Abstract: Microporous glutaraldehyde-crosslinked chitosan sorbents, methods of making and using them, and a generator for the radioisotope 99Mo containing the sorbents.

Abstract: A method for converting depleted uranium tetrafluoride (UF4) to triuranium octaoxide (U3O8), and producing sulfur tetrafluoride, using a two step process. The first step uses heat and a mixture of the uranium tetrafluoride and an alkaline compound, either an alkaline oxide or an alkaline hydroxide, to produce U3O8 and a water-soluble metal halide. The second step uses heat, sulfur and a halogen to produce sulfur tetrafluoride and triuranium octaoxide.

Abstract: A method is presented for producing and isolating an isotopically enriched compound of a desired isotope from a parent radionuclide. The method includes forming, or placing, a precipitate containing a parent radionuclide of the desired daughter isotope in a first reaction zone and allowing sufficient time for the parent to decay into the desired gaseous daughter radioisotope. The method further contemplates collecting the desired daughter isotope as a solid in a second reaction zone through the application of temperatures below the freezing point of the desired isotope to a second reaction zone that is connected to the first reaction zone. Specifically, a method is presented for producing isotopically enriched compounds of xenon, including the radioactive isotope Xe-131m and the stable isotope Xe-131.

Abstract: Provided are methods to separate an isotope from a first solution including uranium. The methods may include (a) cleaning the first solution to form a second solution including the uranium and a third solution including the isotope; (b) oxidizing the third solution to form an oxidized isotope; and (c) separating the oxidized isotope.

Abstract: A solid composition comprises: —MnO2; and—a compound represented by the general formula (I) wherein: R is a polymer; each Y is independently a hydrogen or a negative charge; Z is either hydrogen or is not present; each n is independently 1, 2, 3, 4, 5 or 6; wherein the MnO2 is bound to the compound of formula (I) so as to coat the surface thereof. Such a composition may be used for the separation of polyvalent metal species, such as Mo, from one or more accompanying impurities.

Abstract: The invention is directed to a method for the purification of Radium, in particular 226Ra, for target preparation for an essentially pure 225Ac production from available radioactive sources, using an extraction chromatography in order to separate chemically similar elements such as Ba, Sr, and Pb from the desired Ra; wherein said extraction chromatography has an extractant system on the basis of a crown ether. The invention is related to a method for recycling of 226Ra, for target preparation for 225Ac production from radium sources irradiated with accelerated protons (p,2n), after separation of the produced 225Ac. In this method a combination of the above extraction chromatography and a cation exchange chromatography is used. The obtained 226Ra is essentially free of the following chemical contaminants consisting of Ag, Al, As, Be, Bi, Ca, Cd, Co, Cr, Cu, Fe, Ga, K, Li, Mg, Mn, Na, Ni, Pb, Sr, V, Zn, and Ba.

Abstract: A method used to yield irradiation product with minimal impurity for the solid target for gallium (Ga)-68/germanium (Ge)-68 generator mainly consists of the procedures: first calculate the thickness d for the electroplated gallium (Ga)-69 on the solid target; and then through a graph of decay curves including 69Ga(p, 2n) 68Ge target thickness and incident energy with 5 different incident energy doses, derive the corresponding irradiation energy dose Yi for each group after decay; and through the graph including 69Ga(p,2n)68Ge incident energy and reaction cross-sectional area, derive the nuclear reaction cross-sectional area for each group for germanium(Ge)-68, gallium (Ga)-68, zinc (Zn)-65 and figure out the mean reaction area (MRA) from the reaction cross-sectional area of each group; and select the maximum germanium(Ge)-68 MRA value and the minimum gallium (Ga)-68 and zinc (Zn)-65 MRA values; and generate the required default irradiation energy for the MRA of each group.

Abstract: The present invention relates to a method of obtaining 68Ga from a 68Ge/68Ga radioisotope generator and a method of preparing 68Ga-radiolabelled complexes using the obtained 68Ga. The method comprises elution of the generator with an aqueous chloride ion solution. The invention further relates to an apparatus for carrying out the 68Ga metal complex formation.

