The present disclosure relates to target materials in nuclear reactors and more particularly to fissile target materials and methods for processing fissile target materials.
The production of fission and activation products from fissile target irradiation presents a variety of challenges for follow on use of individual fission products or activation products. In order to use either a fission product or an activation product, a chemical separation must be performed to isolate the product of interest. Isolation of the desirable product, individual fission product or activation product, requires a large-scale separation from the bulk unconsumed fissile target material, separation from fission products, or activation products of non-interest. These chemical separations are costly, time consuming, and often involve high dose samples (and engineered shielding to enable handling of these samples). To address these concerns a variety of technologies and techniques have been invented and deployed. However, each of these prior art embodiments and techniques present various problems and alternatives. The present invention provides a significant advance over these items.
Methods for preparing fissile target materials are provided. The methods can include preparing a target substrate that includes a fissile atom, and layering at least one surface of the substrate with a capturing layer.
Fissile target materials are provided. The fissile target materials can include a target substrate and a capturing layer operably interfacing with at least one surface of the target substrate.
Methods for fissioning fissile target materials are also provided. The methods can include irradiating fissile target material to capture fission products of the irradiated fissile target material in a capturing layer of the target material.
Fission fissile target materials are also provided that can include a target substrate comprising at least one fissile atom and a capturing layer operably interfacing with at least one surface of the target substrate. The capturing layer can include at least one fission product.
Methods for separating fissioned product from fission fissile target materials are also provided. The methods can include separating at least a portion of the captured layer of the fissioned fissile target material from the fissile target material.
Methods for processing fissioned fissile target materials are also provided. The methods can include separating a capture layer from the fissioned fissile target material to remove at least some of the fissioned product atoms. The methods can further include purifying fissile atoms for recycling and/or isolating activation product.
The patent or application file contains at least one drawing executed in color. Copies of this patent or patent application publication with color drawing(s) will be provided by the Office upon request and payment of the necessary fee.
Embodiments of the disclosure are described below with reference to the following accompanying drawings.
This disclosure is submitted in furtherance of the constitutional purposes of the U.S. Patent Laws “to promote the progress of science and useful arts” (Article 1, Section 8).
The methods and materials of the present disclosure will be described with reference to
In accordance with example implementations, target material 12 can have a surface, and upon that surface can be applied a capturing layer 14. Capturing layer 14 can be a metal or organic material. When configured planarly, the capturing layer can be provided along a flat surface of the planar substrate. When configured spherically, the capturing layer can be provided along a curved surface of the sphere. In accordance with example implementations, the capturing layer can be provided about the entirety of the exterior surface of the target substrate. Example implementations can include providing an intermediate layer between the capturing layer and the target substrate and/or providing separable capturing layers, and/or multiple capturing layers of different compositions. Capturing layers can be applied via deposition, coating, veneering, and/or laminating, for example.
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In accordance with example implementations, the fissile target material can be spherical such as a micro capsule like that shown in
Example target substrates can include Th, U, Np, and/or Pu (232Th, 233U, 235U, 238U, 237Np, 238Pu, 239Pu, and/or 240Pu). The target substrates can have a thickness it at least one cross section and this thickness can be from about 2 μm to about 5 μm, making the targets microscale targets.
Example capturing layer materials can include metals such as vanadium, nickel, graphite, ceramics, and/or polymers such as polyethylene glycol (PEG) and other polymers composed at least partially or entirely out of carbon and oxygen. This capturing layer can have an overall thickness in at least on cross section of from about 9 μm to about 35 μm. In accordance with example implementations, multiple and/or separable capturing layers can be provided upon the target substrates and these layers may be of the same or different materials.
Target material 12 can be considered a microscale recoil suppression coated fissile target material that can be polymer coated as the thickness of the material in at least one cross section can be from about 12 μm to about 40 μm.
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Fission products can be created and these fission products can enter the capturing layer. The fission products can be separated by at least about 2 μm in capturing layer. The capturing layer can include multiple layers such as at least two or more layers that are separable from one another. In accordance with example implementations one fission product can be captured within at least one of the capturing layers and another fission product can be captured within another of the capturing layers.
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In accordance with example implementations, the fissioned fissile target materials can include activation products within the target substrate. These activation products can be one or more of 233U, 237Np, 238Pu, 244Pu, 241Am, 243Am, 248Cm, 249Bk, 249Cf, and/or 252Bk. The fission products within the capturing layer can be one or more of 131I, 133Xe, 131Cs, 133Cs, 134Cs, 89Sr, 90Y, 153Sm, 90Sr, 140Ba, 95Zr, 95Nb, 140La, 144Ce, 144Pr, 141Ce, 147Pm, 105Rh, 151Sm, 106Rh, or 106Ru. At least two of these fissioned products may reside in different portions of the capturing layer, and these different portions may be separable from one another and/or of different compositions.
