United States
          Environmental Protection
          Office of
          Radiation Programs
          Washington DC 20460
ORP/CSD 79-2

This report was prepared as an account of work sponsored by the
Environmental Protection Agency of the United States government under
contract No. 68-01-3997.  Neither the United States nor the United
States Environmental Protection Agency makes any warranty,  express or
implied, or assumes any legal liability or responsibility for the
accuracy, completeness or usefulness of any information, apparatus,
product or process disclosed, or represents that its use would not
infringe privately owned rights.

                         Assessment of Waste Management of
                                           Volatile Radionuclides
                             Philip M. Altomare
                             Marcel Barbier
                             Norman Lord
                             Daniel Nainan
                             May 1979
Contract Sponsor: EPA                                           The MITRE Corporation
                                                              Metrek Division
Contract No • 68-01-3997                                         182° Oolley Madison Boulevard
Project No.: 15730                                                McLean. Virgima 22102
Oept.: W-53
                                                         MITRE Technical Reoort

     The Office of Radiation Programs carries out a national program
designed to evaluate the exposure of man to ionizing and nonionizing
radiation, and to promote the development of controls necessary to
protect the public health and safety, and to assure environmental

     Regulations which will become effective in 1983 limit the release of
the volatile radionuclides krypton-85 and iodine-129 into the general
environment from uranium fuel cycle (UFC) operations.  This contract
report considers the problems of, and technologies for, the disposal of
krypton-85 and iodine-129 collected in accordance with the UF£
regulations.  It also considers the disposal of two other volatile
radionuclides, hydrogen-3 (tritium) and carbon-14.  The information in
this report will be used by the Agency in its development of standards
for the management and disposal of high-level radioactive wastes.

     Comments on this report  are welcome;  they may be sent to the
Director, Criteria and Standards Division (ANR-460), Office of Radiation
Programs, U.S. Environmental Protection Agency, Washington  D.C., 20460
                                  William A Mills, Ph.D
                              Criteria Standards Division
                          Office of Radiation Programs (ANR-460)


     This document presents a review of the Technologies for Waste

Management of the Volatile Radionuclides of Iodine-129, Krypton-85,

Tritura, and Carbon-14.  The report presents an estimate of the

quantities of these volatile radionuclides as are produced in the

nuclear power industry.  The various technologies as may be used, or

which are under investigation, to immobilize these nuclides and to

contain them during storage, and in disposal are discussed.  Also,

the alternative disposal options as may be applied to isolate these

radioactive waste from the human environment are presented.

     The report contains information which was available through

approximately January of 1978.

                          TABLE OF CONTENTS


1.0  INTRODUCTION                                                 1

2.0  SUMMARY                                                      5
2.1  Quantities of Waste Produced                                 5
2.2  Available Immobilization Technology                          7
2.3  Disposal Options                                             9

3.1  Introduction                                                 15
3.2  Iodine-129                                                   16
     3.2.1  Quantities Produced                                   16
     3.2.2  1-129 Waste Form                                      17
3.3  Tritium                                                      18
     3.3.1  Quantities of Tritium Produced                        18
     3.3.2  Tritium Waste Form                                    24
3.4  Krypton-85                                                   26
     3.4.1  Quantities of Kr-85 Produced                          26
     3.4.2  Krypton-85 Waste Form                                 27
3.5  Carbon-14                                                    30
     3.5.1  Quantities of C-14 Produced                           31
     3.5.2  Carbon-14 Waste Forms                                 33

4.0  IMMOBILIZATION TECHNOLOGY                                    39
4.1  Iodine-129                                                   40
     4.1.1  Immobilization of Iodine                              41
     4.1.2  Immobilization of Iodine in Zeolite                   50
4.2  Tritium                                                      52
     4.2.1  Polymer Impregnated Tritiated Concrete  (PITC)         53
            Leaching from Concrete
     4.2.2  Organic Compounds                                     60
     4.2.3  Hydrides                                              62
4.3  Krypton-85                                                   66
     4.3.1  Pressure Vessel Containment                           °°
     4.3.2  Zeolite Adsorption                                    ^°
     4.3.3  Ion Implantation/Sputtering
4.4  Carbon-14                                                    „-
     4.4.1  Concreted CaC03                                       „?
     4.4.2  Leaching of C-14 from Concrete

                    TABLE OF CONTENTS (Concluded)
5.1  Disposal Concepts                                            8*>
     5.1.1  Geological Repository                                 86
     5.1.2  Seabed Disposal                                       8S
     5.1.3  Transmutation                                         89
     5.1.4  Extraterrestrial Disposal                             "0
     5.1.5  Other Continental Disposal Options                    92
     5.1.6  Ice Sheet Disposal                                    92
5.2  Disposal Alternatives for Volatile Radionuclides             ^3
     5.2.1  Iodine-129                                            93
     5.2.2  Carbon-14                                            10°
     5.2.3  Tritium                                              104
     5.2.4  Krypton-85                                           106

REFERENCES                                                       11:L

                           LIST OF FIGURES
Number                                                           Page

    1       Schematic of the Direct C02 Fixation Process          35

    2       The Effept of Iodine Content on the Leach             45
            Rate of Barium lodate Concrete, No Additives

    3       Incremental Leach Rate of Barium lodate from          47
            Concrete Containing 9.5 wt% Iodine

    4       Leachability of Iodine into C02 Free Distilled        48
            Water from Type 1 Portland Cement Containing
            9.5 Wt% Iodine as Barium lodate

    5       Projected Tritium Release Versus Time for             58
            Static Leaching of the SRL Lysimeter Testing
            Duplicate Specimen (without Container) in
            Distilled Water

    6       Krypton-85 Heat Generation and Decay Rates as         68
            a Function of Time

    7       Representation of Sodalite Cages Containing           73
            Krypton Atoms

    8       Process for High Pressure Encapsulation of Kr         74
            in Zeolite

    9       Calculated Release of Original Krypton Inventory      77
            from Sodalite at 150°C as a Function of Time

   10       Process for Immobilizing Kr-85 by Ion Implanation/    80

   11       Pressurized Cylinder Storage Facility for Kr-85      109

                           LIST OF TABLES














Preparation of Volatile Radionuclides for             10

Summary of Volatile Radionuclide Disposal             12 12

Annual Tritium Production in a Typical 1 GWe          19
Light Water Reactor

Estimates of Tritium Distribution in Different        21
Pathways at Fuel Reprocessing Plants

ERDA's Estimate of Tritium in Fuel Processing         22

Tritiated Water Produced by 5 MT/Day Reprocessing     23

Carbon-14 Production in Light Water Reactors          32

Solubilities of Selected Iodine Compounds in          42

Characteristics of Chemical Storage Technologies      54
for Tritiated Water

Properties of Engineered Storage Options for          55
Tritiated Water

Summary of Contractor Storage Practices               56

Leaching Data of Polymers                             61

Material Costs for Polymeric Media and Alternate      63
Fixation or Storage Methods

Tritium Activity in Leach Solution ZrHx(T)            65

Annual Kr-85 Storage Requirements for a .1500 MT/      69
Year Reprocessing Plant

Rubidium Production During Storage of Kr-85           71
in High Pressure Steel Cylinders

Sorption of Krypton Gas by Metal                      81



     This study, sponsored by the Environmental Protection Agency

(EPA), Office of Radiation Protection (ORP) investigates control

technologies and waste management for the radionuclides 1-129, Kr-85,

C-14, and H-3, as well as other volatile radionuclides which are ret

leased as gases or in volatile forms from nuclear facilities.  This

report is a survey of existing literature and provides background

information to assist EPA in the preparation of standards for radio-

active waste disposal of these volatile radionuclides.

     In present U.S. defense related programs and proposed commercial

nuclear programs, spent fuel from nuclear power reactors may be re-

processed to recover usable uranium and plutonium.  During these

chemical processing operations, radioactive gases, particulates, and

volatile compounds are released to the off-gas effluent streams.

Control approaches consistingvof treatment of the off-gas streams to

remove particulates and effluent gases, and atmospheric dispersion

have been used in the nuclear industry to maintain radiation levels

below applicable standards.  Although the present control technology

is capable of meeting or exceeding the requirements of maximum allow-

able radionuclide concentration standards, concern remains as to the

release of the long half-life radionuclides.  The long half-life

volatile radionuclides, specifically iodine-129 (1.7 x 10' yr),

krypton-85 (10.44 yr), carbon-14 (5,570 yr), and tritium (12.26 yr),

have the potential to accumulate in the environment, presenting a

long-term radiological hazard.

      EPA recently promulgated standards,  Title 40 Part 190,  Code_£JL

 Federal Regulations, limiting the allowable release of 1-129,  Kr-85,

 and the transuranium elements from the uranium fuel cycle.   These

 uranium fuel cycle standards  include  limits of releases from milling

 operations through fuel element reprocessing operations.*  Standards

 for limits on the allowable releases  of carbon-14 and tritium are

 also under consideration.  EPA is in  the  process of developing

 environmental standards for the "back end"  of the fuel cycle,  namely,

 the disposal of radioactive waste.  Standards for radioactive waste

 disposal  would include  the portions of the  nuclear fuel cycle  fol-

 lowing  reprocessing  through ultimate  disposal—immobilization or

 solidification,  packaging or  containment,  interim storage,  and final


     The  primary emphasis of  the  present  study is to describe  current

 technologies  for immobilization and containment of the collected

 volatile  radionuclides.  To provide a perspective of overall waste

 management,  consideration is  given to the quantities of waste  that

 may be produced,  physical and chemical form of the collected waste,

 immobilization and containment technologies,   alternative disposal

 options, and  environmental transport  at each  waste management stage.

Although  this  report  addresses each of these  areas, the discussion

*Reprocessing  operations may  be interpreted to include additional
 waste treatment  when performed onsite, e.g., immobilization  or
 solidification  of waste and  packaging.

of waste management practices or alternatives remains incomplete and

uncertain in some areas since relevant technologies for waste manage-

ment of the volatile radionuclides are still being developed.

     The primary source of volatile radionuclide release in the

nuclear fuel cycle is from the spent fuel reprocessing operations.

The exception to this would be in the case of a U.S. policy decision

not to perform reprocessing and to directly dispose of spent fuel

elements without recovery of uranium or plutonium.  In this event,

the primary release of the volatile radionuclides would occur some-

time after disposal, if the integrity of the fuel elements and other

engineered containment barriers were to fail.

     At present there are no operating commercial spent fuel repro-

cessing plants in the U.S., although three reprocessing plants have

been constructed.  A reprocessing plant was operated by Nuclear Fuel

Services at West Valley, New York.  This plant was closed when it was

determined that modifications would be required which were uneconomi-

cal for continued operation.  The Midwest Fuel Recovery Plant was

constructed at Morris, Illinois.  Operational problems were encoun-

tered at this facility, requiring major plant modifications.  No

decision has been made to perform these modifications and the plant

is presently being used for storage of spent fuel elements.  A

reprocessing plant was constructed by Allied-General Nuclear Services

at Bamwell, South Carolina and licensing processes were initiated

for this plant.

     A U.S. Administration policy decision has been made not  to pr°"~

ceed with commercial fuel reprocessing until issues of proliferation

of nuclear materials are resolved.*  In view of this decision, the

future of the nuclear fuel reprocessing industry is not clear.  In

the interim, the U.S. Government is developing a program for  the ac-

ceptance and caretaker responsibilities for spent fuel  from privately

owned nuclear reactors.  No decision has been made as to whether such

commercial fuel reprocessing will be performed at some  future date or

whether spent fuel elements will be disposed of as nuclear waste


    Although reprocessing of commercial spent fuel will not occur

unless a U.S. policy change is made, reprocessing is continuing  for

defense-related programs.  Accordingly, the assessment  of the waste

management technology for volatile radionuclides remains of concern

to assure protection of the environment and public welfare.


      The present study is a survey of the existing literature and

provides background information -for the preparation of standards for

radioactive waste disposal.  The present technologies for immobili-

zation and containment of the long-lived volatile radionuclides

(iodine-129, tritium, krypton-85, and carbon-14) are described, and

the quantities produced, physical and chemical forms of the collected

waste, alternative disposal options, and environmental transport  are

reviewed.  Actual experience with control technologies is scarce

because there are no operating commercial spent fuel reprocessing

plants in the U.S.; the present national policy is to postpone com-

mercial fuel reprocessing until issues of proliferation of nuclear

materials are resolved.  Concerning the assessi«ent of waste manage-

ment technology remains of concern to assure protection of the

environment and the public welfare.

2.1  Quantities of Waste Produced

     The quantities of radionuclides produced depend on the installed

nuclear electric power generation capacity.  The estimated gross nu-

clear power capacity in the year 2010 varies from 400 to 1000 GWe.

The high-level radioactive waste from fuel reprocessing amounts to

3.3 percent of the heavy metal weight.  Roughly 3.3 percent of the

spent fuel will have to be disposed of after 10 years' aging.  In

this report, a model fuel reprocessing plant is assumed to be capable

of handling 5MT of spent fuel per day.  From 5 to 14 such plants may

be required by the early part of the next century.


      Iodine-129 is produced primarily as  a fission product  and

 through the decay of other fission products.  A model  reprocessing

 plant of the type referred to  would release  380kg  (or  66 Ci)  of 1-129

 per year,  mixed with about 250kg  of stable 1-127.   Most  of  the iodin«

 is released in elemental  form,  with 1  to  5 percent in  organic form.

 The mercury, iodox,  and chemisorption processes  can be used for the

 collection of iodine.

      Tritium (H-3)  is  produced  primarily  by  ternary fission.   Small

 amounts of H-3  can be  produced  by  (n,n )  reaction  on Li-7 arising

 from the (n,ct)  reaction on  B-10 used  as a neutron  adsorber  in light

 water reactors  (LWRs).  The model  reprocessing  plant releases 1.25 z

 10^  Ci  of  H-3 per  year.  The chemical  state  in  which tritium is

 released is  not well established but  it is known that  tritium can

 occur in elemental  form as  tritiated water and  also in combination

 with organic materials.  The collection of tritium is  accomplished

 through  three processes:

      (a)  head end process, including  voloxidation and pyrochemical

      (b)  process-stream,  controls  followed  by  isotopic  separation;

     (c)  retention  of entire water effluents,  including wastes
          removed  from gas  stream.

     Krypton-85 is  the only noble  gas  radionuclide which  is suffi-

ciently  long-lived  to be important  from a fuel processing standpoint.

The annual production of krypton  from  the reference model reproces-

sing plant is about  143 m3, of  which approximately  six  percent will

be Kr-85 with an activity of 12.7 x  106 curias. Krypton is sepa-

rated from other gaseous effluents by cryogenic distillation.

Adsorbtion in fluorocarbons, liquid carbon dioxide, and on charcoal

are also being investigated.

     Carbon-14 is produced mainly by (n,p) reaction with N-14 (pre-

sent in fuel as an impurity) and (n,c0  reaction with 0-17,  and to a

small extent by neutron capture in C-13.  The amount of activity re-

leased from the model (5MT/day) processing plant is comparatively

small, about 850 Ci/year.  Most of the carbon released from the plant

is in the form of C02»  The best known method for the immobiliza-

tion of carbon dioxide is caustic scrubbing with Ca(OH)2 to form

calcium carbonate.  Adsorption on molecular sieves and in fluoro-

carbons has also been demonstrated on an experimental basis.

2.2  Available Immobilization Technology

     To isolate the volatile radionuclides from the bio-sphere, four

types of barriers are possible:

     •  the chemical form of the waste;

     •  immobilization in a solid matrix;

     •  outer containment;

     •  structural or natural barriers at storage or disposal

     In cases where there is a choice of various chemical compounds

it would be prudent to choose the form which has the lowest solubili-

ty and leachability.  Iodine-129 poses a special problem because of

 its extremely long half-life.   Incorporation  of barium  iodate  in con-

 crete has  been particularly  favored  as  an  immobilization  technique.

 Various  experiments have  been  conducted to develop  a model  of  the

 leaching of barium iodate from concrete, however, a universally ap-

 plicable model is  not  yet available.  Iodine  can  also be  immobilized

 in  zeolites,  of which  silver-exchanged  zeolites have been found to

 have  a high chemisorption capacity for  elemental  iodine.   However,

 because  of  the high cost  of  silver chemisorption, lead  exchanged

 zeolites are  being investigated.  No  leaching tests are available on

 either type of iodine  loaded zeolite.

      For the  containment  of  tritium,  there is a choice  between chemi-

 cal storage and containment.   Chemical  storage technologies  include

 the use  of  polymer-impregnated hydrates, organic  compounds,  and  hy-

 drides.  Polymer-impregnated tritiated  concrete has been  the subject

 of  several  leaching studies  and cost  analyses.  A method  for the con-

 tainment of tritiated  water  mixed with  plaster and  cement  in a poly-

 ethylene drum resulted in very low leach rates. Organic compounds

used  for fixing tritium include bakelite,  poly-acrylonitrile,  and

polystyrene.   Also  zirconium hydride  has been found to  be an adequate

storage mechanism with a  low leak rate.

