Abstract: A dissymmetric RN,N-dialkylamides of formula (I) in which: R1 represents a linear C1 to C4 alkyl, R2 represents a linear C1 to C10 alkyl, and R3 represents a linear or branched C6 to C15 alkyl, where R3 is different from a n-octyl, n-decyl, n-dodecyl, 2-ethylhexyl and 2-ethyloctyl group when R1 represents a n-butyl group and R2 represents an ethyl group. A method for synthesising the N,N-dialkylamides, and uses of same for extracting uranium and/or plutonium from an aqueous acid solution or for fully or partially separating the uranium from the plutonium contained in an aqueous acid solution and a solution resulting from the dissolution of spent nuclear fuel in nitric acid. A method for treating an aqueous solution resulting from the dissolution of spent nuclear fuel in nitric acid, which allows the uranium and the plutonium contained in the solution to be extracted, separated and decontaminated in a single cycle.

Abstract: A composition comprising a Nitrate or a Nitrite of an Amino Acid Compound selected from the group consisting of Arginine, Agmatine, Beta Alanine, Citrulline, Creatine, Glutamine, L-Histidine, Isoleucine, Leucine, Norvaline, Ornithine, Valine, Aspartic Acid, Cysteine, Glycine, Lysine, Methionine, Proline, Tyrosine, and Phenylalanine is provided. The composition may be in a dosage form selected from the group consisting of a capsule, a cachet, a pill, a tablet, a powder, a granule, a pellet, a bead, a particle, a troche, a lozenge, a gel, a liquid, a suspension, a solution, an elixir, and a syrup.

Abstract: A supplement formulation comprising a Nitrate or a Nitrite of an Amino Acid Compound selected from the group consisting of Arginine, Agmatine, Beta Alanine, Citrulline, Creatine, Glutamine, L-Histidine, Isoleucine, Leucine, Norvaline, Ornithine, Valine, Aspartic Acid, Cysteine, Glycine, Lysine, Methionine, Proline, Tyrosine, and Phenylalanine is provided. The supplement formulation may be in a dosage form selected from the group consisting of a capsule, a cachet, a pill, a tablet, a powder, a granule, a pellet, a bead, a particle, a troche, a lozenge, a gel, a liquid, a suspension, a solution, an elixir, and a syrup.

Abstract: The disclosure describes a method of extraction for optimal recovery of rare earth elements (REE). This includes the optimization of a liquid-liquid extraction technique for economic extraction of REE from chemically complex brines. Through the use of a model, operating conditions are altered to increase separation efficiency. The technique achieves >98% recovery of all rare earth elements in one fraction, while using small volumes of sample and reagents. The method of the present invention recovers REE from previously unexploited waste streams, without need for development of new process equipment and while keeping a small footprint because of the volumes involved.

Abstract: Provided is a method of recovering rare-earth elements by performing slight pH adjustment of a leachate, which does not require such significant pH adjustment as in conventional methods, reduces cost and effort, and can efficiently recover rare-earth elements in a bauxite residue in a good yield.

Abstract: The present invention relates to use of an amino-containing neutral phosphine extractant of Formula I in extraction and separation of thorium, and a process of extracting and separating thorium using the amino-containing neutral phosphine extractant of Formula I, wherein, R1 and R2 are each independently selected from the group consisting of C1-C12 alkyl, R3 and R4 are each independently selected from the group consisting of C1-16 alkyl and hydrogen, and n is an integer of 1 to 8.

Abstract: Apparatus and method are provided for the treatment of uranium-contaminated soil by using comprehensive joint technology. The apparatus include the pumping system, the electrokinetic remediation system, elution system, remediation-separation system and recharge system. The remediation technologies (i.e. chemical, photolysis and electrokinetic) are used to remedy the uranium-contaminated soil. First, extract uranium from the contaminated areas and make the ionized uranium extract from the soil phase to the solution phase. Then, use the electrokinetic remediation technology to drive uranium enrichment electromigrate to near the anode. Finally, return the repaired-soil and groundwater back to anode area and recharge well, respectively. This comprehensive joint apparatus can reduce the uranium volume in the contaminated soil or water, and recycle the obtained uranium, which are cleaning processes and have no secondary pollution.

Abstract: A system for extracting uranium from wet-process phosphoric acid (WPA), includes an ion exchange resin or solvent extractor for separating uranium from WPA to produce a loaded uranium solution stream and a uranium depleted WPA stream. An ion exchange resin is positioned to receive the loaded uranium solution stream and bind uranium species thereto. An anion solution stream is positioned to feed a solution comprising anions onto the ion exchange resin to form a loaded uranium eluant stream. The loaded uranium eluant stream may then be treated to provide a uranium containing product.

Abstract: The present invention relates to a process for the extraction of metals from an aqueous phase by means of specific ionic liquids.

Abstract: The invention relates to a separation and recovery method for radioactive waste slag and specifically relates to a separation and recovery method for monazite slag. The separation and recovery method comprises the following steps: acid leaching, pressure filtration, water washing, extraction of valuable components and treatment of filtration slag.

