United States
           Environmental Protection
           Office of
           Radiation Programs
           Washington DC 20460
EPA 520/3-79-002
May 1979
Low-Level Radioactive
Waste Management
           Proceedings of
           Health Physics Society
           Twelfth Midyear
           Topical Symposium

           February 11 -  15, 1979
           Williamsburg, Virginia


           FEBRUARY 11-15, 1979
               CO-HOSTED BY

              VIRGINIA CHAPTER

              JAMES E. WATSON, JR.

     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)

                 FOR 1978 - 1979

                   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

General Chairpersons

William P. Kirk
Sylvester L Meyers
Finance Committee

*John W. Cure III
*Elizabeth P. Katsikis
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


     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
                                                  James E.  Watson,  Jr.

                           TABLE OF CONTENTS

PREFACE                                                                  i

HEALTH PHYSICS SOCIETY ORGANIZATION FOR 1978 - 1979                     11

SYMPOSIUM COMMITTEES                                                   111

ACKNOWLEDGEMENTS                                                        iv

TABLE OF CONTENTS                                                        v



              J. A. Lieberman	     1


    CHAIRPERSON - R. J. Stouky	    10

              A. H. Kibbey and H. W. Godbee	    11


              W. P. Dornsife	    16

              W. F. Holcomb, R. L. Clark, and M. F. O'Connell 	     *

              T.J. Beck, M.R. McCampbell, L.R. Cooley	    27

              R.L. Mundis, M.J. Kikta, G.J. Manner, J.H. Opelka,
              J.M. Peterson, B. Siskind	    38

                                                                  1 >

              J.  Eng, J.  Feldman, and P.A.  Giardina	-    49


    CHAIRPERSON - W. C. McArthur	    61


              S.  Meyers	    62


              R.  L. Clark and W. F. Holcomb		    *


              M.R. Buring and E.E. Gutwein	    67

    METHOD, P/ll

              S.  L. Hwang and C. M. Tsal		    74


              J.  E. Stewart	-—    82

          R. W. Granlund and J.  F.  Hayes	—    91

          D.L. Andrews, J.R.  Gilchrist,  and H.W. Berk	   101

          J.G. Shotts, D.L.  Spate,  and P.K.  Lee			   107

          J. B. Peterson	-	—	-	-	   lie


               W. D. Gregory	-  121


     CHAIRPERSON - A.A. Moghlssi 	  125

     INVITED, P/17

               D.R, Fuhrman, S.A. Black and J.P.  Pasinosky	—  126

     INVITED, P/18

               D, G, Ebenhack	  133

     BURIAL SITE, P/19
               P.K. Lee, J.G. Shotts and D.L. Spate	—  141

               R. Andersen; T.J. Beck, L.R. Cooley and M.  McCampbell -  151

               W. C. McArthur	  159

               W.T. Thompson, J.O. Ledbetter, and G.A. Rohlich  	  168

               N.P. Kirner, A.A. Moghissi and P.A. Blackburn 	  174

               C.E. Roessler, Z.A. Smith, W.E. Bolch, and J.A.
               Wethington, Jr. 	  182

               J. Richardson and D. A. Lee	  196

    DISPOSAL, P/26
              P.O. Macbeth	~   203

    CHAIRPERSON - C. R. Price	   213

              J. A. Adam	   214
              G. W. Kerr	   219
              H. G. Shealy	-	   231
    APPROVAL, P/30
              J. W. N. Hickey	   234
              H. W. Dickson	   238
              J. L. Russell and W. N. Crofford	   252

              G.L. Meyer, S.T. Bard, C.Y. Hung and J. Neiheisel 	  261

              A.E. Desrosiers and E. Njoku				   270

              W.A. Mills	   280


              J,E. Martin, H.J. Pettengill and S. Lichtman	   284


    CHAIRPERSON - S. V. Kaye —	   293

              W.A. Goldsmith, D.J. Crawford, F.F. Haywood, and R.W.
              Leggett	   294

              M.H. Momeni, W.E. Kisieleski, S. Tyler, A. Zielen,
              Y. Yuan and C.J. Roberts	   307

    AREAS, P/39
              L. Bettenhausen and V. Burrows		   329
              A.S. Paschoa, G.B. Baptista, E.C. Montenegro, A.C.
              Miranda, and G.M. Sigaud	   337

              J. Feldman, J. Eng and P.A. Giardina -	   351


              D. R. Fisher and C. E. Roessler		---   356
              P.A. Giardina, J. Eng and J. Feldman	   366

              W.E. Kennedy, Jr., R.B. McPherson and E.C. Watson 	   373

              M. E. Wrenn and N.  Cohen		   385
              Margaret Maxey	—	   40°

    CHAIRPERSON - R. M. Fry	—   420

              T. W. Oakes and K. E. Shank	—   421

              D. Paine, K.R. Price and P.M. Mitchell 	   442

              E.H. Essington, E.B. Fowler and W.L.  Polzer	-   457

              W.J. Smith and A.F. Gallegos	-	    *
              K.R. Price and R. E. Fitzner	   471

              L.L. Cadwell, R.G. Schreckhise and R.E. Fitzner	   485

              F.A. Wenslawski and H.S.  North, Jr. 	   492

              H.S. North, Jr. and F.A.  Wenslawski 		   497


              J.  C. Erb	    504
              W.E. Bolch and W.  MacCready 	


*      Withdrawn

**     Accepted But Unable To Present In Person

***    Added

                             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

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

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,

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

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

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

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

     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.

     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
     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.

              SESSION A
         Session Chairperson
            R. J, Stouky
           NUS Corporation

                         A. H. Kibbey and H. W. Godbee
                         Chemical Technology Division
                         Oak Ridge National Laboratory
                         P. 0. Box X
                         Oak Ridge, Tennessee  37830


     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.


     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


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)  '


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,


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

     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

     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.


     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

     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.
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.


                            William P.  Dornsife
                      Bureau of Radiation Protection
            Pennsylvania Department of  Environmental Resources


     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.


     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.


     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


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).

