Abstract: A method for extracting rare earth metals (e.g., lanthanides and/or actinides) from aqueous solution, the method comprising: (i) acidifying an aqueous solution containing said rare earth metals with sulfuric acid to result in an acidified aqueous solution containing 1-12 M concentration of sulfuric acid; and (ii) contacting the acidified aqueous solution with an aqueous-insoluble hydrophobic solution comprising a rare earth extractant molecule dissolved in an aqueous-insoluble hydrophobic solvent to result in extraction of one or more of the rare earth metals into the aqueous hydrophobic solution, wherein the rare earth extractant molecule has the following structure: wherein R1, R2, R3, and R4 are independently selected from hydrocarbon groups containing 1-20 carbon atoms, provided that the total carbon atoms in R1, R2, R3, and R4 is at least 12; and R5 and R6 are independently selected from hydrogen atom and hydrocarbon groups containing 1-3 carbon atoms.

Abstract: Novel dissymmetric N,N-dialkylamides which meet the following formula (I): where R represents a linear or branched alkyl group at C8 to C15. A method for synthesizing these N,N-dialkylamides, and to the uses of same as extractants, alone or in admixture, in order to extract uranium and/or plutonium from an aqueous acid solution, or to totally or separate uranium from plutonium from an aqueous acid solution and, in particular, an aqueous solution resulting from dissolving spent nuclear fuel in nitric acid. Further, a method for processing an aqueous solution resulting from the dissolution of a spent nuclear fuel in nitric acid, allowing the uranium and plutonium contained in the solution to be extracted, separated and decontaminated in a single cycle, without requiring any plutonium reduction operation, and in which one of the aforementioned N,N-dialkylamides or a mixture of same is used as extractant.

Abstract: A method for the treatment of an aqueous solution resulting from the dissolution of a spent nuclear fuel in nitric acid, allowing the uranium and plutonium contained in the solution to be extracted, separated and decontaminated in a single cycle, without requiring any operation involving a reductive stripping of plutonium. Applications for the method include the processing of uranium-based and/or plutonium-based spent nuclear fuels.

Abstract: A method using diglycolamide for increasing the separation factor between americium and curium and/or between lanthanides during an extraction operation. The operation comprising putting an acid aqueous phase, in which are found the americium, curium and/or lanthanides, in contact with an organic phase non-miscible with water, containing at least one extractant in an organic diluent. The aqueous and organic phases are then separated, and the diglycolamide is added to the aqueous phase.

Abstract: A process which allows separation of americium present in an acid aqueous phase or in an organic phase from the other metal elements also found in this phase, by complexation of the americium with a water-soluble ethylenediamine derivative; and a process for selective recovery of americium from an acid aqueous phase containing, in addition to americium, other metal elements, which comprises the application of this separation process.

Abstract: New compounds which meet general formula (I): where: m=0, 1 or 2; R1 and R2=a C6 to C12, straight-chain or branched, hydrocarbon group; R3=H, a C1 to C12, straight-chain or branched, hydrocarbon group, optionally comprising one or more heteroatoms; a C3 to C8 monocyclic hydrocarbon group, optionally comprising one or more heteroatoms, or a monocyclic (hetero)aryl group; or else R2 and R3 together form a —(CH2)n— group where n=1 to 4; R4=a C2 to C8, straight-chain or branched, hydrocarbon group, or a monocyclic aromatic group; R5=H, or a C1 to C12, straight-chain or branched, hydrocarbon group.

Abstract: Provided herein are processes for recovering molybdenum and/or other value metals (e.g., uranium) present in aqueous solutions from a large range of concentrations: from ppm to grams per liter via a solvent extraction process by extracting the molybdenum and/or other value metal from the aqueous solution by contacting it with an organic phase solution containing a phosphinic acid, stripping the molybdenum and/or other value metal from the organic phase solution by contacting it with an aqueous phase strip solution containing an inorganic compound and having a ?1.0 M concentration of free ammonia, and recovering the molybdenum and/or other value metal by separating it from the aqueous phase strip solution. When the molybdenum and/or other value metal are present only in low concentration, the processes can include an organic phase recycle step and/or an aqueous phase strip recycle step in order to concentrate the metal prior to recover.

