United States
Environmental Protection
Agency
Office of
Radiation Programs
Washington DC 20460
ORP/CSD 79-2
Radiation
xvEPA
ASSESSMENT OF
WASTE MANAGEMENT
OF VOLATILE
RADIONUCLIDES
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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.
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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
MTR-7719
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FOREWORD
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
quality.
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
Director
Criteria Standards Division
Office of Radiation Programs (ANR-460)
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ABSTRACT
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.
111
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TABLE OF CONTENTS
LIST OF FIGURES
LIST OF TABLES
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
3.0 PROJECTED QUANTITIES AND COLLECTED WASTE FORMS 15
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
iv
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TABLE OF CONTENTS (Concluded)
5.0 DISPOSAL OPTIONS FOR VOLATILE RADIONUCLIDES "
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
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LIST OF FIGURES
Figure
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
Sputtering
11 Pressurized Cylinder Storage Facility for Kr-85 109
vi
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LIST OF TABLES
Table
Number
II
III
IV
V
VI
VII
VIII
IX
XI
XII
XIII
XIV
XV
XVI
XVII
Page
Preparation of Volatile Radionuclides for 10
Storage/Disposal
Summary of Volatile Radionuclide Disposal 12 12
Methods
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
Plants
Tritiated Water Produced by 5 MT/Day Reprocessing 23
Plant
Carbon-14 Production in Light Water Reactors 32
Solubilities of Selected Iodine Compounds in 42
Water
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
vii
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L.O INTRODUCTION
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.
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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
disposal.
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.
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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.
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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
products.
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.
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2.0 SUMMARY
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.
5
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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
techniques;
(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
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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
sites.
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
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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
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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
following:
• 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.
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NUCLIDES
TABLE I
PREPARATION OF VOLATILE RADIONUCLIDES FOR STORAGE/DISPOSAL
IODINE-129 TRITIUM KRYI'TON-85
CAKBON-14
QUANTITIES
PRODUCED
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 <*>
COLLECTION
METHOD(S)
Mercurex
lodox
Cheulsorptlon
Voloxldatlon
Pyrochemical processes
Process-Steam treatment
Cryogenic Distillation
Caustic scrubbing.
COLLECTED
FORM
lodate of llg, Ba
Trltlated water
Elemental
CaCO
STATE OF
DEVF.LOPHENT
Experimental
Early etagea
Pilot plant
Well established for
stable carbon.
COST
Not known
$0.15 j.?r kg of CO
fixed.
IHHOBILIZKD
FORM
lodate in concrete
Zeolite
Chemical storage In polymer Im-
pregnated concrete , polyethylene
Organic compounds
Hydrides
Zeolite adsorption
Ion Implantation (sputtering)
CaCO In concrete.
LEACH RATE
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.
STATE OF
DEVELOPMENT
Experiments still In progress.
Impregnated concrete - Advanced
experimental. Organic' compounds,
hydrides - Experimental.
Zeolite -Advanced experimental.
Ion Implantation - experimental
stage.
Well known.
COST
$356.000 In Ag for 600 kg I.
$3.10-$16.90 per gallon of
trltlated water.
CANISTER
CONTAINMENT TYPE
Pressure vessel containment.
(for krypton gas)
55 gallon drums.
CONTAINMENT
LIFETIME
'500 years
STATE OF
DEVELOPMENT
Well established
40c a gull nn
(1)
J./1500 NT/year reprocessing plant
; ..{mixed with a comparable amount of 1-127
(4)
In the forn nf IITO
'C02
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• 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-
85.
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.
11
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TABLE II
SUMMARY OF VOLATILE RADIONUCLIDE DISPOSAL METHODS
Iodine-129
Carbon-14
Krypton-85
Tritium
Geologic
Disposal
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
concept.
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.
Seabed
Disposal
Not likely to contain 1-129
until sufficiently decayed.
Ocean dilution may reduce
concentrations to acceptable
levels.
Appears to be a
satisfactory disposal
concept.
Appears to be a
satisfactory disposal
concept.
Appears to be a
satisfactory disposal
concept.
Extraterrestrial
Disposal
Satisfactory for elimination
of 1-129.
Accident risk and consequences
require careful study.
Probably not warranted
because of cost and
availability of other
concepts.
Probably not warranted
because of cost and
availability of other
disposal methods.
Probably not warranted
because of cost and
availability of other
disposal methods.