     Krypton,  being a  noble  gas, is released  in elemental  form.

Pressure vessel containment  is  the easiest method  of storing krypton.

Five hundred  years  has been  suggested as a minimum for the  useful

life of  each  cylinder  which  can store krypton  at a pressure of  120

atmospheres.  The main reasons for possible  failure of  the vessel  are

heat released during radioactive decay and corrosion caused by the

decay product rubidium.  Krypton can also be stored by  adsorption  in

zeolites and this method of containment has reached an  advanced stage

of development with a possible saturation sorbency of 45 litres per

kilogram.  Another method for immobilization of krypton is through

ion implantation or sputtering on solids such as aluminum, but the

amounts of metal required are too high to  be practical.  Much higher

loadings have been achieved by electro-static acceleration of krypton

ions.  Methods for using metal matrices to hold granules of calcines

containing krypton are also being  studied.  The only technique being

studied for the immobilization of C-14 is the incorporation of

CaC03 into concrete, asphalt, or polymers.

     Table I is a summary for the waste radionuclides of the


     •  quantities produced (for model plant)

     •  available collection technology

     •  available immobilization technology

     •  containment packaging

2.3  Disposal Options

     The following alternatives are considered for the  disposal of

volatile radionuclides:

     •  Geological repositories in salt beds, salt domes, and
        crystalline rock forms such as granite, basalt, shales,
        limestones, and clay beds.  Of these, salt deposits have
        received the most attention because  of their plastic flow
        properties.  The greatest concern is groundwater movement.

                                                                      TABLE  I


                             IODINE-129                          TRITIUM                           KRYI'TON-85
                 66 Cl
                 380 kg
                 .08 .3  (2)
 1.25 x 10  Cl
 0.126 kg
 1.2 lltrea  O
  12.7 x 10  Cl
  530 kg
  143 m3 with 61 (Cr-85
850 Cl
0.192 kg
300 litres <*>
 Pyrochemical processes
 Process-Steam treatment
  Cryogenic Distillation
Caustic scrubbing.
                 lodate of llg,  Ba
                                                       Trltlated water
 Early etagea
  Pilot plant
Well established for
stable carbon.
                    Not known
                                                                                                                              $0.15 j.?r kg of  CO
                 lodate in concrete
 Chemical storage  In polymer Im-
  pregnated  concrete ,  polyethylene
 Organic compounds
  Zeolite adsorption
  Ion Implantation (sputtering)
                                                                                                                              CaCO  In concrete.
                    Concrete leaching
                    A cube 6000 kg leached In
                     17,000 yeara.
                                                    Around 0.0001Z per year
                                                     Zirconium hydride  leaching
                                                      .89  to  1.7  x  10-6 en/day
                                                                                          0.3Z in 8 years  from Zeolite
                                                                                                                              No data.
                 Experiments  still  In  progress.
Impregnated concrete -  Advanced
experimental.  Organic' compounds,
hydrides - Experimental.
Zeolite -Advanced experimental.
Ion Implantation - experimental
                                                                                                                              Well  known.
                    $356.000  In Ag for 600 kg I.
                                                    $3.10-$16.90 per gallon of
                                                     trltlated water.
                                                                                        Pressure  vessel containment.
                                                                                          (for  krypton gas)
                                                                                                                           55 gallon drums.
                                                                                           '500 years
                                                                                           Well established
                                                                                                                              40c a gull nn
J./1500 NT/year reprocessing plant
; ..{mixed with a comparable amount  of  1-127
 In  the forn nf IITO

     •  Seabed disposal,  which involves controlled emplacement in
        deep sea sediments or beneath the bedrock of the ocean
        floor.  Crucial factors for selection of the repository are
        geological stability and the existence of deep sea, sediments.

     •  Transmutation, which is feasible at the present time only in
        fission reactors.  The neutron fluxes available, and neutron
        cross-sections for these volatile nuclides are such that
        there is very little merit in this method.  Fusion reactors,
        however, if eventually developed, may be capable of volatile
        nuclide transmutation due to their high neutron flux.

     •  Extraterrestrial disposal, where the long-lived nuclides are
        launched into space, so they escape the solar system.  This
        is an expensive option and only C-14 and 1-129 have lifetimes
        long enough to warrant the use of it.  The amount of C-14
        produced is of such low magnitude that it does not warrant
        such extreme measures.  Iodine-129 alone should be considered
        as a serious candidate for space disposal, however, there is
        a serious concern regarding accidents during launching and
        possible reentry.

     •  Other continental disposal options such as mined cavities, a
        matrix of drilled holes, super-deep holes, deep well injec-
        tion, and hydrofracture could have an application at appro-
        priate locations where there are no major threats to long
        term containment through groundwater movement.

     •  Ice sheet disposal in Antarctica or Greenland, where thick
        ice formations are available, has been considered.
        Antarctica is subject to international agreements, and
        Denmark has sovereignty over Greenland.  Apart from such
        political considerations, it is desirable to further
        investigate the evolutionary behavior and the effect of
        future climatic changes on the ice sheets.

     •  Storage for a period long enough to decay to non-hazardous
        levels could be a disposal option for tritium and krypton-

     Feasible disposal options for the volatile radionuciides are

shown in Table II.  Data are insufficient on the physics and history

of ice sheets for this concept to be considered practical.

     Radiological health effects have not been considered in relation

to alternative disposal methods.


                                                          TABLE  II

Not likely to contain 1-129
until sufficiently decayed.
Would require identification
of stable, water free, geologic
formations with good iodine
sorption capability.
Appears to be a
satisfactory disposal
Appears to be  satisfactory
but may not be desirable
to place Kr-85 with  other
waste in a repository.
Engineered storage
also possible.
Appears satisfactory
but may not be desir-
able to place tritium
with other waste.
Engineered storage
also possible.
Not likely to contain 1-129
until sufficiently decayed.
Ocean dilution may reduce
concentrations to acceptable
Appears to be a
satisfactory disposal
Appears to be a
satisfactory disposal
Appears to be a
satisfactory disposal
Satisfactory for elimination
of 1-129.
Accident risk and consequences
require careful study.
Probably not warranted
because of cost and
availability of other
Probably not warranted
because of cost and
availability of other
disposal methods.
 Probably not warranted
 because of cost and
 availability of other
 disposal methods.
                    Not a likely removal mechanism
                    unless fusion reactors are
                                   Not a likely approach.
                         Not a likely approach.
                             Tritium could be
                             used as fuel for
                             fusion reactors if


     The long half-life and generally low sorption characteristics of

1-129 are such that it is difficult to assume that geological  dispo-

sal will provide isolation of the waste for the period of time  re-

quired for this radionuclide to decay to innocuous levels.  The

seabed is unlikely for providing complete containment but does pro-

vide an additional time barrier and ocean dilution prior to .reaching

biologically active areas to reduce the biological hazard.

     Dilution of 1-129 in the ocean following release from the seabed

may reduce the concentration to acceptable levels.  Slow release

from containment is necessary to assure that local high concentra-

tions do not occur.

     Extraterrestrial disposal is attractive as a disposal option for

1-129, however, the radiological impact of possible accidents must

be carefully determined.

     Transmutation of 1-129 would only be feasible if fusion reactors

were practical.


     Both geological and seabed disposal appear to be satisfactory

concepts for the disposal of C-14.  The cost of extraterrestrial

disposal is not warranted if other disposal concepts are satis-



     The disposal of Kr-85 in geological and seabed repositories

appears to be a satisfactory option if it is separated from other


waste.  Extraterrestrial disposal and transmutation are not required

due to the relatively short half-life of Kr-85.  In view of the short

half-life, surface or underground engineered storage facilities could

also be considered.


     Geological disposal and seabed disposal of tritium are both

feasible and engineered storage may be more satisfactory.   The

tritium could be used as a fuel in the event fusion reactors were to

become practical.

     Extraterrestrial disposal  of tritium does  not  appear  to be a

satisfactory disposal approach.


3.1  Introduction

     Volatile radionuclides will be created in the process of nuclear

electric power production.  The bulk of these nuclides will be either
released and collected at fuel reprocessing plants or contained and

disposed of within the spent fuel in which they were formed.  The

volatile radionuclides may be packaged and di/posed of in their col-

lected form or further treated to reduce the potential for their

release to the environment.

     The quantities of volatile radionuclides produced will be

affected by the installed nuclear electric power generation capacity.

In a study that supported the EPA in developing environmental stan-

dards for high-level radioactive waste, presented projections of

installed nuclear electric power were presented.^  Projections

estimated an installed gross nuclear electric power capacity in the

range of 400 to 1000 GWe in the year 2010.  A total commercial waste

burden of spent fuel for the lifetime production of installed nuclear

capacity up to the year 2010 was estimated in the range of 3.1-7.7 x

105 MTHM (400-1000 GWe).  The estimated annual disposal requirements

in the year 2000 for commercial spent fuel aged 10 years were 9.7-14.5

x 10^ MTHM, based on a net installed nuclear capacity of 380 GWe to

570 GWe.

     In this study, quantities of waste are referenced to a model

spent fuel reprocessing plant assumed to be capable of handling 5 MT


 per day (1500 MT/year) of spent fuel elements corresponding to 50 GWe

 years of nuclear power generation.3  From 5 to 14 model reproces-

 sing plants are required to handle 400 GWe to 1000 GWe of installed

 nuclear capacity, assuming a 70 percent capacity factor.

 3.2  Iodine-129

      Iodine-129 is produced in  the nuclear fuel  elements as a fission

 product and from the radioactive  decay of  other  short-lived fission

 products such as Te-129,  Sb-129,  or Sn-129.   1-129 decays to the

 stable isotope Xe-129,  emitting beta and  gamma radiations of 120 and

 30 keV, respectively.   The  half-life of 1-129 is 1.7  x 107 years.

      3.2.1   Quantities  Produced

      The projected quantity of  1-129 released from fuel in a 1500

 MT/year model reprocessing  plant  is 380 kg/year.^>^  This released

 1-129 has an activity  of  66 Ci/year (approximately 1.3 Ci/GWe-year).*

 In addition,  approximately  250  kg of stable  1-127 are mixed with the

 1-129 in the  off-gas steam.  Thus a total  of 600-650  kg of iodine  per

 year  has  to  be  treated  in the iodine removal system,  immobilized,  and

 disposed of  for  each model  reprocessing plant.   This  corresponds to

 approximately  12-13 kg/GWe-year.   One to  five percent of the iodine

may be  present as organic iodine  (methyl  iodide  0113!) in the

effluent stream, or as HI and HOI.

*In contrast, the release from  a  typical LWR is  negligible:  10~6

     3.2.2  1-129 Waate Form

     The chemical waste form of 1-129 depends on Che technology

utilized for collection.  Four processes have been investigated for

the collection of iodine:  the Mercurex process, lodox process,

chemisorption process, and caustic scrubbing.'29)

     The Mercurex process will yield 30 m^ of liquid waste per year

per model reprocessing plant.  This is equivalent to 100 liters per

day of 8 molar HNC>3 and 0.4 molar mercuric nitrate containing 1300 g

of iodine in the form of mercuric iodide.  Research and development

are under way to convert the mercuric iodide to solid mercuric iodate

or to barium iodate with recycling of the mercury.  The purpose of

this research is to convert the liquid to a solid waste, thereby

reducing the waste volume and obtaining a less mobile waste form for

further handling.

     The lodox process with nitric acid yields a very low solid waste

volume (0.4 nrvyr per model reprocessing plant) of nonvolatile

mercuric iodate Hg(l03)2«

     The chemisorption process yields 3.5 ton/year per model repro-

cessing plant, occupying 3 cubic meters and containing the 380 kg of

1-129 in the form of chemisorbed silver iodide trapped in silver ex-

changed or silver impregnated adsorbents such as zeolite, silica, and

alumina.  The annual silver costs in this process are $356,000 for the

total 600 kg of iodine captured.  Other metals, such as lead, are

being studied as exchange media to reduce costs.

     Due to its very low efficiency of removal of organic iodine,

caustic scrubbing is no longer being considered for collection  of

iodine from reprocessing plants.

3.3  Tritium

     Tritium has a half-life of 12.2 years.  It decays to stable

helium-3 with the emission of a beta-particle with an energy of 18.6

keV.  There is no gamma radiation.

     Tritium production in light water reactor fuel is mainly by

ternary fission—three fission products instead of the usual two,

the  third one being tritium.  In addition, tritium can be formed

by the (n,n alpha) reaction on Li-7 resulting from a (n, alpha) reac-

tion of reactor neutrons on boron-10.  Boron is present in control

rods of most light water reactors and as a chemical additive in the

reactor coolant of pressurized water reactors.  PWRs usually use

silver-cadmium-indium control rods but boron control rods have been


     3.3.1  Quantities  of Tritium Produced

     Table III shows the annual production rate of tritium in a 1000

MWe LWR.   The tritium produced in the control rods stays in situ until

the end of the life  of  the reactor.  The tritium produced in the

reactor coolant of a PWR appears in the waste of the reactor coolant

treatment  system.

                                             TABLE III

Uranium, plutonium
Lithium-7 resulting from
neutron capture by
Same aa above
Neutron capture
Ternary fission
(n, na)
Same as above
Ci/GWe yr

Reactor coolant
Fuel rod
Boron control rods
(BWR only)
Primary coolant (PWR only)
Source:   Rhinehammer et al.,   p. 352.

     The following discussion is limited to the tritium contained  in

the fuel elements treated at a reprocessing plant.  The number  of

curies of tritium to be expected in the model reprocessing plant is

1.25 x 106 Ci per year.

     There is considerable uncertainty regarding the chemical state

and distribution ratios for tritium in the different pathways it can

take at a fuel reprocessing plant.  This is exemplified in Table IV,

which gives ranges of estimates according to different sources  for the

various possible pathways.  It is noted that the tritium fraction

reclaimed in the cladding hulls after shearing can be high and  depends

on the burn-up of the fuel inside the cladding.  Estimates on the

dissolver operation off-gas also vary widely.  The distribution ratios

used by ERDA as a guideline in 1976 are indicated in Table V^.  The

total tritium is 30 percent lower than in Rhinehammer's evaluation

given in Table III.  In Table V, the tritium from the fuel cladding

hulls is assumed to be recovered by wet processes in the form of

tritiated water (HTO).

     Table VI shows absolute maximum quantities of tritiated water

that could be produced either in concentrated or in diluted tritium

waste,  and the volumes resulting in both categories.  The ratio of the

activities in concentrated or diluted categories vary according to the

elimination process chosen.

                                   TABLE IV

                            AT FUEL REPROCESSING PLANTS
                                  PERCENT OF TOTAL
                                   FISSION YIELD
Shearing operation off-gas

Tritium retained in the
 cladding hulls

Dissolver operation off-gas

Uranium/plutonium bearing
 organic stream from solvent
 extraction process

Aqueous phase after solvent
 extraction process

                                   14b, 20*.
                                     ,  10e, 20f, 458

                                     ,  6h, 20h


Elemental, HTO

In combination
  with organic

a)  Zircaloy 2 cladding; burn up 43,000 MWD/MTHM, (Goode and Vaughen, 1970) 7
b)  ERDA 76-43
c)  Zircaloy 2 cladding; 12,000 MWD/MTHM, (Grossman and Hegland,  1971)8
d)  Zircaloy 2 cladding; 21,000 MWD/MTHM, (Grossman and Hegland,  1971)
e)  Ribnikar and Pupezin
f)  Savannah River Laboratory
    Mus grave
    Hall and Ward

                               TABLE V


Spent fuel input
Gaseous waste (as HT)
Cladding hulls
Dissolver solution (as HTO)
a) High-level waste
Ci/yr «C
1500 MTHM/yr

  b)  Low-level liquid             72           394         591,000

Source:  ERDA 76-43,  Vol. 1,  p. 2.64., Reference 4.

                                TABLE VI

Tritium Ci/liter
Total Tritium Ci/day
Total Tritium Ci/yr4
Total Tritiated Volume
50 kg3

105 (105 kg)
3 x 10~2

3 x 107
3 x 104
^Condensate from head end process such as voloxidation.
^Condensate from evaporators and acid fractionators.
^Volume has not been determined, probably less than 100  liters/day.
 Assuming 300 operation days/year.
Source:  ERDA 76-43, Vol. 2, p. 14.26.t  Reference 4.

     3.3.2  Tritium Waste 'Form

      There are three control systems for the collec-tion of tritium

from reprocessing plants:

     •  Head-end process (voloxidation and pyrochemical techniques);

     •  Process-stream controls (recycle and/or isotopic separation);

     •  Retention of entire water effluent, including water removed
        from gas streams.


     The voloxidation process requires a front-end kiln to heat

chopped fuel elements.  Over 99 percent of the tritium becomes vola-

tilized as tritiated water vapor (HTO) at temperatures ranging from

450 to 650°C.  Off-gls from the chopper (where tritium is released as

HT) is passed through an oxidizer to convert HT to HTO, which may then

be removed by a drier-molecular sieve arrangement, trapping the HTO in

a molecular sieve.  Estimates of  waste quantities from this process-

are less than 100 liters/day for a 5 MT/day reprocessing plant.

