Abstract: The embodiments of the present disclosure provide a method for producing Ac-225 from Ra-226, comprising submitting Ra-226 to a photo-nuclear process, collecting an electrochemical precipitation of an Ac-225 on a cathode in a recipient, removing the cathode from the recipient after the electrochemical precipitation of the Ac-225, transferring the cathode to a hot cell environment, and extracting the Ac-225 from the cathode in the hot cell environment. The Ra-226 may comprise a liquid solution in the recipient, and submitting Ra-226 to the photo-nuclear process may comprise irradiating the Ra-226 to produce Ra-225. The Ra-225 may decay into Ac-225 upon irradiation of the Ra-226.

Abstract: Method and apparatus for analysis and display of fine grained mineral samples. A portion of the sample is illuminated with a charged particle beam. Emitted radiation is detected, and a sample emission spectrum is generated and fit with a plurality of standard emission spectra of minerals in a candidate mineral composition. A mineral composition whose emission spectrum best fits the sample emission spectrum is selected from a plurality of candidate mineral compositions. An assigned color is received for each mineral in the selected mineral composition, and the assigned colors are blended according to the proportion of each mineral in the selected mineral composition. An image pixel corresponding to the portion of the sample is rendered for display.

Abstract: The present invention relates generally to the field of medical isotope production by fission of uranium-235 and the fuel utilized therein (e.g., the production of suitable Low Enriched Uranium (LEU is uranium having 20 weight percent or less uranium-235) fuel for medical isotope production) and, in particular to a method for producing LEU fuel and a LEU fuel product that is suitable for use in the production of medical isotopes. In one embodiment, the LEU fuel of the present invention is designed to be utilized in an Aqueous Homogeneous Reactor (AHR) for the production of various medical isotopes including, but not limited to, molybdenum-99, cesium-137, iodine-131, strontium-89, xenon-133 and yttrium-90.

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 the use of a compound of the formula KmgF3 to trap metals in the form of fluorides and/or of oxyfluorides in a gaseous or liquid phase. It also relates to a compound of the formula KMgF3 which has a surface area at least equal to 30 m2/g and at most equal to 150 m2/g and also to its methods of preparation. The invention notably finds application in the nuclear industry, in which it can advantageously be used to purify uranium hexafluoride (UF6) present in a gaseous or liquid stream, with regard to metal impurities which are also present in this stream.

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: The invention relates to the use of a compound of formula KMgF3 to trap metals present in the form of fluorides and/or of oxyfluorides in a gaseous or liquid phase. It also relates to a compound of formula KMgF3 which has a surface specific area at least equal to 30 m2/g and at most equal to 150 m2/g and also to its methods of preparation. The invention notably finds application in the nuclear industry, in which it can advantageously be used to purify uranium hexafluoride (UF6) present in a gaseous or liquid stream, with regard to metal impurities which are also present in this stream.

Abstract: A process for the extraction of uranium compounds from wet-process phosphoric acid includes lowering the iron concentration of the wet-process phosphoric acid and reducing the valency of any remaining ferric iron in the wet-process phosphoric acid to ferrous iron, and then extracting uranium compounds from the wet-process phosphoric acid. The process can include separating a side stream from a feed stream of wet-process phosphoric acid, wherein the side stream has a greater concentration of the uranium compounds than the feed stream by filtration. Extracting uranium compounds from the wet-process phosphoric acid can be by ion exchange process or by solvent extraction.

Abstract: Compositions are provided that include nuclear fuel. Methods for treating nuclear fuel are provided which can include exposing the fuel to a carbonate-peroxide solution. Methods can also include exposing the fuel to an ammonium solution. Methods for acquiring molybdenum from a uranium comprising material are provided.

Abstract: A process for treating a feedstock comprising tantalum- and/or niobium-containing compounds is provided. The process includes contacting the feedstock with a gaseous fluorinating agent, thereby to fluorinate tantalum and/or niobium present in the feedstock compounds. The resultant fluorinated tantalum and/or niobium compounds are recovered.

Abstract: The present invention provides an apparatus and rapid methods for extracting strontium ions from urine to provide a concentrated and purified strontium-90 extract suitable for scintillation measurements. The methods remove organic compounds, pigments, and alkali metal ions that can interfere with quantitative determination of strontium-90 in urine.

