and various other factors precluded the establishment of
native vegetation in most of the pond area. However, introduced weeds such as
tumbleweeds became well established in some areas and where erosion had
reduced the backfill depth their roots penetrated the contaminated sediment
layer. This resulted in the accumulation of "nuisance quantities" of 137r<;
and 90sr in plant tissues ( C177)
+Pacific Northwest Laboratory, Richland, Washington
-------
443
RHO-SA-99
200 AREA PLATEAU
RADIOECOLOGY
FIELD
LABORATORY
-N-
1000 YDS
'. ' . '
1000 METERS
216-S-16
(REDOX POND ID
SITE
200 W
POWERHOUSE:
Z PLANT
U PLANT
'/*
216-U-10
(U POND)
202-S
(REDOX)
-EAST GATE
216-S-10 DITCH
(REDOX POND I)
SITE
"SURE 1. Map showing the locations of the 200 Area environs and 216-S-17 Pond.
-------
200W
PERIMETER
FENCE
BOUNDARY OF AREA POTENTIALLY
INUNDATED BY
216-S-17
(REDOX POND I SITE)
216-S-16
REDOXPONDI
SITE
11
10
ROAD
500 YARDS
I I I I I 1
500 METERS
SCALE
o
i
CO
VO
FIGURE 2. Detailed map of the 216-S-17 Pond site and surrounding environs
-------
445
RHO-SA-99
In an attempt to control these problems, the western portion of the pond was
seeded to Siberian Wheatgrass, an introduced range species from Eurasia The
revegetation attempt was successful and by 1975 Siberian Wheatgrass covered a
large portion of the original pond area. This new plant was successful in
stabilizing the surface soil and the numbers of tumbleweeds were greatly
reduced through interspecific competition. However, wind erosion during the
peri9d from decommissioning in 1954 until reseeding in 1972 had reduced the
original backfill in some areas to a depth less than 15 cm above the
contaminated sediment layer. Subsequently, laboratory analyses of the
Siberian Wheatgrass collected from these areas indicated that some of the
plants had incorporated radionuclides into their leaf tissues
The objectives of this study were to determine radionuclide inventories
for major biotic and abiotic compartments of the pond and evaluate nondestruc-
tive testing methods utilizing field instrumentation for in situ radionuclide
analyses.
METHODS AND MATERIALS
Field sampling protocol
Sampling grid
Due to the irregular shape of the pond, a modified 30 meter quadrat grid
system was utilized as a guide for vegetative sampling to assess radionuclide
uptake. (Fig. 3) A theodolite was used to survey in the grid lines with
metal posts placed at each 60 meter sampling location. Smaller wooden stakes
marked the 30 meter intermediate locations. Peripheral areas of the actual
pond site were not included in the grid. Only those areas supporting bunch-
grasses with a density of at least 1 plant/100m2 were considered. The grid
comprised an area of approximately 7.6 hectares.
Vegetation sample removal
At each of the sampling locations two nearest Siberian Wheatgrass plants
over 15 cm in diameter were selected for anlysis. Samples were obtained by
cropping the bunchgrass at the base. Each plant was placed into a paper bag,
labeled with the sample location number, and then transported to the labora-
tory for gamma energy analyses.
In situ analyses
Approximately five months following the vegetation sampling, a special-
Purpose mobile van (Dev Van I) designed to support in situ measurements of
environmental radionuclide levels was utilized for additional field measure-
ments. As a test of the unit's capability for providing credible, fast,
inexpensive data concerning surface/subsurface waste, work was initiated to
acquire concentration data for the vegetative sampling grid utilizing those
same Wheatgrass plants sampled earlier (Fig. 3). This provided an exce lent
opportunity to conduct a direct comparison of results, costs, and conclusions.
-------
446
RHO-SA-99
INSITUGeLi AND VEGETATIVE
SAMPLING STATIONS 7 4 .. *)
/
/
/
/
/(
10
8 7 6
1
i 3 2 1
FIGURE 3. Maps showing locations of study area in relation to S-17 Pond
boundaries: (a) location of field sampling locations; (b) location
of the vegetation samples for lab analyses; (c) location of the soil
core samples; (d) location of the small mammal trapping grid.
-------
447
RHO-SA-99
A collimated GeLi detector was employed to obtain in situ quantita-
tive measurements of radionuclides contained in vegetation and soil. Signal
output from the GeLi detector was interphased to a 1024 channel analyzer
Spectral data were transferred directly to magnetic tape and transported'to
the laboratory for further data reduction. A detailed description of the modi-
fied van unit can be found in Bruns (Br77). A 137Cs source was used 1n th
field for instrument calibration before actual sample-site counting was
initiated.
Field measurements were made with the GeLi detector mounted on a conical
lead shield 36 cm high and 36 cm in diameter at the base. Four hundred feet
of coaxial cable connected the detector to the analyzer. All measurements
were made at least 20 to 30 meters distant from the support van. Sample
points were counted for five minutes.
Soil core sampling
A modified 30 meter quadrat grid system was used as a guide for site
characterization work (Fig. 3c). Thirty-one cores were collected utilizing
the stratified random grid sampling technique.
A tripod rig and hammer were used to drive a 6 cm diameter split-tube
core sampler, utilized to reduce cross-contamination, to a depth of 125 cm.
Use of the split-tube sampler permitted acquisition and transport of the
samples as plastic sheathed, undisturbed sections approximately 38 cm in
length. There were three such sections completed per 125 cm core sample.
Drill depth and core length differed due to soil compaction by hammer impact.
Core samples were transported in .5 liter cans to a laboratory hood for
further subsampling. Each 125 cm core sample was divided by placing the 5 cm
surface portion in one can and each ten centimeter portion thereafter in
individual cans. The soil samples were oven-dried and weighed. Each soil
portion was counted in the lab for ten minutes using a collimated GeLi
detector.
Small mammal population
In order to sample the small mammal population of the pond area, a 10 by
10 trap grid was permanently staked. (Fig. 3d). Baited, Sherman folding
live-traps were placed 15 meters apart, utilizing one trap per grid station.
The grid was operated for at least four days per month from March through
December.
The captured animals were examined and data were recorded concerning
species, sex, reproductive condition, abnormalities, and weight. They were
then toe-clipped for permanent identification and released at the point of
capture. Only those animals which suffered mortality as a result ot trap
exposure (N=27) were collected for radionuclide analysis. Whole carcasses
were counted to determine total body burden (pelt and GI tract were not
removed) of 137Cs. Population estimates were calculated using the Jo11y-
Seber method (Jo65) and density estimates were based on the boundary-strip
method (St54).
-------
448
RHO-SA-99
Additional soil samples were collected from 20 mammal burrows and 11 ant
mounds for 137r,s analysis to determine if these animals were transporting
subsurface contamination above ground. These samples were randomly collected
from the mammal or ant burrow nearest to each sample core site (Fig. 3c).
Laboratory analytical methods
Gamma pulse height (GeLi), strontium 89-90 (wet chemistry), and uranium
analyses (spectrophotometer) were performed. Samples of soil, vegetation, and
mammals were analyzed for comparison with the in situ GeLi measurements to
establish appropriate double-sample techniques.
RESULTS AND DISCUSSION
The estimated inventory of principal radionuclides in the S-17 Pond
system calculated from discharge data is presented in Table 1 . The
present study considered all major radionuclides. However, only data for
137cs have been returned from laboratory analyses and are discussed here.
Table 2 presents inventory data for 13?cs calculated from this study for the
major abiotic and biotic components of the pond site. Figure 4 shows the
calculated spatial variability of 137cs. The concentration contour map
(Fig. 4a) was generated by an aerial radiological survey (Ti75) and the sur-
face concentration GeLi detector contours (Fig. 4b) were determined with por-
table instrumentation associated with Dev Van I. The surface water map (Fig.
4c) shows the relationship between the sample grid and the observed surface
water distribution in 1953. The final map (Fig. 4d) depicts the concentration
isopleths for the Siberian wheatgrass samples collected for comparative lab
analyses. The vertical distribution of 137cs in the soil column is pre-
sented in Figure 5. This was the mean depth profile based on weighted means
from each of the concentration contours generated by the in situ field
measurements (Fig. 4b).
Soil
The average surface soil concentration associated with the S-17 sample
grid was approximately 17 pCi/gm (Table 2). The average levels of 137cs in
soils of the 200 Areas from worldwide fallout and plant operations range from
1-3 pCi/gm dry weight . The overall soil inventory for the site was
estimated to be 6.3 Ci (Table 2). This value agrees reasonably well with the
estimated inventory of 17 Ci calculated from discharge information (Table 1),
considering the fact that the study grid comprised only 30 to 40 percent of
the potentially contaminated area (Fig. 4c). About .1 Ci of the soil total
for 137Cs is associated with the top 5 cm of soil. Therefore, assuming that
the top 5 cm are potentially susceptible to wind erosion, a maximum of only
.02 percent of the total soil contamination would be available for atmospheric
resuspension.
The vertical profile data (Fig. 5) show a majority of the contaminated
sediments to be located at the 5 to 35 cm depth. Although the pond site was
reportedly backfilled with 45 to 60 cm of soil , the vertical distri-
bution data indicate that there was insufficient backfill in some areas or
-------
449
RHO-SA-99
Table 1. Inventory of major radionuclides discharged to 216-S-17 Pond
Radionuclide
Pu
90Sr
137Cs
U
Curies
At Time of
Discharge
0.2
40.0
30.0
0.05
Curies Decayed
Through
1/1/79
0.2
21.1
16.5
0.05
-------
Table 2. Cesium - 137 Inventory for Major Compartments of S-17 Pond Study Area (x + SE)
Compartment N
Mean
pCi/gm
ViCi/rr
Total 137cs
Per Organism
(pCi)
Totals
Percent
of Total
Total Soil 381
6.3 Ci t 3.6 Ci
99.998%
Surface Soil* 31 17 + 5.3
1.8 ± .6
1.3 x 10s + 4.2 x 10* uCi .02%
Vegetation 165 4.7 + .7 1.5 x 10~3 + 2 x 10~" 2220 + 330 110 ± 16 uCi
.002%
Small Mairmals** 27 1.9 + .7
26 + 10 0.004 + 0.001 pCi
TOTAL 6.3 + 3.6 Ci
2 x 10-"%
*urface soil values are for only the top 5 cm of soil and are included in the total soil inventory
values for mammals are wet weight
fi
01
o
-------
CONCENTRATION CONTOUR
Cs-137(|lCI/m2)
(a I
451
137 Cs SURFACE CONCENTRATIONJGeLI DETECTOR
(uCI/m2|
OBSERVED SURFACE WATER
FROM JUNE 1453
AERIAL PHOTOGRAPH
137 Cs VEGETATION CONCENTRATION
luCI/m2!
Id)
FIGURE 4 Maps relating spatial variability of 137Cs associated with S-17
Pond and study grid: (a) concentration contours of 13/Cs dis-
tribution calculated from 1973 aerial survey data; (b) 13?Cs con_
centration contours generated from in situ GeLi Measurements; (c) ob-
served distribution of surface water in S-17 Pond based on aerial
photograph; (d) vegetation concentration contours based on field
collection and laboratory analysis of Siberian Wheatgrass.
-------
Q_
LU
Q
.*»
tn
ro
pCi/g
-------
453
RHO-SA-99
considerable wind erosion took place over the period before revegetation. The
spike noted at the 95-115 cm depth was due to a single core located within an
area of the pond which received a greater proportion of soil backfill (106 cm).
It should be noted that surface concentration values (Fig. 4a, b) were
calculated from exposure data utilizing in situ measurements. They do not
represent actual concentrations for surface soil. The inventories provided in
Table 2 and Figure 5 present data on the calculated amounts of 137cs con-
tained in the top 5 cm of soil based on soil core samples. The highest
exposure rate observed on the pond site was .5 mR/hr at 1 meter.
Vegetation
The principle thrust of the vegetation sampling program was directed
toward ascertaining the potential for radionuclide uptake in Siberian
Wheatgrass, since the plant appeared to be an extremely successful perennial
for revegetation of disturbed areas. The average background value due to
fallout in vegetation is 2 t 1.2 pCi/gm in the Hanford environs while
the mean value determined for Siberian Wheatgrass was 4.7 ± .7 pCi/gm dry
weight. Standing crop estimates calculated in early May 1978 provided a total
biomass estimate for Siberian Wheatgrass of 0.3 kg/m2 dry weight. This
resulted in a total !37Cs inventory for the Wheatgrass on the study site of
110 pCi (Table 2). Visual inspection of root depth indicated that the major-
ity of the root mass was associated with the 5-30 cm soil depth, with the
longest roots being less than 60 cm long. Apparently, over a period of time
the combination of wind erosion and shallow soil backfill (45-60 cm) resulted
in optimum conditions for uptake of radionuclides by the Siberian Wheatgrass.
However, only .002 percent of the total calculated site inventory was incor-
porated by the Siberian Wheatgrass. The concentration ratio observed from
this study appeared to be in the order of the 10-1 to 10~2 range. This is
comparable with other vegetative types, such as cheatgrass and tumbleweeds in
the Hanford area (C172). With proper range management techniques, including
at least one meter of backfill and erosion control until seedlings are estab-
lished, Siberian Wheatgrass would be a suitable perennial for revegetation of
disturbed sites in the Hanford area.
Small mammals
A total of 27 Great Basin pocket mice (Perognathus parvus) were analyzed
for total body burden of 137cs. The highest value for all the samples was
13.6 pCi/gm and the sample mean was 1.9 + .8 pCi/gm wet weight (Table 2).
Control samples, affected by fallout levels alone, indicate mean values for
137Cs in small mammals of approximately 2.2 + 1.6 pCi/gm wet weight (Ga77).
Therefore, although the animals on the S-17 Pond are living in a radiation
zone they do not appear to be incorporating much 137r,s into their systems.
A density determination for the pond site indicated an average resident
population of 20 mice/ha. Density estimates for the surrounding sagebrush
(Artmesia tridentata) - Sandburg bluegrass (Poa sandburgii) - cheatgrass
(ttromus tectorum) community are expectedly higher at 25 to 28 mice/ha (He76).
Therefore, the total estimated 137cs inventory for the small mammal
compartment of the study grid is .004 ± .001 uCi (Table 2).
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454
RHO-SA-99
The pocket mice could pick up 137cs by burrowing through contaminated
soil and ingestion of small amounts of soil and attachment of soil particles
to their fur and skin. They could also have incorporated the i-J/Cs into
their bodies by ingestion of the seeds and green tissues of the contaminated
plants. At this time, it is not know whether these animals utilize Siberian
Wheatgrass in their diets. If the body burden is due to ingestion of 13/Cs,
the biological half-life for 137cs in small mammal species has been esti-
mated to range from 2 to 8 days (St70; Ba75).
The behavior of pocket mice on the pond may be more important in terms of
radionuclide availability than the small amounts contained within their bodies.
The concentration values for soil samples collected from the ant mounds were
within background values for surface soil with a mean of 1.1 + .4 pCi/gm; the
highest value being 5.35 pCi/gm. However, the mean value for the mouse burrow
samples was 268.0 + 175 pCi/gm, although this value was principally due to
three samples with high levels (range 682-3490 pCi/gm).
CONCLUSIONS
Construction and other activities associated with radioactive waste site
preparations usually result in the partial or complete denuding of the native
vegetation on these sites. The restablishment of plant cover on these areas
is a difficult task in an environment which receives only 6-8 inches of precip-
itation annually. Natural revegetation in the sagebrush steppe ecosystem is a
slow and arduous process (Da70). Wind erosion of soils becomes a major
problem on disturbed areas until the soil is stabilized by plant growth. As
in the case of the S-17 Pond site, undesirable deep-rooted plants such as
tumbleweeds become established very quickly. The use of Siberian Wheatgrass
in a comprehensive revegetation program offers a solution to these types of
waste management problems.
Visual inspection of Figure 4 (a and b) shows a high degree of correla-
tion between the two in situ measurement techniques. These techniques also
compare favorably with the surface distribution of 137cs -jn Siberian Wheat-
grass based on laboratory analyses (Fig. 4d). The primary reason for this
high degree of correlation is due to the homogenous vertical distribution of
the contaminated sediment layer in the soil column throughout the study grid.
Upon completion of additional radionuclide analyses and further utilization of
the in situ GeLi technique at other waste sites with less homogeneity, appro-
priate double-sampling techniques will be developed. This will provide quick
and relatively inexpensive techniques for estimating inventories at future
waste sites.
In relation to the radionuclide inventory presented in Table 2, biotic
and abiotic compartmental estimates generated by this study are indicative of
the general patterns observed in previous radioecology studies on waste
S't^T^V^in'Lr'lidf'0^ iC172^' In general> 9reate* than 9? percent of
column radlonucllde inventory is associated with the soil or sediment
-------
455
RHO-SA-99
Acknowledgements
We greatly appreciate the invaluable assistance in the field and
laboratory of John Bates, Ron Chandler, and Rick Rollins.
REFERENCES
Ba75 Baker, C. E. and Dunaway, P. B., 1975, "Elimination of 137Cs and 59Fe
and its Relationship to Metabolic Rates of Wild Small Rodents, J. Exp.
Zoology 192; 223 *-
Br77 Bruns, L. E., 1977, Environmental Instrumentation For In Situ Radionuclide
Assay, Rockwell Hanford Operations, Richland, WA, Report No. RHO-SA-1.
C172 Cline, J. F. and Rickard, W. H., 1972, "Radioactive Strontium and Cesium
in Cultivated and Abandoned Field Plots", Health Physics 23; 317.
C177 Cline, J. F. and Uresk, V. A., 1977, Revegetation of a Waste Disposal
Site in the 200 Areas of the Hanford Reservation, Pacific Northwest
Laboratories, Richland, WA, Report No. PNL-2454.
DA70 Daubenmire, R., 1970, Steppe Vegetation of Washington, Wash. Agri. Exp.
Station, Pullman, WA, Tech. Bull. No. 62.
6a77 Gano, K. A., 1977, Small Mammals Inhabiting the Environs of a Low-Level
Radioactive Waste Pond (In preparation).
Ha53 Hanson, W. C., Browning, R. L. and Braymen, W. H., 1953 Waterfowl
Contamination Observed at Redox Swamp During December, 1952, Hanford
Works, Richland, WA, Report No. HW-27104.
He76 Hedlund, J. D. and Rogers, L. E., 1976, Characterization of Small Mammal
Populations Inhabiting the B-C Cribs Environs, Battelle Northwest Labora-
tories, Richland, WA, Report No. BNW-2181.
Jo65 Jolley, G. M., 1965, "Explicit Estimates From Capture-Recapture Data With
Both Death and Immigration Stochastic Mocel", Biometrica 52; 225.
Kr70 Kritzman, E. B., 1970, "Ecological Relationships of Peromyscus maniculatus
and Perognathus Parvus in Eastern Washington", J. Mammalogy 55; 172.
-------
456
RHO-SA-99
St54 Stickel, L. F., 1954, "A Comparison of Certain Methods of Measuring
Ranges of Small Mammals", J. Mammalogy 35; 1.
St70 Stather, J. N., 1970, "An Analysis of the Whole-Body Retention of 137cs
in Rats of Various Ages", Health Physics 18; 43
Ti75 Tipton, W. J., 1975, An Aerial Radiological Survey of the USERDA Hanford
Reservation - 1973-1974, E6&G, Las Vegas, NV, Report No. 1183-1661.
-------
457
RETENTION OF LOW-LEVEL RADIOACTIVE
WASTE MATERIAL BY SOIL*
by
E. H. Essington, E. B. Fowler, and W. L. Polzer
University of California
Los Alamos Scientific Laboratory
Los Alamos, NM
Abstract
Low-level radioactive wastes produced by users of radionuclides are
generally disposed by shallow land burial. Reliance for containment is
placed on characteristics of shallow geologic formations or soils; thus,
effective waste management requires a knowledge of radioactive waste/soil
interactions.
Because of the wide variations in soil and waste characteristics, the
degree of radionuclide retention would be expected to vary; knowledge of
that variation may be of value in predicting radionuclide mobility. This
report discusses results of investigations of radioactive waste/soil inter-
actions as they relate to radionuclide retention and its variability among
soils and radionuclides.
In soil column leaching studies, radioactive waste solutions were
applied to four different soil types; 2ItlAm, 88Y, and 172Hf were retained
in the top four cm of soil with better than 90% retained by a protective
surface sand layer. Less then 50% of the B5Sr, 137Cs, and 83Rb was re-
tained by the surface sand. No 88Y, 172Hf, 85Sr, 137Cs, or 83Rb was de-
tected by gamma counting in the leachate solutions, however, using a more
sensitive analytical technique small amounts of 238Pu, 239»2"°Pu and 21flAm
were found in leachates from all soils. It appears that release of this
small fraction of mobile radionuclide may have a significant long-term
impact on the environment. It also appears that reliance for attenuation
of some radionuclides can not be placed solely on characteristics of the
soil matrix.