Abstract: A method and apparatus for the demilitarization of ammunition. The apparatus is comprised of an elongated tubular ammunition conveying chamber having a wall, an inlet opening, and a discharge opening; a heater in thermal communication with the elongated tubular chamber; and a first discharge barrier obstructing at least a first portion of the discharge opening of the elongated tubular chamber.

Abstract: A mixed extractant solvent that includes at least one dialkyloxycalix[4]arenebenzocrown-6 compound, 4?,4?,(5?)-di-(t-butyldicyclohexano)-18-crown-6, at least one modifier, and, optionally, a diluent. The dialkyloxycalix[4]arenebenzocrown-6 compound is 1,3-alternate-25,27-di(octyloxy)calix[4]arenebenzocrown-6, 1,3-alternate-25,27-di(decyloxy)calix[4]arenebenzocrown-6, 1,3-alternate-25,27-di(dodecyloxy)calix[4]arenebenzocrown-6, 1,3-alternate-25,27-di(2-ethylhexyl-1-oxy)calix[4]arenebenzocrown-6, 1,3-alternate-25,27-di(3,7-dimethyloctyl-1-oxy)calix[4]arenebenzocrown-6, 1,3-alternate-25,27-di(4-butyloctyl-1-oxy)calix[4]arenebenzocrown-6, or combinations thereof. The modifier is a primary alcohol. A method of separating cesium and strontium from an aqueous feed is also disclosed, as are dialkyloxycalix[4]arenebenzocrown-6 compounds and an alcohol modifier.

Abstract: The present application is directed to a method for the purification of Radium, in particular 226Ra, for target preparation for an essentially pure 225Ac production from available radioactive sources, using an extraction chromatography in order to separate chemically similar elements such as Ba, Sr, and Pb from the desired Ra; wherein said extraction chromatography has an extractant system on the basis of a crown ether. The invention is further related to a method for recycling of 226Ra, for target preparation for 225Ac production from radium sources irradiated with accelerated protons (p,2n), after separation of the produced 225AC. In this method a combination of the above extraction chromatography and a cation exchange chromatography is used. The obtained 226Ra is essentially free of the following chemical contaminants consisting of Ag, Al, As, Be, Bi, Ca, Cd, Co, Cr, Cu, Fe, Ga, K, Li, Mg, Mn, Na, Ni, Pb, Sr, V, Zn, and Ba.

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: The invention relates to a process for separating, in an aqueous medium, at least one actinide element from one or more lanthanide elements by using at least one molecule which sequesters the said actinide element to be separated and membrane filtration, the said process successively comprising: a) a step of bringing at least one molecule which sequesters the said actinide element in contact with the aqueous medium, the said molecule not being retained in the non-complexed state by the said membrane and being capable of forming a complex with the actinide element to be separated, comprising the said element and at least two of the said sequestering molecules, which complex is capable of being retained by the membrane; b) a step of passing the aqueous medium over the membrane in order to form a permeate on one side, comprising an aqueous effluent depleted of the said actinide element, and a retentate comprising the said complex.

Abstract: Cd-112 isotope is recycled from a Cd-112 chemical separated solution or a remainder of an electroplating solution having a Cd-112 target. The present invention recycles Cd-112 isotope with a low cost, a high purity and a high recycle rate. The recycled Cd-112 isotope can be easily stored. And, the Cd-112 isotope can be used as an imaging agent in nuclear medicine.

Abstract: At least one system for eluting a radioactive material and a method of eluting a radioactive material is provided. The system for eluting a radioactive material may include an elution column configured to enclose an radioactive material, a first sealing member sealing a first end of the elution column, a second sealing member sealing a second end of the elution column, an elution supply source connected to the first end of the elution column via a first needle, a collection system connected to the second end of the elution column via a second needle, and a filter in the elution column, the filter being configured to support the radioactive material and prevent the radioactive material from contacting the second needle.

Abstract: A method for the extraction and purification of molybdenum, the method comprising the steps of: transferring an irradiated fuel solution to an extraction system, the irradiated fuel solution comprising iodine and molybdenum and other fission products, the extraction system comprising at least one sorbent column; passing the irradiated fuel solution upwards through the at least one sorbent-containing extraction column; directing the irradiated fuel solution towards a fuel management system by means of at least one discharge alignment valve; directing the extraction column eluate towards an iodine removal system; removing the iodine from the extraction column eluate; purifying the extraction column eluate; and collecting the purified eluate. Also disclosed is an apparatus for accomplishing the aforementioned method.