Abstract: A method for the production of UCl3 salt without the use of hazardous chemicals or multiple apparatuses for synthesis and purification is provided. Uranium metal is combined in a reaction vessel with a metal chloride and a eutectic salt- and heated to a first temperature under vacuum conditions to promote reaction of the uranium metal with the metal chloride for the production of a UCl3 salt. After the reaction has run substantially to completion, the furnace is heated to a second temperature under vacuum conditions. The second temperature is sufficiently high to selectively vaporize the chloride salts and distill them into a condenser region.

Abstract: Method of producing anhydrous thorium(IV) tetrahalide complexes, utilizing Th(NO3)4(H2O)x, where x is at least 4, as a reagent; method of producing thorium-containing complexes utilizing ThCl4(DME)2 as a precursor; method of producing purified ThCl4(ligand)x compounds, where x is from 2 to 9; and novel compounds having the structures:

Abstract: A precipitator comprises a counter-current circulation between the reacting substances (8, 6, 7 ?12) and a non-miscible and chemically inert organic confinement phase (10?9) to maintain the phase containing the reagent in an emulsion. The walls of the precipitator are hydrophobic to prevent the adhesion of the precipitate. The emulsion is maintained by a mobile stirrer body (2). The precipitate is removed continually by a scavenging flow rate device (16).

Abstract: Lithium salts with fluorinated chelated orthoborate anions are prepared and used as electrolytes or electrolyte additives in lithium-ion batteries. The lithium salts have two chelate rings formed by the coordination of two bidentate ligands to a single boron atom. In addition, each chelate ring has two oxygen atoms bonded to one boron atom, methylene groups bonded to the two oxygen atoms, and one or more fluorinated carbon atoms bonded to and forming a cyclic bridge between the methylene groups.

Abstract: A process for the extraction of uranium compounds from wet-process phosphoric acid includes lowering the iron concentration of the wet-process phosphoric acid and reducing the valency of any remaining ferric iron in the wet-process phosphoric acid to ferrous iron, and then extracting uranium compounds from the wet-process phosphoric acid. The process can include separating a side stream from a feed stream of wet-process phosphoric acid, wherein the side stream has a greater concentration of the uranium compounds than the feed stream by filtration. Extracting uranium compounds from the wet-process phosphoric acid can be by ion exchange process or by solvent extraction.

Abstract: The present invention provides a process for the production of a uranium trioxide yellowcake from a uranium peroxide precipitate, the peroxide precipitate being in the form of a low solids content, uranium rich feed slurry, the process including the stages of: a. thickening the feed slurry to produce a thickener underflow with a solids content in the range of 15 to 50% w/w and a thickener overflow; b. dewatering the thickener underflow to produce a solids cake with a solids content of at least 50% w/w and a dewater overflow; and c. calcining the solids cake at a temperature in the range of 450° C. to 480° C. to produce a calcined uranium trioxide yellowcake.

Abstract: A process for the extraction of uranium compounds from wet-process phosphoric acid includes lowering the iron concentration of the wet-process phosphoric acid and reducing the valency of any remaining ferric iron in the wet-process phosphoric acid to ferrous iron, and then extracting uranium compounds from the wet-process phosphoric acid. The process can include separating a side stream from a feed stream of wet-process phosphoric acid, wherein the side stream has a greater concentration of the uranium compounds than the feed stream by filtration. Extracting uranium compounds from the wet-process phosphoric acid can be by ion exchange process or by solvent extraction.

Abstract: Systems for treating material are provided that can include a vessel defining a volume, at least one conduit coupled to the vessel and in fluid communication with the vessel, material within the vessel, and NF3 material within the conduit. Methods for fluorinating material are provided that can include exposing the material to NF3 to fluorinate at least a portion of the material. Methods for separating components of material are also provided that can include exposing the material to NF3 to at least partially fluorinate a portion of the material, and separating at least one fluorinated component of the fluorinated portion from the material. The materials exposed to the NF3 material can include but are not limited to one or more of U, Ru, Rh, Mo, Tc, Np, Pu, Sb, Ag, Am, Sn, Zr, Cs, Th, and/or Rb.