Pyrochemical Processing

     In pyrochemical processing the cladding is selectively melted

 (stainless steel at 1450°C, zirconium at 1840°C) and the resulting

 bare fuel is reduced in a solution of zinc, calcium, magnesium, and

 calcium chloride at 800-900°C.  During these steps, tritium is

 released as  a gas together with the volatile radionuclides.  Due to

its small atomic size, tritium can be separated subsequently from  the

other gases.   The complete process has not yet been demonstrated.

Process-Stream Treatment

     In process-steam treatment, two techniques have been proposed to

control the tritium once it has entered the aqueous streams of the

reprocessing plant:  water recycle and isotopic separation, expected

to be used in conjunction.  The most favored simple option (between

total recycle without separation and no recycling, but direct separa-

tion of tritium from the effluent stream) is recycle with bleed stream

separation to reduce the in-plant tritium concentration to a tolerable

level.  According to a study in 1975^ ^', in this option most of the

tritium (94.5 percent) is removed as HTO vapor in the off-gas from the

leacher (high activity side).  Approximately 4.5 percent is removed

from the low activity side in the form of tritiated waste containing

tritiated water.  The tritium removed in the off-gas in the high

activity side for the model reprocessing plant amounts to 735,683

Ci/yr, and 3,500 Ci/yr are removed from the low activity side as

liquid HTO.
     Once recycle has been accomplished, isotopic separation can be

more economically performed on the more concentrated bleeding stream.

Six processes for isotopic separation are envisioned at present:

     •  Catalytic exchange (convert HTO to HT)

     •  Fractional distillation of water

     •  Distillation of hydrogen

     •  Electrolysis of water

     •  Reverse electrolysis (using a palladium diaphragm)

     •  Laser enrichment

The volume of tritiated water released from a recycle  combined  with a.

separation process (isotopic distillation) would be relatively  small—

on the order of 5 to 6 gallons per ton of uranium processed,  i.e. ,  25

to 30 gpd or 7500 to 9000 gallons per year from a 1500 MT/year

reprocessing plant.^

'3.4  Krypton-8 5

     Krypton-85 is the only long half-life noble gas radionuclide

formed as a U-235 fiss.ion fragment that is present in  appreciable

quantities when the LWR spent fuel is reprocessed.  The half-life of

Kr-85 is 10.73 years and the major activity is beta radiation of 0.65,

MeV followed by a gamma ray of energy 0.52 MeV.  The krypton-85 gamma

ray occurs in only 0.4 percent of the disintegrations.  Krypton-85

decays to the stable isotope rubidium-85.

     3.4.1.  Quantities of Kr-85 Produced

     The annual gaseous Kr production from a model reprocessing plant

is about 530 kg, which has a volume of 143 m^ at STP-  About  6  per-

cent of this krypton is Kr-85, for an aggregate radioactive discharge

of 12.7 million curies.^  This is equivalent to 254,000 curies per

year for 1000 MWe of nuclear power generation.

     Krypton gas is released during reprocessing at the chopper and

dissolver steps.  It is accompanied in the off-gas stream by  xenon,

unrecovered oxides of nitrogen, and air and water vapor.  After

treatment for nitric acid recovery -and iodine removal, these off-

gases  pass  through heaters (to avoid condensation), prefilters, and

high efficiency particulate air (HEPA) filters before further treat-

ment of the purely gaseous components.  It is at this stage that tht

krypton containing the radioactive Kr-85 at 6 percent concentration

must be separated for further treatment.  Krypton is only about .003

percent by volume of the total off-gas.

     3.4.2  Krypton-85 Waste Form.

     A number of methods are being developed in U.S. and foreign

laboratories for the collection of krypton from off-gas streams.  Each

of  these methods produces krypton in  the gaseous state.

Cryogenic Distillation

     The noble gases krypton and xenon may be separated from the off-

gas stream by utilizing the widely separated boiling points of the

main components.(*^'  At a pressure of one atmosphere, these boiling

points are N20, -88,5'C; Xe, -108°C;  Kr, -157°C; 02, -183°C;  and

No, -196°C.  The oxide N20 deliquesces at room temperature and

therefore must be removed before any  temperature reduction to prevent

solidification and blocking of the gas-flow lines in the system.

     The process for cyrogenic distillation used at the Idaho

Chemical Processing Plant (ICPP) and  other U.S. installations is as

follows.  In addition to N02, water,  02, and N20, are all

removed prior to cooling since both NoO and water will freeze.

Oxygen is removed to minimize radiolytic ozone formation at cryogenic

temperatures where the excessive oxygen concentration will pose a

severe explosion hazard.  Oxygen is catalytically recombined with

hydrogen to form water over palladium or platinum at 550°G.  This  3tep

is followed by drying either by adsorption or freezeout.  The N20  is

also removed catalytically at 370 to 600"C in a reaction which dis-

sociates N20 to elemental nitrogen and oxygen on rhodium.  The

rhodium is regenerated at 870°C under a reducing stream of


     The dried gas, which is a mixture of krypton, xenon, and nitro-

gen, is precooled to liquefy both xenon and krypton by countercurrent

liquid nitrogen flow.  Operating on a very reduced flow, the separa-

tion column fractionally distills the Xe-Kr mixture to yield mostly

krypton at the top as 75 percent Kr, 25 percent Xe, and almost pure

xenon at the bottom.  Radioactive Kr-85 is confined to the krypton

rich mixture which can be handled remotely and collected as a gas  in

pressurized cylinders.

     At present, the Idaho Chemical Processing Plant (ICPP) system is

operated intermittently as a pilot plant to validate and refine the

technology for a full scale demonstration plant.^

Cryogenic Selective Adsorption - Desorption

     A modification of cryogenic distillation has been proposed by a

consortium in Japan.^''   This system consists primarily of alter-

nate adsorption and desorption at reduced temperature and pressure.

Experiments are presently underway to test the individual steps.    It

is planned  that the system will eventually be used at BWRs on exhaust

gases  and at  all  other nuclear facilities.  A major distinction from

the U.S. cyrogenic distillation, process is that oxygen is not com-

pletely removed prior to> cooling.

     In this process, moisture  and C02 are adsorbed on beds of

synthetic zeolite.  The concentration of-noble gases is achieved on

two adsorption beds of charcoal (A and B), using the following steps:

     •  Selective adsorption on A until a Kr concentration
        limit is reached at output;

     •  Desorption of A by evacuation at high temperature to pass
        noble-gas enriched flow to the next stage.  Since the
        noble gases are not desorbed as easily as carrier gas,
        their concentration in  the bed increases;

     •  Bed B is desorbed when bed A is adsorbing and vice versa.

     The inlet flow to the storage system still contains nitrogen,

oxygen, ozone, and some gaseous impurities.  These are removed

selectively by metal getters.

Fluorocarbon Absorption

     Selective absorption of krypton by liquid fluorocarbon has been

offered commercially in a process applicable for the off-gas from

pressurized and boiling water reactors.^  >*'  In the fuel element

reprocessing, there are some additional problems in the krypton

collection due to the presence  of nitrogen oxides (NO, NC^, ^0),

carbon dioxide, water, iodine,  and methyl iodine.  Recent ERDA results

show that refrigerant-12 (dichlorodifluoromethane) demonstrated the

most overall promise for selective absorption.^20;  ^ fluorocarbon

adsorption process designed for krypton removal can tolerate some

impurities and be equally effective in the impurity removal—notably


iodine, methyl iodide, and C-14 in carbon dioxide.  The process

exploits the difference in solubility of the various gas constituents

in  the solvent and facilitates fractional distillation.

Other Collection Processes

     Other collection processes are being developed for various  types

of  reactor operations.  At Oak Ridge National Laboratory (ORNL),

adsorption in  liquid CO  has been developed for the high tempera-

ture Gas reactor (HTGR).^21^  In West Germany, a process has been

developed for  separating Kr and Xe from dissolver off-gas in repro-

cessing HTGR fuel.  In the West German process, a helium purge-gas

cycle is used  for a coarse fractionation of krypton and xenon by

cold-trapping  at 80°K (-193°C).  At this temperature, xenon is depo-

sited in solid form at low pressures and krypton is deposited at 6

atmospheres.   The separation by freezing is facilitated by the reduced

partial pressure of the two gases with added Jielium.

     Another krypton refraction process has been investigated at

Westinghouse Electric Corporation for use with the liquid metal fast

breeder reactor (LMFBR).  Helium is  being considered as a cover gas

for this reactor and in the proposed process, charcoal would be used

to adsorb krypton from the helium at temperatures between -140°C and

-100°C.   This process is very similar to the Japanese adsorption-

desorption process.

3.5  Carbon-14

     Carbon-14 is  a  low energy beta emitter with a half-life of 5730

years.   It  decays  to the stable isotope nitrogen-14 with the emission


of a beta-ray with a maximum energy of 156 keV.  Carbon, being a

constituent of all organic materials, is easily absorbed into the


      In nuclear power reactors, C-14 is produced by (n,p) reaction

w"ith N-14 and (n,or) reaction with 0-17.  There is also a small

probability o£ neutron capture in C-13.  Oxygen-17 has a natural

abundance of 0.037 percent and C-13 is 1.13 percent of naturally

occurring carbon.

     Carbon-14 is produced in both the fuel elements and in the cool-

ing water in light water reactors.  Nitrogen is present in the fuel

interstices as an impurity and the amount present can vary over a wide

range; twenty parts per million by weight is typical.  Oxygen, of

course, is a major component of oxide fuels used in LWRs.  The reactor

coolant of LWRs is also a source of carbon-14 where nitrogen is pre-

sent as an impurity.  Nitrogen is typically one  part per million by

weight in the reactor coolant water.

    3.5.1  Quantities of C-14 Produced

     Table VII lists the estimates of C-14 production in fuel elements

and coolant water of both types of light-water reactors as published

by Bonka et al.22 and Kelly et al.23

     Between 20 and 30 curies of carbon-14 are formed during the pro-

duction of 1000 MWe-years of electric power.  At the present time

most of the C-14 produced in the fuel is released to the atmosphere

as C02 during the dissolution of the spent fuel at the reprocessing


                                              TABLE VII

                             CARBON-14  PRODUCTION IN LIGHT WATER REACTORS


                                                BWR                           PWR


1 7

Bonka et al. Kelley et a

8.4 2.7

12.9 10.9
21.3 13.6


11.2 16.0
32.5 29.6
1. Bonka et al. Kell








plant.  At Che reacCor site Che isoCope is released mostly in  Che

gaseous form and Che remainder is contained in Che liquid waste.  An

average of approximately 17 Ci/year of C-14 is estimated to be prod-

uced in the spent fuel elements and 11 Ci/year is estimated to be

released from the reactor per 1000 MWe-years of power generation.  A

model 1500 MT/year reprocessing plant would release about 850 Ci/year

from the fuel.

     The C-14 gaseous releases from light water reactors are not

always in the chemical form C02»  Based on various measurements,^

it is estimated that the fraction appearing as CC-2 in BWRs varies

between 66 and 95 percent.  In contrast, over 90 percent of the

gaseous C-14 activity in PWRs appear as CH^ and C^&f,, and only

10 percent as C02«    However, when the fuel elements are dis-

solved in nitric acid, Che excess oxygen in soluCion from U02 and

HN03 convercs mosC of Che carbon Co CO or C02«  Thus ic is esti-

mated thaC aC lease 95 Co 99 percent of the C-14 contained in the fuel

will be released to the off-gas system as C02«

     3.5.2  Carbon-14 Waste Form

     Methods for the collection of C-14 from off-gas streams include

caustic scrubbing, molecular sieve adsorption, and fluorocarbon

absorption.  The most probable chemical form is calcium carbonate,

which  may be incorporated subsequently into concrete or other


     Caustic Scrubbing

     The obvious way to immobilize the carbon dioxide is by reacting

it with a caustic to produce a carbonate.   The most inexpensive such

reagent is CaO (lime), however,  the chemical reaction

                  CaO + C02-*CaC03

is impractical because of the slow reaction rate.  Two aqueous

processes have been considered:   (1) the direct reaction of C02 with

a slurry of slaked lime  Ca(OH)2  where the fixation reaction is

                 Ca(OH)2 + C02-»CaC03  + H20

(2) the double alkali process which involves the reaction of C02

with NaOH to form Na2C03

                 2NaOH + C02 —Na2C03 + H20

followed by the reaction:  Na2C03 + Ca(CH)2~* 2NaOH + CaC03.

It has been contended that the direct fixation is superior to the

double alkali process on grounds of simplicity, smaller corrosion

effects, and better economics.^

     Figure 1 is a schematic of the direct fixation process.  Pebble

lime (CaO) is pulverized and slaked to produce the relatively insolu-

ble Ca(OH)2.  The slaked lime is slurried and pumped to a fixation

tower.  C02 is bubbled through a sparger at the bottom of the tower

and the gas combines with Ca(OH)2 to produce CaC03, which also is

insoluble in water.  The slurry is filtered on a continuous filter and

the filter cake is transported by screw conveyor a disposal system

such as concretion.

Co (OH)2
                                           INDICATES THAT THE PATH
                                           SELECTED DEPENDS  ON THE
                                           PACKAGING / TRANSPORTATION/
                                           DISPOSAL METHODS  USED
                                     FILTER CAKE
                      CONTINUOUS      SCREW
                         VACUUM      CONVEYOR

   Source:   Croff, Reference 25.



                                                     FIGURE 1
                                  SCHEMATIC OF THE DIRECT C02 FIXATION PROCESS

     This process is assumed to remove 99 percent of the C02 ini-

tially present.  It uses a relatively well known technology and is

used industrially to produce CaC03 although very little mass trans-

fer and reaction data have been gathered in the past.  In industrial

calcium carbonate production processes, there is a tendency to lose a

significant portion of the product through leaky pump seals, pipes,

and tanks.  This situation is tolerable in an industrial plant because

the product is relatively inexpensive, but careful attention must be

given to quality assurance and maintenance when dealing with radio-

active materials.

     In the double alkali fixation process, the main difference is

that NaOH is added to the make-up water; further, there is a reaction

vessel wherein the sodium carbonate reacts with calcium hydroxide.

The double alkali C02 fixation process is not used industrially

because the calcium carbonate produced contains residual amounts of

NaOH which should be removed by extensive washing to prevent corro-

sion.  The only difference from .the direct CC>2 fixation process is

the presence of NaOH in the water used for slaking the lime.  It has

an advantage over the direct process with respect to availability of

design data, since data on the reaction rate of C02 with aqueous

NaOH, causticization, and lime slaking are available in the litera-

ture.  Further, a C02 absorption tower contains only soluble sodium

compounds, thus reducing the possibility of scaling.  The presence of

concentrated NaOH, however,  could cause corrosion problems.  The


capital costs of both systems are approximately equal.  The operating

costs are also expected to be about the same except for the additional

sodium hydroxide needed.  Assuming 99 percent recovery of an average

of Ci of ^C02/year, the average annual output of radioactive

calcium carbonate is about 1.36 kg.  However, ten to 100 times as much

CaCOj from atmospheric carbon dioxide would probably be recovered

along with C-14 species.

     Molecular Sieve Adsorption and Fluorocarbon Absorption

     In addition to fixation in CaCC^, two other methods have been

evaluated:  molecular sieve adsorption and fluorocarbon absorption.

In the former, the carbon dioxide is removed by adsorption on a

molecular sieve.  Impurities such as ^0, NO, NC>2, and water vapor

must be removed.  ^0 is removed by catalytic decomposition using

rhodium as the catalyst.  NO and N0£ are removed as nitric acid.

The resultant gas is dried on a 3A molecular sieve and the C02 is

adsorbed on a 5A molecular sieve.

     To collect the carbon, the molecular sieve is regenerated by

heating in a gas purge, and the regenerated gas is scrubbed to produce

CaC03.  The technology for this process is not fully developed at

this time.

     Fluorocarbon absorption utilizes the solubility of carbon dioxide

in fluorocarbons.  It has been developed on a pilot plant scale at Oak

Ridge, but has not been demonstrated on actual dissolver off-gas.


     Various methods can be utilized to isolate the volatile radio-

nuclidea from the biosphere.  The methods for imposing barriers to

the transport of these nuclides into the environment fall into four

general categories of engineered isolation:

     •  selection of the chemical waste form

     •  immobilization in a solid matrix

     •  outer containment (packaging)

     •  structural or natural barriers at storage or disposal sites

     It is desirable that the highly mobile volatile nuclides be con-

verted to a physically and chemically stable nonvolatile, nonsoluble

compound where possible.  Various chemical compounds are under consid-

eration for 1-129, C-14, and tritium.  However, the noble gas krypton

does not combine into a stable physical compound at normal tempera-

tures and pressures and other forms of immobilization are required.