      1 0
                   \   Hazardous Waste
                      U Tailings
         Natural ToxicVHeavy Metal Background
                        Low Level Radwaste (LLW)
         Natural Radium Background
       Figure 1:
102        103        10*        105        106


Relative Toxicity of A Typical Metric Ton of Hazardous
Versus Radioactive Waste


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

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)Assumed to be 0.!
                                                       Quantity Produced
                                                      (Metric Tons/Year -
                                                          Dry Weight)



x 106
x 102
x 105
x 106





                      uranium ore (Nu76)
Table 2:  Quantities of Lifetime Waste Generated by 1000 MW of Various
          Alternate Energy Sources
   Type of Waste and Energy Source
     Spent Fuel'
     Low-level radwaste^J'
     Uranium tailings
     Fly ash and bottom ash
     Scrubber sludge

                                              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
(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)


                            Low-Level Radwaste
                               TIME SINCE DISPOSAL,  YEARS
          Figure 2:   Relative Toxicity of the Total Quantity of Hazardous
                     Versus Radioactive Waste Produced in 1977


                            Low-Level Radwaste
                              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


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.


     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.


     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.

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
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.


                              GENERATED IN 1977

Thomas 3. Beck, Margaret R. McCampbell, Leland R. Cooley; University of Maryland
at Baltimore.

     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.

     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.

     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.


      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


            V = volume in m

            m= 639.74

            X = (year - 1900)

            b = - 41621.63

     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

     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

      •     Dry,  solids - syringes,  vials, test tubes  and other disposable labware,
            absorbent papers, gloves, etc.

      •     Adsorbed liquids - aqueous and organic liquids dispersed  in an adsorbent

      •     Liquid scintillation vials - full  scintillation vials packed  in adsorbent

      •     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)

     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,

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



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





             TABLE 3.  SEALED SOURCES
                     SHIPPED FOR BURIAL
                    BY STUDY POPULATION
          WASTE TYPE







*Percentof  those who disposed of radwaste in 1977
 (total N=292).

41 °/
 * 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.




m3  X  1000    6-




   < 7 DAYS  6-X-&&S21 7.8 %
   X '  fc^ *"t I %-*  ••••••••••••«   ^ "
7-90 DAYS
 > 9O DAYS
                                          I    I  Others



          R.  L. Mundis,  M.  J.  Kikta,  G.  J.  Marmer,  J.  H.  Opelka,
                         J. M. Peterson,  B. Siskind
                        Argonne National Laboratory
                          Argonne, Illinois  60439


     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.


     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


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

     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.

         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.


     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

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).

     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.


     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

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.

     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.


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.
     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.


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.


     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.

                                 TABLE 1
Accelerated Particle
Helium ions
All ions of light
atomic weight
Energy Range
1.67 - 10
Any Energy
Radioactivity In
None None
Limited Very slight
Probable Suspect
                                TABLE  2
Half Life
2.6 a
303 d
270 d
5.26 a
245 d
Produced In
Stainless steel
& copper
                                 TABLE 3



2.6 a
2.6 a
.4xl05 a
303 d
270 d
5.26 a
5.26 a
92 a


% of Saturation
for 25 Year
*Data from (Sw79)

                                     TABLE 4

Type of Accelerator
Cyclotrons <20" diam.
& "~ linacs
20" to 50" diam.
50 to 100" diam.
e~ linacs >_ 600'
Cyclotrons diam.
> 100"
Proton synchrotrons
60' - 200* diam.
Proton synchrotrons
> 200' diam.
proton linac
> 500' length
It of Machines
Estimated Mass
of Active Waste,
tons per
1 30
30 - 50
250 800
800 - 2300
2300 16,500
16,500 220,000
Volume ,
£ 60
260 2200
2200 - 7000
7000 - 20,000
20,000 - 140,000
140,000 - 2 * 106

     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)

1000 C
       Fig.  1. Mass Yield  Curves  for Protons on Bismuth
               Decay  Time  (yrs)
2.  Decay of Accelerator  Induced  Radioactivity

                          LOW-LEVEL RADIOACTIVE WASTE

        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


     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.


     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

     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,


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-


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.


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

     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

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.


     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.


     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

     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-


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.


     Co78  Costle, D., 1978, "Hazardous Waste Proposed Guidelines and Regula-
tions,  and Proposal on Identification and Listing," Federal Register, Vol. 43,

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

     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,

     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,

     Wa78   Wagstaff, D. G., 1978, "NORM -  Problems  in  Oregon," paper presented
at the Region X Radiation Control Meeting, September 26, 1978.

       TABLE 1:  Thorium And Uranium Composition In
          Monazite Concentrates (Weight Percent)
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

 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, OH, SC, DE,  CA,  PA





                                BEACH SANDS
      TITANIUM -      —









— | KAYAr


| GAR*-



— | CHRO

-| Go


[ 100,000
I 80,000



                                                                            r""  L
                                                                         ! Li

                 	  IMPORTS TO U.S.

                • —- U.S. CONSUMPTION
                                                   _  160,000

                                                   -  140,000

                                                   _  120,000
                o      ir>
                UD      1C

             SESSION B
        Session Chairperson
          W. C. McArthur
Hittman Nuclear & Development Corp.



                                Sheldon Meyers
         Office of Nuclear Waste Management, U.S. Department of Energy

     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


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

     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.


     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

     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

      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


          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

    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.


     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

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

     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

     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.



                                M. R. Buring
                         Metropolitan Edison Company
                         Reading, Pennsylvania 19605

                                E. E. Gutwein
                         Reading, Pennsylvania 1960S


     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.


     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

     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.


     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.


     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).

                                   Table 1
                       Nuclide Composition of Effluent
                         Releases  for 1976 and 1977
Co- 5 8
All Others

of Total    Initial Curies
                                  30. ly
Curies After
2 Year Decay


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
                              Curies  After
                              2  Year  Decay

      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.


     (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


     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.

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.


     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/

     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.


These figures do not include the liner surcharge which is currently

     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.


     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

            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


     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.

                   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

     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.

     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-

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
     Radio of volume reduction=   Volume of  feeding sludge
                                Volume of concentrated  sludge

Table 1.  The Comparisons of Different  Dehydration Treatments
          for Radioactive Sludge
 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
 6  Centrifugal
                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

               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.

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.

     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.

                      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


ring $

content of

content of

content of

     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


ring 0

content of

content of

content of

     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.



ring 0

content of

content of

content of

     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

     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 .


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,






            LIQUID  WASTE

            STORAGE  TAMK

                Fig.1  Before  the Improvement of the Dehydration System.



                                         SLUDGE  TANK
                        Fig.2  The Improvement Operation System


                                          by John E.  Stewart
                                   Werner & Pfleiderer Corporation

  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.

