Note: Descriptions are shown in the official language in which they were submitted.
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RADIATION SHIELDING MATERIALS AND
CONTAINERS INCORPORATING SAME
This present invention relates generally to radiation shielding materials,
radiation
shielding containers and methods for preparing the same. More particularly,
the present
invention relates to radiation shielding materials incorporating uranium
dioxide and/or uranium
carbide and containers for radioactive materials incorporating these shielding
materials. This
invention also relates to methods for preparing uranium dioxide and uranium
carbide
microspheres for use in the radiation shielding materials of the present
invention.
Storage, transportation, and disposal of radioactive waste, such as spent
nuclear fuel
("SNF"), high level waste ("HLW"), mixed waste, and low level radiation waste
is a growing
problem in the United States and abroad. In 1995, the Department of Energy
(DOE) estimated
that the commercial SNF inventory was about 30,000 metric tonnes initial heavy
metal
("MTIHM") and is expected to exceed 80,000 MTIHM within two decades. ( 1
tonnes =1 metric
ton = 2,205 pounds). Adding DOE's own inventory of SNF and HLW raises the
domestic total
to nearly 90,000 MTIHM.
Unfortunately, it appears that many U.S. commercial nuclear power plants do
not have
sufficient existing storage capacity to accommodate future SNF discharges.
Moreover, much of
the DOE's SNF and HLW inventory is currently located in unlicensed storage
structures. Many
of these storage structures will have to be upgraded or replaced, and the SNF
and HLW relocated.
Thus, there is a need for improved radiation shielding materials and radiation
shielding
containers incorporating these shielding materials for the storage,
transportation, and disposal of
radioactive materials, including, in particular, SNF waste.
Two principal types of storage methods are generally used for SNF: wet and
dry. In wet
storage, the SNF is typically immersed in a lined, water-filled pool which
performs the dual
functions of shielding and heat removal with the assistance of and reliance on
active systems.
Wet storage of SNF is generally required for a given period of time (about 5
years) after the SNF
has been discharged from a nuclear reactor. Thereafter, the SNF can be placed
into long term dry
storage. Dry storage encompasses a wide spectrum of structures that house the
fuel in a dry inert
gas environment, with an emphasis on passive system design and operation. In
dry storage, the
radioactive material is typically disposed in dry vaults or dry casks. Dry
vault installations
generally utilize a concrete building or other concrete structure for
radiation shielding. Dry cask
storage, on the other hand, utilizes prefabricated containers including an
appropriate shielding
material. Because dry cask storage is usually accomplished more quickly and
cheaply, it is
generally preferred over vault storage. Dry cask storage is also preferred at
sites having an
existing infrastructure for receipt, examination, and loading of SNF for
economic and scheduling
reasons.
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The design and manufacture of a suitable container for the dry storage of SNF
involves
a variety of factors, such as (1) subcriticality assurance, (2) shielding
effectiveness, (3) structural
integrity (i.e., containment), (4) thermal performance, (5) ease of use, (6)
cost, and (7)
environmental impact. Other factors that may affect the selection process are
whether the design
has been previously licensed and actually used to store SNF, or, if the design
has not been
licensed, its perceived ability to meet applicable regulations and standards.
The first factor in designing a storage container is the maintenance of
subcriticality. In
dry storage, the subcriticality design relies on controlling the fissile SNF
and SNF spacing, and
sometimes incorporates the use of neutron-absorbing materials. The
subcriticality control design
of dry storage containers is generally acceptable and does not typically
provide any
discriminating factors for selecting one design over another.
The second factor in designing a storage container is shielding effectiveness.
Shielding
effectiveness affects both onsite worker and public dose rates during the
loading and subsequent
storage of SNF. Both neutron and gamma ray shielding must be provided and
ensured
I 5 throughout the life of the storage system. Dry storage technology relies
on a number of solid
shielding materials, sometimes in combination, to reduce gamma and neutron
dose rates. The
most common solid shielding materials are different forms of concrete (low-
density, high-
density, or hydrogenated), metal (ductile cast iron, carbon steel, stainless
steel, lead), borated
resin, and polyethylene {for neutrons). Often, in order to function
effectively, metal shielding
materials must be combined with additional materials to enhance their neutron
absorbing ability.
The third factor in designing a storage container is structural integrity
(i.e., containment).
Structural integrity ensures that the confinement boundary around the SNF is
maintained under
all operational and postulated accident conditions. All SNF storage
technologies are required to
meet the same standards for structural integrity in accordance with
appropriate codes. Therefore,
the selection of a suitable storage technology will include consideration of
the structural integrity
of the proposed design.
The fourth factor in container design is thermal performance. With the
exception of steel
and cast iron, most shielding materials have inherent limiting temperatures
(i. e., a maximum
allowable temperature that is lower than the fuel cladding temperature limit).
Shielding material
thermal limits include both absolute values of temperature and, in the case of
concrete,
temperature gradients that create thermal stresses. Adequate decay heat
removal is vital to
preventing degradation of the fuel cladding barner to fission product
releases.
Dry storage containers rely on a combination of conduction, convection
(natural or
forced), and radiation heat transfer mechanisms to maintain fuel cladding
temperatures below
appropriate long term storage limits. In particular, metal casks rely on a
totally passive system
for heat removal. The fuel decay heat, in an encapsulating inert gas
atmosphere canister, is
transferred to the canister's walls by a combination of radiation and
conduction heat transfer.
The canister walls, which are in contact with the metal cask wall, transfer
this heat by
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conduction. At the outside of the metal cask, the heat is removed by
conduction and natural
convention to the environment. Metal cask typically are not susceptible to
thermal limits, since
the metals have a higher temperature limit than that of the fuel cladding.
However, in those
embodiments where the metal casks incorporate additional neutron shielding
materials their
S favorable heat-transfer properties may be compromised.
As with metal casks, concrete casks use a passive heat removal system.
Concrete casks,
however, have an inherent vulnerability, because concrete's thermal
conductivity is a factor of
to 40 lower than that of metal. Thus, in order to remove fuel decay heat and
stay below both
the fuel cladding and concrete temperature limits, concrete casks must include
labyrinthine
10 airflow passages that allow natural convection-driven air to enter the
cavity enclosing the canister
inside the concrete and then exit through higher elevation passages in the
concrete to the
environment. The need for these airflow passages introduces the possibility of
an accident in
which adequate heat removal is reduced or eliminated because of inlets and/or
outlets that are
blocked by debris, snow, or even nests and hives. As a result, concrete casks
require surveillance
of their air inlet and outlet flow passages, thereby increasing the associated
life-cycle costs and
personnel radiation exposures.
The fifth factor in designing a storage container is ease of use, which is
defined as the
lack of complexity involved in the operation and maintenance of SNF. As noted
above, the
existence of labrynthine air passages in concrete casks means that additional
operation and
maintenance is required. Ease of use, however, is alsa related to the
complexity associated with
loading, transport, and storage of SNF. Thus, the weight and size of
containers are also of
particuiar importance. For example, since many existing storage sites are
already equipped with
a crane in the storage and receiving facility, it is desirable to utilize
containers with weights that
are within the typical crane capacity of 45 to 91 tonnes. Metal casks
generally cannot be used
with such cranes, because the weight of a fully-shielded metal cask loaded
with a large number
of SNF elements can easily exceed the 91 tonnes limit. Thus, even though metal
casks have
desirable heat transfer characteristics, the additional weight and size
associated with metal
systems limits their applicability.
Additional size and weight limits are imposed when containers are transported.
The U.S.
Department of Transportation and state highway regulations generally limit the
gross weight of
a waste-carrying road vehicle to about 80,000 pounds. Since the typical
tractor trailer weighs
about 30,000 pounds, the weight of a transportation container and its contents
should not exceed
about 50,000 pounds. Heavier weights can be transported by rail, but maximum
container widths
(diameters) are limited to approximately 9 feet to allow for adequate
clearance between tracks.
U.S. Nuclear Regulatory Commission regulations require that the container
provide certain levels
of shielding and be capable of sustaining certain impact stresses without
yielding the waste. The
end result of these regulations is that much of the available weight for the
transportation
container and its contents must be expended in providing adequate shielding
and a shell that can
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withstand the designated impact stresses. The resulting thickness of the
container walls leaves
a relatively small amount of space in the container for SNF.
The sixth factor in designing a storage container is cost. Concrete casks are
generally the
least expensive, with a typical cost of about $350,000 to $550,000, versus $1
million to $1.5
million for their metal cask counterparts.
The seventh factor in designing a storage container is environmental impact.
Over time,
environmental mechanisms can degrade storage containers, possibly exposing the
SNF directly
to groundwater or air. Storage containers and shielding materials that
minimize degradation are
preferred for long term storage and disposal.
In summary, metal casks are desirable because they are known to provide
effective heat
transfer and structural integrity. Unfortunately, metal casks are heavier and
more expensive than
concrete casks. Furthermore, in most SNF applications, metal casks must
incorporate separate
neutron shields, which may compromise their favorable heat transfer
properties.
Thus, there is a significant need for improved, lower weight and higher heat-
transfer
shielding materials and, also, for containers for handling, storage, and
disposal of radioactive
waste that are superior in performance, size and cost, while providing
acceptable structural
strength, shielding effectiveness, and carrying capacity.
In light of the shortcomings associated with existing dry storage containers
and the need
for long term management of existing inventories of SNF, the DOE began to
examine alternative
means for the transportation, storage and disposal of such waste. As a result
of its investigation,
the DOE recommended that the transport and emplacement of commercial spent
fuel into a DOE
waste repository be accomplished using a class of containers known as the
Multi-Purpose Cask
(MPC) and Multi-Purpose Unit (MPU). MPC/MPU containers are intended to perform
the three
functions of storage, transport, and disposal by direct emplacement into a
waste repository. The
MPC is a thin-shelled container, without shielding, which, once filled, is not
intended to be
opened. Proposed MPC/MPU designs use metal canisters requiring massive
fabrication
techniques. As a result, the estimated costs are three to six times greater
than that of concrete
cask designs. Furthermore, the MPC containers hold approximately 12% less SNF
than that of
concrete storage casks. Finally, since the MPC casks do not include shielding,
these casks must
be outfitted with overpacks consisting of thick-walled steel and, typically, a
separate, neutron-
absorbing material to provide shielding.
Meanwhile, the DOE was investigating management options and alternative uses
for large
quantities of depleted uranium hexafluoride ("DUF6") stored at gas diffusion
plants. Among the
various disposal options considered by the DOE was conversion of the uranium
hexafluoride to
3 5 uranium metal, which could be machined for use as a radiation shielding
material. However, the
high costs of uranium metal production (around $10/kg), combined with the
handling, machining,
and environmental costs associated with the use of uranium metal have
historically limited its
use to only a few small applications. In connection with the design of the MPC
and MPU, for
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example, the DOE proposed that depleted uranium metal be used as an axial
shield plug in the
MPC and as a gamma shielding material for the MPU during transport.
Other applications of depleted uranium metal in the fabrication of storage
containers
includes a container made from a composite containing a fibrous mat of
interwoven metallic
a 5 fibers encased within a concrete-based mixture that can include depleted
uranium metal. Another
proposed application includes a depleted uranium metal core for absorbing
gamma rays and a
bismuth coating for preventing chemical corrosion and absorbing gamma rays.
Alternatively,
a sheet of gadolinium may be positioned between the uranium metal core and the
bismuth coating
for absorbing neutrons. The containers can be formed by casting bismuth around
a pre-formed
uranium metal container having a gadolinium sheeting, and allowing the bismuth
to cool.
Still another proposed application incorporates a depleted uranium metal wire
wound on
the inner shell of a cask to create a radiation shield. And yet another
proposed application
utilizes a composite radiation shield made up of rods of depleted uranium
metal. The spaces
between the rods contain smaller rods and are backfilled with lead or other
high-density material.
Still other designs utilize pipes of depleted uranium metal, tungsten, or
other dense metal,
encapsulating polyethylene cores, dispersed in rows of concentric bore holes
around the periphery
of the cask body. None of these existing designs, however, provides a simple,
low-cost, low-
weight radiation shielding system for transportation, storage, and disposal of
radioactive waste.
Uranium compounds have also been proposed for use as shielding materials. For
example, some investigators have proposed that depleted uranium dioxide (DUOZ)
pellets be
mixed with a cement binder to form a material known as DUCRETE, which could be
used as a
shielding material in dry storage containers. The DUOZ pellets replace the
gravel aggregate
normally used in concrete. Due to the increased density of DUO2, however, the
thickness of the
shielding layer can be reduced. Thus, a storage container made from DUCRETE
will have a
greatly reduced weight and diameter compared to conventional concrete casks.
In a typical cask,
for example, the outer shell thickness can be reduced from approximately 2.5
feet for concrete
to approximately one foot with DUCRETE. As a result, the cask diameter is
reduced by
approximately two-thirds, and the weight is reduced from approximately 123
tonnes to
approximately 91 tonnes.
Despite these improvements in size and weight, however, DUCRETE casks systems
suffer from disadvantages similar to those experienced with concrete casks. In
particular, since
DUCRETE has a low thermal conductivity and low temperature limit, DUCRETE
casks must
~ also incorporate labrynthine ventilation gaps. Furthermore, it is not
expected that DUCRETE
will be able to retain the uranium dioxide pellets in its cement matrix for a
long period of time
due to its high porosity of concrete and to the likelihood of water-cement-
uranium dioxide
reactions at warm temperatures (90-300°C). DUCRETE may also be
incompatible with expected
repository requirements. Hence, the use of DUCRETE in significant quantities
for SNF disposal
is questionable.
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Nuclear fuel manufacturing plants produce small particles of uranium dioxide
and
uranium carbide by powdered metallurgical processes. These processes generally
involve
production of a powder of the proper particle size and range, which is then
pressed into pellets,
sintered, and ground to size. Even though powdered processes have shown
success, their
capacity is limited due to mechanical complexity, particle size, reactivity,
and mass transfer
limitations. In practice, line capacities are limited to approximately 100
tonnes/year, and
maximum plant sizes to around 1,000 tonnes/year.
It has been proposed that aqueous processes be used to generate uranium
dioxide and
uranium carbide. Work on aqueous processes, and in particular on aqueous
gelation processes,
began in the late 1960's. By the mid-1970's pilot-scale facilities for
production of uranium oxide
and uranium carbide had been constructed. Experimental and pilot plant studies
focused
primarily on the use of uranyl nitrate solutions. For gelation, these uranyl
nitrate solutions were
dispersed using single nozzles into columns of chlorinated solvents such as
trichloroethylene
(TCE) and perchloroethylene. The resulting microspheres were then processed
using multiple
washing operations with water and ammonium hydroxide. The resulting
microspheres, typically
0.03 mm to 2 mm in diameter, were incorporated into cylindrical pellets.
Unfortunately, these
aqueous processes had small throughputs and the processing was manually
intensive. Thus, for
planned capacities greater than 100 tonnes/yr, these processes were generally
inadequate.
It is anticipated that the demand for shielding materials in accordance with
the present
invention will require the production of 5,000 to 30,000 tonnes/year of
uranium dioxide and/or
uranium carbide. Thus, there is a need for improved process capable of
producing greater than
100 tonnes/year, and preferably 5,000-30,000 tonnes/year, of uranium dioxide
and uranium
carbide in reasonably-sized plants with inexpensive equipment. There is a
further need for a
process for producing microspheres of uranium dioxide and uranium carbide over
a wide size
range (30-1,200 microns). There is also a need for an improved gelation
process for production
of uranium dioxide and uranium carbide directly from uranium hexafluoride.
