TECHNICAL NOTE
ORP/TAD-76-4
Available Methods
of Solidification
for Low-Level
Radioactive Wastes
in the United States
\
LL1
December 1976
U.S. ENVIRONMENTAL PROTECTION AGENCY
Office of Radiation Programs
Washington, D.C. 20460
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Technical Note
ORP/TAD-76-4
AVAILABLE METHODS OF SOLIDIFICATION FOR LOW-LEVEL
RADIOACTIVE WASTES IN THE UNITED STATES
by
William F. Holcomb
Stephen M, Goldberg
December 1976
Technology Assessment Division
Office of Radiation Programs
U.S. Environmental Protection Agency
401 M Street, S. W.
Washington, D.C. 20460
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EPA Review Notice
This report has been reviewed by the Office of Radiation Programs,
U.S. Environmental Protection Agency (EPA) and approved for publica-
tion. Approval does not signify that the contents necessarily reflect the
views and policies of the EPA, nor does mention of trade names or
commercial products constitute endorsement or recommendation for use.
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PREFACE
The mission of the Technology Assessment Division (TAD) is to
provide the primary assessments of the technologies currently
utilized, or being proposed for use in our society in activities which
have a potential radiation impact on man or his environment. TAD
has attempted to fulfill its technology assessment mission by addressing
the major functional responsibilities, as described. in the authorities
and responsibilities gi\en by the reorganization plan under which EPA
was created. One of the major responsibilities for EPA which directly
involves TAD is to render technical assistance to the individual States
within EPA's fields of expertise. This report was prepared in order to
review the currently available and proposed methods of solidification for
low-level radioactive wastes as requested by the National Conference on
Radiation Control Program Directors Task Force on Radioactive Waste
Management.
Readers of this report are encouraged to inform the Office of Radiation
Programs of any omissions or errors. Comments or requests for further
information are also invited.
David S. Smith
Director
Technology Assessment Division
Office of Radiation Programs
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Abstract
This paper reviews the numerous solidificati >n systems and
related matrix materials that are presently beinj offered, or
proposed, to the commercial nuclear power indu ?try. Included,
where possible, is the nature of how these materials and/or systems
are affected by the physical, chemical, and radiolytic character-
istics of the treated radioactive waste materials. Key features of
the equipment used in individual solidification processes are
discussed in order to clarify the relative utility of these processes
for either power plant or fuel cycle application. Finally, a
discussion of current problems facing this phase of the nuclear
industry is presented.
iv
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Table of Contents
Page
Abstract iv
I. Introduction 1
II. General Considerations 5
III. Solidification of Radioactive Wastes in Bitumen 8
IV. Solidification of Radioactive Wastes in Cement 12
V. Absorbents 19
VI. Polymeric Solidification Processes 21
VII. Listing of Waste Solidification and Packaging 25
Systems Suppliers
VIII. Conclusions and Recommendations 31
IX. Cited References 34
List of Figures
Figure 1. Waste Processing Steps 7
Figure 2. Flow Diagram for Cement Incorpor ition Processes 16
Figure 3. Flow Diagram for UF Incorporation Processes 23
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List of Tables
Page
° —
Table I Solidification Systems /Agents 3
Table II Cementation Practices at Various Establishments 14
Table III Burial Site Waste Form Requirements 20
Table IV Comparisons of Leach Rates for Various Solidified 33
Waste Products
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AVAILABLE METHODS OF SOLIDIFICATION
FOR LOW-LEVEL RADIOACTIVE WASTES IN
THE UNITED STATES
I. INTRODUCTION
Radioactive wastes are customarily categorized as low- or high-
level (see page 5 for definitions) depending upon the concentrations of
radionuclides. However, the long-term hazard associated with each
waste is not necessarily proportional to the level of r idioactivity, but
rather to the specific toxicity and decay rate of each radionuclide.
Due to restrictions placed upon the disposal of wastes at the
commercial burial sites and the increasing amount of wastes being
generated both in volume and activity by the nuclear industry, consider-
able interest has been shown during the last several years concerning
various methods and systems for the solidfication. of liquid and solid
radioactive wastes from nuclear power plants and, to a varying degree,
from other fuel cycle facilities.
The existing commercial low-level radioactive waste disposal sites
in the the U.S. were licensed while implementing the concept of contain-
ment of radioactive wastes within site boundaries. The present U.S.
method for the disposal of low-level radioactive wastes is by burial
in shallow trenches dug in the earth's surface.
There are many varieties of solidification materials and techniques
available. Solidification agents include portland cement, concrete,
plaster of paris, asphalt (or bitumen), polymers, and a blend of absor-
bent material and cement or plaster. The method of solidification used
should not be a reversible process which can re urn to the liquid form
after placement of the solid in the disposal trenc h. One benefit of solid-
ification is to remove the liquid and reduce the potential for movement
through the soil of the radionuclides incorporated in the solidified waste.
A second benefit of solidification of radioactive liquids and sludges is
to produce an inert immobilized waste matrix which is safer to handle
during transportation and receiving operations at the burial site.
It is the intent of this report to review the various solidification
systems and matrix materials that are presently employed or proposed;
and to include where possible, how these materials and methods are
affected by the physical, chemical, and radiolytic characteristics of
the waste materials. Presently, most of the solidification systems
used in the United States utilize either cement or an organic polymer,
such as urea formaldehyde, as the basic solidification matrix material.
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Table I lists the various solidification agents and system vendors.
Some of the individual solidification equipment features are included,
along with some of the current problems anticipated in the operational
phase of the alternate systems. This report is primarily a 'state-of-
the-art" survey of the various solidification systems and technology.
It will not present a statement of systems preference.
The first commercially operated burial site began operation in
1962 and since that time, the industry has expanded to include three
private companies operating six sites. The sites are located at Maxey
Flats, Kentucky; Beatty, Nevada; Sheffield, Illinois; Barnwell, South
Carolina; West Valley, New York; and Richland, Washington. Earlier
reports have described these sites in some detail (i, 2,3).
The burial facilities are managed by private industry, but, by
regulation, are located upon Federal- or State-owned land (4). They
are, in general, regulated by the State in which they are located,
according to the provision of agreements between the individual States
and the U.S. Nuclear Regulatory Commission (NRC). The one
exception is the site located at Sheffield, Illinois, which is regulated
by the NRC.
The increasing emphasis upon the environmental considerations
relating to the nuclear industry has focused attention on radioactive
waste management operations. Experience in such operations has
raised questions concerning the environmental acceptability of
current practices and methods used for the disposal of low-level
waste.
EPA is responsible for providing assistance to State agencies
on environmental radiation-related matters. This responsibility,
contained in the Public Health Service Act, was given to EPA under
Reorganization Plan No. 3, provides EPA with authority to assist
States in their efforts to assess the safety of radioactive waste
management activities carried out under State license (5).
The Office of Radiation Programs (ORP) of EPA, at the request
of the National Conference on Radiation Control Program Directors'
Task Force on Radioactive Waste Management, reviewed the various
systems for solidifying wastes and prepared this summary report (6).
