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.

Description

CROSS-REFERENCE TO RELATED APPLICATION

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.

STATEMENT REGARDING FEDERALLY SPONSORED RESEARCH OR DEVELOPMENT

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.

TECHNICAL FIELD

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

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.

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 β-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.

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

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.

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.

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

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:

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

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.

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.

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.

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.

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.

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.

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.

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.

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.

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.

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.

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.

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:

where each of X₁-X₃ is independently selected from an alkyl group, an aryl group, an alkoxy group, an aryloxy group, or combinations thereof. The X₁-X₃ 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.

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-β-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.

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® L or Isopar® M. Isopar® L includes a mixture of C₁₀-C₁₂ isoparaffinic hydrocarbons and is available from Exxon Chemical Co. (Houston, Tex.). Isopar® M includes a mixture of C₁₂-C₁₅ 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.

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.

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⁺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 HNO₃ to approximately 0.5 M HNO₃, such as approximately 0.01 M HNO₃. 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.

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.

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.

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.

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.

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.

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 HNO₃ to approximately 6 M HNO₃, such as from 0.001 M HNO₃ to approximately 0.5 M HNO₃. By way of non-limiting example, the aqueous acid solution may have approximately 0.01 M HNO₃. 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.

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.

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⁺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.

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.

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.

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.

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_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.

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 2.4×10¹³ neutrons per square centimeter per second and 4.2×10¹² 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.

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.

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.