U.S. patent application number 12/770178 was filed with the patent office on 2011-11-03 for methods of producing and recovering plutonium-238.
This patent application is currently assigned to BATTELLE ENERGY ALLIANCE, LLC. Invention is credited to John D. Bess, Roger N. Henry, Leigh R. Martin, David H. Meikrantz, Bruce J. Mincher, James E. Werner.
Application Number | 20110265605 12/770178 |
Document ID | / |
Family ID | 44857204 |
Filed Date | 2011-11-03 |
United States Patent
Application |
20110265605 |
Kind Code |
A1 |
Meikrantz; David H. ; et
al. |
November 3, 2011 |
METHODS OF PRODUCING AND RECOVERING PLUTONIUM-238
Abstract
Methods of producing plutonium-238 are disclosed. One method
includes dissolving neptunium-237 in a nitric acid solution to
produce a neptunium target solution, subjecting the neptunium
target solution to neutron radiation to produce plutonium-238, and
removing the plutonium-238 from the neptunium target solution. A
second method includes exposing a solution of neptunium-237 to
neutron radiation to produce plutonium-238, complexing the
plutonium-238 with an organophosphorus compound, and separating the
plutonium-238/organophosphorus compound complex from the solution
of neptunium-237. A third method includes dissolving neptunium-237
to form a neptunium-237 target solution, exposing the neptunium-237
to thermal neutrons to produce plutonium-238, utilizing an
organophosphorus compound to complex the plutonium-238 and the
organophosphorus compound, extracting the
plutonium-238/organophosphorus compound complex from the irradiated
neptunium target solution, and recovering the plutonium-238.
Inventors: |
Meikrantz; David H.; (Idaho
Falls, ID) ; Werner; James E.; (Idaho Falls, ID)
; Henry; Roger N.; (Idaho Falls, ID) ; Martin;
Leigh R.; (Ammon, ID) ; Mincher; Bruce J.;
(Idaho Falls, ID) ; Bess; John D.; (Idaho Falls,
ID) |
Assignee: |
BATTELLE ENERGY ALLIANCE,
LLC
Idaho Falls
ID
|
Family ID: |
44857204 |
Appl. No.: |
12/770178 |
Filed: |
April 29, 2010 |
Current U.S.
Class: |
75/396 |
Current CPC
Class: |
G21G 1/06 20130101; G21G
2001/0094 20130101; C22B 60/04 20130101 |
Class at
Publication: |
75/396 |
International
Class: |
C22B 60/00 20060101
C22B060/00 |
Goverment Interests
STATEMENT REGARDING FEDERALLY SPONSORED RESEARCH OR DEVELOPMENT
[0002] This invention was made with government support under
Contract Number DE-AC07-05ID14517 awarded by the United States
Department of Energy. The government has certain rights in the
invention.
Claims
1. A method of producing plutonium-238, comprising: dissolving
neptunium-237 in a nitric acid solution to produce a neptunium
target solution; subjecting the neptunium target solution to
neutron radiation to produce plutonium-238; and removing the
plutonium-238 from the irradiated neptunium target solution.
2. The method of claim 1, wherein dissolving neptunium-237 in a
nitric acid solution comprises dissolving from approximately 50 g/l
to approximately 300 g/l of neptunium-237 in the nitric acid
solution.
3. The method of claim 1, wherein subjecting the neptunium target
solution to neutron radiation comprises subjecting the neptunium
target solution to thermal neutrons.
4. The method of claim 1, wherein removing the plutonium-238 from
the irradiated neptunium target solution comprises continuously
flowing the plutonium-238 through at least one separation
device.
5. The method of claim 4, wherein continuously flowing the
plutonium-238 through at least one separation device comprises
flowing the plutonium-238 through a centrifugal contactor or an
annular centrifugal contactor.
6. The method of claim 4, wherein continuously flowing the
plutonium-238 through at least one separation device comprises
flowing the plutonium-238 through at least one extraction
chromatography column.
7. A method of producing plutonium-238, comprising: exposing a
solution of neptunium-237 to neutron radiation to produce
plutonium-238; complexing the plutonium-238 with an
organophosphorus compound; and separating the
plutonium-238/organophosphorus compound complex from the irradiated
solution of neptunium-237.
8. The method of claim 7, wherein exposing a solution of
neptunium-237 to neutron radiation to produce plutonium-238
comprises continuously exposing the solution of neptunium-237 to
neutron radiation.
9. The method of claim 7, wherein exposing a solution of
neptunium-237 to neutron radiation to produce plutonium-238
comprises flowing the solution of neptunium-237 through a liquid
loop and into a neutron source.
10. The method of claim 7, wherein separating the
plutonium-238/organophosphorus compound complex from the irradiated
solution of neptunium-237 comprises separating the
plutonium-238/organophosphorus compound complex from the irradiated
solution of neptunium-237 by liquid-liquid extraction.
11. The method of claim 7, wherein separating the
plutonium-238/organophosphorus compound complex from the irradiated
solution of neptunium-237 comprises separating the
plutonium-238/organophosphorus compound complex from the irradiated
solution of neptunium-237 by extraction chromatography.
12. The method of claim 7, further comprising recovering ions of
the plutonium-238.
13. A method of producing plutonium-238, comprising: dissolving
neptunium-237 to form a neptunium-237 target solution; exposing the
neptunium-237 to thermal neutrons to produce plutonium-238;
utilizing an organophosphorus compound with the exposed
neptunium-237 target solution to complex the plutonium-238 and the
organophosphorus compound; extracting the
plutonium-238/organophosphorus compound complex from the exposed
neptunium-237 target solution; and recovering the
plutonium-238.
14. The method of claim 13, wherein dissolving neptunium-237 to
form a neptunium-237 target solution comprises dissolving
neptunium-237 having a purity of greater than approximately 95% in
a nitric acid solution.
15. The method of claim 13, wherein utilizing an organophosphorus
compound with the exposed neptunium-237 target solution to complex
the plutonium-238 and the organophosphorus compound comprises
forming an organophosphorus extractant solution comprising the
organophosphorus compound and a diluent and combining the
organophosphorus extractant solution with the exposed neptunium-237
target solution.
