Title: Radiation shielding materials and containers incorporating same
Abstract: An improved radiation shielding material and storage systems for radioactive materials incorporating the same. The PYRolytic Uranium Compound ("PYRUC") shielding material is preferably formed by heat and/or pressure treatment of a precursor material comprising microspheres of a uranium compound, such as uranium dioxide or uranium carbide, and a suitable binder. The PYRUC shielding material provides improved radiation shielding, thermal characteristic, cost and ease of use in comparison with other shielding materials. The shielding material can be used to form containment systems, container vessels, shielding structures, and containment storage areas, all of which can be used to house radioactive waste. The preferred shielding system is in the form of a container for storage, transportation, and disposal of radioactive waste. In addition, improved methods for preparing uranium dioxide and uranium carbide microspheres for use in the radiation shielding materials are also provided.
Patent Number: 6,960,311 Issued on 11/01/2005 to Mirsky, et al.
| Inventors:
|
Mirsky; Steven M. (Greenbelt, MD);
Krill; Stephen J. (Arlington, VA);
Murray; Alexander P. (Gaithersburg, MD)
|
| Assignee:
|
The United States of America as represented by the United States Department of Energy (Washington, DC)
|
| Appl. No.:
|
121871 |
| Filed:
|
April 15, 2002 |
| Current U.S. Class: |
252/478; 423/11; 423/261 |
| Intern'l Class: |
C01G 056/00; C04B 035/66 |
| Field of Search: |
252/478
423/261,11
|
References Cited [Referenced By]
U.S. Patent Documents
| 3087781 | Apr., 1963 | Levey, Jr et. al.
| |
| 3313602 | Apr., 1967 | Flack et. al.
| |
| 3518065 | Jun., 1970 | Triggiani.
| |
| 3617584 | Nov., 1971 | Flack et al.
| |
| 3697441 | Oct., 1972 | Petit.
| |
| 3748273 | Jul., 1973 | Smith.
| |
| 3862908 | Jan., 1975 | Fitch et al.
| |
| 4038202 | Jul., 1977 | Votocek.
| |
| 4119563 | Oct., 1978 | Kadner et al.
| |
| 4152395 | May., 1979 | Borner et al.
| |
| 4367184 | Jan., 1983 | Stinton.
| |
| 4663093 | May., 1987 | Haas et al.
| |
| 4671927 | Jun., 1987 | Alsop.
| |
| 4963294 | Oct., 1990 | Yato et al.
| |
| 6599490 | Jul., 2003 | Mirsky et al.
| |
Primary Examiner: Tucker; Philip C.
Attorney, Agent or Firm: Daubenspeck; William C., Gottlieb; Paul A.
Parent Case Text
This application is a divisional of Ser. No. 08/826,088 filed Mar. 24, 1997,
now U.S. Pat. No. 6,372,157.
Claims
1. A method for production of microspheres of uranium dioxide, comprising:
dispersing a solution of uranyl fluoride in hydrogen peroxide whereby uranyl
peroxide precipitates as a microsphere;
converting the uranyl peroxide microsphere to uranium dioxide microspheres.
2. The method of claim 1, where in the conversion of the uranyl peroxide microsphere
to uranium dioxide microspheres, comprises
drying the uranyl peroxide precipitate;
sintering the precipitate to produce uranium dioxide microspheres.
3. A method of production of uranium dioxide microspheres, comprising:
vaporizing uranium hexafluoride solid to produce uranium hexafluoride gas;
reacting the uranium hexafluoride gas with steam to produce uranyl fluoride and
hydrogen fluoride;
separating the uranyl fluoride and hydrogen fluoride;
quenching the uranyl fluoride;
reacting the uranyl fluoride with aqueous nitric acid to form a uranyl nitrate solution;
chilling the uranyl nitrate solution;
dispensing the uranyl nitrate solution in hydrogen peroxide whereby uranyl peroxide
precipitates as a imicrosphere microsphere;
separating the uranyl peroxide precipitate;
sintering the uranyl peroxide precipitate to produce dense uranium dioxide microspheres.
4. The method of claim 3 wherein the method further includes dissolving uranium
metal in the uranyl nitrate solution.
5. The method of claim 3 wherein the method further includes dissolving uranium
oxides in the uranyl nitrate solution.
6. The method of claim 1 wherein the uranyl fluoride solution is chilled to a
temperature between about 0° C. and about 25° C.
7. The method of claim 2 wherein the uranyl fluoride solution is chilled to a
temperature between about 0° C. and about 25° C.
8. The method of claim 2 wherein the uranyl fluoride solution is dispersed into
a peroxide solution having a concentration between about 0.5 and 50%.
9. The method of claim 3 wherein the uranyl nitrate solution is dispersed into
a peroxide solution having a concentration between about 0.5 and 50%.
10. The method of claim 1 wherein the uranyl nitrite fluoride solution is dispersed
into a peroxide solution having a temperature between about 0° C. and about
25° C.
11. The method of claim 2 wherein the uranyl fluoride solution is dispersed into
a peroxide solution having a temperature between about 0° C. and about 25° C.
12. The method of claim 3 wherein the uranyl nitrate solution is dispersed into
a peroxide solution having a temperature between about 0° C. and about 25° C.
13. The method of claim 3 wherein the peroxide for washing the uranyl peroxide
precipitate has a concentration from about 0.001 to 5 molar.
14. The method of claim 2 wherein the precipitate is dried with warm nitrogen.
15. The method of claim 3 wherein the precipitate is dried with warm nitrogen.
16. The method of claim 2 wherein the precipitate is sintered under nitrogen.
17. The method of claim 3 wherein fluorboric acid is added to the quenched solution.
18. The method of claim 3 wherein the urea is added to the quenched solution.
19. The method of claim 3 wherein aluminum nitrate is added to the quenched solution
to facilitate partial complexation of the fluoride ions.
20. The method of claim 19 wherein aluminum nitrate has a concentration from
about 0.001 to 1.25 molar.
Description
BACKGROUND OF THE INVENTION
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.
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 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
barrier 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 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
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 10 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 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
also related to the complexity associated with loading, transport, and storage
of SNF. Thus, the weight and size of containers are also of particular 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 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 ("DUF
6")
stored at gas diffusion plants. Among the various disposal options considered by
the DOE was conversion of the uranium hexafluoride to 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 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 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 (DUO
2)
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 DUO
2
pellets replace the gravel aggregate normally used in concrete. Due to the increased
density of DUO
2, 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.
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 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 (DUO
2) or depleted uranium carbide (DUC
or DUC
2). 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 (1) 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 form
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 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 limit 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.
BRIEF DESCRIPTION OF THE FIGURES AND TABLES
The present invention will be more clearly understood for a reading of the following
detailed description in conjuction with the accompanying figures.
FIGURES
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. 4.1
a 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. 4.2
a 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 UO
2F
2 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. 4.6.2
a 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;
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;
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 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
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
FIG. 6 is a process flow diagram for a peroxide gelation process 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 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. 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 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 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 wall
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
30 can be a one
inch perforated support plate constructed from materials such as steel, lead, and
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 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.
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 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 U
3O
8 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 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 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 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 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 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
1H
0-2,
with C
1H
0.5 most preferred. An advantage of this combination
is that it forms an environmentally inert shielding material. When pyrolytic uranium
dioxide is mixed with a 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 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 and/or pressed. Examples of such
resins include polytetrafluoroethylene (sold under the tradename TEFLON), polyamides,
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 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 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,
aluminium/boron 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 B
2O
3 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 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 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 throug
