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MICROCOPY RESOLUTION TEST CHART
NATIONAL BUREAU OF STANDARCS -1963
in
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CESTI PRICES
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NOV 2 9 1968
HC $ 1.00 50
Conf-660920-9
DISPOSAL OF RADIOACTIVE WASTES IN GEOLOGIC FORMATIONS
W. J. Boegly, Jr., F. L. Parker, and E. G. Struxness,
Health Physics Division
Oak Ridge National Laboratory
Oak Ridge, Tennessee
MASTER


RELEASED FOR ANNOUNCEMENT
IN NUCLEAR SCIENCE ABSTRACTS
Introduction
In September 1955, the Atomic Energy Commission requested the Earth
Sciences Division of the National Academy of Sciences--National Research
Council to 'organize a meeting of geologists and engineers to discuss the possi-
bilities of permanent disposal of radioactive wastes in geol.ogic formations.
At this meeting , the disposal of high-level wastes in salt was recommended
as having the greatest present potential. Disposal of diluted high-level.
wastes into deep porous formations was' suggested as a possible method for the
future. In subsequent discussions with representatives of the petroleum
industry, it was agreed that it might be possible to dispose of radioactive
wastes by pumping them into formations using hydraulic fracturing techniques.
Research and development, leading up to demonstration experiments using radio-
active materials, have been carried out on the disposal of high-level radio-
active solids in salt formations and the disposal of intermediate-level liquid
wastes by hydraulic fracturing. To date, no field demonstration has been
performed on injection of low-level liquid wastes into deep permeable formations.
..
For publication in the Proceedings of the First International Congress of
the International Radiation Protection Association, Rome, Italy, September 5-10,
1966.
"Research sponsored by the U. S. Atomic Energy Commission under contract with
the Union Carbide Corporation.

T-
2
Disposal in Salt Formations
Salt was sufgested as a possible medium for high-level waste disposal
was
because of its availability, geographic distribution, thermal conductivity,
plasticity, impermeability, and low cost of mining. 12-5) Initially, studies
on the direct disposal of liquid radioactive wastes in salt were performed,
but in 1961 the liquid waste studies were terminated and solid waste studies
were initiated.
The solid wastes to be stored in salt are those produced by
the solidification or calcination of first-cycle wastes from reactor fuel
reprocessing.
In the United States a Waste Solidification Pilot Plant is
currently under construction to demonstrate the potential methods for achieving
the desired solidification. The solids produced are highly radioactive and
generate significant amounts of heat.
There are two ways in which the was e containers can be stored in the mine;
above the mine floor in racks, or below the floor in drilled holes. For storage
above the mine floor, cooling can be obtained : y convection but all handling
operations would have to be performed remotely. Storage in the mine floor allows
access to the rooms for transfer operations, but requires drilling of individual
holes for each waste container. After detailed study, storage in the mine floor
with heat dissipation by conduction thru the salt adjacent to tie containers was
deemed most suitable. A computer program was written to determine the optimum
spacing of waste containers to dissipate the heat in the salt and prevent over-
heating of the waste cans or the salt. 167 The computer results were checked by
field experiments using electrical heaters, and the calculated results agreed
favorably with the experimental observations.
Allowable salt temperature rises are limited by the shattering of the
salt due to increased vapor pressure in small bubbles containing brine ("negative
crystals") located within the salt, and by an increased rate of plastic flow of
the salt. Laboratory and field studies have shown that the problems of salt
MY.PL
shattering and plastic flow can be minimized if the salt temperature is not allowed
to exceed 200°c.
As a result or the theoretical calculations and experimental studies it
appeared that it would be possible to dispose of high-level solid wastes in a
sult mine. In order to determine the equipment and handling operations necessary
in an actual disposal operation and the most economical design of the disposal
facility, a full-scale field demonstration was needed. This demonstration, called
Project Salt Vault, is currently in operation in a salt mine in Lyons, Kansas."
The engineering and scientific objectives of Project Salt Vault are: :
mination of the possible production and release of radiolytically produced chlorine;
(3) determination of the possible gross effects of radiation (up to 10% rad) on
in the renge of 100-200°C; and (4) collection of information on creep and plastic
flow of salt at elevateo. temperatures which car be used later in the design of
: actual disposal facilities.
The Iyons mine was operated for a number of years before it was placed in
an inactive status in 1948. In the existing mine space the marketable salt has
been mined leaving the more impure salt in the floor and ceiling. Since the
impure salt remaining in the mined out areas contains significant moisture bear-
IS
the existing mine floor such that the fuel assembly canisters would be located in
pure salt. Excavation of the new experimental area involved the mining of 19,000
re
tons of salt. The new mine level, which is about 14 feet above the old mine floor,
is connected by a ramp having a 10 per cent grade.
A schematic cross section of Project Salt Vault is shown in Figure 1.
The
waste solids, after canning in Idaho, are loaded into a carrier and shipped on
LEGAL NOTICE
This report was prepared as an account of Governduent sponsored work. Neither the United
States, nor the Commiosion, nor any person acting on behalf of the Commission:
A. Makes any warranty or representation, expressed or implied, with respect to the accu-
racy, completeness, or usefulness of the informatiou contained in this report, or that the use
of any information, apparatus, method, or proceso disclosed in this report may not infringe
privately owned righto; or
B. Assumes any liabilities with respect to the use of, or for damages resulting from the
uso of any information, apparatus, method, or proceso disclosed in this report.
As usod in the above, "person aoting on behalf of the Commission" includes any em-
ployee or contractor of the Commission, or employee of such contractor, to the extent that
auch employee ng contractor of the Commission, or employee of such contractor prepares,
disseminates, or provides access to, any information pursuant to his employment or contract
with the Commission, or his employment with such contractor.