Abstract: A method of extracting uranium from spent nuclear fuel (SNF) particles is disclosed. Spent nuclear fuel (SNF) (containing oxides of uranium, oxides of fission products (FP) and oxides of transuranic (TRU) elements (including plutonium)) are subjected to a hydrogen plasma and a fluorine plasma. The hydrogen plasma reduces the uranium and plutonium oxides from their oxide state. The fluorine plasma etches the SNF metals to form UF6 and PuF4. During subjection of the SNF particles to the fluorine plasma, the temperature is maintained in the range of 1200-2000 deg K to: a) allow any PuF6 (gas) that is formed to decompose back to PuF4 (solid), and b) to maintain stability of the UF6. Uranium (in the form of gaseous UF6) is easily extracted and separated from the plutonium (in the form of solid PuF4). The use of plasmas instead of high temperature reactors or flames mitigates the high temperature corrosive atmosphere and the production of PuF6 (as a final product).

Abstract: The invention relates to a method for the removal of uranium(VI) species from waters by means of weakly basic, polyacrylic-based anion exchangers, said uranium(VI) species being present in the form of uranyl complexes as dissolved uranyl.

Abstract: Devices, systems, and methods relating to advanced, high pressure oxidation are described. The devices, systems, and methods can be used to decontaminate ground water in a well or opening in a ground water table, and to recover minerals and hydrocarbons from subterranean deposits.

Abstract: Method for coprecipitation (or simultaneous precipitation) of at least one actinide in oxidation state (IV) with at least one actinide in oxidation state (III), wherein: a solution i.e. mixture of actinide(s) in oxidation state (IV) and actinide(s) in oxidation state (III) is prepared by adding to it a singly charged cation whose presence makes it possible to stabilize the aforementioned oxidation states in the mixture, or a singly charged cation which does not act to stabilize the aforementioned oxidation states in the mixture; a solution containing oxalate ions is mixed with the said mixture of actinides in order to carry out coprecipitation, i.e. simultaneous precipitation, of the said actinides in oxidation states (IV) and (III) and a fraction of the singly charged cation. According to another embodiment, a solution i.e.

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: A process for separation of no-carrier-added thallium radionuclide from no-carrier-added lead and mercury comprising providing a solution of no-carrier-added thallium radionuclide and no-carrier-added lead and mercury to dialysis. By this method separation of 199Tl radionuclides has also been achieved in presence of macro quantity of inactive thallium, which is as high as 10 mM. The method is capable of being used in Medical industry, diagnosis of cardiac diseases by 201Tl or 199Tl and all other industries where trace amount of thallium separation is required from mercury and lead.

Abstract: Methods and systems for removing copper minerals from a molybdenite concentrate. One embodiment provides leaching copper from the molybdenite concentrate with a leaching solution comprising ferric chloride, removing molybdenite from the leaching solution, introducing an acid into the leaching solution and introducing O2, O3, or a combination of both, into the leaching solution. A method for regenerating ferric chloride in a leaching solution is also provided. One embodiment provides adding a leaching solution comprising Fe(II) ions, Fe(III) ions, or a combination of both, to a mixture of mineral sulfides, introducing an acid into the leaching solution, and introducing O2, O3, or a combination of both, into the leaching 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 product includes actinium-225 (225Ac) and less than about 1 microgram (?g) of iron (Fe) per millicurie (mCi) of actinium-225. The product may have a radioisotopic purity of greater than about 99.99 atomic percent (at %) actinium-225 and daughter isotopes of actinium-225, and may be formed by a method that includes providing a radioisotope mixture solution comprising at least one of uranium-233 (233U) and thorium-229 (229Th), extracting the at least one of uranium-233 and thorium-229 into an organic phase, substantially continuously contacting the organic phase with an aqueous phase, substantially continuously extracting actinium-225 into the aqueous phase, and purifying the actinium-225 from the aqueous phase. In some embodiments, the product may include less than about 1 nanogram (ng) of iron per millicurie (mCi) of actinium-225, and may include less than about 1 microgram (?g) each of magnesium (Mg), Chromium (Cr), and manganese (Mn) per millicurie (mCi) of actinium-225.

Abstract: The invention relates to a process for the chemical beneficiation of raw material containing tantalum-niobium such as wastes, scoria, concentrates and ores.

Abstract: A process for the drying of yellowcake, the yellowcake initially being in the form of a low solids content, uranium rich feed slurry, the process including the stages of: a. dewatering the feed slurry to produce a first dewatered solids cake with a solids content higher than the feed slurry; b. re-slurrying the first dewatered solids cake with sufficient water to dissolve at least some impurities and to produce an intermediate slurry with a solids content lower than the first dewatered solids cake; c. dewatering the intermediate slurry to produce a second dewatered solids cake with a solids content higher than the feed slurry; and d. drying the second dewatered solids cake to produce dried yellowcake.