Finally, there is
a need for an improved gelation process that avoids the necessity of
converting uranium
hexafluoride to uranyl nitrate in order accomplish gelation. The present
invention addresses
these and other needs.
SUMMARY OF THE INVENTION
Briefly, and in general terms, the present invention resides in an improved
radiation
shielding material and storage systems for radioactive materials incorporating
the same. The
shielding material is preferably formed from a PYRolytic Uranium Compound
("PYRUC") and
provides improved radiation shielding in comparison with other shielding
materials. In
accordance with the invention, the shielding material can be used to form
containment systems,
container vessels, shielding structures, and containment storage areas, all of
which can be used
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to house radioactive waste. The preferred embodiment of the shielding system
is in the form of
a container for storage, transportation, and disposal of radioactive waste.
The precursor for the PYRUC shielding material is preferably a mixture of a
uranium
compound and a binding material. In the preferred embodiment, the uranium
compound is
depleted uranium dioxide (DUOZ) or depleted uranium carbide (DUC or DUCZ). The
uranium
compound is preferably in the form of small particles, and more preferably in
the form of pellets
or microspheres, which can be coated or uncoated. The present invention
incorporates a number
of improvements over prior art methods for producing uranium dioxide and
uranium carbide
microspheres, whereby 5,000-30,000 tonnes/year of these microspheres can be
produced in
reasonably-sized plants and with inexpensive equipment. The improved gelation
process of the
present invention permits the use of oil in the gel forming column, deliberate
carryover of oils
to the sintering steps for supplying carbon and hydrogen, use of nitrogen as
the sintering carrier
gas, and use of peroxide for gelation of both uranium oxides and carbides.
In some cases, the precursor material can simply be cured to form a radiation
shielding
material. However, in preferred embodiments, the particles are immersed in a
matrix of a
binding material, so that the binding material fills the interstitial spaces
and also provides
additional neutron shielding. In accordance with the present invention, the
binder is
advantageously comprised of ( I ) a carbonaceous material (such as pitch); (2)
a high-temperature
resin (such as a polyimide); (3) a metal (such as aluminum powder); and/or (4)
a metal-oxide
(such as alumina). In addition, materials such as hydrogen, boron, gadolinium,
hafnium, erbium,
and/or indium in their non-radioactive isotopes, can be added in the mixture
in the appropriate
chemical form (usually the oxide) to provide additional neutron shielding
effectiveness. The
shielding materials are formed by applying sufficient heat to the mixture to
cause a pyrolytic
reaction that forms a solid material.
The present invention also resides in an method for manufacturing storage
containers
utilizing PYRUC shielding materials. In accordance with the invention, the
precursor mixture
can be poured or extruded into the container and then pyrolyized to foam a
solid shield. In a
particularly preferred embodiment, the precursor starting materials are poured
or extruded into
a space formed by the inner and outer wall of a container and then pyrolized.
The manufacturing
process provides maximum flexibility in designing shielding shapes. The walls
of the container
provide the shape, structural support, and missile and drop protection, and
also function as the
secondary confinement barrier for the depleted uranium. The use of PYRUC
simplifies shield
manufacture and avoids the massive metal forging and machining activities
associated with metal
casks.
PYRUC shielding materials in accordance with the present invention offer
superior
gamma and neutron radiation shielding with the desirable thermal properties of
metal at a much
lower thickness, weight, and life-cycle cost than conventional materials.
Furthermore, the
PYRUC shielding materials can be optimized for specific circumstances and
source terms. The
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use of depleted uranium reduces the assay (enrichment) level of the overall
package, which
provides for criticality mitigation. Furthermore, since PYRUC shielding
materials have high
thermal conductivities, the need for labyrinthine air passages and daily
inspections is avoided.
Similarly, PYRUC materials have higher thermal conductivities and temperature
limits than
concrete or DUCRETE and, thus, do not Iimit the design. In particular, the
thermal
conductivities of PYRUC materials exceed DUCRETE values by 25-100%. The
temperature
limits of carbonaceous PYRUC materials exceed 1000°C and PYRUC
materials using other
binders have temperature limits above 300°C. Moreover, the high thermal
conductivity and the
high material temperature limit of PYRUC eliminate the need for a separate,
inner canister for
containing SNF. As a result, the PYRUC shielding materials can be used in SNF
containers with
direct contact between the shield's inner annulus and the basket containing
the SNF, which
further reduces size and weight.
It is believed that PYRUC-shielded SNF containers will cost about $600,000 to
$700,000
each, with the PYRUC component accounting for about $200,000 of the cost. The
PYRUC
container, although having an initial capital cost slightly greater than the
concrete cask, is
expected to be significantly less expensive than the metal cask while having
similar advantages.
Lower life-cycle costs are also expected for the PYRUC container as compared
with either
concrete or DUCRETE containers, since PYRUC's superior heat transfer
properties will preclude
the need for frequent inspection and subsequent maintenance activities. Thus,
PYRUC
containers should be cost-competitive with traditional containers.
PYRUC is also environmentally desirable because it utilizes a waste product
from the
nuclear industry (depleted uranium) and, in one form, a waste product from the
petrochemical
industry (carbonaceous binder material) and converts them to environmentally
stable forms. The
PYRUC shielding material is also environmentally desirable because it is both
microencapsulated
and macroencapsulated, and has enhanced leach resistance. As a result, the
material is potentially
stable for geologic time periods. Thus, by virtue of its composition and
expected behavior in a
disposal environment, PYRUC is an environmentally friendly material.
Thus, the present invention satisfies the need for a shielding material having
combined
shielding performance, high temperature resistance, high thermal conductivity,
and
environmentally desirable characteristics, and for smaller, lighter containers
for storage,
transportation, and disposal of radioactive materials. While the primary
applications for PYRUC
are containers for SNF and HLW storage, transport, and disposal, PYRUC
shielding materials
can also be utilized in radiopharmaceutical containers, ion exchange resins,
reactor cavity
shielding and activated materials (i. e., made radioactive by neutron
absorption) among others.
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The present invention will be more clearly understood from a reading of the
following
detailed description in conjunction with the accompanying figures and tables.
S FIG. 1 is a cross-sectional view of a container for storage, transport, and
disposal of
radioactive material which includes a PYRUC shielding material in accordance
with the present
invention;
FIG. 2 is a cross-sectional view of the container shown in FIG. 1 along the
line 2-2 in
accordance with the present invention;
FIG. 3 is a flow diagram setting forth the overall process for manufacture of
a container
incorporating PYRUC shielding materials in accordance with the present
invention;
FIG. 4.1 is an overview in block form of the gelation process for producing
uranium
dioxide microspheres in accordance with the present invention;
FIG. 4.1a is an overview in block form of the gelation process for producing
uranium
carbide microspheres in accordance with the present invention;
FIG. 4.2 is an overall process flow diagram and material and energy balances
for the
production of uranium dioxide microspheres in accordance with the present
invention;
FIG. 4.2a is an overall process flow diagram and material and energy balances
for the
production of uranium carbide microspheres in accordance with the present
invention;
FIG. 4.3 is a process flow diagram for a depleted uranium hexafluoride
receiving and
volatilization station in accordance with the present invention;
FIG. 4.4 is a process flow diagram for a the UOZFZ production station in
accordance with
the present invention;
FIG. 4.5 is a process flow diagram for an uranyl nitrate formation station in
accordance
with the present invention;
FIG. 4.6.1 is a process flow diagram for a carbon suspension formation station
utilized
in connection with the production of uranium carbide microspheres in
accordance with the
present invention;
FIG. 4.6.2 is a process flow diagram for an uranyl nitrate solution adjustment
station for
manufacture of uranium dioxide in accordance with the present invention;
FIG. 4.6.2a is process flow diagram for an uranyl nitrate solution adjustment
station for
production of uranium carbide in accordance of the present invention;
FIG. 4.7 is a process flow diagram for a gel solution preparation station in
accordance
with the present invention;
FIG. 4.8 is a process flow diagram for a gel formation station for production
of 1,200
micron spheres in accordance with the present invention;
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FIG. 4.9 is a process flow diagram for a gel formation station for 300 micron
spheres in
accordance with the present invention;
FIG. 4.10 is a process flow diagram for an oil purification system in
accordance with the
present invention;
5 FIG. 4.11 is a process flow diagram for a 1,200 micron sphere
setting/washing station in
accordance with the present invention;
FIG. 4.12 is a process flow diagram for a 300 micron sphere setting/washing
station in
accordance with the present invention;
FIG. 4.13 is a process flow diagram for a 1,200 micron sphere drying station
in
10 accordance with the present invention;
FIG. 4.14 is a process flow diagram for a 300 micron sphere drying station in
accordance
with the present invention;
FIG. 4.15 is a process flow diagram for a 1,200 micron sphere conversion and
sintering
station in accordance with the present invention;
FIG. 4.16 is a process flow diagram for a 300 micron sphere conversion and
sintering
station in accordance with the present invention;
FIG. 4.17 is a process flow diagram for a calcium nitrate reconstitution
station in
accordance with the present invention;
FIG. 4.18 is a process flow diagram for a ammonium hydroxide solution
purification
station in accordance with the present invention;
FiG. 4.19 is a process flow diagram for a vertical tube furnace gas
purification station in
accordance with the present invention;
FIG. 4.20 is a process flow diagram for a ammonium hydroxide reconstitution
station in
accordance with the present invention;
FIG. 4.21 is a process flow diagram for an urea and HMTA recovery station in
accordance with the present invention;
FIG. 4.22 is a process flow diagram for a cylinder decontamination station in
accordance
with the present invention;
FIG. 4.23 is a process flow diagram for a waste management station in
accordance with
the present invention;
FIG. 4.24 is a process flow diagram for an uranium carbide sintering station
in accordance
with the present invention;
FIG. 4.25 is a process flow diagram for an uranium carbide coating station in
accordance
with the present invention; and
3 S FIG. 5 is a process flow diagram for a graphite route for production of
the uranium
carbide microspheres in accordance with the present invention; and
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FIG. 6 is a process flow diagram for a peroxide gelation process in accordance
with the
present invention.
Table 1 sets forth the material properties and estimated costs for various
shielding
materials;
~ Table 2 sets forth the shielding properties for various shielding materials;
Table 4.1 sets forth the list of assumptions for the exemplary gelation
processes of the
present invention;
Table 4.2 sets forth the overall material and energy balances for production
of uranium
dioxide microspheres in accordance with the present invention;
Table 4.2a sets forth the overall material and energy balances for the
production of dense
uranium carbide microspheres in accordance with the present invention;
Table 4.3 sets forth the material and energy balances for the depleted uranium
hexafluoride receiving and volatilization station for production of uranium
dioxide microspheres
in accordance with the present invention;
Table 4.3a sets forth the material and energy balances for the depleted
uranium
hexafluoride receiving and volatilization station for production of uranium
carbide microspheres
in accordance with the present invention;
Table 4.4 sets forth the material and energy balances for the uranyl fluoride
production
station for production of uranium dioxide microspheres in accordance with the
present invention;
Table 4.4a sets forth the material and energy balances for the uranyl fluoride
production
station for production of uranium carbide microspheres in accordance with the
present invention;
Table 4.5 sets forth the material and energy balances for the uranyl nitrate
formation
station for production of uranium dioxide microspheres in accordance with the
present invention;
Table 4.5a sets forth the material and energy balances for the uranyl nitrate
formation
station for production of uranium carbide microspheres in accordance with the
present invention;
Table 4.6.1 sets forth the material and energy balances for the carbon
suspension
formation station for production of uranium dioxide microspheres in accordance
with the present
invention;
Table 4.6.2 sets forth the material and energy balances for the uranyl nitrate
solution
adjustment station for production of uranium dioxide microspheres in
accordance with the
present invention;
Table 4.6.2a sets forth the process flow diagram for a uranyl nitrate solution
adjustment
station for production of uranium carbide microspheres in accordance with the
present invention;
Table 4.7 sets forth the material and energy balances for the gel solution
preparation
station for production of uranium dioxide microspheres in accordance with the
present invention;
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Table 4.7a sets forth the material and energy balances for the gel solution
preparation
station for production of uranium carbide microspheres in accordance with the
present invention;
Table 4.8 sets forth the material and energy balances for the gel formation
station for the
production of 1,200 micron spheres in accordance with the present invention;
Table 4.8a sets forth the material and energy balances for the gel formation
station for
production of 1,200 micron spheres of uranium carbide in accordance with the
present invention;
Table 4.9 sets forth the material and energy balances for the gel formation
station for the R
production of 300 micron spheres in accordance with the present invention;
Table 4.9a sets forth the material and energy balances for the gel formation
station for
production of uranium carbide microspheres in accordance with the present
invention;
Table 4.10 sets forth the material and energy balances for the oil
purification station for
production of uranium dioxide microspheres in accordance with the present
invention;
Table 4.10a sets forth the material and energy balances for the oil
purification station for
production of uranium carbide microspheres in accordance with the present
invention;
Table 4.11 sets forth the material and energy balances for the 1,200 micron
sphere
setting/washing station for production of uranium dioxide microspheres in
accordance with the
present invention;
Table 4.lla sets forth the material and energy balances for the 1,200 micron
sphere
setting/washing station for production of uranium carbide microspheres in
accordance with the
present invention;
Table 4.12 sets forth the material and energy balances for the 300 micron
sphere
setting/washing station for production of uranium dioxide microspheres in
accordance with the
present invention;
Table 4.12a sets forth the material and energy balances for the 300 micron
sphere
setting/washing station for production of uranium carbide microspheres in
accordance with the
presentinvention;
Table 4.13 sets forth the material and energy balances for the 1,200 micron
sphere drying
station for production of uranium dioxide microspheres in accordance with the
present invention;
Table 4.13a sets forth the material and energy balances for the 1,200 micron
sphere
drying station for production of uranium carbide microspheres in accordance
with the present
invention;
Table 4.14 sets forth the material and energy balances for the 300 micron
sphere drying
station for production of uranium dioxide microspheres in accordance with the
present invention;
Table 4.14a sets forth the material and energy balances for the 300 micron
sphere drying
station for production of uranium carbide microspheres in accordance with the
present invention;
Table 4.15 sets forth the material and energy balances for the 1,200 micron
conversion
and sintering station for production of uranium dioxide microspheres in
accordance with the
present invention;
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13
Table 4.15a sets forth the material and energy balances for the 1,200 micron
conversion
and sintering station for production of uranium carbide microspheres in
accordance with the
present invention;
Table 4.16 sets forth the material and energy balances for the 300 micron
conversion and
sintering station for production of uranium dioxide microspheres in accordance
with the present
invention;
Table 4.16a sets forth the material and energy balances for the 300 micron
conversion and
sintering station for production of uranium carbide microspheres in accordance
with the present
invention;
Table 4.17 sets forth the material and energy balances for the calcium nitrate
reconstitution station for production of uranium dioxide microspheres in
accordance with the
present invention;
Table 4.17a sets forth the material and energy balances for the calcium
nitrate
reconstitution station for production of uranium carbide microspheres in
accordance with the
present invention;
Table 4.18 sets forth the material and energy balances for the Ammonium
hydroxide
solution purification station for production of uranium dioxide microspheres
in accordance with
the present invention;
Table 4.18a sets forth the material and energy balances for the ammonium
hydroxide
solution purification station for production of uranium carbide microspheres
in accordance with
the present invention;
Table 4.19 sets forth the material and energy balances for the vertical tube
furnace gas
purification station for production of uranium dioxide microspheres in
accordance with the
present invention;
Table 4.19a sets forth the material and energy balances for the vertical tube
furnace gas
purification station for production of uranium carbide microspheres in
accordance with the
present invention;
Table 4.20 sets forth the material and energy balances for the ammonium
hydroxide
reconstitution station for production of uranium dioxide microspheres in
accordance with the
present invention;
Table 4.20a sets forth the material and energy balances for the ammonium
hydroxide
reconstitution station for production of uranium carbide microspheres in
accordance with the
' present invention;
Table 4.21 sets forth the material and energy balances for the urea and hmta
recovery
station for production of uranium dioxide microspheres in accordance with the
present invention;
Table 4.21a sets forth the material and energy balances for the urea and HMTA
recovery
station for production of uranium carbide microspheres in accordance with the
present invention;
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14
Table 4.22 sets forth the material and energy balances for the cylinder
decontamination
station for production of uranium dioxide microspheres in accordance with the
present invention;
Table 4.22a sets forth the material and energy balances for the cylinder
decontamination
station for production of uranium carbide microspheres in accordance with the
present invention;
Table 4.23 sets forth the material and energy balances for the waste
management station
for production of uranium dioxide microspheres in accordance with the present
invention;
Table 4.23a sets forth the material and energy balances for the waste
management station
for production of uranium carbide microspheres in accordance with the present
invention;
Table 4.24 sets forth the material properties and energy balances for the
uranium carbide
and sintering station for production of uranium carbide microspheres in
accordance with the
present invention; and
Table 4.25 sets forth the material and energy balances for the uranium carbide
coating
station for production of uranium carbide microspheres in accordance with the
present invention.