Estimation of the rate of leaching from a solidified matrix during
burial is one of the important considerations in the assessment of a
solidification method as it will strongly influence the amount of treat-
ment, containment, and surveillance that will be needed. Low matrix
solubility will improve the safety of waste management by reducing the
likelihood of an unplanned release.
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Table i
SOLIDIFICA'"ION SYJSTHI'.IS
TYPE
I. Cement System
Portland cement and sodium silicate
Cement /Vermlculite
Portland Cement Type II
Portland Cement Type I with ActiyJUe:
Cement with organic polymers
Cement/vermiculite
Cement / shale / clay
Cement
II. Urea Formaldehyde Systems
III. Bitumen (Asphalt) Systems
IV. Organic Polymer Systems
'.' LJM DOR
.''/o'. M'jeJaar Industries
' OK
•:.'••- Equipment Company
:k ':idge National J^ab a/
i;"khti^eu National Lab~a/
.'.-.lip-aii^use Electric Corp.
;;,,7-,;.s.j-e Custom Materials
r (\r.:i; Knergy Conversion Co.
:en.-Nuclear Systems
• ''ir>an Nuclear & Develop-
\enf' Corp
•HO? ai Electric Company
'er?1/ lac.
Urs'i.'-d Muclear Industries
Chen: - Nuclear Systems
Hittinau Nuclear & Develop-
ment Corp.
H-oi.eetive Packaging Inc.
ANEFCO, Inc.
Stock Equipme it Company
A.C; reject Energy Conversion Co.
Eiie'rv Trie.
-'eine.' t.nd Pf eiderer Corp.
^c-;j ojei: Dnerg; Conversion Co.
Division
and Technical
ical Company
^/ Non-commercial applications
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This becomes increasingly important when long-term disposal is
considered.
Present regulations involving the solidification of reactor-gen-
erated wastes are currently not the result of burial c onsiderations,
but of transportation considerations. The transport of radioactive
wastes is done in compliance with existing Department of Transpor-
tation (DOT) and Nuclear Regulatory Commission rules fo3 the safe
transport of radioactive materials. This will normally ensure that
radioactive waste transportation will not result in an unacceptable
radiation hazard to man and the environment.
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II. GENERAL CONSIDERATIONS
Wastes from he nuclear fuel cycle and particularly from power
plants can be solid, liquid or gaseous with varying chemical, physical,
and radiological characteristics. Since most solidification systems
process special liquids which produce wet solids such as evaporator
concentrates, resins, etc., it is necessary to also categorize or
classify those liquid wastes which produce the wet solids. The
following terms which are used in this report are defined.
1. Absorb - to immobilize by a method in which the liquid is totally
retained by physical means (e. i*., by use of such processes as
absorption or nicrocellular capture).
2. Immobilize - to treat the radioactive liquid wastes in such a
manner as to eliminate characteristics of fluidity, dispersibility,
or freedom of movement within the receptacle.
3. Low-Level Wastes - all those wastes from the fuel cycle other
than the transuranium-contaminated and high-level wastes.
They usually consist of contaminated paper, cloth, filters,
clothing, filter material, demineralizer resins, evaporator
concentrates, sludges, activated structural components, etc.
4. Hjgh-Level Wastes - (a) high-level liquid waste or (b) the
products from solidification of high-level liquid wastes, or
(c) irradiated fuel elements, if discarded without processing.
5. High-Level Liquid Waste - the aqueous waste resulting from
the operation of the first - cycle extraction system, or equivalent
concentrated wastes from subsequent extraction cycle, or equiv-
alent wastes from a process not using solvent extraction, in a
facility for processing irradiated reactor fuels.
6. Receptacle - the primary containment receptacle, into which the
radioactive liquid waste and the immobilizant are placed for
immobilization.
7. Solidified Radioactive Wastes - products resulting from the
immobilization or chemical fixation of liquid, semi-liquid or
solid radioactive wastes.
8. Transuranium-Contaminated Wastes - any wastes containing
significant amounts of long-lived alpha emitters such as
plutonium (this number has not been resolved, but is considered
by some to be greater then 10 nanocuries per gram).
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The radioactive liquid wastes from the power reactors, such as,
primary system blowdown, equipment drains, resin sluicing water,
evaporator condensates, decontamination solutions, demineralizer
regenerative solutions, laundry and laboratory wastes, evaporator
concentrates, spent ion exchange resins, filter precoat and caike
materials (powdex and solka-floc), cartridge filter units, and
diatomaceous earth are suitable for immobilization. These wastes
contain the bulk of the volume and radioactivity of the solidified
wastes sent to the commercial low-level burial facilities.
The processing of these wastes can be broken down into five steps
(7): (a) waste collection; (b) solids pretreatment; (c) the solidification
process; (d) mixing and pack;iging; and (e) final handling (Figure 1).
Collection usually takes place in sumps or tanks; the contents are then
processed on a batch or semi-continuous basis. The solids pretreatment
operation consists of reducing the volume of the wet solids by using an
evaporator or other volume reduction device. The solidification and
mixing steps involve the use of an agent, such as cement or an organic
polymer with additives or a catalyst, to produce an immobilized,
monolithic, inert matrix. The container handling operations include
inspection to ascertain that solidification took place, capping the container
and adding appropriate shielding, decontamination, marking and labeling,
container testing, and storage awaiting shipment to a burial facility.
There are several physical, chemical and radiological properties
which should be considered as of primary importance in assessing the
potential environmental impact from the solidified wastef on the shallow-
land burial facilities. Some of these properties include: (a) leachability;
(b) thermal conductivity; (c) chemical stability; (d) radiation resistance;
(e) mechanical ruggedness; ([') noncorrosiveness of shipping container;
(g) total solidified volume, (h) pH of solidified matrix, (i flammability;
(j) density; etc.
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HI. SOLIDIFICATION OF RADIOACTIVE WASTES
IN BITUMEN
The use of bitumen to solidify low-level radioactive wastes
has been successfully applied on an industrial scale for many years
in different countries (8-21). Bitumen or asphalt, a mixture of
high-molecular weight hydrocarbons, is a by-product residue from
the petroleum refining processes. Various grades of bitumen are
commercially available with a wide range of physical properties.
Bitumen processes generally operate in the range of 150 to 230 ° C,
at which temperature water originally present in the waste can be
potentially volatilized.
Basically^ all of the bitumen processes consist of mixing the
waste solution, slurry, or solids with commercial emulsified asphalt
or molten base-asphalt and raising the temperature to evaporate
the waste fluid. The solids remain intimately dispersed in the asphalt
and the product flows out of an evaporator into a receptacle. The
process has been successfully demonstrated in both continuous and
batch mixing operations (8,9,10). Four different processes have
been developed for the bitumen-waste incorporation process:
(a) stirred evaporation, (b) film evaporation, (c) the emulsified
bitumen process, and (<0 the screw extrusion.
Operational Experience
Several versions of the bitumen processes for incorporating
radioactive wastes for disposal have been utilized on an industrial
scale in Europe (8-15). The first plant scale bitumen process,
using the stirred evaporator method was started in 1964 with an
evaporation rate of 100 liters per hour at Mol, Belgium (9, 15).