Abstract: A process for the extraction of uranium compounds from wet-process phosphoric acid includes lowering the iron concentration of the wet-process phosphoric acid and reducing the valency of any remaining ferric iron in the wet-process phosphoric acid to ferrous iron, and then extracting uranium compounds from the wet-process phosphoric acid. The process can include separating a side stream from a feed stream of wet-process phosphoric acid, wherein the side stream has a greater concentration of the uranium compounds than the feed stream by filtration. Extracting uranium compounds from the wet-process phosphoric acid can be by ion exchange process or by solvent extraction.

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: 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 of separating isotopes from a mixture containing at least two isotopes in a solution is disclosed. A first isotope is precipitated and is collected from the solution. A daughter isotope is generated and collected from the first isotope. The invention includes a method of producing an actinium-225/bismuth-213 product from a material containing thorium-229 and thorium-232. A solution is formed containing nitric acid and the material containing thorium-229 and thorium-232, and iodate is added to form a thorium iodate precipitate. A supernatant is separated from the thorium iodate precipitate and a second volume of nitric acid is added to the thorium iodate precipitate. The thorium iodate precipitate is stored and a decay product comprising actinium-225 and bismuth-213 is generated in the second volume of nitric acid, which is then separated from the thorium iodate precipitate, filtered, and treated using at least one chromatographic procedure.

Abstract: Medical isotope generator systems are disclosed according to some aspects. In one aspect, a 90Y generator system comprises a generator column, a concentration column, and a flow control system, through which the generator column and the concentration column are in fluid communication. The flow control system provides a plurality of flow configurations for delivering a milking solution to the generator column, the concentration column, or both, and for delivering an eluent solution to the concentration column in either a forward or a reverse flow direction. The generator column can comprise a 90Sr stock adsorbed on a sorbent. The milking solution preferentially elutes 90Y from the generator column. The concentration column comprises a sorbent that captures 90Y from the milking solution without altering the milking solution. The eluent solution elutes 90Y from the concentration column.

Abstract: The invention provides a series of techniques for processing uranium containing feed materials such as uranium ores, reprocessed uranium, uranium containing residues and uranium containing spent fuel. The processes described involve fluorination of uranium containing material, separation of the uranium containing material from other materials based on ionization thereof with the non-ionized fluorine containing material being recycled. Metallic uranium and/or plutonium and/or fission products may result. The technique offers advantages in terms of the range of materials which can be reprocessed and a reduction in the number of complexity of stages which are involved in the process.

Abstract: A protective cover and radionuclide generator assembly. The cover protecting inlet and outlet connections disposed on the radionuclide generator. The radionuclide generator having a distal end having a generally flat top surface. The protective cover removably fixed at the distal end and positioned over the generally flat top surface. The protective cover engaging with the radionuclide generator and providing an interference fit. The interference fit preventing the removal of the cover from the radionuclide generator, unless so desired by a user. The cover being made from a radioactive resistant polypropylene.

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: Fluorine or a fluorine compound is subjected to a reaction with a spent oxide fuel to produce fluorides of uranium and plutonium, and recovering the fluorides using a difference in volatility behavior. The method includes steps of: subjecting a mixture of UO2 and PuO2 with hydrogen fluoride mixed with hydrogen to HF-fluorinate uranium and plutonium into UF4 and PuF3; subjecting UF4 and PuF3 with a fluorine gas to F2-fluorinate uranium and plutonium into UF6 and PuF6; and fractionating UF6 and PuF6 using a difference in phase change of obtained UF6 and PuF6, removing a part of UF6, and volatilizing the remaining UF6 and PuF6 at the same time. By such a reprocessing method, PuF4 hard to undergo a reaction is prevented from being formed as an intermediate fluoride, the material of a reactor is hard to be corroded, and a consumption of expensive fluorine gas is reduced.

Abstract: This invention is provided for improvement of corrosion-resistant property of a crucible and for promotion of safety in a pyrochemical reprocessing method for the spent nuclear fuel. The spent nuclear fuel is dissolved in a molten salt placed in the crucible. In a pyrochemical reprocessing method, the nuclear fuel is deposited, and the crucible (2) is heated by induction heating. Cooling media (5, 6) are supplied to cool down, and a molten salt layer (7) is maintained by keeping balance between the heating and the cooling, and a solidified salt layer (8) is formed on inner wall surface of the crucible.