Transmutation
Not a likely removal mechanism
unless fusion reactors are
developed.
Not a likely approach.
Not a likely approach.
Tritium could be
used as fuel for
fusion reactors if
developed.
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Iodine-129
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.
Carbon-14
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-
factory.
Krypton-85
The disposal of Kr-85 in geological and seabed repositories
appears to be a satisfactory option if it is separated from other
13
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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.
Tritium
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.
14
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3.0 PROJECTED QUANTITIES AND COLLECTED WASTE FORMS
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
15
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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
Ci/yr.5
16
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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.
17
-------
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
used.
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.
18
-------
TABLE III
ANNUAL TRITIUM PRODUCTION IN A TYPICAL 1 GWe LIGHT WATER REACTOR
SOURCE ELEMENT
Deuterium
Uranium, plutonium
Lithium-7 resulting from
neutron capture by
Boron-10
Same aa above
NUCLEAR
REACTION
Neutron capture
Ternary fission
(n, na)
Same as above
TRITIUM
Ci/GWe yr
12
25,000
10,500
1,100
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.
20
-------
TABLE IV
ESTIMATES OF TRITIUM DISTRIBUTION IN DIFFERENT PATHWAYS
AT FUEL REPROCESSING PLANTS
PATHWAY
PERCENT OF TOTAL
FISSION YIELD
FORM
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
9.3a
14b, 20*.
, 10e, 20f, 458
, 6h, 20h
Remainder
Elemental
Elemental
Elemental, HTO
In combination
with organic
materials
HTO
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
g)
h)
Mus grave
10
Hall and Ward
11
21
-------
TABLE V
ERDA'S ESTIMATE OF TRITIUM IN FUEL REPROCESSING PLANTS
Spent fuel input
Gaseous waste (as HT)
Cladding hulls
Dissolver solution (as HTO)
a) High-level waste
PERCENT
100
5
15
SO
8
Ci3 H/MTHM
547
28
82
437
43
Ci/yr «C
1500 MTHM/yr
820,500
42,000
123,000
655,000
64,500
concentrate
b) Low-level liquid 72 394 591,000
waste
Comprehensively:
As HT
As HTO
5
93
28
519
42,000
778,500
Source: ERDA 76-43, Vol. 1, p. 2.64., Reference 4.
22
-------
TABLE VI
TRITIATED WATER PRODUCED BY A 5 MT/DAY REPROCESSING PLANT
SOURCE
Liters/day
Tritium Ci/liter
Total Tritium Ci/day
Total Tritium Ci/yr4
Total Tritiated Volume
Liters/year
m-Vyear
CONCENTRATED
TRITIUM WASTE
50 kg3
60
3000
900,000
15,000
15
DILUTED
TRITIUM WASTE
105 (105 kg)
3 x 10~2
3000
900,000
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.
23
-------
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.
Voloxidation
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.
24
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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
25
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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
26
-------
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
hydrogen.^
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
28
-------
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
29
-------
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
30
-------
of a beta-ray with a maximum energy of 156 keV. Carbon, being a
constituent of all organic materials, is easily absorbed into the
biocycle.
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
31
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Fuel
U)
Coolant
TOTAL
TABLE VII
CARBON-14 PRODUCTION IN LIGHT WATER REACTORS
(Ci/GWe-yr.)
BWR PWR
17
0
14
N
1 7
'o
14
N
Bonka et al. Kelley et a
8.4 2.7
12.9 10.9
21.3 13.6
9.9
1.3
11.2 16.0
32.5 29.6
1. Bonka et al. Kell
7.1
12.2
19.3
9.8
1.3
U.I
30.4
2.7
10.9
13.6
6.0
19.6
-------
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
material.
33
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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.
34
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COz
FIXATION
TOWER
BURNER
OFF-GAS
FROM
KALC
TO
ATMOSPHERE
LIME
SLAKING
TANK
PEBBLE
LIME
Co (OH)2
t_n
CoC03
SLURRY
SLURRY
FILTRATE
LIME
STORAGE
INDICATES THAT THE PATH
SELECTED DEPENDS ON THE
PACKAGING / TRANSPORTATION/
DISPOSAL METHODS USED
MAKE-UP
HoO
MOIST
FILTER CAKE
PACKAGING
CONTINUOUS SCREW
VACUUM CONVEYOR
FILTER
Source: Croff, Reference 25.