Introduction
Shallow land burial has been used for disposal of low-level radioactive
waste material for more than three decades. Recently, investigators have
observed small amounts of radioactivity in the environs of several of those
sites, signaling the need for a more thorough evaluation of the potential and
mechanisms of long-term redistriubution of buried radionuclides both on site
and off site.
*Work performed under the auspices of the Nuclear Regulatory Commission.
-------
458
Field experiments to describe the physical and chemical interactions
affecting radionuclide migration are very expensive and in some cases cannot
yield the information required for prediction. Laboratory tests are, then,
generally conducted to derive the necessary data for use in flow or migration
prediction models.
Over the past three decades many studies of radionuclide migration have
been conducted in soil and geologic materials to assess the degree of sorption
or retention the materials have for the radionuclides. However, until recently,
most studies of that type relied on single prepared radioisotope solutions in
carefully controlled systems and did not represent the very complex conditions
existing in radioactive waste interactions with soil .or geologic media.
A study was designed to provide information on radionuclide retention
and migration in soils as a function of soil type and radionuclide species.
Preliminary results show the nature of low-level radioactive waste material
interactions with several widely varying soil types. Based on information
gained in this study, future experiments will be designed to more precisely
evaluate factors affecting retention or migration and to develop a measure
of the confidence one might expect in retention values. This information
will be incorporated into radionuclide migration and dose assessment models
used in low-level waste management and licensing.
This study is one phase of a larger study to derive some confidence in
predictions of radionuclide retention by soils and is sponsored by Nuclear
Regulatory Commission, Division of Safeguards, Fuel Cycle and Environmental
Research. Figure 1 depicts the rationale behind the study. A great deal of
variability in the degree of radionuclide retention does exist because of the
wide variety of radioactive sources and species in each source and the large
variability in the evironmental materials available for interaction. The
variety of waste sources is indicated in the first column of Fig. 1; examples
of characteristics of the waste/soil solution believed to influence retention
are listed in the second column. The degree of radionuclide retention as
determined by soil types for each waste/soil characteristic will define an
envelope from which some confidence relative to retention can be projected.
Correlation of the retention with one or a combination of more specific
parameters (characteristics), such as, soil texture, pH, clay type, etc.,
may yield a smaller degree of predicted retention variability. In other
words, the more that is known about the system the better one can predict
both short- and long-term radionuclide retention .
Materials and Methods
Four soils were reacted with a low-level radioactive industrial waste
solution in order to evaluate the radionuclide retention capacity of the
soils. Several major physical and chemical soil characteristics likely to
be related to radionuclide retention are listed in Tables 1 and 2. The Ap
horizon of Fuquay loamy sand represents the top 15 cm of soil collected from
the Barnwell area of South Carolina. The Ap horizon of Fayette silt loam repre-
sents the top 18 cm of soil collected from Sheffield, Illinois. The B horizon
of Car jo loam and the C horizon of Puye sandy loam are from Los Alamos, New
Mexico and were collected from depths of 5 to 15 cm and 25 to 51 cm respec-
tively. The four soils range in pH (saturated paste) from 5.5 to 7.1 and have
similar cation exchange capacities (CEC) ranging from 15 to 23 meq/100 g of
soil. Although the clay content of the Fuquay is very low, enough clay is
-------
459
present to Identify the major clay types as vermiculite and kaolinite. One
of the characteristics believed to be very important in mediating the chelate
complexation of radionuclides is the available (extractable) iron. Large
amounts of extractable iron were present in Fayette and Fuquay soils as meas-
ured by the synthetic chelating agent diethylenetriaminepentaacetic acid (DTPA).
Leaching columns, 3.8 cm in diameter were prepared in triplicate using
300 g (air dry) of each soil type. A 1-cm plug of washed silica sand (20-50
mesh) was placed at the top and bottom of each column. After packing, each
column was saturated with distilled water from the bottom to minify the trap-
ping of air and then allowed to drain to field capacity prior to commencement
of leaching.
Soil column setup and leaching were conducted according to the procedure
of Essington and Nishita (Es66). The waste solution was added to the top of
the soil column in 10-cm irrigation increments (115 ml each) until the infil-
tration rate of the waste solution was seriously impaired. At that point 2.5 1
of waste solution had been passed through the soil column. Each 10-cm leach
was collected and a selected few were analyzed for major gamma emitting radio-
nuclides and 238Pu, 239.21t0Pu, and 21|1Am. The gamma emitting radionuclides
were measured with a germanium (lithium drifted) detector and pulse height
analyzer. The other radionuclides were measured using radiochemical separations
and alpha pulse analysis. Electrical conductivity, Eh, pH, and alkalinity were
also determined on each leachate. Upon completion of the leaching experiment
each column was fractioned into approximately 2-cm increments. Each increment
was analyzed for major gamma emitting radionuclides and 238Pu, 239'21<0Pu, and
mAm.
The leaching solution was a two-week composite of a liquid radioactive waste
material accepted by a waste treatment facility. Table 3 lists several of the
many measured characteristics of a similar waste material as collected and after
centrifuging to remove particulates greater than about 0.05 \sm in diameter.
Material remaining in the supernatant solution after centrifugation was design-
ated as "soluble" although the supernatant probably contained small particulates,
polymers, and ions. Centrifugation removed substantial quantities of phosphate,
aluminum, iron and organic matter, suggesting that a predominance of these
materials was associated with particulates larger than 0.05 ym. Those materials
are also suspected of being involved in interactions with some of the radionuclides
in the waste.
Some of the more prevalent radionuclides present in the waste are listed
in Table 4. The ratio of radionuclide concentration before and after centri-
fuging are given to indicate their degree of partition and to suggest a degree
of complexity of the waste solution. Three groupings are suggested in Table 4
based on the amount of radionuclide contained in the soluble fraction relative
to the total. The group including 83Rb, involves the largest amount of soluble
radionuclide, whereas the group including 88Y involves the largest amount of
insoluble radionuclide. Note particularly that the partition of the two iso-
topes of plutonium is significantly different in the two fractions (standard
deviation/ mean of 4 replicates = C.V. - 0.1).
Results and Discussion
The distributions of various waste radionuclides in soils and leachates
indicate the complexity of the chemical and physical interactions between the
-------
460
waste and the soil matrix. That complexity is indicated in Fig. 2 by the
presence of at least two forms of 21tlAm shown by the accumulation of a large
portion in the sand layer and by the small but constant amount in the deeper
soil fractions. Figure 2 also shows the 241Am distribution in the three re-
licate soil columns of Carjo soil as well as the variability of Am in the
various soil fractions. The vertical bars represent the plus or minus one-
sigma counting error for each fraction. Note that in the shallow portion of
the soil columns the counting error is small and data are reproducible. In
the deeper portions of the column the counting error is large, however, the
21tlAm levels for the three replicates appears to fall well within the counting
error. A non-linear least-squares procedure was used to fit the "* Am values
in the soil column fractions to an exponential equation of the type shown on
the figure. Subsequent figures will use those fitted curves where possible
for clarity of presentation; however, several of the curves were hand fitted.
Figure 3 shows the distribution of 21tlAm in the four soils studied. The
distribution of 241Am in Puye, Carjo, and Fuquay, did not appear to be differ-
ent, whereas 241Am in Fayette appeared to be distributed somewhat differently,
particularly in the shallower parts of the column. Note that in all cases the
sand layer retained about 90% of the 21tlAm in the column. Additionally, the
level of 2l|1Am in the deeper column fractions approached a constant. That
level of 241Am was calculated to be nearly the same concentration as that
found in the final leachates and therefore, can be attributed to the soil
column water remaining at field capacity.
Hafnium-172, 21flAm, and 88Y appear to be similarly distributed in all
soil types as illustrated by the data for the Puye soil in Fig. 4. As with
21tlAm, the sand layer accumulated better than 90% of the 172Hf and 88Y.
Hafnium-172, 88Y, and 241Am were not detected in the leachate solutions with
the gamma counting technique initially employed. However, the lower levels
of 21tlAm were easily detected by radiochemical separations and alpha pulse
counting.
The distribution of 137Cs in the four soils is shown in Fig. 5. In this
case only about 40% of the 137Cs was removed by the sand layer; however, 137Cs
was not detected in the leachate solutions. The differential distribution of
137Cs in the 5- to 12-cm region of the Puye soil may be related to soil char-
acteristics. In the shallower depths the 137Cs had distributed in a manner
similar to that for 2tflAm, 172Hf, and 88Y as indicated by the slopes of the
distributions in the 0- to 5-cm region.
Strontium-85 distribution in the four soils is shown in Fig. 6. Only
about 40% of the 85Sr was removed by the sand layer and no 85Sr was detected
in the leachate solution. Strontium-85 appears to distribute in the soils
differently from the radionuclides discussed earlier. This conclusion is
based on the observation that the slopes of the distribution curves in the
upper portion of the columns are not as steep as for the other radionuclides,
indicating that greater amounts of 85Sr have moved deeper into the soil col-
umns. Similar distribution patterns were reported by Essington et ai. (Es66)
and Nishita and Essington (N167) in columns of agricultural soils containing
radioactive strontium and leached with irrigation or distilled water. Movement
or retention of the strontium was believed to be governed by the sorption
mechanisms of the soils and the competition of calcium ions for sorption sites.
Since no Sr was found in the leachate solutions, that portion of the 85Sr
not removed from the waste leaching solution by filtration appears to be
attenuated by sorption mechanisms.
-------
461
Figure 7 shows the distribution of 83Rb in the four soils. The distri-
bution of 83Rb in Puye, Carjo and Fayette was similar to that of 241Am; however,
only 10% of the Rb was retained by the sand layer and no 8 ^b was found in
the leachates.
Rubidium-83 distribution in Fuquay was quite different from that in the
other soils in that concentrations were constant with depth to about 7 cm and
then steadily decreased to a non-detectable level. It appears that the 83Rb was
retained by Fuquay to a lesser degree than was 85Sr (Fig. 6). However, based
on the similarity of the 83Rb distribution to that of 85Sr similar sorption
mechanisms may be involved. No effort was made to determine stable rubidium
in the waste solution; stable rubidium could have a significant effect on 83Rb
retention by the soil matrix.
Plutonium-238 and 239»21*0pu were measured in the soil column leachates
only. Figure 8 shows the amount of 238Pu added with each increment of waste
solution and the amount of 238Pu found in each leachate; data are presented as
averages of the leachates from the three soil columns. The appearance of
238Pu in the first leach indicates a rapid breakthrough. The attainment of a
relatively constant level after the second waste addition indicates that portion
of 238Pu in the waste solution that was not filtered or sorbed by the soil. A
comparison of the 238Pu in the leachate with that added to the soil columns
shows that 83 to 97% of the 238Pu was retained by the soil columns. Those re-
sults indicate that there was a small but highly mobile fraction of 238Pu in
the waste solution. Although 239»21t0Pu was found in the same leachate,
239,2ifOpu reacte(j somewhat differently from 238Pu. This is shown in the ratios
of Z39'21|0Pu to 238Pu in the waste solution, centrifuged waste, and the soil
leachates (Fig. 9). In general, the ratios for soil leachates fell between
those of the source waste and centrifuged waste solutions. In the early stages
of leaching there were rapid changes in the leachate plutonium ratios, but as
leaching progressed the ratios tended to converge at about 0.1. Those data
show that isotopes of the same element can react differently in the soils.
Plutonium in the waste solution originates from a number of widely different
operations. Those operations dealing primarily with 238Pu tend to dispose
of a higher percentage of soluble 238Pu than soluble 239'2"*°Pu. Those oper-
ations dealing primarily with 239'2l|0Pu and 241Am (Table 4) tend to dispose
of particulate material or material prone to be associated with particulates.
Thus the nature of the source material may dominate the retention pattern in
soils and the differences in plutonium isdtopic distributions are not necessarily
due to basic differences in behavior of the isotopes.
The waste/soil system is complex as indicated by the data presented. That
complexity may be a result of a number of interactive mechanisms. The major
mechanisms, in addition to filtration and ion exchange, include: precipitation
and dissolution of calcium carbonate; complexation; and microbiological growth.
Those mechanisms as well as evaluation of changes in the waste source upon aging
and results of batch sorption studies with the same soils were reported by
Fowler et al. (FO78).
Waste radionuclides interacting with calcium carbonate, the primary inor-
ganic mineral in the waste, should behave in a manner similar to that of the
calcium carbonate. For example, changes in the carbonate system which result
in either a precipitation or dissolution of solid carbonates should result in
-------
462
a change in the radionuclide concentration in the soil solution. Such changes
have been noted to occur in the waste solution upon aging, and undoubtedly occur
upon contact with the soil. The changes could continue to occur during the pro-
gress of leaching as the leaching solution aerates or changes upon encountering
different materials in the soil column. The extent of the carbonate effect on
radionuclide retention was not specifically investigated in the experiment.
However, the presence of CaC03 in the waste solutions was confirmed by X-ray
diffraction. Significant decreases in the concentration of soluble plutonium
and americium were observed upon formation of CaCOs in the waste material,
whereas, changes in the soluble cesium and uranium concentrations did not occur.
Those observations may account, in part, for the differences in the amounts of
241 Am and 137Cs retained by the sand layer on top of the soil column.
The small but significant amount of plutonium and americium found in the
leachate solutions may be due to carbonate complexing or the presence of che-
lated species. The charge of some soluble actinides in the waste was shown to
be predominatly negative whereas the charge of the 137Cs was predominantly
positive (Po79). This was accomplished by passing the soluble waste fraction
through cation and anion exchange resins and observing the amount of radio-
nuclide not held by the resin (Fowler et al. Fo78). The work of Alberts et
al. (A177) on the identification of the charge of plutonium in samples of
Lake Michigan water indicated that the charge of the plutonium species was
negative. They attributed the negative nature of the charge to the formation
of a stable carbonate complex in the CaC03 super-saturated water having a pH
of approximately 8. Ames et al. (Am76) present stability diagrams for pluto-
nium and uranium which indicate that at pH 8.0 carbonate complexes could account
for the predominantly negative charge of plutonium and uranium species.
Although the waste solution was not analyzed for the presence of chelating
agents it is strongly suspected that significant quantities were present. The
sources of the waste solution include decontamination fluids containing deter-
gents and chelating agents and chemical laboratory wastes where chelating agents
and other organic complexers are used and discarded. The effect of the chelating
agents would be to form a very stable complex with the radioactive rare earth
or actinide ions present and at pH 8 the resulting complexes are likely to be
neutral or negatively charged. Those species would migrate through the soil
rapidly with the possibility of exchange with metal ions in the soil solution.
The degree of exchange of the radioactive ion with a metal ion would be
dependent upon their concentrations in the soil solution and upon their relative
stabilities with the chelate ligand.
During the conduct of the column leaching experiment an algal growth was
allowed to be established in the columns. Normally the leaching experiment
would be conducted with care not to allow unnatural microbiological activity;
however, waste solutions allowed to impact the open environment would surely be
subject to microbiological activity. The effect of the algae on radionuclide
retention in the soil columns is a valid parameter. A separate experiment was
conducted to test the effect of algae on their capacity to assimilate or tieup
waste radionuclides. An algal bloom was allowed in the waste solution in
which the pH was adjusted to 6.0; the algae were removed, washed, and analyzed.
The effect of algae on retention of waste radionuclides depended on the radio-
nuclide; a large percent of the actinides (86-95%), but only 30% of the cesium
was associated with the algae. Those results are consistent with the large
-------
463
accumulation of 21tlAm, 172Hf, and 88Y and the somewhat smaller degree of
accumulation of 3 Cs, 85Sr, and 83Rb in the sand layer on the soil column.
The information presented thus far is indicative of the complexity of
the waste/soil systems. Reactions of liquid waste materials from different
low level waste streams, shallow waste burial pits or trenches, or the high
level waste processing streams accidentally released to the soil environment
may be even more complex.
Relating the retention data to migration rates and distance is generally
accomplished by use of the distribution coefficient (K
-------
464
number of different forms of a given radionuclide were shown to exist e.g.,
filterable, sorbable, and highly mobile. Each form may consist of a number of
species; each specie could be described by a separate K
-------
465
References
A177 Alberts J.J., Wahlgren M.A., Nelson D.M. and Jehn P.J., 1977, "Submicron
Particle Size and Charge Characteristics of 239.21*0pu in Natural Waters,"
Environ. Sci. Tech. 11: 7, 673.
Am76 Ames L.L., Rai D. and Serne R.J., 1976, "A Review of Actinide-Sediment
Reactions with an Annotated Bibliography," BNWL-1983, Battelle Pacific
Northwest Laboratories, Richland, Washington.
Es66 Essington E.H. and Nishita H., 1966, "Effect of Chelates on the Movement
of Fission Products Through Soil Columns," Plant and Soil XXIV: 1, 1.
Fo78 Fowler E.B., Essington E.H., Polzer W.L., 1978, "Differential Attenuation
of Waste Radionuclides in Soil," LA-UR-78-1670, Los Alamos Scientific
Laboratory, Los Alamos, NM, Presented: American Nuclear Society Winter
Meeting, November 12-17, 1978, Washington, DC.
Ma47 Mayer S.W. and Tompkins E.R., 1947, "Ion Exchange as a Separations Method:
A Theoretical Analysis of the Column Separation Process," J. Amer. Chem.
Soc. 69, 2866.
Ni67 Nishita H. and Essington E.H., 1967, "Effect of Chelating Agents on the
Movement of Fission Products in Soil," Soil Science 103: 3, 168.
Po79 Polzer W.L., Fowler E.B., and Essington E.H., 1979, "Characteristics of
Wastes and Soils Which Affect Transport of Radionuclides through the
Soil and their Relationship to Waste Management," Annual Progress Report
for October 1, 1977, to September 30, 1978, Los Alamos Scientific
Laboratory, Los Alamos, NM, In preparation.
Tr78 Travis C.C., 1978, "Mathematical Description of Adsorption and Transport
of Reactive Solutes in Soil: A Review of Selected Literature," ORNL-5403,
Oak Ridge National Laboratories, Oak Ridge, TN.
-------
46i
pH
Eh (mV)
P04 (*)
Co (*)
Al (*)
Fe (*)
COD (*)
TOTAL
7.8
+410
29
75
2
12
170
SOLUBLE
8.1
+ 400
2
51
0.2
0.4
60
(*)=(ppm)
TABLE 1. SELECTED MAJOR PHYSICAL CH AR ACTE Ft ISTICS O F FOUR
SOILS USED IN LEACHING STUDIES.
Total Ca
Ammonium
Acetate
Extractable
Ca
Mg
Fe
Mn
Al
DPTA Extr. Fe
Puye
(ppm)
10 000
1800
170
4.8
24
6
6
Car jo
(ppm)
10 000
1000
200
5.4
54
9
7
Fayette
(ppm)
10 000
1400
330
36
24
24
28
Fuquay
(ppm)
1000
230
22
1
16
4
50
FOUR SOILS
-------
467
Pw at F.C.1
CEC
pH
SAND %
SILT %
CLAY %
CLAY TYPE2
•i— . . _
Puye
43
23
7.1
45
48
7.5
KM
Car jo
40
18
6.4
53
37
10
KM
Fayette
40
15
6.6
19
73
8.5
KMV
Fuquay
30
16
5.5
i
86
14
~ 0
VK
Vercent Moisture at Field Capacity
2V=Vermiculite; K = Kaolinite; M = Montmorillonite
TABLE 3. SELECTED PHYSICAL AND CHEMICAL PROPERTIES OF SOURCE WASTE
(TOTAL) AND SOLUTION PHASE (SOLUBLE) AFTER CENTR IFUG ATION.
83Rb
85Sr
137Cs
238U
88Zr
60Co
88y
172Hf
241Am
238pu
239,240pu
Tl/2
83d
64d
30y
4.51X109y
85d
5.26y
108d
S/T1
.67
.54
.60
.46
.15
. 1 1
.018
5y .001
458y
86. 4y
24390y,
6580y
.012
.03 1
.01 1
L
T = Source waste, S = Soluble fraction
TABLE 4. MAJOR RADIONUCLIDES IDENTIFIED IN SOURCE "ASTE THE
RADIOACTIVE DECAY HALF-LIFE, T%. AND THE PARTITION UPON CENTRIFUG ATION.
-------
WASTE
SOURCE
SPILLS
WASTE /
BURIAL';
EFFL. '•
WASTE/
SOIL
CHAR.