DETAILED DESCRIPTION OF THE PREFERRED EMBODIMENTS
With reference now to the exemplary drawings, and particularly to FIGS. 1-2,
there is
shown, in cross-section, a container 10 in accordance with the present
invention. The container
includes a lid 12, a base 14 and a body 16 defining a central cavity 18. The
container 10 is used
to store waste material, including, in particular, radioactive waste
materials, such as SNF. In this
regard, a plurality of pressurized water reactor ("PWR") assemblies housing
waste material are
fitted inside of a basket assembly 20 disposed within the container 10, as
best seen in FIG. 2.
The container 10 can have a variety of geometries. In the embodiment shown in
FIGS. 1 and 2,
the container is cylindrical, having a circular cross-section. Alternatively,
the container could
have a cross-section that can be square or hexagonal, among other geometries,
in order to
facilitate various packing and storing configurations.
The body 16 includes an inner wall 22a and an outer wall 24a thereby defining
cavity
26a. A PYRUC shielding material 28a is disposed within the cavity 26a. The
shielding material
advantageously absorbs neutrons from neutron-emitting waste materials and
gamma rays from
gamma-emitting waste materials. As described in detail, below, during
manufacture of the
container, the PYRUC precursor material is prepared and poured or extruded
into the cavity
between the inner wall and outer wall of the body and then pyrolized to form a
solid radiation
shield. Alternatively, the solid radiation shield may be formed by several
sequential castings,
forming successive axial and radial rings, thereby allowing the shield to be
tailored to a variety
of requirements. For example, it may be desirable to utilize two radial layers
of different
PYRUC shielding materials, such as a more dense inner layer which will absorb
neutrons more
effectively in combination with a less dense outer layer that will absorb
gamma rays.
The inner wall 22a and outer wall 24a are formed from forged steel from about
0.10 to
about 3.00 inches thick, preferably from about 0.5 to about 1.0 inches thick.
The preferred
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embodiment shown in FIGS.1 and 2, is an MPU designed to hold twenty-four PWR
assemblies.
In this particular embodiment, the body 16 is 160 inches in height, the
diameter of the central
cavity 18 formed by the inner wall is 65.8 inches, the outer diameter of the
outer wall is 81.8
inches, and the inner wall 22a and outer wall 24a of the body 16 define an
eight-inch cavity 26a.
' S It will be understood, however, that the thickness of the inner and outer
walls 22a and 24a and
size of the cavities 18 can vary according to the strength and shielding
requirements of the
container 10 and the size of the waste to be contained. Forged steel is
desirable because it is
economical, easy to manufacture, and a reasonably good conductor of heat.
Alternatively, other
materials such as carbon steel, stainless steel, titanium, aluminum, or the
like can be used. While
10 stainless steel would be generally more expensive, it provides the
additional advantage of
corrosion resistance.
The lid 12 and base 14 are attached to body 16 and each includes an inner wall
22b and
22c and an outer wall 24b and 24c which define a cavity 26b and 26c,
respectively. In this
particular embodiment, both cavities are about thirteen inches high and
incorporate a PYRUC
15 shielding material 28b and 28c. The lid and base are constructed from the
same materials as are
used to construct the body.
The container 10 or any of its components, body 12, base 14 and lid 16, can be
manufactured with an inner wail 22 and outer walls 24 that are coated.
Coatings can be used, by
way of example, to decrease permeability or to enhance radioactivity absorbing
characteristics
of the container or for corrosion resistance. Typical permeability coatings
include glass coatings,
epoxy coatings, and inorganic coatings (such as those containing silica),
galvanizing materials
(zinc) and zirconia, among others. The coating thickness is typically from
about 1.0 to 2,000
microns. As best seen in FIG. 1, a liner 30 is located adjacent to the inner
wall 22a. This liner
can be a one inch perforated support plate constructed from materials such as
steel, Lead, and
25 the like.
Turning now to the details of the basket 20, as shown in FIG. 2, the basket 20
is
dimensioned to hold multiple PWR assemblies. The central cavity 18 is equipped
with a means
(not shown), such as a locking pin, which secures the basket in an upright,
centralized position.
The basket is a removable compartmentalized structure, preferably made of
metal, which is
30 designed to hold assemblies of the radioactive material in a segregated
manner. In a preferred
embodiment, a number of baskets having different configurations are
interchangeable so that
both large (24 or 21 PWR) and small (12 PWR) assemblies can be accommodated.
It is
also desirable to equip the container 10 with a lifting trunnion 34 attached
to the body 16. This
lifting trunnion advantageously facilitates handling of the container 10.
3 5 In use, the base 14 is attached to the container 10 and the container is
filled with SNF by
wet or dry methods. After loading, the lid 12 is seal welded to the body 16 of
the container.
Alternately, bolt closures with flexitallic, elastomeric, or metallic o-
ring/groove sealing (not
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16
shown) can be used to seal the lid. If the container was loaded under water,
the water is removed
via a drain valve (not shown) and the container dried with warm nitrogen gas
by circulation
through a top vent (not shown). Subsequently, nitrogen or helium is
introduced, and the vent and
drain are welded to the container 10.
In some embodiments, suitable granular material is added to fill the spaces 36
between
the basket and the inner wall 22 of the container 10, thereby improving heat
transfer and
shielding. For storage applications, this granular material includes carbon
spheres and sand,
particularly colemonite sand, which includes boron and bound water. For MPC
and related
applications, uranium oxide and uranium carbide could be added, although
adjustments may be
necessary to account for varying crane weight limits at particular storage or
disposal sites.
Referring now to FIG. 3, an overview of the process for preparation of PYRUC
shielding
materials is shown. In accordance with the preferred embodiment of the
invention, depleted
uranium hexafluoride is converted by an improved gelation process, discussed
below, into
microspheres of a pyrolytic uranium compound, most preferably, into uranium
dioxide, uranium
monocarbide, and/or uranium dicarbide microspheres (collectively "uranium
carbide or "UC").
In some embodiments, at least two sizes of microspheres are utilized to
promote higher spatial
densities. Also, in some embodiments, the particles are coated with materials
such as carbon,
silica, pitch, metal, or the like. During the gelation process, other uranium-
containing materials,
such as uranium metal and U30g can be incorporated and processed to produce
microspheres.
Binding materials are sized and classified to match the size of the
microspheres. Two
sizes of binding material can be used to maximize the density and minimize
pore volume of the
shielding material. The microspheres and binding material are then mixed and
homogenized to
form a precursor mixture. The precursor mixture is poured or extruded into the
cavity 26 defined
by the inner wall 22 and outer wall 24 of the container 10. Heat treatment and
pressure are
advantageously used to pyrolize the microspheres and form a solid shielding
material.
Inspections and sealing complete the assembly of the container 10.
The precursor mixture contains from about 5 to 100% of a particulate pyrolytic
uranium
compound. Preferred mixtures contain uranium dioxide and/or uranium carbide.
microspheres.
The size of the particles can all be the same size (uniform), can be
distributed over a range of
sizes (distributed), or can be classified into several discrete size ranges
(classified). Preferred
particle sizes range from 0.030 mm to 2.0 mm. Smaller particles can be used,
but are generally
too fine for easy handling and create environmental concerns. Larger particles
can also be used,
but require long times for densification, as by sintering, and do not pack as
well.
The preferred particle shape is spherical, but particles can be any suitable
shape, including
3 5 cylindrical, rectangular, and/or irregular. The preferred embodiment uses
spherical particles of
two discrete size ranges: 300 to 500 microns and 1,000 to 1,300 microns in
diameter, including,
in particular, a mixture of 300 micron and 1,200 micron spheres. It is
believed that these
particles provide_a suitable combination of packing, handling, environmental
and densification
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17
requirements. In a particularly preferred embodiment, the precursor mixture
contains 80%
pyrolytic uranium microspheres. Various binders or additives make up the
remaining portion of
the material. The microspheres, in turn, are preferably comprised of 70%
uranium monocarbide
coated with pyrolytic carbon, as a 1,000 to 1,300 micron diameter particle,
and 30% uranium
~ 5 dioxide coated with pyrolytic carbon, as a 300 to 500 micron particle.
As noted above, in preferred embodiments, the binding materials are added to
fill the
interstitial spaces, provide additional shielding, and enhance the overall
performance of the
shielding material. The binding materials generally constitute up to 95% of
the precursor
mixture. Typically, a binding materials is selected based upon an assessment
of the radiation
I 0 spectrum of the material requiring shielding.
The main categories of precursor mixtures in accordance with the present
invention are
classified by the binding material utilized in their production: ( 1 )
carbonaceous binders; (2) resin
binders; (3) metal binders; and (4) metal oxide binders. Suitable carbonaceous
binders.are
formed by the low temperature pyrolysis (heating) of pitch, tar, polyvinyl
alcohol and related
I 5 compounds, graphite, coke byproduct or the like. The preferred form of
carbonaceous binder is
pitch, because it mixes well with the pyrolytic uranium compound and forms a
continuous
structure upon pyrolysis. The carbonaceous binders are preferably pyrolized to
the empirical
formula C,Ha2, with C,Ho.s most preferred. An advantage of this combination is
that it forms an
environmentally inert shielding material. When pyrolytic uranium dioxide is
mixed with a
20 carbonaceous binder, it is preferred that the uranium dioxide first be
coated with, for example,
pyrolytic carbon, for better carbon-uranium dioxide adhesion.
Resin binders are polymers and include mixtures of polymers, such as
polyethylene,
polypropylene, polyurethane, polyimides, and polyamides. Resin binders provide
the advantage
of excellent neutron shielding, albeit with some heat transfer penalties. The
resin binder can be
25 a thermoplastic resin, such as polyethylene, polypropylene, or
polyurethane, that can be melted
and extruded as a paste or viscous liquid. Advantageously, however, resin
binders are
comprised of non-thermoplastic resin binders, delineated herein as thermoset
resins, which do
not melt readily, but which bond when the precursor mixture is heated andlor
pressed. Examples
of such resins include polytetrafluoroethylene (sold under the tradename
TEFLON), polyamides,
30 polyimides, teflon analogues, FEP (fluorinated ethylene-propylene, which is
a copolymer of
tetrafluoroethylene and hexafluoropropylene), polyvinylidene fluoride (sold
under the tradename
KYNAR), and a copolymer of chlorotrifluoroethylene and ethylene (sold under
the tradename
HALAR), and PFA (perfluoralkoxy), among others. Polyamides include materials
such as nylon-
6 and nylon-6,6. Polyimides, on the other hand, have a phthalimide structure
and are typically
3 5 formed from dianhydrides and diamines containing aryl groups. Polyimides
generally have high
strength, stability, and thermal resistance, in some cases greater than
500°C. Typical polyimides
include the reaction products of benzophenone tetracarboxylic dianhydride
(BTDA) and 4,4'-
diaminodiphenyl ether (DAPE) (sold under the tradenames KAPTON, TORAY, PYRO-
ML, and
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18
PYALIN), a BDTA/m-phenylenediamine (MPD) derivative (sold under the tradenames
MELDIN
and SKYBOND), and trimellitic anhydride (TMA)/MPD (sold under the tradenames
KERIMID,
KERMEL, and ISOMID). In addition, it is believed that both thermoplastic and
thermosetting
polyfunctional resins will be advantageously utilized in accordance with the
present invention.
Polyfunctional resins contain at least two chemical functional groups in each
repeating polymer
unit. In addition to polyurethane and the polyimides and polyamides identified
above, other
suitable polyfunctional resins include acetonitrile butadiene styrene (ABS),
polyphylene sulfide
(PPS), polysulfones, polyesters (including dacron-type polyesters), phenolic
plastics, and
fiberglass reinforced plastic combinations. The preferred resin is both
thermosetting and
polyfunctional. In the preferred embodiment, the resin binder is a 100%
polyimide resin.
Suitable metal binding materials include copper, zinc, nickel, tin, aluminum,
aluminiumlboron mixtures and the like. Preferred metal binders contain
aluminum powder.
Most preferred is an aluminum/boron mixture, because it exhibits both high
heat transfer and
neutron shielding effectiveness.
Metal oxide binders include both ceramic and refractory materials. Suitable
metal oxides
include alumina, magnesia, silica, hafnia, hematite, magnetite, silica, and
zirconia, among others.
Alumina is the generally preferred metal-oxide binder. A castable alumina
material, with 6%
boric/and acid added, is the most preferred, because of its neutron shielding
effectiveness and
adhesion to uranium dioxide.
While any one or any combination of the binding materials can be used, the use
of one
binding material will be preferred for simplicity and greater mechanical
robustness. By way of
example, high heat load waste is advantageously shielded using a shielding
material containing
a binder having high heat transfer properties, such as a metal binder. In
contrast, mixed uranium
plutonium oxide waste is advantageously shielded by a shielding material
containing a binder
optimized for neutron shielding.
The composition of the precursor mixture varies with the category of binder
material used
and application. While the precursor can contain up to 100% of the uranium
material (essentially
close packing of the microspheres or pellets), optimum shielding weight is
achieved with 55-80%
pyrolytic uranium compound and 45-20% binder; based on the weight of the
precursor mixture.
The precursor mixture also advantageously includes additives, comprising
typically up
to 20% of the binding material, for enhanced shielding, heat transfer, or
stability. Typical
additives include hydrogen, boron, gadolinium, hafnium, erbium, indium and the
like. These
additives are included in the appropriate chemical forms. For example, an
alumina binder can
be combined with boric acid and/or gadolinium oxide. A particularly preferred
additive is boron-
10, which can be added as granular boric acid and converted to B203 when the
precursor mixture
is pyrolized. Alternatively, sodium borate can be utilized. In addition, for
gadolinium, halfnium,
erbium and indium, the oxide form is generally preferred. Mechanical additives
such as steel
shot or glass beads may also be added to the PYRUC mixture. Alternatively,
additives such as
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19
gadolinium, hafnium, erbium, and indium can be added to the gel-forming step
of the gelation
process, so that they reside within the spheres of uranium dioxide/carbide as
their respective
oxides.