DR
YE
R
TRANSPORTATION
TO
DISPOSAL
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
36
-------
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.
37
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4.0 IMMOBILIZATION TECHNOLOGY
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
39
-------
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;
40
-------
• 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
41
-------
TABLE VIII
SOLUBILITIES OF SELECTED IODINE COMPOUNDS IN WATER*
Solubilities3 (10~3 kg-mole/ at 298°K)b
Cation
Iodide
lodate
Sodium
Potassium
Magnesium
Calcium
Barium
Lead
Silver
Mercuric
Mercurous
8.2
4.7
5 molalc
4.9 at 20°C
4.0
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
0.47
0.42
0.25
7.9 x 10~3
8.1 x 10~4
(3.6 to 5.5)
1.8 x 10~4
"Insoluble"d
(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+).
2
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.
42
-------
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.
43
-------
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.
44
-------
-2
CN
I
W
w
o
I— (
H
Q
w
a
o
8 x 10
7 x 10~2L
6 x 10
-2
5 x 10
-2
4 x 10
~2
3 x 10
2 x 10
"2
io
o
-2
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.
F1GURE2
THE EFFECT OF IODINE CONTENT ON THE LEACH RATE
OF BARIUM IODATE CONCRETE, NO ADDITIVES
-------
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 \
\o.i3j2
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.
46
-------
-O
-.J
tfl
T3
"e
o
rH
X
U
ss
,-t
3
H
t5
g
e
SPECIMEN D-l
— SPECIMEN D-2
10
JO
0 20 40 60 80
AVERAGE LEACHING PERIOD
2
Source: Clark, 1977, IU:1 tirenre 29.
(L is the time period between changes of leachant)
200 300 400
(day)
FIGURES
INCREMENTAL LEACH RATE OF BARIUM IODATE
FROM CONCRETE CONTAINING 9.5 WT% IODINE
-------
6.0 8.0 10.0 12.0
0.00
0.0
2.0
SQ. RT. ACCUM. TIME (DAYS)
Source: Moore, 1977, Reference 26.
FIGURE4
LEACHABILITY OF IODINE INTO CO, FREE DISTILLED WATER FROM
TYPE1 PORTLAND CEMENT CONTAINING 9.55 WT% IODINE AS BARIUM IODATE
(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.
49
-------
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
50
-------
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
gram.
The annual disposal of 600kg of iodine per model reprocessing
plant would require about 1900 kg or 2.2 cubic meters of zeolite per
year.
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
51
-------
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,
52
-------
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:
-5
^IL pL) / 1) = 8.61 . 10" cm/day
n " '"' "-n
53
-------
TABLE IX
CHARACTERISTICS OF CHEMICAL STORAGE TECHNOLOGIES FOR TRITIATEO WATER
1. Hydrates
•) Drying agent*
Celcium eulfate
SI lie* gal
Activated alula*
Molecular sieves
b) Uydraullc Cement*
Portland cement
!. Organic Compounds
Polyacetylena
•akallte analog polymers
Polyscryloaltrlle
•
FolyNtyrena
1. Hydrides
Zirconium. ZrHj ^
'
Titan !*•
tlranitio^UHj
Water Loading wt Z
6.2
40
20
10-20
21
Hydrogen (fern 0.6Bg of
water can b*j Incorpor-
ated to Ig of poly-
acetylene
Hydrogen from O.lSg of
water can be fixed In
Ig of polyacrylonltrlli
O.Sg of water per g of
polymer
l.S wtl II (O.Ug of
trltiated water)
Tritium Release Bate
10~3-10~*/day
-1 -4
10 -10 /day when en-
capaulated In poly-
etyreoe
•t
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
further
21 lose la rinsing.
none further
51 la Initial rinsing
tot SMaaurable
5 « 10~*/year la NaOU
4 x 10-S/yesr la diet.
I120
2 x 10-*/yeer la HC1
leurka
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
solvents
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
1970
Stable below 30O»C
Avallahlllty of
Technology
Available for removing
UTO vapor
Cncepvulatlon In con-
crete and poly more also
available
Polymer Isfiregnatioa
requires developnent
Available, coatings
available; polymer Im-
pregnation requlrea
development
Available
Acetaldehyde production
available, polymeriza-
tion also requires
Improvement
Available. Furlflca-
catlon neceasary.
complex, expansive
Available! problems la
tritium-control
Available; development
of facility able to fix
JO Kg UTO/ Jay require*
^ y«
-------
TABLE X
PROPERTIES OF ENGINEERED STORAGE OPTIONS FOR TRITIATED WATER
High Preasure Steel Cylinders
Type IH (for high level concentrated
waste
Above Ground Storage Tanks
(for low level diluted waste)
VOLUME
LITERS
IB
1.2 x 107
LIFE TIME
1EARS
>AO
REMARKS
uated.