OM
CEC
SOIL
TYPE
CLAY
LOAM
SAND
ENVELOPE OF
CONFIDENCE
468
o
sP
2
w
K
FIG. 1; FLOW DIAGRAM RELATING RADIOACTIVEWASTE
SOURCE AND SOIL CHARACTERISTICS TO RADIONUCLIDE
RETENTION AND THE DEVELOPMENT OF STATISTICALLY
BASED CORRELATIONS'
1U =
:
10'S
10"=
o :
< io_;
:
io"=
m"5
Iw
a
S
SAND
5 C
a Replicate-1
o Replicate-2
A Replicate-3
V
>^ 1 IT# I 5
^» T I f Tiff T
I II1
Y=Ae-BX+C
SOIL
i i i I i
) 5 10 15 20 25
DEPTH (cm)
FIG. 2. DISTRIBUTION OF241 Am IN COLUMNS OF CARJOSOIL
AFTER LEACHING WITH RADIOACTIVE WASTE. A/AQ REPRESENTS
FRACTION OF TOTAL COLUMN 241 Am FOUND IN EACH INCREMENT.
10%
10 -
10"
PUYE-C
CARJO-B
FAYETTE-Ap
FUQUAY-Ap
SAND; SOIL
-5
r~
20
25
0 5 10 15
DEPTH (cm)
FIG. 3. DISTRIBUTION OF241 Am IN FOUR SOILS.
-------
10° 3
< 10 "i
ICT-s
10"
0
SAND
D 24lAm
0 '»Hf
A »«Y
SOIL
1 i i i
-5
0 5 10 15 20 25
DEPTH (cm)
FIG. 4. COMPARISON OF THE DISTRIBUTION OF
241A
m,
241
Am.
172
Hf AND
IN
PUYESOIL,
io"=
<
<
io-=
10"
PUYE-C
CARJO-B
FAYETTE-Ap
FUQUAY-Ap
SANDi SOIL
1 1 1 1 1 1
-50 5 10 15 20 25
DEPTH (cm)
FIG. 5. DISTRIBUTION OF 137Cs IN FOUR SOILS.
469
10%
10" =
10
o
10"
SAND SOIL
PUYE-C
CARJO-B
FAYETTE-Ap
FUQUAY-Ap
1 —I 1 1 1 I
-50 5 10 15 20 25
DEPTH (cm)
FIG. & DISTRIBUTION OF ^Sr IN FOUR SOILS.
-------
10
10"=
SAND
10"
-5
PUYE-C
CARJO-B
FAYETTE-A,
FUOUAY-Ap
SOIL
10
0 5 10 15 20 25
DEPTH (cm)
470
FIG. 7 DISTRIBUTION OF uRb IN FOUR SOILS.
SOURCE WASTE
0 10 •=
:
\
\ioS
5- :
^ lo-1
> H
< 10""=
in-^
8— — — __^
}T
L " PUYE-C
* FAYETTE-Ap
0 5 10 15 20 25
LEACHATE NO.
FIG. t. PLUTONIUM-231 IN SOURCE WASTE AND IN LEACHATES FROM FOUR SOILS.
a,
0.25
0.20 -
0.15-
0.10 -
0.05-
c PUYE-C
o CARJO-B
a FUQUAY-Ap
+ FAYETTE-AP
SOURCE WASTE
CENTRIFUCED
WASTE
0.00 -] 1 1 1 1 1
0 5 10 15 20 25
LEACHATE NO.
FIG. 1 RATIO OF 231- 240Pu TO M*P» IN SOURCE WASTE. IN CENTRIFUGEO
WASTE . AND IN LEACHATES FROM FOUR SOIL!
-------
471
THE USE OF HANFORD WASTE WATER PONDS BY WATERFOWL
by
K. R. Price and R. E. Fitzner
Battelle, Pacific Northwest Laboratory
Richland, Washington
Abstract
Census and environmental surveillance information on waterfowl that use the
Hanford Site 200 Area waste water ponds are described and evaluated. Physical
features of the ponds are discussed in relation to their use and suitability for
waterfowl. Seasonal distributions observed for the years 1971 through 1974 in-
dicate that the highest use by waterfowl occurs during the spring and fall migra-
tory periods. Base population estimates are 300 to 400 resident waterfowl with
a few tens of pairs nesting during the summer. Environmental surveillance data
on 137Cs in muscle tissue are presented for the years 1971 through 1977. Com-
parisons are made between Columbia River and waste water pond waterfowl, between
waterfowl groups, and among ponds. Waterfowl collected from ponds frequently
have easily detected levels of 137Cs in muscle tissue. However, those water-
fowl collected from the Columbia River seldom show a 137Cs level above that
expected from worldwide fallout. Waterfowl collected from the pond with the
smallest 137Cs inventory and the poorest waterfowl habitat contained the lowest
levels of 137Cs in muscle tissue.
Introduction
The purposes of this study were to:
0 identify the waterfowl species that use the Hanford Site
waste water ponds
0 establish the abundance of waterfowl during all seasons of
the year
0 evaluate features of the ponds (vegetation, size, amount of
human activity) for their attractiveness to waterfowl
0 compare concentrations and variations of 137Cs in muscle
tissue between dabbling ducks and diving ducks and among
ducks using different ponds and the Columbia River.
Study Areas
The study areas consisted of three shallow ponds located in the 200 Area of
the Hanford Site. The waste water ponds are commonly referred to as U Pond,
Gable Mountain Pond, and B Pond. A diagram of vegetation zonation around the
ponds is shown in Figure 1.
U Pond is the smallest of the ponds studied (6 ha), and has received coolant
water and other waste water since July 1944. In July 1972, a chemical herbicide
was applied and immediately defoliated approximately 80% of the cottonwoods
(PoEulus spp.) and willows (Salix spp.). Other species of vegetation, both
terrestrial and aquatic, were less affected. Some of the trees, mostly cotton-
woods, showed new growth by the next spring, and now about one-half of the
-------
472
original live canopy has returned. The pond is characteristic of most ponds of
the region in later stages of ecologic development. In addition to the usual
patches of cattails (Typha latifolia) and reeds (Scirpus spp.), large willow trees
and cottonwoods (10 m tall) grow along the shoreline. A dense cover of weedy
vegetation ranging from 0.5 to 1.0 m tall and composed mostly of Russian knapweed
(Centurea repens) and cudweed (Gnaphalium margaritacea) grows in the moist soil
not occupied by trees. Human activity and disturbance from nearby heavy con-
struction has been extensive.
Gable Mountain Pond is the largest of the ponds studied (29 ha) and has
been used since December 1957. The water source is primarily coolant water that
is discharged continuously. The pond has only a partial dike and has not been
treated with herbicides to control riparian vegetation. Human disturbance is
limited. This pond, in time, is expected to show the same advanced stages of
vegetation zonation as is currently found at U Pond. Cattails and patches of
reeds are extensive, and peachleaf willow trees (Salix amygdaloides) are widely
scattered but smaller and younger than trees at U Pond.
B Pond, the second largest pond (19 ha), has been active since April 1945
and receives water from a coolant source and a chemical sewer ditch. Shore-
line dike construction and herbicide applications in 1971 and 1972 destroyed
most riparian and emergent aquatic plants. Reestablishment of shoreline
vegetation has been slow. Some isolated reed patches occur in the pond shallows,
and the terrestrial shoreline is ringed by a narrow zone of barnyard grass
(Echinochloa crusgallii). Human activity at the pond is limited; however, the
close proximity to the 200-E Area (about 1 km) adds a noise factor that is
absent at Gable Mountain Pond.
Methods Employed
Weekly observations at the ponds were conducted from September 1971 to
March 1974. Generally, waterfowl were observed during the morning hours while
driving slowly around each pond, stopping at prescribed points, and scanning
open water and border vegetation with binoculars. This technique, although not
a complete census, revealed trends in abundance and species diversity. Also,
human disturbance near each pond and meteorological conditions were noted in
order to assess their possible effects on observed bird counts. Observation
records for each census date were compiled for computer storage and retrieval.
Waterfowl are routinely collected as part of the Hanford environmental
surveillance program. A 500-g sample of breast muscle is taken from each bird
and a gamma spectroscopic analysis is performed on the fresh tissue. Various
comparisons of the levels of l37Cs in duck muscle tissue were made with the
use of computer generated log-normal probability plots (Wa76).
Results and Discussion
Waterfowl belong to the order Anseriformes and are represented by the ducks,
geese and swans. Ducks are often subdivided into dabbling ducks, diving ducks
and mergansers. The dabbling ducks dabble and tip as they feed mostly on vege-
tation in shallow water. Diving ducks dive underwater when they feed on vege-
tation and varied forms of animal life. Mergansers dive for their food like .
the diving ducks but feed primarily on invertebrates and fish. For purposes
of this discussion, the American Coot, order Cruiformes. has been included as
-------
473
a waterfowl species. Coots may feed like dabbling ducks, but also will dive for
submerged aquatic plants and invertebrates. All of these waterfowl are legally
hunted, but geese and dabbling ducks are preferred game.
Seasonal Distribution and Abundance of Waterfowl
During the 33 months of weekly waterfowl censuses, a total of 20 duck species,
the Canada goose, the Whistling Swan and American Coot were observed to use the
Hanford waste ponds (Table 1).
The American Coot was the most abundant species observed to use the waste
ponds; 19,141 of these birds were counted during the study. The Mallard was the
most abundant dabbling duck (5,858 total observed), while the Ringnecked Duck was
the most common diving duck (4,255 total observed). The total numbers of water-
fowl observed were: 9,853 dabbling ducks, 12,437 diving ducks, 538 mergansers,
4,446 Canada geese, and 25 Whistling Swans.
Figures 2 and 3 show the weekly totals of dabbling and diving ducks, respec-
tively. Figures 4 and 5 show weekly totals of Canada geese and coots, respectively.
Note that the numbers of waterfowl in all groups display seasonal ups and downs
typical of migratory waterfowl. Weekly fluctuations are also apparent and in-
dicate that ducks, geese and coots frequently move on and off the ponds. The
seasonal and weekly changes in duck and coot numbers indicate that few remain
on the ponds as permanent residents.
Large numbers of dabbling ducks, geese and coots visit the Hanford waste
ponds in the fall and midwinter (October through December) and in the spring
(March through May) . The Common Merganser is a visitor only in the fall and
winter. During these migration periods, most individual waterfowl are likely
to spend less than a few days or weeks associated with the ponds. It is equally
likely, however, that some birds spend over a month associated with waste ponds.
As shown by Figures 2, 3 and 5, a base population of about 100 dabbling ducks,
100 diving ducks and 150 to 200 coots occur on the waste ponds during the fall-
early winter and spring migration periods. Nesting surveys each year indicate
that between 10 and 30 pairs of dabbling ducks (Mallard, Pintail, Cinnamon Teal),
no more than 10 pairs of diving ducks (Ruddy Duck, Redhead, and Lesser Scaup),
and no more than 40 pairs of coots nested on the ponds.
The numbers of waterfowl that used the individual waste ponds can be esti-
mated from Table 1. Gable Mountain Pond received the greatest use by all water-
fowl groups. However, some species were observed more frequently at other waste
ponds. Of the dabbling ducks for instance, shovelers preferred B Pond. In the
diving duck group, goldeneyes and Bufflehead were most abundant on B Pond.
The reasons for these differences in pond use are unclear, but may be re-
lated to available food types, physical features, and human activity at each pond.
The openness of B Pond and scarcity of emergent aquatic plants may be attractive
to the three diving duck species. Canada geese may prefer Gable Mountain Pond
because of nearby cheatgrass (Bromus tectorum) fields in which the geese were
observed to forage. Coots and many of the dabbling ducks and diving ducks may
have been more abundant on Gable Mountain Pond because of the large water surface
area and the low degree of human activity. The Common Merganser's nearly total
association with Gable Mountain Pond may be directly related to a relatively
abundant food supply. This merganser feeds on fish, and Gable Mountain and
U Ponds were the only ponds containing fish. Goldfish were somehow introduced
-------
474
TABLE 1. Total Count of Waterfowl Observed on Hanford Waste
Ponds, September 1971 Through March 1974
Dabbling Ducks
Mallard
Gadwall
American Wigeon
Green-Winged Teal
Blue-Winged Teal
Cinnamon Teal
Shoveler
Pintail
B Pond
1390 (26)*
8 (1)
80 (5)
271 (48)
24 (28)
29 (18)
326 (54)
137 (17)
Total Dabbling Ducks 2265 (23)
Diving Ducks
Redhead 199 (25)
Canvasback 0
Greater Scaup 293 (26)
Lesser Scaup 429 (42)
Ring-Necked Duck 951 (23)
Common Goldeneye 626 (65)
Barrow's Goldeneye 39 (91)
Bufflehead 1870 (62)
Old Squaw 1 (9)
Ruddy Duck 108 (14)
Total Diving Ducks 4516 (36)
Mergansers
Hooded
American
Canada Goose
Whistling Swan
American Coot
0
1 (1)
926 (21)
0
1257 (7)
U Pond
Gable Mountain
Pond
1373 (26)
32 (5)
482 (27)
131 (23)
39 (45)
37 (23)
120 (20)
70 (9)
2284 (23)
25 (3)
4 (1)
120 (11)
12 (1)
113 (3)
24 (3)
4 (9)
61 (2)
0
62 (8)
425 (3)
2 (67)
3 (1)
0
0
330 (2)
2494 (48)
578 (94)
1214 (68)
164 (29)
23 (27)
93 (59)
159 (26)
579 (74)
5304 (54)
568 (72)
570 (99)
694 (63)
572 (57)
3107 (74)
299 (32)
0
1097 (36)
10 (91)
579 (77)
7496 (61)
1 (33)
531 (98)
3520 (79)
25(100)
17352 (91)
Percent distribution among ponds
-------
475
into these two ponds and a large supply is seasonally available to fish-eating
waterfowl.
The rank of puddle duck and diving duck abundance observed at all ponds is
shown in Table 2. A similar relative abundance has been reported in several
studies of nearby ponds located off of the Hanford Site (Je48, Ha54, Jo56 and
Yo60).
TABLE 2. Relative Abundance (Rank) of Ducks
Observed on Hanford Waste Water Ponds
Puddle Ducks Diving Ducks
1. Mallard 1. Ring-Necked Duck
2. Wigeon 2. Bufflehead
3. Pintail 3. Greater and Lesser Scaup
4. Gadwall 4. Common and Barrow's Goldeneye
5. Shoveler 5. Redhead
6. Green-Winged Teal 6. Ruddy Duck
7. Blue-Winged and 7. Canvasback
Cinnamon Teal
Cesium-137 in Pond Environments
The concentration of 1^7Cs in pond water at each of the study sites was
estimated from averaging results of environmental surveillance samples collected
quarterly from 1971 through 1977. (a' These values are shown in Table 3. Ana-
lytical standard errors for individual water samples may have been as large as
±100% and typical detection levels ranged from 30 to 50 pCi/fc through 1973.
Detection levels were reduced to 3 to 5 pCi/&» and standard errors were improved
for later samples because of improved analytical techniques.
TABLE 3. Average Annual Concentrations of 137Cs
at Hanford Waste Water Ponds, pCi/£
Gable Mountain
Year B Pond Pond U Pond
51 79
42 220
56 73
31 BDL
52 ' 42
31 8.0
14 13
BDL - Below Detection Limit
1971
1972
1973
1974
1975
1976
1977
36
BDL*
BDL
16
6.0
BDL
3.5
See Ho78 for the most recent Hanford environmental surveillance report.
-------
476
The majority of the 137Cs inventory present at each pond is in the upper 5 cm
of bottom sediments. Values reported for pond sediments are summarized in Table 4.
A great deal of spatial variability was reported as noted in the range of concen-
trations. Very little information has been reported on the concentrations of 137Cs
in waterfowl food sources available from pond environments, but the levels are con-
sidered to be low.
TABLE 4. Sediment Concentrations of 137Cs at
Hanford Waste Water Ponds, pCi/g Dry-Wt
Pond Average Concentration (Range)
U (Em74) 12000 (6000 - 21000)
Gable Mountain (Cu74) 30000 (13000 - 59000)
*
B 42
*
Single value
Cesium-137 in Waterfowl
Cesium-137 normally occurs in the flesh of waterfowl collected from Hanford
waste water ponds. A small amount can be attributed to worldwide fallout, but
some is undoubtedly contributed by the waste pond environment. The data presented
in Figures 6 and 7 show that dabbling ducks on Hanford waste ponds (n = 53) had
higher body burdens of 137Cs than similar ducks on the Columbia River (n = 290).
Dabbling ducks from waste ponds had body burdens an order of magnitude greater
than dabbling ducks collected from the Columbia River. The solid-line curves
are computer-generated least squares regression lines and 0.1 pCi/g is the
approximate analytical detection level (ADL). The geometric mean (50%) for
ducks collected on the Columbia River is below the ADL. However, the similarity
in slopes of the regression curves for pond and river ducks indicate a similar
degree of variability in each population.
Surveillance data for waste water ponds were not extensive enough to compare
the 137Cs levels in mergansers, geese or coots. However, ducks collected from
the waste ponds were compared according to the ponds they used and according to
dabblers versus divers. The most frequently collected dabbling duck was the Mal-
lard, whereas Buffleheads and goldeneyes were the most common diving ducks. Figures
8, 9 and 10 are log-normal plots of 137Cs concentrations for all ducks collected
on Gable Mountain Pond (n = 32), B Pond (n = 24) and U Pond (n = 25) . Again,
0.1 pCi represents the approximate analytical detection level. The slopes of
the regression curves are virtually the same and indicate similar variability
from pond to pond. The geometric means were approximately 39 pCi/g for Gable
Mountain Pond, 4.6 PCi/g for B Pond, and 36 pCi/g for U Pond. Clearly, ducks
collected from B Pond-were less radioactive than ducks collected from Gable
Mountain or U Ponds. The maximum !37Cs levels reported for a duck from each
pond were 210 pCi/g at Gable Mountain Pond, 260 pCi/g at B Pond, and 130 pCi/g
f r rTr' ^ IT* I8 VfU6S (FlgUre 9) f°r B Pond do not follow the pattern
for other collected ducks and may represent ducks that came from another waste
pond. The two low values for Gable Mountain Pond (Figure 8) also are unusual •
and may represent recent immigrants from offsite.
-------
477
Differences of 137Cs in dabbling ducks and diving ducks for all ponds com-
bined can be compared using Figures 7 and 11. The steep slope of the regression
curve for diving ducks (Figure 11) indicates large variability (large spread) in
the data and may be the result of a small sample size (n = 14). However, the
maximum concentration levels for 137Cs in dabbling ducks (=130 pCi/g) is virtually
the same as for the divers (=120 pCi/g).
Conclusions
The abundance of surface water in the region of the Hanford Site attracts
waterfowl. In addition, the isolation and lack of hunting pressure is conducive
to waterfowl use of the Hanford waste water ponds. However, the ponds probably
provide a better food source for diving ducks and coots than for dabbling ducks
or Canada geese. Census data show that pond use by diving ducks and coots was
greater than use by any other category of waterfowl. The reason for this greater
use may be that dabbling ducks and geese have the option of feeding on terrestrial
plants. Diving ducks and coots may use other local bodies of water, but the
amount of interplay is unknown. Census data taken at the waste ponds indicate
seasonally high use by waterfowl coinciding with fall and spring migratory periods.
Base population estimates indicate 300 to 400 resident waterfowl, but nesting
surveys noted only several tens of pairs.
Surveillance data on 137Cs concentrations in dabbling ducks show a difference
between those collected from the Columbia River and those collected from waste
ponds. For example, a dabbling duck would be much more likely to contain a
higher than normal 137Cs concentration if it were collected from a waste pond.
Seasonal differences in 137Cs concentrations for any waterfowl were not evident.
Waterfowl use of ponds and 137Cs concentrations in muscle tissue seem to
be related to pond size, degree of human disturbance, ecological stage, and
sediment concentration of 137Cs. The concentration of 137Cs in waterfowl flesh
was greatest at U Pond and Gable Mountain Pond where 137Cs sediment concentrations
are highest. The high degree of human disturbance at U Pond may offset the ad-
vantage of an advanced stage of ecological development and result in lower water-
fowl use than at other ponds. B Pond represents a poor habitat, but its large
size seems to attract waterfowl, especially diving ducks. The fact that there
was little difference in the maximum level of 137Cs in birds collected at each
pond indicates that waterfowl probably fly between ponds.
-------
478
References
Cu74 Gushing C. E. and Watson D. G., 1974, Aquatic Studies of Gable Mountain
Pond, Battelle Pacific Northwest Laboratory, Richland, WA, BNWL-1884.
Em74 Emery R. M., Klopfer D. C. and Weimer W. C., 1974, The Ecological Behavior
of Plutonium and Americium in a Freshwater Ecosystem. Phase I: Limnological
Characterization and Isotopic Distribution, Battelle Pacific Northwest
Laboratory, Richland, WA, BNWL-1876.