16. The method of claim 13, wherein utilizing an organophosphorus
compound with the exposed neptunium-237 target solution to complex
the plutonium-238 and the organophosphorus compound comprises
binding the organophosphorus compound to a solid support and
contacting the solid support with the exposed neptunium-237 target
solution.
17. The method of claim 13, wherein extracting the
plutonium-238/organophosphorus compound complex from the exposed
neptunium-237 target solution comprises adding an organophosphorus
extractant solution comprising the organophosphorus compound and a
diluent to the exposed neptunium-237 target solution and
partitioning the plutonium-238/organophosphorus compound complex
into the organophosphorus extractant solution.
18. The method of claim 13, wherein extracting the
plutonium-238/organophosphorus compound complex from the exposed
neptunium-237 target solution comprises binding the
organophosphorus compound to a solid support and complexing the
plutonium-238 to the organophosphorus compound.
19. The method of claim 18, further comprising eluting the
plutonium-238 from the organophosphorus compound bound to the solid
support.
20. The method of claim 13, further comprising exposing the
extracted neptunium-237 target solution to additional thermal
neutrons to produce additional plutonium-238.
Description
CROSS-REFERENCE TO RELATED APPLICATION
[0001] This application is related to co-pending U.S. patent
application Ser. No. 12/468,679 to Meikrantz et al., entitled
"METHODS OF PRODUCING CESIUM-131," filed on May 19, 2009, the
disclosure of which is incorporated by reference herein in its
entirety.
TECHNICAL FIELD
[0003] The invention, in various embodiments, relates generally to
methods of producing and recovering radioisotopes. More
specifically, the invention, in various embodiments, relates to
methods of producing and recovering plutonium-238 (Pu-238).
BACKGROUND
[0004] Pu-238 is used to generate electrical power, such as in
radioisotope thermoelectric generators and radioisotope heater
units, in situations where solar energy or internal power sources
can not be used. Since Pu-238 is an alpha emitter, the Pu-238
produces high energy radiation with low penetration. One use of
Pu-238 is in batteries used in space missions. The batteries
convert heat from the decaying Pu-238 into electricity. However,
the supply of Pu-238 is decreasing because there has been no
domestic production in over twenty years, and foreign supplies of
Pu-238 are limited and will be consumed in the near future. No
alternative production processes are currently available.
[0005] In one conventional process of producing Pu-238, a solid
neptunium-237 (Np-237) target is irradiated with neutrons,
producing neptunium-238 (Np-238), which decays to Pu-238 by
.beta.-decay with a half life of 2.12 days. The Pu-238 is recovered
by chemical extraction. This process has limited efficiency
because, during solid target irradiation times, the Pu-238 can
capture one or two neutrons, forming plutonium-239 (Pu-239) or
plutonium-240 (Pu-240), which decrease the purity of the Pu-238.
The purity of the Pu-238 produced by this process is approximately
72%. The solid form of Np-237 utilized as the target is required to
be manually loaded by personnel into a nuclear reactor for
irradiation. The irradiated solid target is then manually removed
from the nuclear reactor for isolation and purification of the
Pu-238. However, the preparation, loading, unloading, target
dissolution, and plutonium separation from the Np-237 target is
time consuming, costly, and exposes personnel to radiation.
[0006] It would be desirable to be able to produce and recover high
purity Pu-238 at an improved efficiency. It would also be desirable
to eliminate the time, cost, and hazards to personnel associated
with using a solid Np-237 target.
BRIEF SUMMARY
[0007] An embodiment of the present invention includes a method of
producing plutonium-238 that comprises dissolving neptunium-237 in
a nitric acid solution to produce a neptunium target solution,
subjecting the neptunium target solution to neutron radiation to
produce plutonium-238, and removing the plutonium-238 from the
irradiated neptunium target solution.
[0008] Another embodiment of the present invention includes a
method of producing plutonium-238 that comprises exposing a
solution of neptunium-237 to neutron radiation to produce
plutonium-238, complexing the plutonium-238 with an
organophosphorus compound, and separating the
plutonium-238/organophosphorus compound complex from the irradiated
solution of neptunium-237.
[0009] Yet another embodiment of the present invention includes a
method of producing plutonium-238 that comprises dissolving
neptunium-237 to form a neptunium-237 target solution, exposing the
neptunium-237 to thermal neutrons to produce plutonium-238,
utilizing an organophosphorus compound with the irradiated
neptunium-237 target solution to complex the plutonium-238 and the
organophosphorus compound, extracting the
plutonium-238/organophosphorus compound complex from the irradiated
neptunium target solution, and recovering the plutonium-238.
BRIEF DESCRIPTION OF THE SEVERAL VIEWS OF THE DRAWINGS
[0010] While the specification concludes with claims particularly
pointing out and distinctly claiming that which is regarded as the
present invention, the advantages of this invention may be more
readily ascertained from the following description of the invention
when read in conjunction with the accompanying drawings in
which:
[0011] FIGS. 1-3 are schematic illustrations of isotope production
systems and methods of producing and recovering Pu-238 according to
embodiments of the invention.
DETAILED DESCRIPTION
[0012] Methods of producing and recovering Pu-238 are disclosed.
The Pu-238 is produced by neutron capture on Np-237. The Np-237 is
dissolved to produce a neptunium target solution containing the
Np-237. The neptunium target solution is continuously circulated
through a neutron field and radiated with neutrons to produce
Np-238, which decays to Pu-238. A liquid loop is used to circulate
the neptunium target solution through the neutron field. As used
herein, the term "liquid loop" means and includes means for
transporting or circulating the neptunium target solution, an
irradiated neptunium target solution, or an extracted neptunium
target solution throughout the neutron field. The Pu-238 is
selectively removed from the irradiated neptunium target solution
using an organophosphorus extractant and subsequently recovered.
The methods of the present invention are cost effective and provide
Pu-238 at a higher purity than conventional processes of producing
Pu-238. By utilizing a neptunium target solution, the methods of
the present invention also eliminates the time and cost associated
with preparing a solid neptunium target. In addition, the methods
produce fewer undesirable radioactive byproducts, such as Pu-239.