Carbon steel construc-
tion, protection froa
Ice formation by In-
sulation and tank
heaters; can be an
AVAILABILITY
control required.
Available, corrosion
control by pH adjust-
ment
POSSIBLE
ACCIDENTS
Corrosion, 'material
fatigue
Radiolysla producing
hydrogen, oxygen and
recombination
Leakage
Sudden rupture
PRECAUTIONS
Temperature control,
•onitorlng, chem-
ical pll adjustment
Monitoring, cata-
lyst to promote
recombination
reaction
Pumping to reserve
tank
Double contain-
ment, transfer lines
Secondary contain -
ment
Ln
Source I ERDA 76-43, Vo. 2
-------
TABLE XI
SUMMARY OF CONTRACTOR STORAGE PRACTICES
Form
Container Disposal Method
Solid
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
<10Ci/ft3
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
drum
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
Gaseous
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
beds
Elemental tritum: SS tanks at pressurs <2 atm
Source: Rinehammer, 1973, pp. 329-336
-------
u
on
3
10
-1
o
M
a
2
fa
w 10
-2
I 10
-3
10
-4
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
10'
Source: Colombo (1976), Reference 36.
FIGURES
PROJECTED TRITIUM RELEASE VERSUS TIME FOR STATIC LEACHING
OF THE SRL LYSIMETER TESTING DUPLICATE SPECIMEN
(WITHOUT CONTAINER) IN DISTILLED WATER
58
-------
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
59
-------
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.
60
-------
TABLE XII
LEACHING DATA OF POLYMERS
Polymer
Bakelite
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
Hydrogen-
ated Poly- -1-1
styrene 690 d min" g"
Polyurey-
lene/poly-
methane —
Polystyrene
Tritiated
on Rh/Al203 -
1430 d min
-1
5130 d min
-1
794 d min
-1
Z 3 H Leached
Following
Initial Rinse
(a)
(b)
(c)
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),
39,41
61
-------
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
62
-------
TABLE XIII
MATERIAL COSTS FOR POLYMERIC MEDIA
AND ALTERNATE FIXATION OR STORAGE METHODS
Medium
f
Resorcinol- formaldehyde
acetaldehyde polymer
Phenol- formaldehyde-
acetaldehyde polymer
Polyacrylonitrile
Polystyrene
Polyacetylene
Polyureylene/polymethane
Polymer impregnated
Polymer Weight/kg
Water Disposed
5.9 kg
5.6
2.9
2.2.0
1.5
5.2-5.5
Cost/
kg Water
$ 4.50
2.20
1.60
2.20
0.80
12.00
5.27
Cost/
Gallon Water
$16.90
8.25
5.90
8.25
3.10
43.00
19.00
concrete
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.
63
-------
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
64
-------
TABLE XIV
TRITIUM ACTIVITY IN LEACH SOLUTIONS ZrHx (T)
Control
Solutions
(No T)
Distilled Water
Sat. KC1
Sat. NaCl
Test
Solutions
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
35.6
35.5
35.5
31.8 31.8
31.8 31.6
30.4 30.9
30.9 31.3
35.8
6/17/75
37.7
—
—
53.1
34.7
40.0
103.3**
—
*Total Tritium inventory equivalent to 1.3 x 10^ CPM/ml
**Sample pulverized after October 1, 1974.
Source: Colombo (1975), reference 36
65
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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,
66
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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.
67
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2000 \ ->
I
w
EH
1500 _
1000 _
Ed
500 _
20 30
TIME (Years)
-3.0
-2.5
-2,0
a
o
<
Source: Christensen, Reference 48.
FIGURES
KRYPTON-85 HEAT GENERATION AND DECAY RATES
AS A FUNCTION OFTIME
68
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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.