Ha54 Harris S. W., 1954, "An Ecological Study of the Waterfowl of the Potholes
Area, Grant County, Washington," Amer. Midi. Nat. 52, 403-432.
Ho78 Houston J. R. and Blumer P. J., 1978, Environmental Status of the Hanford
Site for CY-1977, Battelle Pacific Northwest Laboratory, Richland, WA, BNWL-2277.
Je48 Jeffrey R. G., 1948, Notes on Waterfowl of Certain Pothole Lakes of
Eastern Washington. Unpublished Master of Science Thesis. Washington State
University, Pullman, Washington.
Jo56 Johnsgard P. A., 1956, "Effects of Water Fluctuations and Vegetation Changes
in Bird Populations, Particularly Waterfowl," Ecology 37, 689-701.
Wa76 Waite D. A. and Bramson P. E., 1976, Interpretation of Near-Background
Environmental Surveillance Data by Distribution Analysis, in Biological and
Environmental Effects of Low-Level Radiation Vol. II. pp. 291-303 (Vienna:
IAEA) IAEA-SM-202/706.
Yo60 Yocum C. and Hanson W. C., 1960, "Population Studies of Waterfowl in
Eastern Washington." j;. Wildl. Manage. 24, 237-250.
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479
20
15
uJ 10
UJ
"• 5
0
-5
'U-POND"
TREE WILLOWS
AND COTTONWOODS
REEDS
WATER
LEVEL
15
UJ
UJ
5
0
-5
10
0
-5
"GABLE MOUNTAIN POND"
SHRUB WILLOWS
CATTAILS AND REEDS
- WEEDS CN
WATER
LEVEL
BARNYARD
GRASS
CATTAILS
AND REEDS
"B-POND"
DEAD TREES
WATER
LEVEL
FIGURE 1. Vegetation Zonation Around
Hanford Waste Water Ponds
-------
480
700 \-
DABBLING DUCK ABUNDANCE
HANFORD WASTE PONDS
J FMAMJJASONDJ FMAMJJASOND
S O N D
1971
J F M
1974
FIGURE 2. Weekly Totals of Dabbling Ducks Observed
at Hanford Waste Water Ponds
700 \-
DIVING DUCK ABUNDANCE
HANFORD WASTE PONDS
1974
FIGURE 3. Weekly Totals of Diving Pucks Observed
at Hanford Waste Water Pon^S^
-------
481
SUU
700
<
Q
> 500
Q
Z
LL
O
c
UJ
CD 300
Z
100
50
0
CANADA GOOSE
—
-
-
-
•
1
o
ill
N
D
J
F
II
M
A
M
1971
J
J
A
s!o
1
NlD
1972
ll
J
F Ml AlMl J
J 1 A
S
O
N
1973
ll
O J F
1974
FIGURE 4. Weekly Totals of Canada Geese Observed
at Hanford Waste Water Ponds
1971
1972
1973
FIGURE 5. Weekly Totals of Coots Observed
at Hanford Waste Water Ponds
-------
1000
100
(9
UJ
5
I:
O)
10
0.1
: "'Cs DABBLERS ALL PONDS
1971-1977
.x
I
20 40 60 80
CUMULATIVE PERCENT
98
FIGURE 6. Cesium-137 Concentration in Dabbling Ducks
Collected from Hanford Waste Water Ponds
1000
100
P
O
UJ
t 10
£
O)
0.1
137Cs DABBLERS COLUMBIA RIVER
1971-1977
20 40 60 80
CUMULATIVE PERCENT
98
FIGURE 7. Cesium-137 Concentration in Dabbling Ducks
Collected from the Columbia River
-------
483
1000
- 100
(9
UJ
fc
s ALL WATERFOWL GABLE MTN POND
1971-1977
10
a
s
X X
X X
0.11
J 1—I—I I I I I
20 40 60 80
CUMULATIVE PERCENT
98
FIGURE 8. Cesium-137 Concentration in All Ducks
Collected from Gable Mountain Pond
1000
_. 100
= 137Cs ALL WATERFOWL B POND
1971-1977
(9
10
0.1
x x
20 40 60 80
CUMULATIVE PERCENT
FIGURE 9. Cesium-137 Concentration in All Ducks
Collected from B Pond
-------
484
1000
1«Cs ALL WATERFOWL U POND
1971-1977
100
x
-------
485
CESIUM-137 IN COOTS (FuUoa amerioana) ON HANFORD
WASTE PONDS: CONTRIBUTION TO POPULATION DOSE
AND OFFSITE TRANSPORT ESTIMATES
L. L. Cadwell
R. G. Schreckhise
R. E. Fitzner
Ecosystems Department
Pacific Northwest Laboratory
Richland, Washington 99352
Abstract
American coots (Fulioa amerioana) were periodically collected from ponds
receiving low-level radioactive waste on the Hanford Site and from ponds on a
control area. Gut contents and selected tissues were removed and analyzed for
1-^Cs, 90gr an
-------
486
Although the coot is not necessarily a preferred game species, it is rela-
tively abundant, easy to bag, and a part of the waterfowl hunters' harvest.
During the 1973 hunting season, the "retrieved kill" data for Washington State
waterfowl hunters included 8,104 coots, which comprised 1.3% of the total bag
for all waterfowl species (US78).
The Study Area
The ponds are used for disposing of process waste water by percolation
and evaporation. They have undergone ecological succession since their forma-
tion and have shoreline vegetation that provides suitable nesting habitat and
food supply for waterfowl (Fi75).
Gable Mountain Pond, with a 29 hectare surface area, is the largest of
the three ponds. Constructed in 1957, this pond contains the greatest
quantity of fission products, primarily as a result of an accidental release
in 1964 (An74). B-Pond, in existence since 1945, has a surface area of
approximately 19 hectares. It contains the smallest quantities of radioactive
materials as the result of minor unplanned releases. U-Pond is the smallest
with an area of about 6 hectares. It was formed in 1944 and receives waste
water from a laundry facility that contains low levels of radionuclides. Of
the three ponds it contains the smallest quantity of fission products but the
greatest amount of plutonium.
Methods
One hundred three coots were collected from Gable Mountain Pond, 18 from
U-Pond and 31 from B-Pond. An additional 13 coots were collected from the
Columbia Wildlife Refuge in central Washington to serve as controls to
determine fallout concentrations of radionuclides. Collections were made at 1
to 2 month intervals from June, 1974 through March, 1976. An additional
collection was made in January, 1977. Five to nine birds were collected
during each sampling period.
The birds were whole-body counted to determine !37Cs content, then they
were weighed, dissected and selected samples were removed for radiochemical
analysis. The samples included muscle, liver, bone and contents of the
intestinal tract. Samples were analyzed for gamma-emitting radionuclides with
a sodium iodide detector and multichannel analyzer. Selected samples were
also sent to a commercial laboratory where they were analyzed for 90gr an(j
gross Pu.
Results and Discussion
The concentrations of 90Sr, 137Cs and gross Pu for selected coot
sample types from birds collected at Gable Mountain Pond are summarized in
Table 1. These data show clearly that 137Cs occurred in greatest concentra-
tion in tissue from Gable Mountain Pond coots. Additional plutonium analyses
from U-Pond coot samples were made because of the high Pu content in U-Pond
relative to the other two ponds. These analyses showed the Pu concentrations
in muscle from U-Pond coots were only slightly greater than for Gable Mountain
Pond birds, but still insignificant in comparison with !37Cs (Table 1).
-------
Table 1. Average Concentration of 90Sr, and 137Cs and Gross Pu in Selected
Tissues of Coots Collected at Gable Mountain Pond
Concentration (pCi/g dry weight)
90Sr 137cs
Sample
Type
Bone
Liver
Muscle
Gut Contents
Average Dry Weight
per Coot (g)
24.0
4.3*
59.0*
3.0
(n =
X
2.6
0.53
0.28
4.3
12)
SE
0.6
0.20
0.12
1.5
(n =
X
200.
440.
570.
3,400.
103)
SE
30.
40.
40.
200.
Gross Pu
(n =
X
0.023
0.052
0.019
0.14
24)
SE
0.006
0.031
0.015
0.03
*Multiply by 2.9 to convert to wet weight.
n = 16 for muscle only.
Table 2. Average Concentration of
from the Study Areas
in Samples of Coots
Concentration (pCi/g dry weight)
Gable Mountain
Sample Type
Bone
Liver
Muscle
Gut Contents
Pond (n
X
200.
440.
570.
3,400.
= 103)
SE
30.
40.
40.
200.
U-Pond
(n =
X
70.
220.
360.
1,300.
18)
SE
10.
20.
30.
200.
B-Pond
(n =
X
5.7
16.
30.
85.
31)
SE
0.8
2.
4.
11.
Columbia Wildlife
Refuge
X
1.0
0.7
0.02
0.8
(n = 13)*
SE
0.3
0.2
0.05
0.5
*Columbia Wildlife Refuge concentrations were near or below detection limits
Ov>0.5 pCi/g), which varied with sample size.
oo
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488
Concentrations of !37Cs £n sampies from all three ponds are shown in
Table 2. Gable Mountain Pond coots were highest for all sample types. The
significance of the radionuclide inventory in birds from the other two ponds
is reduced when one also considers that over 90% of the total number of coots
observed on the Site were on Gable Mountain Pond (Fi75).
Coots from Gable Mountain Pond were sampled periodically to determine
whether there were any seasonal trends in their l"Cs content. Figure 1
shows no clear-cut pattern for !37Cs concentration in muscle, but there was
also a high degree of variability in concentrations among samples for any one
collection date. Part of the reason for the variability is probably due to the
fact that some coots sampled were recent arrivals and thus had not attained
muscle concentrations near equilibrium. A ratio (^3'Cs concentration in the
gut/!37Cs concentration in the muscle) was calculated as an index of whether
or not the birds sampled were recent arrivals. Since the gut content reflects
the concentration of 137Cs in the food, it readily equilibrates and will show
a high concentration within two or three days. Muscle, on the other hand,
requires longer to reach an equilibrium concentration. The time required is
proportional to the biological half-time for !37Cs £n muscle. Although the
muscle half-time for !37Cs in coots is unknown, it is probably similar to
that reported for other waterfowl. Fendley et al. (Fe77) reported a 137Cs
half-time of 5.6 days for wood ducks, and Halford et al., (Ha78) reported a
2000
1800
1600
WOO
1200
1000
800
600
400
200
0
JU
-T
1 1 '
1
1
N SEP DEC
1974
1
• = MEAN CONCENTRATION
T RANGE OF CONCENTRATION
1 OBSERVED
i ,
<
1
MAR JUN SEP DEC
1975
1
1 1 1
MAR JUN SEP DEC
1976
COLLECTION DATE
Figure 1. 137Cs concentration in muscle of coots from
Gable Mountain Pond (June 1974 - January 1977)
-------
489
value of 11.7 days for mallards. Therefore, one would expect that it would
take from 1 to 2 months, or approximately 5 half-times, for the concentration
of Ij/Cs in coot muscle to equilibrate. Newly arrived birds, with high gut
to muscle ratios can be distinguished from birds which have been on the pond
long enough for muscle concentrations to approach equilibrium. A plot of the
gut to muscle ratio for all Gable Mountain Pond coots (Figure 2) shows that the
ratios for most of the birds was clustered in the range of 3 to 10. The ratios
from about 10% of the coots was considerably higher (above 20). Our interpre-
tation is that those birds were relatively new arrivals. The highest ratios
occurred during the months of September and October which correspond with the
expected influx of migrant birds. These months also overlap the hunting season
for waterfowl in Washington. Thus, average muscle concentrations for the year,
(used here for dose to man calculations) should provide upper-limit dose
estimates since part of the population harvested during the hunting season may
not have muscle concentrations at equilibrium.
The total-body 50-year dose commitment (D) from 137Cs to an individual
ingesting all edible tissue from one coot (0.03 u£i) was calculated according
to Morgan and Turner (Mo67) using parameters from ICRP #10 (In68):
M
2.1 mrem
18250
/
0.693(18250 days)
TE
" dt
where, E = effective energy for 137Cs (0.59 Mev/dis);
M = total body mass (7 x 10^ g);
fw = fraction of ingested 137Cs reaching organ of reference (1.0); and,
TE = effective half-time of 137Cs in total body (114 days).
A similar calculation of a one-year dose gives approximately 1.9 mrem. ^Thus,
a member of the general public consuming one coot would receive about 2/4 of the
dose an individual receives from natural background sources of approximately
100 mrem/year (Ei73). This would also be equal to 1.1% of the U.S. Department
of Energy's radiation protection standard of 170 mrem/year for individuals and
population groups in uncontrolled areas (US77).
Harvest data for coots in the State of Washington for the 1973 hunting
season show that hunters bagged approximately 8,100 birds (US78). Breeding
season population estimates for the same year put the state coot population at
37,240 birds. The ratio of retrieved harvest to population estimates suggest
that approximately 22% of the state population may have been harvested. An
average of 300 coots were observed on the Hanford area waste ponds during the
1973 hunting season (Fi75). If one assumes that 22% were bagged then
approximately 70 coots might have been harvested. Since about 9U or tne
Hanford area coots were on Gable Mountain Pond (Fi75), then the harvest
-------
490
215
200
135
120
105
90
75
60
45
30 -
15
n i I;
JUN SEP DEC
1974
MAR JUN SEP DEC
1975
COLLECTION DATE
MAR JUN
1976
137.
Figure 2. Ratio (gut content/muscle) for "'CB concentration
in coots from Gable Mountain Pond
estimate for that pond for 1973 is 63 birds. The 50-year dose commitment to
the general public (waterfowl hunters and their families) from consuming the
edible portion of 63 Gable Mountain coots (1973 data) was 0.13 person-rem
The harvest estimate and subsequent dose calculations from Gable Mountain'pond
coots is very likely high. It is unlikely that coots using the Hanford area
ponds will be harvested at a rate as high as the state average Coots must
leave the confines of the Hanford Site to be available to hunters
The movements of coots from Hanford area waste ponds have not been
investigated extensively enough to accurately determine the quantity of
1J'Cs transported offsite. However, the amount of 137Cg avaiuble for
annual export is the product of the coot whole-body burden and the number of
coots that leave the ponds each year. If one examines tS ^ number of
Fitzner and Rickard (Fi75) and assumes that the reduct ion ,' K' A***?** ^
succeeding observations reflects the movemen of birS away"^™ th "7* *
then the number of birds leaving the ponds averaged about 500 I / P '
during 1972 and 1973. The coot body burdens werfo to (I 0 O^TosT n
and 0.004 (± 0.001) LtCi for Gable Mountain, U- and B-P ( °'
The average body burden, weighted according to'rSltive
the three ponds, was 0.092 KCi. Therefore! the annual eort r
the ponds could be about 46 ,Ci/year (0.092 .Ci/^ot x SO^coot /year)
f
"
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491
Acknowledgements
We thank Mr. H. A. Sweany for his technical assistance during the study
The study was supported in part by Rockwell Hanford Operations' Long-Term
Management of Low-Level Waste Program and in part by the Department of Energy
Office of Health and Environmental Research.
References
An74 Anderson J. D., 1974, Radioactive Liquid Wastes Discharged to Ground
in the 200 Areas During 1973. ARH-2806, Atlantic Richfield Hanford Company
(NTIS, Springfield, VA).
Cu74 Gushing C. E. and Watson D. G., 1974, Aquatic Studies of Gable
Mountain Pond, BNWL-1884, Battelle, Pacific Northwest Laboratories.
(NTIS, Springfield, VA).
Ei73 Eisenbud M., 1973, Environmental Radioactivity. 2nd Ed. (New York:
Academic Press).
Fe77 Fendley T. T. Manlove M. N. and Brisbin I. L. , Jr., 1977, "The
Accumulation and Elimination of Radiocesium by Naturally Contaminated Wood
Ducks", Health Phys. M, 415-422.
Fi75 Fitzner R. E. and Rickard W. H., 1975, Avifauna of Waste Ponds ERDA
Hanford Reservation Benton County, Washington, BNWL-1885, Battelle, Pacific
Northwest Laboratories. (NTIS, Springfield, VA.)
Fi77 Fix J. J. and Blumer P. G. , 1977, Radiochemical Analysis of Game Birds
Collected from the Hanford Environs 1971-1975. BNWL-2089, Battelle, Pacific
Northwest Laboratories. (NTIS, Springfield, VA.)
Fi79 Fitzner R. E. and Schreckhise R. G., 1979, The American Coot (Fulica
americana) on the Hanford Site, Part I: Nesting Biology. PNL-2462,
Pacific Northwest Laboratories. (NTIS, Springfield, VA.)
Ha78 Halford D. K. Millard J. B. and Schreckhise R. G., 1978, "Retention of
Activation and Fission Radionuclides by Mallards from the Test Reactor Area
Radioactive Leaching Pond", Ecological Studies on the Idaho National
Engineering Laboratory Site. 1978 Progress Report. IDO-12087.
In68 International Commission on Radiological Protection, 1968, Evaluation
of Radiation Doses to Body Tissues from Internal Contamination Due to
Occupational Exposures. ICRP Publication 10. (New York: Pergamon Press).
Mo67 Morgan K. Z. and Turner J. E. (Editors), 1967, Principles of Radiation
Protection. (New York: John Wiley and Sons, Inc.).
US77 Department of Energy, 1977, "Standards for Radiation Protection" DOE
Manual Chapter 0524 Appendix, Part 3, approved March 3, 1977.
US78 US Department of the Interior, Fish and Wildlife Service, 1978,
Waterfowl Status Report 1974, Special Scientific Report-Wildlife No. 211,
.Washington, DC.
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492
RESPONSE TO A WIDESPREAD, UNAUTHORIZED DISPERSAL OF
RADIOACTIVE WASTE IN THE PUBLIC DOMAIN
F. A. Wenslawski and H. S. North
U. S. Nuclear Regulatory Commission, Region V
1990 N. California Boulevard
Walnut Creek, California 94596
Abstract
In March 1976 State of Nevada radiological health officials became aware
that radioactive items destined for disposal at a radioactive waste burial
facility near Beatty, Nevada had instead been distributed to wide segments of
the public domain. Because the facility was jointly licensed by the State of
Nevada and the federal Nuclear Regulatory Commission, both agencies quickly
responded. It was learned that over a period of several years a practice
existed at the disposal facility of opening containers, removing contents and
allowing employees to take items of worth or fancy. Numerous items such as
hand tools, electric motors, laboratory instruments, shipping containers, etc.,
had received widespread and uncontrolled distribution in the town of Beatty
as well as lesser distributions to other locations. Because the situation might
have had the potential for a significant health and safety impact, a comprehen-
sive recovery operation was conducted. During the course of seven days of
intense effort, thirty-five individuals became involved in a comprehensive door
by door survey and search of the town. Aerial surveys were performed using a
helicopter equipped with sensitive radiation detectors, while ground level scans
were conducted using a van containing similar instrumentation. Aerial recon-
naissance photographs were taken, a special town meeting was held and numerous
persons were interviewed. The recovery effort resulted in a retrieval of an
estimated 20-25 pickup truck loads of radioactively contaminated equipment as
well as several loads of large items returned on a 40-foot flatbed trailer.
Discussion
In February 1976, the Radiation Safety Officer (RSO) at a low-level
radioactive waste land disposal area near Beatty, Nevada noticed that the
company's stockpile of bagged cement was diminishing disproportionately to the
quantity of liquid waste being solidified. The RSO, who was relatively new
to the site, found on investigation that some company employees had been using
the facility's cement mixing truck and materials to pour concrete on several
different jobs in the nearby town of Beatty. The truck was routinely used
to solidify liquid radioactive waste onsite and was internally contaminated.
The company reported this matter to the State of Nevada, radiological health
officials.
It was not uncommon for uncontaminated site equipment, road graders, etc.,
to be used in assisting area residents and on civic projects. To better under-
stand the events which occurred at and around the disposal site, it is essential
that one understand the somewhat unique characteristics of the locale. Beatty
is a small, remote desert town of approximately 500 residents located about
125 miles north of Las Vegas, Nevada. The area is bordered on the west by the
-------
493
Death Valley National Monument and on the east by the U.S. government's Nevada
Test Site. Employment in the area generally comes from the Nevada Test Site,
a few tourist-related businesses, ranching, mining and the waste burial facility.
Many residents of the town were employees or former employees of the waste
facility or were relatives or friends of these people. Because of the difficulty
and cost of obtaining materials in such a remote location, the people of Beatty
appeared reluctant to part with any item that could have salvage value. Many
residences were observed to have accumulated stockpiles of salvageable items
including lumber, tires, automobiles, appliances and a myriad of other materials.