Abstract: In a preferred embodiment, a process for extracting uranium from wet-process phosphoric acid (WPA), comprises separating uranium from WPA to produce a loaded uranium solution stream and a uranium depleted WPA stream. The loaded uranium solution stream is then contacted by with an ion exchange resin. Uranium species bound to the ion exchange resin are eluted by contacting the resin with a solution comprising anions to produce a loaded uranium eluant stream. The loaded uranium eluant stream is treated to provide a uranium containing product.

Abstract: A method for extracting a radioisotope from an aqueous solution, the method comprising: a) intimately mixing a non-chelating ionic liquid with the aqueous solution to transfer at least a portion of said radioisotope to said non-chelating ionic liquid; and b) separating the non-chelating ionic liquid from the aqueous solution. In preferred embodiments, the method achieves an extraction efficiency of at least 80%, or a separation factor of at least 1×104 when more than one radioisotope is included in the aqueous solution. In particular embodiments, the method is applied to the separation of medical isotopes pairs, such as Th from Ac (Th-229/Ac-225, Ac-227/Th-227), or Ra from Ac (Ac-225 and Ra-225, Ac-227 and Ra-223), or Ra from Th (Th-227 and Ra-223, Th-229 and Ra-225).

Abstract: A method with which uranium from a natural uranium concentrate may be purified, including a) extracting the uranium present as uranyl nitrate in an aqueous phase A1 resulting from the dissolution of the natural uranium concentrate in nitric acid, by means of an organic phase which contains an extractant in an organic diluent; b) washing the organic phase obtained at the end of step a), with an aqueous phase A2; and c) stripping the uranyl nitrate of the organic phase obtained at the end of step b), by circulating this organic phase in an apparatus, as a counter current against an aqueous phase A3. The extractant is an N,N-dialkylamide and the ratio between the flow rate at which the organic phase obtained at the end of step b) and the aqueous phase A3 circulate in the apparatus where step c) occurs, is greater than 1.

Abstract: The invention relates to a process for reprocessing spent nuclear fuel which, among other advantages, does not require a plutonium-reducing stripping operation. This process finds particular application in the processing of uranium oxide fuels and uranium and plutonium mixed oxide fuels.

Abstract: The invention relates to a process which makes it possible to separate together all the actinide(III), (IV), (V) and (VI) entities present in a highly acidic aqueous phase from fission products, in particular lanthanides, also present in this phase by using a solvating extractant in a salting-out medium. Applications: reprocessing of irradiated nuclear fuels, in particular for recovering plutonium, neptunium, americium, curium and possibly uranium, present in the form of traces, in a pooled but selective fashion with regard to lanthanides, from a solution for the dissolution of an irradiated nuclear fuel, downstream of a cycle for the extraction of uranium.

Abstract: A method with which americium may be selectively recovered from a nitric aqueous phase containing americium, curium and fission products including lanthanides and yttrium, but which is free of uranium, plutonium and neptunium or which only contains these three last elements in trace amounts. The method is applicable for treatment and recycling of irradiated nuclear fuels, in particular for removing americium from raffinates stemming from methods for extracting and purifying uranium and plutonium such as the PUREX and COEX™ methods.

Abstract: A process which allows separation of americium present in an acid aqueous phase or in an organic phase from the other metal elements also found in this phase, by complexation of the americium with a water-soluble ethylenediamine derivative; and a process for selective recovery of americium from an acid aqueous phase containing, in addition to americium, other metal elements, which comprises the application of this separation process.

Abstract: Provided are methods using a ketoxime in metal extraction. One aspect of the invention relates to a method for the recovery of metal from a metal-containing aqueous solution at an elevated temperature using a ketoxime. Another aspect relates to a method of separating iron/copper using a specific ketoxime. Aldoximes may also be added to the reagent compositions used in these methods.

Abstract: A method of processing spent TRIZO-coated nuclear fuel may include adding fluoride to complex zirconium present in a dissolved TRIZO-coated fuel. Complexing the zirconium with fluoride may reduce or eliminate the potential for zirconium to interfere with the extraction of uranium and/or transuranics from fission materials in the spent nuclear fuel.

Abstract: Reagent compositions, methods for their manufacture and methods of their use are described. In particular, provided are reagent compositions comprising an aldoxime and ketoxime with an alkyl substituent. Also provided are methods of metal recovery using these reagent compositions.