70
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Table XVI
RUBIDIUM PRODUCTION DURING STORAGE OF KR-85
IN HIGH PRESSURE STEEL CYLINDERS
500 psi cylinder 2000 psi cylinder
Year
0
10
25
50
100
150
Kr-85 (Ci)
128,000
67,200
25,600
5,090
294
8
Rb (g)
0
155
262
314
327
327
T (8C)
60
45
32
24
21
21
Kr-85 (Ci)
419,000
220,000
83,800
16,700
6,660
27
Rb (g)
0
509
857
1030
1070
1070
T (°C)
127
84
50
21
25
21
71
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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
o
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
Q
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
72
-------
FIGURE?
REPRESENTATION QF SODAL1TE CAGES
CONTAINING KRYPTON ATOMS
73
-------
ACTIVATED
ZEOLITE
Kr
HIGH
PRESSURE
7ESSEL
HEAT
•*•
STORAGE
CONTAINER
Kr
ENCAPSULATED
IN ZEOLITE
STORAGE
CYLINDER
Source: Knecht, Reference 13.
FIGURES
PROCESS FOR HIGH PRESSURE ENCAPSULATION
OF Kr IN ZEOLITE
74
-------
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
75
-------
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
equation
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
76
-------
01
>
c
p
o
CO
PI
•H
C*
01
60
co
CO
q
o
•H
4-1
a
100 150 200
Time (Years)
250
300
Source: Knecht, ICP-1125, kelerence 13.
FIGURE9
CALCULATED RELEASE OF ORIGINAL KRYPTON INVENTORY
FROM SODALITE AT 150°C AS A FUNCTION OF TIME
-------
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
storage.
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
loading.
*Pascal (Pa) = Newton/meter2
**MPa = 106 Pa
78
-------
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.
79
-------
A. IMPLANTATION PHASE
kV
o
„/ j.
RECEPTOR
O 0 O O 0 O 0
0 0 O
00000© 0
TARGET
IONIZER
I I
85T
© 85Kr IONS
o IMPLANTED Kr
B. SPUTTERING PHASE
/777/v7/77
RECEPTOR
oooooooooooooo
v\\\\\\\\\\\\\\\\\\\\\\\\\\\\\\\\\V
r
IONIZER
I
TARGET
.. -9 ]rV
85
© 85Kr IONS
• SPUTTERED METAL ATOMS
o IMPLANTED Kr
Source: Check, et.al., Reference 50.
FIGURE10
PROCESS FOR IMMOBILIZING 85Kr BY
ION IMPLANTATION/SPUTTERING
80
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TABLE XVII
SORPTION OF KRYPTON GAS BY METAL
Metal
Process
Wt.-Fract.
Loading
Required Wt.
of Metal*
00
Cu
Al
Fe
Ni
Au
Ag
Ni
Gd.llCo.73Mo.16
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
800°C
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
82
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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
83
-------
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
4.1.
84
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5.0 DISPOSAL OPTIONS FOR VOLATILE RADIONUCLIDES
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
85
-------
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
86
-------
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-
istics.
The effectiveness of geological repositories to isolate volatile
radionuclides depends on two factors:
(1) the form of the waste material and its resistance to
transport;
(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
87
-------
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
water
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
88
-------
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
half-lives.
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
89
-------
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
90
-------
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.
91
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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.
92
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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
sheets.
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,
93
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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.
5.2.1.1 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
94
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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
strata.
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.
5.2.1.2 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
95
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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.
96
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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
where
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
/chain.
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
98
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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.
5.2.1.3 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
release.
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
99
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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*
5.2.1.4 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
feasible.
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.
100
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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
kg/year.
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
101
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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.
5.2.2.1 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.
5.2.2.2 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
disposal.
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
102
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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
disposal.
103
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5.2.2.3 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
104
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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.
5.2.3.2 Transmutation. The transmutation of tritium is not
feasible because the cross-sections for both thermal and fast flux
105
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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.
5.2.3.3 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.
5.2.3.4 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 5.2.4.4) would
probably be used. Capability for recovery of tritium and
repackaging, and decontamination in the event of leaks would be
required.
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
106
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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.
5.2.4.1 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
107
-------
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.
5.2.4.2 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.
5.2.4.3 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.
5.2.4.4. 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.
108
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LEAK TESTING
TRANSFER
PORTS
SEALED STORAGE
AREA
STRADDLE CARRIER
SOURCE: ChrLstensen, 1CP-1128, Reference 48.
FIGURE 11
PRESSURIZED CYLINDER STORAGE FACILITY FOR 85Kr
-------
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
facility.
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.
110
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