Abstract: A method for recovering heavy metals and rare earth elements from fly ash, coal ash, and unrefined mineral ores containing rare earth metals using an ionic liquid and an organic acid to solubilize the metals. The solubilized components are removed from the ionic liquid by electrochemical deposition. The heavy metals and rare earth elements are deposited onto an electrode, and then purified via electrochemical processing.

Abstract: A multi-layered filtration medium for detoxification of chemical contaminants and disinfection of biological contaminants. The filtration medium includes two disinfection nanofiber layers and one detoxification nanofiber layer disposed between the two disinfection nanofiber layers. The filtration medium is loaded with high content of a detoxifying material which capable of achieving about 95% detoxification efficiency with no leaching of the detoxifying material. The filtration medium may also include a medicinal substance for medical applications and a sensor for indicating the lifetime of the filtration medium.

Abstract: The present invention relates to a process of separating and purifying thorium by solvent extraction, comprising: 1) mixing a thorium enrichment with an inorganic acid to produce a feed; 2) mixing a neutral phosphorus extractant with an organic solvent to obtain an organic phase; 3) extracting from the feed with the organic phase to obtain a loaded organic phase; 4) scrubbing the loaded organic phase with a scrubbing solution and then back-extracting thorium with a stripping solution to obtain a thorium solution; 5) mixing the thorium solution with an oxalate to obtain a precipitate, which is then sintered to obtain thorium oxide. The present process allows to increase the purity of thorium from 80%-99% to 99.99% or more with a yield of more than 98%.

Abstract: An apparatus for automated inspection and repair of a tube sheet. The apparatus has a rotating gripper pod, comprising at least one tube gripper, a sliding body portion containing the gripper pod; a housing portion comprising at least one tube gripper and a tool head coupling. The tool head coupling swapably attaches to a eddy current test probe and at least one kind of tube repair tool. Novel, auto-locking tube grippers are also disclosed. A serial bus connects electronic modules within the apparatus and also connects the apparatus to an external controller.

Abstract: A process for extracting uranium compounds from wet-process phosphoric acid (WPA) includes lowering iron content of WPA to produce a lowered iron WPA, reducing valency of any remaining iron in the lowered iron WPA to produce a reduced iron valency WPA. Uranium compounds are extracted from the reduced iron valency WPA via a solvent extraction process.

Abstract: An improved process for the preparation of high purity rare metal compounds such as oxides utilizing TBP (Tri-Butyl Phosphate)-nitrate solvent extraction technique adapted to manufacture nuclear grade rare metal compounds such as zirconium oxide wherein the said process substantially aids in reducing the specific generation of ammonium nitrate effluent volume thereby increasing its concentration when the said effluent comprising ammonium nitrate and ammonium sulphate are utilized for stripping of the said rare metal compound from the organic solvent in the said process of production of high purity rare metal oxide powder.

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: A process for extracting uranium compounds from wet-process phosphoric acid (WPA) includes lowering iron content of WPA to produce a lowered iron WPA, reducing valency of any remaining iron in the lowered iron WPA to produce a reduced iron valency WPA. Uranium compounds are extracted from the reduced iron valency WPA via a solvent extraction process.

Abstract: A method for the selective recovery of uranium from a sulphate-based acidic aqueous solution of uranium containing iron and other metals by means of solvent extraction, in which the extractant used in the organic extraction solution is bis(2-ethylhexyl) phosphate and a liquid branched trialkyl phosphine oxide is the modifying agent. It is typical of the method that the uranium concentration in the feed solution is less than 50 mg/l and a reducing agent is introduced into the aqueous and/or extraction solution to prevent the permanent oxidation of iron to trivalent. In the method the majority of the extraction solution is circulated in a circuit consisting of the extraction stage and the storage tank and only a small part of the uranium-loaded extraction solution is routed to scrubbing and stripping.

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: 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 present invention relates to a process of separating and purifying thorium by solvent extraction, comprising: 1) mixing a thorium enrichment with an inorganic acid to produce a feed; 2) mixing a neutral phosphorus extractant with an organic solvent to obtain an organic phase; 3) extracting from the feed with the organic phase to obtain a loaded organic phase; 4) scrubbing the loaded organic phase with a scrubbing solution and then back-extracting thorium with a stripping solution to obtain a thorium solution; 5) mixing the thorium solution with an oxalate to obtain a precipitate, which is then sintered to obtain thorium oxide. The present process allows to increase the purity of thorium from 80%-99% to 99.99% or more with a yield of more than 98%.

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 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: The method of dissolving the solids formed in the apparatus and pipework of a nuclear plant, in which the solids are brought into contact with an aqueous dissolving solution chosen from aqueous solutions of carbonate ions having a concentration of greater than or equal to 0.3M, aqueous solutions of bicarbonate ions, and solutions of a mixture of nitric acid and of a polycarboxylic acid chosen from oxalic acid and triacids.