Abstract: Provided are methods using ketoximes and/or aldoximes, including 3-methyl-5-alkylsalicylaldoxime and/or 3-methyl-5-alkyl-2-hydroxyacetophenone oxime, in reagent compositions for metal extraction/isolation. One such method is of extracting a metal from a nitrate-containing aqueous solution. Another such method is of extracting a metal from an aqueous ammoniacal solution. A third method is of multi-metal extraction based on a predetermined pH.

Abstract: The invention relates to the use of butyraldehyde oxime as an anti-nitrous agent in a plutonium stripping operation based on a reduction of this element from oxidation state (IV) to oxidation state (III). Applications: any nuclear fuel reprocessing process in which employing a compound that has the twofold property of being extractable into an organic phase and of being capable of destroying the nitrous acid therein may be useful and especially any process including one or more operations for the reductive stripping of plutonium.

Abstract: A method for treating spent nuclear fuel, which includes first decontaminating the uranium, plutonium and neptunium found in a nitric aqueous phase resulting from dissolving the nuclear fuel in HNO3. The uranium, plutonium and neptunium found in the solvent phase is then split in a first aqueous phase and a second aqueous phase. Next, the first aqueous phase is stored. Following, the plutonium or other mixtures found in the first aqueous phase is purified relative to the fission products still found in said phase, in order to obtain, at the end of said purification, an aqueous solution containing a mixture of Pu and U or Pu, U and Np. Finally the resulting mixture of Pu and U or the mixture of Pu, U and Np is co-converted into a mixed oxide.

Abstract: The process of the invention is the separation of minor actinides from lanthanides in a fluid mixture comprising, fission products, lanthanides, minor actinides, rare earth elements, nitric acid and water by addition of an organic chelating aid to the fluid; extracting the fluid with a solvent comprising a first extractant, a second extractant and an organic diluent to form an organic extractant stream and an aqueous raffinate. Scrubbing the organic stream with a dicarboxylic acid and a chelating agent to form a scrubber discharge. The scrubber discharge is stripped with a simple buffering agent and a second chelating agent in the pH range of 2.5 to 6.1 to produce actinide and lanthanide streams and spent organic diluents. The first extractant is selected from bis(2-ethylhexyl)hydrogen phosphate (HDEHP) and mono(2-ethylhexyl)2-ethylhexyl phosphonate (HEH(EHP)) and the second extractant is selected from N,N,N,N-tetra-2-ethylhexyl diglycol amide (TEHDGA) and N,N,N?,N?-tetraoctyl-3-oxapentanediamide (TODGA).

Abstract: The addition of a compatible metal salt crystal to the organic solution entering the mixer(s) in the solvent extraction stage(s) and/or the stripping stage(s), or to the emulsion mixture of the organic solution and the aqueous solution in the mixer(s), or to the mixture of the organic solution and the aqueous solution in a settler tank(s) following the mixer(s) in the solvent extraction and/or stripping stage(s) following the leaching of metal values from the ore containing that/those value(s) into an aqueous solution, and prior to the further refining of those values in processes, such as electrowinning, during mining operations for those metal values in order to improve the phase separation of the organic phase and the aqueous phase, and to promote the removal of contaminants from the organic phase.

Abstract: Reagent compositions, methods for their manufacture and methods of their use are described. In particular, provided are reagent compositions comprising an aldoxime and ketoxime with an alkyl substituent. Also provided are methods of metal recovery using these reagent compositions.

Abstract: Provided are methods using a ketoxime in metal extraction. One aspect of the invention relates to a method for the recovery of metal from a metal-containing aqueous solution at an elevated temperature using a ketoxime. Another aspect relates to a method of separating iron/copper using a specific ketoxime. Aldoximes may also be added to the reagent compositions used in these methods.

Abstract: Provided are methods using ketoximes and/or aldoximes, including 3-methyl-5-alkylsalicylaldoxime and/or 3-methyl-5-alkyl-2-hydroxyacetophenone oxime, in reagent compositions for metal extraction/isolation. One such method is of extracting a metal from a nitrate-containing aqueous solution. Another such method is of extracting a metal from an aqueous ammoniacal solution. A third method is of multi-metal extraction based on a predetermined pH.