Once the components of the precursor mixture have been selected, they are
combined,
and then homogenized. Mixing is advantageously accomplished by either batch or
continuous
methods, such as twin-screw auger extruders, and slight heating may be
applied.
The homogenized mixture is placed within the cavity 26a formed by the inner
wall 22a
and outer wall 24a of the body 16 by extrusion/pumping (preferred for viscous
binder
combinations) and/or vibratory methods (preferred for powder blends). Slight
heat and pressure
may be applied. After filling, sufficient heat (100-1000°C) and
pressure (0-20 atmosphere) are
applied to the container, to pyrolize and form a solid shielding material. An
end closure is
attached to the body 16 by suitable means, such as tungsten inert gas welding
in order to seal the
body 16. Thereafter, the container 10 is brushed and polished. Gamma
radiography and other
non-destructive examination (NDE) methods are used check the body 16 prior to
use. The lid
12 and base 14 can be similarly manufactured.
In those embodiments where a combination of carbonaceous binding materials are
employed, the pyrolitic uranium component, carbon powder, additives, and pitch
are mixed in
an extruder. The extruder then deposits a first annular layer of the precursor
mixture into the
cavity 26a, Next the layer is pyrolized in an inert atmosphere of nitrogen,
argon or similar gases
to form the solid shielding material. Pyrolysis typically requires from about
0.1 to about 24 hours
at temperatures of from about 300-800°C. Thereafter, additional annular
layers precursor
material are extruded into the and pyrolized under similar conditions. Each
layer is from 1 to 4
meters thick. Thus, the shield, typically, consists of several annular layers,
each individually
pyrolized and bonded together.
Alternatively, the inner wall 22a can be removed for better heat transfer and
heat- treated
in one step. If carbon powder is used by itself as the binder material, then
the mixture is dry-fed
into the cavity 18a. Heat is applied as before and the material is pressed,
thus forming the shield.
In the preferred embodiment, heat is supplied by electrical resistance
inductance or radiance.
Heat may also be supplied by direct or indirect fired equipment.
For resin-based PYRUC materials, powdered resins are dry blended by mechanical
and
vibratory means with the uranium form and loaded by vibratory means into the
cavity 18a.
Electric heating is preferably used to heat the material to 400-600°C,
typically for 0.1 to 24
- hours, to form the PYRUC monolith. If thermal resins are used, they are
mixed in an extruder
under heat. Thereafter, the mixture is extruded into the container 10 as a
viscous fluid. Heat and
pressure are then applied to form the solid monolith in a manner similar to
the carbon forms.
For metal-based PYRUC materials, the container 10 is heated electrically or by
a fired
furnace under an inert cover gas to the melting point of the metal binder. For
the typical metals
cited, this temperature will fall between 400 and 1,000°C, preferably
below the melting point of
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the container's materials of construction. Thereafter, an initial amount of
molten metal binder
is added to the container 10 to form a layer 1 cm to 4 m thick, followed by an
initial quantity of
preheated uranium material. Due to its density, the uranium material will sink
through and to the
bottom of the molten metal layer, forming a packed bed of the particles with
the metal filling the
5 interstitial points. The process is repeated until the cavity 18a is filled.
Thereafter, the heat
source is removed, and the shield cools and solidifies. Alternatively, where a
powdered metal,
such as a copper or nickel powder, is used, the metal powder, uranium form,
and any additives
are dry blended by mechanical and vibratory means and vibratorily loaded into
the cavity 18.
Heating is used to melt the powder, causing the matrix to congeal and fuse
together into a
10 monolith. It is particularly preferred to heat the material by induction,
utilizing induction coils.
As before, typical temperatures of 400-1,000°C and times of 0.1 to 24
hours are required.
For metal-oxide PYRUC materials, the metal oxide and uranium form can be
combined
with water (0-40 wt %), mixed, and then pumped into the cavity 18a. It is
believed that water
hydrates the metal oxide binder and, therefore, assists in bonding of the
material. The material
15 is preferably allowed to harden for 2 to 96 hours and then heat treated for
0.1 to 24 hours at
temperatures up to 400°C. In the preferred embodiment, the shielding
material is formed in
sequential layers in order to facilitate heat and mass transfer.
Alternatively, the inner wall of the
body 16 can be removed and replaced with a temporary, combustible wall (e.g.,
manufactured
from wood products) for casting as before. This allows the number of casting
steps to be reduced
20 significantly, in some cases, allowing a single step. The thermal step
burns away the combustible
inner wall.
The choice, mix, and awangement of the shielding materials used in the PYRUC
mixture
will vary with the type and quantity of radioactive material being transported
or stored. Thus,
the thickness, diameter, number and arrangement of the shielding materials
will be varied to
provide optimum protection against the neutrons and gamma radiation emitted by
the particular
type and quantity of radioactive material.
The use of uranium dioxide and uranium carbide advantageously facilitates
simpler and
less expensive manufacturing routes for both the uranium material and the
shielding cask. It
essentially involves the direct casting of the PYRUC material from a mixer or
an extruder into
the cavity formed between the two metal walls of the cask. Only low
temperatures are involved,
and the casting and machining of uranium metal are eliminated. Finally,
uranium dioxide and
coated uranium carbide have good heat transfer and thermal characteristics.
Thus, their use
eliminates the need for the labyrinthine air cooling passages present in
concrete-shielded storage
containers, thereby reducing monitoring requirements and costs.
Table 1 presents a comparison of material properties and estimated costs for
various
shielding materials, including PYRUC. As summarized in Table 1, PYRUC is a
shielding
material that provides superior thermal conductivity and temperature limits at
a competitive cost,
while offering superior neutron and gamma shielding. Similarly, Table 2
presents a comparison
CA 02284942 1999-09-24
_ WO 98142793 PCT/US98/05493
21
of the properties of PYRUC with other shielding materials. As shown in Table
2, casks
incorporating PYRUC typically offer thermal performance and gamma shielding
capabilities
approaching that of metal. Meanwhile, PYRUC materials provide low temperature
ease of
fabrication, and chemically non-reactive forms which are not susceptible to
combustion or
~ 5 chemical interaction above ground or in an SNF repository.
Other potential applications for PYRUC include radiopharmaceutical containers,
ion
exchange resins, reactor cavity shielding, and activated materials. PYRUC may
also have utility
in other applications as a shielding material for utility resin shields,
reactor cavities, naval
reactors, spacecraft, and Greater Than Class C (GTCC) materials.
As discussed above, in the preferred embodiment, it is desirable to utilize
substantially
spherical uranium dioxide or uranium carbide particles of generally less than
1,300 microns.
There are several types of processes known in the art that can be used to
produce such particles:
( 1 ) Powder Metallurgy Processes (Granulation processes); (2) Melting
Processes (Arc glazing,
Plasma burner glazing, Suspension melting, Glazing of hydrate salts); and (3)
Fluid Processes
(Synthetic resin condensation, Emulsion processes, and Gelation processes). In
accordance with
the present invention, the small particles of uranium dioxide and uranium
carbide can be
generated by any suitable means.
In the past, however, only power metallurgy processes have provided the basis
for
commercial production and only for commercial production of uranium dioxide
particles.
Processes for the production of uranium dioxide and uranium carbide are also
described in
Controlled Nuclear Chain Reaction: '~J~e First 50 Years, American Nuclear
Society, 1992, La
Grange Park, Illinois, and M. Benedict, T. Pigford, and H. Levi, Nuclear
(,',hemical En~ineerine,
Second Edition, McGraw-Hill, New York, NY, 1981, incorporated herein by
reference.
Such processes are mechanically intensive. They typically start with a low
density
uranium dioxide powder, produced in a rotary kiln from depleted hexafluoride,
followed by
mixing, granulation, pressing a pellet, sintering, and pellet-grinding to
produce dense uranium
dioxide particles. Furthermore, powder process-based plants are generally
modular, small
throughput operations. Thus, scale up to the requirement for uranium materials
contemplated
for use in accordance with the present invention would necessitate hundreds of
process lines.
Other nonfluid methods for the manufacture of uranium dioxide were
investigated in the
late 1970s and early 1980s, which avoid the mechanically intensive, powder
processes. These
alternative processes are described in S.M. Tiegs, et al., "The Sphere-Cal
Process: Fabrication
~ of Fuel Pellets from Gel Microspheres," ORNL/TM-6906, September, 1979;
"Fuels Recycle and
Development," (FRAD) Program Review, Battelle Northwest, September 13-15,
1978, and J.M.
Pope, "Spherical UC Fuel Via Gel-Precipitation," American Nuclear Society,
Annual Meeting,
Miami, June 7-11, 1981, incorporated herein by reference. Additional
information is available
in "Fuels Refabrication and Development (FRAD) Program Review," Battelle
Pacific Northwest
Laboratories, September 13-1 S, 1978; "NPR-MHTGR Fuel Development Program,"
Idaho
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National Engineering Laboratory (INEL), EGG-NPR-8971, June 1990; and R.H.
Perry and C.H.
Chilton, Chemical Engineers' Handbook, Fifth Edition, New York, NY, 1973, also
incorporated
herein by reference.
Furthermore, granulation processes are suitable for dense uranium carbide
particles, but
cannot produce dense particles of uranium oxide. Melting processes have the
drawback of being
expensive and yielding an excessively-large range of particle sizes.
Accordingly, for economic and capacity reasons, it is preferably most
desirable to
generate the uranium particles using gelation processes. An overview of the
conversion of
depleted uranium hexafluoride into spheric, dense uranium dioxide particles by
gelation is
presented in A.P. Murray, S. Mirsky, P. Hogroian, and S. Krill, "Gelation
Conversion Of
Depleted Uranium Hexafluoride Into Dense Uranium Dioxide Microspheres," Third
International
Uranium Hexafluoride Conference Proceedings, November 28-December 1, 1995,
Paducah,
Kentucky, incorporated herein by reference. These processes include variously
as sol-gel,
gel-precipitation, internal gelation, external gelation, particle fuel,
microsphere, and solution
precipitation processes. In gelation processes, hydrodynamics is used to form
spheres of
ammonium diuranate ("ADU"), which are subsequently cured, dried, and sintered
into dense
uranium dioxide microspheres typically ranging from 30 to 1,500 microns in
size. Furthermore,
for a specific size, a narrow-size distribution can be obtained.
Gelation processes are based on the fact that if a colloidal solution ("sol"
or "broth") of
a uranium dioxide precursor (e.g., uranyl nitrate) is dispersed into a fluid
with which it is
immiscible, or only slightly miscible, spherical droplets are formed which
solidify by gelling
(hence, the "gel"). The critical part of the processes occurs when the
colloidal solution is
dispersed in the fluid. In order to promote gelling, while maintaining droplet
integrity, it is
necessary to remove the positive charge on the droplets for greater
immiscibility and precipitation
potential. This can be advantageously accomplished by either (a) extraction of
water; (b)
extraction of acid; or (c) addition of alkali.
Gelation methods are generally classified as either external or internal
gelation routes.
In external gelation routes, microspheres of uranium dioxide or uranium
carbide are produced
by introducing droplets of a uranyl nitrate solution into a column containing
ammonia gas. As
the droplets fall through the gas, surface tension effects cause them to form
spheres of ADU.
Due to size effects upon mass transfer, external gelation generally requires
careful design for
production of spheres larger than about 800 microns. In contrast, internal
gelation uses aqueous
phase immiscibility in an organic liquid as the basis for sphere formation.
The gel formers in -
internal gelation are typically organic oils or solvents containing ammonia-
releasing compounds
such as amines that release ammonia (e.g. hexamethylenetetramine ("HMTA")).
Due to better
mass transfer, internal gelation can typically produce larger, more uniform
microspheres.
Furthermore, since there is better heat transfer between the gel former
solution and the droplet,
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23
shorter columns with longer residence times can be used. Thus, in the present
invention, internal
gelation is preferred.
MANUFACTURE OF DENSE URANIUM DIOXIDE AND EXEMPLARY PROCESS
FIG. 4.1 provides an overview of the preferred gelation process for producing
uranium
dioxide (UOZ) microspheres according to the present invention from depleted
uranium
hexafluoride ("DUF6").
Depleted uranium hexafluoride gas is reacted with steam ("H20") to produce
solid uranyl
fluoride ("UOZFz") and gaseous hydrogen fluoride ("HF"). The hydrogen fluoride
gas is
recovered in the anhydrous form, and the uranyl fluoride solid is collected,
quenched, and
dissolved in water. Thereafter, any residual hydrogen fluoride in the uranyl
fluoride solution can
optionally be removed by distillation. However, as discussed in detail below,
the resulting uranyl
fluoride solution can be used directly in the gelation process for the
production of uranium
dioxide microspheres. The presence of residual hydrogen fluoride does not
significantly affect
the gelation steps, and any residual fluoride hydrogen can be removed from the
final uranium
dioxide product in the subsequent steps of aging and washing. Alternatively,
the uranyl fluoride
so produced can be further reacted and converted to uranyl nitrate which in
turn is used to make
uranyl dioxide. As shown in FIG. 4 conversion of uranyl fluoride is
accomplished by adding
calcium nitrate ("Ca(N03)3") to the aqueous uranyl fluoride solution, thereby
precipitating
calcium fluoride ("CaF2") and forming aqueous uranyl nitrate ("U02N03)Z").
Prior to gelation,
the resulting uranyl nitrate solution is adjusted by evaporation, urea is
added, and the solution
chilled.
In the preferred internal gelation routes in accordance with the present
invention,
vibrating nozzles are used to disperse the uranyl nitrate solution into
droplets which are then
introduced into a vertical column of an immiscible oil, gel-forming solution.
The size of the
nozzles and the vibration frequency determine the droplet sizes, and, thus,
the microsphere sizes.
As the uranyl nitrate droplets fall vertically within the gel-forming
solution, heat transfer between
the gel-forming solution and the droplets causes the uranyl nitrate to form
ADU from the
ammonia produced by the decomposition of hexanethylenesracmine ("HMTA").
Preferably, the
gel-forming solution will flow in the opposite direction of the droplets to
slow the descent of the
droplet and permit additional time for the uranyl nitrate solution to form
microspheres with
sufficient strength to avoid sphere deformation at the bottom of the column.
Column heating to
50-100°C advantageously increases the formation of ADU. The ADU gel
spheres are collected
at the bottom of the column. Typically, the gel spheres are fragile and
require careful handling
to avoid breakage.
The gel spheres are then aged in an ammonium hydroxide solution. After aging,
the
"green" gel spheres are dried at low temperatures to remove water and excess
ammonia.
Subsequently, a vertical tube furnace converts and sinters the microspheres
under an inert
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gas-hydrogen atmosphere. While argon and helium are acceptable inert gases, it
is preferable to
utilize nitrogen due to its low cost and availability. Furthermore, since
nitrogen is slightly
reactive, it will advantageously form uranium nitrides in concentrations up to
10,000 ppm, and
more typically in the range of 300 to 1,000 ppm, which will also function as a
radiation shielding
material.
