Patent Application
20120103900
The present invention relates to the separation of actinides, including for example americium, from radioactive waste solutions.
A number of nuclear processes generate spent nuclear fuel with high levels of radioactive decay. At present, spent nuclear fuel is stored among reactor sites in approximately 39 different states across the U.S., and largely without recycling or reprocessing. For example, as much as 59,000 tons of spent nuclear fuel are in temporary storage, growing at a rate of 2,300 tons per year. Over the past several decades, however, efforts have been underway to centralize the disposal of current and future spent nuclear fuel into a single repository. To maximize the capacity of a single repository, methods have been proposed to recycle portions of spent nuclear fuel, thereby minimizing the quantity of spent nuclear fuel intended for long term storage.
For example, processes such as Uranium Extraction (UREX) and Plutonium Uranium Recovery by Extraction (PUREX) have been developed to extract uranium and/or plutonium from spent nuclear fuel for use in commercial reactors using liquid-liquid reprocessing techniques. After these extraction processes, however, a number of highly radioactive constituents remain. After UREX reprocessing, for example, the remaining constituents include plutonium (Pu) and several minor actinides, including americium (Am), neptunium (Np) and curium (Cm). Am constitutes approximately 73% of all the minor actinides present after fifty years of storage. In addition, Am-241 generates approximately 75% of the emitted heat after 250 years of storage, while its decay daughter, Np-237, is the principal source of radiotoxicity among the remaining minor actinides.
The amount of spent nuclear fuel that can be stored in any single repository is generally constrained by the amount of decay heat and radiotoxicity generated by the constituent actinides. As the decay heat and radiotoxicity of constituent actinides increase, the spacing between storage units within the repository must also increase. Thus, it is generally desirable to eliminate or severely reduce the presence of actinides, and most significantly Am, from spent nuclear fuel.
Accordingly, there remains a need for an improved system and method for the separation of actinides from spent nuclear fuel. In particular, there remains a need for an improved system and method for the separation of americium from an aqueous waste solution to minimize the heat decay and radiotoxicity of spent nuclear fuel for long term storage solutions.
Systems and methods for the extraction of americium from radioactive waste solutions are provided. The systems include the transfer of oxidized americium from an aqueous feed through an immobilized liquid membrane to an organic receiving phase. The receiving phase is subsequently stripped of americium and recycled at the immobilized liquid membrane for the continuous extraction of oxidized americium. The remaining constituent elements in the aqueous feed can be processed for long term storage substantially free of americium, and the sequestered americium can be used as a nuclear fuel, a nuclear fuel component or a radiation source.
The immobilized liquid membrane includes a supporting layer and a separating layer. Both layers are generally thin, having relatively high porosity and low tortuosity. The immobilized liquid membrane is optionally cylindrically-shaped, having an inner separating layer and an outer supporting layer. The aqueous feed flows through the cylinder core while the organic receiving phase is directed along the cylinder exterior. An organic immobilized solvent is retained within pores in the inner separating layer and within pores in the outer supporting layer to extract highly oxidized americium from the aqueous feed. The optional single channel, tubular geometry of the supporting and separating layers facilitates the cross-flow extraction of americium across the immobilized liquid membrane.
The aqueous feed can include an acidic solution containing oxidized americium. For example, the aqueous feed can initially include nitric acid containing Am(III). The Am(III) can be oxidized to Am(VI) by bubbling ozone through UREX and/or PUREX waste streams. In some applications the aqueous feed can be pressurized up to and including 14.5 psig, while in other applications the aqueous feed can be pressurized to higher pressures if desired (e.g., 50 psig or higher depending on the pore characteristics). The aqueous feed can be singularly or repeatedly treated at the immobilized liquid membrane before being processed for long-term storage substantially free of americium.
The immobilized solvent and the organic receiving phase include a suitable solvent or solvent mixture adapted to extract oxidized americium. In some embodiments the immobilized solvent and the organic receiving phase include tributyl phosphate, while in other embodiments the immobilized solvent and the organic receiving phase include an N,N-dialkylamide. A high concentration gradient of oxidized americium is maintained across the immobilized liquid membrane to provide for the continuous transfer of americium from the acidic aqueous feed and into the organic receiving phase. This process improves the separation efficiency over existing methods by maximizing the removal of Am(VI) from the acidic aqueous feed while avoiding equilibrium limitations.