Additional advantages of the methods of the present invention
include the ability to bring the methods online years faster than
conventional processes and at a cost of approximately one-fourth of
the capital costs and one-tenth the operating costs than a solid
target based process.
[0013] As used herein, the terms "comprising," "including,"
"containing," "characterized by," and grammatical equivalents
thereof are inclusive or open-ended terms that do not exclude
additional, unrecited elements or method steps, but also include
the more restrictive terms "consisting of" and "consisting
essentially of" and grammatical equivalents thereof. As used
herein, the term "may" with respect to a material, structure,
feature or method act indicates that such is contemplated for use
in implementation of an embodiment of the invention and such term
is used in preference to the more restrictive term "is" so as to
avoid any implication that other, compatible materials, structures,
features and methods usable in combination therewith should or must
be, excluded.
[0014] The Np-237 used in the methods of the present invention may
be of high purity, such as a purity of greater than approximately
95%. The Np-237 may include Np-237 recovered from irradiated
uranium fuel, such as by the Plutonium/Uranium Extraction (PUREX)
process, which is stored at Department of Energy sites in solid
form and may be present in large quantities as a result of nuclear
reactors and plutonium and tritium production processes. Before use
in the methods of the present invention, the Np-237 may be
dissolved and purified to remove daughter isotopes. Once purified,
the Np-237 may be dissolved in a nitric acid solution to form the
neptunium target solution. Initially, the neptunium target solution
may be non-irradiated. As used herein, the term "non-irradiated"
means and includes Np-237 that has not been substantially exposed
to neutron radiation. The neptunium target solution may include
Np-237 at a concentration of between approximately 50 g/L and
approximately 300 g/L in the nitric acid solution. The nitric acid
solution in which the Np-237 is dissolved may be an aqueous
solution having a nitric acid concentration of from approximately 1
M to approximately 5 M, such as from approximately 2 M to
approximately 3 M.
[0015] After dissolving the Np-237 in the nitric acid solution, the
neptunium target solution may be subjected to neutron radiation. To
irradiate the Np-237, the neptunium target solution 2 may be
introduced into an isotope production system 4, as shown in FIGS.
1-3. The illustrations presented herein are not meant to be actual
views of any particular isotope production system, but are merely
idealized representations which are employed to describe the
present invention. Additionally, elements common between figures
may retain the same numerical designation. In one embodiment, shown
in FIGS. 1 and 2, the isotope production system 4 includes an inlet
(not shown) through which the neptunium target solution 2 is
introduced, a neutron source 6, a liquid loop 8, a separator 10,
and an outlet through which a Pu-238/extractant complex solution 12
or Pu-238 solution 16 exits the isotope production system 4. In
FIGS. 1 and 2, the separator 10 is at least one centrifugal
contactor or ACC. In another embodiment, shown in FIG. 3, the
isotope production system 4 includes the inlet, neptunium target
solution 2, the neutron source 6, the liquid loop 8, and the
separator 10, which is at least one extraction chromatography
column. In the embodiment of FIG. 3, the Pu-238 is recovered from
the separator 10 as described in more detail below.
[0016] The neptunium target solution 2, irradiated neptunium target
solution 2', or extracted neptunium target solution 2'' may be
transported or circulated in the liquid loop 8 throughout the
isotope production system 4. The structure of the liquid loop 8 may
be formed from a material that is configured to contain the
neptunium target solution 2, irradiated neptunium target solution
2', and extracted neptunium target solution 2'', and is
substantially non-reactive with the components of the neptunium
target solution 2, irradiated neptunium target solution 2', and
extracted neptunium target solution 2''. By way of non-limiting
example, components of the liquid loop 8 may be formed from
stainless steel, aluminum, zirconium, or other corrosion-resistant,
neutron-transparent alloy or material. The liquid loop 8 may be
configured as tubing or a plate, which provides an increased
surface area. To assist in circulating the neptunium target
solution 2, irradiated neptunium target solution 2', and extracted
neptunium target solution 2'', the isotope production system 4 may
include additional, optional features, such as a pump (not shown).
The pump, if present, may be a conventional device that is
configured to transport the neptunium target solution 2, irradiated
neptunium target solution 2', and extracted neptunium target
solution 2'' through the liquid loop 8 and neutron source 6. The
pump may also circulate the irradiated neptunium target solution 2'
to the separator 10. The pump may be formed from a material
compatible with the neptunium target solution 2, irradiated
neptunium target solution 2', and extracted neptunium target
solution 2''. The isotope production system 4 may also include a
heat exchanger (not shown) if heating or cooling of the neptunium
target solution 2, irradiated neptunium target solution 2', or
extracted neptunium target solution 2'' is desired. The heat
exchanger, if present, may be a conventional device that transfers
heat away from the neptunium target solution 2, irradiated
neptunium target solution 2', or extracted neptunium target
solution 2''. The isotope production system 4 may also include
openings (not shown), such as vents, to enable offgasing of
byproducts. These additional features of the isotope production
system 4 are not shown in FIGS. 1 and 2 for simplicity and
clarity.
[0017] The neptunium target solution 2 and extracted neptunium
target solution 2'' may be subjected to neutron radiation by
continuously flowing the neptunium target solution 2 or extracted
neptunium target solution 2'' through the neutron source 6. The
neutron source 6 may be a device configured to produce neutron
radiation, such as a nuclear reactor. The neutron source 6 may be
configured to produce thermal neutrons or fast neutrons for
irradiating the neptunium target solution 2 or extracted neptunium
target solution 2''. In one embodiment, the neptunium target
solution 2 or extracted neptunium target solution 2'' is exposed to
thermal neutrons. The neutron source 6 may be a conventional
nuclear reactor configured to produce neutrons and continuously
irradiating the neptunium target solution 2 or the extracted
neptunium target solution 2'' with neutrons. By way of non-limiting
example, the nuclear reactor may be a pool-type reactor including,
but not limited to, a TRIGA reactor. The nuclear reactor may be of
a sufficient size to produce the desired amount of Pu-238 from a
neutron capture process utilizing Np-237. Up to approximately 300
g/L of the neptunium target solution 2 may be circulated through
the neutron source 6 to provide the desired amount of Pu-238.