Krypton and the chemical compounds of the other radionuclides can be

incorporated into a solid matrix thus providing an immobilization

barrier and delaying their release into the environment.  The solid

matrix material selected must be capable of withstanding the nuclear

radiation and high temperatures that may result from radioactive decay

heat.  The release of the nuclides under consideration can be further

restricted by using an outer packaging, e.g., metal containers both to

contain the radioactive nuclides and to resist and delay the effects

of corrosion, erosion, and leaching which could eventually release the


radioactive elements.  Finally, the storage or disposal method or site

can be engineered or selected in such a way that if the nuclides

escape the packaged containment, there will be secondary containment,

i.e., the transport into the biosphere will be sufficiently delayed so

as to allow natural radioactive decay to reduce the biological hazard.

An additional safety factor can be obtained by selecting the storage

or disposal site such that releases into the environment are diluted

prior to reaching areas of human exposure (e.g., atmospheric

dispersion or dilution in the sea in the case of seabed disposal).

     This section discusses chemical waste form, immobilization

technologies, and research that has been conducted to measure the

leach rate from different immobilization forms of the radionuclide.s..

Containment packaging is also briefly addressed.  Alternatives for the

storage and disposal of the volatile radionuclide wastes are included

in Section 5.0.

4.1  Iodine-129

     Features of Iodine-129 which present particular problems i-n the

waste management of this radionuclide are as follows:

     •  A long half-life (16 million years) such that complete
        isolation from the biosphere until quantities of this
        radioisotope decays to innocuous levels cannot be assured;

     •  The iodine compounds are not stable at temperatures above a
        few hundred degrees, thus incorporation into glass matrices as
        proposed for other radioactive waste is not practical;

     •  Most iodine compounds are soluble to some extent in water;

     •  Iodine ion exchange with most soils  is not as  favorable
        as that of other elements.

     The radioactive iodine can be chemically combined into several

iodide and iodate compounds.  Table VIII gives the solubility of these

various iodine compounds.  Among the iodates, the mercurous salt is

the most insoluble (1.1-1.6 x 10~12 kg mole/liter) but the few pre-

liminary experiments with the mercurous salt have not been successful

in obtaining lower leach rates than with the barium iodate^.

Barium iodate is particularly favored for incorporation in concrete.

The chemisorption of iodine on silver and lead exchanged zeolites,

forming the silver and lead iodides, has also been investigated.

     The inclusion of any compound of radio-iodine in glass does not

appear feasible because iodine compounds dissociate at the temperature

where glass melts.  Various researchers state that 1-129 cannot be

easily immobilized in glass matrices and'that no development work is

currently in progress.26,27,28  xhere is a possibility of introduc-

ing iodine into glass by use of a pressurized system, but this is not

considered practical at present.

     4.1.1  Immobilization of Iodine

     Barium Iodate in Concrete

     Barium iodate has been investigated for immobilization of iodine

in concrete for three reasons. 9>30 jj its ]_ow solubility (8.1 x

10"^ kg-mole/liter at 25°C); (2) it can be prepared directly from

partially soluble barium hydroxide without using any superfluous

                                TABLE VIII


                  Solubilities3 (10~3 kg-mole/  at 298°K)b

5 molalc
4.9 at 20°C
1.65 x 10~3
1.1 x 10~8
9.7 x 10~5 to
1.3 x 1(T4
3 x 1(T10 (Hg2D
7.9 x 10~3
8.1 x 10~4
(3.6 to 5.5)
1.8 x 10~4

(1.1 to 1.6)

x 10~5

x 10~9(HgI03)
*The cations selected include those most abundant in Portland cement
 (Ca2+, Mg2+, Na+, K+,  omitting A13+, Fe3+, and Si4+ and those
 that form the most insoluble,  simple compounds of iodine (Ba2+, Ag+
 Pb2+, Hg2+, and Hg2+).

aMolar solubilities were calculated from data obtained from A. Seidell,
 W.F. Linke, Solubilities of Inorganic and Metallorganic Compounds,
 4th ed»,  American Chemical Society, 1958, as indicated.

bExcept where indicated.

cLack of solution density data  prevented conversion to molarity.

dFrom N.A. Lange, Handbook of Chemistry, revised 10th ed.,  pp 278
 -279,  McGraw Hill, New  York, 1967.

ion species and all of the resulting slurry of barium iodate can b«

incorporated in the concrete product so that no liquid waste; and (3)

the iodate form is compatible with the lodox process, which is under

development at ORNL for collecting the 1-129 waste.  This process

yields essentially 100 percent iodic acid HI03 ^rom which barium

iodate can easily be prepared.

     It has been found that approximately 10 weight percent iodine in

the form of barium iodate can be incorporated into concrete.  It is

estimated that the fission product iodine, after decay of the short-

lived iodine isotopes, will consist of 75 percent 1-129, the remainder

being 1-127.  In this form, the heat generation amounts to 3.6

microwatt/kg of concrete.  This heat generation is considered negli-

gible for all practical purposes and cannot give rise to temperatures

which would impair the long term stability of the iodine compounds in

the concrete.

     Approximately 6000 kg of concrete of volume 2.6 cubic meters is

required for the disposal of 600 kg of iodine from a model reproces-

sing plant.

     Iodine Leaching from Concrete

     It appears that the rate at which water penetrates concrete is

slow, so only a small fraction of the radioiodine compound is in

contact with water at a given time at the solid/liquid interface

inside the concrete.  This effect slows down the dissolution rate of

the radioiodine.

     The solubility of a chemical compound is not  related  in a  simple

way to the concentration of this compound in water when  the  compound

is embedded in concrete.  Tests in stagnant water must be  made  to  find

the upper limit of concentration of the compound in water.   When the

water is flowing or changed at regular intervals, the conditions are

different from those with stagnant water.

     For barium iodate incorporated in concrete, some leach  tests  have

been conducted under dynamic conditions simulated by changes  of leach-

ant at regular intervals.

     Small, but finite, leach rates were measured by Clark and Moore

on concrete cylinders 50 nun x 50 mm, including up to 11.9 weight

percent iodine. °»29

     In one experiment, the leachant (distilled water) was changed

every day at the beginning, every week later on, and every month at

the end of the test period.29

     The cumulative leach rates obtained are shown in Figure  2.  The

ordinate is the function

          Z(a /A ) (V/S)
          n  n  o

where a  is the weight of iodine leached per leaching period;

      A  is Che initial iodine content of the specimen;

      V  is the specimen volume;

      S  is the apparent specimen area.


 I— (
       8 x 10
       7 x 10~2L
       6 x 10
      5 x 10
       4 x 10
       3 x 10
       2  x 10
              0   2  4  6  8  10  12  14  16  18  20

          SQUARE ROOT OF LEACHING PERIOD,   t*  (day)'

 SOURCE:  W.  E.  Clark, 1977,  Reference  29.



This function is plotted versus the square root of the total leaching

period for various iodine contents in weight percent 'in the concrete.

The leach rates in this experiment were found to decrease with time

(Figure 3),  and the leach curves (Figure 2) exhibited a bend toward a

saturation effect.  When the experiments were repeated, was changed

every day.^9  Moore repeated Clark's experiments,  this time

changing the leachant every day.  This curve (Figure 4) shows no

saturation effect.

     Figure 4 can be extrapolated to determine when all of the

radioiodine will leach out of a concrete block.  While not appropriate

in the practical case for reasons noted below, the exercise is of

interest in indicating the difficulties of containing 1-129 for the

periods required for the radioactivity to decay to innocuous levels.

     The 600 kg yearly output of iodine from a model reprocessing

plant can be mixed with cement to form 6000 kg of concrete.  This

quantity is assumed to be cast into a cubic block.  Assuming a

density of 2.3 grams/cm^, a side of the cube would be 140 cm

and the volume to surface ratio a/b = 23 cm.  Assuming a linear

relationship i Figure 4.

                           Vt = (23) /14  \
or   t = 6.13 x 106 days = 1.68 x 10^ years.  This time period is

insignificant compared to the half-life of 1-129 (1.7 x 107 years)

i.e.,  nearly all of the 1-129 would exist at the time of complete

leaching from the concrete block.


                      	  SPECIMEN D-l

                      —  SPECIMEN D-2
 0         20        40       60        80

                    AVERAGE LEACHING  PERIOD      	
Source:  Clark, 1977,  IU:1 tirenre 29.

(L   is  the  time period  between  changes of leachant)
                                                                    200   300  400
                             INCREMENTAL LEACH RATE OF BARIUM IODATE
                             FROM CONCRETE CONTAINING 9.5 WT% IODINE

                               6.0    8.0    10.0   12.0
                        SQ. RT. ACCUM. TIME   (DAYS)
    Source:  Moore, 1977, Reference 26.
             (WATER/CEMENT RATIO = 0.89; CURED 56 DAYS)

     This example is not a good representation of what is likely to

occur in the practical case.  Reasons for the inadequate representa-

tion are discussed below.

     First, the leach test that was conducted extended over a period

of a few hundred days, during which the cumulative leached fraction

increased linearly with the square root of time.  Other effects,

such as those due to the penetration of the leachate in concrete,

or the diffusing of the iodine, are not yet apparent.  Concrete

dams are known to hold water and recent Japanese work has shown

that it takes one year for sea water to penetrate 1 centimeter deep

into a concrete block. *•  Obviously, the leaching experiments

did not extend over a time period sufficient to penetrate the 5 cm

height x 5 cm diameter samples that were used.  Further studies are

necessary to define the long term leach rate of iodine from concrete

in order to arrive at meaningful conclusions.

     Second, the leachate used in laboratory tests was distilled

water.  Tests are currently in progress with leachates comparable to

those that are found in nature, using salt water, brines with various

                                          •5 r\
concentrations of minerals and well water.    These tests show that

the leach rates are much smaller than with distilled water.

     Third, there are doubts on the inverse dependence of the leach

rate on the volume to surface ratio.  It is deemed possible that this

dependence breaks down at small volume to surface ratio, due to depth

of penetration effects.


     Fourth, the leaching process is different in stagnant or

flowing waters.  The influence of the flow speed of the leachant has

to be considered.

     It is not inappropriate to conclude that current leach rate data

cannot be used to adequately predict 1-129 concentrations in ground

water streams leaching the concrete blocks.  Much work is necessary to

define the various physical phenomena involved in the leach process.

However, there are inherent difficulties and uncertainties of project-

ing leaching or any other effects extending over time periods compar-

able to the half-life of 1-129.  Accordingly, the pursuit of further

research must be weighed against the benefits to be derived.  It is

recommended that the methods of immobilization of 1-129 be considered

only as an interim means of containment and not by themselves as a

form of isolation of this radionuclide from the biosphere.

     4.1.2  Immobilization of Iodine in Zeolite

     As noted in section 3, zeolites can be used to separate iodine

from the waste.  The zeolites can also be used to immobilize iodine

collected by other processes.

     Silver-exchanged zeolites were found to have a high chemisorption

capacity for elemental iodine in gaseous streams.32  The maximum

iodine chemisorption capacity of silver zeolite at 150°C was found to

be 214 mg of iodine per gram of zeolite, based on a dry bed density

of 0.85 g/cm .  This is 60 percent of the stoichiometric capacity

based on, the number of silver sites per gram.  Because silver is a

valuable commodity, tests on desorption of iodine from the silver for

recycling used silver beds and of chemisorption of iodine on lead

exchanged zeolites have been performed.33,34  jn this latter pro-

cess, it appears that a compound with a chemical formula approaching

Pbl2 is formed, which is then chemisorbed in the zeolite and is

stable at 150°C and remains kinetically stable once cooled to room

temperature in the presence of air.  A loading'of 317 mg. 12/g °f

lead exchanged zeolite (PbX) is possible and represents 88 percent of

the stoichiometric capacity based on the number of lead sites per


     The annual disposal of 600kg of iodine per model reprocessing

plant would require about 1900 kg or 2.2 cubic meters of zeolite per


     The zeolites proposed for use for both silver and lead are

silica mixed with alumina with a silica/alumina ratio of 5 to 1.

As an example, when combined with Pbl2, 40 percent of the resulting

weight is Pbl2«

     No research on the leaching of iodine from zeolites was identi-

fied in this study.  It is noted, however, that current tests to fix

iodine in zeolite are made with X-type zeolite.33  This zeolite is

not acid resistant.  Since ground water streams are often acidic, it

would appear desirable to investigate the Z-type zeolite (Zeolon)

which resists attack by acids.  The zeolite could be utilized as the

waste form for the disposal of 1-129, or zeolites?could be mixed with

cement to form a concrete.  Zeolite usually appears in the form of

10/20 mesh granules, i.e., they have an average diameter of about 2mm.

These granules have a crystalline structure, are hard, and are ex-

pected to mix readily with cement.  The casting of the.-iodine contain-

ing zeolites in concrete may provide a worthwhile additional barrier

t'o the transport of iodine.  However, experiments to investigate this

method of containment do not appear to have been made to date.

     No leach tests are available on iodine loaded zeolite incorpor-

ated in concrete.  It is estimated, however, that the solubility of

lead iodide (Pb^)  and silver iodide (Agl) , when adsorbed on

zeolite, is about one half of the solubility of these salts when

pure.-"  Research to investigate the immobilization of' iodine loaded
zeolite in concrete and tests on the leachability of the iodine com-

pound from the concrete matrix may be desirable in developing an

improved waste form.

4.2  Tritium

     In principle, there is a choice between chemical storage and

containment.  In chemical storage, a hydrogen-containing compound is

utilized as a solid storage medium.  In containment, tritiated

water is stored in its unaltered form.

     Chemical storage technologies include the use of polymer

impregnated hydrates (such as drying agents or hydraulic cements),

organic compounds (such as polyacetyl^ne, bakelite analog polymers,

polyacrylonitrile, polystyrene) and hydrides (especially zirconium

and titanium hydrides).  Characteristics of these compounds are

presented in Table IX.

     Containment can be accomplished by utilizing high-pressure steel

cylinders or large above-ground storage tanks.  The storage can be

either interim storage for low level waste (it takes 100 years  for

  _ Q
10   Ci/liter water to decrease to the maximum permissible

concentration (MFC) in water of 3 x 10"^ Ci/ml) or final storage for

high-level concentrated waste.  Table X shows the properties of

contained storage options.  Table XI gives a summary of current

storage practices.

     4.2.1  Polymer Impregnated Tritiated Concrete (PITC)

     Leaching of Tritium from Concrete

     Leaching tests have been conducted at the Savannah River Plant

Plant on PITC prepared at Brookhaven National Laboratory.36  xhe

block was lowered into the ground and leached by rain water.  Leach

rates obtained were less than those obtained in static leaching when

the block was fully immersed in distilled water.  To be conservative,

the results of the static leaching test (which was more severe) will

be cited, as these results are more applicable to concrete immersed in

a continuously flowing stream of water.  Figure 5 shows the tritium

release as a function of time and the same curve corrected for decay.