Abstract: A chemical element to be very efficiently separated from uranium starting from an acid aqueous phase, in an extraction cycle for the uranium, when this chemical element is present in said phase at a concentration less than that of the uranium, or even as a trace element, and when it is moreover less extractable by the extractant used in this extraction cycle than is the uranium. The chemical element can notably be neptunium(IV) or thorium 228.

Abstract: A method for separating and recovering trivalent and tetravalent actinoids in a simple and less costly manner without using an organophosphorus compound is provided. This method selectively separates and recovers the tetravalent actinoid plutonium Pu (IV) and the trivalent actinoids americium Am (III) and curium Cm (III) from trivalent lanthanoids Ln (III), etc. with the use of an extractant having a functional group with neutral multidentate ligand activity which is a hybrid donor type organic compound having both of donor atoms, i.e., an oxygen atom and a nitrogen atom.

Abstract: The present invention provides a method for extraction of metals selected from Cr, Mo, Pd, Tc, W, Re, and Pu using a new extractant of methyliminobisalkylacetamide represented by a formula (I): CH3—N—(CH2CONR2)2??(I) wherein R represents an alkyl group having 8-12 carbon atoms.

Abstract: The invention relates to a process for reprocessing a spent nuclear fuel and for preparing a mixed uranium-plutonium oxide, which process comprises: a) the separation of the uranium and plutonium from the fission products, the americium and the curium that are present in an aqueous nitric solution resulting from the dissolution of the fuel in nitric acid, this step including at least one operation of coextracting the uranium and plutonium from said solution by a solvent phase; b) the partition of the coextracted uranium and plutonium to a first aqueous phase containing plutonium and uranium, and a second aqueous phase containing uranium but no plutonium; c) the purification of the plutonium and uranium that are present in the first aqueous phase; and d) a step of coconverting the plutonium and uranium to a mixed uranium/plutonium oxide. Applications: reprocessing of nuclear fuels based on uranium oxide or on mixed uranium-plutonium oxide.

Abstract: The invention relates to a process which makes it possible to separate together all the actinide(III), (IV), (V) and (VI) entities present in a highly acidic aqueous phase from fission products, in particular lanthanides, also present in this phase by using a solvating extractant in a salting-out medium. Applications: reprocessing of irradiated nuclear fuels, in particular for recovering plutonium, neptunium, americium, curium and possibly uranium, present in the form of traces, in a pooled but selective fashion with regard to lanthanides, from a solution for the dissolution of an irradiated nuclear fuel, downstream of a cycle for the extraction of uranium.

Abstract: A spent fuel reprocessing method contacts an aqueous solution containing Technetium(V) and uranyl with an acidic solution comprising hydroxylamine hydrochloride or acetohydroxamic acid to reduce Tc(V) to Tc(II, and then extracts the uranyl with an organic phase, leaving technetium(II) in aqueous solution.

Abstract: An apparatus for the removal of uranium from a body of material is provided. The apparatus has at least one ultrasonic extractor, having a bottom and a top. The at least one ultrasonic extractor is configured to accept solids at the bottom and acid at the top, and has a mixing screw and at least one source of ultrasonic energy. The mixing screw is configured to transport the solids in a direction countercurrent to the acid in the at least one ultrasonic extractor; and the source of ultrasonic energy is configured to impart ultrasonic energy into the solids and the acid, as the solids and the acid traverse the at least one ultrasonic extractor countercurrently.

Abstract: A method is disclosed for separating trivalent americium from trivalent curium, coming from an aqueous solution containing at least these cations, wherein, at an acid concentration of 0.01 mol/l-0.3 mogl/l, the aqueous solution is brought into contact with an organic solvent containing a bis(aryl)dithiophosphinic acid having the formula (4) where R1=phenyl or naphthyl R2=phenyl or naphthyl, and radicals of R1 and R2 substituted by at least one methyl, ethyl, propyl, isopropyl-, cyano, nitro, or halo substituent, and containing a synergist having the formula (5) where X and/or Y and/or Z is R or RO, wherein R is branched or unbranched alkyl.

Abstract: A spent fuel reprocessing method contacts an aqueous solution containing Technetium(V) and uranyl with an acidic solution comprising hydroxylamine hydrochloride or acetohydroxamic acid to reduce Tc(V) to Tc(II, and then extracts the uranyl with an organic phase, leaving technetium(II) in aqueous solution.