It was in this isolated realm of self-sufficiency, that items destined for
disposal at the waste facility found their way into the public domain.
The waste facility itself is a low-level radioactive waste burial operation,
which in 1976 was jointly licensed by the State of Nevada and the U.S. Nuclear
Regulatory Commission (NRG). The State regulated the disposal of source, by-
product and natural occurring materials while the NRC's regulatory activities
were limited to special nuclear materials (SNM). Materials shipped for disposal
came from a wide variety of users of radioactive materials including military,
medical, industrial and research facilities. A large quantity of material
arriving at the site for disposal consisted of equipment that became contaminated
during usage. Typical items included protective clothing, tools of all kind,
laboratory equipment and instruments, electric motors, piping, wiring and other
paraphernalia. In addition, a large quantity of radium dial clocks, watches,
compasses, assorted gauges and other similar obsolete military equipment were
received at the site for disposal over the years.
During their investigation of the cement mixer occurrence, State officials
discovered that other possibly contaminated materials had been removed from the
disposal site in past years and were distributed in the town of Beatty. The
State quickly responded by suspending the company's license on March 8, 1976
and halting all State licensed operations at the site (similar action was sub-
sequently taken by NRG). The State also requested radiological assistance from
the Environmental Protection Agency's (EPA) Las Vegas facilities to help locate
contaminated materials in the town of Beatty. The request was made to EPA
because of their proximity, their monitoring capabilities and because of the
existence of professional acquaintances. The State also notified NRC which
promptly dispatched an inspector to the site.
NRC's initial role was in an investigative capacity only, because there
had been no direct evidence of the release from the site of articles contam-
inated with SNM. However, early survey efforts did identify the presence of
a Plutonium contaminated triple beam balance and weights in a private residence.
It was disclosed that the scale and weights had come from the disposal site,
thereby involving the SNM license and creating a situation under joint State
and NRC jurisdiction. The State and NRC subsequently joined their resources
in a cooperative effort.
Through detailed investigation it was disclosed that during the period
of 1967 to 1973 it was common practice at the site to open containers of
radioactive waste, remove the contents and allow employees to take what they
wanted, it was disclosed that over a period of time, mostly through trial and
error, employees were able to determine which containers held desirable items.
This could be determined in large measure by the location from which the con-
tainers were received. Another method employed to determine whether a container
-------
494
was to be opened was to roll it over and if it rattled, it was a good indication
that the contents were of value. Such containers came to be referred to as
"rattlers". From information gathered, it was apparent that a myriad of items
destined for disposal had actually been diverted into the public domain.
Plywood shipping containers were routinely disassembled to salvage the wood.
Numerous radium dial objects were taken. Contaminated tools of all kinds were
diverted. It was apparent that any item of practical use or of just human
interest had the potential for being diverted. Interrogation of involved in-
dividuals revealed that only minimal radiation surveys were performed on
materials removed from the site.
One item of particular interest which had been routinely diverted was a
large (about 200 cubic ft.) cylindrical steel shipping container referred to
as a "Bennett Bucket." These containers were fabricated and used for waste
shipments by a large laboratory involved in the use of large quantities of
radioactive materials including transuranics. Upon receipt at the disposal
site, the contents of these containers were usually dumped into a disposal
trench and the containers were diverted for such uses as septic tanks, animal
grain and water storage and as storage tanks for both potable and nonpotable
water. From detailed investigation it was determined that approximately 141
Bennett Buckets were shipped for disposal during the 1967 to 1973 time period.
It was estimated that about 95% of these were opened and diverted from disposal.
On the afternoon of March 10, EPA personnel began to conduct radiation
surveys in the town of Beatty and on outlying ranches. Systematic assign-
ments were made to assure complete survey coverage. Because the majority of
the town's streets were unnamed and maps did not exist, aerial photographs
were used to identify locations. A "Statement Of Authority For Investigation"
was issued by the Nevada State Health Officer. The Statement delineated
applicable Nevada statutes that authorized the entrance upon private property
and authorized the impoundment of materials found to be radioactive. Copies
of the Statement were carried by personnel performing surveys. However, the
legal authority was not frequently imposed as the vast majority of residents
were cooperative in allowing surveyors to enter their property and, if applicable,
remove contaminated or radioactive items.
The principal survey instrument used was a sensitive, fast response gamma
scintillation instrument (He76). Other survey instruments, including ones
with alpha detection capability were available and used as circumstances dictated.
Surveyors worked on an individual basis, carefully surveying each room of resi-
dences, as well as yards, garages, storage sheds, business establishments and the
piles of salvable items prevalent throughout the town. General ground rules were
to perform a thorough gamma survey, make note of any suspicious looking items
for followup alpha surveys, only pick up items exhibiting positive indications of
radioactivity, and make note of identified contaminated items which were not
picked up in order to allow later pick up. By the evening of March 10, it became
apparent that large quantities of contaminated material were present in the town
as pickup truck loads of items were identified and recovered.
On March 11, 1976, NRC issued an order to the disposal company suspending
NRG licensed activities at the facility. In addition, under terms of the
Interagency Radiological Assistance Plan (IRAP), NRC requested radiological
assistance from the Energy Research and Development Administration (ERDA, now
-------
495
Department of Energy DOE) to assist in the comprehensive radiation surveys
already initiated in the town of Beatty. Additional NRG personnel were dis-
patched to the site to assist. NRG inspectors were also dispatched to three
other disposal sites to determine whether a similar problem might exist at those
sites. The State of Kentucky was notified because of a site there under sole
State jurisdiction. It was found that the situation was unique to the Beatty
site.
On the afternoon of March 11, an NRG representative in Beatty met with
ERDA representatives to brief them on the scope of the problem. ERDA responded
with a team consisting of ERDA employees and personnel from three ERDA contrac-
tors (EG&G, Inc., Lawrence Livermore Laboratory (LLL), and Reynolds Electrical
and Engineering Co. (REECO)). On the morning of March 12, EPA and ERDA survey
efforts were quickly intergrated with overall responsibility being jointly
shared by the State of Nevada and NRG. Three-man survey teams were formed con-
sisting of two personnel performing alpha surveys and one performing gamma
surveys. In addition, an EG&G helicopter and van, both equipped with a large
array of sodium iodide gamma detectors, were used to survey outlying ranches
and dispersed areas as well as locations in town (Jo76). A LLL van equipped
with a portable GeLi detector and multichannel analyzer was used for isotopic
identification of samples and to perform in situ analyses where necessary (An76).
A REECO sodium iodide gamma detector with a multichannel analyzer and a portable
gas proportional beta/alpha detector with a sealer were set up in a Beatty motel
being used as a "command post" (Re76). A base of radio communications was estab-
lished at the "command post" and walkie-talkies were used for field communications.
Overall, approximately 35 individuals had become involved in the survey effort.
By Friday, March 12, survey teams began to hear rumors that some material
was being buried, disposed of in the desert or otherwise hidden by residents,
apparently out of fear of being "caught with materials." To counter this and
to establish better public relations, State officials called a town meeting that
evening. Notice of the meeting was spread by the county sheriff's office.
About 130 residents attended the meeting which was held in the school gymnasium.
The local residents were informed of the situation and told that the agencies
were principally interested in recovering the materials and assuring that there
was no health hazard. The townspeople were assured that no action would be
taken against them. Residents were afforded the opportunity to ask questions.
Generally, the tone of questions from townspeople was serious and concerned.
The atmosphere was good natured and friendly. The meeting lasted about one
hour. Results of the town meeting were evident on the ensuing weekend as people
voluntarily returned miscellaneous items.
Survey efforts continued on March 13 and 14 and by the evening of March 14,
the bulk of work had been completed and the survey teams were disbanded. Approx-
imately 20 to 25 pickup truck loads of materials were recovered. In addition,
several loads of large, bulky and heavy items were returned on a 40-ft. flatbed
truck. A skeleton crew of personnel remained in Beatty through Friday March 19
Performing followup surveys and generally wrapping up loose ends. On Monday,
March 15, representatives from NRG, EPA and the waste disposal company met with
the Governor of Nevada and other State officials to hold a debriefing. The
Governor had maintained a keen interest and had requested frequent updates
during the course of the survey effort. A joint State/NRG press conference was
held following the meeting with the Governor. The press conference was attended
by several Las Vegas radio and television stations as well as the printed press.
News coverage waned quickly after the initial coverage.
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496
Isolated survey efforts continued through March 26. Previously unavailable
locations in and around Beatty were surveyed. It was necessary from information
gathered during investigations in Beatty to followup leads and conduct surveys
in Indian Springs, Pahrump and Fallen, Nevada as well as Richland Washington
and a few locations in California. A few additional items were recovered as a
result of these surveys.
During the course of surveying in Beatty and the vicinity, about two
dozen Bennett Buckets were located. By best estimates it was considered that
a little over one hundred unaccounted for Bennett Buckets were still in the
public domain. Since many of these containers were used to ship significant
quantities of the more hazardous radionuclides, NRG concluded that further
effort was necessary to locate the containers. On March 30, the NRG Region V
office issued 1500 postal patron addressed letters covering the geographical
areas which could reasonably be affected. The letter briefly described the
circumstances, provided a sketch of a Bennett Bucket and provided a convenient
reply form with a self-addressed postage paid envelope. The letter assured
residents that NRC's interest was to locate and survey the containers or
other materials believed to have come from the disposal site.
Thirty responses were received to the letters (only responses with positive
information were requested). Of the thirty, the majority reported negative
information, a few responded with uncomplimentary remarks and a few reported
containers or other materials which may have originated at the disposal site.
Followup surveys were made on these items and, as a result, a few more items
were recovered.
An attempt was made to quantify the cost of responding to this radiological
incident. It was estimated that approximately $50,000 combined cost was
incurred by the various organizations involved in investigating the incident,
conducting surveys and generally evaluating the health and safety significance.
The specific impact of this occurrence on the public health and safety is the
subject of another paper entitled "Health and Safety Implication of a Widespread,
Unauthorized Dispersal of Radioactive Waste in the Public Domain" (No79).
References
An76 Anspaugh L.R., 1976, Private communication.
He76 Hendricks D.W., and Fort Jr. C.W., 1976, Radiation Survey in
Beatty, Nevada and Surrounding Area (March 1976), ORP/LV-76-1,
Office of Radiation Programs - Las Vegas Facility (Las Vegas, NV:
U.S. Environmental Protection Agency).
Jo76 Jobst J.E., 1976, A Radiological Survey of Beatty, Nevada and the
Amargosa Valley, EG&G/Aerial Surveillance Department Report,
Las Vegas, NV.
No79 North H.S., and Wenslawski F.A., 1979, "Health and Safety,
Implication of a Widespread Unauthorized Dispersal of Radio-
active Waste in the Public Domain."
Re76 Reynolds Electrical and Engineering Co., Inc., 1976, Radiation
Survey - Beatty, Nevada, REECO Report, Mercury, Nevada.
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497
HEALTH AND SAFETY IMPLICATION OF A WIDESPREAD
UNAUTHORIZED DISPERSAL OF RADIOACTIVE WASTE '
IN THE PUBLIC DOMAIN
H. S. North, Jr. and F. A. Wenslawski
U. S. Nuclear Regulatory Commission, Region V
1990 N. California Boulevard
Walnut Creek, California 94596
Abstract
As a result of the distribution of contaminated and potentially contaminated
articles and containers in the public domain, assessment of the possible threat
to public health and safety was required. An extensive effort was made to
identify radioisotopes and assess the hazard associated with the large number of
items recovered during surveys. Exposure by direct radiation appeared minimal,
and the majority of effort was devoted toward identifying potential internal
exposures. Many of the recovered articles were intact obsolete military equip-
ment with luminous radium facings. Other typical items included a plutonium
contaminated balance scale and weights, a strontium-90 contaminated stainless
steel plate, a centrifuge with carbon-14 and cobalt-60 contamination and a hoist
assembly with plutonium contamination. A large number of specially fabricated
waste containers which had previously been used to ship significant quantities
of americium, plutonium and other nuclides were found in use for grain storage,
water supply tanks and other applications. Sensitive gamma radiation surveys and
smear and water samples identified no significant contamination in those tanks
which had been located. Based on available information of isotopic identifica-
tion and quantification, and location and use of contaminated items individuals
with the highest potential for internal deposition were identified for whole body
counting and bioassay. The results of whole body counting, including phoswich
counts for plutonium and bioassays for tritium, carbon-14, strontium 89-90 and
plutonium-238 and 239 did not yield any evidence of internal deposition. Two
employees of the waste disposal facility indicated higher than usual tritium
urine levels which returned to normal on resampling. Based on consideration of
all factors, this uncontrolled release of contaminated waste in the public
domain had no identifiable effect on the exposed population.
Discussion
The disclosure of the distribution of contaminated or potentially con-
taminated articles or radioactive materials discussed in the "Response to a
Widespread, Unauthorized Dispersal of Radioactive Waste in the Public Domain,"
(WE79), made necessary a prompt evaluation of the resulting threat to the
health and safety of the public. The purpose of this paper is to identify
the factors which resulted in the conclusion that this uncontrolled release
of contaminated waste in the public domain had no identifiable effect on the
Population at risk.
The early phase of the investigation was conducted concurrently at three
levels. The first involved field and laboratory measurements of radiation
and contamination levels in Beatty and the surrounding area. Second, during
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498
discussions with employees and former employees of the disposal company,
efforts were made to identify items or types of items which had been diverted
and, where possible, the recipients of such items. As information concerning
the recipients or the types of items diverted became available, the information
was provided to the field survey teams in order to permit the teams to be more
specific and alert for certain types of items. The third level involved
inquiries with certain organizations which had sent waste for disposal. The
latter was done in an effort to identify the level of hazard represented by
certain of the waste shipments that were routinely opened. Following the early
phase, individuals were selected for whole body counting and bioassay based on
the findings during the early phase.
The first field surveys were associated with slabs poured with concrete
from the transit-mix truck used at the site to solidify liquid waste. Inter-
views identified three, and a possible fourth, locations in Beatty where slabs
had been poured. The slabs were to be used as the foundation of a residence,
a patio behind a saloon, the foundation and floor of the town jail and court-
house building, and possibly the floor of a chicken coop. Portable instruments
used for these surveys included Baird Atomic NE-148A, Gamma Scintillators, with
ranges to 3 mR per hour and a low range of 0 to 30 uR per hour. This instrument
was capable of detecting a 5 uCi radium-226 source at six feet in air. The six
foot measurement was approximately 2 uR per hour above background in a 4 uR per
hour background field. The response of this instrument was sufficiently rapid
to make it a good "fast search" gamma detector. Eberline Instrument Corporation
(EIC) PAC-1SA, Alpha Scintillators, were used to monitor suspect locations and
items that had no detectable gamma levels, and EIC E-500 B, Beta/Gamma Meters
were used to survey gamma radiation levels above the range of the gamma
scintillation instruments (HE76).
Surveys of the slabs identified three localized areas on the saloon slab
which ranged from 8 to 45 mrad per hour, beta plus gamma, and a relatively
uniform, 3 to 6 uR per hour, above background for the residence slab. These
measurements were made with the detectors in contact with the slabs. Con-
tamination in the residence slab appeared to be relatively homogeneous. The
other two slabs were found to be background. Gamma spectrum analysis of
concrete samples from the saloon identified the presence of localized concentra-
tions of cobalt-60 with trace quantities of cobalt-58 and manganese-54. Samples
of left over concrete from the residence slab pour showed trace levels of
cobalt-58 and 60 and manganese-54. Concrete samples from the courthouse pour
showed only background on analysis. Samples from the chicken coop pour were
not analyzed. The saloon patio slab was removed and returned to the waste
disposal site for burial. No action was taken with respect to the residence
slab.
The continuing investigation identified an apparent widespread dispersal
of potentially contaminated or radioactive items in the community and the
surrounding area. In order to evaluate the threat to health and safety and
to recover contaminated or radioactive items for safe disposal, a door-to-door
survey including all rooms in each residence or business, all out-buildings,
and all piles of salvaged or construction materials was conducted.
The monitoring teams were provided with instruments of the type
previously described or with Ludlum Measurements, Inc., Model 125, Count Rate
Meter (Micro-R-Meter), a gamma scintillation survey instrument with three
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499
scale ranges to 3 mR per hour full scale. Also available were Reuter-Stokes
RSS-111, Environmental Radiation Monitors with ranges to 200 uR per hour- '
Ludlum Measurements, Inc., Model 2200 Sealer and Alpha Scintillator for swipe
counting (HE76) ; ETC, E-520 Beta/Gamma Survey Meters;and PAC-4G, Gas
Proportional Alpha Counters. A counting laboratory was set up in a local motel.
Gamma spectroscopy and proportional counting for swipe analysis was provided
at that location. Portable EIC, Model MS-3, Miniscalers with Ludlum
Measurement, Inc., Gas Proportional Chambers were used for swipe evaluation by
some field monitoring teams. Swipe samples collected during the surveys which
showed contamination were in most cases analyzed promptly. Certain samples
were retained for later more detailed analysis. The majority of samples were
analyzed by Reynolds Electrical and Engineering Company, Inc. at the Nevada
Test Site Laboratories (Re76). Additional analyses were performed by
Lawrence Livermore Laboratory's (LLL) portable Ge(Li) system (An76); EPA,
Environmental Monitoring Support Laboratory, Las Vegas (He76); and Health
Safety Laboratory, Idaho.
Contaminated or radioactive articles identified by the monitoring teams
were either collected during the surveys or, where large numbers of items were
involved, were marked for later collection. Cans of spray paint were used
for marking large items. It should be noted that background radiation in this
area is variable and quite high, considerably above that found at Mercury,
Nevada (An76). In addition, some of the residents are "rock hounds" and
collect interesting and occasionally naturally radioactive rocks. Some of
these materials had been incorporated into buildings (e.g., a rock fireplace)
which did not simplify the monitors' work.
In response to NRC's request for assistance under the provisions of the
Interagency Radiological Monitoring Assistance Plan (IRAP) , ERDA made available
three mobile detector systems. Two of the systems operated by EG&G were
mounted in a helicopter and a van, respectively. The third system operated by
Lawrence Livermore Laboratory (LLL) was also mounted in a van. This equipment
was of particular value in the survey effort in that it was found that material
from the waste disposal site had been transported substantial distances and
distributed over wide areas. One of the ranches where many items were found
covered 72,000 acres. Much of the area is broken and cut with dry washes.
Roads other than highways are rudimentary if they exist.
The EG&G van and helicopter were used to evaluate ranches approximately
15 miles north of Beatty and along the roads and highways near Beatty. The
van was also used in surveys along the streets and near piles of construction
materials and debris in Beatty. The helicopter was also used for visual
examination of remote or otherwise inaccessible areas.
The van was equipped with both neutron and gamma detector pods. The
neutron pod contained three Harshaw Type B4-72S/50 neutron detector tubes.
Each tube has a thermal neutron sensitivity of 1550 cps/n/cm sec. The gamma
Pod contained twenty 12.7 cm diameter by 5 cm thick Nal (Tl) crystals. The
gamma pod was oriented so that it was most sensitive to radiation from the side
of the van.. The minimum detectable point source gamma activity ranged from 10
"Ci of cobalt-60 at 20 feet to 90 uCi of cesium-137 at 50 feet. The helicopter
was equipped with gamma detectors similar in size and number to those contained
in the van but in two separate pods. The minimum detectable point source gamma
activity covered the same range as the van for direct fly overs. Lateral
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500
displacement of the detectors equal to the altitude in the range of 20 to 50
feet increased the minimum detectable point source activities by factors of
approximately 2.5 to 3 (Jo76). The ILL van was equipped with a portable
Ge(Li) spectrometer and analyzer. The Ge(Li) system was used to analyze
specific samples and was mounted on a one meter tripod for measurements in
certain buildings or near large structures (An76).
The surveys performed by the monitoring teams, vans and helicopter
established that there were no significant unshielded sources of gamma or
neutron radiation in Beatty or on the surrounding ranches (Jo76). The maximum
radiation level observed during these surveys was 250 mrad per hour (beta plus
gamma) at contact with a box containing 40 radium luminized D.C. ammeters
(Re76). Measurements at about six feet, while detectable with the instruments
used, were insignificant when compared with the normal background of 10 to 15 uR
per hour. This was true for all radiation levels observed.
Although direct radiation did not represent a hazard, the potential for
internal exposure did appear to exist. The type of items identified and
recovered during the surveys established the possibility for ingestion of
radioactive materials due, in some cases, to direct and continuing contact with
members of the public. While it was generally true that the contaminated items
recovered did not appear to present a significant hazard, there were exceptions.