Abstract: A process in which isotopes of the same element belonging to the alkaline earth metals, transition elements and heavy metals having an atomic mass of less than 209, in particular lanthanide metals, are separated in an aqueous medium by treating an aqueous medium.

Abstract: Sulphidation method for a UO2 powder, in which said powder is sulfurated by bringing it into contact with a gaseous sulphidation agent. Method for manufacturing nuclear fuel pellets based on uranium oxide, or mixed oxide of uranium and plutonium, from a load of totally or partially sulfurated UO2 powder or UO2 powder and PuO2 powder, by lubrication, pelletizing and sintering, in which: the load of powder subjected to the lubrication, pelletizing and sintering is prepared by the following successive steps: sulphidation of a UO2 powder by the above sulphidation method; optionally mixing, said sulfurated powder in a matrix comprising a UO2 powder, or of a UO2 powder and a PuO2 powder; and, subjecting said load, formed from said sulfurated powder or said mixture, to lubrication, pelletizing and sintering operations.

Abstract: A method of stabilizing nuclear material is disclosed. Oxides or halides of actinides and/or transuranics (TRUs) and/or hydrocarbons and/or acids contaminated with actinides and/or TRUs are treated by adjusting the pH of the nuclear material to not less than about 5 and adding sufficient MgO to convert fluorides present to MgF2; alumina is added in an amount sufficient to absorb substantially all hydrocarbon liquid present, after which a binder including MgO and KH2PO4 is added to the treated nuclear material to form a slurry. Additional MgO may be added. A crystalline radioactive material is also disclosed having a binder of the reaction product of calcined MgO and KH2PO4 and a radioactive material of the oxides and/or halides of actinides and/or transuranics (TRUs). Acids contaminated with actinides and/or TRUs, and/or actinides and/or TRUs with or without oils and/or greases may be encapsulated and stabilized by the binder.

Abstract: A method of operating a fuel reformer includes advancing a first air/fuel mixture having a first air-to-fuel ratio into the fuel reformer. The method further includes determining if a soot purge is to be performed and generating a purge-soot signal in response thereto. Further, a second air/fuel mixture having a second air-to-fuel ratio is advanced into the fuel reformer in response to generation of the purge-soot signal. The second air-to-fuel ratio is greater than the first air-to-fuel ratio in order to burn soot present within the fuel reformer. A fuel reformer system is also disclosed.

Abstract: Fluorine or a fluorine compound is subjected to a reaction with a spent oxide fuel to produce fluorides of uranium and plutonium, and the fluorides are recovered using a difference in volatility behavior. The spent oxide fuel is subjected to a reaction with an HF gas, whereby uranium, plutonium and most impurities are converted into solid fluorides having low valences or remained as oxides to inhibit volatilization thereof, and then in an F2 fluorination step, the HF fluorination product is subjected to a reaction with a fluorine gas in two stages: one at a low temperature and the other at a high temperature, whereby a certain amount of gaseous uranium and volatile impurities are separated with plutonium kept in a solid form in the first stage, and mixed fluorides of remaining uranium and plutonium are fluorinated into hexafluorides at the same time in the second stage. By such a reprocessing method, plutonium enrichment can be adjusted, uranium and plutonium can be purified, and steps are simplified as well.

Abstract: The invention relates to a process for the preparation of a product based on a phosphate of at least one element M(IV), for example of thorium and/or of actinide(IV)(s). This process comprises the following stages: a) mixing a solution of thorium(IV) and/or of at least one actinide(IV) with a phosphoric acid solution in amounts such that the molar ratio PO 4 M ? ? ( IV ) ?is from 1.4 to 2, b) heating the mixture of the solutions in a closed container at a temperature of 50 to 250° C. in order to precipitate a product comprising a phosphate of at least one element M chosen from thorium(IV) and actinide(IV)s having a P/M molar ratio of 1.5, and c) separating the precipitated product from the solution. The precipitate can be converted to phosphate/diphosphate of thorium and of actinide(s). The process also applies to the separation of uranyl ions from other cations.

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: The present invention generally relates to the preparation of mixed actinide oxides, such as mixed oxides of uranium and plutonium (U, Pu) O2, by simultaneously coprecipitation and then calcinations.

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.