Neutron sources 6 configured to irradiate the neptunium target
solution 2 and extracted neptunium target solution 2'' are
conventional and, therefore, specific details of its design and
configuration are not described or illustrated herein. The neutron
source 6 may, optionally, be located in proximity to a Department
of Energy facility capable of handling Pu-238 so that recovered
Pu-238 may be easily and quickly transported for additional
processing.
[0018] As the neptunium target solution 2 or extracted neptunium
target solution 2'' flows through the isotope production system 4,
a portion of the neptunium target solution 2 or extracted neptunium
target solution 2'' may enter the neutron source 6 at a given time
and be radiated with thermal neutrons or fast neutrons, forming the
irradiated neptunium target solution 2'. Since the neptunium target
solution 2 or neptunium target solution 2'' continuously circulates
throughout the isotope production system 4, the entire volume of
the neptunium target solution 2 or extracted neptunium target
solution 2'' may, over time, be radiated with neutrons. Even though
only a portion of the neptunium target solution 2 or extracted
neptunium target solution 2'' passes through the neutron source 6
at a given time, for simplicity, the process is described herein as
applying to the neptunium target solution 2 or extracted neptunium
target solution 2''. As the neptunium target solution 2 or
extracted neptunium target solution 2'' passes through the neutron
source 6, the neptunium target solution 2 or extracted neptunium
target solution 2'' may be exposed to radiation of a sufficient
energy for the Np-237 to capture neutrons, forming Np-238, which
decays to Pu-238. Neptunium has a sufficient neutron capture
cross-section to capture a neutron and convert Np-237 to Np-238.
The energy conditions utilized for radiating the neptunium target
solution 2 or extracted neptunium target solution 2'' are
conventional and, therefore, are not described in detail herein. By
way of non-limiting example, the neptunium target solution 2 or
extracted neptunium target solution 2'' may be exposed to thermal
neutrons having a mean energy of approximately 0.025 eV and a
velocity of approximately 2200 m/s. The neptunium target solution 2
may be flowed through desired locations of the neutron source 6 to
efficiently radiate the Np-237. By way of example, the neptunium
target solution 2 may be circulated through G-hole positions in the
nuclear reactor.
[0019] Once irradiated, the irradiated neptunium target solution 2'
may be circulated by way of the liquid loop 8 to the separator 10
of the isotope production system 4. The neutron source 6 may be
close-coupled to the separator 10. Since the irradiated neptunium
target solution 2' is not manually transported from the neutron
source 6 for separation and purification of the Pu-238, exposure of
personnel to radiation is greatly reduced in comparison to
conventional techniques for producing Pu-238. The irradiated
neptunium target solution 2' may include Np-237, Np-238, and
Pu-238, and minor amounts of other radioisotopes, such as Np-239 or
Pu-239, depending on the amount of time that has elapsed since the
irradiation of the neptunium target solution 2 or extracted
neptunium target solution 2''. As the irradiated neptunium target
solution 2' is continuously circulated through the liquid loop 8,
the Np-237 in the irradiated neptunium target solution 2' is
converted to Np-238, which subsequently decays to Pu-238 with a
half-life of 2.12 days.
[0020] The rate of production of the Pu-238 may depend on the
initial concentration of Np-237 in the neptunium target solution 2,
the neutron fluence, the neutron capture cross section of
neptunium, and the radiation time. Immediately before irradiation
of a fresh volume of the neptunium target solution 2, which
contains Np-237, limited radioactivity may be present in the
neptunium target solution 2. After irradiation, the radioactivity
in the irradiated neptunium target solution 2' may be substantially
due to Np-238. In addition, trace amounts of other radioisotopes
may be present. As the Np-238 decays, Pu-238 may begin to appear in
the irradiated neptunium target solution 2'. Therefore, the
irradiated neptunium target solution 2' may include the radioactive
isotopes Np-237, Np-238, and Pu-238 before the Pu-238 is
selectively removed in the separator 10.
[0021] As the Pu-238 begins to accumulate in the irradiated
neptunium target solution 2', the Pu-238 may be continuously
removed using the separator 10. The separator 10 may include at
least one separation device 10A that utilizes an organophosphorus
compound as an extractant to selectively remove the Pu-238. The
Pu-238 is removed relative to the Np-237 and Np-238, which remain
in the irradiated neptunium target solution 2'. The
organophosphorus compound may form a complex with the Pu-238,
enabling its selective removal from the irradiated neptunium target
solution 2'. The Pu-238 in the irradiated neptunium target solution
2' is set at a valence state of +4 while the Np-237 and Np-238 are
set at a valence state of +5 and, therefore, the Pu-238 may
selectively coordinate or complex with the organophosphorus
compound utilizing conventional PUREX process chemistry. The
chemistry in the liquid loop 8 may be adjusted to maintain the
valences stated above using oxidation/reduction methods and
reagents, which are known in the art and, therefore, are not
described in detail herein.
[0022] The separation device 10A may be a device configured to
conduct a liquid-liquid extraction using the organophosphorus
compound as the extractant, as shown in FIGS. 1 and 2, or an
extraction chromatography device configured to use the
organophosphorus compound as a stationary phase, as shown in FIG.
3. By way of non-limiting example, the separation devices 10A, 10B,
10C may be liquid-liquid extraction devices, such as centrifugal
contactors or annular centrifugal contactors ("ACC"), or extraction
chromatography columns. Examples of ACCs include those described in
U.S. Pat. Nos. 5,571,070, 5,591,340, and 7,157,061 to Meikrantz et
al. and U.S. Pat. No. 4,959,158 to Meikrantz, the disclosure of
each of which is incorporated by reference herein in its entirety.
ACCs are commercially available, such as from CINC Industries Inc.