The final, sintered spheres have individual densities usually exceeding 95% of
the
theoretical density for uranium dioxide. Coarse microspheres (e.g., about
1,000 microns in
diameter) might provide spatial densities of 65-70% of theoretical, and the
addition of a finer
microsphere (e.g., about 300 microns in diameter) might provide spatial
densities of 80-85% of
theoretical. As a result, two or three size fractions are typically preferred
in order to achieve
spatial densities approaching 90%. For example, a 60 wt. % fraction of 1,000
micron spheres,
a 20 wt. % fraction of 300 micron spheres, and a 20 wt. % fraction of 30
micron spheres can be
used to achieve spatial densities in the 90-95% range.
Each of the following steps are now addressed in greater detail:
A. Uranium Hexafluoride Receiving And Volatilization;
B. Uranyl Fluoride Production By Reaction Of Uranium Hexafluoride And Steam;
C. Uranyl Fluoride Collection, Quenching And Addition Of Water To Form An
Aqueous Uranyl Fluoride Solution;
D. Uranyl Fluoride Solution Distillation To Adjust Residual Hydrogen Fluoride
Concentration;
E. Uranyl Nitrate Solution Formation By Addition Of Calcium Nitrate And
Precipitation Of Calcium Fluoride;
F. Uranyl Nitrate Solution Adjustment By The Addition Of Urea And Increase In
Acidity;
G. Gel Solution Preparation And Addition To Uranyl Nitrate Solution;
H. Gel Sphere Formation By Internal Gelation Techniques;
I. Oil Purification;
J. Gel Sphere Aging By Setting/Washing With Ammonium Hydroxide;
K. Gel Sphere Drying And Liberation Of Ammonia And Water;
L. Gel Sphere Conversion And Sintering;
M. Gel Sphere Collection;
The following additional steps are advantageously undertaken in connection
with the
overall process design:
N. Calcium Nitrate Reconstitution;
O. Ammonium Hydroxide Solution Purification;
P. Vertical Tube Furnace Gas Purification;
Q. Ammonium Hydroxide Reconstitution;
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R. Urea And HMTA Recovery;
S. Cylinder Decontamination; and
T. Waste Management.
A. Uranium Hexafluoride Receiving and Volatilization
5 FIG. 4.3 depicts the flow sheet for the Depleted Uranium Hexafluoride
Receiving and
Volatilization Station. Depleted uranium hexafluoride is obtained from
enrichment plants in
standard, 14 tonne cylinders; a typical 14 tonne cylinder will contain about
12.5 tonnes of solid
uranium hexafluoride. The extra space provides room for expansion, when the
solid uranium
hexafluoride is heated. These uranium hexafluoride cylinders 1 are received by
truck and rail in
10 the Cylinder Shipping and Receiving Station and stored until needed. It is
desirable to utilize a
storage building having capacity for enough cylinders for one month's
operation (i.e.
approximately 300 cylinders) and additional capacity for storage of an
equivalent number of
empty cylinders 4 while they await shipment to the Cylinder Disposal Facility.
Prior to use, the
uranium hexafluoride cylinders are transferred to the Full Cylinder Temporary
Storage Station.
1 S A feed of uranium hexafluoride gas is obtained by heating the uranium
hexafluoride
cylinders in an autoclave. Heating causes the solid uranium hexafluoride to
sublime, so that the
pressurized uranium hexafluoride vapor above the phase can be extracted.
Heating of the
cylinders is achieved by heating air within the autoclave with steam 5 from a
Steam Plant. The
oven is heated to a temperature sufficient to cause sublimation but, below the
liquefaction
20 temperature for uranium hexafluoride (about 150°F). The heating rate
is preferably selected to
maintain the uranium hexafluoride under subatmospheric pressure. In a
preferred embodiment,
the oven is heated to about 140°F in one hour. At this temperature, the
uranium hexafluoride
temperature will be lower, e.g., about 125°F, and the corresponding
cylinder pressure will be
below atmospheric pressure, e.g. 10 psia. At these pressures, stresses on the
cylinders are
25 avoided.
The pressure of the uranium hexafluoride gas 4 is then slightly increased (S -
10 psig) to
near atmospheric conditions using a feed compressor. This increase in pressure
also causes the
temperature of the uranium hexafluoride gas to increase to about 212°F.
The uranium
hexafluoride feed 4 is then directed sent to the UOZF, Production Facility for
further processing.
In the exemplary process set forth in FIG. 4.3, uranium hexafluoride feed
rates of about 28,000
' tonnes/year, i.e. 3,200 kg/hr, are employed. At these feed rates,
approximately 88 tonnes of
uranium hexafluoride can be processed on a daily basis, requiring
approximately 7 cylinders of
- - depleted uranium hexafluoride per day. A typical gas diffusion plant has a
gaseous feed station
consisting of three ovens: (1) a first oven heating a full cylinder of uranium
hexafluoride
(requiring approximately two hours); (2) a second oven supplying gaseous
uranium hexafluoride
to the enrichment operations at typical feed rates of 1,500-2,000 kg/hr
(requiring approximately
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four to six hours); and (3) an third oven cooling down (requiring
approximately 1-2 hours).
Thus, a typical gas diffusion plant can process about four cylinders per day.
Nevertheless, the
preferred feed rates of 3,200 kg/hr can be achieved by utilizing five ovens,
sequenced as follows:
( 1 ) Oven 1: Heating, two hours; (2) Oven 2: Feeding, first two hours; (3)
Oven 3: Feeding,
second two hours; (4) Oven 4: Feeding, third two hours; (5) Oven 5: Cooling, 1-
2 hours (i.e., hot
cylinders have to be cooled prior to moving). It would also be desirable to
have an additional
oven as a spare for use during maintenance or in the event of a breakdown.
Adding two
additional ovens, one as a spare for the heating phase and one as a spare for
the feeding operation
would also assist in assuring an uninterrupted supply of uranium hexafluoride
at the design flow
rate.
The approach set forth above represents "hot feeding" of the cylinders, with
pressures
exceeding 25 psig. Where concerns exist regarding the pressure rating of the
uranium
hexafluoride cylinders, cylinder pressurization can be avoided by "cold
feeding" the uranium
hexafluoride at temperatures below 147°F using a withdrawal compressor
with sublimation from
the solid uranium hexafluoride. Since cold feeding is generally restricted to
lower feedrates,
typically in the 360-450 kg/hr (800 -1,000 lb/hr) range per cylinder, it would
take approximately
3 5 hours to empty a cylinder containing 12.5 tonnes of uranium hexafluoride.
(The feedrate
typically drops to 180-200 kg/hr about 400 lb/hr when a withdrawal compressor
is not used.)
Thus, in order to obtain the desired overall feedrate of 3200 kg/hr at 360
kg/hr per oven, the
gelation plant would require simultaneous feeding of about 9 cylinders. The
following oven
sequence would be advantageous in this situation: (a) Oven l: Heating, two
hours maximum; (b)
Oven 2-10: Feeding, probably on a 3-4 hour sequence/changeout schedule; (c)
Oven 11: Cooling,
one-two hours; (d) Oven 12: Spare, for heating/feeding; and (e) Oven 13:
Spare, for cooling.
Once again, it would be desirable to have an additional oven as a spare for
use during
maintenance or in the event of a breakdown.
In order to minimize or avoid cylinder pressurization, the flow sheets and
mass balances
set forth in this exemplary process utilize the 100% baseline case for cold-
feeding of uranium
hexafluoride using thirteen autoclaves and one spare.
After the uranium hexafluoride is discharged, a "heeling" compressor (not
shown) is used
to reduce the empty cylinder pressure to less than 1 psia and the residual
uranium hexafluoride
"heel" to 4.5 kg ( 10 lb). Heel compressor requirements shown in Table 4.3 are
estimated as 10%
of the main compressor. Empty cylinders 4 are sent to Empty Cylinder Temporary
Storage
before shipping to Cylinder Shipping and Receiving and, eventually, forwarded
to the Cylinder
Disposal Facility.
Condensate 6 from the Autoclave Facility is fed to the Condensate Return and
recycled
to the steam plant. Facility waste 7, such as personnel protective clothing
and equipment, is sent
to the waste treatment station.
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B. Uranyl Fluoride Production By Reaction Of Uranium Hexafluoride and Steam
FIG. 4.4 depicts the flow sheet for the UOzFz Production Station, including,
'inter ~i , the
reaction of depleted uranium hexafluoride gas with steam to produce UOZF2. The
material and
energy balances for the exemplary process are shown in Table 4.4.
S Uranium hexafluoride is reacted with steam according to the following
equation:
UF6 + 2 H20 -- UOZFZ + 4 HF~B~h> + 256 KW~t~
In the preferred embodiment, uranium hexafluoride gas 1 from the DUF6
Receiving and
Volatilization Station and steam 2 from the steam plant 25 are combined in a
reactor vessel, such
as a pyrolysis reactor or kiln. In the exemplary process shown in FIG. 4.4,
the uranium
hexafluoride gas 1 and steam 2 are introduced concurrently into a pyrolysis
reactor at about
200-300°C. This reaction is exothermic and proceeds spontaneously. At
temperatures over
about 150°C, no excess steam is required and this reaction produces
essentially anhydrous
hydrogen fluoride. Thus, stoichiometric efficiencies are assumed in the flow
sheet shown in FIG.
4.4. In order to prevent runaway temperatures and provide steam for subsequent
use in the Steam
Plant, the heat of reaction is advantageously used to generate steam 4 from
deionized water
stream 3.
The gas 5 exiting the reactor vessel consists primarily of anhydrous hydrogen
fluoride gas
with traces of entrained uranyl fluoride powder. The uranyl fluoride powder is
removed
downstream of the reactor vessel using cyclone separators and filters. The
hydrogen fluoride gas
24 can be subsequently condensed and collected in an HF Storage Facility. The
stored hydrogen
fluoride 9 can be dispensed as a saleable product.
C. Uranyl Fluoride Collection, Quenching and Formation Of An Aqueous Uranyl
Fluoride Solution
FIG. 4.4 depicts the flow sheet for the U02F2 Production Station, including,
j~ ~, the
collection of UOZFZ and the subsequent quenching and formation of an aqueous
uranyl fluoride
solution. The material and energy balances for the exemplary process are shown
in Table 4.4.
The uranyl fluoride powder 10 formed in the pyrolysis reactor is collected and
removed
by a screw auger device (not shown). Since the uranyl fluoride powder formed
in the reactor
vessel is discharged at elevated temperatures, the powder 10 is preferably
quenched in a water
spray and dissolved in a water solution made up of deionized water 23 and wash
water 28 from
the Cylinder Decontamination Station. The hydration reaction is represented by
the following
equation:
UOzF2 ~,~ + 6 H20 -- U02Fz ~ 6 H20 + 5 kCah,~
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The heat of hydration is removed by the circulation of cooling water 17
through coils (not shown)
disposed in the quenching system. The resulting uranyl fluoride solution 13 is
transmitted to the
Distillation Station for further processing.
D. Uranyl Fluoride Solution Distillation To Adjust Residual Hydrogen Fluoride
Concentration
FIG. 4.4 depicts the flow sheet for the UOZFZ Production Station, including,
inter ,
distillation of the uranyl fluoride solution to adjust the residual hydrogen
fluoride concentration.
The material and energy balances for the exemplary process are shown in Table
4.4.
A small fraction, approximately 5%, of the anhydrous hydrogen fluoride
produced in the
pyrolysis reactor may be entrained with the uranium fluoride powder and, thus,
become hydrated
in the Quench Reactor. It is expected that the hydrated hydrogen fluoride in
the quenched uranyl
fluoride solution 13 can be removed by distillation in the Distillation
Station. Heat for the
distillation is provided by steam 26 provided from the steam plant 25. The
distilled product 14
can be returned to the pyrolysis reactor as the azeotrope, HF~2H20. The sizing
of the distillation
column shown in exemplary process in Table 4.6.2 is based upon 5% carryover of
hydrogen
fluoride. Facility waste 6 is sent to Waste Management.
The distilled uranyl fluoride solution 15 is shipped to the UOZFZ Storage
facility where
it is available for shipping to the UOZFZ Precipitation Station.
E. Uranyl Nitrate Solution Formation By Addition Of Calcium Nitrate and
Precipitation Of Calcium Fluoride
FIG. 4.5 depicts the flow sheet for the Uranyl Nitrate Formation Station. The
material
and energy balances for the exemplary process are shown in Table 4.5.
In the past, gelation methods for producing dense uranium dioxide utilized
nitrate
solutions generated from uranium oxides as chemical substitutes for nitrate
solutions from
reprocessing nuclear fuels. Nevertheless, it is believed that uranium fluoride
solutions can be
used directly for gel formation without subsequent processing. The direct use
of uranyl fluoride
solutions uses concentrations between 0.1 and 40%, and preferably 1 S-25% in
uranyl fluoride.
However, since nitrate solutions have previously found favorable application
in the past, the
exemplary process shown in FIG. 4.5 includes the additional steps for forming
the nitrate
solution. The material and energy balances relating to this conversion are
contained in Table 4.5.
A calcium nitrate solution 14 from the Calcium Nitrate Reconstitution Station
and, as
needed, fresh calcium nitrate powder 3 are combined with deionized water 4
obtained from the '
Deionized Water Supply 2 in a mixing vessel. The resulting calcium nitrate
solution 7 and uranyl
fluoride solution 1 from the UOZFZ Storage facility are introduced into a
precipitator vessel.
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Tramp and washwater 6 from the Slurry Washer/Dryer, described below, may also
be introduced
into the Precipitator. The aqueous calcium nitrate reacts with the uranyi
fluoride solution to form
a slurry 8 of uranyl nitrate and calcium fluoride according to the following
stoichiometric
relation:
S UOzFZceq~ + Ca(N03)ZC,q~ ~ UOz(NO3)2(89) + CaF2cS~
The heat of reaction is expected to negligible. The slurry 8 is transferred to
a liquid cyclone
where the calcium fluoride precipitate is removed. A fraction of the calcium
fluoride can be
recirculated to function as seed crystals in the Precipitator.
The calcium fluoride precipitate 9 is washed and dried in the Slurry
Washer/Dryer using
deionized water 5 from the Deionized Water Supply. The resulting dry calcium
fluoride product
11 is transmitted to a CaF2 Storage Facility. Since the uranium concentration
associated with the
calcium fluoride is anticipated to be sufficiently low, the stored calcium
fluoride 13 can be
dispensed as a saleable product.
The uranyl nitrate solution 10 is sent to the Uranyl Nitrate Solution
Adjustment Station.
Waste 12 from the Uranyl Nitrate Formation Station is forwarded to Waste
Treatment.
F. Uranyl Nitrate Solution Adjustment By The Addition Of Urea, Concentrating
The
Solution, And Increasing The Acidity
FIG. 4.6.2 depicts the flow sheet for the Uranyl Nitrate Adjustment Station.
The material
and energy balances for the exemplary process are shown in Table 4.6.2.
It has been found that more desirable gelation occurs if the uranyl nitrate
solution is
adjusted prior to gelation. In particular, it is desirable to decrease the
acidity of uranyl nitrate
solution and to add urea (CO(NHZ)2) in order to stabilize the uranyl ion.
Recycled urea 7 from the Urea Recycle Station, described below, is combined
with
deionized water 6 obtained from the Deionized Water Supply in a mixing vessel.
The urea is
dissolved and, as necessary, urea powder 8 is added to form a solution having
a molar ratio of
1 to 1.5, and preferably 1.25, urea to uranium.