The initial operation was directed at the bitumen incorporation of
radioactive chemical sludges. Subsequently, concentrated solutions,
incinerator ash, vermiculite, ion-exchange maierial and sand have
been processed. Difficulties were experienced with sand and ash
owing to the abrasive nature of the material, phase separation
occurring due to density differences between the ash and bitumen,
and foam formation with organic ion-exchange material. However,
these problems were solved using a slow-mixing device, particle
separation, and resin incineration (9, 15), Certain oxidizing salts,
particularly nitrates, produce an undesirable hardness in the final
product; however, this difficulty can be solved by use of reducing
agents or different bitumen mixes. Finally, boric acid solutions can
be incorporated into bitumen if the solutions are first neutralized to
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prevent acid volaiilization and water leaching of final product. The
waste treatment plant at Harwell in the United Kingdom was designed
on the same principles as that at Mol and was intended primarily
for the incorporation of chemical sludges, but is presently not in
routine use (9,11, 15).
The emulsion bituminization process for radioactive sludges
was pioneered at the Marcoule Centre, France (9, 15). This process
requires a much higher bitumen content where greater than 1 Ci/m3
material is processed. Various surf ace-active agents, (a surface-
active agent is a soluble compound that reduces the surface tension
of liquids, or reduces interfacial tension between two liquids or a
liquid and a solid), variable reaction times, and different bitumen
varieties are used depending on the sludge to be treated. Experi-
mental work on this process was also done at the Oak Ridge
National Laboratory (ORNL), Tennessee (9). In this process the
initial mixing of liquid waste and bituznen could be affected readily
at any convenient temperature below the boiling point of the waste
solution with the water and/or volatiles removed by heating. The
liquid wastes of special interest to ORNL were the evaporator
concentrates and solutions of sodium metaborate, nitrate and nitrite.
No significant difficulities were experienced in the incorporation
of 60 weight per cent (w/o) of solids in bitumen from evaporator
concentrates, however, the test final products were produced at
solids contents of 45-50 w/o. Boi on compounds required higher
temperatures, neutralization, am, low sodium to boron ratios,
especially with tetraborates as the bitumen hardens making stirring
impossible. ORNL never established this method for commercial
or routine use (9).
Studies were initiated for incorporation of concentrated solutions
using a thin-film and screw extruder evaporators; and the screw
extruder evaporator concept was developed at the Eurochemic Plant
at Mol, Begium and the Karlsruhe Nuclear Research Center in West
Germany with apparently good success at both facilities (lO, 19). The
industrial bituminization plant for evaporator concentrates utilizing
the screw-extruder-evaporator has been sucessfully operated and
commercialized by Werner & Pfleiderer Corporation at an evaporation
capacity of 140 kg of water per hour with the final product containing
50% salts (i6, 17, 19). ORNL has also investigated the film evaporator
process, along with Marcoule, for incorporation of industrial, urban
and radioactive wastes in asphalt (8, 9, 15).
Additional research and operational work involving radioactive
wastes in bitumen has been done by other countries and nuclear facilities;
specifically, the USSR, Bulgaria, Japan, Hungary, and Austria (9, 15).
The Hungarian Mineral, Oil and Natural Gas Research Institute developed
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a technique for the incorporation of nitrate-containing wastes. The
technique utilizes an inert gas being introduced during the process
which eliminates the danger of fires and explosions; the gas also ensures
adequate mixing which results in a more satisfactory dispersion of the
salts in the bitumen. Workers in India and Japan (9) prefer incorporation
of radioactive wastes in bitumen emulsions with the addition of suitable
surface active agents. The ease of mixing, low water content in
product and reduced leaching rates in sea water are main reasons
for this preference. The Soviet Union has a great deal of interest
in the solidification of radioactive concentrates by incorporation in
bitumens; they have had a pilot plant in operation since 1969 to
evaluate the conditions and feasibility of incorporating wastes
enriched in sodium nitrate. Also, both the British and Russians
have developed methods for immobilizing radioactive wastes attached
to netural and synthetic sorbents such as vermiculites, zeolites,
clinoptilolite, and ion exchange resins, into a bitumen matrix (9, 15).
The main reported advantages and disadvantages of using
bitumen for the insolubilization of radioactive wastes are as
follows (9, 12, 15):
Advantages
(a) The leach r^te of th^e final product can be expected to be
between 10 and 10 times lower than similar cement mixes.
(b) There are a large number of types of bitumen with a wide
variety of properties; thus it is usually possible to obtain
a suitable material for any waste.
(c) Bitumen has good coating characteristics and good adhesion
to incorporated material.
(d) The solubility of bitumen in water is negligible.
(e) Bitumen possesses a degre e of plasticity and elasticity which
are of benefit during the incorporation process.
(f) Bitumen, is resistant to attack by microorganisms.
(g) There is some evidence that bitumen is more suitable than
cement for wastes which ei dt emanations.
10
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Disadvantages
(a) There is always an inhere nt risk in working with organic
material at elevated temperatures; however, there is no
evidence that incorporation of inert material into bitumen
increases the risk of fire or explosion.
(b) There is some evidence t iat the presences of nitrates or
nitrites and other oxidizing agents can increase the fire
risk.
(c) It is obvious that no substance should be added to bitumen
which decomposes at the working temperature. Difficulties
may be experienced with certain plastics and compounds
such as sodium citrate.
(d) Heating of bitumen mixe.c can result in the releases of oils,
fumes and mercaptans.
(e) It is necessary to work at high temperatures to obtain
efficient mixing.
(f) To obtain the best final product it is necessary to remove
as much as possible of the water present in the waste.
Leachability increases significantly with increasing amounts
of retained water.
(g) Strict temperature control is required in the bituminization
process.
(h) Mixing tetraborates and iron and alu ninium salts with
bitumen causes hardening to an extent which can interfere
with discharge of the final product from the equipment.
(i) Irradiation of bitumen modifies the chemical and physical
properties. In some cases the effect is negligible and in
others considerable. Irradiation up to an integrated dose
of 10*° rad by an external source or up to 10 rad due to
the incorporation of radionuclides can be accepted
provided a suitable type of bitumen is chosen.
(]) Phase separation in bitumen mixes is likely to occur more
readily than in cement-waste products, particularly
during accidental fires in transport or storage.
(k) Experiments at Marcoule and Karlsruhe have shown that
swelling of certain bitumen products can occur in water.
11
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IV. SOLIDIFICATION OF RADIOACTIVE WASTES
IN CEMENT
The cement solic ification process with and without additives
has been in common f ractice at nuclear waste treatment facilities
for many years in difierent countries on an industrial scale (' , 12,
13, 14, 20)'. In 1968, the IAEA concluded that low-level sludg •
incorporation into cement is considered an at equate treatme! t (9).
As recent as 1974, both bitumen and cement incorporation wore
considered by IAEA to be acceptable techniques (12).