The release rate is:

  ^IL  pL)  / 1)  = 8.61 . 10"  cm/day
n  "  '"'  "-n

                                                      TABLE IX

1. Hydrates
•) Drying agent*

Celcium eulfate
SI lie* gal

Activated alula*

Molecular sieves

b) Uydraullc Cement*
Portland cement

!. Organic Compounds

•akallte analog polymers



1. Hydrides
Zirconium. ZrHj ^

Titan !*•
Water Loading wt Z





Hydrogen (fern 0.6Bg of
water can b*j Incorpor-
ated to Ig of poly-

Hydrogen from O.lSg of
water can be fixed In
Ig of polyacrylonltrlli
O.Sg of water per g of

l.S wtl II (O.Ug of
trltiated water)

Tritium Release Bate


-1 -4
10 -10 /day when en-
capaulated In poly-
10 /day f first month)
2.5 » 10 /day (when
aepholt coated)
3 x 10-4/day (wlien 1"
asphalt cast around)
10-%/day when polyner
la^regnated (5-15 wtZ)

41 Initial., none

21 lose la rinsing.
none further

51 la Initial rinsing

tot SMaaurable

5 « 10~*/year la NaOU
4 x 10-S/yesr la diet.
2 x 10-*/yeer la HC1


Low cost
Only at low loading*

Stable with respect
to beating, water.
various cliesilcale

Thermally stable to
)2S°Ci Insoluble In
all solvents

Stable to 230°C; In-
soluble In various

Therul condones t Ion
above 200°C; degraded
by alkali
Degradation above 2SO°C

Stable below 30O°C
Zirconium sponge
5.SO-1J.OO »/lb la
Stable below 30O»C

Avallahlllty of

Available for removing
UTO vapor

Cncepvulatlon In con-
crete and poly more also
Polymer Isfiregnatioa
requires developnent

Available, coatings
available; polymer Im-
pregnation requlrea


Acetaldehyde production
available, polymeriza-
tion also requires
Available. Furlflca-
catlon neceasary.
complex, expansive
Available! problems la

Available; development
of facility able to fix
JO Kg UTO/ Jay require*
^ y«
                                                     TABLE X


High Preasure Steel Cylinders
Type IH (for high level concentrated
Above Ground Storage Tanks
(for low level diluted waste)

1.2 x 107




Carbon steel construc-
tion, protection froa
Ice formation by In-
sulation and tank
heaters; can be an


control required.
Available, corrosion
control by pH adjust-

Corrosion, 'material
Radiolysla producing
hydrogen, oxygen and
Sudden rupture

Temperature control,
•onitorlng, chem-
ical pll adjustment
Monitoring, cata-
lyst to promote
Pumping to reserve
Double contain-
ment, transfer lines
Secondary contain -
     Source I  ERDA 76-43, Vo. 2

                                           TABLE XI

  Container Disposal  Method

 Uranium titride

 Tritiated materials  and wastes
    .11         ii        it     ii
     (long terra  storage)

 High level Q200 Ci/drum  to
  >1000 Ci/drum)
 Intermediate  <1000 Ci drum
 Low level  <10Ci/drum
SS containers  with  valves

Double 0-ring  sealed  anodized aluminum continers

Welded SS cylindrical continers

>20,000 Ci HTO absorbed  on  2 kg dessicant in sealed
metal cans and concrete  and _>50,000 CiHTO in welded
5" x 11" SS vessels coated  with asphalt and packaged
with vermiculite in an asphalt coated 30 gallon 17H

Sources cast in plastic, molecular sieves in 6" dia-
meter aluminum conduit

Plastic bags,  cardboard  boxes, drums burial in ground
 High  level  MOO Ci/drum
            >10 Ci/liter
  (Vacuum pump oils), water from
  inert  gas  purification systems)
 Intermediate  and  low
  >1 Ci/liter; >AO Ci/liter
Collection of HTO on absorbent,  packaging  in  plastic
bags and cand for burial,  or solidification of HTO  >1000
Ci/liter on adsorbent or with plaster-cement  mixture
in polyethylene containers inserted  in 30  gal. metal
drum, and filled with asphalt

Solidification with cement, absorption on  vermiculite,
all in metal drums.  Absorption  on absrobents.   Disposal
of contaminated objects such as  pumps,  in  asphalt lined
concrete filled drums.  Solidification of  low level
(  0.1 Ci/liter) on pallatized corn  cobs collected  in
polyethylene lined streel  drums  for  burial

                                      TABLE XI  (Concluded)
    Form                                    Container Disposal  Method
                                  Converted to HTO for absorption on drying agent and disposal
                                  as a solid
                                  Discharge of low level waste gases (  <0.01Z  H) through zeolite
                                  to collect HTO  before exhaust
                                  Tritium oxidation and collection on zeolite  or molecular sieve
                                  Elemental tritum:  SS tanks at pressurs  <2 atm
Source:  Rinehammer,  1973, pp. 329-336

w  10
I  10
            1  i  I  I i i i
                                              i   i  i
                         8.61 x 10" cm/day
                                      NO DECAY
                       CORRECTED FOR DECAY
            i   i  i  i i i i
                       10               10
                         LEACH TIME,  years
Source:   Colombo (1976), Reference 36.

with the same definitions as in section 4.1.  The specimen volume

to surface ratio is 4.545 cm.  Total release of tritium from the

specimen is estimated to be complete after 145 years.  The maximum

activity present in the environment occurs after 17.7 years and is

equal to 4.45 percent of the initial tritium activity fixed in the

block when the immersion in water is accomplished.

     The costs of high-level tritiated waste fixation in PITC have

been estimated.^7  A hypothetical installation disposing of 1200

liters/yr of high-level tritiated water in PITC has been estimated

to have operating costs of $6,327/yr including containers, formu-

lation polymerization, labor, freight, burial, and handling.  This

amounts to $5.27/liter of tritiated water.  Capital costs have not

been estimated.  The major parts of the facility include a tritium

storage and filling station, a drum tumbling station, a monomer

storage and filling station, a water bath curing station (if

required), a drum filling station, a transport cart with load cell,

and an overhead crane, which is normal equipment in the chemical

and building industries.

     Monsanto Tritiated Liquid Waste Packaging

     An improved method for packaging tritiated liquids for burial

was developed at Monsanto Mound Laboratory.™  The burial package

is prepared by inserting a 27-gal polyethylene drum into an asphalt-

coated 30-gal steel drum.  The polyethylene drum is filled with either

81 kg (90 liters) of a 3 to 1 dry mixture of plaster and cement


for tritiated water waste, or 9.5- kg (90 liters) of vermiculite or

absorbal for organic wastes (pump oils).  A recommended maximum of 35

liters of tritiated water (or 28 liters of tritium contaminated pump

oils) can be enclosed.  The polyethylene drum is sealed and the void

volume above it is filled with asphalt.  The steel drum lid is then

sealed in place using a sealant and a bolted clamp ring.

     These packages were tested in running water to determine the

tritium permeation rate.  Based on the test results, it is concluded

that the amount of tritium released from the package to the ground

water each year would not exceed 0.0001 percent of the total tritium

contained in the package.  Since there is a 5.5 percent natural decay

each year, the projected maximum tritium released during 85 years of

burial would be 0.002 percent of the total tritium in the package, or

1.6 Ci from the 70,000 Ci (recommended maximum) package.

     4.2.2  Organic Compounds

     Methods for industrially fixing tritium in bakelite (resorcinol

or phenol acetaldehyde formaldehyde), polyacrylonitrile, and polysty-

rene have been studied.3'  Leaching^tests (Table XII) show initial

release of tritium of the order of a few percent during rinsing,

except for polystyrene.  No further loss was detectable over a 4 to 6

week period of testing (the sensitivity of the measuring equipment was

not indicated).  Although this shows the organic compounds route is

promising in this respect, more extended tests with large samples

should be conducted.

                                TABLE  XII

                        LEACHING DATA OF POLYMERS
                 Total 3H Content          2  Total 3H
3HContent  of    Exposed  per/ml  Rinse    Content  Lost
 Polymer (a)     _ Solution _   During Rinse (b
            2874 d
Poly (aery-            .
lonitrile)  7650 d min  g

ated Poly-           -1-1
styrene     690 d min" g"

methane           —

on Rh/Al203       -
                   1430  d  min
                                5130 d min
                                 794 d min
                                                                      Z 3 H Leached
                                                                    Initial Rinse
   Determined from combustion analysis of polymer and scintillation counting
   of resulting water.
   Determined by scintillation counting of rinse water.

   Only during first three days, none thereafter.
Source:  Franz and Burger  (1975,1976),

     Unfortunately,  more recent work reports difficulties in develop-

ing tritiated bakelite and acrylonitrile;41 therefore, two new

methods of preparing polymeric media were investigated. ' One method

uses polyurethane/polymethane copolymer, which gives a loss of 6

percent of activity  in rinsing with' no further release 3 days there-

after.  The second method tritiates polystyrene with a rhodium-on-

alumina catalyst, which offers a one-step fixation procedure.  The

resulting material is inert to exchange of hydrogen and is already in

a polymeric form for storage.

     Material costs  have been estimated for polymeric media by the and

are shown in Table XIII.  The estimated cost of isotope separation of

tritium is also included in the table.  Polyacetylene costs have been

estimated by Colombo^ (based on laboratory experiments) and the

polymer impregnated  concrete costs are more recent estimates.-^'  Deep

well injection costs are estimated ^ ag we^ ag long term tank

storage and isotope  separation costs. ^    There are no process

cost estimates.

     4.2.3  Hydrides

     At Battelle, zirconium hydride has been investigated as a storage

medium for tritium.  5  Conditions of preparation have been developed

such that hydrogen to zirconium ratios in the range of 1.5 to 2 are

obtained.  Pure hydrogen gas at 760 Torr pressure was used in the

reaction, which involved temperatures of the order of 630°C.  Samples

                                  TABLE XIII

Resorcinol- formaldehyde
acetaldehyde polymer
Phenol- formaldehyde-
acetaldehyde polymer
Polymer impregnated
Polymer Weight/kg
Water Disposed
5.9 kg

kg Water
$ 4.50
Gallon Water

Isotope separation^             	              0.18-0.23       0.68-0.87

(a'Isotopic concentration of ^H by a factor of 100.

Sources:  References 37, 39, 41, 42 and 43.

of irradiated zircalloy 2 and zircalloy 4 cladding from nuclear reac-

tors were also found adequate for hydrogen storage.  There appears to

be practically no combustion hazard for zirconium hydride. ^

     Results have been obtained of tritium leach tests on zirconium

hydride in various solutions extending over a period of 1 year (Table

XIV)34.  There was no detectable release during the first 6 months.

During the period from the sixth to the twelfth month, releases of 2.3

x 1CT5 to 6 x 10~5 (for distilled water) and 1.6 x 10~4 (for

HC1) of the tritium inventory were measured.  In another case (NaOH

solution), the zirconium hydride sample fell to the bottom of the

vessel during the sixth month and was crushed by the stirring bar.

Subsequent to this accident, which greatly increased the surface to

volume ratio, the fraction of activity released in the next 6 month

period was 5.5 x 10~4 of the initial tritium inventory.  It appears

that hydrides are an adequate storage medium for tritium, yielding

small leak rates.

     The atomic weight of Zr is 30 times that of H-3.  It has been

shown that each atom of Zr can adsorb between 1.5 and 2 tritium atoms.

Thus the weight of zirconium needed to fix tritium is in the range 15

to 20 times the weight of tritium to be fixed.  One curie of tritium

weighs 1.03 x 10~4 grams as T2.  One curie weighs 1.37 x 10~4

grams as HT.  Assuming that 10 reprocessing plants are operating in

the U.S. and generating comprehensively 1.25 x 107 Ci/year, the

weight of tritium released would be 1.3 kg/year.  This amount of


                                TABLE XIV
(No T)
Distilled Water
Sat. KC1
Sat. NaCl
Sat. KC1
HC1 (pH 4)
Dist. H20
Dist. H20
NaOH (pH 11)
Sat. NaCl
Counts per Min. per ml of Liquid*
6/1/74 6/13/74 10/1/74
31.4 31.5 31.5
31.8 31.8
31.8 31.6
30.4 30.9
30.9 31.3


 *Total Tritium inventory equivalent to 1.3 x 10^ CPM/ml

**Sample pulverized after October 1, 1974.

Source:  Colombo (1975), reference 36

tritium could be fixed with 20 to 26 kg of zirconium, but to achieve

this practically all tritium should be in chemically pure gaseous

form.  In practice most of the waste is tritiated water mixed with

ordinary water from which it is impractical to chemically and

isotopically separate the tritium.   In those cases where the tritium

can be isolated, fixation in hydrides appears to be very promising.

4.3  Krypton-85

     As a result of collection procedures, krypton gas would be

available in almost pure form, with at most a small admixture of

xenon.  The gas can be immobilized  by a number of physical and

chemical fixation technologies being developed.

     4.3.1  Pressure Vessel Containment

     One technology that already exists is the pressurized cylinder-

which has been used to store compressed industrial gas for at least 50

years.^°  Several thousand of these cylinders have been tested by

Union Carbide in normal usage over  many years and extrapolation of/

their results indicates a useful life of 500 years.  It was learned

from Union Carbide that current use of as many as 5 million cylinders

indicates that their failure rate,  including leakage, is probably far

less than the rate of one in 500 per year, thus the 500 year life can

be regarded as a minimum.^'  This assumes, however, normal room-

temperature use.  For long term storage of Kr-85, it is estimated that

a period of 100 years is sufficient in that the krypton released would

be a small fraction of the allowable quantity in 40 CFR 190.  However,

the steady radioactive decay would liberate an  appreciable amount of
thermal energy.  Figure 6 shows the course of this heat generation in
a mole (85 gms) of Kr-85 or about 1.42 kg  (375 liters) of krypton gas
with a 6 percent Kr-85 constituent.^
     The low-alloy high-strength steels used for high pressure cylin-
ders exhibit strain aging in the temperature range 150 to 370°C. that
results in a tensile strength increase.  HoweVer,  above 370°C., the
yield and tensile strength decrease with increasing temperature.
Hence, the maximum temperature of the storage cylinders must be held
well below 370°C.
     Estimates of the storage capacity were made under the assumption
of presently available technology.^6  Heat transfer was calculated
for cylinders, cooled only by natural convection (21°C ambient air),
containing 6.0 percent Kr-85 in krypton gas.  The results for 50-
liter volume cylinders are shown in Table XV.
     The number of storage cylinders required for the annual krypton
production from the 1500 metric ton/yr model reprocessing plant is 100
at 500 psi and 29 at 2000 psi.13  At 500 psi, 12800. curies of Kr-85
are contained in each cylinder and at 2000 psi 41900 curies of Kr-85
are contained in each cylinder.  Hence, the higher pressure poses a
greater adverse effect associated with the risk of cylinder failure.
At the higher pressure, between 1500 and 2000 such cylinders could be
in use by the year 2000.  If there is an admixture of 25 percent xenon
by volume in the gas as anticipated, these estimates would be raised
by one-third.

   2000 \	->
   1500 _
   1000 _
    500  _
                      20      30

                     TIME (Years)
 Source:   Christensen,  Reference 48.

                                          TABLE XV

                              ANNUAL Kr-85 STORAGE REQUIREMENTS
                            FOR A 1500 MT/YEAR REPROCESSING PLANT

Storage Method           Pressure    Amount of Sodalite  Storage Temp.  Number of
                            Psi               Kg               °C       50 liter Cylinders

                            500                                60*               100
Pressured Cylinder         2000                               127*                29

Sodalite Encapsulation*                     2800              120++               82
 *Wall temperature
 ^Assumed loading 1.8x10"^ mole (40 ml STP) Kr per gram of sodalite
*"+Mean temperature:  Center-line temperature is 150°C

     An uncertain danger posed by simple gas storage is the steady

accumulation of the alkali metal rubidium, the Kr-85 decay product.

Table XVI indicates rubidium production of 327 gm and 1070 gm,

respectively, in the 500 psi and 2000 psi cylinders after 150 years'

decay.  As long as the cylinder is intact, the inert krypton atmos-

phere will prevent the rubidium chemical reactions to which it is

normally prone when in contact with moisture or 02«  However, the

liquid rubidium may still attack the steel of the cylinder by removing

carbon and nitrogen from the grain boundaries.  This effect has been

identified for liquid sodium in the stainless steel tubes of LMFBR

cooling system, but these effects are detectable only at temperatures

above 450°C.  Hence there is a reasonable likelihood that at the low

storage temperatures which define mechanical stability for the

cylinders, the chemical corrosion danger of rubidium will be absent.

By the same token, the danger of excessive temperatures poses a hazard

both in the potential for chemical reactions by the rubidium and the

increased risk of containment failure through increased pressure.

     4.3.2  Zeolite Adsorption

     Additional safety benefits such as sharply reduced adverse

effects of a cylinder failure and lower storage costs are obtained if

krypton is immobilized in a solid form prior to encapsulation.

Containment of krypton in zeolites has reached an advanced stage of

development for this purpose.

                 Table XVI

   500 psi cylinder                2000 psi cylinder
Kr-85 (Ci)
Rb (g)
T (8C)
Kr-85 (Ci)
Rb (g)
T (°C)

     A zeolite suitable for krypton is basic sodalite which normally

contains some NaOH as part of the crystalline atomic array.  Its

formula is:

               Na6 (Al6Si6024)  x NaOH- (8 - 2x) H20

with each NaOH replacing 2 ^0  molecules of the ideal sodalite

hydrate.  The sodalite interstitial cages (Figure 7) for krypton sites

are truncated octahedra with about, 6.6 A free diameter for an

inscribed sphere.   At room temperature, the krypton atom occupying the

site has a diameter which can be inferred from classical gas theory to

be 3.5 A.  The "cage" site position represents a relatively strong

free energy minimum for the krypton atom in the zeolite crystal

lattice.  Activation energy for mobility of the atom across cage sites

is high enough so  that after diffusion of krypton into the lattice at

the moderately elevated temperature range of 300°-400°C, the krypton

is effectively trapped when the zeolite crystal is quenched below

150°C.  Storage requirements for this kind of kryptoi. immobilization

have been  compared to those for the pressurized steel cylinder in

Table XV.49

     The process of loading krypton into a zeolite is shown in Figure

S.^3  Activated zeolite (interstitial water removed) is loaded into

the pressure vessel and heated  to the adsorption temperature.  Krypton

is introduced from a cylinder at the encapsulation temperature, the

temperature is lowered after the desired approach to equilibrium

adsorption in the  zeolite, and  the unadsorbed krypton is left in the


                       IN ZEOLITE
Source:  Knecht,  Reference 13.
                          OF Kr IN ZEOLITE

storage cylinder.  Tests are being conducted on this process to

determine the amount of krypton gas which can be so encapsulated as a

function of time, pressure, and temperature, and also to determine the

leakage rate for inference of leakage at storage temperatures over

long periods of time.  The most obvious physical advantage of this

kind of trapping in a solid is the continued isolation of the possibly

corrosive rubidium decay product.