Abstract: The present invention relates to a method for the volumetric decontamination of radioactive metals. The method includes the step of precipitating out radioactive gamma and beta emitting nucleotides and then recovering non-radioactive metal compounds.

Abstract: Disclosed is a method for rapidly separating radioactive copper from nickel that contains radioactive copper and radioactive cobalt, which comprises dissolving nickel that contains radioactive copper and radioactive cobalt in an acid solution and leading it to pass through a chelating-ion exchange resin-filled column to thereby make nickel, radioactive copper and radioactive cobalt held by the chelating-ion exchange resin, and then leading an acid solution to pass through the chelating-ion exchange resin-filled column to elute nickel and radioactive cobalt, and thereafter leading an acid solution having a higher concentration than that of the previous acid solution to pass through the chelating-ion exchange resin-filled column after nickel and radioactive elution therefrom to thereby elute radioactive copper.

Abstract: The present invention relates to a treatment of high-level waste of radiochemical production containing radionuclides and macro-admixtures including sodium. The method of extraction of radionuclides by processing acidic aqueous waste solutions by extractants containing macrocyclic compounds selected from the group of crown ethers having aromatic fragments containing alkyl and/or hydroxyalkyl substituents of a linear and/or branched structure, and/or cyclohexane fragments containing alkyl and/or hydroxyalkyl substituents of a linear and/or branched structure, and/or fragments of —O—CHR—CH2O—, where R is the normal or branched alkyl or hydroxyalkyl in organic solvents containing polyfluorinated telomeric alcohol 1,1,7-trihydrododecafluoroheptanol-1 having the formula H(CF2CF2)nCH2OH, where n=3, and a mixture of polyoxyethylene glycol ethers of synthetic primary higher aliphatic alcohols of a fraction C12-C14 of a general formula CnH2n+1O(C2H4O)mH, where n=12-14, m=2 is proposed.

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 solid composition comprises: —MnO2; and—a compound represented by the general formula (I) wherein: R is a polymer; each Y is independently a hydrogen or a negative charge; Z is either hydrogen or is not present; each n is independently 1, 2, 3, 4, 5 or 6; wherein the MnO2 is bound to the compound of formula (I) so as to coat the surface thereof. Such a composition may be used for the separation of polyvalent metal species, such as Mo, from one or more accompanying impurities.

Abstract: The present invention is directed to a generator apparatus for separating a daughter gallium-68 radioisotope substantially free of impurities from a parent gernanium-68 radioisotope, including a first resin-containing column containing parent gernanium-68 radioisotope and daughter gallium-68 radioisotope, a source of first eluent connected to said first resin-containing column for separating daughter gallium-68 radioisotope from the first resin-containing column, said first eluent including citrate whereby the separated gallium is in the form of gallium citrate, a mixing space connected to said first resin-containing column for admixing a source of hydrochloric acid with said separated gallium citrate whereby gallium citrate is converted to gallium tetrachloride, a second resin-containing column for retention of gallium-68 tetrachloride, and, a source of second eluent connected to said second resin-containing column for eluting the daughter gallium-68 radioisotope from said second resin-containing column.

Abstract: The present application is directed to a method for the purification of Radium, in particular 226Ra, for target preparation for an essentially pure 225Ac production from available radioactive sources, using an extraction chromatography in order to separate chemically similar elements such as Ba, Sr, and Pb from the desired Ra; wherein said extraction chromatography has an extractant system on the basis of a crown ether. The invention is further related to a method for recycling of 226Ra, for target preparation for 225Ac production from radium sources irradiated with accelerated protons (p,2n), after separation of the produced 225AC. In this method a combination of the above extraction chromatography and a cation exchange chromatography is used. The obtained 226Ra is essentially free of the following chemical contaminants consisting of Ag, Al, As, Be, Bi, Ca, Cd, Co, Cr, Cu, Fe, Ga, K, Li, Mg, Mn, Na, Ni, Pb, Sr, V, Zn, and Ba.

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: 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 radioisotope TI-201 is produced. The process includes electroplating, irradiating, dissolving, precipitating, ion exchanging, decaying and filtering. The TI-201 obtained is a liquid having a high purity.