Cement has been described as an adhesive substance, lime being
the principal constituent, capable of uniting fragments of solid matter
into a compact monolithic structure. The most commonly used
cement for the incorporation of r idioactive wastes is the "Portland"
variety. It is obtained by intimaiely mixing silica-, alumina-and
ferric oxide-bearing materials tc the lime and burning these
materials to a very hard brick ai d grinding the resulting brick.
There are various types of portl; nd cement depending on the
fineness of the grinding and on th 3 addition of certain additives
or the amount of various constitu mts. Portland cement Type I
has been most commonly used, but other Types, such as Type V
which is resistant to sulfate salt deterioration, have also been
employed (22, 23).
Basically all the cement solidification processes consist of
mixing the cement wit i waste solution, slurry, or solids within the
receptable. The actual kinetic process leading to the curing of cement
is not known. However, the mechanism for the setting of cement,
according to the literature (9, 22, 23), has been postulated to include
a reaction between water and cement which formed solid particles
and causes crystallization of the calcium hydroaluminate, hydro-
ferrite, and hydrosilicate with tb : crystals giving the strength to
the hardened cement. If the cem mt product is to be in satisfactory
condition for transportation or b\ rial it must have adequate compressive
strength. It is common practice to neutralize the acid wastes before
the cementation process and to control the salt content. Poorly cured
cement will crack anc spall, cau: ing more surface area to be exposed
to leaching conditions.
The mixing of the cement with the various radwaste forms, i.e. ,
sludges, resin beads, etc., affecis the properties of the product. The
strength of the cement will be a function of the total salt content in
the sludges, resins, etc. , where there is a narrow range in the accept-
able values for the ratio of basic and acidic oxides in the final product.
In this regard, waste to cement ratios recommended for proper curing
12
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vary significantly among both foreign and domestic suppliers (see
Table II) (9). USSR studies have shown that in order to produce
cement of acceptable structural strength the concentration of
sodium nitrate salt should preferably not exceed 130 g per kg of
cement (9).
Some of the U.S. utilities employ a combination of vermiculite
and cement to solidify their radwaste. The expanded vermiculite
is porous, permitting the infiltration of dry cement into the ver-
miculite structure. This would act like a sponge absorbing the
liquid and giving a better final product than when cement is used
alone (i4). Two cement solidification vendors, United Nuclear
Industries (UNI) and Delaware Custom Materials, have developed
a process which utilizes sodium silicate as an additive with portland
cement (24,25). However, t le addition of sodium silicate to
cement-waste mixtures incr ases the volume of waste per volume
of solid formed.
In the non-commercial area of solidification ORNL blends their
radioactive wastes with a dr> mixture of cementitious materials
and clays. The dry solids consist of a mixture of portland cement
Type I, with a variety of clays including grundite which has a high
retention capacity for cesium (14).
The presence of water, nitrates, sulfates, borates, and other
unstable (in a radiation environment) compounds in the cement
could give rise to gaseous radiolytic products (13). Gases also
could result from volitalization of compounds by elevated tempera-
ture in the cement-waste mixture causing voids to form within
the crystalline structure.
Brookhaven National Laboratory (BNL) found that the leach-
ability properties of cement could be improved by developing a
polymer impregnated concrete (PIC) matrix (26, 27). PIC composites
containing tritiated aqueous waste, solid calcine, incine -ator ash,
aqueous and solid sodium nitrate, reactor waste, acidic and neu-
tralized fuel reprocessing wastes, and ion-exchange an
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Table II
Cementation Practices at Various Establishment (9)
Establishment
France
F. R. Germany
USA
Czechoslovakia
Nature of Waste
(a) Evaporator
(400 g/liter)
(b) sludge
Evaporator
concentrate
(a) Evaporator
concentrate
(20% solids)
(b) Neutralized
concentrated
USSR
(c) Evaporator
Resins
(d) Evaporator
R esins
(a) Sludge with solids
content of 20 to 25%
(b) Evaporator concen-
rate neutralized to
pH 6 to 8 (200 g/liter
Evaporator concentrates
(Max. 150 g/liter)
Composition of Mixture
250 liters sludge
300 kg cement
40 kg vermiculite
83 kg sludge
55 kg cement
100 to 110 liters
150 to 200 kg cement
Vermiculite (2.7 m3)
and Portland cement
(0. 68 m3)
(i) 75 liters of concen-
trate, 128 kg cement,
4 kg vermiculite
(ii) 20 to 35 liters/min
concentrate, 60 to
65 kg/min of cement
91 kg of cement
100 liters of waste
3 to 1 ratio cement tt
waste with a sodium
silicate additive
35 liters of sludge
110 kg of cement
10 kg of sludge
5 kg of evaporator
concentrate
22 kg of cement
130 •» of salt of the
sodium nitrate-type
per kg of cement
14
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Operational Experience in the United States
As exhibited in Figure 2, the cement and radwaste could be
mixed either within the shipping container or prior to loading the
shipping container. For exampje, ATCOR performs all its mixing
with an in-line dynamic or mecl anically driven mixer (28). Cement
and the liquid radwaste are dri\ 3n into one end of the mixer and a
homogeneous mix is discharged into the shipping container (which
has vermiculite added) where in-container solidification occurs.
Often the cement will pre-hardea causing the mixer to jam.
Batch mixers have also been employed. The components are
introduced into a mixer; a mixing blade blends the constituents
and the mixture is drained into u receptacle. Earlier plants employed
roller and tumbler mixers. When the cement is initially loaded in
a drum with a definite mixing weight, a measured quantity >f waste
is injected into the drum, and the drum is physically rolled and/or
tumbled (9, 11,12).
Another system that can be characterized as an in-drum mixing
process has been developed by Stock Equipment Company (S-E Co. )
(29). Since transport of fresh cement has historically presented
difficulties due to the premature hardening and resultant incomplete
curing of the waste, S-E Co has developed a process to overcome
this difficulty, by having mixing take place in the final storage
drum at the rate of 50 to 200 kg of waste per hour. S-E Co has
concluded that the quantity of cement and/or additive in each 208
liter (55 gal) drum averages about 91 kg (200 Ibs) and the amount of
radwaste averages about 106 liters (28 gal). For this system, cap
removal, filling, cap replacement, and mixing is an automatic
operation. Therefore, the operator does not have to estimate the
correct prescription for solidification.
The UNI and Delaware Custom Material (which uses the Chem-
Fix process) systems for solidifying radwaste use an in-line
batch mixer for waste and cement which is then mixed with sodium
silicate in the shipping container. The UNI and DCM systems
provide: (a) proportional pumps for metering waste feed; (b) in-line
mixer to assure homogeneity; (c) single fill port for wastes; and
(d) in-container solidification (24, 25, 30, 31).
The main reported advantages and disadvantages of using cement
for insolubilization are as follows (9, 11, 12, 13):
Advantages
(a) No complex equipment; it is often possible to carry out the
incorporation in the disposal receptacle.