During the surveys, an alpha contaminated, laboratory type, triple beam balance
and weights were located in a private residence. The balance and weights were
used as decorations which minimized the potential for exposure to the
contaminant which was subsequently identified as plutonium-239. Swipe analysis
identified as readily removable 3.8 x 10~ uCi of plutonium-239 and 3.8 x 10~
uCi of carbon-14 and lesser quantities of uranium-235, 238 and radium-226.
Investigation identified the waste disposal facility as the original local
source of the balance. At another residence, 74 miles from Beatty in Pahrump,
Nevada, a stainless steel balance pan was found. The pan was in daily use by
a former disposal site employee as a catch-all for keys, change and other
small items. An open-window GM survey of the pan indicated 70 mrad per hour,
beta plus gamma,.on contact with the pan. Analysis of a swipe of the pan
showed 1.3 x 10 uCi of strontium-90 and lesser quantities of cesium-134,
137, europium-154, 155 and cobalt-60. Two other items recovered at one
ranch were of special interest. One was a hoist and support structure which
was identified by NRC inspectors as having come from a source fabrication
facility which had experienced a large spill of plutonium-238. The hoist had
been used during the decontamination of that facility. Swipes of the hoist
showed 2 x 107 uCi of plutonium-238 or americium-241; 10 uCi of cesium-137;
and 4.5 x 10 uCi of plutonium-239. The other item was a disassembled metal
room, the door still showing the radiation symbol and the words "Anti C's
Required Beyond This Point" (Re76>5 The door and walls indicated up to 3 mrad
per hour, beta plus gamma, and 10 uCi beta and 10 uCi alpha on swipes.
Approximately 400 swipe samples were analyzed by REECo. Only swipes
which indicated contamination in the field were submitted for analysis.
Nuclides identified included plutonium-238 and 239, uranium-235 and 238,
neptunium-239, radium-226, strontium-89 and 90, cobalt-60, cerium-139 and 144,
europium-154 and 155, ruthenium-rubidium-106, cesium-134 and 137 and
antimony-125. The samples were collected at approximately 30 different
locations; however, approximately 280 samples came from the five locations at
which most of the contaminated items were found.
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The previous paper, "Response to a Widespread, Unauthorized Dispersal of
Radioactive Waste in the Public Domain," (We79), briefly discussed "Bennett
Buckets" and mentioned their routine diversion from the site. Out of the 134
that were estimated to have been diverted only 22% were found, half of them
on one ranch. Not all of these containers were returned to the disposal site.
If the containers in use were uncontaminated and the present owner wished,
they were left in place. It is believed that many of the missing containers
were used as septic tanks and were therefore effectively removed from contact
with the public. Of particular concern, were two "Bennett Buckets" used as
water tanks at residences, one as a potable water supply and the other for
showers and watering pets. Samples of water from these two "Bennett Buckets"
were analyzed by the EPA Las Vegas laboratory which reported no plutonium,
tritium or gamma emitting nuclides in either sample (He76). Direct surveys
and swipes of the two water tanks showed only background activity (Re76) .
Information provided by the original shipper disclosed that some
"Bennett Buckets" had contained significant quantities of hazardous materials.
For example, three of the "Bennett Buckets" had contained, when originally
shipped, 600 millicuries of californium-252 and americium-241, 30 curies of
promethium-147, and 900 millicuries of plutonium-239, respectively. Original
container serial numbers and content records were available; however, most
of the "Bennett Buckets" recovered showed no identifying markings. It should
be noted that none of the recovered "Bennett Buckets" exhibited any
significant activity. However, the soil near the lip of one of the "Buckets"
contained a 350 uR per hour "hot spot." Analysis of the soil "hot spot"
identified cobalt-60, antimony-125, chromium-51 and silver-108m and 110m (He76).
During the investigation, it was learned that one of the major items
salvaged from waste shipped to the disposal site was plywood from disassembled
waste packing boxes. The salvaged plywood or portions of boxes had been used
fairly widely for construction or were stockpiled at various homes and
ranches. Some of the wood still retained labels or markings used to identify
radioactive materials (e.g., "LSA") or the name of the consignee. Most of the
wood was found to be at background levels; however, contact readings up to
3 mrad per hour were identified on some piled plywood. Monitors were alerted
to the use of plywood in construction and surveyed plywood wherever it was
observed. No significant use of contaminated plywood was found.
Many of the items recovered were military instruments, compasses and
clocks with radium luminized markings. These items were originally disposed
by the military services as a part of the military's radium removal programs.
The highest radiation level associated with such items was 250 mrad per hour
from the 40 D.C. ammeters previously mentioned. Ships' clocks or chronometers
were probably the most widely distributed single item of salvage. These
items, as well as any others that were found to be radioactive were returned
for disposal. The monitoring teams' efforts also identified several old
clocks which contained radium but which obviously had not come from the
disposal site. In several cases, the owners requested that the clocks be
disposed when informed that they were radioactive.
The NRC identified a total of eight representative individuals for whole
tody counting and bioassay sampling. The individuals were selected on the basis
°f the highest probability of internal exposure due to the possession of
contaminated articles or the use of "Bennett Buckets" for potable and nonpotable
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water supplies. Those selected included representatives of the families that
possessed the contaminated balance, the stainless steel pan, hoist and
contaminated disassembled "hot lab." Those selected also represented the
locations that were the origin of most of the contaminated swipe samples.
The State of Nevada identified seven present and former waste disposal
site employees for whom evaluation of internal deposition was desirable. The
analyses were performed by the Environmental Monitoring and Support
Laboratory (EM&SL), US EPA, Las Vegas, Nevada. The EM&SL had engaged in the
whole body counting and bioassay evaluation of residents from communities and
ranches surrounding the test site for a number of years, in support of test
site activities. Selected families and individuals from these localities have
been examined on a continuing basis. Coincidentally, three of the eight
individuals identified by NRC for evaluation were former or current
participants in the test site evaluation program.
Whole body counts were performed using two systems. One system consisted
of a single 11 inch diameter by 4 inch thick sodium iodide crystal coupled
to seven photomultiplier tubes. The detector output was supplied to 200
channels of a 400 channel TMC analyzer. This system had a sensitivity for
cesium-137 of approximately 5 pCi per kg. This system could identify
approximately 1 nCi of most gamma emitters.
EPA reported that for the 15 individuals counted cesium-137 burdens
ranged from less than 5 to 32 pCi per kg. The normal range for cesium-137
in area residents is from less than 5 to 50 pCi per kg with an average of
approximately 20 pCi per kg. No other gamma emitting nuclides were
identified (Ka76-77).
The second system, for plutonium-239 screening, used a cesium iodide-
sodium iodide phoswich (each segment is one inch by three inches in diameter)
coupled to a single photomultiplier. The signal was supplied to 200 channels
of the 400 channel TMC analyzer. The detector was held in the armpit and one
count was made on each side of the chest. The phoswich was used for screening
for bioassay since EM&SL had no method for evaluating and correcting for chest
wall thickness and had not established the sensitivity of the system. The
individuals counted were compared with counts of 100 EPA employees, none of
whom had known plutonium exposure. None of the 15 individuals indicated a
statistical probability of internal deposition (Ka76-77).
Urine samples were collected from the 15 individuals and analyzed for
tritium. Thirteen of those sampled showed tritium levels of 350 to 6400
pCi per liter. The two others had levels of 24 and 29 nCi per liter. When
resampled in several months, these two individuals showed tritium levels of
2100 and 2900 pCi per liter. It should be noted that the individuals with
increased tritium levels were current employees at the disposal site and had
been involved in the processing and disposal of large quantities of tritium
contaminated water from a reactor. Strontium-89 and 90 analysis was performed
on a urine sample obtained from the user of the stainless steel balance pan.
Results were less than the minimum detectable activity. Composite analysis
on the urine samples from all 15 individuals for plutonium-238 and 239 were
negative (Ka76-77)-
In summary, radioactive or contaminated items were recovered from
approximately 45 locations in Beatty or from surrounding ranches or other
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communities. The bulk of the recovered materials came from approximately
four or five locations, each of which was associated with a present or former
employee of the waste disposal site. The great majority of the recovered
items consisted of radium luminized military surplus equipment. Direct
radiation was excluded as a problem as a result of extensive surveys at many
locations. Whole body counting and bioassay provided reasonable assurance that
ineestion had not occurred in the population with the greatest exposure
otential. Based on consideration of the factors discussed in this paper, it
was concluded that there had been no significant effect on the population at
risk.
References
An76 Anspaugh L.R., 1976, Private Communication
He76 Hendricks D.W., and Fort Jr. C.W., 1976, "Radiation Survey in
Beatty Nevada and Surrounding Area (March 1976)," ORP/LV-76-1, Office of
Radiation Programs - Las Vegas Facility (Las Vegas, NV: U.S. Environmental
Protection Agency)
Ka76-77 KayM.E., 1976 and 1977, Private Communications
Jo76 Jobst J.E., 1976, "A Radiological Survey of Beatty, Nevada and
the Amargosa Valley," EG&G/Aerial Surveillance Department Report,
Las Vegas, NV
Re76 Reynolds Electrical and Engineering Co., Inc., 1976 "Radiation Survey
Beatty, Nevada," REECo Report, Mercury, Nevada
We79 Wenslawski F.A., and North H.S., 1979, "Response to a Widespread,
Unauthorized Dispersal of Radioactive Waste in the Public Domain
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AN ASSESSMENT OF THE ENVIRONMENTAL TRANSPORT OF RADIOIODINE IN THE
AIR-GRASS-COW-MILK PATHWAY USING REPORTED ENVIRONMENTAL MONITORING DATA
John C. Erb
U.S. Air Force Hospital/SGPH, Patrick Air Force Base, FL 32925
ABSTRACT - The environmental transport of radioiodine has been a much studied
and sometimes controversial subject. This paper presents an attempt to analyze
the air-grass-cow-milk pathway by using data routinely reported to the U.S.
Nuclear Regulatory Commission from the effluent and environmental monitoring
programs maintained at nuclear power stations. This data is used in currently
accepted transport models to make a comparison of computer predicted versus
actual measured values of the resultant concentration of 1-131 in milk samples.
Introduction
With the advent of nuclear power and its use to generate electrical energy,
there has been much concern about the effects of radioactivity on the environ-
ment and man. Thus, the U.S. Nuclear Regulatory Commission (NRC) has, in Title
10 of the Code of Federal Regulations, placed constraints on the amount of
radioactivity that can be routinely released by a nuclear-powered generating
station. Perhaps the most limiting pathway for radio-nuclides through the en-
vironment to man, is the air-grass-cow-milk path taken by radioiodine. Iodine
in man is concentrated in the thyroid gland, and thus exiguous amounts of radio-
iodine can give a disproportionately large dosage of radiation to this small
organ.
This report is an attempt to utilize routine environmental monitoring data
to assess the reasonableness of the U.S. Nuclear Regulatory Commission's models
for the environmental transport of radioiodine in the milk-food chain by com-
paring predicted versus actual measured levels of radioiodine in milk.
The method chosen to evaluate the NRC's predictive models was to identify
particular cases where radioactivity had been detected during the operation of
routine environmental monitoring programs conducted by the utilities which
operate the nuclear power stations. Once such cases had been found, the appli-
cable NRC models could be used to determine if there was a relationship between
the predicted levels of radioactivity and the actual measured levels. Because
of atmospheric testing of nuclear weapons, many instances where radioactivity
was detected had to be eliminated because the activity could not be directly
attributed to the operation of the nuclear plant.
Iodine-131 in milk was chosen as the measurement to consider in this re-
port because it offers several advantages over other environmental measure-
ments. First, the 1-131 can be readily measured at the end of the pathway,
just before man; furthermore, the analytical procedures for the detection and
measurement of 1-131 in milk are sensitive and usually yield good results.
Another advantage is the relatively long half-life of iodine-131 (8 days) as
compared to the time required to travel through the pathway. Iodine-133 and
iodine-135 are also released by nuclear plants, but their half-lives of 21
hours and 6.7 hours respectively, are not long enough to allow measureable
amounts of these isotopes to reach the milk stage of the pathway. In addition,
the iodine-131 pathway is recognized as a primary mode by which radioactivity
released to the environment can reach and affect man. The particular pathway
investigated was that of gaseous iodine-131 in the reactor-air-grass-cow-milk
chain.
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^g Meteorological and Cow-Milk Models
The modeling of the environmental transport of radioiodine is accomplished
in two separate parts. The first part consists of using the meteorological
data (i.e., wind speed, wind direction, atmospheric stability class, etc.)
during the period of concern to predict atmospheric dispersion and relative
deposition. That is, the model predicts the amount of radioiodine that will
be deposited on the ground at some point of interest due to the release of
iodine by a nuclear plant. The second part is the cow-milk portion of the
pathway which uses the relative deposition value predicted by the meteorologi-
cal model to determine the concentration of radioiodine in the milk of a cow
grazing on the contaminated pasture. A detailed discussion of the meteorolog-
ical and cow-milk pathway models follows.
The meteorological model used by the U.S. Nuclear Regulatory Commission
in its evaluation of the atmospheric transport and dispersion of radioactive
material is the XOQDOQ model and is entitled "Program for the Meteorological
Evaluation of Routine Effluent Releases at Nuclear Power Stations" (Ref. 5).
This model is of the constant-mean-wind-direction type, and uses "straight-line"
airflow to calculate average relative air concentrations (X/Q's) and average
relative depositions (D/Q's) for effluent material. The basic equation given
in the XOQDOQ mode'l for calculating the relative air concentration for elevated
releases is:
X~(x,k) = 2.032 • RFk(x)
ZDEPL.. (x) • DECi(x) • f,,v exp - i ^e (Eq. 1)
1Jk 3' ' I 12~QQJ
ij u± 0Vj(x)
where
= average effluent concentration normalized by
source strength at distance x and direction k;
^i = mid-point values of the ith wind speed class;
^VjW = vertical (z) spread of effluent at distance x for the jth
stability class;
fijk = Joint probability of the ith wind speed class, jth stability
class, and kth wind direction;
x = downwind distance from release point or building;
^e = effective plume height;
DECi(x) = reduction factor due to radioactive decay at distance x for the
ith wind speed class;
DEPLijk(x) = reduction factor due to plume depletion at distance x for the ith
wind speed class, jth stability class, and kth wind direction; and
RFk(x) . correction factor for air recirculation and stagnation at distance
x and kth wind direction.
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This equation requires that a three-way, joint distribution be calculated for
the wind speed, wind direction, and atmospheric stability class.
The wind speed can be distributed into a maximum of fourteen velocity
categories. Various strategies can be employed for determining the breakdown
of the categories. For example, the wind speed divisions could be equally
spaced (e.g., 2 mph, 4 mph, 6 mph, etc.), or each division could contain an
equal number of values. Also a finer breakdown of the lower wind speed cate-
gories could be used since the effluent does not travel as far in one increment
of time for the lower velocity winds.
Wind direction is distributed into the sixteen cardinal compass sectors of
22.5 degrees each. These divisions of wind direction are shown in Figure 1.
The meteorology is also distributed by atmospheric stability. In this
report the stability of the atmosphere was classified according to the tem-
perature lapse rate (i.e., the change in the ambient temperature of the air
with a change in altitude); specifically, Pasquill stability categories were
used (refer to Table 1).
The meteorological data, once distributed into the proper form, is used by
the XOQDOQ computer code, along with other necessary information, to calculate
the atmospheric dispersion and the subsequent deposition of the effluent.
TABLE 1
PASQUILL STABILITY CLASSES
Lapse Rate (°c/100m) Class Stability
< - 1.9 A Very unstable
-1.9 to -1.7 B Unstable
-1.7 to -1.5 C Slightly unstable
-1.5 to -0.5 D Neutral
-0.5 to +1.5 E Slightly stable
+1.5 to +4.0 F Stable
> 4.0 G Very stable
Other parameters taken into consideration by the XOQDOQ model are: terrain.
(allowance being made for effluent recirculation), height and velocity at which
effluent was released, height at which wind data was measured (for possible
extrapolation to release height), plume decay and depletion; as well as allow-
ing for possible building wake correction.
There are several inherent limitations in the XOQDOQ computer code, as
well as some imposed by the methodology of this study. The model can use mete-
orological data for a single station only, and thus cannot take into considera-
tion spatial variation in wind speed, direction, etc. Wind directions near the
border of two sectors are credited to only one and is added into the total
average of that sector; thus, any contribution of the effluent to the neigh-
boring sector is neglected. This effect could be enhanced where local topog-
raphy may bias the wind direction; for example, the wind may preferentially
blow in the direction of 75 degrees, and therefore be credited to the ENE sector,
and not to the E sector lying 3.75 degrees away. Also, the meteorological data
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FIGURE 1.
WIND DIRECTION DISTRIBUTION
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508
input to this model usually consists of hourly values that are calculated from
fifteen-minute averages, as opposed to the average being taken over the whole
hour.
As previously mentioned, the methodology of this study necessarily im-
poses some additional limitations to accuracy of the XOQDOQ model. With the
purpose of the study being to compare model predictions with actual measured
cases of 1-131 appearing in milk samples it was necessary to use only a short
time period of data to describe the meteorological conditions concomitant with
the milk sample. The XOQDOQ model, however, is structured primarily to calcu-
late annual averages of relative deposition and air concentration values . Thus
instead of the usual 8760 hourly units of data for a year, only approximately
a week's worth of data, which generally contains 168 hourly units, is used.
This tends to amplify the effect of any wind direction biasing that would
probably cause much less of an effect for a full year of data. In addition,
it should be emphasized that the effects of the wind blowing into a particular
sector were averaged over the whole sector.
Although the XOQDOQ computer code was not specifically designed to compute
effluent air concentration and deposition values for short time periods, it can
still be expected to yield reasonable results for this study.
After the radioiodine had been deposited on the pasture, it was assumed
that the animal would consume the contaminated grass, metabolize the iodine-131
and secrete it into the milk. The mathematical representation for this model
was taken from Section C.3.b of the NRC Regulatory Guide 1.109, "Calculation
of Annual Doses to Man from Routine Releases of Reactor Effluents for the Pur-
pose of Evaluating Compliance with 10 CFR Part 50, Appendix I" (Ref. A).
To calculate the amount of 1-131 appearing in the animal's milk from a
given amount of 1-131 deposited on the pasture grass, the following relation-
ship was used:
Cm(t) = (D/Q) (Q) (f „) (Qf ) (Fm) exp (-^t) (Eq. 2)
Yu
where
= Concentration in milk of 1-131 (pCi/,?), at time t;
D/Q = Normalized deposition on pasture grass for 1-131 (m~2) ;
Q = Amount of 1-131 released by the nuclear plant (pCi) ;
fs = Fraction of animal's daily intake of food which consists of
pasture grass (default = 1) ;
Qf = Animal's daily food intake (50 kg/day for cows, 6 kg/day for
goats ;
Fm = The average fraction of the animals daily intake of 1-131 which
appears in each liter of milk (0.006 days/liter for cows,
0.06 days/liter for goats);
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509
Yu = Crop yield — the amount of pasture grass per unit area
(0.75 kg/mz);
X = Radioactive decay constant for 1-131 (0.086 day -1);
t = Average transfer time between the intake of 1-131 and its
appearance in the milk (approximately six hours, normally
taken as zero), (Ref. 6).
The normalized deposition (D/Q) was calculated using the XOQDOQ model
for the location where the animal grazes. This normalized deposition was
then multiplied by the amount of 1-131 released (Q) , giving the amount of
iodine per unit area of pasture grass. The significance of this calculation
was that the total iodine released from the nuclear plant, for the week prior
to the occurrence of the 1-131 contaminated milk sample, was assumed to be
deposited on the pasture grass. No radioactive or weathering decay correc-
tions were considered in the deposition calculation. In the absence of site-
specific data, the values of fs, Qf, and Yu were assumed to have the values
indicated above. The values for the transfer factor (Fm) were taken from NRC
Regulatory Guide 1.109.
As with the meteorological model, there are several assumptions that
must be made which limit the accuracy of the cow-milk model. The principal
uncertainty of this model is the amount of 1-131 consumed by the cow. This
iodine value is based on estimates of food intake by the animal (Qf = 50 kg/
day for cows; 6 kg/day for goats), and the areal grass density (0.75 kg/m^);
both of these estimates are subject to wide variations.
How The Study Was Carried Out
As stated 'earlier, the purpose of this study was to attempt to utilize
the data gathered by the environmental monitoring programs at nuclear power
stations to evaluate the reasonableness of the NRC's predictive models for
the environmental transport of radioiodine. To effect this study, it was
necessary to identify cases where measurable quantities of iodine-131 appeared
in milk samples, and then obtain the pertinent meteorological and source term
data.