In some embodiments, the organic receiving phase is subsequently stripped of americium. For example, the organic receiving phase is stripped of americium in a back-extraction process at a second immobilized liquid membrane. The second immobilized liquid membrane can be in series with the first immobilized liquid membrane, or can be nested around the exterior of the first immobilized liquid membrane. The second immobilized liquid membrane can contain immobilized diluted acid and can be surrounded by a diluted nitric acid receiving solvent. Neat tributyl phosphate can be recirculated to the first immobilized liquid membrane for reuse as an extractant of oxidized americium. Other methods for back-extraction include a mixer/settler containing a weak nitric acid solution and/or a centrifugal contactor containing a weak nitric acid solution. In addition, the supporting layer for the first and second immobilized liquid membranes is optionally formed of stainless steel, and the separating layer for the first and second immobilized liquid membranes is optionally formed of a metal-oxide.
The systems and methods of the present invention employ robust immobilized liquid membranes to facilitate the removal of americium from an aqueous feed under highly acidic and radiolytic conditions, potentially achieving the removal of over 99% of the americium in the aqueous feed. The systems and methods overcome removal limitations caused by equilibrium effects and can be operated using only the solvents employed elsewhere in recycling processes. Additionally, phase mixing of solvents does not occur, eliminating the generation of secondary toxic waste and the potential for emulsion problems. The net result of the removal of americium is the substantial expansion of repository storage capacity. When used in combination with UREX reprocessing, the present invention can potentially remove in excess of 95% of the mass of nuclear waste, significantly reducing close-packing constraints caused by long-term heat generation and radiotoxicity.
These and other features and advantages of the present invention will become apparent from the following description of the invention, when viewed in accordance with the accompanying drawings and appended claims.
The invention as contemplated and disclosed herein includes systems and methods for the extraction of americium from spent nuclear fuel using an immobilized liquid membrane. As set forth below, the immobilized liquid membrane of the present invention can substantially remove americium from spent nuclear fuel to potentially increase the storage capacity of radioactive waste repositories while also providing a source of americium for nuclear fuel, nuclear fuel components and other applications.
After the separation and recovery of uranium and/or plutonium, the remaining spent nuclear fuel includes several constituent transuranic elements and fission products. The constituent transuranic elements typically include approximately 79.3% plutonium, 15.1% americium, 5.5% neptunium and 0.1% curium after UREX reprocessing as generally show in
With reference to
A wide selection of available and suitable materials can ensure long-term chemical stability of the inorganic composite structure 12 under strongly acidic and radiolytic conditions. In particular, the separating and supporting layers 20, 22 are generally selected from inorganic materials adapted to withstand the radiotoxicity of americium and the acidity of the aqueous feed (e.g., 3M HNO3) that would otherwise degrade an organic (e.g., polymeric) membrane. In the present embodiment, the separating layer 20 is formed of a ceramic or metal oxide, for example Al2O3 or TiO2, while alternative materials may be utilized in other embodiments. In addition, the supporting layer 22 is formed of stainless steel in the present embodiment, while in other embodiments the supporting layer 22 can include other materials (e.g., ceramic oxides or carbides) as desired. In addition, the pore size within the separating layer 20 is generally at least an order of magnitude smaller than the pore size in the supportive layer 22. For example, the metal oxide separating layer 20 can include an average pore size in the range of 2-200 nm, optionally 5-20 nm, while the stainless steel supportive layer 22 can include an average pore size in the range of 0.5-50 μm, optionally 1-10 μm (not shown to scale).
An immobilized organic solvent 14, for example tributyl phosphate (TBP), is supported and retained in pores in the tubular structure 12 by capillary forces and wetting. Typically the immobilized organic solvent 14 is first completely incorporated into both layers 20, 22 of the composite structure 12 through a wetting step. A number of alternative organic solvents can also be utilized. For example, the immobilized solvent can include an N,N-dialkylamide, as well as hydrocarbons and various other mineral spirits that are suitable for use as extractants.
The composite structure 12 is well suited for the cross-flow transfer of oxidized americium (and oxidized cerium i.e. Ce(IV)) due at least in part to its optionally thin and uniform cross-section having relatively high permeability, narrow pore size distribution, as well as a high porosity and low tortuosity. As generally depicted in
In operation, Am(III) is oxidized to Am(VI) by bubbling ozone through an acidic aqueous feed. Unlike cerium and most of the trivalent fission products, Am(III) can be oxidized to its higher oxidation state to facilitate extraction. The oxidized feed flow is directed through the core of the composite structure 12, and a receiving flow including an organic receiving phase is directed along the exterior of the composite structure 12. TBP contained within the separating layer 20 extracts Am(VI) solutes from the aqueous feed. Am(VI) is then transferred into the supporting layer 22 at the interface 24 between the separating and supporting layers 20, 22. The concentration of Am(VI) in the separating and supporting layers 20, 22 declines as one moves radially outward from the first major surface 16 toward the second major surface 18. In addition, there is a change in the concentration gradient across the tubular structure 12 at the interface 24.