15
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SYSTEM FOR ENCAPSULATING
RADWASTE
CEMENT
MIXES RADWASTE & CEMENT
WITHIN THE SHIPPING CONTAINER
MIXES RADWASTE & CEMENT PRIOR
TO LOADING SHIPPING CONTAINER
ROLLER
MIXERS
TUMBLER
MIXERS
PADDLE
MIXERS
IN-LINE
MECHANICALLY
DRIVEN MIXER
REMOVEABLE
PADDLE MIXER
BATCH
MIXER
DISPOSABLE
PADDLE MIXER
BASICALLY LIMITED TO
SMALL SHIPPING
CONTAINERS
(eg. • 55 GAL DRUMS)
L
BASICALLY LIMITED TO
RELATIVELY SMALL SHIPPING
CONTAINERS WITH SPECIFIC
CONFIGURATIONS & FEATURES.
SMALL OR LARGE SHIPPING
CONTAINERS OF VARIOUS
CONFIGURATIONS & WEIGHTS
(eg. SHIELDED OR UNSHIELDED)
CAN BE USED.
FIGURE 2
FLOW DIAGRAM FOR
CEMENT INCORPORATION PROCESSES
16
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(b) Low capital investment and low running costs; power require-
ments minimal.
(c) No applied heat required; low operating temperature means
no fire risk and eliminates difficulties with off-gas purification.
(d) Most systems fully automatic; and therefore, operators can
be trained easily.
(e) Waste-cement mixes are not grossly affected by pi .
(f) Cement is relatively cheap, but this is often off-set by the
greater quantity required.
(g) Chemical and physical properties of cement well known.
(h) Cement imparts good shielding properties.
(i) Natural alkalinity of cement is useful in helping to neutralize
acidity in waste solutions.
(j) Little reported trouble with phase separation in the mix.
(k) Water is required for s itting the mix so there is no need for
extensive dewatering pi ovided a satisfactory water/cement
ratio is maintained.
(1) Presence of nitrates an 1 nitrites and other oxidizing agents
do not have the same detrimental affects as they can have
when mixed with an organic material such as bitumen.
(m) Less subject to irradiation damage than bitumen.
Disadvantages
(a) The concentration of ce -tain salts, .^uch as borates, may
cause the cement-wasti matrix some difficulty in curing
and causing deteriorati >n over time thus leaching at an
abnormally high rate.
(b) The weight and volume of the final product will normally be
about twice that for other corresponding solidification
processes. Experience at Fez, Czechoslovakia (a govern-
ment laboratory), is that 1.5 m3 of filtered sludge, containing
20 to 25% of solids, wot Id result in a volume of 4. 8 m3 after
incorporation into cemr at. The weight and volume increase
is mainly due to the arr )unt of cement which must be added
to react with the residual water in the waste.
17
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(c) If mixing equipment experiences operational trouble and
frequently breaks down, this could require frequent cleaning
of the equipment, particularly the blades.
(d) Nonautomated systems require several manual operations
during the solidification process.
(e) Most studies have shown that when buried and after the
container rusts away the cement will leach if in contact with
ground water.
(f) Cement is relatively cheap, but this is often off-set by the
greater quantity required.
(g) Chemical and physical properties of cement well known.
(h) Cement imparts good shielding properties.
(i) Natural alkalinity of cement is useful in helping to neutralize
acidity in waste solutions.
(j) Little reported trouble with phase separation in the mix.
(k) Water is required for setting the mix so there is no need for
extensive dewatering provided a satisfactory water/cement
ratio is rr aintained.
(1) Presence of nitrates and nitrites and other oxidizing agents
do not have the same deterimental affects as they can have
when mixed with an organic material such as bitumen.
(m) Less subject to irradiation damage than bitumen.
18
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V. ABSORBENTS
Absorbents are used to eliminate free standing liquids by virtue
of their ability to hold water molecules within their pores. The
absorbent is, however, not chemically bound to the waste nor does
it represent a free standing monolithic solid; therefore, the absor-
bents should not be considered as solidification ager ts. Further,
they do not provide or enhanc e resistance to leaching, if water comes
in contact with the absorbed radioactive materials. The tbsorbents
are stored in a diiy environment and are placed in the shipping
container prior to adding radioactive liquids. Some comr lonly used
absorbents are vermiculite, clays, silica gel, plaster of paris,
microcell and/or diatomaceous earth filter aid (13, 20). The prime
use of absorbents is at older plants that do not have installed solidi-
fication systems. Not all the burial facilities will accept wastes
shipped with an absorbent, because of its unacceptable properties,
such as leachability after burial (see Table IIC).
Vermiculite, dehydrated clay granules, and diatomite absorbents
have been routinely used for liquid wastes, with perhaps vermiculite
the most widely used. The absorl ent method, when properly applied,
will physically entrap the waste liquid so that no appreciable free
liquid will leak out if the container is breached. With most of the
absorbents, the liquids are physically entrapped and can be displaced
readily by the addition of water.
In preparation the receptable is filled with vermiculi';e and
liquid waste equivalent to about 1/3 to 1/2 or the volum? . For
some materials such as the diatomaceous earth, physical mixing
of the liquid and absorbent may be necessary. Care must be taken
with all absorbent to avoid supersaturation.
19
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Table III
BURIAL REQUIREMENTS AT THE SIX COMMERCIAL BURIAL SITES
Requirements N.Y. * KY. ILL. NEV. WASH. S.C.
A. ACCEPTANCE OF RADWASTE TYPES
1. DEWATERED RESINS NO YES NO YES YES YES
2. DEWATERED POWDEX NO YhS ? YES YES NO
3. DIATOMACEOUS EARTH ? YES ? ? YES NO
4. DEWATERED SLUDGES NO ? NO YES YES NO
5. FREE LIQUIDS NO NO NO YES YES YtS
B. SOLIDIFICATION AGENTS
1. ALL TYPES OF CEMENT NO YES YES YES YES YES
2. UF SYSTEMS NO ? YES YES YES YES
3. OTHER ORG. POLYMERS NO ? ? ? ? NO
4. ASPHALT NO ? ? ? ? ?
C. SELECTED REQUIREMENTS
1. RLTRIEVABILITY NO NO NO NO NO YES
2. PU LIMITATIONS YES YES YES YES NO YES
* PRIOR TO ITS CLOSING ON 3/19/75.
? UNCERTAIN ABOUT ACCEPTANCE
20
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VI. POLYMERIC SOLIDIFICATION PROCESSES
Incorporation of radioactive wastes into polymeric fixation agents
is a relatively new solidification process when compared to incor-
poration in cement or bitumen (13,14). The solidification process
can take place either at ambient temperatures or with hot evaporator
concentrates (up to 60°C). Presently, several U.S. companies sell
urea-formaldehyde (UF) solidification systems (as shown ir Table
I). All the organic processes are essentially batch processes where
a catalyst is generally mixed with the wastes and polymer either in
a premixer vessel or in the receptacle itself (14, 32, 33). The poly-
meric processes do not really solidify the wastes; the long chain
molecules of the organic polymer are linked together to form a
multi-voided sponge that "traps" the waste. Not all U.S. burial
sites at this time, however, will accept radwaste solidified with
an organic polymer (see Table III).