(Carson City, Nev.), and provide a high throughput method of
performing the liquid-liquid extraction. The Pu-238 may be
continuously separated from the Np-237 and Np-238 by flowing the
irradiated neptunium target solution 2' through the separation
devices 10A, 10B, 10C of the separator 10.
[0023] If the separation devices 10A, 10B, 10C are liquid-liquid
extraction devices, such as ACCs, the irradiated neptunium target
solution 2' may be contacted with an organophosphorus extractant
solution 14 that includes the organophosphorus compound. The
components of the organophosphorus extractant solution are
described in detail below. The organophosphorus extractant solution
14 may function in the liquid-liquid extraction device as an
organic phase, while the irradiated neptunium target solution 2'
may function as an aqueous phase. When the irradiated neptunium
target solution 2' and the organophosphorus extractant solution 14
are contacted and agitated with one another, the Pu-238 may
partition into the organophosphorus extractant solution 14 (organic
phase), while the Np-237 and Np-238 remain in the irradiated
neptunium target solution 2' (aqueous phase). The Pu-238 is removed
or forward extracted from the irradiated neptunium target solution
2' and into the organophosphorus extractant solution 14.
[0024] The extractant in the organophosphorus extractant solution
14 may be an organophosphorus compound that is configured to
selectively bind to the Pu-238 and form a Pu-238/extractant
complex. The organophosphorus compound may be selected to optimize
Pu-238 recovery, minimize Np-237 extraction, and provide required
overall process efficiency. The organophosphorus compound may
selectively bind to the Pu-238 due to the difference in oxidation
state between Pu-238, which is in the +4 oxidation state, and
Np-237, which is in the +5 oxidation state. The organophosphorus
compound may have the following general chemical structure:
##STR00001##
where each of X.sub.1-X.sub.3 is independently selected from an
alkyl group, an aryl group, an alkoxy group, an aryloxy group, or
combinations thereof. The X.sub.1-X.sub.3 groups may also include
at least one heteroatom. As used herein, the term "alkyl" used by
itself or as part of another term means and includes a saturated or
unsaturated, branched, straight-chain, or cyclic monovalent
hydrocarbon radical derived by the removal of one hydrogen atom
from a single carbon atom of a parent alkane, alkene or alkyne. The
alkyl group may include from one carbon atom to twelve carbon
atoms. By way of example, the alkyl group may include, but is not
limited to, methyl; ethyl, such as ethanyl, ethenyl, ethynyl;
propyl, such as propan-1-yl, propan-2-yl, cyclopropan-1-yl,
prop-1-en-1-yl, prop-1-en-2-yl, prop-2-en-1-yl (allyl),
cycloprop-1-en-1-yl, cycloprop-2-en-1-yl, prop-1-yn-1-yl,
prop-2-yn-1-yl; butyl, such as butan-1-yl, butan-2-yl,
2-methyl-propan-1-yl, 2-methyl-propan-2-yl, cyclobutan-1-yl,
but-1-en-1-yl, but-1-en-2-yl, 2-methyl-prop-1-en-1-yl,
but-2-en-1-yl, but-2-en-2-yl, buta-1,3-dien-1-yl,
buta-1,3-dien-2-yl, cyclobut-1-en-1-yl, cyclobut-1-en-3-yl,
cyclobuta-1,3-dien-1-yl, but-1-yn-1-yl, but-1-yn-3-yl,
but-3-yn-1-yl, etc. As used herein, the term "aryl" used by itself
or as part of another term means and includes a monovalent aromatic
hydrocarbon group derived by the removal of one hydrogen atom from
a single carbon atom of a parent aromatic ring system. The aryl
group may include from one carbon atom to twenty carbon atoms. By
way of example, the aryl group may include, but is not limited to,
a group derived from aceanthrylene, acenaphthylene,
acephenanthrylene, anthracene, azulene, benzene, chrysene,
coronene, fluoranthene, fluorene, hexacene, hexaphene, hexylene,
as-indacene, s-indacene, indane, indene, naphthalene, octacene,
octaphene, octalene, ovalene, penta-2,4-diene, pentacene,
pentalene, pentaphene, perylene, phenalene, phenanthrene, picene,
pleiadene, pyrene, pyranthrene, rubicene, triphenylene,
trinaphthalene, etc. The organophosphorus compound may be a
phosphate compound, a phosphonate compound, a phosphinate compound,
or a phosphine oxide compound, such as a trialkyl phosphine
oxide.
[0025] By way of example, the organophosphorus compound may be
tri-n-butyl phosphate (TBP), tri-n-hexyl phosphate, tri-isoamyl
phosphate (TAP), diethyl butyl phosphate, diethyl isobutyl
phosphate, diethyl amyl phosphate, diethyl decyl phosphate, dibutyl
methyl phosphate, dibutyl ethyl phosphate, tributyl phosphate,
dibutyl hexyl phosphate, dibutyl octyl phosphate, dibutyl decyl
phosphate, butyl octyl phenyl phosphate, butyl diphenyl phosphate,
dibutyl ethoxybutyl phosphate, tri-butoxy ethyl phosphate,
tri-.beta.-chloroethyl phosphate, diethyl methyl phosphonate,
diethyl butyl phosphonate, diethyl hexyl phosphonate, diethyl octyl
phosphonate, diethyl hexadecyl phosphonate, diethyl phenyl
phosphonate, dibutyl methyl phosphonate, dibutyl ethyl phosphonate,
dibutyl butyl phosphonate, dibutyl hexyl phosphonate, dibutyl
isoctyl phosphonate, dibutyl decyl phosphonate, dibutyl hexadecyl
phosphonate, dibutyl phenyl phosphonate, dioctyl phenyl
phosphonate, diisooctyl phenyl phosphonate, dibutyl hydrogen
phosphonate, diallyl phenyl phosphonate, dioctyl styrene
phosphonate, diethyl trichloromethyl phosphonate, diethyl benzoyl
phosphonate, diamylamyl phosphonate (DAAP), di-butyl
N,N-diethylcarbamoyl methyl phosphonate, methyl dibutyl
phosphinate, ethyl dibutyl phosphinate, butyl dibutyl phosphinate,
ethyl dihexyl phosphinate, ethyl dichlorophenyl phosphinate,
tributyl phosphine oxide, dioctyl hydrogen phosphine oxide,
tri-n-octyl phosphine oxide, triphenyl phosphine oxide, or
combinations thereof.