     If one krypton atom occupies each cage of sodalite, the satura-

tion capacity for ideal sodalites is 52.6 cm^ (at STP) of krypton

gas per anhydrous gram of sodalite.  Tests of equilibrium isotherms of

Kr, as a function of pressure for the amount of Kr sorbed on sodalite,

conform closely to the shape of the Langmuir formula for fraction of

saturation capacity sorbed in a perfect sorbent.  Test data at diffu-

sion temperatures between 326°C and 544°C^" indicated a saturation

sorbency of 45 cnr/gm, in fair agreement with the theoretical value.

The results were similar both for sodalite with intercalated NaOH and

for the NaOH removed by extraction.  The equilibrium capacities did

not change, indicating that krypton could occupy a cage containing

NaOH.  It is therefore anticipated that 1 kg of zeolite can trap 45

liters of krypton at STP for a total requirement of 3180 kg/year

zeolite for one reprocessing plant.

     Diffusion data at high temperatures was used to measure the

activation energy and diffusion constant to evaluate that leakage

rates at low temperatures.  For a fractional leakage Qt/Qco at

temperature T, the diffusivity:
                 rr 2          d
                   o           _

                  36      '        dt
where ro is an average diffusion path length related to the size  of

zeolite or sodalite crystals (Qt = quantity diffused in time  t; Qg., =

quantity diffused after an elapse of an infinite period).  The values

of Qt/Qoo were measured as a function of time at temperatures  between

413°C and 560°C for a number of samples using a mass spectrometer  to

detect the leakage of krypton.  This determines the temperature

dependence of D which indicates both the diffusion constant DQ and

the activation energy E via the equation.

                  D = D  exp (-E/RT)

were then used to calculate the long term leakage at 150°C shown  in

Figure 9.  The graph shows the effect of radioactive decay of the

entrapped krypton upon the leakage of the radioactive species from the

sodalite.  The shape of the curve for Kr-85 is represented by the

                                        exp  /_  E  _ \t
where the factor exp(-Xt) accounts for radioactive decay.  The

fractional leakage of krypton at 150°C and 100 yrs is 2 x 10~2.

If the cylinder containing sodalite ruptures at this time due


                           100       150        200

                                   Time (Years)
        Source: Knecht,  ICP-1125, kelerence 13.



to Kr-85 decay, only 3 x 10~^ of the original Kr-85 will escape.

This is shown on the Kr-85 line.  After eight years the fractional net

leakage of Kr-85 is a maximum of 3 x 10~3, which rep-resents a safety

factor of 200 for sodalite encapsulation relative to pressurized tank


     4.3.3  Ion Implantation/Sputtering

     Another method for immobilization of krypton in a solid is ion

                                                              1 ^
implantation or sputtering on crystalline or amorphous solids.10

Krypton can diffuse into a metal surface (at high temperature and

pressure) to a depth of _£10   cm and occupy a lattice position.^

At loadings far below saturation, the amount of Kr sorbed, V, varies

linearly with:

     •  pressure

     •  t, where t is time

     •  exp (1/T), where T is the temperature

     More than 25 different powders, foils, and other solid forms have

been used to sorb krypton (including copper, aluminum, iron, nickel,

gold, silver,  etc.).  Loadings achieved at tens of megapascals* (200-

800°C and  5-90 hours) were typically far from equilibrium in the

range 10~2 to 10~6 cm3 g"1.  Tests at the Idaho National

Engineering Laboratory (INEL) using } x 10~6 meter aluminum powder

at 510°C, 190 MPa** and 24 hours resulted in less than 1 cm3 g"1

 *Pascal (Pa) = Newton/meter2

**MPa = 106 Pa

     The results of the INEL test are shown in Table XVII.  Total
amounts of metal required are in the range of millions to billions of
metric tons.
     In high temperature/high pressure sorption, krypton is imbedded
by diffusing into a solid under a concentration gradient.  An alter-
nate method of imbedding krypton uses electrostatic energy to accel-
erate krypton ions (at low pressures,  0.1 to 100 Pa) into a receptor
surface as shown in Figure 10.^0  After the receptor surface has
been loaded, the polarity can be reversed and krypton ions are accel-
erated to strike a target surface.  The metal atoms which are sput-
tered from the target by the energy of ion bombardment are deposited
on the receptor, yielding a "clean" surface into which additional Kr
ions can be implanted.  The process can be operated to maximize
implantation in one target and sputtering in the other.  The trapping
efficiency is a function of Kr ion penetration in the receptor and of
the probability that the Kr ion is accommodated in the crystal struc-
ture (in voids or vacancies) and increases with increasing energy,
approaching a maximum in the kilovolt region.  The atomic number and
structure of the substrate have a variable effect on trapping, and its
temperature has an inverse effect on trapping.  Purity requirements of
krypton have not yet been determined; potential effects of the xenon
impurity must be measured.  Xe with ionization potential of 12.1 eV
can interact with krypton ions (Kr ionization potential is 14.0 eV)
resulting in xenon ions and krypton atoms; since Kr must be in ionic
form to be trapped, a decrease in efficiency could occur.


	 „/ j.
O 0 O O 0 O 0
0 0 O
 00000©  0
                        I	I
© 85Kr  IONS



.. -9 ]rV
                            © 85Kr  IONS
                            • SPUTTERED METAL ATOMS
                            o IMPLANTED  Kr
Source:  Check,  et.al., Reference 50.

                                                TABLE XVII

                                     SORPTION OF KRYPTON GAS BY METAL
Required Wt.
 of Metal*
Diffusion under high temperature     3.7 x 10"^     1.4 x 10*> Tons

and pressure                            to              to

10       100  MPa*                   3.7 x 10~8     1.4 x 109 Tons


5        90 hrs.

Ion Implantation/Sputtering at       5.98 x 10~2    8.91 Tons

low pressures (.1      100 Pa)

               "                         .475       1.12 Tons
     *Required  for 1500 Mg/yr reporcessing

     +Meghpascal  = MPA = 106 N/m2 = 9.87 atmospheres

     Studies of this method of trapping Kr-85 are underway in
foreign countries (including the U.K. and Germany) as well as
in the U.S. at Battelle-Pacific Northwest Laboratories.  Load-
ings of up to 4 atom percent (16 STP cm3 Kr g"1) in kilogram
quantities of Ni have been achieved at Battelle using a high
density sputtering system.  Higher loadings (of up to 30 atom
percent or 127 cm3 g"^) have been achieved in thin films of
the amorphous material, GdQ^-Q • COQ^J^ ' MoQ^jg, possibly
due to the pressure of larger interstitial voids in the dis-
ordered materials.  As indicated in Table XVII, for both of
these cases the total weight required for each reprocessing
plant is only a few metric tons.
Other Immobilization Technologies
     Since metals have a large thermal conductivity, good mechanical
strength, and high radiation resistance, they have been-used to form
matrices containing high-level wastes.  One possible process for
forming a metal matrix uses molten metal casting.  Metal matrices that
have been formed by this process contain different combinations of
calcine with aluminum, iron, zinc, lead-tin, aluminum-titanium, or
iron-titanium.-'2  The metal matrix volume is about the same as the
volume of the calcine.  Metal matrices have also been formed by
compacting a mixture of calcine and metal powders and then sintering
the mixture.  Both processes form products with compression strengths
above 26.7 MPa and thermal conductivities 10 to 60 times that of glass
matrices.  However, the process temperatures range between 280°C and

980°C.  As with glass, the high temperatures  (>400°C) of the molten

metal casting process are not suitable for immobilizing solids con-

taining krypton.  It has been speculated that if metals with low

melting points are used, ^ the sintering process could-advan-

tageously immobilize granules containing krypton.

4.4  Carbon-14

     The only technique currently being studied for the immobilization

of C-14 is the formation of calcium carbonate (CaCO^) and subsequent

incorporation into concrete or other material.

     4.4.1  Concreted CaCOit

     The concreted CaC03 product can be considered to be low speci-

fic activity (LSA) material.  Such materials have only the very mini-

mum of package requirements, usually referred to as Type A, which can

use a wide varity of readily available metal or fiber drums, wooden

boxes or fiberboard boxes, and must meet conditions principally

consisting of various drop tests.

     The concrete holds about 30 weight percent CaCC>3.  Thus the

annual output of 1.36 kg CaCC>3 from a model reprocessing plant will

be incorporated in 4.5 kg of concrete  (O.OOSrn^) as a yet undeter-

mined multiplication factor due to the absorption of carbon from the

atmosphere which would raise this by a factor between 1 to 100.

      The standard 55-gallon steel drums (which sell for about $10

each) would seem to satisfy the Type A packaging requirement for

CaC03 containing the radioactive C-14.  The concretion of CaCC>3

and packaging in steel drums has been estimated to cost approximately

40 cents/gallon or approximately $1.00/m3.

     4.4.2  Leaching of C-14 from Concrete

     Experimental data for the leaching of C-14 from concreted CaC03

are not available.  However, the solubility of CaC03 in pure water

is about  1.4xlO~7 kg mole/litre.  This is about a fifth of the

value for 83103, whose leaching properties were discussed in section



     Radioactive waste is the inevitable byproduct of nuclear elec-

tric power generation.  The annual projected nuclear power capacity

in the U.S. varies from 400 to 1000 GWe by the year 2010.  While

radioactivity is encountered at most  stages of the nuclear fuel

cycle, the largest quantities and those of potentially greatest con-

cern are those present in spent fuels.  Using the projected nuclear

capacity in the U.S. by the year 2010, the corresponding commercial

waste burden of spent fuel is estimated to be between 3.1 and 7.7 x

10^ metric tons of heavy metal (MTHM) per year.^

     Under the reprocessing option, spent fuel elements from nuclear

reactors are reprocessed to recover usable uranium and plutonium.

During these chemical processing operations, radioactive particles

and volatile materials are released to the off-gas effluent streams.

Control approaches have been utilized by the nuclear industry to

maintain radiation emmission levels of these volatile radionuclides

below applicable standards.  Although present control technology is

capable of meeting or exceeding required standards, the disposal of

volatile radionuclides which will be  produced in evergrowing

quantities will be a matter of concern.

     The primary source of volatile radionuclide release in the

nuclear fuel cycle is spent fuel reprocessing operations.  The

exception to this would be in the case of U.S. policy decision not  to

perform reprocessing and to directly dispose of spent fuel elements

without recovery of uranium or plutonium.  In this event, the primary

release of the volatile radionuclides would occur sometime after

disposal if the integrity of the fuel elements and other engineered

containment barriers were to fail.

5.1  Disposal Concepts

     In previous sections various methods of collecting and immobil-

izing volatile nuclides were discussed from a technical and

environmental impact perspective.  This section discusses final

disposal options which are feasible for volatile radionuclides.  A

brief description of each disposal option is summarized below.

Radioactive waste disposal options are discussed in detail in

references 3 and 54.

     5.1.1  Geological Repository

     Disposal of radioactive wastes in deep, stable geologic forma-

tions has long been the preferred method for isolation of wastes from

contact with man's environment.  A number of possible geologic media

have been considered for such disposal.  These include salt beds,

salt domes, crystalline rock forms such as granite or basalt, shales,

limestones, and certain types of clay beds.  To date, salt deposits

have received the most attention as a suitable medium, 'especially in

the U.S., because of their demonstrated stability over very long

time periods, their homogeneity, and their property of plastic flow

and selfhealing in the presence of stress.  The self-healing

properties of salt effectively eliminate  the possibility of extensive

cracking, thereby preventing the opening  of pathways to radionuclide

migration in the environment*

     An alternative to salt is stable crystalline rock, such as

basalt or granite.  Again, crystalline rock is a suitable candidate

with demonstrated seismic stability.  Crystalline'rack, however, does

not have the self-healing characteristics of salt but possesses other

advantages, including resistance to water intrusion, that make it a

desirable medium for geologic disposal of radioactive waste.

     Shales and clay deposits have also been considered for geologic

disposal.  They  have the advantage of low water permeability, but

the disadvantage of indeterminate  long term stability character-


     The effectiveness of geological repositories to isolate volatile

radionuclides depends on two factors:

     (1)  the form of the waste material  and its resistance to


     (2)  the location and design  of the  geologic disposal  facility

          to achieve maximum isolation from the environment.

More specifically, there are several important factors which deter-

mine the effectiveness of geological repositories to  isolate  the

volatile radionuclidea:

     •  the form of the  waste products

     •  the type of containment

     •  the resistance of the waste matrix to leaching.

     »  the solubility of the leached radioactive elements in ground

     All of these factors affect the rate at which water might trans-

port radioactivity from the repository.

     To date, the major  thrust of analyses of engineering controls of

geological repositories  has been limited to salt deposits, because

this is the only type of geological medium for which extensive

information is available and because of their demonstrated stability

for long periods of time.  For other media such as shale, basalt, or

granite, the data available are limited in scope.

     The greatest concern for the migration of volatile radionuclides

through the geosphere to areas of immediate significance to mankind

appears to be related to groundwater movement.  The initial ground-

water flow conditions at potential disposal sites of geological

respositories are extremely important.  The leaching characteristics

of volatile radionuclides are also important to the migration route.

     5.1.2  Seabed Disposal

     Seabed disposal involves the controlled emplacement of vola-

tive radionuclides in deep sea sediments or beneath the bedrock of

the ocean floor.  The effectiveness of seabed disposal in contain-

ing volatile radionuclides depends upon demonstrating that seabed


emplacement can contain the volatile radionuclides long enough for

them to decay to relatively innocuous levels.

     Under the seabed, physical and environmental barriers exist that

may prevent the migration of radionuclides to parts of the ocean that

are of immediate significance  to mankind.  On the other hand, several

mechanisms may act singly or in combination  to compromise the inte-

grity of the physical and environmental barriers:

     •  corrosion of the canister;

     •  leaching of the waste  material;

     •  upward transport through the upper sediment layers to the
        lowest water layers;

     •  advection and diffusion through the  water column;

     •  biological transport of incorporated isotopes across the
        seabed or upward through the water column.

For seabed disposal, it is  crucial  to select an  ocean repository

which has demonstrated geological stability  and  which consists of

deep sea sediments that can act as  an effective  barrier to isotope

migration for geological time  periods.  These requirements are

especially crucial for 1-129 and C-14 because of their  long


     5.1.3  Transmutation

     One of the possible alternatives being  considered  for the

management of long-lived radioactive wastes  is  to  transriute  them into

short-lived or  stable isotopes.  If this  concept is  demostrated  to  be

technically feasible, the quantity of waste containing long-lived

radionuclides could be reduced significantly, and the time required

lor the storage of the waste shortened.

     The process of transmutation is accomplished by using some

nuclear device.  Four types of such devices have been discussed in

the literature:  particle accelerators, thermonuclear or fission

explosives, fusion reactors, and fission reactors.  Each type of

device has to be judged on several criteria including overall energy

balance, overall waste balance, and the rate of transmutation.  A

favorable overall energy balance implies that the energy required to

dispose of the waste should be less than the energy furnished by the

reactor which produced the waste, preferably by an order of magnitude

or better.  A conceivable exception would be when the era of nuclear

fission power comes to an end and there are other plentiful energy

sources available which can be used for the disposal of wastes left

over from that era.  The criterion of overall waste balance is self-

evident:  the waste disposal program should not create more waste

than it removes.  The rate of transmutation depends not only on the

particular device that is utilized, but also the properties of the

target nuclides.

     5.1.4  Extraterrestrial Disposal

     The concept of extraterrestrial disposal involves .launching

radioactive nuclear waste into space or for placement on planetary

bodies without any possibility of return  to earth.  The long-lived

wastes, with half-lives of thousands to millions of years, may thus

be disposed of without concern for  the lifetime integrity of their

containers and respositories.

     Extraterrestrial disposal is expensive.  The  feasibility of

using it for the disposal of tritium and  Kr-85 may be questionable

because  these  isotopes have relatively short half-lives and other

disposal options may be more practical.   Containment for a period of

about 200 years may be adequate  to  ensure that these volatile

nuclides decay to relatively innoccous levels.

     Both 1-129 and C-14 have  lifetimes which are  long enough to

warrant  consideration of methods of disposal other than long term con-

tainment.  Therefore, space disposal should be considered as a possi-

bility for the disposal of  1-129 and C-14.  At the same time, it must

be mentioned that the estimated  levels of exposure from the possible

release  of C-14 from the nuclear power industry are of low magnitude

and  extreme measures to limit  its release may not  be implemented.

Thus, it would appear that  among the volatile radionuclides, 1-129

alone should be seriously considered as  a candidate  for space dis-

posal at the present time.

     By  far  the most serious concern associated with space  disposal

involves launching  accidents and space vehicle re-entry.  An accident

involving  concentrated amounts of C-14 and 1-129  could pose radiation

contamination  hazards  to man and the environment.

     5.1.5  Other Continental Disposal Options

     Disposal of radioactive wastes in deep,  stable geologic forma-

tions is currently the preferred method for isolation of these wastes

from contact with man's environment.  However, there are alternative

concepts to deep geological respositories which are under considera-

tion for different forms of radioactive waste:

     •  solution mined cavities

     •  waste disposal in a matrix of drilled holes

     •  waste disposal in super deep holes

     •  deep well injection

     •  hydrofracture

Long term containment is a major concern in the disposal of the vola-

tive radionuclides and must be assured in all concepts. The major

threat to long term containment is groundwater movement.  The con-

cepts must preclude contact of the was,te with groundwater movement to

minimize waste migration to the biosphere.