As part of the environmental monitoring program required of the licensees
by the U.S. Nuclear Regulatory Commission, periodic collection and analysis of
milk samples from nearby dairy farms is performed. Milk sampling is usually
done on a weekly basis during the pasturing season. (In the northern states
the pasturing season covers the period of May through September, whereas in
the southern states it may extend throughout the whole year.) Although the
major part of the analysis of the milk is for the detection and measurement
of iodine-131, some licensees analyze for radiostrontiums and gamma-emitting
nuclides. Because gamma spectroscopy techniques are not sensitive enough to
achieve the lower limits of detection required by the NRC for iodine-131 activ-
ity, radiochemistry procedures are utilized to concentrate the iodine-131 for
beta or beta-gamma coincidence counting. The results of the milk sample testing
are reported to the NRC in the licensees' periodic environmental reports.
Because boiling water reactors (BWR's) release effluents continuously, a
review of the environmental reports of all BWR's was performed. The criteria
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for identifying a specific case of interest was that the measured iodine-131
value in milk (pCi//) was larger than the associated error at the 95% confi-
dence level (2 ); however, in certain instances exceptions to this criterion
were made. After obtaining a list of cases, a determination was made as to
whether or not the source of the iodine-131 was fallout from the atmospheric
testing of nuclear weapons. Several "clear periods" were established when
the appearance of iodine-131 in the milk samples was attributable to the
operation of the nuclear plants, and not to the nuclear weapons testing. The
summer months of 1975, and the whole of 1976, were determined to be two such
"clear periods". Following these allowances, approximately thirty cases were
identified for further analysis.
Using The Data In The Models
Because of the large uncertainties inherent in the XOQDOQ computer code,
it is generally used to compute annual averages. Therefore, it should be
recognized that the use of the XOQDOQ model in this study, with the limited
data available, has significant deficiencies. However, the model provides a
basis for "normalizing" the evaluations of the cases, that is, all the cases
(with the exception of case #17 where a puff model was used) were analyzed by
the same procedure.
With the milk sampling having been done on a weekly schedule, and since
most of the cases were singular, that is, there was usually no iodine-131 de-
tected in the previous week's milk samples, the meteorological data for the week
prior to the iodine-131 containing milk sample was used to calculate the iodine-
131 concentrations. Furthermore, the iodine-131 source term measurements were
also generally made at weekly intervals. Thus, the normalized deposition
values predicted by the XOQDOQ model were based on a week of hourly meteoro-
logical data. The total iodine-131 released during the week was deposited at
the rate predicted by the XOQDOQ model at the point of interest. This depo-
sition was then used in the cow-milk pathway model to predict the amount of
iodine-131 content in the milk. The predicted iodine-131 value was then com-
pared with the actual measured value to determine the amount of agreement
between the two values.
Summary and Conclusions
The results of the analyses are sumarized in Table 2 and from these data
there is no simple and easily distinguishable correlation between the predicted
milk values and the actual measured values. However, even though the analyses
of the twenty-eight cases cannot either substantiate or invalidate the meteoro-
logical and/or cow-milk models used in the study, certain interesting observa-
tions can be made which may provide possible answers to anomalies previously
discussed in the iodine pathway literature, as well as some insights into the
processes involved.
If cases 10 through 16 are considered separately from the others, the models
values are in reasonable agreement with the measured levels. These predicted-to-
measured ratios are in the range of 0.11 to 17 (excluding case #12). During the
time periods associated with these cases, the reactor was operating normally
(56% to 94% full power) with no elevated release rates of iodine-131. Al£hough
these cases were unique because the milk samples were obtained from goats, and
not cows as in all the other cases, there is no evidence to indicate that'such
-------
511
TABLE 2
RESULTS OF THE ANALYSES
CASE #
1
2
3
1*
5
6
7
8
9
10
11
12
13
14
15
16
17
18
19
20
21
22
23
24
25
26
2?
28
MILK VALUES (pCi/1)
j^KfiuiTTTEIT
0.038
26.7
3-7
0
1.28
0.15
58
39
430
0.26
0.79
0
0.6l
0.24
18.2
4.8
1.7
85
0
120
20
36
246
20
^7
120
69
16
1VUSASUK.ED
1.89+0.30
5.26+0.84
1.47+0.24
0.78+0.38
1.08+0.19
2.61+0.29
1.70+0.14
0.04+0.11
2.99+0.20
2.3
0.1
0.53
1.40
0.18
3.*
0.29
10+1
4 . 7±0 . 2
0 . 58+0 . 1 9
0.97±0.12
2.8+0.2
1 . 2+0 . 2
0.70+0.17
0.68+0.16
0.81+0.16
4 . 2+0 . 2
0.64+0.16
1 . 4+0 . 2
PREDICTED TO
MEASURED RATIO
0.02
5.1
2.5
0
1.2
0.06
34
970
143
0.11
7-9
0
0.44
1-3
5.4
17
0.17
18
0
124
7-1
30
350
29
58
29
108
11
PRECIPITATION*
1-75+
0.59+
1.50
1.09+
1.09+
1.09+
0.91
0.91
0.91
0.79+
0.79+
0
0.38
0.38
0.26+
0.26+
0.15+
0.52+
0.52
0.52+
0.42
0.42+
0.42+
0.42
0.38+
0.38
0.38+
0.38+
* TOTAL INCHES FOR THE WEEK PRIOR TO THE SAMPLE
"+" INDICATES WASHOUT WAS POSSIBLY A FACTOR IN DEPOSITION
-------
512
a difference is significant once the food intake and lactation parameters are
adjusted appropriately.
For cases 18 through 28 the predicted values were very high, having an
average predicted-to-measured ratio of seventy. The major portion of the
food intake of the cows at the local dairy farms was from stored feed and
not pasture grass since the cows were pastured for exercise only and on an
irregular basis. This would provide part of the reason for the high ratios,
but was probably not the entire cause. (Note: the food intake parameters,
fs and Qf, of Equation 2 were not adjusted from their normal values of one and
fifty respectively.
Also, it can be noted from Table 2 that high values for the predicted-to-
measured ratios were usually associated with high release rates. In thirteen
cases (7, 8, 9, 18, and 20 through 28) high ratio values were concomitant with
high release rates of iodine-131, although the cause-effect relationship was
not obvious.
As discussed earlier, the effects of rain on deposition are not consid-
ered in the meteorological model used in this study. However, whereas pre-
cipitation may occur for a very small fraction of a period of interest, pre-
cipitation scavenging appeared to be one of the more critical factors in
determining whether or not radioiodine was detected in milk. In only one of
the twenty-eight cases analyzed there was no rainfall associated with the milk
sample. In seventeen cases washout by rain seemed to be a major cause. (These
cases are indicated by a "+" on Table 2). This determination was made by con-
sidering the wind's direction during the rainfall. If the wind was blowing
in the general direction of the farm (within 100°), while the rain was falling,
then washout was considered to be a probable major factor in deposition. The
effect of washout could be to either increase or decrease the amount of deposi-
tion at the dairy farm. If the washout occurred before the effluent reached
the farm, the amount of iodine-131 consumed by the cows (or goats) would be
significantly less than if the washout occurred directly over the farm.
In cases one and seventeen the washout effect was particularly critical.
For case #1, 0.38 inches of rain fell while the wind was blowing toward the
dairy farm involved. Another 1.07 inches of rain occurred while the wind was
blowing in the approximate direction of the farm (circa 30° relative to the
location of the farm). Washout also was a major factor in case No. 17. The
scenario for this case was 0.15 inches of rain concomitant with a wind in the
general direction of the farm (circa 50° relative to the farm). The amount of
iodine-131, 10+1 pCi^C, measured in the milk sample would have undoubtedly
been significantly less if not for the effect of precipitation scavenging.
The analyses of the twenty-eight cases provided a basis on which several
general observations were made: 1) no good correlation was found between the
predicted and measured values; 2) short term, high iodine-131 effluent release
rates associated with reactor refueling did not significantly affect the radio-
iodine concentration in milk at local dairies unless there were mitigating cir-
cumstances such as rain; 3) precipitation scavenging appeared to have played
a major role in the deposition of the iodine-131 in the air-grass portion of
the pathway.
These observations indicate that precipitation scavenging is a primary
mechanism in the environmental transport of radioiodine and should be considered
-------
513
nv modeling of the air-grass-cow-milk pathway. However, it should be noted
h t precipitation scavenging is a function of many factors, such as: the
mint of rain, the size of the droplets, cloud height, temperature, etc.
These and other factors are discussed fully in Meteorology and Atomic Energy
(Kef. 2).
The results of this study have strongly supported the feasibility of
ine data from effluent release and environmental monitoring programs for
"nalyzing environmental transport mechanisms for radionuclides, and further
studies should be, accomplished using this available data.
-------
514
REFERENCES
1. F. 0. Hoffman, "Environmental Variables Involved with the Estimation of
the Amount of 1-131 in Milk and the Subsequent Dose to the Thyroid",
Institute fur Reaktorsicherheit, Cologne, West Germany, IRS-W-6, June
1973.
2. Meteorology and Atomic Energy - 1968, edited by D. H. Slade, U.S. Atomic
Energy Commission, June 1968.
3. Regulatory Guide 1.111, "Methods for Estimating Atmospheric Transport
and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-
Cooled Reactors", U.S. Nuclear Regulatory Commission, Washington, D.C.,
July 1977.
4. Regulatory Guide 1.109, "Calculation of Annual Doses to Man from Routine
Releases of Reactor Effluent for the Purpose of Evaluating Compliance with
10 CFR Part 50, Appendix I", U.S. Nuclear Regulatory Commission,
Washington, D.C., October 1977.
5. J. Sagendorf and J. Goll, "XOQDOQ - Program for the Meteorological Evalu-
ation of Routine Effluent Releases at Nuclear Power Stations", Draft,
U.S. Nuclear Regulatory Commission, Washington, D.C., 1976,
6. Radioactivity and Human Diet, edited by R. Scott Russell, Pergamon Press,
Oxford, England, 1966.
7. "Title 10 of the Code of Federal Regulations", Office of the Federal
Register, National Archives and Records Service, General Services Adminis-
tration, Washington, D.C., January 1977.
8. Y.C. Ng, et al., "Transfer Coefficients for the Prediction of the Dose to
Man via the Forage - Cow-Milk Pathway from Radionuclides Released to the
Biosphere, Lawrence Livermore Laboratory, University of California,
Livermore, California, July 1977.
9. Environmental Surveillance in the Vicinity of Nuclear Facilities, edited by
W. C. Reinig, Charles C. Thomas Publishers, Springfield, Illinois, 1970.
-------
515
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Dev-Yu Hsia
Institute of Nuclear Eng.
119 Harvard St. # 8
Cambrdige, MASS. 02139
Cheng Hung
USEPA
401 M. Street
Washington, D.C.
Alun James
Ontario Government
Apt. 325 80 St. Patrick St.
Toronto, Ontario,Canada
Johnny D. James
University of Oklahoma
1207 Columbia Ct.
Norman, OKLA. 73071
J.P. Jarrell
James F. MacLaren Ltd.
435 McNicoll Ave.
Willowdale, Ontario,Canada M2H2R
Jenny M. Johansen
University of Delaware
417 Academy St.
Newark Delaware 19711
James E. Johnson
Colorado State University
Dept. Rad. Biology
Fort Collins, CO. 80523
Robert U. Johnson
Harvard University
EH&S 75 Mt. Auburn St.
Cambridge, MASS. 02138
Edward Johnston
USF Medical School
8606 Leeward Dr.
Tampa, Fla. 33614
Alan Jones
Nuclear Diagnostics Lab.
Peekskill, N.Y.
-------
524
Charles C. Jones
United Nuclear Inc.
714 S. Conway
Kennewick, WA. 99336
David Barrett Jorgensen
Duke University
2711 Oberlin Drive
Durham, North Carolina 27705
Jack-H-N-Jow
Yankee Atomics
20 Turnpike Rd.
Westboro, MASS. 01581
Thomas L. Junod
NASA Lewis Research Center
320 Bogart Road
Huron, Ohio 44839
Arthur Lewis Kaplan
General Electric Company
FOB 780
Wilmington, NC 28401
Jacob Kastner
USNRC
Washington, DC 20555
Elizabeth P. Katsikis
Burroughs Wellcome Co.
3030 Cornwallis Road
Research Trinagle Park
North Carolina 27709
Alan J. Kawaters
R.S. Landauer, Jr.&co.
Glenwood Science Park
Glenwood, 111. 60425
Stephen V. Kaye
Oak Ridge National Lab.
POB X
Oak Ridge, Tenn. 37830
E.W. Kendall
Reynolds Electrical & Eng.
POB 14400
Las Vegas, Nevada 89114
William E. Kennedy, Jr.
Battelle Northwest
POB 999
Richland, WA. 99352
Edward R. Kerr
RAD Services Inc.
POB 599
Laurel, MD. 20810
Wayne Kerr
USNRC
Office of State Programs
Washington, D.C. 20555
Arlene Kibbey
Oak Ridge National
POB X
Oak Ridge, Tenn. 37830
Stephen M. Kim
Radiation Managment Corp.
3508 Market St.
Philadelphia, PA. 19104
Hazel Kimmel
Wyeth Labs. Inc.
POB 8299
Philadelphia, PA. 19101
Caleb Kincaid
FDA-BRH
Rockville, MD
20857
Lynn J. Kirby
Pacific Northwest Laboratory
POB 999
Richland, WA. 99352
Nancy Kirner
State of Washington
4425 Green Cove N.W.
Olympia, WA. 98502
William P. Kirk
EPA Health Effects Lab
Research Triangle Park
North Carolina 27711
Walter J. Knapp
Philadelphia Electric Co.
2301 Market St.
Philadelphia, PA. 19101
Conrad M. Knight
Duke University
Box 3155
Durham, North Carolina 27710
George Kniazewycz
Tera Corp.
2150 Southwick Ave.
Berkeley, CA. 94794
Harold Kohn
Ohio Power Siting Commission
361 East Broad Ct.
Columbus, Ohio 43215
-------
525
Steve C. Kouba
Mason & Hanger
FOB 30020
Amarillo, Texas 79110
Barry Kreiling
University of Connecticut
Radiation Safety Office
Storrs, Connecticut 06268
Donald J. Kvam
Lawrence Livermore Laboratory
FOB 5505
Livermore, CA. 94550
Thomas Labenski
New England Nuclear
601 Treble Cove. Rd.
N. Billerica, MASS. 01862
Robert Ladd
Automation Industries
2361 Jefferson Davis HW
Arlington, Virginia 22202
Paul Lamberger
Monsanto Research Corp.
Mound Lab FOB 32
Miamisburg, Ohio 45342
Edward Landa
U.S. Geological Survey
Reston, VA. 22092
Harold Vincent Larson
Battelle Northwest
904 Cottonwood
Richland, WA. 99352
L. Todd Leasia
Northwestern University
303 E. Chicago Ave. Ward Bldg.
Chicago, 111. 60611
Philip K. Lee
University of Missouri
Health Physics Services
Columbia, MO 65211
David Leigh
NL Industries Inc.
2414 Village of Pennbrook
Levittown, PA. 19054
Jack Lentsch
Portland General Electric
121 SW Salmon
Portland, Oregon 97204
Ronald E. Lerch
Westinghouse Hanford
FOB 1970 W/C - 35
Richland, WA 99352
Thomas B. Lewis
Vacco Industries
140 Canterbury Rd.
Mt. Laurel, N.J. 08054
Leon Leventhal
LFE ENV ANAL LABS
2030 Wright Ave.
Richmond, Calif. 94804
Sam Levin
Mass. Institute of Tech.
MIT Room 20B-238
Cambridge, MA 02139
Joseph A. Lieberman
Nuclear Safety Associates
5101 River Road
Bethesda, MD 20016
Gert Linderoth
Swedish State Power Board
JAMTLANDSGATAN 99
S-16287 VALLINGBY
SWEDEN
Craig Little
Oak Ridge National Lab.
FOB X
Oak Ridge, TN 37830
Thomas P- Loftus
National Bureau of Standards
Rockville, MD 20853
James A. Lonergan
SAI
FOB 2351
La Jolla, California 92038
Walter R. Lorenz
Burns & Roe Inc.
690 Kinderkamack Rd.
Oradell, New Jersey 07649
Phillip M. Lorio
Columbia University
520 W. 120th. St.
New York, New York 10027
Gene Loud
Tri-State Motor Transit Co.
Rt. 3 Box 305
North East, ND 21901
-------
526
Paul Clark Lovendale
Fermi National Accel.
FOB 500
Batavia, 111. 60510
Lab.
Paul MacBeth
Ford, Bacon, & Davis
POB 8009
Salt Lake City, Utah 84109
Teresa Ann Mack
Minnesota Dept. of Health
717 S.E. Delaware
St.Paul, Minn. 55440
William J. Madia
Battelle ONWI
505 King Ave.
Columbus, Ohio 43210
Gail Magenis
F.X. Masse Associates Inc.
POB 95
Middleton, MA. 01949
Gerald Maestas
Los Alamos Sci. Lab.
POB 1663 MS 517
Los Alamos, NM 87545
John Mahon
Schering=Plough Corp.
60 Orange Street
Bloomfield, N.J. 07003
Adam H. Malik
Union Carbide Corp.
61 E. Park Dr.
Tonawanda, N.Y. 14150
Charles W. Mallory
Hittman Nuclear & Devel.
9190 Red Branch Rd.
Columbia, Maryland 21045
Marjorie Malmberg
NUS Corp. Consulting
3300 Gregg Rd.
Brookeville, MD. 20729
Thomas James Maloney
ISO-TEX
Box 909
Friends Wood, Texas 77546
Philip Manly
Gamma Corporation
POB 430
Wahiawa HI 96786
Sally Mann
USDOE
Chicago Operations
9800 South Cass Ave.
Argonne, Illinois 60439
Doyle Markham
USDOE
550 2nd. St.
Idaho Falls, Idaho 83401
George A. Marquardt
Houston Lighting & Power Co.
Box 1700-South Texas Project
Houston, Texas 77001
Don W. Marshall
EG&G Idaho, Inc.
POB 1625
Idaho Falls, ID 83401
Randy W. Marshall
Tenn. Valley Authority
400 Commerce Ave.
Knoxville, Tenn. 37902
Frank Masse
MIT BATES LINAE
Maple Street
Middleton, Mass. 01949
Joseph Massey
Chem-Nuclear Systems, Inc.
POB 726
Barnwell, S.C. 29812
Ronald D. Maxson
Office Code 105
Noffolk Naval Shipyard
Portsmouth, VA. 23709
Cynthia Mayes
Argonne National Laboratory
Health Physics 9700 S. Cass Ave.
Argonne, 111. 60439
Robert Leslie Mayton
Carolina Power & Light
POB 1551
Raleigh, N.C. 27602
Wilson C. McArthur
Tera Corp.
2150 Southwick Ave.
Berkeley, CA. 94794
Margaret McCampbell
660 W. Redwood Street
Baltimore, Maryland
Room 105 Howard Hall
-------
527
Richard N. McGrath
1000 Prospect Hill Road
Dept. 9487-423
Wendsor, CT. 06095
Milton E. McLain
University of Arkansas
Mechanical Engineering
Fayetteville, AR. 72701
James E. McLaughlin
USDOE
376 Hudson Street
New York, New York 10014
Robert C. McMillan
USA MERDC
DRDME-VR
Ft. Belvoir, VA. 22060
Charles B. Meinhold
Brookhaven National Lab.
Upton, New York 11719
Lewis Meyer USEPA
Office of Radiation Programs
Washington, DC 20460
Sheldon Meyers
USDOE
20 Mass. Ave.
Washington, D.C.
Sylvester Beyers
YPI & SU
1403 Valleyview
Blacksburg, VA. 24060
Douglas L. Michlink
Tenn. Valley Authority
400 Commerce Ave.
Knoxville, Ten.. 37902
C.H. Miller
OREL
POB X
Oak Ridge, TENN. 37830
David W. Miller
Sargetn & Lundy Engineers
55 E. Monroe St.
Chicago, Illinois 60603
Frank P. Miller
VEPCO
POB 402
Mineral, Virginia 21137
Henry T. Miller
Gulf Oil Corp.
POB 3240
Pittsburgh, PA. 15238
William A. Mills
Office of Radiation Programs
401 M. St. SW
Washington, DC 20460
Steve N. Millspaugh
Georgia Tech
900 Atlantic DR. N.W.
Atlanta, Georgia 30318
Barry Mingst
USNRC
12303 Charles Rd.
Wheaton, MD. 20906
Ronald Mitchell
Rockwell International
POB 800
Richland, WA. 99352
A. Alan Moghissi
USEPA
401 M. St. S.W.
Washington, D.C. 20460
Michael Momeni
Argonne National Lab.