C
O RML-DWG 63-3918 R2
mont
HEAD FRAME-
f
SHIPPING
TRAILER

FUEL ASSEMBLY SHIPPING CASK-
A
FUEL ASSEMBLY CHARGING SHAFT —
-1000 ft I ---
FUEL ASSEMBLY TRANSPORTER
DOUBLE CONTAINMENT
ENTRY 5
14
Ini
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--SHIELD
14ft
-- ENCAPSULATED FUEL ASSEMBLY
110 % GRADE
-HOLE LINERS
1.
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Figure 1
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Schematic Cross-Section of Project Salt Vault
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Schematic Cross Section of Demonstration:
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a specially designed truck trailer to Lyons, Kansas. At Lyons the carrier is
l'emoved from the trailer ind placed vertically over å steel-cased charging shaft
OW
which extends from the surface to the mine level, approximately 1000 feet be .ow.
The waste canisters are lowered, one at a time down the shaft into a shieldea
container mounted on a mobile vehicle. This vehicle (Figure 2) moves from the
Iuel assembly charging shaft to the experimental area where the canisters are
to be lowered into the storage holes.'°)
Since high-level packaged solids were not in production at the time the
demonstration was proposed, irradiated fuel assemblies from the Engineering Test
Reactor (ETR) were used as the heat and radiation sources. In order to achieve
a peak dose to the salt of about 109 rad, it is necessary to replace the fuel
assemblies every six months witli freshly irradiated assemblies. Each shipment
from the Idaho Chemical Processing Plant (ICPP) consists of 14 ETR fuel assemblies
contained in seven stainless steel canisters. The canisters supply the secondary
containment system and are about 5-in. in diameter by 7 1/2 feet long. A depleted
uranium shield plug is loated at the upper end of the canisier and thermocouples
are installed to monitor fuel assembly temperatures during the experiment. When
Project Salt Vault has been completed all canisters will be returned to Idaho
for recovery of the fuel assemblies.
Project Salt Vault is composed of 4 experiments: (i) an array of seven fuel
assembly canisters located in pure salt in the newly mined area; (2) a non-
radioactive control array using electrical heaters to study the effects of
radiation on plastic flow and chlorine release; (3) a heated pillar experiment
for plastic flow and mine stability studies; and (4) an array of seven canisters
in the old mine floor to determine operational problems related to the use of
abandoned mines for radioactive waste disposal. Their location in the mine is
shown in Figure 3.