The urea solution 3 is added to the uranyl nitrate solution 1 from the Uranyl
Nitrate
Formation Station in a Vapor Recompression (VR) Evaporator. The VR Evaporator
provides
the benefits of multistage evaporation in a single-stage unit and achieves
typical evaporation
efficiencies of 0.0452 KW-hr per kg of water (35 BTU/lb of water evaporated,
as compared to
normal values of around 1,000 BTU/lb). The evaporator advantageously performs
three
functions: ( 1 ) mixing the urea with the uranium solution to form the
urea/uranium complex; (2)
concentrating the uranium solution; and (3) rendering the solution slightly
acid deficient (i.e.,
having an anion (nitrate and fluoride) to uranium molar ratio of 1.5 instead
of 2). The uranyl
nitrate solution . generated by the VR Evaporator contains uranium in the 4.8-
3.0 molar
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concentration range (Table 4.6.2 utilizes a value of 2.2 molar) and is dense,
with a specific
gravity of approximately 4.8. The overhead product evolved from the VR
Evaporator is a dilute
nitric acid solution 2 (approximately 2%), which is transferred to the Calcium
Nitrate
Reconstitution Station, described below. As the uranyl nitrate solution is
generated in the VR
5 Evaporator it is transferred to a Uranyl Nitrate Storage Tank. The uranyl
nitrate solution 4 is then
chilled to approximately 0°C before the chilled solution 9 is
transferred to the Gel Solution
Preparation Station. Since the solution boiling point elevation data for this
solution is not readily
available, the evaporation energy shown in the material and energy balance in
Table 4.6.2 is
estimated at 0.1292 KW-hr/kg (100 BTU/lb).
10 Facility waste 14 from the Uranyl Nitrate Adjustment Station is sent to
Waste Treatment.
In the preferred gelation process, the uranium feed material is obtained in
the form of
uranium hexafluoride. However, Uranium and, in particular, depleted uranium is
available in
a variety of forms, such as uranium metal, low density uranium oxides, and
uranium tetrafluoride.
Thus, it is desirable to incorporate theses forms into the gelation process.
In accordance with the
15 present invention, the uranyl fluoride solution beneficially assists the
dissolution of alternative
feed materials. In particular, the uranyl fluoride solution can be used to
dissolve alternative
uranium feed materials. It is believed that this alternative route will be
particularly advantageous
because it utilizes fewer reagents, requires less precipitant, and generates
less waste. Thus, it is
expected that 80-100% of its total uranium feed will be obtained as uranyl
fluoride, derived from
20 the uranium hexafluoride, with the remaining 0-20% of the uranium feed
obtained from
alternative sources of uranium. It may be desirable to facilitate the process
by adding less than
stoichiometric amounts of nitric or hydrofluoric acid and 0.0001-0.5% a
catalyst such as
fluorboric acid (HBF4), fluorboric acid as a catalyst. Urea at a 1-1.5 molar
ratio (to the total
uranium) is also added prior to dissolution of the alternative uranium form.
After dissolution,
25 0.01-10% of aluminium, as the fluoride or nitrate, is added to complex the
fluoride ion.
G. Preparation Of Gel-Former Solution Having An Ammonia Releasing Compound
FIG. 4.7 depicts the flow sheet for the Gel Former Preparation Station. The
material and
energy balances for the exemplary process are shown in Table 4.7.
As discussed above, in the preferred embodiment, gelation is preferably
accomplished
30 using an internal gelation technique. Internal gelation of the chilled
uranium nitrate solution
from the Uranyl Nitrate Adjustment Station preferably utilizes a gel formation
solution
comprising an ammonia releasing compound, such as hexamethylenetetramine
("HMTA"),
(CH2)6N4. Other amines can also be used, such as ethylene diamine ("EDA").
HMTA powder 4 is dissolved in deionized water 2 from the Deionized Water
Supply in
a Mixing Tank. Recycled HMTA solution 3 from the HMTA Recycling Station,
described
below, is introduced to form a solution having about a 3 molar concentration.
The HMTA
solution 5 is subsequently chilled to around 0°C. The chilled HMTA
solution 7 is combined with
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the chilled urea-uranyl nitrate solution from the Uranyl Nitrate Adjustment
Station in a Static
Mixer. The resulting HMTA-containing urea-uranyl nitrate broth is chilled to
about 0°C in order
to avoid HMTA decomposition and premature precipitation of ADU. Even at these
reduced
temperatures, however, the solution has a limited shelf life, on the order of
1-3 days. Stabilizers,
such as surfactants and aliphatic hydrocarbons can be added to the solution to
extend its shelf
life. For example, a low concentration (0.001 to 1%) of a surfactant can be
added to the solution
to aid sphere formation and to inhibit particle agglomeration during gel
formation.
In the exemplary process shown in FIG. 4.7 and Table 4.7, a first portion 9
containing
approximately 70% of the chilled broth solution 8 is transferred to a 1,200
Micron Gel Formation
Station, and a second portion 10 containing approximately 30% of the chilled
broth solution 8
is transferred to a 300 Micron Gel Formation Station. Both of these facilities
are described in
additional detail below. As described previously, a minimum of two particle
sizes are desirable;
one size relatively coarse ( 1000 to 2000 microns, and preferably 1000 to 1300
microns diameter)
and one size relatively fine (30 to 1000 microns, preferably 300 to 500
microns in diameter).
This allows for closer packing and higher densities, resulting in better
shielding.
Facility waste 11 from the Gel Former Preparation Station is sent to Waste
Treatment.
H. Gel Sphere Formation Of Uranium/Ammonium Diuranate Precipitate By Internal
Gelation Techniques
FIG. 4.8 depicts the flow sheet for the 1,200 Micron Gel Formation Station.
The material
and energy balances for the exemplary 1,200 micron process are shown in Table
4.8.
In the Gel Formation Station, spheres of uranium dioxide are preferably formed
by an
internal gelation process. In this process, small dmps of the chilled broth
solution 1 from the Gel
Solution Preparation Station are dispersed using vibrating feed nozzles (not
shown) into a Gel
Forming Column containing oil. In the past, gelation was accomplished using
columns of
chlorinated solvents such as trichloroethylene (TCE) and perchloroethylene.
The preferred
embodiment, however, uses an oil, such as kerosene or fuel oil nos. 1, 2 or 3,
which are relatively
non-toxic and nonflammable. These oils also have low ash and residue contents
and gum less
than other oils, such as heat transfer oils. Further, use of oils eliminates
the production of acid
gases that is associated with the use of halogen-containing solvents. Still
further, when oil is
used, washing operations can be advantageously scaled back, because any oil
carried over to the
sintering step can be burnt away. Alternatively, oil that is carned over to
the sintering process
can be pyrolized to carbon for carbide production, dissociated into a coating,
or reformed into
hydrogen gas.
While the prior art uses high velocity or air impact/impingement nozzles. The
preferred
embodiment uses low kinetic energy nozzles, such as ultrasonic nozzles, which
reduce energy
consumption, reduce gas handling, reduce deformation, increase homogeneity,
and provide better
control with wider operational ranges. The vibrating nozzles fragment the
chilled broth solution
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into droplets 0.1-6 mm diameter in an air space above the Gel Forming Columns.
Fragmentation
of the chilled broth solution can optionally be undertaken using a nitrogen
purge 2. Nozzle
flowrates will vary, but are relatively small; on the.order of liters of
solution per hour. As the
immiscible droplets fall through the oil, surface tension effects form each
droplet into a sphere,
the diameter of which is determined by the size and vibration frequency of the
feed nozzles. The
higher temperature of the oil, ranging from 50 to 100°C, initiates the
dissociation of the
ammonia-releasing compound (HMTA) and the formation of a uranium/ADU-like
precipitate.
Typically, the "green" gel spheres formed in the Gel Formation Station are
about three times the
desired diameter of the final product. Therefore, for a final uranium dioxide
microsphere
diameter of 1,200 microns, the green gel sphere diameter should be about 3,600
microns.
The typical minimum free-fall residence time for gel sphere formation in the
Gel Forming
Column is in the 20-30 second range, after which the gel spheres are
sufficiently well-formed to
avoid sticking and deformation. This residence time is achieved by having a
column of sufficient
height, by countercurrent flow of the oil. or by a combination of both. The
preferred embodiment
uses both.
In the exemplary process described in Table 4.8, the following system
characteristics are
utilized: (1) nozzle flow rate: 4 liters/min; (2) green sphere diameter: 3,600
microns (0.36 cm);
(3) column residence free-fall time: one minute. Column height is predicated
upon one-minute
residence time, which translates into an active height of approximately 22
meters without any
credit for the oil flow. Using an upwards oil flow velocity of 30 cm/sec (1
ft/sec) produces a
column height of 4 meters (13 feet). For piping and plumbing connections, flow
disengagement
sections, and sphere aging, Table 4.8 utilizes a 6.1 meter height (20 feet). A
four-inch diameter
represents the full-length column diameter required per nozzle (i.e., no
converging/diverging
sections). Setting residence time is sixty minutes. Thus, a nominal column
diameter of 30 cm
( 1 foot) would accommodate approximately seven nozzles, have an aqueous feed
rate of
approximately 28 liters per hour (0.12 gpm), and an oil flow rate of 79,500
liters per hour (350
gpm). In contrast, a nominal column diameter of 51 cm (20 inches) would
accommodate
nineteen nozzles, have an aqueous feed rate of 76 liters per hour (0.33 gpm),
and an oil flow rate
of 223,000 liters per hour (980 gpm). These analyses assume a nominal column
diameter of 31
cm (i.e., 7 nozzles) for producing the 1,200 micron gel spheres.
After the green gel spheres have been formed and fallen to the bottom of the
column, they
will have developed sufficient strength to resist deformation under their own
weight. These
green gel spheres remain in the hot oil for thirty to sixty minutes for
setting. Setting
advantageously permits additional precipitation to occur and hardens the
spheres. Even after
setting, however, only about 1% of the HMTA will typically decompose, and,
thus, only about
S% of the precipitation reaction will have gone to completion. As much as
about 30 cm of a
static bed of green spheres will accumulate at the bottom of the Gel Forming
Column during the
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setting process. These green gel spheres 5 are removed from the Gel Forming
Column and
transferred to the 1,200 Micron Sphere Aging/Washing Station.
Oil from the column overflows to an Oil Overflow Tank. a portion 7 of the of
the oil 8
from the Oil Overflow Tank is recirculated through a heat exchanger and the
heated oil 6 is
' S returned to the Gel Forming Column. Since the oil in the Gel Forming
Column absorbs some
water from the gel spheres (up to its saturation limit), a fraction of the oil
10 is sent to the Oil
Purification Station, described below, for water removal and the purified oil
4 returned to the Gel
Forming Column.
Facility waste 13 from the 1,200 Micron Gel Formation Station is sent to Waste
Treatment.
FIG. 4.9 depicts the flow sheet for the 300 Micron Gel Formation Station. The
material
and energy balances for the exemplary 300 micron process are shown in Table
4.9.
The gel formation column for the 300 micron spheres is sized in the same
manner as the
1,200 micron spheres, only the nozzle flow rates are different the values set
forth above for the
1,200 micron case.
For a 300 micron final diameter, the green, gel sphere diameter would be
approximately
900 microns (0.09 cm). This results in a terminal velocity of approximately
18.3 cm/sec.
Assuming the same droplet velocity as in the 1,200 micron column (6.6 cm/sec),
then the upward
oil velocity would have to be 14.7 cm/sec. This translates into an oil flow
rate of 31,000 liters
per hour (136 gpm) and 86,300 liters per hour (380 gpm) for the 31 and 51 cm
columns,
respectively. The number of nozzles and the aqueous feedrates would be the
same as for the
1,200 micron case. For simplicity and to bound the case, the columns for the
300 micron
diameter spheres are the same height as the columns for the 1,200 micron
microspheres. A S 1
cm column diameter is used as the basis because it provides similar oil flow
rate characteristics
as for the 1,200 micron columns, and, thus, would require the same sized
equipment. This
column size translates into 19 nozzles.
I. Oil Purif cation
FIG. 4.10 depicts the flow sheet for the Oil Purification Station. The
material and energy
balances for the exemplary process are presented in Table 4.10.
As discussed above, the oil used in the Gel Formation Columns will absorb
water from
the gel spheres, up to the solubility limit of water in the oil. While exact
limits are not defined,
gel formation precedes more advantageously with oils that are not water
saturated. Therefore,
it is desirable to dry the oil.
' 35 Drying can be accomplished by several methods. Chilling of the oil
followed by phase
separation is the most straightforward method, but requires equipment and
piping. Alteratively,
molecular sieves and other adsorbents could be used to remove water from the
oil , but would
require a regeneration system and handling of the tramp oil. Membrane systems
could also be
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used, but are probably not cost effective. Therefore, in the preferred
embodiment, chilling
followed by phase separation is used to dry the oil.
The wet oil stream 1 from the 1,200 Micron Gel Formation Station, the wet oil
stream
2 from the 300 Micron Gel Formation Station, and the wet oil stream 3 from the
Ammonium
Hydroxide Setting/Washing Station, described below, are fed to the Oil
Purification Station. Wet
oil streams 1-3 will likely contain around 1,000 ppm dissolved water, a
typical value for oils
around 70°C. The wet oil streams 1-3 are combined into a single stream
4 which is introduced
into a Heat Exchanger and cooled. The oil 5 is then passed through a Chilling
System and cooled
to around 5°C. At this low temperature, the water solubility is only
about 100 ppm. As a result,
phase separation occurs. The resulting oil/water mixture 6 is sent to an
Oil/Water Separator
which coalesces and recovers the water. The dried oil 7 is reheated using the
heat exchanger.
Makeup oil 9 is added to the heated, dried oil. The resulting mixture 11 is
divided into a first
portion 12 that is returned to the 1,200 Micron Gel Formation Station and a
second portion 13
that is returned to the 300 Micron Gel Formation Station. The water 8, which
is saturated with
oil, is pumped to Waste Treatment.
J. Gel Sphere Aging By Setting/Washing With Ammonium Hydroxide
FIG. 4.11 depicts the flow sheet for the 1,200 Micron Sphere Setting/Washing
Station.
The material and energy balances for the exemplary 1,200 micron process are
shown in Table
4.14.
Additional ammonia necessary to complete the ADU-like precipitation reaction
is
supplied by washing the spheres in a setting solution of 10% ammonium
hydroxide solution. The
stoichiometry for this reaction is:
2 UOZ (N03)z + 2 NH4 OH + H20 ~ (NH4)2 UZ O, + 4 HN03
During this process, the spheres 1 are periodically discharged from the 1,200
Micron Gel
Formation Column into a Setting Washing Column having the same diameter as the
formation
column (31 cm). A setting solution 4 of 10% ammonium hydroxide is formed by
combining the
ammonium hydroxide stream 3 from the Ammonium Hydroxide Purification Station,
described
below, and a portion 8 of the ammonium hydroxide 6 from the Ammonium Hydroxide
Overflow
Tank. The setting solution 4 is circulated through the bed of spheres in the
Setting/Washing
Column, causing them to harden. Typical washing times are about one hour.
Typical flow rates
are on the order of five to ten times the initial broth solution feed rate to
the column. Ambient
temperatures are believed to supply sufficient heat for the reaction.
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During washing, impurities such as urea and HMTA are removed from the spheres
by
leaching (i.e., preferential absorption into the aqueous phase), typically
with efficiencies of
90-95%, because of favorable thermodynamics.