Argonne, 111. 60439
Mary Moore
Cooper Medical Center
Radiological Physics
Camden, New Jersey 08103
Patrick J. Moore
Pima Community College
7321 E. Calle Managua
Tuscon, Arizona 83710
John George Morand
UNC
B-5 Venable Hall 045
Chapel Hill, N.C. 27514
William Morgan
North Carolina State U.
2609 Cataline Drive
Raleigh, North Carolina 27607
Richard A. Moyer
Dupont
1207 Eisenhower Dr.
Augusta, Georgia 30904
-------
Eric Muller
Dept. for Ent.
EPS-Nuclear Programs
Ottawa,Ontario,Canada KIAiCS
Robert L. Mundis
Argonne National Lab.
9700 S. Cass Ave.
Argonne, 111. 60439
Laurence Munnikhuysen
Newport News Shipbuilding
4101 Washington Ave.
Newport News, Virginia
Glenn Lee Murphy
University of Georgia
110 Riverbend Road
Athens, Georgia 30602
Robert 0. Murphy
The Aston Company
1800 Montreal Circle
Tucker, Georgia 30084
Michael Musachio
NIH
Bethesda, MD.
Lowell Muse
University of Georgia
Public Safety Bldg.
Athens, GA. 30602
Charles S. Myser
Ohio State University
410 W. 10th. Ave.
Columbus, Ohio 43210
J.R. Naidu
Brookhaven National Lab.
Safety & Environmental
Upton, New York 11973
Dan Victor Neagu
Becktel Power Corp.
1421 Patwood Dr.
La Habra, CA. 90631
Thomas P. Neal
Palisades Nuclear Plant
Rt. # 1
South Haven, MI 49090
James N. Neel
Nuclear Engineering
FOB 7246
Lousiville, KY 40207
Robert H. Neill
New Mexico Health
POB 968
Santa Fe, New Mexico 87503
Seth Nelson
Safety & Supply Co.
5510 East Marginal Way South
Seattle, Washington, 98134
Regis M. Nicoll
Tenn. Valley Authority
Rt. 3 Box 86-A
Tuscumbia, AL. 35674
Lawrence H. Norris
Phillips Petroleum
82 F-TRC
Bartlesville, OK. 74004
Harry S. North
USNRC Region V
97 Sonora Way
Corte Madera, CALIF. 94925
Thomas Wyatt Oakes
ORNL
POB X
Oak Ridge, TN. 37830
Steven Gary Oberg
Utah State University
1077 Rose Street
Logan, UT 84321
James A. O'Brien
Bettis Atomic Power Lab.
POB 79
West Mifflin, PA. 15122
Fearghus O'Foghludha
Duke University
Durham, North Carolina 27710
Larry Richard Olden
Ontario Hydro
700 University Ave.
Toronot, Canada, Ontario
George Oliver
Carolina Power & Light
Box 1552
Raleigh, N.C. 27602
Robert P. Olson
Old Dominion University
333 Briarfield Drive
Chesapeake, VA. 23320
528
-------
529
Alan L. Orvis
Mayo Clinic
Radiation Control
Rochester, MINN. 55901
Richard V. Osborne
Chalk River Nuclear Labs.
Atomic Energy of Canada
Chalk River, Ontario,Canada
Jacques Ovadia
Michael Reese Hospital
Chicago, 111. 60616
Oktay Oztunali
Dames & Moore
20 Haarlem Ave.
White Plains, NY 10603
Claire C. Palmiter
President, IRPA
714 University Blvd. W.
Silver Spring, MD. 20901
Garris Dudley Parker
C.I.I.T.
POB 12137
Research Triangel Park
North Carolina, 27709
Anselmo Salles Paschoa
PUC/RJ, Physics Dept.
Rua Marques des Vicente
Rio de Janeiro, RJ 22453
Brazil
John W. Peel
USDOE
Washington, D.C. 20545
Nelson Perry
University of So. Alabama
MSB Room 2144
Mobile, Al. 36688
William Larry Petcpvic
Radiation Management Corp.
6437 Oaken Door
Columbia, MD. 11045
Charles B. Peteler
Safety & Supply Company
5510 East Marginal Way South
Seattle, Washington 98134
R.W. Peterson
Battelle ONWI
505 King Ave.
Columbus, Ohio 43210
Charles R. Phillips
EPA
POB 3009
Montgomery, AL. 36106
John R. Polli
Newport News Industrial
230 41st. St.
Newport News,, VA. 23607
T. Jordan Powell
Lawrence Livermore Lab.
POB 5505
Livermore, CA. 94550
Charles R. Price
VA. Dept. of Health
109 Governor St.
Richmond, VA. 23219
Keith R. Price
Battelle Northwest
Tichland, WA. 99352
Robert J. Prince
Dairyland Power
POB 135
Genoa, WIS. 54632
Dennis M. Quinn
Power Authority of N.Y.
POB 215 Indian Point
Buchanan, NY 10511
Bro. Jerome Rademacher
St. Mary's College
Physics Dept.
Winona, MINN. 55987
Glen A. Rae
Chem-Nuclear Systems, Inc.
POB 726
Barnwell, S.C. 29812
Theodore E. Rahon
NL Industries
1130 Central Avenue
Albany, New York 12205
P.O. Randolph
EG&G Idaho
POB 1625
Idaho Falls, Idaho 83401
Louis E. Reynolds
Chem-Nuclear Systems
POB 1866
Bellevue, WA. 98009
-------
530
John Richardson
Ontario Hydro
BNPDS Box 1540
Iverton,Ontario,Canada
John H. Riley
Charleston Naval Shipyard
Radiological Control Off.
Charleston, S.C. 29408
Kenneth Brown Ritchie
UNC
B-5 Venable Hall
Chapel Hill, N.C.
Barney Roberts
Southwest Nuclear Co.
FOB 43046
Lousiville, KY 40243
Carlile J. Roberts
Argonne National Lab.
EIS Division Nldg. 10
Argonne, 111. 60439
Charles E. Roessler
University of Florida
114 BLK
Gainesville, FL. 32611
Genevieve S. Roessler
University of Florida
237 NSC
Gainesville, FL. 32611
Gilbert Rosenberger, Jr.
Babcok & Wilcox
FOB 785
Lynchburg, VA. 24505
Leon Rothman
Columbia University
520 W. 120th. St.
New York, New York 10027
Gary Warren Rowlett
Ark. Power & Light
900 Center Street
Little Rock, Ark. 72203
Romas Leonard Rupinskas
Sargetn & Lundy Eng.
55 East Monroe
Chicago, 111. 60603
John L. Russell
USEPA
400 M. St. S.W.
ANR-459
Washington, D.C. 20460
Andrew T. Sabo
Westinghouse Elec. Co.
Box 355
Pittsburgh, PA. 15230
Jean St. Germain
Memorial Sloan Rettering
1275 York Ave.
New York, New York 10021
Thomas J. St. Jean
Medical College of VA
FOB 112
Richmond, VA 23298
Bradley J. Salmonson
Exxon Minerals Co.
FOB 2180
Houston, Texas 77001
Ronald Sanacore
American Nuclear
Farmington Avenue
Farmington, CT. 06032
Philip Sandel
Texas A&M University
Radiological Safety Office
College Station, TEXAS 77843
Lorion J. Sanders
University of Illinois
343 McKinley Hospital
Urbana, 111. 61801
Keith Jerome Schiager
ALARA, Inc.
FOB 860
Lyons, Colorado 80540
Rudolph Max Schletter
EG&G Idaho Inc.
BOX 162S
Idaho Falls, Idaho 83401
Charles T. Schmidt
Lawrence Berkley Lab.
Bldg. 26
Berkeley, CA. 94720
Ronald Charles Schrotke
Battelle-Northwest
Battelle Blvd.
Richand, WA. 99352
Jacob Sedlet
Argonne National Lab.
9700 S. Cass Ave.
Argonne, 111. 60439
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531
Jack M. Selby
Battelle
Box 999
Richland, WA. 99352
Kenneth E. Shank
ORNL
FOB X
Oak Ridge, TN. 37830
Heyward Glenn Shealy
SC DHEC
2600 Bull St.
Columbia, South Carolina
Dillard Shipler
Battelle ONWI
505 King Ave.
Columbus, Ohio 43201
Jamieson C. Shotts
University of Missouri
Health Physics Services
Columbia, MD. 65211
Jay S. Silhanek
EPA
401 M. St. S.W.
Washington, D.C. 20460
Dale L. Sillyman
Hittman Nuclear
9190 Red Branch Road
Columbia, Maryland 21045
John D. Simchuk
Nuclear Packaging Inc.
1733 Fawcett Ave.
Tacoma, Washington 98402
John E. Simek
Texas A & M University
Radiological Safety
College Station, Texas 77843
Bernard Singer
USNRC
Washington, D.C. 20555
Gerald James Sinke
Kerr McGee Nuclear Corp.
Kerr McGee Center
Oklahoma City, OKLA. 73125
Glen Sjoblom
Naval Sea Systems Command
Washington, D.C.
Benjamin Sklar
Boston Edison Co.
800 Boylston St.
Boston, MASS. 02199
Lester A. Slaback
Armed Forces RRI, DNA
NNMC Bldg. 42
Bethesda, MD. 20014
William Lyle Slagle
Allied Chemical Corp.
550 2nd. St.
Idaho Falls, Idaho 83401
Benjamin Slone III
Newport News Shipyard
4100 Washington Ave.
Newport News, VA. 23601
Dale Smay
B & W
Warren Ave.
Apollo, PA. 15613
Leonard R. Smith
NENC
549 Albany St.
Boston, MASS. 02130
Leslie Charles Smith
Newport News Shipbuilding
4101 Washington Ave.
Newport News, VA. 23607
Walter H. Smith
Argonne National Lab.
9700 S. Cass Avenue
Argonne, ILL. 60439
William L. Smith
U.S. Geological Survey
Box 25046
Denver, Colorado 80225
John F. Sommers
EG&G Idaho
Box 1625
Idaho Falls, ID. 83401
Earl R. Sorom
Reynolds Electrical
POB 14400
Las Vegas, Nevado 89114
David L. Spate
University of Missouri
Research Park Devel.
Columbia, MD. 65211
-------
Daniel D. Sprau
East Carolina University
Radiation Safety Office
Greenville, N.C. 27834
Michael J. Steffensmeier
University of Nebraska
Room 128 501 N. 10th. St.
Lincoln, Nebraska 68588
James G. Steger
LASL
Environmental Sciences
Las Alamos, NM 87545
Fred J. Steinbrenner
Law Engineering Test Co.
2749 Delk Rd. S.E.
Marietta, GA. 30067
Dwane H. Stevens
Ludlum Measurements Inc.
501 Oak St. FOB 248
Sweetwater, Texas 79556
John E. Stewart
Werner & Pfleiderer
160 Hopper Ave.
Waldwick, NJ 07463
Jathan N. Stone
Naval Research Lab.
Washington, DC 20375
Jon Stoukey
NYS Corp.
#4 Research Place
Rockville, MD. 20850
Daniel J. Strom
Old Dominion University
Radiation Safety Office
Duckworth Hall 123
Norfolk, VA. 23508
Al Stueber
ORNL
Environmental Sci. Div.
Oak Ridge, TENN. 37830
Norman R. Sunderland
University of Missouri
413 Clark Hall
Columbia, Missouri 65211
Tsuneo Tamura
ORNL
FOB X
Oak Ridge, TENN. 37830
532
James T. Tanner
Food & Drum Administration
200 C. Street
Washington, D.C. 20204
Edmund C. Tarnuzzer
Yankee Atomic Electric Co.
20 Turnpike RD.
Westboro, MA. 01581
Richard D. Terry
Victoreen Inc.
10101 Woodland Ave.
Cleveland, Ohio 44104
Paul E. Theiss
Catholic University
POB 951
Washington, DC 20064
Bill R. Thomas
DOW Chemical
1803 Bldg. Indus^. HY. Lab.
Midland, MI. 48640
Ralph H. Thomas
University of California
2771 Doverton Square
Mountain View, CA. 94040
Walter Thomasson
USDOE
Germantown Maryland Office
Washington, DC 20545
Warren T. Thompson
Environmental Sci.
ORNL
Oak Ridge, Tenn. 37830
Bliss Tracy
Rad. Protec. Bureau
Brooklied Road
Ottawa, Ontario, Canada KIAICI
Richard J. Traub
North Dakota State University
College of Pharmacy
Fargo, N.D. 58105
Milt Trautman
Eberline
245 Roosevelt Rd.
West Chicago, 111. 60185
Rice Terrill Trolan
Lawrence Livermore Lab.
659 Park Hill Rd.
Danville, CA. 94526
-------
Keith Uhland
Dupont
132 Hitching Post Dr.
Wilmington, Del. 19803
Carl M. Unruh
Battelle Northwest
Box 999
Richland, WA. 99352
Edward J. Vallario
USDOE
Washington, DC. 20585
Robert Van Wyck
Con Edison
208 Radcliff Drive
Upper Nyack, N.Y. 10960
Dominic A. Versage
Cornell University
Radiation Safety Office
935 Warren Road
Ithaca, N.Y. 14850
John C. Villforth
Bureau of Rad. Health FDA
5600 Fishers Lane
Rockville, MD. 20350
E.L. Vinecour
Allied Nuclear Inc.
39187 Liberty St. Suite F
Fremond, CA. 94538
Milo Voss
Ames LAB.
DOE
3309 Ross Road
Ames, Iowa 50010
William W. Wadman III
University of California
FOB 4085
Irvine, CA. 92716
Lewis J. Walker
LA.SL
578 Todd Lane
Los Alamos, NM 87544
Andrew Wallo III
The Aerospace Corp.
20030 Century Blvd.
Germantown, MD. 20767
James Warden
TRI Nuclear Corp.
FOB 178
Ballston Lake, New York
John L. Warren
Los Alamos Sci. Lab.
FOB 1663
Los Alamos, NM 87545
Bruce A. Watson
Inst. for Resource Med.
7815 Old Georgetown Rd.
Bethesda, MD. 20014
Edwin C. Watson
Battelle Northwest
FOB 999
Richland, WA 99352
James E. Watson
UNC
517 Yorktown Dr.
Chapel Hill, NC 27514
Robert L. Watters
USDOE
Off. of Health & Env. Res.
Washington, D.C. 20545
Finley Clay Watts
Bowman Gray School of
Medicine
Winston Salem, N.C. 27103
Billy H. Webster
Carolina Power & Light
FOB 1551
Raleigh, N.C. 27602
Richard A. Weetman
United Nuclear Industries
FOB 490
Richland, WA. 99352
Walter F. Wegst, Jr.
Calif. Inst. of Tech.
1201 E. Cal. Blvd.
Pasadena, CA. 91125
George Wehmann
Ford, Bacon, & Davis
FOB 8009
Salt Lake City, Utah 84108
Richard A. Welch, Jr.
Catalytic Inc.
1500 Market St.
Philadelphia, PA. 19104
Frank A. Wenslawski
USNRC Region V
1990 N. Calif. Blvd.
Walnut Creek, Calif. 94596
533
-------
Robert V. Wheeler
R.S. Landauer, Jr. & Co.
Science Road
Glenwood, 111. 60425
John B. Whitsett
DDE-Idaho Operations
550 Second St.
Idaho Falls, Idaho 83401
D.H. Willard
Battelle Northwest
FOB 999
Richland, WA 99352
Donald G. Willhoit
UNC-CH
Health & Safety Office
B-5 Venable Hall 045A
Chaptel Hill, NC 27514
Edward F. Williams, Jr.
RADEF Instr. Test Facility
6626 Reynard Drive
Springfield, VA. 22152
Robert A. Williams
Westinghouse Electric
493 Longvue Drive
New Kensington,PA 15068
Bob Wilson
Univ. of Lousiville
FOB 35260
Louisville, KY. 40232
Clifford E. Winters, Jr.
Todd Research & Technical
FOB 1600
Galveston, Texas 77553
Robert G. Wissink
3 M Company
3 M Center Bldg. 220
St. Faul, MINN. 55101
Richard L. Woodruff
USNRC
101 Marietta St. N.W.
Atlanta, GA. 30075
McDonald E. Wrenn
NYU Medical Center
550 First Ave.
New York, New York 10016
Robert A. Wynveen
Argonne National Lab.
9700 S. Cass Ave.
Argonne, ILL. 60441
Kenyon D. Yoder
Miles Laboratories Inc.
1127 Myrtle St.
Elkhart, Indianna 46514
Melvin C. Young
U. of Arizona
Radiation Control Office
Tucson, Arizona 85724
Paul Ziemer
Purdue University
Bionucleonics Dept.
W. Lafayette, IN. 47907
SPEAKER
Margaret N. Maxey
University of Detroit
8801 Kingswood #201
Sherwood Heights
Detroit, Michigan 48221
534
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AUTHOR INDEX
535
Author
Adam, J.A.
Andersen, R.L.
Andrews, D.L.
Baptista, G.B.
Bard, S.T.
Beck, T.J.
Berk, H.W.
Bettenhausen, L.
Black, S.A.
Blackburn, D.A.
Bolch, W.E.
Buring, M.R.
Burrows, V.
Cadwell, L.L.
Cohen, N,
Cooley, L.R.
Crawford, D.J.
Crofford, W.N.
Desrossiers, A.E
Dickson, H.W.
Dornsife, W.P.
Ebenhack, D.G.
Eng, J.
Erb, J.C.
Essington, E.H.
Feldman, J.
Fisher, D.R.
Fitzner, R. E.
Fowler, E.B.
Fuhrman, D.R.
Giardina, P. A.
Gilchrist, J.R.
Godbee, H.W.
Goldsmith, W.A.
Granlund, R.W.
Gregory, W.D.
Gutwein. E.E.
Paper
Number
27
20
14
40
33
5
20
14
39
17
23
24
10
39
52
45
5
20
37
32
. 34
31
3
18
7
41
43
55
49
7
41
43
42
51
52
49
17
7
41
43
14
2
37
13
-
in
Page
Number
214
151
101
337
261
27
151
101
329
126
174
182
67
329
485
385
27
151
294
252
270
238
16
133
49
351
366
504
457
49
351
366
356
471
485
457
126
49
351
366
101
11
294
91
121
67
Author
Hayes, J.F.
Haywood, F.F.
Hendricks, D.W.
Hickey, J.W.N.
Hung, C.Y.
Hwang, S.L.
Paper
Number
13
37
7
30
33
11
Kennedy, W.E., Jr. 44
Kerr, G.W.
Kibbey, A.H.
Kirner, N.P.
Kisieleski, W.E.
Kitka, M.
Ledbetter, J.O,
Lee, D.A.
Lee, P.K.
Legget, R.W.
Lichtman, S.
Liebennan, J.A.
Macbeth, P.J.
Mariner, G.
Martin, J.
Maxey, M.N.
McArthur, W.C.
Me Campbell, M.
McPherson, R.B.
Meyers, G.L.
Meyers, S.
Mills, W.A.
Miranda, A.C.
Mitchell, R.M.
Moghissi, A. A.
Montenegro, E.G.
Momeni, M.H.
Mundis, R.
Neiheisel, J.
Njoku, E.
North, H.S., Jr.
Oakes, T.W.
Opelka, J.
Paine, D.
Paschoa, A.S.
28
2
23
38
6
22
25
15
19
37
36
1
26
6
36
46
21
5
20
44
33
8
35
40
48
23
40
38
6
33
34
53
54
47
6
48
40
Page
Number
91
294
49
234
261
74
373
219
11
174
307
38
168
196
107
141
284
284
1
203
38
284
400
159
27
151
373
261
62
280
337
442
174
337
307
38
261
270
492
497
421
38
442
337
-------
536
Author
Pasinosky, J.
Peterson, J.
Peterson, J.B
Pettengill, H
Polzer, W.L.
Price, K.R.
Richardson, J
Roberts, C.J.
Roessler, C.E
Rohlich, G.A.
Russell, J.L.
Schreckhise,
Shank, K.E.
Shealy, H.G.
Shotts, J.G.
Sigaud, G.M.
Siskind
Smith, Z.A.
Spate, D.L.
Stewart, J.E.
Thompson, W.T
Tsai, C.M.
Tyler, S.
Watson, E.G.
Wens laws ki, F
Wethington , J
Wrenn, M. E.
Yuan, Y.C.
Zielen, A.
Paper
Number
17
6
16
.J. 36
49
48
51
25
38
24
42
22
32
R.G. 52
47
29
15
19
40
6
24
15
19
12
22
11
38
44
.A. 53
54
.A. 24
45
38
38
Page
Number
126
38
116
284
457
442
471
196
307
182
356
168
252
485
421
231
107
141
337
38
182
107
141
82
168
74
307
373
492
497
182
385
307
307
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