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In accordance with example implementations, irradiated fissile cores can be recycled for thermal neutron irradiations, 14 MeV neutron irradiations, and accelerator irradiations. To recycle the target materials the capture layer can be removed, isolating the fissile core. The fissile core can be recoated with a fresh recoil capture layer and the irradiation repeated.
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A shallow dive exploration of fission product ejection from uranium metal and uranium oxide can be performed using SRIM model for uranium and recoil capture materials. The input parameters for this modeling can be the density and composition of the target material or recoil capture material. For the mass and element of the ion (fission product) and the initial kinetic energy for the ion (fission product) Equation 1 below can be used, for conservation of energy in A-symmetric fission and 165 MeV as the kinetic energy split between the fission products. The greatest and least recoil ejections of one of the smallest fission products 72 Ga (˜115 MeV) and one of the largest fission products 161 Tb (˜50 MeV) can be determined.
Where E1 is the kinetic energy for the fission product of mass A1, A2 is the mass of the other fission product, and Ekinetic is the total fission product energy imparted into each fission product.
The stopping range of fission products can be determined for uranium metal, uranium oxide, and for an outer recoil capture layer.
Referring to
Fission product recoil capture into a polyethylene glycol capture layer can be determined for the 161Tb fission product transmission through 3 μm of uranium oxide into polyethylene glycol and displayed in
Additionally, transmission of 72Ga, the fission product with the longest recoil ejection can be determined. The penetration of 72Ga through 3 μm into an additional 17 μm of polyethylene glycol capture layer are displayed in
Referring again to the target material of
Accordingly, two separate assemblies were constructed, each beginning with electrodeposition of 93% enriched uranium onto a 50-μm thick graphite rods. The mass of uranium deposited onto the graphite rod was roughly 10 μg for each rod and equated to a thickness of less than 1 μg onto the graphite rods. Vanadium metal was chosen for the outer recoil capture layer because of its thermal resistivity and neutron transparency. The vanadium recoil capture layer was 100 μm thick. This target assembly is shown in
The assemblies were irradiated over a course of 1.78 hours, producing roughly 1010 fissions, and then allowed to cool for 16 hours before gamma screening and disassembly. High-purity germanium detectors were used to perform gamma spectroscopy on the complete assembly and each individual portion of the assembly.
The results of the gamma analysis are displayed in Table 1. The vanadium recoil capture layers were manually removed from the assembly, prior to fission product quantification. The uranium target layer was removed from the graphite by agitation mixing with 10 mL of concentrated hydrochloric acid. The graphite recoil capture layers were analyzed after the uranium target material was removed.
131I (Bq)
133Xe (Bq)
239Np (Bq)
Quantification of the activation product, 239Np, in the recoil capture layer was not possible because the quantity of 239Np produced during irradiation was below the gamma analysis detection limit. For the uranium target layer, the 239Np activation product was detected in the target layer during irradiation and after disassembly of the target material.
In accordance with another example, the target material can be a 3 layered target with the inner layer Ni, the middle layer 235U, and an outer V capsule. After irradiation, fission products in Ni layer and the V capsule could be quantified with no detectable 235U present. The fission product and 235U for target type is shown below in Table 2.
235U-
235U-
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Accordingly, the method can include production of 236Np by proton bombardment of 238U at roughly 25 MeV. During the production of 236Np, a side product of 1×1018 fissions are produced, which approximately correlates to 500 Rad an hour after 7 days of cooling. With the fissile target materials of the present disclosure, the fission products can be ejected from the target material and captured in the capture layer. In a single step performed in a hot cell, the capture layer, containing the fission products, can be removed from the target material, containing the 238U and 236Np. A follow-on purification procedure can be performed in a glove box. The cost, time, and purity of the final 236Np product will be superior to prior art methods.
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In accordance with example implementations, fuel cells can include the target materials of the present disclosure, and these fuel cells can be packed with microscale or planar materials, and the microscale or planar materials can be processed upon irradiation in conventional nuclear reactor assemblies.
In compliance with the statute, embodiments of the invention have been described in language more or less specific as to structural and methodical features. It is to be understood, however, that the entire invention is not limited to the specific features and/or embodiments shown and/or described, since the disclosed embodiments comprise forms of putting the invention into effect. The invention is, therefore, claimed in any of its forms or modifications within the proper scope of the appended claims appropriately interpreted in accordance with the doctrine of equivalents.
This application claims priority to and the benefit of U.S. Provisional Patent Application Ser. No. 62/880,746 filed Jul. 31, 2019, entitled “Advanced Targets for Isotope Production”, the entirety of which is incorporated by reference herein.
This disclosure was made with Government support under Contract DE-AC0576RL01830 awarded by the U.S. Department of Energy. The Government has certain rights in the invention.
Number | Date | Country | |
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62880746 | Jul 2019 | US |