Abstract: An extraction agent for the separation of trivalent actinides from lanthanides in an acidic media and a method for forming same are described, and wherein the methodology produces a stable regiospecific and/or stereospecific dithiophosphinic acid that can operate in an acidic media having a pH of less than about 7.

Abstract: The invention relates to a method for separating uranium(VI) from one or more actinides selected from actinides(IV) and actinides(VI) other than uranium(VI), characterized in that it comprises the following steps: a) bringing an organic phase, which is immiscible with water and contains the said uranium and the said actinide or actinides, in contact with an aqueous acidic solution containing at least one lacunary heteropolyanion and, if the said actinide or at least one of the said actinides is an actinide(VI), a reducing agent capable of selectively reducing this actinide(VI); and b) separating the said organic phase from the said aqueous solution. Applications: reprocessing irradiated nuclear fuels, processing rare-earth, thorium and/or uranium ores.

Abstract: An apparatus and method to remove uranium from a body of material wherein the method includes the steps of depositing the body of solid material in an ultrasonic extractor and depositing an amount of acid in the ultrasonic extractor. The method also provides for the steps of heating a jacket of the ultrasonic extractor, transporting the body of solid material in the ultrasonic extractor and the amount of acid such that the body of solid material and the acid contact each other inside the heated ultrasonic extractor while the ultrasonic extractor provides ultrasonic energy to both the body of solid material and the amount of acid, wherein the amount of acid strips uranium from the body of solid material. The method further provides for collecting the amount of acid and the body of solid material in the ultrasonic extractor in different positions, transporting the amount of acid with the stripped uranium to an extraction mixer settler, and settling uranium product from the extraction mixer settler.

Abstract: Extraction mixture capable of recovering actinide elements such as U, Pu and transplutonium elements from radioactive liquid waste in reprocessing of spent nuclear fuel. One embodiment of the extraction mixture includes a solution of bidentate organophosphorus extractant, dihexyl-N, N-diethylcarbamoyl phosphonate in a polar diluent, wherein bis-tetrafluoropropyl ether of diethylene glycol is used as the polar diluent at the following ratio of components: 0.1-1.2 M/L of bidentate extractant and the rest of diluent. Another embodiment of the extraction mixture includes a solution of bidentate organophosphorus extractant, phenyloctyl-N,N-diisobutylcarbamoylphosphine oxide in a polar diluent, wherein a mixture of metanitrobenzotrifluoride with trialkylphosphate is used as the polar diluent at the following ratio of components: 0.1-1.2 M/L of bidentate extractant, 0.3-1.1 M/L of TBP, and the rest of MNBTF.

Abstract: Most part of an amount of uranium contained in the spent nuclear fuel is removed by making fluorine or a fluorochemical act on the spent nuclear fuel to convert the uranium into UF6, and the uranium is purified through a simple method of distilling the UF6 together with a absorbent. After removing the most part of the amount of uranium, the remaining nuclear fuel material is dissolved and then transferred to an extraction process to recover plutonium. By doing so, a small sized dry process can be employed as a uranium purification process. Since the nuclear fuel material is dissolved and extracted after removing most part of an amount of uranium, a volume of processing solution can be reduced and the machine installation scale can be made small. Accordingly, the reprocessing facility can be extremely downsized.

Abstract: A method for separating a metal (a) from a metal (2), preferably zirconium from hafnium, which consists in dissolving said metals in an aqueous solution wherein said metals are in a state preventing them from passing through a nanofiltration membrane; treating the aqueous medium with a ligand, for example EDTA, which is complexed with metal (1) and/or metal (2), then in passing the resulting treated medium on a filtering membrane allowing through the ligand-metal complexes, but retaining the metals not complexed with the ligand.

Abstract: The present invention provides a burner for use in a combustion-type waste gas treatment system for combusting waste gases emitted from semiconductor manufacturing system, particularly, a deposition gas containing SiH4 and a halogen-base gas, simultaneously at a high efficiency of destruction, making it difficult for a powder of SiO2 to be attached and deposited, performing a low-NOx combustion, and maintaining a desired level of safety. The combustion-type waste gas treatment system has a flame stabilizing zone (15), which is open toward a combustion chamber (11), surrounded by a peripheral wall (12), and closed by a plate (14) remotely from the combustion chamber. A waste gas, an auxiliary combustible agent, and air are introduced into and mixed with each other in the flame stabilizing zone (15), and the mixed gases are ejected toward the combustion chamber (11) perpendicularly to the plate (14).