About 95% of the tramp oil carried over from the Gel Formation Station is
skimmed from
5 the Ammonium Hydroxide Overflow Tank. This oil 10 is then recycled to the
Oil Purification
Station. A portion of the oil is believed to dissolve to ammonium hydroxide
solution,
approximately 100 ppm, while the remainder clings to the spheres. The ammonium
hydroxide
also causes reformation of any HMTA that initially reacted with the ADU.
After washing, a majority 7 of the ammonium hydroxide 6 collected in the
Ammonium
I 0 Hydroxide Overflow Tank is sent to the Ammonium Hydroxide Purification
System to remove
impurities (i.e., the urea and HMTA) and to allow its recycle. The purified
ammonium hydroxide
solution 3 is then returned to the 1,200 Micron Sphere Setting/Washing Station
and combined
with a minority 8 of the ammonium hydroxide solution 8 collected in the from
the Ammonium
Hydroxide Overflow Tank.
15 As set forth in the material and energy balances in Table 4.11, it is
believed that the
remaining 95% of the precipitation reaction occurs in this step. Furthermore,
oil carryover is
believed to be equivalent to the void space in the packed bed volume of the
spheres. The
recirculation rate is calculated based upon five times the broth solution feed
to a 31 cm diameter
column; this corresponds to approximately 2.3 liters per minute (0.62 gpm) per
column.
20 After setting and washing of the gel spheres is complete (i.e., the uranium
is completely
converted to ADU and no more ammonium hydroxide reacts), the aged spheres 9
are transferred
to the 1,200 Micron Drying Station for further processing.
Facility waste lI from the 1,200 Micron Sphere Setting/Washing Station is sent
to Waste
Treatment.
25 FIG. 4.12 depicts the flow sheet for the 300 Micron Sphere Setting/Washing
Station. The
material and energy balances for the exemplary 300 micron process are shown in
Table 4.12.
The 300 Micron Setting/Washing Station is identical to the 1,200 Micron
Setting/Washing Station. As with the 1,200 micron spheres, the remainder of
the ammonia
required to complete the ADU-like precipitation reaction is supplied by
washing the spheres in
30 a setting solution of 10% ammonium hydroxide. Furthermore, as with the
1,200 micron
particles, typical washing times are on the order of one hour and typical flow
rates are around five
times the flow of the initial broth solution fed to the column. These
conditions correspond to a
large excess of ammonium hydroxide, since only around 10% or so is actually
consumed by the
setting reactions. Hence, the reactions are rapid.
35 K. Gel Sphere Drying And Liberation Of Ammonia and Water
FIG. 4.13 depicts the flow sheet for the 1,200 Micron Sphere Drying Station.
The
material and energy balances for the exemplary 1,200 micron process are shown
in Table 4.13.
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In the preferred embodiment, the spheres are subjected to slow, low
temperature drying
in order to remove the volatile species, such as water, nitric acid, and
ammonia, while avoiding
cracking. Non-volatile species, such as urea and HMTA, are not removed in this
process.
Preferably, the aged spheres 1 from the 1,200 Micron Sphere Setting/Washing
Station are
introduced into a commercial Tray Dryer. The Tray Dryer is equipped with a
moving screen
which provides a contacting arrangement for drying with warm nitrogen gas 7.
Warm nitrogen
gas 7 is purified in a commercial Nitrogen Condenser/Dryer and recycled.
Adiabatic drying is
assumed.
The Nitrogen Condenser/Dryer uses refrigeration to condense volatile species 9
which
are transferred to the Ammonium Hydroxide Purification Station. A recuperative
heat exchanger
(not shown) minimizes energy consumption. An adsorbent bed of molecular sieves
(not shown)
is also used to further dry the nitrogen. The molecular sieves are
periodically regenerated by a
purge air stream (not shown) The material and energy balances for the
exemplary process in
Table 4.13 utilize 99% condensation of the water in the Tray Dryer and 100%
condensation of
the nitric acid and ammonium hydroxide in the water. The remaining 1 % of the
water is removed
by the adsorbent bed. Although leak-free operation has been attained
commercially, the analyses
in Table 4.13 utilize a nitrogen makeup stream 4 represents the reconstituted
nitrogen stream.
Stream 5 is the supply air used to regenerate and purge the dryer, and becomes
stream 8.
The dried spheres 6 are removed from the Tray Dryer and transferred to the
1,200 Micron
Sphere Conversion/Sintering Station.
Facility waste 10 from the 1,200 Micron Sphere Drying Station is sent to Waste
Treatment.
FIG. 4.14 depicts the flow sheet for the 300 Micron Sphere Drying Station. The
material
and energy balances for the exemplary 300 micron process are shown in Table
4.14.
This station performs the drying for the 300 micron sized product. Apart from
the flow
rates, the operations are identical to those for the 1,200 micron spheres
discussed in Section
4.4.14. FIG. 4.14 shows the flow sheet, and Table 4.14 summarized the mass and
energy
balances.
L. Gel Sphere Conversion and Sintering By Heating
FIG. 4.I5 depicts the flow sheet for the 1,200 Micron Sphere Conversion and
Sintering
Station. The material and energy balances for the exemplary 1,200 micron
process are shown
in Table 4.15.
The dried spheres 1 obtained from the 1,200 Micron Drying Station are
introduced into
a Vertical Tube Furnace (VTF) where the spheres are converted to uranium
dioxide and sintered
{i.e., densified). The dried spheres react in the VTF according to the
following equation:
{NH4)ZU~O, + 2 HZ ~ 2 UOZ + 3 Hz0 + 2 NH3
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The VTF consists of a vertical tube containing a slowly moving packed bed of
ADU spheres
within an electrically heated furnace. Dry spheres 1 are added at the top of
the VTF, while a
discharge valve discharges dense product 3 from the bottom.
Typical residence times in the VTF approach twelve hours at maximum
temperatures of
I ,100 to 1,300°C. In normal conversion and sintering, hydrogen is
added to an argon diluent,
typically at concentrations of 3.5-4 ppmv, and the gas mixture 2 converted to
uranium dioxide
and sintered (i.e., densified) is circulated upwards through the VTF.
Approximately 50% of the
hydrogen is consumed in this process, and the surviving off gas 4 is sent to
the Gas Purification
Station for recycle. The concentrations of hydrogen in the argon/hydrogen
mixture are below the
I 0 lower flammability limit for hydrogen in air, and, thus, prevent hydrogen
fire hazards within the
VTF system itself.
Meanwhile, urea, HMTA, ammonium hydroxide, nitric acid, and the oil thermally
crack
and reform in the VTF as follows:
CO(NHZ), + H,O ~ COZ + 2 NH3 (urea)
(CHZ)6N4 + 12 H20 ~ 4 NH3 + 12 HZ + 6 COZ (HMTA)
NH40H ~ NH3 + H20
2 HN03 + HZ ~ 2 NOZ + 2 H20
(CHZ),° + 20 Hz0 ~ 10 COZ + 30 Hz
The enthalpy balance set forth in Table 4.15 for the exemplary process
utilizes a value of 30 KW
per VTF, based upon average values reported in the literature. This enthalpy
is approximately
twice that required for volatilization of the chemical species. For uranium
dioxide flow rates of
15,000 tonnes/year, approximately 20 VTF's are required. Unlike normal
conversion and
sintering operations, there is a net hydrogen production generated from the
cracking and
reforming of the trace impurities. Also, since the reactions result in a net
consumption of water,
steam (not shown) is added to the argon diluent in the VTF feed.
Alternatively, steam could be
added internally within the VTF. These hydrogen and carbon production effects
accrue from the
' carryover of residual oil upon the spheres. The argon flowrates are based
upon maintaining a 4%
hydrogen in argon concentration at the exit of the VTF.
- The preferred embodiment uses nitrogen in place of argon because of its
lower cost and
the residual nitrogen in the dense uranium (as uranium nitride) provides
additional neutron
shielding.
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Facility waste 5 from the 300 Micron Sphere Conversion and Sintering Station
is sent to
Waste Treatment.
FIG. 4.16 depicts the flow sheet for the 300 Micron Sphere Conversion and
Sintering
Station. The material and energy balances for the exemplary 1,200 micron
process are shown
in Table 4.16.
Conversion and sintering of the 300 micron particles occur in a manner
analogous to the
1,200 micron spheres. Approximately 8 VTF's are required for the 300 micron
spheres.
M. Gel Sphere Collection
As discussed previously, the final dense spheres are approximately 30% the
size of the
initial droplets (gel spheres). It is desirable to have at least two size
ranges for better packing,
higher bulk densities, and, consequently, better shielding. These size ranges
are 1-2 mm for the
coarse fraction and 0.030-1 mm for the fine fraction. The coarse spheres
contain the majority of
the uranium on a mass basis.
N. Calcium Nitrate Reconstitution for the Calcium Nitrate Reconstitution
Station
FIG. 4.17 depicts the flow diagram for the Calcium Nitrate Reconstitution
Station. The
material and energy balances for the exemplary process are shown in Table
4.I7.
The overhead product 3 consisting of a dilute nitric acid stream from the
Uranyl Nitrate
Solution Adjustment Station is added to a Mixing Tank along with dry lime 2 to
form calcium
nitrate. Additional nitric acid i is added to the output 3 of the Mixing Tank,
and the mixture 4
is sent to a Vapor Recompression (VR) Evaporator where Excess Water 5 is
removed and sent
to the Deionized Water System. The Reconstituted Product 6 from the VR
Evaporator containing
the calcium nitrate is transferred to the Uranyl Nitrate Formation Station for
reuse.
O. Ammonium Hydroxide Solution Purification
FIG. 4.18 depicts the flow sheet for the Ammonium Hydroxide Solution
Purification
Station. The material and energy balances for the exemplary process are shown
in Table 4.18.
Ammonium hydroxide solution from the 1,200 Micron Setting/Washing Station 1
and
from the 300 Micron Setting/Washing Station 2 are sent to Oil/Water
Separators, which remove
any excess oil from the respective ammonium hydroxide solutions. The separated
oil portions
3 and 4 are sent to the 1,200 Micron and 300 Micron Gel Forming Stations,
respectively. In a
large plant, significant quantities of oil would be expected to be carried
over due to the size of
the equipment. Next, the ammonium hydroxide streams from the Oil/Water
Separators are
merged into a single stream 6 and combined with the condensate 15 formed by
the combination
of the condensates 13 and 14 from the 1,200 Micron Dryer Station and the 300
Micron Dryer
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Station, respectively. A 50% Sodium Hydroxide Solution 5 is added to the
condensate/ammonium hydroxide mixture 16 to neutralize the nitric acid,
displace ammonia, and
raise the solution pH above 10. A VR Evaporator processes the resulting
mixture 7. The
overhead stream 8 from the VR Evaporator is a dilute, 10% ammonium hydroxide
solution,
which is sent to the Ammonium Hydroxide Solution Reconstitution Station,
described below,
for further processing. The bottoms 9 from the UR Evaporator contain sodium
nitrate, urea, and
HMTA, with a total dissolved solids concentration approaching 40%. Traces of
oil are removed
from the bottoms 9 by activated carbon in a Carbon Guard Bed. A membrane
system, such as
a reverse osmosis membrane, separates sodium nitrate permeate 10 from the urea
and HMTA in
the de-oiled solution 17. The sodium nitrate permeate 10 is sent to Waste
Treatment, while the
retentate 11 reports to the Urea and HMTA Recovery Station, described below.
An optimized
design for this portion of the system would preferably evaporate less water
and use solubility
limits to the utmost advantage. Facility waste 12 from the Ammonium Hydroxide
Solution
Purification Station is sent to Waste Treatment.
P. Vertical Tube Furnace Gas Purification
FIG. 4.19 depicts the flow diagram for the Vertical Tube Furnace (VTF) Gas
Purification
Station. The mass and energy balances for the exemplary process are shown in
Table 4.19.
The hot gases 2 and 3 from the VTF's in the 1,200 Micron and 300 Micron
Conversion
and Sintering Station, respectively, are merged into a single stream 4 and
filtered. (Although it
is expected that little or no material will be captured due to the size,
narrow diameter distribution,
and density of the microspheres.) The filtered gas is then passed through a
Heat Exchanger
where it is cooled. Thereafter, the cooled gas 10 is passed to a Spray Tower.
The Spray Tower
condenses the ammonia and nitrogen dioxide in an aqueous solution 11 which is
sent to the
Ammonium Hydroxide Purification Station, for further processing.
A 50% solution of sodium hydroxide 8 is diluted in a Mixing Tank with
deionized water
7 from the Deionized Water Supply 5. The resulting solution 9 having a pH of
10-13.5 is
introduced into a COz Scrubber Tower where it is used to scrub the carbon
dioxide from the gas
stream 12 emerging from the Spray Tower, according to the following equation:
2 NaOH + COZ ~ Na2C03 + Hz0
The sodium carbonate product solution 14 discharged from the COZ Scrubber
Tower is sent to
Waste Treatment for disposal.
The remaining hydrogen-inert gas (nitrogen in the preferred route) mixture 13
emerging
from the COZ Scrubber Tower is filtered. A membrane system is then used to
separate the excess
hydrogen 15 from the remaining hydrogen-argon gas mixture 13. This separation
is
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accomplished with relative ease due to the large molecular weight differences
between the
hydrogen and argon gases and the selectivity of the membranes; hydrogen
diffuses rapidly
through membranes, and the other gases do not diffuse at all. Typical
membranes include
polysulfone and related, hollow fiber designs. The excess hydrogen 15 flows to
the Utility
5 Supply Building where it can be used as fuel in the incinerator. The
purified hydrogen-argon
mixture 16 is enriched with additional argon 20 from the Argon Supply and
passed through the
Heat Exchanger. The heated and enriched hydrogen-argon mixture 17 is divided
into two
portions 18 and 19 that are sent to the VTF's in the 1,200 and 300 Micron
Conversion and
Sintering Stations, respectively. Steam is added to reform (destroy) tramp oil
into hydrogen,
10 methane, and carbon. The flow sheet assumes a closed system, but includes a
10% per year
makeup for the argon.
Facility waste 21 from the VTF Gas Purification Station is sent to Waste
Treatment.
Q. Ammonium Hydroxide Reconstitution
FIG. 4.20 depicts the flow diagram for the Ammonium Hydroxide Reconstitution
Station.
15 The material and energy balances for the exemplary process are shown in
Table 4.20.
The ammonium hydroxide solution 1 from the Ammonium Hydroxide Purification
Station is reconstituted by mixing in a Mixing Tank with a fresh, 50% ammonium
hydroxide
solution 2 and deionized water 3. The resulting 10% solution of ammonium
hydroxide 4 is
divided into two fractions 5 and 6 that are sent to the 1,200 Micron and 300
Micron
20 Setting/Washing Stations, respectively.
R. Urea and HMTA Recovery
FIG. 4.21 depicts the flow diagram for the Urea and HMTA Recovery Station. The
material and energy balances for the exemplary process are shown in Table
4.21.
Since urea and HMTA effectively function as complexing agents and catalysts
the
25 preferred embodiment, they are present in sizable quantities. Therefore, it
is desirable to separate
and recycle them. In particular, it is desirable to have a urea product that
is relatively free of
HMTA in order to avoid premature precipitation of uranium dioxide. It is
noted, however, that
the recovered HMTA can contain urea without significant effects upon the
overall process.
It is well known that urea and HMTA form crystals from concentrated solutions.
Thus,
30 this property is often used in their manufacture and purification. Urea
possesses a solubility of
approximately 50% at 17°C in water, and 17% at 20°C in alcohol.