The extraction or transfer of Am(VI) from the separating layer 20 and into the supporting layer 22 is followed by the continued transfer of Am(VI) into the sweeping receiving flow. The receiving flow includes neat TBP in the present embodiment, becoming a TBP solution of americium as shown in
As also shown in
The above steps can occur rapidly in series, providing high flux and selectivity. Because TBP extracts oxidation states IV and VI and does not extract oxidation state III, Am(VI) is extracted into the TBP-filled inorganic separating and supporting layers 20, 22. With a sufficient driving force or concentration gradient generated from the americium-stripped receiving solvent, Am(VI) continues to be extracted into the TBP receiving phase through the thin and selective ILM.
While depicted as being cylindrical in
As noted above, the receiving flow can be stripped of americium as part of a continuous process. Referring now to
As optionally depicted in
The ILM of the present invention can be operated with or without a pressure differential across the composite structure 12. For example, the feed flow can be at a higher pressure than the receiving phase. In other embodiments, the feed flow and the receiving phase do not include a pressure differential. Generally, the capillary actions that retain the immobilized solvent (e.g. TBP) within the composite structure 12 should be sufficiently strong to withstand a pressure differential across the composite structure. More specifically, the capillary actions prevent the aqueous feed solution from permeating through the composite structure 12 at pressures at least 15-20 psi below the displacement pressure or bubble point pressure (i.e., the pressure at which the organic solvent can not be maintained integrally within the pores in the composite structure) for the system. The displacement pressure is related to the properties of the composite structure 12 (pore size, surface properties such as contact angle) and surface tension at the liquid-liquid interface. For example, membranes with a lower pore diameter (<0.5 μm) show relatively high bubble point pressures (>20 psig), which is desirable for the ILM.
In addition, the feed solution can be any americium-loaded aqueous solution. For example, the feed solution can include a 3M nitric acid solution from uranium or plutonium separation processes such as UREX, PUREX and TRUEX. Americium in the feed solution is initially oxidized to the +6 oxidation state according to any suitable method, including ozonation for example. The feed solution can also be pressurized as noted above. Pressures up to and including 14.5 psig have been tested, and the ILMs are expected to be capable of maintaining integrity at higher pressures, including for example up to and including 30 psig, 40 psig and/or 50 psig depending on the size of the separative layer.
Because TBP is also used in UREX extraction processes, the TBP receiving solvent does not introduce additional solvents into the industrial process. Additionally, since phase mixing does not occur, the extraction process eliminates the generation of secondary waste and the potential for emulsion problems. The immobilized solvent can also be different from the receiving phase. For example, the immobilized solvent can include TBP, while the receiving phase includes N,N-dialkylamide or other suitable extractants.
Americium sequestered by the system and method noted above can be of a sufficient quality for use as a nuclear fuel, a nuclear fuel component, or as a radioactive source. The separation of americium according to the present invention is superior over other methods by eliminating equilibrium limitations and reducing waste streams. The removal of americium can potentially eliminate close-packing constraints caused by the large amount of heat produced by americium. Thus, spent nuclear fuel can be stored in much smaller spaces, lessening the need to find new storage solutions.
The above description is that of current embodiments of the invention. Various alterations and changes can be made without departing from the spirit and broader aspects of the invention as defined in the appended claims, which are to be interpreted in accordance with the principles of patent law including the doctrine of equivalents. Any reference to elements in the singular, for example, using the articles “a,” “an,” “the,” or “said,” is not to be construed as limiting the element to the singular.
This application claims the benefit of U.S. Provisional Application No. 61/408,054, filed Oct. 29, 2010, the disclosure of which is hereby incorporated by reference in its entirety.
This invention was made with government support under Contract No. DE-AC05-00OR22725 awarded by the U.S. Department of Energy. The government has certain rights in the invention.
| Number | Date | Country | |
|---|---|---|---|
| 61408054 | Oct 2010 | US |