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.
                     Original ORNL dwg 78-18217
1962   1966   1970    1974

 Figure  1. Annual solid  waste volume shipped from
         U.S. nuclear  power plants. Reference 5,


                 ASPHALT  COLL   LUBE
                 FEED  \  &FILT   OIL    v
                 PUMP \SYS_\CpOLER\
                            CRANE TRAVEL AREA




                                                I   OOOOOO
                                CRANE  CONTROL

                                                    CONTROL  PANEL

                                ASPHALT STORAGE TANK

                                 ASPHALT  GEAR   FEED
                                 PUMPS  /BOX  /PUMPS
                        Figure 2. Typical design approach to VRS layout.

  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
                              DOSE RATE

1.0 to 10.0
1.0 to 5.0"

0.1 25 to 0.25

           *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.


  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
      7,500 HOURS WITHOUT

  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.
                            MIXING AND
                       KNEADING ELEMENTS
 Figure 4. Co-rotating screws which operate with 1 mm
         of clearance to provide self-cleaning action
         inside extruder-evaporator.


  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


  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
  Table III compares the shipments from a VRS system
with those from a non-volume reducing cement system.
                                               TABLE III
                                        SHIPMENT COMPARISONS
 *CPS — Condensate Polishing System

  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

                                                      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
/CN m v
, r




i p i
n „
, 150
m m. _

10 iO




Specific Activity of Concentrate
Spez. Aktivitdt des Konzentrats: 15 Ci/m3


                                           100     120

                                            »     4
V 1 PJ



, .1


ill LU



                                          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.

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.


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.

 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
 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.

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

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

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



 Figure 6. Recommended pH values for VRS system

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

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.

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
               600 F IGNITION POINT
             _ 550 F
    FLASH POINT (min.)
    ASTM TEST D-92
             _ 410 F STEAM INLET TEMP.
                ion  9noF
             l_ 190 • zoo t-
 Figure 7. Thermal data of asphalt as used in the VRS

 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


  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

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
Zero (Safe and non-
One (Requires significant
  preheating before ignition
  can occur)
Zero (Safe and non-
   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.


 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


                                              IN PROCESS
                           WATER IN FEED
                           AS WATER EVAPORATES,
                           SOLIDS TAKE ITS

                           VRS OPERATING TEMP. 300F
                           STEAM INLET TEMP. 410F
                           IGNITION TEMP. 600F
                             WATER ISA FIRE
                             EXTINGUISHER, THUS
                             INHIBITS IGNITION
                             SOLIDS GENERALLY
                             RAISE IGNITION

                             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

                             MORE THAN 300F
                              SAFETY MARGIN

                             MORE THAN 500F
                              SAFETY MARGIN

                             SIGNIFICANTLY REDUCED
                              ENVIRONMENTAL IMPACT &
                              CONTAINER CORROSION


  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
  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.


  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
• 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
    -4  _
                         TAP WATER

                 =—	~
                  SEA WATER


 i  i   i   i   i  i   i   i   i   i
          5    15      30      45
                              60    70
 Figure 8. Leach rate test results from AERE, Harwell,
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.

 1. W. Kluger, H.  Krause, O. Nentwich, FIXING OF
   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
   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
   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.
   reprinted  by Werner  & Pfleiderer  Corporation,
 5. A.H.  Kibbey,  H.W. Godbee, E.L.  Compere, A
   ORNL/NUREG-43 (Revision of ORNL-4942), Oak
   Ridge National Laboratory, Oct.  1978. .
   1, KFK 2104,  Gesellschaft fur Kernforschung
   M.B.H., Karlsruhe, West Germany, April, 1975.
 7. E. Backof, W. Diepold, STUDY OF THE THERMAL
   OXYGEN SALT MIXTURES,  Translated by Ralf
   Friese,  KFK-tr-450, July 1975,  Gesellschaft fur
   Kernforschung M.B.H., Karlsruhe, West Germany.
   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
   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
   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
   REPORT, July-Dec. 1977, HEDL-TME 78-48 UC-
   70,  published  July 1978, Hanford Engineering
   Development Laboratory.
   77-74 UC-70, Jan.-June 1977, published July 1977,
   Hanford Engineering Development Laboratory.
14. W. Bahr, W. Hild, W. Kluger, BITUMINIZATION OF
   Karlsruhe,  West  Germany,  presented  at ANS
   Winter Meeting, Oct. 27-31, 1974.



                         Rodger W.  Granlund
 Pennsylvania  State  University,  Health Physics  Office, University Park, Pa.
                              John  F.  Hayes
               Delaware Custom Materiel,  State College, Pa.

    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.


    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

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.


    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-

    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.

    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

                            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


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

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.

    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


    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.


    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

    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-

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.

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.

                         FIGURE  1
                         7 LEACH TESTS  FOR
            T7           SOLUTION SOLIDIFIFD

                         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.


     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.

     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

     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.

y   so


Z   40


|   -

             69 70 71  72 73 74 75 76  77  78

                     CALENDAR YEAR

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

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.


                  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

     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.



                 CALENDAR YEAR


                    I	I
           69 70  71 72 73 74 75 76 77  78

                  CALENDAR YEAR


                          FROM A LARGE  UNIVERSITY

Jamieson G. Shotts, David L. Spate,  and Philip K. Lee.  Health Physics Services,
University of Missouri,  Columbia, Missouri


    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.


    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

    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).


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


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

    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.


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.

                                  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
No. of



Vials (Gallons)*
Animals (Bags)              -


*100 filled vials - 0.25 gallon
                                     1974-75      1975-76      1976-77     1977-78















                                  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

  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

  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

                           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

                                FIOURE 1
                         UNIVERSITY OF MISSOURI
CAMPUS Columbia
Pick up
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
Disposition (Date)





                         DO NOT REMOVE THIS TAO
                        WITHOUT AUTHORIZATION OP
                           09-874      Pilule* in U 5 •
                          f Aflinelflltt Inc.  0  tafto Placa. N.r

          FIGURE 2


                       LOW-LEVEL TRU WASTE CONTAINERS

                              J. Bruce Peterson

                               Mound Facility*
                              Miamisburg, Ohio

     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.

     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

     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.


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

     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-

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


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


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.


     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.


     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.


     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

     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.