     Of the alternative geological disposal concepts considered, the

technology for super-deep holes is not yet developed and the specific

heat of the volatile radionuclides is insufficient for the rock melt-

ing concepts.  The remaining concepts could have application to the

disposal of those radionuclides.

     5.1.6  Ice Sheet Disposal

     Continental ice sheets have been considered as an alternative

approach to the final disposal of high-level radioactive waste.

Theoretically, ice sheets could provide  the necessary 'geographic

isolation for some of the short-lived fission product wastes, how-

ever, the feasibility of ice  sheets' long term containment capabil-

ities is presently uncertain.  Before ice sheets could seriously be

considered for waste disposal  applications, certain-areas should be

further investigated:

     •  the evolutionary processes  in ice sheets;

     •  the relationships of  ice sheets  with climatic changes;

     •  the effect of future  climatic changes on the stability of ice

     Because of  these factors, ice  sheet disposal will not be consi-

dered as a feasible alternative for the  disposal of the volatile

radionuclides in this report.

5.2   Disposal Alternatives for Volative Radionuclides

     5.2.J  Iodine-129

     Iodine-129  is produced in the  nuclear fuel elements as a fission

product and from the xadioactive decay of other short-lived fission

products.  The projected annual release  of 1-129 from spent fuel in a

1500 Mt/yr model reprocessing plant is 380 kg  (66 curies).  The

half-life of 1-129 is 1.7 x 107 years.   It would take approximately

1.7 x 108 year  (10 half-lives ) for 1-129 to decay  to a  low  level'

(i.e.,  less than 0.1 percent  of its original activity).

     The  chemical waste  form  and encapsulation  technology of  1-129

depends upon  the designated disposal alternative.   Currently,

three processes have been investigated for the collection of iodine:

the Mercurex process,  the lodox process,  and cheaisorption process.

These processes chemically combine radioactive iodine into several

iodide and iodate compounds.   It is important to select an iodide or

iodate compound which has demonstrated low leachability.  Among these

compounds—mercurous iodate,  barium iodate, and zeolite—each mixed

with concrete has been shown  to have fairly low leach rates.

     With respect to iodine leaching from concrete, the rate at which

water penetrates concrete is  slow.  Since only a small fraction of

the radioiodine compound is in contact with water at a given time at

the solid/liquid, interface inside the concrete, the dissolution rate

of radioiodine is predicted to be slow.  Experiments have indicated

that 600 kg of iodine in the  form of barium iodate incorporated in a

concrete cube side of 1.4 meters would leach completely after approx-

imately 1.6 x 10^ years.  This time period is only one thousandth

of the half-life of 1-129 (1.6 x 107 years) and as such, over 99.9

percent of the 1-129 activity would exist at the time of complete

leaching from the concrete.  Data on probable periods for complete

leaching of other compounds in concrete is not available, but it can

reasonably be assumed that the immobilization form of 1-129 would be

insufficient containment, and additional barriers are required.  Geological Disposal of 1-129.  If 1-129 wastes were to

be disposed of in a geological repository, it is very important to

have minimal or no natural water movement in the strata and interbeds

of any geologic formation.  Since it is highly probable that in deep

geological repoaitories there will be water-bearing strata either

above or below the potential repository area, water movement in the

potential repository strata and interbedded strata will occur.   Even

if highly selective siting criteria are applied, there can be

extremely slow natural migration of water through the repository


     Because iodine is not sorbed in ground strata as well as other

radionuclides (it has a retardation factor of 1) and because of the

long half-life of 1-129,  there .is some concern that geological

repositories may not serve as adequate barriers for 1-129, particu-

larly for geological time periods.  Most of the 1-129 will still

exist at the time of complete leaching from the containment and may

cause localized radioactive contamination to the land and water food

chain systems.  The 1-129 would, of course, be removed from the human

environment during the period it remained in containment and the time

required to be transported to the biosphere.  Since most of the 1-129

would eventually be released, geological repositories cannot be con-

sidered to offer permanent isolation of 1-129.  The  acceptability of

this concept must, therefore, be determined on  the basis of the

radiological hazard.  The rate of release will be a  significant

factor in this evaluation.  Seabed Disposal.  If 1-129 wastes were  to be  em-

placed in the ocean sediments, the principal requirement  for their

effectiveness  is  that  deep sea clay sediments act as  a suitable


barrier to 1-129 migration for a period long enough for it to decay

to relatively innocuous levels.  As with geological repositories,

1-129 is likely to be incorporated in an iodide or iodate compound in

a stable form such as concrete.  However,  concrete may pose some dif-

ficulties concerning its effects on emplacement techniques in the

ocean sediments.  Specially designed canisters and, accordingly;

waste forms, need to be developed which will penetrate the ocean

sediments in a manner that will ensure hole closure.

     Because of the corrosive nature of sea water, the probable

period for containment failure and leaching of 1-129 is estimated to

be from 100 to 1000 years.  This period becomes important if the

1-129 waste is mixed with short-lived radionuclides for burial.  Most

of the heat generated by short-lived radionuclides occurs during the

first 1000 years of burial.  The heat produced by fission product

decay during this time period may be high enough to affect the effe£-

tiveness of the waste form and canister as a barrier to migration.

 In general, if radioactive wastes leach through the canister while

the deep sea sediments surrounding the canister are at a high tem-

perature, the thermal and hydraulic gradients created by this tem-

perature may cause rapid upward transport of 1-129 through the

sediments.  If the policy is to isolate 1-129 from other wastes and

encapsulate and bury the 1-129 separately, then this problem is

insignificant since the heat generated by 1-129 alone is small.

     Once 1-129 has leached from containment, the two remaining

barriers are the deep sea sediments and dispersion into the ocean.

Specific information on the migration rate of 1-129 in deep sea

dediments is not available.  However, assuming no retardation by the

sidiment itself due to sorption, an estimate of the migration time

can be obtained based on diffusion of the 1-129 through the sediment

process.  The diffusion time, t, is give by

     t * diffusion  time, see
     L » depth of sediment, cm
     D =• diffusion  coefficient  cm2/sec

For a burial depth  of  100 meters  (1 x 104 cm) and using an aver-

age value  for deep-sea sediment  siffusion coefficient of 3 x

10~6 cm2/sec.

     t « (1 x 104)2 ,  3.3 x 1013  gec or  i x  io6  years
           3 x 1
 effects on aquatic and marine organisms may be  small.   However,  there

 is  the risk  that uniform dispersion will not take  place,  and  that  the

 1-129 may become localized in certain regions of the ocean.   Under

 these conditions, marine organisms can accumulate  I-»i29 from  contami-

 nated food,  water and suspended sediments and can  enter man's food


     Seabed  disposal, like geological repository disposal, may not

 assure the isolation of the 1-129 over the period  of time required

 for this isotope to decay to relatively innocuous  levels.  Seabed

 disposal does, however, offer an additional barrier to  transport to

 man.  Further, the diluting potential of the sea is such  that if

 highly dispersed, the 1-129 from radioactive waste disposal may be

 significantly less, approximately 1/100, of the total radioactivity

 in  sea water (3 x lO'11 Ci/cm3). 55

     Seabed  disposal is a feasible option for disposal  of 1-129 in

 that fully diluted was-te wotud be a small fraction of background but

 would require an evaluation of the transport pathways to  determine a

 low risk of  concentrated exposure to the ecosystem.  A  low concentra-

 tion of 1-129 could be attained if a low release of 1-129 is  assured.

     Studies have been conducted on the retention  of iodine in marine

 clay soils.57  The clay fraction constituted 68 percent of the soil

 and contained minerals such as mica, chlorite, quartz,  feldspar; and

 amphobile.   Results showed clearly that at low pH  values, iodine is

absorbed on clay particles.  This property may be utilized to some
extent for the retention of iodine in geological formations as well.
as a marine environment.  Quantitative measurements on various typee
of clay would be useful.  Extraterrestrial Disposal.  Extraterrestrial dispos*!
of 1-129 is technically feasible.  Space disposal offers the long
term benefit of permanent disposal of 1-129 with no interaction with
the ecosystem of the earth.  However, two factors must be considered
in the overall practicality of this concept:   (1) space disposal is
expensive relative  to other alternatives; and  (2) space disposal
poses a short term  risk of accidents with the  potential of radiation
      Assuming that a suitable waste form and  encapsulation can be
developed, the major technological concern  in  extraterrestrial dis-
posal is the potential of accidents.  With  improved launch vehicles
such as the space shuttle and highly developed recovery capabilities
in the event of an  accident, it  is estimated  that  the  probability  of
loss of a capsule containing the  1-129 waste  could be  reduced  to  less
than 10~2.  With highly advanced  encapsulation technology,  the
probability of prompt release of  the 1-129  from vehicle  explosion and
fire or  intact re-entry could conceivably be  in the  range of  10~5
to 10~6-
     The important  factors  in the extraterrestrial  disposal option
are  the  probabilities  of  accidents and  the  recovery  of the waste
capsule  in the event  of  an accident.  Whether or not  space disposal

represents an acceptable risk with the inherent hazards of uncon-

trolled loss of a waste package requires a detailed system concept

evaluation and risk and consequence assessment.  Without such an

analysis, the acceptability (aside from economic consideration) of

extraterrestrial disposal cannot be determined*  Transmutation.   Transmutation of 1-129 is considered

impractical in fission reactors.  1-129 has a thermal neutron cross-

section of 34.5 barns and an effective fast neutron cross-section of

0.24 barns.  At a thermal flux of 3 x 10^ neutrons/cm^sec, it

could take over fifty years to achieve a reduction to 10 percent of

the original 1-129 activity.

     Fusion reactors are capable of producing high neutron flux

levels and it is conceivable  that these devices could be used to

dispose of 1-129.  The fusion reactor is not developed and requires a

major technical breakthrough  before this concept can be considered


     5.2.2  Carbon-14

     Carbon-14 is in both the fuel elements and the cooling water in

light water reactors.  At the present time, most of the C-14 in

spent fuel is released to the atmosphere as C02 during the disso-

lution of spent fuel at the reprocessing plant. A model 1500 MT/year

reprocessing plant releases approximately 0.19 kg of C-14 with an

activity level of 850 Ci.

     Methods for collection of C-14 from off-gas streams include

caustic scrubbing, molecular sieve adsorption, and fluorocarbon

adsorption.  The most probable chemical form is calcium carbonate,

which may subsequently be incorporated into concrete or other mater-

ials.  The caustic scrubbing method has been shown to remove 99 per-

cent .of the C02 initially present in the spent fuel and produces

CaC03.  Assuming 99 percent recovery of an average 850 Ci of

  C02/year, the average annual output of radioactive calcium car-

bonate is around 1.36 kg.  However, 10 to 100 times as much CaCC>3

from atmospheric carbon dioxide would probably be recovered along

with ^CC^, bringing the total amount of waste to be disposed of

per 1500 MT reprocessing plants from 13.6 to 136 kg/yr.  The calcium

carbonate (CaC03) is subsequently incorporated into concrete com-

prising 30 percent CaC03 and 70 percent concrete.  The annual

production of 1.36 kg of CaC03 from a model reprocessing plant  is

incorporated in 4.5 kg of concrete.  Due to the recovery of CaC03

from atmospheric carbon dioxide, the total amount of concrete con-

taining CaC03 disposed of is estimated from between 58 to 580


     Experimental data for the leaching from concentrated CaCC>3 are

not available, thus it is difficult to provide estimates on the prob-

able time period for total leaching of CaCC^.  It is speculated

that this period ranges from 30 to 500 years, depending on the  dis-

posal medium.  Experiments indicate that if CaC03 is suspended  in


pure water in the presence of C02, a small quantity dissolves.  If

this C02 is absent, the CaC03 progressively decomposes to form

Ca(OH)2, which is much more soluble than CaC03.

     C-14 has a half-life of 5.57 x 103 years and will decay to a

low level (0.1 percent of original activity) after 5.57 x 10^

years.  Several disposal options may be applicable to C-14.  Transmutation.  The transmutation of C-14 in fission

reactors is not feasible because the cross-section of carbon-14 for

both thermal and fast flux neutrons is of the order of microbarns.

Even at a high flux of lO^ neutrons/cm^sec, the transmutation

rate is about 10~13sec-l compared to the natural decay constant

of 10~12sec-l for relatively long-lived C-14.  Thus transmuta-

tion may be ruled out as a disposal option for C-14.

      Transmutation in fusion reactors may h,ave some potential, but

the future availability of fusion reactors is uncertain.  Geological Disposal and Seabed Disposal.  Two disposal

options appear to be capable of maintaining long term integrity and

isolation of CaC03 wastes:  geological disposal and seabed


     For the geological disposal option, the most important barrier

to C-14 transport to the geosphere is the proposed salt beds and

underground strata.  Since it would take approximately 5 x 10^

years for C-14 to decay to 0.1 percent of its original activity,  it

must be demonstrated that geological repositories could potentially


act as an effective barrier for this  time period.  A large degree of
uncertainty is associated with values of the geological parameters-
important in geosphere  transport.  All  that can be established is a
range in values for important parameters.  Experiments oa the
migration potential of  C-14 in various  media indicate that C-14 is
transported fairly rapidly relative to  other radionuclides (retarda-
tion factor » 10).  More experimentation on CaC03 is necessary to
establish quantitative  estimates of leacb rates, hydrodynamic disper-
sivity, water migration ratio, and dilution in nuclide concentration
that occurs during migration.  However, analytical models of geo-
sphere transport  indicate that only insignificant amounts of C-14
would be transported  to areas of danger to humans during the period
C-14 decay to 0.1 percent of original activity.^°  Assuming that
the transport models  are confirmed, the conclusion is that geological
repositories are  capable of maintaining integrity and providing
isolation of C-14 for the duration of its significant activity.
     For seabed disposal of CaCC^, it is important to demonstrate
that deep sea sediments (clays) can contain C-14 and prevent migra-
tion to the ocean for 5 x 10^ years.  The retardation factor of the
migration of carbon has not as yet been determined.  It is likely,
however, that deep sea  clays will act as an effective barrier to C-14
migration.  The potential problems of containment in the seabed as
previously discussed  also apply to C-14.  Further research and
development is obviously required to  support the concept of s'eabed

-------  Extraterrestrial Disposal.  Extraterrestrial disposal

of C-14 is technically feasible.  Space disposal offers the long term

benefit of permanent disposal of C-14 but with a short tern risk of

some accidental release as discussed for 1-129.  The costs of space

disposal are high compared to geological and seabed disposal.  If

these latter alternatives are shown to be effective in isolating O14

wastes, is no overriding reason to recommend that C-14 be disposed of

in space.

     5.2.3  Tritium

     Tritium is produced in fuel elements, control rods,  and in the

primary coolant of light water reactors.  The tritium contained in

fuel elements will be treated at reprocessing plants and disposed of

accordingly.  A model 1500 MT/year reprocessing plant releases

approximately 3 x 10  kg of tritiated water from the voloxidation

method with an activity level of 1.06 x 10^ Ci.

     Because tritium has a relatively short half-life (12.3 years),

it can either be stored above ground in large cylinders or chemically

treated for final disposition.  For final disposition, a probable

waste form is polymer impregnated tritiated concrete (PITC).  Based

on a polymer weight/kg of tritiated water of 0.5, the total amount of

PITC which will be disposed of is 1.5 x 10^ kg/year (based on a

1500 MT/year reprocessing plant).  For storage, unaltered tritiated

water must be contained for approximately 200 years (it takes 191

years for 1 x 10~^ Ci/liter of tritiated water to decay to the

allowable drinking water concentration of 2 x 10~8 ci/l (20,000

pci/1).  Leach experiments have indicated that the probable period

for total leaching of PITC ranges from 30 to 200 years.

     5«2.3.1  Geological and Seabed Disposal.  Because of the rela-

tively short half-life of tritium, both geological and seabed

disposal provide adequate barriers to migration for periods long

enough to allow tritium to decay to relatively innocuous levels.

This occurs after approximately 200 years.

     Tritium reacts very little or not at all with sediments and

soils  (retardation factor - 1).  Even if there is total leaching of

PITC within 30 years, the decay period for active tritium is small

compared to the time period it would take tritium to migrate to areas

of significance to man from deep geologic or seabed disposal.  It

decays before it reaches the biosphere.  It is desirable to isolate

the tritium from other waste simply to avoid an additional water

source which could accelerate leaching of other emplaced waste.

     For seabed disposal, deep sea clays act as an effective barrier

to migration of tritium.  It is postulated that it would take tritium

buried 100 meters below the deep sea sediments approximately 10*

years  to migrate to the ocean surface.  This time period is far

greater than the time necessary for tritium  to decay  to relatively

innocuous levels.  Transmutation.  The transmutation of tritium  is not

feasible because the cross-sections for both thermal  and fast flux

neutrons are of the order of microbarns, which is too small to

achieve an appreciable gain over tritium's natural decay rate.