For HMTA, the solubility is
about 45% at 15°C in water and 3% at i 5 °C in alcohol. However,
urea possesses better
crystallization properties, including the capability to form clathrate-type
crystals (essentially
double compound crystals) in the presence of low concentrations of paraffinic
hydrocarbons.
35 Thus, in the preferred embodiment, urea is selectively crystallized away
from the HMTA, via a
combination of thermal and chemical effects, and both are recycled.
Alternatively, tailored
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membranes may be effective for the urea-HMTA separation. The crystallization
process may be
advantageously operated in the batch mode for enhancement of the separation.
S. Cylinder Decontamination
FIG. 4.22 depicts the flow diagram for the Cylinder Decontamination Station.
The
material and energy balances for the exemplary process are shown in Table
4.22.
The cylinder decontamination station removes the residual uranium ("heel")
from the
cylinder via a three step operation. In the first step, the empty cylinders 1
are filled continuously
rinsed with deionized water 2 from the Deionized Water Supply 3 for
approximately four hours.
Since most of the remaining uranium fluorides and oxyfluorides are dissolved
in this rinsing step,
low decontamination factors (DF's) of 10-20 (i. e., 90-95% removal) and
production of the highly
soluble, uranyl fluoride is expected. The rinse water 6, which becomes a
dilute uranyl fluoride
solution, is periodically recycled to the Uranyl Fluoride Formation Station as
part of the quench
water requirements. In the second step, the empty cylinders are filled and
rinsed with a dilute
(5%) nitric acid solution 4 for four hours. The nitric acid solution 4 allows
ion exchange
recovery of the uranium in a side column, thus regenerating the acid and
allowing its reuse. This
provides for very high DF's of 50-1,000. Finally, in the third step, the empty
cylinders are filled
and rinsed with deionized water 5 from the Deionized Water Supply 3 for
approximately four
hours, to remove residual chemicals and traces of uranium. This provides an
additional DF of
2-10.
Facility Waste 11 from the Cylinder Decontamination Station is sent to Waste
Management.
T. Waste Management
FIG. 4.23 depicts the flow sheet for the Waste Management Station. The
material and
energy balances for the exemplary process are shown in Table 4.23.
Liquid waste streams 9 from the Oil Purification Station and the VTF Gas
Purification
Station are carbon filtered to remove traces of oil and organics The filtered
waste stream 9 is
then passed through an Ion Exchange to remove the traces of ionic species.
Meanwhile, liquid
waste stream 10 from the reverse osmosis is filtered. Permeate produced in the
HMTA Recycle
Station is similarly filtered. Waste streams 9 and 10 are then merged into a
single stream 14
which is essentially pure water, and are discharged to Publicly Owned
Treatment Works
("POTW") or National Pollution Discharge Elimination System ("NPDES") Point.
No liquid
radioactive wastes are generated by this process.
Solid wastes from the operations are collected, and sorted. Incineration is
used to treat
combustible wastes, while non-combustible waste is compacted and solidified.
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MANUFACTURE OF DENSE URANIUM CARBIDE AND EXEMPLARY PROCESS
In the preferred embodiment, microspheres of depleted uranium carbides are
advantageously utilized as shielding materials, because they have the highest
densities of any
uranium compounds. More particularly, uranium monocarbide (UC) is preferred,
because it has
slightly better physical properties than uranium dicarbide (UCZ), including a
higher density and
thermal conductivity. In either form, however, uranium carbides react slowly
in moist air to form
uranium dioxide. Thus, in the preferred embodiment, it is desirable to apply
an impervious
coating to the depleted uranium carbide microspheres that will render them
inert under normal
conditions.
There are two principal routes for manufacturing uranium carbide materials: (
1 ) graphite
reduction; and (2) gelation. Historically, the production of uranium carbide
has been
accomplished using the reduction of uranium dioxide with carbon (e.g., from
graphite). This
process is described in M. Benedict, T. Pigford, and H. Levi, Nuclear Chemical
Engineering,
Second Edition, McGraw-Hill, New York, NY, 1981, incorporated herein by
reference. The
uranium dioxide starting materials utilized in the reduction process can be
manufactured from
uranium hexafluoride or uranyl nitrate solutions using a variety of known
methods. In the
preferred embodiment, however, uranium carbide materials are manufactured
directly utilizing
a gelation process that is substantially similar to the uranium dioxide
gelation process discussed
previously.
The graphite reduction and gelation routes for production of uranium carbide
are
discussed in detail below.
A. Graphite Roate
FIG. 5 prevents an overview of the graphite route for production of depleted
uranium
carbides.
As discussed above, the reduction of uranium dioxide to form uranium carbide
is known
in the art. In the preferred embodiment, a uranium dioxide solid is mixed with
carbon powder
and an polyethylene binder to form a slurry. This slurry is then oven dried
and ball-milled to
sand-sized particles (0.03-2 mm). The oxides are then converted into carbides
in a vacuum
heating operation, in which the oxygen is replaced by carbon, and,
consequently, carbon
monoxide and carbon dioxide are released. The amount of carbon in the initial
mixture
determines whether the uranium monocarbide or the dicarbide are formed. The
resulting small
particles of uranium carbide are fed through a furnace operating in excess of
the normal melting
point of the carbide, and the microspheres are formed. Surface tension effects
produce the
spherical shape.
3 5 The depleted uranium dioxide starting materials used in the reduction
process can be
generated using. any of a variety of known methods. In the preferred
embodiment, however,
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43
uranium dioxide is produced by gelation. While the gelation route for
production of uranium
dioxide can be advantageously used to generate uranium carbide by reduction of
uranium dioxide
microspheres, it is more desirable to produce the uranium carbides directly in
the gelation process
as described in detail below.
Once the depleted wanium carbide microspheres have been produced, it is
desirable to
apply a coating which will effectively isolate the uranium carbide from the
environment at the
microscopic level. While mufti-layer coatings are known in the art, the
preferred embodiment
utilizes a single coating of carbon. This coating can be applied in a
fluidized bed furnace in
which a stream of inert gas (usually argon) is introduced into the furnace to
levitate and heat the
carbide microspheres. A mixture of hydrocarbons is then introduced into the
gas stream. The
hydrocarbons dissociate when they come in contact with the outer surface of
the microspheres,
and form a dense, pyrolytic carbon layer.
B. Gelation Route
FIG. 4.1a depicts the modified flow diagram for the manufactwe of depleted
uranium
carbide microspheres by gelation. The overall material and energy balances are
for the
exemplary process are shown in Table 4.1a.
As is readily apparent from FIG. 4.1a, the process for production of wanium
carbide
microspheres is substantially similar to the process for production of wanium
dioxide shown in
FIG. 4.1. There are four main changes to the gelation process when uranium
carbide
microspheres are the desired end product: (1) a carbon powderJsurfactant
solution is prepared in
a Carbon Suspension Formation Station (FIG. 4.6.1 and Table 4.6.1); (2) the
carbon
powder/surfactant solution is added to the wanyl nitrate solution in the
Uranyl Nitrate Solution
Adjustment Station (FIG. 4.6.2a and Table 4.6.2a) (3) sintering of the uranium
carbide spheres
in the Uranium Carbide Coating Station is accomplished using a two-step
sintering process {FIG.
4.24 and Table 4.24); and (4) the wanium carbide microspheres are coated in a
Uranium Carbide
Coating Station (FIG. 4.25 and Table 4.25).
As shown in FIG. 4.1a, depleted uranium hexafluoride is reacted with steam to
produce
wanyl fluoride and hydrogen fluoride; the latter being recoverable in an
anhydrous form. The
solid uranyl fluoride is collected, quenched, and dissolved in water.
Adjustment of the residual
hydrogen fluoride concentration is optionally undertaken utilizing
distillation methods. As with
the wanium dioxide gelation process, uranyl fluoride can be utilized directly
for the formation
of microspheres. Nevertheless, in a conservative approach, the wanyl fluoride
is converted to
wanyl nitrate using calcium nitrate to precipitate the fluoride and form
uranyl nitrate in solution.
Meanwhile, a carbon suspension is formed in a Carbon Suspension Formation
Station.
- 35 FIG. 4.6.1 depicts the flow sheet for the Carbon Suspension Formation
Station. The material and
energy balances for the exemplary process are shown in Table 4.6.1.
The manufactwe of wanium carbide microspheres requires the addition of carbon
to
convert the uranium dioxide produced by the gelation process into uranium
carbide. In the
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44
preferred embodiment, carbon is introduced to the uranium solutions, prior to
gelation, as part
of an aqueous suspension of fine carbon particles. As shown below, the carbon
substitutes for
the oxygen in the uranium dioxide and generates carbon dioxide:
UO~ + 2C -- UC + COZ
UO~ + 3C ~ UCZ + COZ
The required quantity of carbon is estimated from the stoichiometry of the
conversion and
sintering reactions; the monocarbide requires less carbon than the dicarbide.
For the 100%
capacity case, 1,900 tonnes/year would be required for the monocarbide and
2,850 tonnes/year
for the dicarbide. The particulate form of carbon, such as carbon black or
fine graphite, is
preferred because of its small size, consistency, and ease of making a
suspension. The quantity
of added carbon depends upon the desired final carbide (monocarbide or
dicarbide), and is
usually between 14% and 25% of the broth solution (weight basis).
Consequently, surfactants
are added to the solution to stabilize the particles and keep them in
suspension. The
carbon/surfactant suspension is added to the uranium nitrate solution in the
Uranyl Nitrate
Adjustment Station, as shown in FIG. 4.6.2a and Table 4.6.2a.
In the preferred embodiment, a 20% carbon suspension is prepared by combining
appropriate amounts of carbon pigment 1 and deionized water 3 in a Mixing
Tank. About 1000
ppm of surfactant 2 is added to the Mixing Tank to facilitate dispersion of
the fine carbon
particles in the water. Gentle heating of the suspension to about 50°C
using a Heat Exchanger
also assists the dispersion. The resulting suspension 5 is pumped to the
Uranyl Nitrate Solution
Adjustment Station. FIG. 4.6.2a depicts the flow sheet for the modified Uranyl
Nitrate Solution
Adjustment Station for uranium carbide production. The material and energy
balances for the
exemplary process are shown in Table 4.6.2a.
Adjustment of the uranyl nitrate solution for production of uranium carbide is
virtually
identical to the uranium dioxide process described previously. In the modified
process, however,
the carbon/surfactant suspension 14 from the Carbon Suspension Formation
Station is introduced
into the VR Evaporator along with the uranyl nitrate solution 1 and the urea
solution 3. The
resulting solution is processed in the same manner as before.
As with the uranium dioxide routes, the carbon-containing, adjusted uranyl
nitrate
solution is used for gel sphere formation. In the preferred embodiment,
internal gelation routes
are preferred. Following gel sphere formation, the gel spheres are aged in an
ammonium
hydroxide solution. After aging, the gel spheres are dried at similar
temperatures to remove
water and excess ammonia.
Subsequently, a vertical tube furnace ("VTF") sinters the microspheres under
an
argon-hydrogen.atlnosphere. In the preferred embodiment, a two zone furnace
with an inert gas,
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such as nitrogen, is utilized in order to avoid over-reduction of the uranium.
FIG. 4.24a depicts
the flow sheet for the modified Uranium Carbide Sintering Station. The
material and energy
balances for the exemplary process are shown in Table 4.24a.
The sintering process for uranium carbide spheres is substantially the same as
uranium
5 dioxide sintering, except that a Two-Zone VTF is utilized to avoid over
reduction of the uranium
carbide. Too much hydrogen causes conversion the uranium carbide into uranium
metal and
methane. The first zone of the VTF utilizes an argon cover gas 2 containing 2-
4% hydrogen.
During this stage of the sintering process, most of the reaction and
generation of carbon
monoxide and dioxide occur. It should be noted that sintering of uranium
carbide spheres
10 produces substantially more carbon monoxide and carbon dioxide than with
uranium dioxide.
The second zone of the VTF operates at higher temperatures, using only argon
as the cover gas,
and results in sintering and the high densities desired. The final sintered
spheres have densities
usually exceeding 95% of the theoretical density for uranium carbides. If two
or three sizes of
microspheres are produced, then spatial densities exceeding 90% of theoretical
can be obtained
15 by vibratory loading methods.
Subsequently, as with the graphite route, fluidized bed furnaces apply
coatings to the
microspheres which effectively isolate the uranium carbide from the
environment at the
microscopic level. While the mass and energy balances in Table 4.24 are set
forth for two
coatings, it is expected that a single coating will be preferred.
20 PEROXIDE GELATION
Peroxide gelation is an alternative gelation process contemplated in
accordance with the
present invention. FIG. 6 provides an overview of the peroxide gelation
process. As shown in
FIG. 6, uranium hexafluoride is vaporized and defluorinated to produce
anhydrous hydrogen
fluoride and uranyl fluoride powder. However, the steam required for the
reaction comes from
25 a recycle stream 1 containing the azeotrope (HF~2H20) plus traces or uranyi
fluoride, nitric acid,
and aluminium nitrate. The uranyl fluoride powder is quenched and dissolved in
diluted,
aqueous nitric acid and used to dissolve uranium metal and low density oxide
feed materials.
Dissolution is aided by a nitric acid recycle stream 2. Fluorboric acid and
urea may be added to
the solution in the quench/dissolving step. Aluminum nitrate can be added in
the molar ratio of
30 0.001 to 1.25 to facilitate partial complexation of the fluoride ions. The
stream is chilled to 0 to
25 °C, and dispersed using nozzles into a hydrogen peroxide solution
bath or column. The
peroxide solution has a concentration between 0.5 to 50%, and is maintained
between 0-45 °C.
Uranyl peroxide (UO4-2H20) precipitates as a microsphere. The solids are
separated by screens
and filters washed with a dilute peroxide stream (0.001 to 5 M), dried in warm
nitrogen and
35 sintering under nitrogen to produce dense uranium dioxide microspheres.
Uranium carbides can be manufactured by adding a carbon suspension to the
uranium
stream prior to chilling and droplet formation. The uranyl peroxide
precipitate particle retains
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the carbon pigment, but allows the soluble species such as nitrates,
fluorides, and urea to diffuse
into the bulk solution. The two-step sintering procedure, described
previously, is used to produce
the dense carbides, but no hydrogen is needed.
The peroxide solution recycles between the precipitator and the
filter/separator. The
recycled solution contains hydrogen fluoride, nitric acid, aluminum nitrate,
fluorboric acid, and
traces of uranyl fluoride and peroxide. Once the peroxide is consumed by
reaction or otherwise
depleted by decomposition the stream is periodically or continuously withdrawn
and distilled.
The bottoms product from distillation contains the hydrogen fluoride-water
azeotroph (boiling
point circa 110°C), with aluminium nitrate, fluorboric acid, and traces
of uranyl fluoride/nitrates.
This stream is recycled to the defluorinator. The distillate product from
distillation contains
nitric acid, at temperatures of 50 to 75 °C. Molecular sieves,
distillation, or related means can
be used to remove excess water prior to recycle of the nitric acid stream 2 to
the quencher.
Although a particular form of the invention has been illustrated and
described, it will be
appreciated by those of ordinary skill in the art that various modifications,
alterations, and
substitutions can be made without departing from the spirit and scope of the
invention.
Accordingly the scope of the present invention is not to be limited by the
particular embodiments
set forth above, but is only to be defined by the following claims.