                              WITH  A HAMMERMILL


                              Winborn D. Gregory

                           University of Rochester
                           Health Physics,  Box  RB&B
                              601 Elmwood Avenue
                             Rochester, NY  14642
     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.


     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.



     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.


     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.


     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.

                 SESSION C
            Session Chairperson
              A. A. Moghissi
   U.S. Environmental Protection Agency

                              Teledyne Isotopes
                                D.R. Fuhrman
                                S.A. Black
                                J.P. Pasinosky
     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.


     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.


     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


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.


     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.

     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

     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)


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
     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

     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


               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.


     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.


(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.


                     BARNWELL  FACILITY


                     David  G.  Ebenhack

                 Chem-Nuclear Systems,  Inc.
        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.


        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.


        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.


        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.


        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


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

        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.

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

Personnel                6                 6
(Mechanics, Welder


New Jersey
New York
North Carolina
North Dakota
New Hampshire
Puerto Rico
Rhode Island
South Carolina
South Dakota
Washington, D.C.
West Virginia


1971     1972    1973     1974    1975     1976    1977
               VOLUME BURIED - BARN WELL
1979   £



Philip K. Lee, Jamieson G.  Shotts, and  David  L.  Spate.  Health Physics Services,
University of Missouri, Columbia, Missouri


     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.


     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


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


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

    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


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.

            TABLE 1
Summary of Radioactive Burials
    University of Missouri
(Cubic Meters)
4. 08x10 10

                                  TABLE  2
                      Residual Activities at Burial
                              (December 29, 1978)
Unknown Beta Emitters
   Activity (becquerels)

        i-ao-nfe*   SCALE r-io'

                       FIGURE 2
                VOLUMES OF  BURIED
                  RADIOACTIVE WASTES
                (corned forward)
           1972   1973    1974   1975   1976    1977   1978
                        CALENDAR YEAR

                          FIGURE 3






>   4x10
                  ACTIVITY  OF  BURIED WASTES

                                 PREVIOUS  ACTIVITY
                                    (carried  forward)
             1972    1973    1974    1975    1976    1977    1978

                           CALENDAR  YEAR

                      FIGURE 4

^  4xlO'v .
    2x10  .

                    OF BURIED WASTES
           1972   1973   1974   1975    1976   1977   1978

                       CALENDAR  YEAR


                               R. Andersen
                          University of  Colorado
                 T. Beck,  L. Cooley,  and  M. McCampbell
                  University of Maryland  at Baltimore
     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.


     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.


     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       ' '


    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


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".



     A comprehensive assesment of alternatives is beyond the scope
of this paper.  However, three alternative do warrant a brief dis-

     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.


     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.


     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  ..."


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.


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
Bur. Space
Chem Exp.
Bur. Trench
Rad . Exp

 Figure  1.  Matrix of Radwaste Parameters and Disposal Processes

           with Specific Impacts


                     PH o
                     W ft
                     u e
                     D O
                     Q U
                              SPECIFIC IMPACT

                              (Burial Space)


  (12% of Volume
   Buried in 1975) u
w C
en 0
W -H
U -P
O (0
« M
o u


           RESOURCES          ENVIRONMENT         HEALTH & SAFETY

           Personnel          Burial Trench       Chem. Exposure
           Facilities         Integrity           Fire/Expl.
           Containers                             Accidents
           Burial Space

Figure 3.  Impacts of Institutional Radioactive Waste Disposal

                            HOW TO  PLAN A DISASTER!

                               Wilson  C.  McArthur
        Hittman Nuclear and Development Corporation,  Columbia, Maryland

     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.


     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.


     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


     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.


     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

     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.


     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.


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.


     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


     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.


     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.

                                    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
West Valley
Maxey Flats
New York
South Carolina
Operations suspended
in March, 1975
Operations suspended
in December, 1977
Operations suspended
in April, 1978
135,000 ft3 per month
limitation imposed

Chem Nuclear
New Mexico
Licensing effort

Licensing application

                                   TABLE III
          Plant to Sheffield, Illinois
          Plant to Barnwell, South Carolina
          Plant to Beatty, Nevada
                          Approximate Cost

                               $  400
           TABLE IV

               Percent since
Overall living costs
Food, including meals out
Housing and operations
Clothing and upkeep
Medical care
Personal care
           Percent  since
             Year Ago


            TABLE V



                    *Ebasco Services, Inc.

                   **Regulatory changes account for 78 percent
                     of soaring nuclear plant costs.

CU.  FT.
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
                                                                                                                        SITE CLOSED
                                                                                                                       APRIL 9,  197S


S2.50 "1
SI. 80

Dec. 12
S'-75 April
     JUNE 19,  1976

                                                                                                   SITE OFFICIALLY  CLOSED
                                                                                                       DECEMBER 1977
                                                                                                               Dec. 6,
                                                                                                                            Apr!I  2
      JULY  I.JAN.  1,  APRIL  1.
       197«.    1975     1975
                                         JAN.  15,
   MAY 15,
JAN.  1.
JAM.  1.
JUNE  1.
                                                         EFFECTIVE DATE OF INCREASE

                                   COST OF TRANSPORTATION

                     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,
Joe 0. Ledbetter, Civil Engineering, The University of Texas at Austin
Gerard A. Rohlich, Civil Engineering, The University of Texas at Austin


     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.


     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

      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.)


     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

     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;

     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


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

     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).


     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).


     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.


     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

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.


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.

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

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

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,

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.



                         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

     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.

     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.


     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.


     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.

     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).

     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

     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

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:


     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).

     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).

     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
     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.


     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

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).

     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).

     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.

     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.


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",

US76  US Department of the Interior, Bureau of Indian Affairs, Final
  Environmental  Statement, Sherwood Uranium Project, Spokane Indian

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.

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.

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.

     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.


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.


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.


     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

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.


     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.


     The work of Mr. Bruce Butler in performing emanating fraction measurements
is greatly appreciated.


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,


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
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
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
(CRa> .
~/-i,' /„ a)
Fraction , N
fv\ b>
Emanating C for
Radium (cf )
Ka r>e - 1 «/-,• /„
    Range through profile
    Leached zone
    Mixed overburden lands
Sand Tailings
Debris Lands
Misc.  Sediment,  Scale
                                         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
32 (21-65)
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)
3.8 (2.5-7.8)
                                              C.   THERMAL PROCESS
69 (45-101)
3.5 (2.2-5.1)
 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.