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The first transfer of fuel assemi.lies to Project Salt Vault occurred on
November 17-19, 1965. Seven canisters were lowered into the mine and placed in
the main radioactive array. 'The fuel assemblies were about 105 days out of reactor
and contained approximately one million curies. The maximum radiation dosage
received by the operating personnel during the transfer operation was 200 mr. '!
Transfer of the seven canisters from the main radioactive array to the
array in the existing mine floor, followed by the insertion of seven new canisters
in the main array, took place June 13-15, 1966. The seven new canisters contained
assemblies that were 105 days out of the reactor. The second set of assemblies
received higher irradiation in the ETR and their heat generation rate was 800 w
initially and they contained about 1 1/2 million curies. As in the case of
the initial loading operation, no handling problems were encountered, and radiation
exposures of operating personne... were minimal.
Temperatures in both the electrical and radioactive arrays are reasonably
close to those predicteå by the theoretical calculations. Figure 4 shows the
temperature rises in the salt 1 1/2 feet from the center line of the center
holes in the arrays along with the calculated temperature rises. It can be
seen that most of the temperatures are somewhat lower than the calculated
values. This is due in part to heat loss from the salt to the mine room. The
peak fuel element temperatures reached about 300°C soon after placement in the
mine.
.
In Fig. 5a is shown the uplift profiles for Room 1 (radioactive array), along
the north-south and east-west axes of the room. It is apparent that thermal
expansion of the material in the floor extends to 40 or 50 ft from the center of
the array.
The north-south uplift profile shows the restraining effect produced
by the presence of the adjacent pillars.
Vertical thermal expansion of the floor in the center of the arrays had
reached nearly an inch by the end of December, 1965. The floor uplift (measured
.
..
Figure 4
-
TV-
.
.-
.-
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.
--
: Comparison of Actual and Theoretical Temperature Rises in Radioactive
and Electrical Arrays
ORNL-DWG 66-7392
-
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CANISTERS IN PLACE-
* PROPERTIES
THEORETICAL FOR 20°C SX
TER OF ARRAY
TEMPEL STURE RISE (°C).
ft FROM CENTER OF
4
-
1. MAIN ARRAY – HOLE 4 tt
HOELECTRICAL ARRAY - HOLE 4.
99 NU-
xin
10°
2
5
10' 2
5 102 2
TIME (hr)
5
10.3 2
5
104

ORNL-DWG 66-648

FLOOR UPLIFT (ft)
& F NORTH-SOUTH PROFILE
EAST-WEST PROFILE
CENTER OF
ral EXPERIMENTAL AREA
O SOUTHERN NORTHERN ENTRY TUNNEL
EDGE OF ROOM
EDGE OF ROOM
15 5 5 15 25 35 45
DISTANCE FROM CENTER OF ARRAY (ft)
25

0.100
0.075
| CENTER OF ARRAY
LN
FLOOR UPLIFT (ft)
0.050
10 ft FROM CENTER
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5.
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O ROOM 4 (RADIOACTIVE ARRAY)
• ROOM 4 (ELECTRICAL ARRAY)
I
0. 10 20 30 40 50 60 70
Figure 5 Floor Uplifts - Project Salt Vault DARS
la) Floor Uplift Profiles in Room 1 as of December 30, 1965.
(6) Floor Uplift in Array Rooms.
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-
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in feet) as a function of time for each array at the center and 10 ft from the
center is shown in Fig. 56. It may be observed that the rate of rise in and
near the array is slowing down. This is due to the fact that the rate of rise
of the salt temperature is also slowing down. The total vertical expansion had
reached about 1 1/4 inch in May, 1966.
Project Salt Vault will continue, with fuel assembly changeouts every six
months until about November 1967. Following completion of the demonstration
essentially all basic data necessary for the design of an actual disposal facility
will have been obtained.
Disposal by Hydraulic Fracturing
•
Hydraulic fracturing has been used for many years in the petroleum industry
to increase production from oil wells. Normally, a mixture of sand and water or
sand and oil is forced out into a hydraulically produced crack in the formation
where the sand is deposited, allowing the oil to drain into the sand layer and
then out into the well. In the case of radioactive waste disposal, however,
the problem is not one of increased oil production but rather one of pumping
the material requiring disposal into the fracture.
The disposal well is possibly the most critical part of the disposal
facility. The well must be drilled into the formation selected for disposal
operation, cased, and cemented to prevent ground water from entering the well.
When an injection is to be performed, this casing is slotted and water is pumped
into the well until the pressure builds up producing a fracture or crack in the
formation (see Figure 6). The waste-cement mixture 1.s then injected into this
fracture. In the case of radioactive waste disposal, solid ingredients (such
as cement, clay, and a retarder are mixed with the waste to produce a slurry
which will harden in the fracture and retain the fission products.
An essential prerequisite in the use of hydraulic fracturing for radioactive