Paraffin and polyethylene based solidification agents can also be
used to solidify wastes. These agents must be liquified by heating
prior to mixing with the wastes (20, 21, 34).
The only industrial experience with polymeric solidification systems
to date has been with the UF process. The process description, advan-
tages and disadvantages are based on systems using UF, whereas there
are other organic polymer processes, such as the Dow Chemical and
the Todd Research processes (35, 36), which are either not operational
or have not been in operation long enough to provide operational infor-
mation comparisons.
Urea Formaldehyde
The physical method of organic polymeric mixing depends upon
the type of solidification agent and receptacle used. In general,
there are three types of UF mixing:
1. In-container disposable paddle mixer
2. In-line static mixer
3. In-line mechanically driven mixer
The in-container mixer is generally used for cement (f-ee Chapter
IV) but is also used for mixing resin beads with UF. The in-line
dynamic mixer is used by UNI to mix liquid waste and UF prior o
to discharging into a receptacle. UF systems generally employ ,in
21
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in-line static mixer which contains stationary helical vanes to mix
the fluids as they flow through the mixer. Just as the mixed polymer
and waste are injected into a container, the acidic catalyst is added
to initiate solidification. Figure 3 is a flow diagram for UF indicating
the incorporation steps.
The UF solidification process system developed by Protective Pack-
aging, Inc., (PPI) is called Tiger Lock. Tiger Lock is a registered
trade mark for a proprietary augmented agent of urea formaldehyde
resin that is manufactured to strict physical and chemical specifications.
The PPI system currently being sold includes Tiger Lock and catalyst
(usually sodium bisulphate-Na2 S04), associated processing equipment,
and container liners for use in the transportation and burial of the
solidified radwaste.
PPI suggests that the desired ratio o.' Tiger Lock to radwaste
is 2:1 by volume. For this system the operator has the task of
estimating the correct amount of catalyst for solidification which
is highly dependent upon the quantity and type of radwaste that would
be solidified by the PPI systom.
Gel time of the product can be adjusted from minutes to hours by
the catalyst concentration (normally about 1 to 3%). The UF catalyst
is used at a pH of 3 (32). If UF is used after its shelf life has been
exceeded or at low temperatures or low viscosity, the "cottage cheese"
effect will occur, i.e., little solidifying and essentially a settling of
materials of different density within the container.
The main reported advantages and disadvantages of using urea-
formaldehyde for insolubilization are as follows (13, 14):
Advantages
(a) The amount of waste capable of incorporation in a receptacle
with UF is about 30% by volume more than with cement.
(b) For shipping not requiring radiation shields, shipping cost
with UF or polymerics is less than for cement and bitumen
due to the ability to put more waste in a given container and
a Lower density for the mixture.
(c) Mixt ire of UF and radwaste are not combustible. Further,
no detectable exothermic reaction occurs with UF.
22
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UF
MIXES RADWASTE, UF & CATALYST
PRIOR TO"LOADING SHIPPING
CONTAINER
MIXES RADWASTE, UF & CATALYST
WITHIN THE SHIPPING CONTAINER
IN-LINE STATIC
MIXERS
IN-LINE
MECHANICALLY
DRIVEN MIXER
PADDLE MIXER
SMALL OR LARGE SHIPPING
CONTAINERS OF VARIOUS
CONFIGURATIONS & WEIGHTS
(eg. SHIELDED OR UNSHIELDED)
CAN BE USED.
FIGURES
FLOW DIAGRAM FOR UF
INCORPORATION PROCESSES
23
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Disadvantages
(a) For shipments requiring radiation shields, UF or polymeric
solidified materials, due to its lower bulk density and higher
activity, requires more shielding than materials solidified
with cement.
(b) It aptpears that routine attainment of the complete elimination
of free standing water is a problem with encapsulated UF rad-
waste, particularly those having lower concentrations of the
polymer (ratios of 1 to 3 or less) (37).
(c) According to the utility operators, it is difficult to work
with UF because of the relatively low viscosity of the mixture,
which permits settling or floating (segregation) of materials
of different densities.
(d) Solidification time is affected by both the pH of the mixture,
which is regulated by the amount of catalyst, and the temperature
of the mixture. Operating experience at the utilities indicates
that the optimum conditions for solidifying are 29 C (85 °F) and
apH of 3 (37,38).
(e) The UF shelf life is limited and is dependent upon slorage
conditions.
(f) Equipment must be designed to eliminate fume problems
with UF; the odor is disagreeable even in small concentrations.
(g) Some manufacturers of UF have stated that this product is
biodegradable; also the catalyst is corrosive to most metals.
(h) During the solidification process when the UF-radwaste
mixture is exposed to air, water vapor evaporates from the
mass, but if the matrix remains in an air-tight container, the
mixture will remain semi-liquid.
Polyethylene Process
Polyethylene agents are not used commercially in the U.S. Poly-
ethylene is a superior solidifying agent for most organic liquid wastes.
The waste is combined with molten polyethylene inside a heated chamber
in which the water and other volatile constituents are evaporated. The
mixed and dehydrated liquid product is discharged to a container where
it solidifies upon cooling. The final product is a solid plaster block,
which is relatively inert at room temperature and is insoluble in water.
It has good freeze-thaw characteristics and a storage life of several
years. Polyethylene is completely combustible and can be incinerated.
It is flammable with a flash point of 250°C (20, 21, 34).
24
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VII. LISTING OF WASTE SOLIDIFICATION AND PACKAGING
SYSTEMS SUPPLIERS
The previous sections discussed the various solidification methods
while briefly mentioning some of the companies offering the process.
This section is intended to provide a listing of those companies which
supply partial or complete waste solidification and packaging systems.
These systems may or may not include options for using either cement
or polymerics and in many cases these systems do not have actual on-
line processing experience in tire United States nuclear industry. The
coverage of each company in this listing varies in length and detail,
but in most cases is representative of their process information
brochures. The reason for this is the lack of actual use of the
systems, and in many cases the process ingredients are proprietary.
1. Werner & Pfleiderer Corporation (..0,16,17, 19, 28, 58)
Werner & Pfleiderer Corporation, (WPC) offers a waste solidi-
fication system that includes a volume reduction system. With asphalt
as the solidification agent, WPC indicates the end product is stable,
particularly against leaching. Although it has only been recently
introduced in North and South America, it has been operating routinely
in Europe since 1965. The WPC radwaste solidification process yields
a liquid-free (0. 5%) solid using a continuous, fully automatic process
with a multi-screw compounding extruder. The extruder-evaporator
simultaneously provides homogenous mixing (including reagent additives),
liquid evaporation and solidification in one machine. The extruder
evaporator normally discharges the asphalt/salts mix into standard
DOT 208 liter (55 gal) drums at a rate from 1 to 114 liters (1/4 to 30
gal) per hour, depending on the the size and speed of the extruder and
the concentration of the feed stream. The entire process, complete
with interlocks, can be controlled remotely.
The WPC extruders, originally developed for the plastics industry,
are designed and built to operate a full year without maintenance.
The operating record established in Europe bears this out and is
quite impressive (over 134, 000 hours operation at Marcoule without
mechanical failure).