TABLE 2.  Estimated Inventory of Florida Phosphate Solid Wastes and By-products*
(Form, 226Ra)
Facility Florida
Annual Annual
(spoils, 5 pCi/g)
Clays ;+
(settling ponds,
28 pCi/g)
Sand Tailings:
(spoils, 5 pCi/g)
(gypsum piles,
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
Volume, m3
Mass, tons (metric)
Area, hectares
226Ra activity, Ci
9.9 x 106 1
1.4 x 107 1
2.0 x 106 2
2.8 x 106 3
2.3 x 106 3
4.2 x 106 5
.3 x 108
.9 x 108
.5 x 107
.6 x 107
.0 x 107
.5 x 107

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
2.8 x 108
4.0 x 108
1.1 x 109
6.0 x 108
1.3 x 108
1.4 x 108
(slag pit, 60 pCi/g)
Volume , m3
Mass, tons (metric)
226Ra activity, Ci
1.1 x 105 2
1.4 x 105 2
.1 x 105
.7 x 105
6.7 x 106
8.6 x 106
 *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.


     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.


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).

         Table  4.   General  Models  for Soil Surface Radon Flux*
Radon flux at the soil-air
Radium concentration in the soil
media (grams, dry weight)
Bulk density, volume per unit
of dry weight
Emanating power, amount of radon


  released to void space per
  unit of radium (or radon)

Decay constant for radon-222


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

Depth of media
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
                                -  2)e
                               ml =
                                                   kl =
           y =
                                                   k2 = CRa2P2E2/P2
                                                    4    m D, tanh z
                                                                       .  ,
* From Bolch et al.(Bo78).

                                                                          To overburden
                                                                          spoil piles

                                 Weight:  100%
                               226 Ra: 38pCi/g
| Benef iciation
I Plant

\  22
  Weight:  30%
226 Ra: 26pCi/g
                                   (Flotation     \
                                     ^     J
        Weight: 10%
        226Ra: 57pCi/g

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
                            To sand
                                  To dryer,
                                chemical plant,
                                 or customer
                Fig. 1.  Simplified Flow Diagram for Phosphate Mining and Benef iciation.
                            Radium Concentrations are for Central Florida.

Phosphate Rock
226 Ra:  37 pCi/g
238 U:   32pCi/g
                                                                 Sulfuric Acid
[A. Phosphoric \
Acid \
Plant \


f Phosphoric Acid \
I 226 Ra: <1pCi/g
V 23B U: 30pCi/g >

By- („* Gypsum A
product V 238 U; <, pCJ/g 7
To gypsum
                           To further processing,
                        fertilizer production, and/or
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  )
                                        226 Ra:  1 pCi/g
                                        238 U:   2pCi/g
           f       Slag
           (  226 Ra: 64pCi/g
           V  238 U:  63pCi/g
                                                          Ferrophosphorus  \
                                                          226 Ra:  2pCi/g   )
                                                          "" U:  41pCi/g   7
        Fig. 3. Simplified Flow Diagram for Thermal Process Elemental Phosphorus
                Production. Radionuclide Concentrations are for Central Florida.

 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.



     J. Richardson, D.A. Lee; Bruce Nuclear Power Development,
Ontario Hydro, Tiverton, Ontario, NOG 2GO


     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.


     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
            . Stage 2 - waste volume reduction facility.
            . Stage 3 - concrete trench and tile hole storage
            . 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
was tes

12%- volume
98.4% - activity
88%- volume
1. 6%- ac tivity
Combus t ibles
- Incineration

                               63% - volume
                               0.05%- activity
                tibles  (PVC,
               - Compaction
    - trenches   /Storage
      - trenches
                                                         25% - volume
                                                         1.55%- activity

     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
Type 1
Type 2
Type 3
(Nominal Ci
< 0.1
0.1 to
> 100
/m3 <" )


     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-

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
           RELIEF  '
                                            1600F - 1800F
                               IGNITION BURNERS
                                                                          DRAFT FAN
                                                                                 CONTAINER FOR
                                                                                RADIOACTIVE ASH
                                   AIR BLOWER
               IGNITION BURNER;;
                  AIR BLOWER



                                            FIGURE   2   INCINERATOR SCHEMATIC

     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

     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).

                (yCi/burn)     (%)
Incinerator Ash
Baghouse Ash
Stack Effluent
ASH (%)
ASH (%)
< 1
< 1
< 1
< 1
< 1
< 1
< 1
   (Gd-153,  Ce-141,  Sn-113,
   Sb-125,  Mn-54,  Fe-59,  Ba-140)

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 .



                         Paul J. Macbeth
                  Ford, Bacon & Davis Utah,  Inc.

       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.


      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.

      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

      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


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.


      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-

      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


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.

    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

Reclamation Events
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
Hell Hater Single Cununulative
Inhalation/Direct Gamma/Food Transportation/Consumption/Container Accidents Effect
110 340








                      TABLE   2


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



Crew Size

Injuries Fatalities


Crew Size





                      TABLE    3











Shallow-Land Burial-Eastern
Shallow-Land Burial-Western
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





























& Profit






























101. SO
Total Total Unit,
Costs Costs ($/m )









































uompatiDility/Site/Saf eguards/Env./Availability/Inst . /Public /Individual /Industry Weighted
with Waste/Selection/ /Effects/of Technique/Control/Accept . /Costs /Costs /Coinparisti
Weighting Factor
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 .



. 129
. 164
. 164
. 152
. 140
. 140


. 059
. 078
0. Ill


. Ill
. 067
. 189
. 189





0. 128
0. 128
0. 112
0. 176
0. 176
0. 144
0. 144
0. 224

0. 300
0. 143
0. 157
0. 229
0. 429
0. 458


. 125
. 125
. 138
.275 '
. 200

1. 00
0. 97
1. 16

            SOLAR & PLANET
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                                                                    USA. NEA, IAEA



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                                 • CONCRETE STRUCTURES
                                 • METAL STRUCTURES
                                 • TANKS
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                                   USA, SJ.LGm.M. CANAUA,
                                  • UK. FRANCE" FOR,! INDIA.
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                                                                    USA. JAPAN. BELGIUM,
                                                                   , ITALY, KiR, IAEA, NEA,
                                                                    AUSTRALIA, FRANCE

Session Chairperson
    C.  R. Price
 State of Virginia


                                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-


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.

     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

     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

     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.
     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.