Figure 6
Schematic Cross-Section of Hydraulic Fracturing Operations for
Radioactive Waste Disposal
.
ORNL-ER-DWG 76752)
FLUID WASTE CEMENT MIXTURE
PUMPED IN UNDER PRESSURE
-STEEL CASING CEMENTED INTO ROCK
- SHALE PERM. 0.0002 Md
1
NEW FRACTURE FORMING
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PLUG
OLD FRACTURES FILLED
WITH SOLID -
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MIXTURE
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waste disposal is the need for producing horizontal fractures in the formation.
Considerable controversy exists in the petroleum industry over the conäitions
necessary to produce horizontal rather than vertical fractures. Since experi-
ence in the petroleum industry is mainly with permeable formations and not the
relatively impermeable formations proposed for radioactive waste injections it'
was decided to perform a series of test injections in the shale at Oak Ridge.
The first experimental injection was made at a depth of 290 feet and consisted
of 27,000 gallons of 1 mixture of water, cement, and diatomaceous earth tagged
with 35 curies of +Cs. Subsequent coring and gamma ray logging verified that
the grout sheet had followed the bedding planes and that the fracture was
essential.ly horizontal (See Figure 7). The second experiment Included two
injections at greater depths. The final injection was made at a depth of
934 feet and consisted of 91,500 gallons of water, cement, and bentonite, tagged
with 25 curies of t5cs. Several days later a second injection was made in the
same well at a depth of 100 feet. Core drilling and logging again verified that
the grout sheets followed the bedding planes in the formation (Ses Figure
During the course of these experimental injections, measurements were made on
surface uplift and wellhead and observation well pressures in order to develop
an understanding of the mechanics of fracture I'ormation.
Desirable characteristics of a waste-cement slurry for the disposal of
radioactive waste by hydraulic fracturing are: (1) low viscosity for the period
of time the waste is injected; (2) sorption and retention of the radioactive
liquid after the slurry sets; and (3) a mixture that is relatively cheap.
Studies of the waste-cement slurries for use with ORNL intermediate-level
waste have shown that it is possible to develop mixtures with these properties.
For long pumping times a "retarder" such as calcium signosulfonate (CIS) must
be added; if the cement content is reduced a "suspender" such as attapulgite is
required. For improved retention of radiocesium a clay material such as illite