2. ATCOR, Inc. (28)
ATCOR has developed a process system to mix liquid and solid
wastes with a cement or a cement and vermiculite mixture to produce
a solidified product within a disposable receptacle. Basically, the
system mixes separate feeds of moist radioactive waste or evaporator
25
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concentrates and dry cement in a small volu rte continuous mixer.
Solid waste materials are preconditioned within the radioactive waste
feed tank to provide sufficient moisture when mixed with the dry cement
to achieve an acceptable cement mixture. The system not only solidifies
resins, sludges and evaporator concentrates, but it can also be used
to fix spent filter cartridges within a solidified matrix. In this case
drums or large volume liners containing spent filter cartridges could
be filled with a cement mixture that contains radioactive wastes.
The cemented waste mixture can be loaded directly into standard
208 liter (55 gal) drums or larger receptacle. Where waste is to be
packaged in drums, drum capping and decontamination can also be
provided. There is no preparation of drums required prior to filling.
3. Stock Equipment Company (28,29)
The Stock Equipment Company (S-E-Co. ) Solid Radwaste System
is designed and manufactured as a completely integrated system
utilizing components which are designed specifically for the service
expected rather than attemping to modify standard equipment. The
S-E-Co. System is furnished complete for placement into the radwaste
building and interfacing with liquid system piping, utilities, etc.
The S-E-Co. design uses cement as a solidification agent and
packages the solid radwaste into standard DOT, 208 liter (55 gal)
drums.The S-E-Co. system is also easily adaptable to other types
of solidification agents such as urea formaldehyde. The S-E-Co.
waste system consists of: a cement storage hopper; a storage
tank to hold liquid wastes that conta n concentrated solutions of
dissolved solids; a decant tank for f]lter media, resins, and/or
the solid waste slurry from the storage tanks; a drum processing
unit which is fully automatic for uncapping the drums, filling the
drums with cement, filling from the decant tank, reinsertion of
the cap, and for the mixing/tumbling operation.
4. United Nuclear Industries, Inc. (24,28,30,31)
United Nuclear Industries (UNI) offers radwaste solidification
systems utilizing as the solidification agent either urea formalde-
hyde (UF) or Portland cement with sodium silicata as an additive.
The use of the UF material permits the use of in-line static
mixers with no moving parts. Solidification of the waste - UF
mixture is accomplished using either a sodium bisulfate catalyst
(pH range of 3 to 7. 5) or a phosphoric acid catalvst (pH range of 3
to 10).
26
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5. Aerojet Energy Conversion Company (28, 39-43)
The Aerojet 3nergy Conversion Company (AECC) has marketed
a VR-20 Radioactive Waste Management System which reduces the
volume and encapsulates the waste. The volume reduction is achieved
by conversion of all liquid wastes into anhydrous calcined solids and
drying of dewatered spent re >ins and t ludges. These solid wastes
can then be encapsulated in cement, U-F or bitumen as the solidifi-
cation agent for subsequent shipment ;oid burial. Using the VR-20
process, the volume of liquic waste could be reduced by a factor
of 10-20, while the volume o! liquid spent resins and sludges could
be reduced by a factor of 2-4 when compared with other solidification
methods.
The main feature of this system is a fluidized bed calciner which
receives the radioactive liquid waste feed containing the dissolved
chemical solids and processes these aqueous solutions into free-
flowing anhydrous particles. Concentrated radioactive liquid waste
(evaporator bottoms, etc.) is pumped from the concentrated liquid
waste storage tank to a heated fluidized bed calciner concentrator.
The volatiles exist with the water vapor at the top of the fluidized
bed concentrator, leaving behind the dissolved solids. The granular
solid produced can then be encapsulated.
6. Chem-Nuclear Systems (28,44)
Chem-Nuclear Systems Iiic., offers either portions of or a
complete waste system design, component selection, procurement,
fabrication, construction, installation and operation of solidification
systems using either cement or urea formaldehyde as the solidification
agent. Chem-Nuclear also has available a mobile .solidification unit
using the UF system.
7. Protective Packaging, Inc. (28,32,33)
Protective Packaging, Inc. (PPI), a wholly owned subsidiary of
Nuclear Engineering Co., developed and was the first company to
design and sell, a system using a chemical solidification agent other
than Portland Cement. Since then, they have filed several patent
applications on the system and trademarked the name "TIGER-LOCK".
Their patent applications cover both the use of the liquid solidification
agent (TIGER-LOCK, a type of urea formaldehyde polymer), and
also all the related hardware that makes up a TIGER-LOCK Radwaste
Solidification System. This includes the process equipment, control
panel, power panel, and associated material handling equipment.
The key aspects of the PPI design are: (a) three separate tanks,
27
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pumps, different size liners for radwaste, TIGER LOCK, and catalyst
respectively; (b) a premixer for the radwaste and UF to homogenize
the slurry prior to contact with the catalyst; (c) a manual decoupling
device to seal the liners that contain the cured waste; and (d) an
automatic level detector to indicate filling to 90% volume.
8. ANEFCO, Inc. (28)
ANEFCO, Inc., offers a waste solidification syst jm using urea
formaldehyde and sulfuric acid, or an equivalent catalyst, as a
solidifying agent. Their process system uses a 3785 liter (1000
gal) batch tank, a static mixer aid a disposable polyethylene liner
in the disposal container.
9. Hittman Nuclear & Development Corporation (28, 45)
Hittman Nuclear & Development Corporation (HNDC) offers
radioactive waste solidification systems using cement or a polymer
such as urea formaldehyde as the solidifying agent. Chemical
additives are used with both ag( nts to enhance the efficiency, i. e.,
volume of waste per unit volum i of solidified product. The
disposable containers used to package radwaste vary in size from
a standard 208 liter (55 gal) drum up to 5. 7 m (200 ft ) capacity.
10. General Electric Company (28, 46)
The present General Electric Solid Radwaste Systems use cement
as the solidification agent with a disposable mixer and large disposal
containers.
11. Westinghouse Electric Corp. (47)
The Westinghouse Waste Encapsulation System is basically a
vacuum packaging process in which spent radioactive resins and waste
evaporator bottoms are encapsulated using a cement vermiculite
mixture in standard DOT 17 H drums.
12. Delaware Custom Materials (25)
The Delaware Custom Materials company utilizes the Chemfix
process which offers a complete service of equipment and chemicals
for solidification of radioactive wastes, including a variety of
inorganic and organic sludges. The process uses a combination of
cements, shales and clay as the solidification agent.
28
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13. Dow Chrmical Company (35)
Dow Chemical Company has developed a radwaste solidification
system that produces radwaste free of liquid, reasonably hard, and
free standing. The solidification system is usable for all anticipated
chemical decontamination solvents and regular wastes from nuclear
power stations.