     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.


     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.


     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.


            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


     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

     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.

                           G. Wayne Kerr
                 U. S. Nuclear Regulatory Commission
                      Washington, D. C.  20555

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.


     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


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

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.


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.


     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

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.


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


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


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

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.

     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


     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-


     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.


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."

                           Table 1.  Volume of wastes and amount of activity buried at low level sites;

                                     cumulative through 1977.
Year Licensed   Volume buried X10 ft
 BPM (Ci)*
 SM (Ibs)
SNM (kg)
Pu (kgm)
Beatty, NV
Maxey Flats, KY
Hanford, WA
Sheffield, IL
West Valley, NY
Barnwell, SC
*    BPM, byproduct material; SM, source material; SNM, special nuclear material, including plutonium.

**   Cumulative through 1976.

***  1970 through 1977 only.


                            Heyward G. Shealy

          S. C. Department of Health and Environmental Control

                             Columbia, S. C.
     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.


     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.

     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

     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

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.

                        BURIAL OF SMALL QUANTITIES
                               NRC APPROVAL

                             John W. N. Hickey
                    U.S. Nuclear Regulatory Commission
                          Washington, D.C.  20555
     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.


     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


     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

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.

     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


                                                            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


     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.


     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.


     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.


                             H.  W.  Dickson
                  Health and Safety Research Division
                     Oak Ridge National Laboratory
                      Oak Ridge, Tennessee  37830

     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.

     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

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-
     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.

     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.

     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

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
     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.

     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

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-
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

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

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
      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.

     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

                           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.

     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.
     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;

     3.    radioactive waste disposal by conventional methods such as
          dumps, landfills, and on-site burial;
     4.    misapplication of waste products containing radioactive
     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.

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.

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
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.

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,
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.

     Decommissioning Standards - the Radioactive Waste Impact
                    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.


     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
     -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

     -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


     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

     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,


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

     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.


     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.

     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

     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.


     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

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.

     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.


     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.

     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.


(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.


                                G.  Lewis Meyer,
                                Stephen  T. Bard,
                                 Cheng Y. Hung
                                James Neiheisel

                          Office of Radiation Programs
                      U.S.  Environmental Protection Agency
                            Washington,  D.C.  20460


     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

     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

     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

     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.


' "






     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.


     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

 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

 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.

     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.
     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.

 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.

     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

     Environmental assessment of ocean disposal will require a considerably
different 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

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.

     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.


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,
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).


                       A. E. Desrosiers
                       Battelle, Pacific Northwest Laboratories
                       P.O. Box 999
                       Richland, WA  99352

                       Edwin Njoku
                       University of Kansas
                       Lawrence, KS  66044

     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.


     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


number of health effects per unit  of radiation dose and per person-year at

     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.


     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

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.


     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

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)	
                                  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.


                     W (137Cs) = 4.1 x 10-^ R-l  (yCi/£)

                     W (239Pu) = 2.5 x 10~3 ^  (yCi/£)


     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.


     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,

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.


     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.


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.,

                             TABLE 1
                        Risk Parameters
Red Marrow
Lower Large
Factor (W )
	 x —
Risk x 10-9
Risk x ID'8
from 58Co in-
gestion (yCi"1)
Stomach, Small
Upper Large
Intestine, Liver    0.24

TOTAL               1.00


*Risk of genetic effect in first two generations

                                        TABLE 2

                      Total Risk Coefficients and Allowable Intakes



Effective Whole
Body Dose Equiva-
lent per Unit
Ingestion (mrea/yCi)

Standard: 7
mrem/yr Effec-
tive Whole Body
Dose Equivalent
Allowable Intake (pCi)
Standard; 7 mrem/yr
Whole Body Dose Equiva-
lent or Critical Organ
Dose Equivalent Commitment*

(3) (4)
*I-131 intake calculated at 21mrem/yr


      1000 yr ACCUMULATION OF 239Pu
                           TABLE  3


                               William A. Mills
                         Criteria & Standards Division
                         Office of Radiation Programs
                      U.S. Environmental Protection Agency
          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

      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

exposures, working levels, to risk of lung cancer, bypassing the rad/rem

     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


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!


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).

                           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.


     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

     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

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


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


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

     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

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

     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

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

     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.


     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.


     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

     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

     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

     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


     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

     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.

               SESSION E
          Session Chairperson
              S.  V. Kaye
     Oak Ridge National Laboratory


                   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
     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.


     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.

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


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


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

     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

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

     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

     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

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

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.

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.

         Table 1.  Summary of current on-site exposures at the
                      Canonsburg, Pennsylvania, site
    Exposure source
   Range of average
    values observed
Maximum value
Radon in air inside

Radon daughters in air
inside structures

External gamma radiation
inside structures

Airborne 230Th inside

Radon in air outside

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
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

           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
0.52 pCi/£
0.0009 WL
11 uR/hr
<0.1 mrad/hr
Maximum values
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.

ota	  1+0	z+o	'tlffe^—4+o	s+o	6+0	 7+0-—	e-K>   >9


L. 	 _^



, 	 I 	 I 	 I 	 I 	 I

T 	 1




i 	 —


  Fig.  1.  Layout of  the present setting of the site at Canonsburg, Pennsylvania.

                                                                                   ORNL-DWG 78-7737
                                                I    I     I     I    1    I    I     I

                                                        .<^f - H	---I--
                                                                                          200 L
                                                                                          300 L
                                                                                    -1    -2
                                                                                O  50 100
                                                                                I  I   I
     Fig.  2.   Layout of landfill  site at Burrell Township, Pennsylvania.  Shaded
areas are  those where subsurface  contamination has been found.

                                                    ORNL DWG 78-20630
 200' L
 100' L
 100* R
 600' R
                                      Fig. 1
Fig.  3.   Present setting of the Middlesex, New  Jersey,  landfill

 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


     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

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:

                              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)

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

     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

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

     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).

     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:
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)

where £ is the depth of the soil cover over tailings which release  flux
   = 0) = $t.  The attenuation function f (5) is:
where  k =  rc  , h =/ rt  ,  and T = Tanh  (h  • z)
     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,'

(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

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

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).

     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

     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.

     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

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

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-

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).

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) .