ORNL-LR-DWG 64048
SOUTH
SOUTH
350f
350 ft
200
200 ft
100
100 ft
100 fi
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200
200 ft
356
350 ft
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Location of Grout Sheet - First Experiment
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pproximate N-S Section Along Line B-B'
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must be added. From the studies completed to date it is apparent that mixes can
be designed having any range of physical properties and cost for any composition
of radioactive waste. (12)
Based on the results obtained in the experimental injections, an experimental
plant has been built at ORNL to dispose of intermediate-level wastes and
evaporator concentrates. A view of the Fracturing Plant is shown in Figure 9.
The four large bins contain the mixture of solids used in the slurry. The
.
solids are transferred pneumatically to the mixing cell in the concrete block.
structure where they are mixed with the liquid waste and the resulting slurry
is pumped into the well by the high pressure pump. shown in the foreground. In
order to provide radiation protection, it is necessary to enclose the waste
pumps, mixer, wellhead, and injection pump in concrete cells.!))
Geologic conditions at the site are shown in Figure 10. Shale formations
exist at several depths that are believed to be suitable for waste disposal
.
injections. Economic considerations suggested the use of the shale located at
a depth of 720 to 950 feet. All of the formations are well below the deepest
known water bearing formation (about 200 feet).
Experimental operation of the plant during the past two years has resulted
in the safe disposal of approximately 430,000 gallons of waste containing 11,500
curies. A total of seven experimental injections ranging in size from 40,000 to
148,000 gallons have been made. During these injections it has been shown that
it is possible to halt the injection, clear the well and equipment, make repairs,
and resume operation without undue hazard to the operating personnel. Core
drilling has shown that the grout sheets from the first five experimental injections
conformed to the bedding planes of the shale.
Cost estimates have shown that the cost of injecting 400,000 gallons per
year of "radioactive waste would be 13€ per gallon including solid ingredients,
depreciation, and operating costs.
The estimated costs are based on one
1
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Geology of Shale Fracturing Site
CRNL-DWG 64-4726
SUBSURFACE GEOLOGY ORNL
FRACTURING PLANT SITE :
-
GRAY SHALE -
CONASAUGA
FORMATION
FEET
700
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1000
1360
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FAULT
LIMESTONE
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.
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3100
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DOLOMITE
KNOX FORMATION
injection of 100,000 gallons per slot and a well life of ?0 years.
.: . The major unknown in hydraulic fracturing is that of well life. As
successive injections are performed and layers of the waste are built up in
the formation, the earth's surface is slowly pushed up and stresses build up
in the overlying formations. Exactly how many injections can be made until
the stresses in the rock will produce failure in the system (vertical
fractures) is currently under study.
At the present time the plant used at ORNL for experimental hydraulic
fracturing is being upgraded to an operating facility for injecting intermediate-
Level wastes and evaporator concentrates on a routine basis. Routine operation
of this facility is scheduled to begin this fall.
Injection into Porous Formations
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--...-.
For a number of years the petroleum industry has used deep well injection
as a means of disposal of waste brines. In the East Texas field alone, 7 x 109
gallons of brine were injected in 1958, and the total volume of brine injected
since operations were initiated in 1935 was 1 x 10+1 gallons. (15) At the
present time, a number of companies are also using deep well disposal methods
for their industrial wastes. (16)(17)
. Before using deep well injection for radioactive wastes there must be an
adequate disposal site and the waste must be compatible with the formation and
its fluid. (10) An ideal site would be one in which a brine saturatei permeable
formation is bounded above and below by impermeable strata. The original concept
was to store the wastes in the interstices of the formation, but the current
philosophy allows injection into flowing formations since the normal flow is
usually very slow, sufficient decay time should be available to reduce the
radionuclide concentrations to safe levels. Radionuclides have been found to
move much slower in the formation than the liquid due to ion-exchange and
adsorption. However, nonhomogeneities in the formation could produce higher
.
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10
velocities in given areas and this may prove to be limiting.
In 1959 the American Association of Petroleum Ceologists (AAPG) was
requested to make a survey of potential sites where deep well injection could be
performed. Their report to describes six basins which apparently would supply
the necessary permeable formations. Three of the basins selected also contain
shale reservoirs and salt deposits which might be used for intermediate- and
high-level waste disposal.
The movement of the liquid and radionuclides in homogeneous isotropic
formations can be calculated. (20) Laboratory studies using a Berea sandstone
block have verified the calculations. (21) Figure Il shows the rate of movement
of Osr from a single injection well. It can be seen that the measured
distribution agrees quite well with the predicted behavior,
Experience in the deep well injection of brines and other wastes has
shown that unless the waste is compatible with the formation and its fluid the
well will eventually plug. Plugging can also be caused by suspended solids or
by bacterial growth. If the, well were to plug by these mechanisms it might be
possible to remove the plugged area by acid treatment, but the resulting waste
containing radionuclides washed from the formation would present an additional
disposal problem. It would appear simpler to pretreat the waste and filter
prior to injection. The type and degree of pretreatment would be dependent on
the characteristics of the waste and the injection formation and would have to
be determined for each specific site.
At the present time there are no firm plans for a field scale demonstration
of deep well injection. This is due, in part, to the large volumes of waste
required to carry out a meaningful field study. It would also require a considera-
ble period of time to complete a field study of any significant size. The results
of a single field sutdy could not necessarily be extrapolated to other sites due
ORNL DWG. 64-4601