To date, Dow has solidified the following simulated wastes in the
laboratory and in 208 liter (55 gal) drums, containing no detectable
free liquid: (a) spent decontamination solutions at pH's of 3 to 5
and 9 to 10 with 40% solids; (b) filter aid and slurries, 90/10 by
volume; (c) ion exhange resins, 90/10 by volume; (d) PWR evaporator
bottoms with a pH of 2. 5 and 7% solids; and (e) BWR evaporator
bottoms with a pH of 10.6 and 6% solids. After casting, the drums
solidified within one hour. The radwaste to agent ratio is at least
1. 25 to 1 and as high as 2. 5 to 1. A field demonstration was carried
out by successfully solidifying 3400 liter (900 gal) of radioactive
decontamination solvent at a nuclear power plant.
To simulate disposal conditions, Dow evaluated the solidifi-
cation product for the following: (a) compressive strength;
(b) temperature cycling; (c) radiation stability; (d) leachability;
(e) impact testing; (f) heat exposure; and (g) free liquid.
14. Todd Research and Technical Division (36)
Todd Research and Technical Division is marketing a solidifi-
cation agent called SAFE-T-^ET, which is a long chain linkage
organic polymer. The agent can be used with concentrated low-level
liquid radioactive wastes fro n filtration, precipitation, ion-exchange
or evaporacion. The set-up time varies from one minute to several
hours depending on the amount used in proportion to the volume of
waste. One-half kilogram (1 Ib) of SAFE-T-SET will solidify 3. 8
liters (1 gal) of liquid material. The agent can be tailored to any
particular system or circumstances including pumping the waste and
SAFE-T-SET mixture and can be adapted to molds of any type. The
solidified matrix remains stable under conditions of freezing, high
temperature and leaching.
15. United Technologies (48)
United Technologies - Chemical Systems Divison offers the
Inert Carrier Process which handles ingredients in an inert liquid.
The operating concept is based on dispersal of the reactants in
an inert carrier to provide maximum surface area for solid -
liquid reaction mechanism. In addition the process provides for
29
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a clean separation of the reaction product from the inert carrier.
The waste materials are low viscosity dispersions in an inert
carrier. The inert carrier i,s a fluid selected so that neither the
starting materials nor the products are solub.e in it or chemically
reactive with it. The process has particular advantages in operations
which require (a) preparation of compositions which are too viscous
to mix by ordinary methods; (b) extremely intimate mixing of solids
with small quantities of liquids; (c) safe control of highly exothermic
chemical reactions; or (d) a closed system and/or remote controlled
processing of hazardous, toxic, or explosive materials.
16. Energy Incorporated (49)
Energy Incorporated and Newport News Industrial Corporation has
developed a Radioactive Waste Reduction (RWR) system to convert
all low and medium level liquid and solid combustible radioactive
wastes to solids by a fluidized bed calcining process. The system
produces a granular, anhydro is solid which may be placed directly
in burial containers or incorporated into matrices such as concrete,
urea formaldehyde or bitumen for burial.
30
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VIII. CONCLUSIONS AND RECOMMENDATIONS
In the foregoing sections brief descriptions have been presented
of several established and proposed processes for the solidification
of low-level radioactive wastes. Each of the processes as well as
each of the solidified waste products (cement, bitumen, UF, etc.)
have a number of advantages and disadvantages. Table I listed the
solidification agents and/or systems with each supplier. In general,
the following conclusions can be drawn concerning the three major
systems:
. Bitumen - some question concerning thermal stability,
particularly above ambient temperature, but
possesses good mechanical ruggedness and
radiation resistance.
. Cement - good thermal stability, mechanical ruggedness,
and radiation resistance but questionable ability
to properly fix certain salts with the sludges or
resins; also some system operation problems.
. Urea Formaldehyde- questionable radiation resistance,
thermal stability, and biodegr ad ability
properties but the best waste to agent
ratio; also some process operation
problems.
. Other Organic Polymers - not enough actual on-line
experience to provide an
adequate conclusion.
The most likely mechanism of radionuclide release to the
surroundings is by solution in the water existing in the environs of
the burial site. Therefore, measurements are usually attempted
to indicate the rate at which radionuclides are leached from the
solidified products. The leachability properties of a radwaste
solidification matrix will strongly influence the amount of treatment,
containment and surveillance that will be required.
A review of the various It ach rate tests conducted by industry,
(50-57) indicates that the res ilts were not developed under stan-
dardized laboratory conditions and were not correlated to simulate
current burial conditions at the various commercial shallow-land
31
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sites. Studies currently underway which are investigating the
properties of solidified wastes include: (a) a study by Brookhaven
National Laboratory (under contract with NRC) to evaluate various
solidified wastes generated by commercial nuclear power plants and
establish a standardized leach test for solidified matrixes (57); (b) a
study by the Army Corps of Engineers (under contract to EPA-SWRL-
Cinn) to evaluate solidification agents for solidifying hazardous waste
sludges (52); and (c) a study of leach rates for various solidified
wastes by the Oak Ridge National Laboratory (50, 51, 54).
The studies are intended to provide information on the leach
resistance of various solidified waste products. Leach rates for
alkali and alkaline earth, rare earth, and actinide elements from
various waste matrices are compared in Table IV (20). A comparison
shows: (a) cement has wide ranging leach rates; (b) calcines are
extremely leachable; that for a given waste matrix the leach
rates for rare-earth and actinide elements are about a factor of
1,000 less than those for alkali and alkaline earth elements; and
(d) the leach rates for rare-earth and actinide elements from
cements and grouts are as low as those from glasses. Additional
leachability work needs to be accomplished in the near future to
permit a more complete understanding of the environmental impact
of solidified radioactive waste.
Table III indicated some of the burial requirements at the six
commercial sites. The requirements have been arranged into
three categories: (a) the type of radwaste shipped by the utilities
that are being accepted; (b) the solidification systems that are
being accepted; and (c) other requirements. The question marks
indicate uncertainties the burial sites have concerning certain
requirements. Several key aspects which can be pointed out are:
(a) there are inconsistencies concerning what each burial site
will accept; and (b) there is a lack of information concerning the
merits of the various solidification agents, so that the licensing
agencies can pass judgement on what matrix media would be
acceptable for burial.
The quality assurance requirements for the various solidifi-
cation processes should also be improved at the nuclear plants.
It is important that each waste shipping container have a consistent
composition of solidified matrix. This condition would improve the
reliability of the leachability measurements.
In conclusion, this summary has drawn together in a list the
various vemdors of solidification processes and systems. There is
definitely a need for specific standards for leach testing the solidified
matrices and developing standardized burial requirements for the
sites
32
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Table IV
COMPARISONS OF LEACH RATES FOR VARIOUS SOLIDIFIED WASTE PRODUCTS (20)
LEACH RATES GRAMS/CM2-DAY
Calcines
Ceramics
Phosphate
Devitrified
Phosphate glass
Glasses
Borosilicate
Phosphate
Aluminosilicate
Bi tumens
Cements
Grouts
Alkali and
Alkaline-Earth
10-4-10-2
io-7-io-5
io-8-io-5
10-8-1Q-7
io-7-io-4
Rare
Earth
io~4-io-3
10-9-1Q-6
10-9-10-7
10-9-1Q-6
Actinide
10-8-10-7
10-9-10-7
10-7
33
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39
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