[0026] The organophosphorus extractant solution 14 may include the
organophosphorus compound dissolved in a diluent. The diluent may
be an inert diluent, such as a straight chain hydrocarbon diluent.
For instance, the diluent may be an isoparaffinic hydrocarbon
diluent, such as Isopar.RTM. L or Isopar.RTM. M. Isopar.RTM. L
includes a mixture of C.sub.10-C.sub.12 isoparaffinic hydrocarbons
and is available from Exxon Chemical Co. (Houston, Tex.).
Isopar.RTM. M includes a mixture of C.sub.12-C.sub.15 isoparaffinic
hydrocarbons and is available from Exxon Chemical Co. (Houston,
Tex.). The organophosphorus compound may be present in the
organophosphorus extractant solution at a concentration of from
approximately 0.0025 M to approximately 1.0 M. The organophosphorus
extractant solution 14 may be prepared by combining the
organophosphorus compound with the diluent to form a mixture.
[0027] After extracting the Pu-238, the irradiated neptunium target
solution 2' (aqueous phase) may be substantially depleted of
Pu-238, while the organophosphorus extractant solution 14 includes
substantially all of the Pu-238. The irradiated neptunium target
solution 2' (aqueous phase) and the organophosphorus extractant
solution 14 (organic phase) containing the Pu-238 may then be
separated from one another. Since the organophosphorus extractant
solution 14 includes one predominant isotope, Pu-238, minimal
recovery and purification acts are used to recover the Pu-238
compared to conventional Pu-238 processes. As shown in FIG. 1, the
organophosphorus extractant solution 14 containing the Pu-238 may
be removed from the separation device 10A once sufficient
radioactivity has accumulated, and is referred to herein as
Pu-238/extractant complex solution 12. To increase the amount of
Pu-238 removed from the irradiated neptunium target solution 2',
the irradiated neptunium target solution 2' from separation device
10A may be passed through separation devices 10B, 10C in which
additional liquid-liquid extractions are conducted using fresh
volumes of organophosphorus extractant solution 14. The
Pu-238/extractant complex solutions 12 that exit the separation
devices 10A, 10B, 10C may be collected and combined and the Pu-238
recovered by conventional techniques, such as by using a dilute
nitric acid solution to remove or strip the Pu-238. The
organophosphorus extractant solution 14 lacking the Pu-238 may be
reused in subsequent extractions.
[0028] The Pu-238/extractant complex solution 12 from the
liquid-liquid extraction may be further processed outside the
separator 10 to recover the Pu-238 in the form of Pu-238 ions, such
as Pu-238.sup.+4. The Pu-238 may be removed or stripped from the
Pu-238/extractant complex solution 12 by adjusting the pH of the
Pu-238/extractant complex solution 12 with the dilute nitric acid
solution. The dilute nitric acid solution may have from
approximately 0.001 M HNO.sub.3 to approximately 0.5 M HNO.sub.3,
such as approximately 0.01 M HNO.sub.3. The Pu-238/extractant
complex solution 12 and the dilute nitric acid solution may be
contacted and agitated such that the Pu-238 partitions from the
Pu-238/extractant complex solution 12 and into the dilute nitric
acid solution. The Pu-238/extractant complex solution 12, which is
now depleted of Pu-238, and the dilute nitric acid solution, which
now contains the Pu-238, may then be separated by conventional
liquid-liquid separation techniques. While Pu-238 is being stripped
from the Pu-238/extractant complex solution 12, a fresh volume of
the organophosphorus extractant solution 14 may be contacted with
the irradiated neptunium target solution 2' in the isotope
production system 4 to provide a continuous process for recovering
the Pu-238.
[0029] Alternatively, as shown in FIG. 2, one of the additional
separation devices (i.e., 10B, 10C) may be utilized to remove or
strip the Pu-238 from the organophosphorus extractant solution 14'
containing the Pu-238 by conventional techniques, such as by
contacting the organophosphorus extractant solution 14' with a
nitric acid solution 16. The nitric acid solution 16 used to strip
the Pu-238 may have a nitric acid concentration of from
approximately 0.01 M to approximately 0.2 M. After the Pu-238
partitions into the nitric acid solution 16, the organophosphorus
extractant solution 14 may be recycled and reused to extract Pu-238
on a continuous basis within the separator 10. The nitric acid
solution 16, which contains the Pu-238, may exit the separation
device 10C as Pu-238 solution 18 once sufficient radioactivity has
accumulated.
[0030] The Pu-238/extractant complex solution 12 or Pu-238 solution
18 may be periodically removed from the separation devices 10A,
10B, 10C, such as hourly, daily, weekly, or monthly. Since
plutonium has a sufficient capture cross-section to capture a
neutron and become Pu-239, the Pu-238 may be removed from the
irradiated neptunium target solution 2' to minimize the formation
of Pu-239. By continuously removing the Pu-238, the Pu-238 is no
longer exposed to neutrons, which substantially prevents the
production of Pu-239. Therefore, continuously removing the Pu-238
from the irradiated neptunium target solution 2' may maximize the
Pu-238 recovery rate and purity.
[0031] The irradiated neptunium target solution 2' including the
Pu-238 may be continuously passed through the liquid-liquid
extraction device to continuously remove the Pu-238 as it is
produced. The Pu-238 may be continuously removed from the
irradiated neptunium target solution 2' by continuously contacting
the irradiated neptunium target solution 2' with the
organophosphorus extractant solution 14, enabling the Pu-238 to
distribute into the organophosphorus extractant solution 14. Once
desired levels of Pu-238 are achieved in the separator 10, the
Pu-238/extractant complex solution 12 or Pu-238 solution 18 may be
removed from the separator 10 and further purified, if desired.