OOO 0
-|-144 HOURS
44-72 HOURS
0
0
0
0
0
CONTOURS OF MEAN 86Sr MOVEMENT IN
SANDSTONE BLOCK AS A FUNCTION OF TIME
.; 0 = EXPERIMENTAL POINTS
: THEORETICAL CURVES
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Figure 1l
Figure 11
Contours of Me
of Time
contours of Mean
Movement in Sandstone Block as El Function
11
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to local geologic peculiarities because the imperfections or non-homogeneities
01 the formation would probably control the degree of leakage,
Conclusion
• Field scale demonstrations have shown that it is possible to safely and
economically store high-level solidified wastes in selt formations and internediate-
1 vel wastes in the slightly permeable shales at Oak Ridge. Disposal of low-level
liquid wastes in permeable formation has not yet been demonstrated in field scale
plants but the concept has been proven on laboratory scale models.
Acknowledgment
The authors would like to acknowledge the contributions by tre following
individuals to the work described in this paper: R. L. Bradshaw, W. de Laguna,
F. M. Empson, D. G. Jacobs, H. Kubota, W. C. McClain, W. F. Schaffer, Jr.,
R. C. Sexton, T. Tamura, and H. 0. Weeren.
References
L
1.. Committee on Waste Disposal, Division of Earth Scieżces, Disposal of
Radioactive Waste on Land, National Academy of Sciences, National Research
Council Publication 51), April 1957, P. 6.
2. Mineral Resources of the United States, p. 180. Public Affairs Press (1958).
3. F. Birch and H. Clark, "The Thermal Conductivity of Rocks and its Dependence
on Temperature and Composition," Amer. Jour. of Sciences, Vol. 238, 1940,
pp. 529-558, 613-635.
4. R. L. Bradshaw, F. M. Empson, W. J. Boegly, Jr., H. Kubota, F. L. Parker,
and E. G. Struxness, "Properties of Salt important in Radioactive Waste
Disposal," Proceedings of the International Conference on Saline Deposits,
Houston, Texas, November 12-17, 1962. (In Press).
5. A. E. Inman, Salt, An Industrial Potential for Kansas, p. 22. University of
Kansas Research Foundation (1951).
6. W. J. Boegly, Jr., et al., "Disposal in Salt Formations," Health Physics
Division Annual Progress Report for Period Erding July 31, 1962, pp. 10-20.
7. W. J. Boegly, Jr., et al., "Project Salt Vault: A Demonstration Disposal
of High-level Radioactive Solids in Iyons, Kansas, Salt Mine," Health Physics,
Pergamon Press, 1966. Vol. 12, pp. 417-424.
8. W. F. Schaffer, Jr., et al., "Project Salt Vault: Design and Demonstration
of Equipment," Proceedings of International Symposium on the Solidification
SS
and long-term Storage of Highly Radioactive Wastes, Richland, Washington,
February 14-18, 1966. (In Press).
9. Health Physics Division Annual Progress Report for Perioa Ending July 31, 1966.
(In Press).
10. W. de Laguna, "Disposal of Radioactive Wastes by Hydraulic Fracturing, Part I.
General Concept and First Field Experiments," Nucl. Eng. Design 3 (1966), p. 338.
- mar, compare anorama proporoouse
21. W. de Laguna, "Disposal of Radioactive Wa:tes by Hydraulic Fracturing,
Part II, Mechanics of Fracture Formation and Design of Observation and
Monitoring Wells," Nuc... Eng. Design 3 (1.965) p. 432.
12. T. Tamura, "Disposal of Radioactive was tes by Hydraulic Fracturing, Part IV,
Chemical Development of Waste Cement Mixés," Nucl. Eng. Design 3 (1966). :
13. H. O. Weeren and J. O. Blomeke, "Disposal of Radioactive Wastes by Hydraulic
Fractaring," Nucl. Eng. and Design 4 (1966), pp. 108-117.
14. H.O. Weeren, "Cost Estimates for Hydrofracture Facilities," Letter to
E. G. Struxness dated June 21, 1963.
15. W. S. Morris, "Sub-Surface Disposal of Salt Water from 011 Wells,"
'J. Water Pollution Control Federation, 1960, Vol. 32, pp. 41-51.
16. R. F. Selm and B. T. Hulse, "Deep. Well Disposal of Industrial Wastes,"
Chem. Engrg. Progr., 1960, Vol. 56, No. 5, pp. 138-144.
17. R. S. Stewart, "Proposed Underground Disposal of Industrial Wastes in the
Northeast U. S.," Proc. 12th Industrial Waste Conf., Purdue Univ. Engrg.
Extn. Ser. No. 94, 1957, pp. 494-501.
Moore, T. V. et al., "Problems in the Disposal of Radioactive Wastes in
Deep Wells": Report of Subcommittee on Disposal of Radioactive Waste, Central
Committee on Drilling and Production Practice, Division of Production,
American Petroleur. Institue, Dallas, Tex. (1958).
19. American Association of Petroleum Geologists, Inc. Radioactive Waste-
Disposal Potentials in Selected Geologic Basins, SAN-413-2 (1964)
20. D. G. Jacobs, "Ion Exchange in the Deep-Well Disposal of Radioactive Wastes,"
International Colloquium on Retention and Migration of Radioactive Ions in
Soils, October 16-18, 1962, Center d'Etudes Nucleaires de Saclay, France (1963).
21. D. G. Jacobs and M. U. Shaikh, "Liquid Injection into Deep Permeable
Formations," Health Physics Division Annual Report for Period Ending
July 31, 1964, ORNL-3697, pp. 3-10, (1964).
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