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: A separation medium, a method for using that separation medium and an apparatus for selectively extracting multivalent cations such as pseudo-lanthanide, prelanthanide, lanthanide, preactinide or actinide cations from an aqueous acidic sample solution is described. The separation medium is preferably free-flowing and comprises particles having a diglycolamide (DGA) extractant dispersed onto an inert, porous support.

Abstract: The present invention provides a method of separating and purifying Cesium-131 (Cs-131) from Barium (Ba). Uses of the Cs-131 purified by the method include cancer research and treatment, such as for the use in brachytherapy. Cs-131 is particularly useful in the treatment of faster growing tumors.

Abstract: An extractant composition comprising a mixed extractant solvent consisting of calix[4] arene-bis-(tert-octylbenzo)-crown-6 (“BOBCalixC6”), 4?,4?,(5?)-di-(t-butyldicyclo-hexano)-18-crown-6 (“DtBu18C6”), and at least one modifier dissolved in a diluent. The DtBu18C6 may be present at from approximately 0.01M to approximately 0.4M, such as at from approximately 0.086 M to approximately 0.108 M. The modifier may be 1-(2,2,3,3-tetrafluoropropoxy)-3-(4-sec-butylphenoxy)-2-propanol (“Cs-7SB”) and may be present at from approximately 0.01M to approximately 0.8M. In one embodiment, the mixed extractant solvent includes approximately 0.15M DtBu18C6, approximately 0.007M BOBCalixC6, and approximately 0.75M Cs-7SB modifier dissolved in an isoparaffinic hydrocarbon diluent. The extractant composition further comprises an aqueous phase. The mixed extractant solvent may be used to remove cesium and strontium from the aqueous phase.

Abstract: The present invention provides a method of separating and purifying Yttrium-90 (Y-90) from Strontium-90 (Sr-90). In addition, a zirconium (Zr) clean-up step for the Y-90 is provided. Uses of the Y-90 purified by the method include cancer research and treatment. Y-90 is particularly useful in cell directed therapy, e.g., where the Y-90 is attached directly or indirectly to a targeting molecule such as an antibody.

Abstract: The present invention provides a method for improving the recovery of cesium-131 (Cs-131) from barium (Ba) carbonate. Uses of the Cs-131 purified by the method include cancer research and treatment, such as for the use in brachytherapy. Cesium-131 is particularly useful in the treatment of faster growing tumors.

Abstract: A new method to strip metals from organic solvents in a manner that allows for the recycle of the stripping agent. The method utilizes carbonate solutions of organic amines with complexants, in low concentrations, to strip metals from organic solvents. The method allows for the distillation and reuse of organic amines. The concentrated metal/complexant fraction from distillation is more amenable to immobilization than solutions resulting from current practice.

Abstract: The present invention provides a method of separating and purifying Cesium-131 (Cs-131) from Barium (Ba). Uses of the Cs-131 purified by the method include cancer research and treatment, such as for the use in brachytherapy. Cs-131 is particularly useful in the treatment of faster growing tumors.

Abstract: Sodium nonatitanate compositions, a method using the composition for recovery of 82Sr from irradiated targets, and a method using the composition for generating 82Rb. The sodium nonatitanate materials of the invention are highly selective at separating strontium from solutions derived from the dissolution of irradiated target materials, thus reducing target processing times. The compositions also have a very low affinity for rubidium, making it an ideal material for use as a 82Rb generator. Sodium nonatitanate materials of this type both improve the recovery of 82Sr and provide a safer, more effective 82Rb generator system.

Abstract: The present invention provides a method for producing a sealed 210Pb—210Po ? source (? particle emitter) and an apparatus thereof, which can be used as an ? particle source for a random pulse generator. The method for producing a sealed 210Pb—210Po ? source (? particle emitter) includes the steps of: collecting 210Pb—210Po with a 210Pb collector using radon collection; precipitating the hydroxides of the collected 210Pb—210Po and collecting the precipitates by a polycarbonate (PC) filter; dissolving the 210Pb—210Po hydroxide precipitate to form a 210Pb—210Po radioactive thin film; and sealing the 210Pb—210Po radioactive thin film for protection.