Additional purification of the Pu-238 from the Pu-238/extractant
complex solution 12 or Pu-238 solution 18 may be conducted outside
the separator 10, such as by passing the Pu-238/extractant complex
solution 12 or Pu-238 solution 18 through extraction chromatography
columns or ion exchange columns. The Pu-238 may then be
concentrated by evaporation or precipitation.
[0032] After the irradiated neptunium target solution 2' and
organophosphorus extractant solution containing the Pu-238 are
separated in the liquid-liquid extraction device, the extracted
neptunium target solution 2'', which lacks the Pu-238, may be
circulated through the isotope production system 4 for an amount of
time sufficient for Np-237 remaining in the extracted neptunium
target solution 2'' to be activated to Np-238 and for additional
Pu-238 to grow in. Since neptunium is expensive, the extracted
neptunium target solution 2'', which contains Np-237, may be
recirculated through the isotope production system 4. By
continuously reusing the extracted neptunium target solution 2'',
the methods of the present invention may have less loss of Np-237
than conventional processes using solid neptunium targets. Since
the Np-237 continuously circulates through the liquid loop 8 and is
never removed, the loss of Np-237 with the methods of the present
invention is minimal. The extracted neptunium target solution 2''
may be flowed through the neutron source 6 and exposed to neutron
radiation, producing the irradiated neptunium target solution 2'.
As the Pu-238 accumulates in the irradiated neptunium target
solution 2' and is continuously removed by the separator 10, as
described above, the Pu-238/extractant complex solution 12 or
Pu-238 solution 18 may be removed outside the continuous process
system for further purification whenever the radioactive quantity
desired is reached. As Np-237 is depleted from the neptunium target
solution 2, additional Np-237 may be introduced into the isotope
production system 4, as indicated by reference numeral 2 shown
entering the separator 10 of the isotope production system 4, to
produce additional Pu-238. By way of non-limiting example,
additional Np-237 may be dissolved into the extracted neptunium
target solution 2'' and circulated through the isotope production
system 4.
[0033] As shown in FIG. 3, the separation devices 10A, 10B, 10C may
also be extraction chromatography columns that contain the
organophosphorus compound coated on a solid support. For instance,
the organophosphorus compound may be used as a stationary phase in
an extraction chromatography column. The solid support may be
silica or an organic polymer. Extraction chromatography columns and
techniques for immobilizing the organophosphorus compound on the
solid support are known in the art and, therefore, are not
described in detail herein. As the irradiated neptunium target
solution 2' passes through the extraction chromatography column,
the Pu-238 may come into contact with the organophosphorus compound
and form a complex with the organophosphorus compound. Since the
Pu-238 is to be continuously removed, the irradiated neptunium
target solution 2' may be continuously flowed through extraction
chromatography columns that function as separation devices 10A,
10B, 10C. The extracted neptunium target solution 2'' that exits
the extraction chromatography column is substantially depleted of
the Pu-238. If, however, additional Pu-238 remains in the extracted
neptunium target solution 2'', the extracted neptunium target
solution 2'' may be flowed through additional extraction
chromatography columns as desired.
[0034] The extracted neptunium target solution 2'', which lacks the
Pu-238, may be circulated through the isotope production system 4
for an amount of time sufficient for Np-237 remaining in the
extracted neptunium target solution 2'' to be activated to Np-238
and for additional Pu-238 to grow. The extracted neptunium target
solution 2'' may be flowed through the neutron source 6 and exposed
to neutron radiation, producing the irradiated neptunium target
solution 2'. As the Pu-238 accumulates in the irradiated neptunium
target solution 2' and is passed through the extraction
chromatography columns, the Pu-238 complexes with the
organophosphorus compound in the extraction chromatography columns.
The Pu-238 may be eluted from the extraction chromatography columns
using an aqueous acid solution as a mobile phase. To reduce
exposure of personnel to radiation, the extraction chromatography
columns having the Pu-238 complexed to the organophosphorus
compound may be removed from the separator 10 and transported to a
different location for elution of the Pu-238. For instance, if the
separation devices 10A, 10B, 10C are extraction chromatography
columns, the extraction chromatography columns may be removed from
the isotope production system 4 before eluting the Pu-238. By way
of example, after a sufficient amount of Pu-238 has complexed to
separation device 10A, the separation device 10A may be removed
from the separator 10 and the Pu-238 eluted therefrom. Separation
devices 10B and 10C may be moved into locations where separation
devices 10A and 10B were previously positioned, and a new
extraction chromatography column may be located in the separator 10
where separation device 10C was previously positioned. The aqueous
acid solution used to elute the Pu-238 may be an aqueous nitric
acid solution having from approximately 0.001 M HNO.sub.3 to
approximately 6 M HNO.sub.3, such as from 0.001 M HNO.sub.3 to
approximately 0.5 M HNO.sub.3. By way of non-limiting example, the
aqueous acid solution may have approximately 0.01 M HNO.sub.3. The
aqueous acid solutions exiting the extraction chromatography
columns include substantially all of the Pu-238 and may be
collected and combined. The aqueous acid solution containing the
Pu-238 may be concentrated, such as by evaporating the aqueous acid
solution or precipitating the Pu-238. The aqueous acid solution
containing the Pu-238 may also be further purified and recovered by
subjecting the aqueous acid solution to filtration, ion exchange
chromatography, extraction chromatography, or other conventional
techniques.
[0035] In addition, combinations of ACCs and extraction
chromatography columns may be used as the separation devices 10A,
10B, 10C. While three separation devices 10A, 10B, 10C are
illustrated in FIGS. 1-3, the separator 10 may include more than
three or less than three separation devices 10A, 10B, 10C depending
on the desired purity of the Pu-238. For instance, the irradiated
neptunium target solution 2' may be flowed through one or two
separation devices 10A, 10B. To protect personnel from the
radiation emitted by the irradiated neptunium target solution 2',
the separation devices 10A, 10B, 10C may be enclosed in a
containment device, such as a glove box and/or shielded cell.