Abstract: A method of recovering zirconium values from an ore containing zircon, baddeleyite, and uranium is disclosed. The method includes fusing the ore with soda ash and contacting the resulting fused ore with sulfuric acid, which generates an acid leach liquor that contains zirconium and uranium values and solids that comprise baddeleyite and silica. Baddeleyite is recovered by contacting it with sulfuric acid to convert baddeleyite to zirconium sulfate, which can be dissolved in the acid leach liquor. The method also includes separating zirconium from uranium by solvent extraction followed by precipitation.

Abstract: The invention relates to new calixarenes according to the formula: wherein R1 and R2, which may be identical or different, are alkyl, alkoxy or aryl groups, and n is an integer equal to 2, 3 or 4. Said calixarenes may be used to extract actinides and lanthanides from aqueous solutions.

Abstract: A spent fuel reprocessing method including the steps of partitioning U and Pu(III) in a solvent by solvent extraction and subsequently polishing the solvent in a neptunium rejection operation for removing Np therefrom. The solvent obtained from the neptunium rejection operation (the polished solvent or NpA solvent product) is then recycled to a U/Pu partitioning operation. The method enables a reduction in solvent feed and solvent effluent volumes.

Abstract: A method is describe for removing metal ions of interest from aqueous solutions containing low concentrations of the metal ions of interest. The method involves the use of a microbially produced polymer, &ggr;-glutamic acid. The polymer is used in conjunction with tangential flow filtration and results in substantial removal of metal ions.

Abstract: The present invention relates to solvents, and methods, for selectively extracting and recovering radionuclides, especially cesium and strontium, rare earths and actinides from liquid radioactive wastes. More specifically, the invention relates to extracting agent solvent compositions comprising complex organoboron compounds, substituted polyethylene glycols, and neutral organophosphorus compounds in a diluent. The preferred solvent comprises a chlorinated cobalt dicarbollide, diphenyl-dibutylmethylenecarbamoylphosphine oxide, PEG-400, and a diluent of phenylpolyfluoroalkyl sulfone. The invention also provides a method of using the invention extracting agents to recover cesium, strontium, rare earths and actinides from liquid radioactive waste.

Abstract: The spent nuclear fuel reprocessing method includes contacting an organic phase containing Np(VI) with a hydrophilic substituted hydroxylamine in which at least one N—H hydrogen of the hydroxylamine is replaced by an organic substituent, there being at least one such organic substituent having one or more hydroxyl groups. Np(VI) is reduced to Np(V), which is then backwashed into an aqueous phase.

Abstract: A spent nuclear fuel reprocessing method includes contacting an organic phase containing NP(VI) with an oxime of the formula R2C═NOH, where each R is independently H or an organic substituent. Np(VI) is reduced to Np(V) which is then backwashed into an aqueous phase.

Abstract: A method of dissolving in an ionic liquid a metal in an initial oxidation state below its maximum oxidation state, characterized in that the ionic liquid reacts with the metal and oxidizes it to a higher oxidation state. The initial metal may be in the form of a compound thereof and may be irradiated nuclear fuel comprising UO2 and/or PuO2 as well as fission products. The ionic liquid typically is nitrate-based, for example a pyridinium or substituted imidazolium nitrate, and contains a Bronstead or Franklin acid to increase the oxidizing power of the nitrate. Suitable acids are HNO3, H2SO4 and [NO+]. Imidazolium nitrates and certain pyridinium nitrates form one aspect of the invention.

Abstract: The invention involves a process for separating actinides and lanthanides by liquid-liquid extraction by means of calixarenes. These calixarenes have the formula: with R1 and R2 being alkyl groups or o-nitrophenoxy alkyl groups and R3 and R4 being aryl groups, and they are used in an organic liquid phase containing an organic diluent. The diluent and the calixarene concentration of the organic phase are chosen so as to ensure an enrichment of the organic phase with the actinide(s) and/or lanthanide(s) to be separated from an aqueous acid or saline solution.