         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.5 and ALKO.5 are covered with soil and Ra-226 concentration has not been
 determined yet.
             TABLE 2.  Physical Characteristics of Tailings Solids
 Diameter (ym)
Velocity (m/sec)
Velocity (m/sec)
Radon release rate:
of tailings:
area : *
3 x
x 10
x 10
x 10
x 10
x 10


 *This area was subdivided into 16 areas, each with an area of 1.88 x 10~2 km2,


TABLE 3.  Annual Relative Frequency of Occurrence (Sum of All Stability
                 Classes) Metset Bluewater, New Mexico
Wind Speed, knots
Wind Direction
Column totals:
Over 21
Row To
     TABLE 4.  Total Inhalation Dose Commitment (mrem/year)  from
               Particulates at the Direction of Maximum
                  Dispersion and Base Case Condition
Distance (km)
2 . 31E+00
7 . 51E-

        TABLE  5.  Total  External*  Dose  Commitment  (mrem/year)  from
             Ground  Contamination  and Airborne  Radionuclides
Whole Body
Bone Marrow
Distance (km)







2 . 60E-03





*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)
1 . 30E+00

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.

                                   •  CORRECTED FOR 226Ro IN SOIL COVER
                                   *  AVERAGE MEASURED FLUX
                                      PREDICTED FLUX
                                 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.
                                            ALKO TAILINGS
                                   •  CORRECTED FOR 226Ra IN SOIL COVER
                                   *  AVERAGE MEASURED FLUX
                                      PREDICTED FLUX
                    200      300     400      500
                        DEPTH OF SOIL COVER,  cm
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
  o  2
  8 •
                                                                         TAILING  MANAGEMENT
                 20th y«or  of operation
                 at maximum dl*p*r«lon
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

O • BOB* ea««
O • Mol*t b*och«*
A - W«t b«och«*
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.

  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.



              Lee Bettenhausen* and Veronica Burrows
                U. S. JUivironmental Protection Agency
                  Philadelphia, Pennsylvania
     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

     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
     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

"presently with U. S. Nuclear Regulatory Commission, King of  Prussia,  Pa.


     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

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.

     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


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

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


                         hot spot source   averaged source
                         "   "     Met 2
          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
Met 2
Met 1

Met 2
         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

   FIGURE  1.
   airport met. data
      underlined figures are
      radii in statute miles

      figures in sectors
      are person-rem/year


n-site  met. data
   underlined figures are
   radii in statute miles

   figures in sectors
   are person-rein/year


                    FIGURE  3.
                    POPULATION  DOSE
                    DISPOSAL  SITE
                    hot spot source
                    underlined figures are
                    radii  in statute miles

                    figures in sectors
                    are person-rem/year

                                  FIGURE  4.
                                  POPULATION DOSE
                                  DISPOSAL  SITE
                                  average source
0   underlined figures are
    radii in statute miles

    figures in sectors
    are person-rein/year


              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.


     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

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

     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 «* «-^« ..-„-,•,	,•„,•„„	„

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

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 «-^~ „-„„«._„„,..;	_*

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

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.


    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

    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

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

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

     (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"!


     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).


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  L. Cullen and Eduardo Penna Franca, editors), 5.

An74 J.R. Andrade Ramos and M.O. Fraenkel. 1975. Main Uranium Occurrences
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AP78 American Physical Society (APS). 1978. Report to the APS by the
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Br78 A.J. Brownscombe, D.R. Davy, M.S. Giles, and A.R. Williams. 1978.
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Ei67 M. Eisenbud. 1967. Radionuclides in the Environment. In Proceedings
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Fe78 P.A.M. Ferreira. 1978. personal communication.

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Fr50 R. Frahya. 1950. Relatorio da Diretoria - 1948 - Divisao de Fomento
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Ha74 P.L. Hainberger, I.R. de Oliveira Paiva, H.A. Salles Andrade, G.
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Ha67 J.H. Harley. 1967. Manual of Standard Procedures, Health and  Safety
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Ha68 B. Havlik, J. Grafova, and A. Nycova. 1968. Radium-226. Liberation
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Ha70 B. Havlik. 1970. Radioactive Pollution of Rivers of Czechoslovakia.
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IC59 International Commission on Radiological Protection (ICRP). 1959.
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IC74 International Commission on Radiological Protection (ICRP). 1974.
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IC77 International Commission on Radiological Protection (ICRP). 1977.
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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&
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  Ore Deposit at Zirovski Vrh. Journal of Radioanalytical Chemistry, 44,

NP77 Nuclear Power Issues and Choices. 1977. Report of the Nuclear Energy
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Pa77 A.S. Paschoa and G.B. Baptista.  1977. Environmental Impact of a Nu-
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  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.

Table 1.  Ra-226 concentrations -in waters of the Pogos de Caldas region.
A1- 6**
A- 7
Jan 77
wet season

1.8± 0.7 t
0.6± 0.2
0.4 + 0.2
ND tt
5.0± 1.0
Jul 77
dry season
1.2± 0.7
ND tt
4.1± 0.7
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
1.7± 0.4
6.0± 1.0
5.0± 1.0
1.3± 0.5
      * 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/£.

                 A,/ ^ ''NCOS  "\
                          «       \
                            ilOAS   V\
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.

              ANTAS RIVER
     50 -
LU 40
< 30

S 20
u. |0

* X
x b)

X x .
* X
* X x X
1 1 1 1 1 1 1 1 1 1 1 1 1


o, 40

CM 3.0



o E
C) E
o o Q
o o t—
o M-4 "

•A-l £

o A V - 2 2
o —
°' or
o o o

- O _J
• ' X
9 ^""
• • • 2
•*••*** * °
A 2
I I I I I. 1 1 1 I I I 1 1

e •

200 '!





1 1
' 1
1 1 I I 1 1 Y I 1 1 I 1
         JFMAMJJASONDJ (1978)
                  YEAR'- 1977
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).

                   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.

                        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


     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


     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

     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-

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
     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

          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.

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

     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

     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

     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.

   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-


                    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
     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.


     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.

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

     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

     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.


     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.

     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.

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


          Nz = P0 Z(10-5)    2-*   (1 + i)n.                             (8)

and the anticipated number of additional lung cancers due to the radiation
component is

          Nadd a K Wa Z P0(lCTe) \t^2 +


                                            (t1 - n) (t2 - n) (• .         (h)

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).