Abstract: A method of producing and purifying promethium-147 including the steps of: irradiating a target material including neodymium-146 with neutrons to produce promethium-147 within the irradiated target material; dissolving the irradiated target material to form an acidic solution; loading the acidic solution onto a chromatographic separation apparatus containing HDEHP; and eluting the apparatus to chromatographically separate the promethium-147 from the neodymium-146.

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 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: 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 mixed extractant solvent including calix[4]arene-bis-(tert-octylbenzo)-crown-6 (“BOBCalixC6”), 4?,4?,(5?)-di-(t-butyldicyclo-hexano)-18-crown-6 (“DtBu18C6”), and at least one modifier dissolved in a diluent. The mixed extractant solvent may be used to remove cesium and strontium from an acidic solution. The DtBu18C6 may be present from approximately 0.01 M to approximately 0.4M, such as from approximately 0.086 M to approximately 0.108 M. The modifier may be 1-(2,2,3,3-tetrafluoropropoxy)-3-(4-sec-butylphenoxy)-2-propanol (“Cs-7SB”) and may be present from approximately 0.01M to approximately 0.8M. In one embodiment, the mixed extractant solvent includes approximately 0.15M DtBu18C6, approximately 0.007M BOBCalixC6, and approximately 0.75M Cs-7SB modifier dissolved in an isoparaffinic hydrocarbon diluent. The mixed extractant solvent may form an organic phase in an extraction system that also includes an aqueous phase. Methods of extracting cesium and strontium as well as strontium alone are also disclosed.

Abstract: Undesirable metal cation contaminants, including hafnium, can be reduced from a solution containing Lutetium-177 on a bed of an anion exchange resin. The thus purified solution can be stored and transported in a polypropylene vial, to prevent the possibility of hafnium entering the solution from a glassware wall and of Lutetium from being lost from the solution by chemically reacting with a glassware wall. The vial can be sealed with a rubber stopper which has a polytetrafluoroethylene coating facing the solution to prevent the possibility of contaminants which could interfere with later uses of the Lutetium-177 from leaching into the solution from the rubber stopper. Recipients of the Lutetium-177-containing vial can be provided with a prepackaged column of the anion exchange resin to enable such recipients to remove the hafnium which accumulated in the solution as a product of radioactive decay during shipment.

Abstract: A bismuth-213 generator comprising an insoluble composition having the general formula Zr(Phosponate)x(HPO4)2?x.nH2O, wherein x is between 0 and 2; and n is the number of waters of hydration; and wherein cations of radioactive isotopes selected from radium, actinium and combinations thereof are immobilized on the composition. The value of x may be between about 0.2 and about 1. The phosphonate may be n-phosphonomethyl-miniodiacetic acid (PMIDA), wherein x may be between about 0.1 and about 1.9. The phosphonate may be one or more phosphonate having the formula: H2O3P—(CH2)a—N—((CH2)bCO2H)—((CH2)cCO2H), wherein a, b, and c are numbers from 1 to 3 that may or may not be equal. The value of x may also be between about 0.1 and 1.9.

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: 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 separation medium, a method for using that separation medium and an apparatus for selectively extracting multivalent cations such as pseudo-lanthanide, prelanthanide, lanthanide, preactinide or actinide cations from an aqueous acidic sample solution is described. The separation medium is preferably free-flowing and comprises particles having a diglycolamide (DGA) extractant dispersed onto an inert, porous support.

Abstract: Methods and apparatus for separating ions of metallic elements are provided. Preferred methods utilize a hydrophobic chelating extractant, such as an organophosphorus compound, adsorbed onto carbon or graphite fibers in the form of felt. Also described is a new thallium-201 generator that comprises a column containing an acidic organophosphorus extractant adsorbed on carbon or graphite fibers, and a yttrium-90 generator system comprised of two extraction columns designed to selectively absorb yttrium-90 at different pH, to enable the separation of yttrium-90 from strontium-90. The two columns are connected in series for stepwise separation. The yttrium-90 product is freed from residual strontium-90 and metal contaminants and can be eluted from the second column with dilute acid, acetate buffer, water or saline for labeling biological targeted molecules. The new generator system provides rapid and efficient separation of yttrium-90 and is amenable to both scale-up and automation.