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: 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 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: Methods of separating actinides from lanthanides are disclosed. A regio-specific/stereo-specific dithiophosphinic acid having organic moieties is provided in an organic solvent that is then contacted with an acidic medium containing an actinide and a lanthanide. The method can extend to separating actinides from one another. Actinides are extracted as a complex with the dithiophosphinic acid. Separation compositions include an aqueous phase, an organic phase, dithiophosphinic acid, and at least one actinide. The compositions may include additional actinides and/or lanthanides. A method of producing a dithiophosphinic acid comprising at least two organic moieties selected from aromatics and alkyls, each moiety having at least one functional group is also disclosed. A source of sulfur is reacted with a halophosphine. An ammonium salt of the dithiophosphinic acid product is precipitated out of the reaction mixture. The precipitated salt is dissolved in ether.

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: Methods of separating actinides from lanthanides are disclosed. A regio-specific/stereo-specific dithiophosphinic acid having organic moieties is provided in an organic solvent that is then contacted with an acidic medium containing an actinide and a lanthanide. The method can extend to separating actinides from one another. Actinides are extracted as a complex with the dithiophosphinic acid. Separation compositions include an aqueous phase, an organic phase, dithiophosphinic acid, and at least one actinide. The compositions may include additional actinides and/or lanthanides. A method of producing a dithiophosphinic acid comprising at least two organic moieties selected from aromatics and alkyls, each moiety having at least one functional group is also disclosed. A source of sulfur is reacted with a halophosphine. An ammonium salt of the dithiophosphinic acid product is precipitated out of the reaction mixture. The precipitated salt is dissolved in ether.

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: 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: Disclosed is a method of removing nitric acid from an aqueous liquid containing various components such as vegetable extract, and nitrate ion is removed selectively without spoiling the taste or other components, by subjecting the aqueous liquid to chromatographic treatment with an amphoteric ion exchanger to separate nitrate ion from other components contained in the aqueous liquid. A nitric acid-reduced drink is produced by preparing a raw drink material comprising an extract or juice of plant tissue; removing nitric acid from the raw drink material with use of the method of removing nitric acid from an aqueous liquid as described above; and preparing a drink using the raw drink material after the removing of nitric acid.

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 and apparatus for removing uranium (IV) and uranium (VI) from sands and soils. The method and device assays a volume of soil and determines a presence and position of uranium enrichment. A concentration of uranium enrichment is determined, and compared to a threshold concentration. The volume of soil is processed and transported to a coarse screen, deck screen and classified in a classifier. In the classifier, the volume of soil is separated into a washed and fines fraction, assayed and transported to a uranium recovery facility, then transported to a silicon removal process and subsequent ammonia and metals removal processes.

Abstract: A process for selectively extracting metal values, including, uranium, thorium, scandium and zirconium, from starting materials which include the metal values. The process is particularly well suited to extracting metal and recovering metal values from tantalum/niobium production process ore residues.

Abstract: A two-cycle countercurrent extraction process for recovery of highly pure uranium from fertilizer grade weak phosphoric acid. The proposed process uses selective extraction using di-(2-ethyl hexyl) phosphoric acid (D2EHPA) and tri-n-butyl phosphate (TBP) with refined kerosene as synergistic extractant system on hydrogen peroxide treated phosphoric acid, and stripping the loaded extract with strong phosphoric acid containing metallic iron to lower redox potential. The loaded-stripped acid is diluted with water back to weak phosphoric acid state and its redox potential raised by adding hydrogen peroxide and re-extracted with same extractant system. This extract is first scrubbed with sulfuric acid and then stripped with alkali carbonate separating iron as a precipitate, treated with sodium hydroxide precipitating sodium uranate, which is re-dissolved in sulfuric acid and converted with hydrogen peroxide to highly pure yellow cake of uranium peroxide.

Abstract: A method of recovering daughter isotopes from a radioisotope mixture. The method comprises providing a radioisotope mixture solution comprising at least one parent isotope. The at least one parent isotope is extracted into an organic phase, which comprises an extractant and a solvent. The organic phase is substantially continuously contacted with an aqueous phase to extract at least one daughter isotope into the aqueous phase. The aqueous phase is separated from the organic phase, such as by using an annular centrifugal contactor. The at least one daughter isotope is purified from the aqueous phase, such as by ion exchange chromatography or extraction chromatography. The at least one daughter isotope may include actinium-225, radium-225, bismuth-213, or mixtures thereof. A liquid-liquid extraction system for recovering at least one daughter isotope from a source material is also disclosed.

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: An improved process of extraction of uranium from phosphoric acid and in particular uranium VI from phosphoric acid especially strong phosphoric acid using a selective synergistic extractant mix of an organo-phosphorous acid and a neutral extraction agent. The process basically involves the steps of extraction comprising contacting said acid with a selective synergistic extractant system of di-nonyl phenyl phosphoric acid (DNPPA) and a neutral agent selected from di-butyl butyl phosphonate (DBBP) and tri-n-octyl phosphine oxide (TOPO); and recovering the uranium values from the loaded organic phase. The above process would provide for an improved process for recovery of uranium both from weak and strong phosphoric acids using a stable and relatively cheap extractant system. The process is directed to improved recovery of U-VI from phosphoric acid by way of a simple, industrially applicable and cost-effective process.

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.