     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
              Land Type
   Crawl Space Homes
    and Mobile Homes

         overburden and debris

        Lower activity


        Unmined, near-surface
         radioactive deposits
         and fill

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)  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
   (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

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
1 . S 1 ab-on-g rade :
High-activity overburden
and debris lands
Low-activity overburden
2. Crawl space and mobile
homes: all reclaimed
1. Slab-on-grade
2. Crawl space and mobile
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

     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,

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.


     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.


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.          ~	~~	

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).

                         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


     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.


     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


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

     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

     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.


     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


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
     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

     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.


     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

     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.


     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

     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

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


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.

     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.


            W.  E.  Kennedy, Jr., R. B. McPherson and E. C. Watson
                        Pacific Northwest Laboratory
                         Richland, Washington 99352
     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.

     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.


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

   • 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.


   •  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
       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;

        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.


     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.

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,

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.


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.

             Table 1.   Estimated Maximum Radioactivity Deposited
              on the FRP Site Over a 30-Year1 Operating Lifetime
Deposited Radioactivity (yCi/m2)
Selected Times After Shutdown
5 . 5E-5
2 . 4E-3
2. OE-5
9 . 4E-6
10 Years
— (a)
— _
3. OE-5
2. OE-5
30 Years
2 . OE-5
3 . 1E-5
100 Years
— —
2. OE-5
3. OE-5

(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)
Deposited Radioactivity (yCi/m2)
at Selected Times After Shutdown
2 . 8E-5
10 Years
9 . 1E-8
7 . OE-4
7 . OE-4
30 Years
— _
100 Years
7 . 1E-5
7 . 1E-5
(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
Deposited Radioactivity (pCi/m2)
at Selected Times After Shutdown
23 5n
23 GU
7 . OE-6
7 . 2E+0
3 . 5E+0
10 Years
7. OE-6
7 . 2E+0
3 . 5E+0
30 Years

7 . OE-6
7 . 2E+0
3 . 5E+0
100 Years
2 . 4E-6
7. OE-6
2 . 4E-6
7 . 2E+0
3 . 5E+0
1 . 5E+0
6 . 8E+0
(a)  A dash indicates  values  less  than  10 7 pCi/m2.

              Table 4.  Disposition Criteria on the Reference
                FRP, PWR3 and MOX Sites Corresponding to a
                 Maximum Annual Dose of 1 mrem per



Class Y

Time After


239pu _ 2«tlAm
239Pu _ 2klM
239Pu - 241^
(a)  Includes doses from  inhalation,external  exposure,  and  ingestion  of
    food products.

                            • CONTAMINATION CHARACTERISTICS
                            'LAND AND FACILITY CONDITIONS
                            'POSTULATED SCENARIOS FOR
                          STEP 1:   COMPUTE THE MAX I MUM ANNUAL
                                  DOSES FOR THE REFERENCE
                                  RADIONUCLIDE INVENTORIES
LIMIT OF 1 mrem/yr
       VALUES OF 1 mrem/yr;
                           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


                       McDonald E.  Wrenn and Norman Cohen

                       Institute of Environmental Medicine
                       New York University Medical Center
                                550 First Avenue
                            New York, New York  10016
     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.


     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.


     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.

     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

     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)


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.

     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

     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
                             Dc = (DihVh + DibPbV I
             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


            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
           Oh  =  occupancy factor in halls (2 m from detector) taken as 1 hour/
          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:
     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

     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.


     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

     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:


             (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

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


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

     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
                                         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


Incineration (inhalation)
Landfill (ingestion/food)
Activity (yCi)
2x10" 3
3xlO~ 5
      ,The collective estimates of potential harm (risk) are summarized
       in Table 4.


       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.


     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.

     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.

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.

Ba75   Battelle Northwest Laboratories, 1975, "DACRIN - A Computer Program for
  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-


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

In72   International Commission on Radiological Protection, 1972, ICRP-19,
  "The Metabolism of Compounds of Plutonium and Other Actinides," Pergamon

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.

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'

                                     TABLE 1

                50-Year Committed Dose Equivalent and Risk Values
                     Adopted for 2l>1Am.  0.5 Micron Particles
Estimate of
Risk/yCi Inhaled
  or Ingested
Ingestion (fi
Class Y
= 10-")
Class W

Cancer Risk/rem
Class Y Class W
l.lxlO"1 l.lxlO"2
5x10- 3 ID'2
3xlO~3 0.6xlO~2
l.ZxlO-1 2.7x10-^
* The dose to lung is to the pulmonary region where lung cancers are most
often found in inhalation studies with actinides in experimental animals

                                     TABLE 2

                     Calculation of Collective Dose and Risk
            	from Incineration of Smoke Detectors	

                 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

            Then, the total estimate of risk from one year's operation
            until complete disposal would be:


            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

               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

             Composition	Typical Composition
                         TABLE  4

                Collective Disposal Risks
                                     Lifetime Committed
Incineration                               2xlO_*
Burial                                     2xlO_5
Firefighters                                 10 5
  External	person-rem

    home             340
    transport          2
    warehousing        6
    retailing         17
Total                365                   7x10


       Margaret N. Maxey, Ph.D.
         University of Detroit
       Detroit, Michigan  48221


     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
     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.

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

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

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

        (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


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.

        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."

II.  EXPERTS AND STANDARDS FOR "SAFETY":   How we have institutionalized dissensus
                                          and value conflicts.
        Aaron Wildavsky has recently observed, "Experts are used to disagreeing,
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  .
        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;

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.


        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.


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

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)


       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


       "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.

       . . . 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,


          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.

       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."

        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.

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
       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."

        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.



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.

                SESSION F
           Session Chairperson
                R. M. Fry
            State of Kentucky


                     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
         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.

     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

 Operated by Union Carbide Corporation under contract W-7405-eng-26 with
 the U.S. Department of Energy.

                                                                               ORNL-DWO 65-12157*7*
     AREA NO. 3
                                                                           ,SOLIO WASTE DISPOSAL
                                                                                 AREA NO. 2

                       SEWAGE TREATMENT PLANT

                     SOLID WASTE DISPOSAL
       AREA NO. I
                                      l.L.W. TRENCHES
                                        (NOT IN USE)
                SAMPLING STATION
                OF ORNL COMPLEX
      1000     2000
       1     ,    I

     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
Operating Dates
1943 - 1944
1944 - 1946
1946 - 1951
1951 - 1959
Land Used
       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

the east entrance of the Laboratory.   Selection of the site was most
likely based on consideration for the reduction o