[0036] Since only a portion of Np-237 decays to Pu-238 at a given
time, the Pu-238/extractant complex solution 12, the Pu-238
solution 18, or the aqueous acid solution containing the Pu-238 as
Pu-238.sup.+4 ions may be a dilute solution of Pu-238. Since the
pure Pu-238 is produced and recovered in small amounts, the
irradiation of the neptunium target solution 2 and the recovery of
the Pu-238 may be conducted under minimal security requirements. If
a concentrated form of Pu-238 is desired, at least a portion of the
aqueous acid solution or nitric acid solution may be removed. For
instance, the Pu-238 solution 18 or the aqueous acid solution
containing the Pu-238 may be concentrated by evaporating the
aqueous acid solution or nitric acid solution. The aqueous acid
solution containing the Pu-238 or nitric acid solution containing
the Pu-238 or solid Pu-238 may be transported to a suitable
facility, such as a Department of Energy facility, and stored or
used in radioisotope thermoelectric generators and radioisotope
heater units. Radioisotope thermoelectric generators and
radioisotope heater units containing the Pu-238 may be formed by
conventional techniques, which are not described in detail herein.
The methods of the present invention may be used to produce Pu-238
at an amount of approximately one kg/year using Np-237 as the
starting material. Calculations of the Pu-238 yearly production
rate indicate that from approximately 50 g to approximately 90 g
are formed per MW of reactor power using a liquid loop 8 irradiated
in only the peripheral positions utilizing Np-237 at a
concentration of from approximately 100 g/L to approximately 300
g/L. Thus a 10 MW reactor may be used to provide from approximately
500 g to approximately 900 g of Pu-238 per year by these methods.
Increased production may be possible using more central irradiation
positions in TRIGA type reactors having from approximately 10 MW to
approximately 15 MW of power.
[0037] Since Pu-238 is the only isotope to be removed by the
separation devices 10A, 10B, 10C (liquid-liquid extraction device
or extraction chromatography column), Pu-238 having higher purity
may be achieved by the methods of the present invention compared to
conventional techniques. By way of non-limiting example, the purity
of the Pu-238 at the time of production and separation/recovery may
be greater than or equal to approximately 80%, such as greater than
or equal to approximately 99%, with the remainder including Pu-239
or Pu-240. In addition, since the Pu-238 is continuously removed
from the irradiated neptunium target solution 2' before subsequent
neutron capture can occur, the resulting Pu-238 is substantially
free of Pu-239. The methods of the present invention of producing
Pu-238 also provide isolating Pu-238 with fewer separation
acts.
[0038] The Pu-238 may also be intermittently removed from the
isotope production system 4, such as if the isotope production
system 4 is taken offline for maintenance. Before restarting the
neutron source 6 after the isotope production system 4 has been
offline, any Pu-238 that has accumulated in the irradiated
neptunium target solution 2' and extracted neptunium target
solution 2'' may be removed, as described above, to prevent the
formation of Pu-239. After removing the Pu-238, the isotope
production system 4 may be put back online.
[0039] By utilizing a liquid target of Np-237, the irradiated
neptunium target solution 2' may be continuously circulated through
the isotope production system 4 until substantially all of the
Np-237 is depleted and has been converted to the recovered Pu-238.
In contrast, conventional processes of producing Pu-238 use a solid
neptunium target, which leads to incomplete use of available
Np-237. In addition, since the irradiated neptunium target solution
2' is a liquid, the irradiated neptunium target solution 2' may be
easily transported between the neutron source 6 and the separator
10 by way of the liquid loop 8, with minimal exposure of personnel
to radiation. The irradiated neptunium target solution 2' may be
circulated through a single system, the isotope production system
4, to achieve both radiation of the Np-237 and separation of the
Pu-238. This is in contrast to conventional processes of producing
Pu-238 where the neptunium target is a solid material that is
manually loaded into the nuclear reactor for radiation. The
irradiated solid target is then manually removed from the nuclear
reactor for isolation and purification of the Pu-238. However, the
loading and unloading of the solid target is time consuming,
costly, and exposes personnel to radiation.
[0040] A preliminary model of a TRIGA core has been developed for
use in Monte Carlo N-Particle Transport Code (MCNP5). MCNP5
software is known in the art and is not described in detail herein.
The design of the model is based upon a 1 MW nuclear reactor that
utilizes U--Zr--H fuel (8.5 wt %, 20% enriched U--Zr--H fuel) and
three boron carbide control rods. The central core position (A1) of
the nuclear reactor was left empty, as well as the entire G ring
and most of the F-ring. The k.sub.eff for the model is
approximately 0.99. The temperature of the model is room
temperature until appropriate calculations can be performed to
estimate the operational temperature of the reactor fuel and
coolant.
[0041] An average neutron flux was determined for the A1 center
core position, as well as the nuclear reactor perimeter, G-ring
positions, of the TRIGA model for preliminary estimates of Pu-238
production. The thermal neutron flux for the A1 and G-ring
positions was 2.4.times.10.sup.13 neutrons per square centimeter
per second and 4.2.times.10.sup.12 neutrons per square centimeter
per second, respectively. Assuming a continuous flow of fresh,
neptunium target solution 2 around the core and that all the Pu-238
produced by irradiations is promptly removed from the irradiated
neptunium target solution 2' passing through the liquid loop 8,
initial calculations estimate that at 1 MW, the TRIGA reactor can
produce approximately 290 g/yr of Pu-238. Conventional TRIGA
nuclear reactors may readily be operated up to 3 MW, which is
estimated to produce approximately 880 g/yr of Pu-238. This model
provides the basis for an affordable, isotope production system 4
to be built that will allow production of Pu-238 to begin for
approximately $100 million total ($50 million for the TRIGA nuclear
reactor and $30 million for the separator) with annual costs of
less than approximately $5 million per year.
[0042] While the invention may be susceptible to various
modifications and alternative forms, specific embodiments have been
shown by way of example in the drawings and have been described in
detail herein. However, it should be understood that the invention
is not intended to be limited to the particular forms disclosed.
Rather, the invention is to cover all modifications, equivalents,
and alternatives falling within the scope of the invention as
defined by the following appended claims and their legal
equivalents.
* * * * *