     Tritium is utilized as a fuel in fusion reactors and could be

beneficially used if this technology is developed.  Extraterrestrial Disposal.  Extraterrestrial disposal

of tritium is technically feasible; however, the costs of space

disposal would be high compared to geological and seabed disposal.

These latter alternatives are more suitable for tritium disposal.  Engineered Storage Facilities.  The 12.3 year'half-life

of tritium is such that above ground engineered storage during

radioactive decay is feasible.  The engineered facility must assure

public and occupational safety for normal operation and in the event

of accidents.  A facility design similar, but of less complexity, to

that discussed below for Kr—85 storage (Section would

probably be used.  Capability for recovery of tritium and

repackaging, and decontamination in the event of leaks would be


     5.2.4  Krypton-85

     Krypton-85 is the only noble gas radionuclide formed as a U-235

fission fragment present in appreciable quantities when LWR spent

fuel is reprocessed or in unreprocessed spent fuel stored for ten

years.  The half-life of K-85 is 10.73 years and it takes

approximately 100 years for it to decay to 0.1 percent of its ori-

ginal activity.  The annual gaseous krypton production from a model


reprocessing plant is about 530 kg, which has a volume of 143 m^ at

standard temperature and  pressure.  About six percent of this kryp-

ton is Kr-85 with an activity level of 1.3 x 107 Ci/year.

     A number of methods are being developed for collection of kryp-

ton from off-gas streams:

     •  cryogenic distillation;

     •  cryogenic selective adsorption;

     •  fluorocarbon adsorption.

     These collection methods are described in detail in other

sections of this report.  As a result of these collection procedures,

krypton gas would be available in almost pure form.  One technology

that already exists is the containment of krypton gas in pressurized

cylinders.  Several thousand of these cylinders have been tested in

normal usage over many years and extrapolation of their results

indicates a usefiil life of 500 years; however, these tests have been

conducted at room temperatures.  Radioactive decay of krypton would

liberate an appreciable amount of heat.  The corrosion properties of

the daughter product rubidium must also be determined.  Experiments

indicate that the alloy steels used for these high pressure cylinders

must be held well below 370°C because above this temperature the

yield and tensile strength of the cylinders decrease rapidly.  Geologic and Seabed Disposal.  Because of their rela-

tively short half-lives, both geological and seabed disposal  provide

adequate protection against migration for periods long enough to

allow krypton-85 to decay to relatively innocuous levels.  Krypton-85


should be stored separately from high-level radioactive, waste to

avoid high temperatures which affect containment*  Disposal of Kr-85

in a geological repository could present problems if gas leakage

occured during the operating phase.  Transmutation.  Transmutation of Kr-85 is not feasible

because its cross-section for both thermal and fast flux neutrons is

too small to achieve an appreciable gain over its natural decay rate.  Extraterrestrial Disposal.  Extraterrestrial disposal

of Kr-85 is technically feasible; however the costs of space dis-

posal are high compared to geological and seabed disposal options.  Engineered Storage Facilities.  The storage time for

permanent disposal of krypton-85, as well as tritium, is relatively

low (on the order of 100-200 years, i.e., 10 lifetimes).  Both steel

cylinders and encapsulated zeolites are the two most promising tech-

nologies for Kr-85.  The principal danger to be avoided is high tem-

perature from the accumulated decay.  Accordingly, the engineered

storage facility, as illustrated in Figure 11, /+8  is of interest

for the Kr-85.  It consists of two parts, a remote handling transfer

cave and a sealed storage area.  Pressurized cylinders or

encapsulated zeolite units are transported to the cask in a shielded

transfer cave and then moved to a storage slot in the sealed storage

area.  All this is easily accomplished with remote handling equipment

and viewing windows.  The other methodxof promise, deposition in

metal by ion implantation, requires very simple storage.

                               LEAK TESTING
                                                        SEALED STORAGE
               STRADDLE CARRIER
SOURCE:  ChrLstensen,  1CP-1128, Reference 48.
                                       FIGURE 11

     The leak tight storage vault prevents any sudden accidental

release of Kr-85, if a pressurized cylinder develops a leak.  In the

event of failure of one or more pressurized cylinders, the released

krypton must be confined to the storage facility until it is recycled

through an adjoining krypton recovery unit.  The encapsulated zeolite

is not vulnerable to such release and could be stored in a simpler


     The engineered storage option is also attractive in this case

because with the decay of Kr-85 it may be useful to retrieve older

capsules or cylinders for testing and possible reuse.

     Tritium could be stored in a similar type of facility.  Large

quantities of tritium are required if thermonuclear (fusion) reactors

reach the operational stage of development and the tritium waste

could, at that time, be gainfully used.

(1)  Statement by President Carter on Nuclear Power  Policy,  the  White
     House, April 7, 1977.

(2)  "Technical Support for Radiation Standard for High  Level  Radio-
     active Waste Management."  Task A Draft Report  to Office  of
     Radiation Programs, U.S.  Environmental Protection Agency.
     Arthur D. Little, Inc., Cambridge, MA, July 1977.

(3)  "Technical Support for Radiation Standards  for  High Level Radio-
     active Waste Management."  Task B Draft Report  to Office  of
     Radiation Programs, U.S.  Environmental Protection Agency.
     Arthur D. Little, Inc., Cambridge, MA, July 1977.

(4)  "Alternatives for Managing Waste from Reactors  and  Post-Fission
     Operations in the LWR Fuel Cycle."  U.S. Energy Research  and
     Development Administration, ERDS-76-43, Section 3.2.  May 1976.

(5)  D.T. Pence and B.A. Staples.  "Solid Adsorbents for Collection
     and Storage of 1-129 from Reprocessing Plants." Proceedings  of
     the Thirteenth AC Air Cleaning Conference,  San  Francisco, CA,
     CONF-740807, U.S.A.E.G.,  August, 1974.

(6)  T.B. Rhinehamer and P.H.  Lamberger.  Tritium Control Technology,
     WASH-1296 LIC-70 Monsanto Research Corporation, Mound
     Laboratory, Miamisburg, OH, December 1973.

(7)  J.H. Goode and V.C.A. Vaughen.  "Experiments on the Behavior  of
     Tritium During Head-end Processing of Irradiated Reactor  Fuels."
     ORNL-TM-2703, Oak Ridge National Laboratory, Oak Ridge, TN,
     February 1970, p. 5.

(8)  L.V- Grossman and J.O. Hegland.  "Tritium Distribution  in High
     Power Fuel Elements."  GEAP-12205, General  Electric Co.,
     Pleasanton, CA, June 1971, p. 11.

(9)  S.V- Ribnikar and J.D. Pupejin.  "Possibilities of  Tritium
     Removal from Waste Water  of Pressurized Water Reactors  and  Fuel
     Reprocessing Plants."  Proc. 13th Air Cleaning  Conference CONF
     740807.  San Francisco, CA, August 1974.

(10) B.C. Musgrave.  "Tritium Distribution in the Nuclear Industry."
     ICP-1041 Allied Chemical  Corp., Idaho Falls, ID, January, 1974.

(11) N.E. Hall and G.N. Ward.   "Tritium Control by Water Recycle in a
     Nuclear Fuel Reprocessing Plant."  NEDG-11342,  General  Electric
     Co., San Jose, CA, June 1975.


(12)   Savannah River laboratory Quarterly Report,  "Light Water
      Reactor Fuel Cycle."   DPST-LWR-76-1-2,  E.I.  DuPont de Nemours
      and Co., Aiken,  SC, April-June 1976.

(13)   D.A. Knecht.  "An Evaluation of Method  for Immobilizing
      Krypton-85."  Idaho Noatio'nal Engineering Laboratory ICP-1125,
      Idaho Falls, ID,,1977.

(14)   C.M. Slansky.   "Separation Processes  for Noble Gas Fission
      Products from the Off-gas of Fuel  Reprocessing Plants."  Atomic
      Energy Review 9,  pp 423-440, 1971.

(15)   C.L. Bendixsen and F.O.  German. "1974  Operation of the ICPP
      Rare Gas Recovery Facility."  ICP-1057,  Idaho  National
      Engineering Laboratory,  Idaho Falls,  1C,  1975.

(16)   C.J. Barton.  "Separation and Containment of Noble Gases."
      Nuclear Safety 15, pp  302-305,  1974.

(17)   T.  Kanazawa, et  al. "Development of the Cryogenic Selective
      Adsorption - Desorption  Process or Removal of  Radioactive Noble
      Gases."  Proc. 14th ERDA Air Cleaning Conference.   Sun Valley,
      ID, 2-4 August 1976, published February 1977.

(18)   G.  Griffith.  "99% Cleanup of Nuclear Gaseous  Wastes."  Power
      Engineering, pp 62-64, March 1973.

(19)   R.M. Hogg.  "New Radwaste Retention System."  Nuclear
      Engineering International, pp.  98-99, February 1972.

(20)   M.J. Stephenson and R.S. Eby.  "Development  of the Faster
      Process for Removing Krypton-85, Carbon-14 and Other
      Contaminants from the  Off-Gas of Fuel Reprocessing Plants."
      Proc. 14th ERDA Air Cleaning Conference,  Sun Valley, ID, 2-4
      August 1976, published February 1977.

(21)   R.W. Glass, H.W.R. Beaufeau, V.L.  Fowler, T.M. Gilliam, O.J.
      Inman and D.M. Levins.   "Experimental Studies  on the Krypton,
      Adsorption in Liquid C02 (FALC) Process." Ibid.

(22)   H.  Bonka, K. Brussermann, G. Schwarz.  "Unweltbelastung durch
      Radiskohlenstoff aus kerntechnischen Anlagen."  Reaktortagung,
      Berlin, April 1974.

(23)  G.N. Kelly, J.A. Jones, P.M. Bryant and F. Morley.  "The
      Predicted Radiation Exposure of the Population of the European
      Community Resulting from Discharges of Krypton-85, Tritium,
      Carbon-14 and Iodine-129 from the Nuclear Power Industry to, the
      Year 2000."  Commission of the European Communities,
      Luxembourg, Doc V/2627/75, September 1975.

(24)  R.L. Blanchard.  "Radiological Surveillance Studies at the
      Oyster Creek BWR Nuclear Generating Station."  U.S.
      Environmental Protection Agency, Office of Radiation Program,
      EERF, RNEB, Cincinnati, OH, June 1976.

(25)  A.G. Croff.  "An Evaluation of Options Relative to Fixation and
      Disposal of 14C-Contaminated C02 as CaO^"  ORNL/TM-5171,
      Oak Ridge National Laboratory, Oak Ridge, TN, April 1976.

(26)  J.G. Moore.  Oak Ridge National Laboratory, Private
      Communcation, 1977.

(27)  J. Mandel.  Battelle Northwest Laboratories, Private
      Communication, 1977.

(28)  L. Hydes.  Savannah River Laboratory, Private Communication,

(29)  W.E. Clark.  "Immobilization of 1-129 as Barium lodate with
      Portland Cement."  Transactions of the American Nuclear society
      23, pp. 264-265, June 1976^

(30)  W.E. Clark.  "The Isolation of Radioiodine with Portland Cement
      Part I - Scoping Leaching Studies."  Nuclear Technology (to be

(31)  H. Hatta and H. Ono.  "Experimental Study of Leaching of
      Radioactive Materials from Radwaste Solidified in Cement by
      Seawater, Part II."  Central Research Institute of the Electric
      Power Industry: Translation from the Japanese, ORNL-TR-4226.
(32)  B.A. Staples, L.P. Murphy and T.R. Thomas.  "Airborne Elemental
      Iodine Loading Capacities of Metal Zeolites and a Dry Method
      for Recycling Silver Zeolite."  Proc. 14th ERDA Air Cleaning
      Conference, Sun Valley, ID, 2-4 August 1976, published February
(33)  C.M. Slansky, Ed.  "Alternative Fuel Cycle Technology Progress
      Report."  ICP-1131, Idaho National Engineering Laboratory,
      Idaho Falls, ID, October, 1977.

(34)   T.R.  Thomas,  B.A.  Staples,  L.P.  Murphy, J.T.  Nichols.
      "Airborne Elemental Iodine  Loading Capacities of Metal Zeolites
      and a Method for Recycling  Silver Zeolite, ICP-1119, Jiaho
      National Engineering Laboratory, Odaho Falls, ID, July 1977.

(35)   T.R.  Thomas.   Private Communication,  1977.

(36)   P. Colombo and M.  Steinberg.   "Tritium Storage Development,
      Progress Report No* 4," Brookhaven National Laboratory
      BHL-20421, Upton,  NY, July  1975.

(37)   P. Colombo and M.  Steinberg.   "Tritium Storage Development
      Progress Report No. 9."  Brookhaven National Laboratory,
      BNL-50625, Upton,  New York, September 1977.

(38)   H.F.  Anderson and C.J. Kershner.  "Tritium Waste Control
      Project - October 1976-March 1977."  Monsanto Research
      Corporation MLM-2451, Miamisburg, OH, October 1977.

(39)   J.A.  Franz and L.L Burger.   "Polymeric Media for Tritium
      Fixation," BNLB-430/UC-70,  Battelle Pacific Northwest
      Laboratories, Richland, WA, May 1975.

(40)   P. Colombo.  "Tritium Storage Development Progress Report #8."
      Brookhaven National Laboratory,  BNL-50583, Upton, NY, June

(41)   J.A.   Franz and L.L. Burger.  "Polymeric Media for Tritium
      Fixation."  BNWL-B-430, SUPI, Battelle Pacific Northwest
      Laboratories, Richland, WA, February 1976.

(42)   P. Colombo, M. Steinberg, and B. Manowitz.  "Tritium Storage
      Development Progress Report 1."  Brookhaven National
      Laboratory, BNL-19408, Utpon, NY, October 1974.

(43)   W.D.  Arnold.  "Preliminary  Evaluation Method for the Disposal
      of Tritiated Water from Nuclearly Stimulated Natural Gas
      Mills."  ORNL-TM-4024, Oak  Ridge National Laboratory, Oak
      Ridge, TN, April 1973.

(44)   L.L.  Burger and J.L. Ryan.   "The Technology of Tritium Fixation
      and Storage."  Battelle Northwest Laboratories, BNWL-1807,
      Richland, WA, January 1974.

(45)   R.D.  Scheele and L.L. Burger.  "Zirconium Hydride as a Storage
      Medium for Tritium."  Battelle Northwest Laboratories,
      BNWL-2083, Richland, WA, July 1976.

(46)  B.A. Foster and D.T. Pence.  "An Evaluation of High Pressure
      Steel Cylinders for Fission Product Noble Gas Storage."  Idaho
      National Engineering Laboratory, Idaho Falls, ID, February

(47)  R. Neary, Linde.  Molecular Sieves Division, Morristovra, NJ,
      Private communication, December 1977.

(48)  A.B. Christensen.  "Physical Properties and Heat Transfer
      Characteristics of Materials for Krypton-85 Storage."  Idaho
      Chemical Program Operations Office, ICP-1128, September 1977.

(49)  R.A. Brown, H. Hoza, and D.A. Knecht.  "85Kr Storage by Zeolite
      Encapsulation."  Proc. 14th ERDA Air Cleaning Conference, Sun
      Valley: ID, 2-4, 1976.  Published February 1977.

(50)  D. Check, R. Meahl, 0. Cucciata and E. Camevale.  "Radioactive
      Kryptonates, I. Preparation."  Int. J . Rad Isotopes 14 pp.
      581-591 (1963).

(51)  G. Carter and J.S. Colligan.  Ion Bombardment of Solids,
      American Elsevier, pp. 547-599, New York, 1968.

(52)  J.R. Berreth, H.S. Cole, E.G. Samsel and L.C. Lewis.  "Status
      Report:  Development and Evaluation of Alternative Treatment
      Methods for Commercial and ICPP High-Level Solidified Wastes,
      ICP-1089, Idaho National Engineering Laboratory, Idaho Falls,
      ID, May 1976.

(53)  R.W. Benedict.  "An Evaluation of Methods for Immobilizing
      Solids Loaded with Krypton-85,"  ICP-1130, Idaho National
      Engineering Laboratory, August 1977.

(54)  "Alternative Disposal Concepts for High-Level and Transuranic
      Radioactive Waste,"  Draft Report, MTR 7718, MITRE Corp., March

(55)  Assessment of the Radiological Protection Aspects of Disposal
      of High-Level Waste on the Ocean Floor,"  P.O. Grimwood and
      G.A.M. Webb, National Radiological Protection Board, NRPG-R-48,
      October 1976.

(56)  Technical Support for Radiation Standards for High-Level
      Radioactive Waste Management," Task C Draft Report ice of
      Radiation Programs, U.S. Environmental Protection Agency,
      Arthur D. Little, Inc., Cambridge, MA, January 1978.

(57)  A. Hamid and B.P. Warkentin.  "Retention of 1-131 Usea as
      Tracer in Water - Movement Studies" Soil Science 104 4, p.  279,
                                              US GOYtRMItPlI WmilNC OfJTCE 1979-261-147 .'12*