Abstract: Methods for purifying the molybdenum-99 isotope are disclosed. Molybdenum-99 is loaded onto an anion exchange column and extracted. In some embodiments, the extraction solution may include nitric acid and nitrate salts. In other embodiments, a two stage elution is performed in which a nitric acid containing eluent and a hydroxide containing eluent are used in succession to extract molybdenum-99.

Abstract: The invention relates to the use of an organic-inorganic hybrid material, comprising an inorganic solid support on which are grafted organic molecules having the general formula (I) below: in which: x, y and z=0 or 1, with at least one of x, y and z different from 0; m=1 to 6; v and w=0 or 1, with v=1 when w=0, and v=0 when w=1; if x=0, R1?H or a saturated or unsaturated, linear or branched, C1 to C12 hydrocarbon group, whereas, if x=1, R1=a group bound to the inorganic solid support by at least one covalent bond; if y=0, R2?H or a saturated or unsaturated, linear or branched, C1 to C12 hydrocarbon group, whereas, if y=1, R2=a group bound to the inorganic solid support by at least one covalent bond; if z=0, R3?H or a saturated or unsaturated, linear or branched, C1 to C12 hydrocarbon group, whereas, if z=1, R3=a group bound to the inorganic solid support by at least one covalent bond; R4 and R5?H, a saturated or unsaturated, linear or branched, C2 to C8 hydrocarbon group, or a monocyclic aromatic group

Abstract: A method of forming a composite material for use in multi-modal transport includes providing three-dimensional graphene having hollow channels, enabling a polymer to wick into the hollow channels of the three-dimensional graphene, curing the polymer to form a cured three-dimensional graphene, adding an active material to the cured three-dimensional graphene to form a composite material, and removing the polymer from within the hollow channels. A composite material formed according to the method is also provided.

Abstract: Apparatuses, processes and methods for the separation, isolation, or removal of specific radioactive isotopes from liquid radioactive waste, these processes and methods employing isotope-specific media (ISM). In some embodiments, the processes and methods further include the vitrification of the separated isotopes, generally with the ISM; this isotope-specific vitrification (ISV) is often a step in a larger scheme of preparing the radioactive isotopes for long-term storage or other disposition. A variety of ISM are disclosed.

Abstract: Oxine ligands placed on styrene base ion exchange resins selectively remove iron and gallium from acidic solutions. After loading, the oxine resin is stripped of the loaded metals and used again for further metal removal. The resins can be used for process streams, acid rock drainages, or any other iron or gallium containing solution.

Abstract: An apparatus for processing a radionuclide including a parent radionuclide that decays over time into a daughter radionuclide, a separation column that separates the daughter radionuclide from the parent radionuclide, a plurality of valves and at least one pump that operate to separate the daughter radionuclide from the parent radionuclide and deliver the daughter radionuclide into the daughter radionuclide container by alternately connecting at least two of the parent radionuclide container, the daughter radionuclide container, the separation column container and the plurality of processing containers, a plurality of RFID tags including an RFID tag of the plurality of RFID tags affixed to each of the daughter radionuclide container and the separation column and a programmed processor that reads an identifier of each of the plurality of RFID tags, an identifier and position of each of the plurality of valves and pump and saves the identifiers, positions and operations into a tracking file.

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 present invention relates to a calcium carbonate having a surface charge, the preparing process thereof and filler for producing a paper using the same. The present invention provides the process of preparing a calcium carbonate having a surface charge, characterized in that it comprises the following steps of: preparing a solution mixture of one or more monomer selected from the group consisting of calcium hydroxide, acrylic acid, acrylamide, or [2-(acryloyloxy)ethyl]-trimethylammonium chloride; removing dissolved gases such as oxygen and carbon dioxide in the said solution by injecting nitrogen (N2) gas; adjusting pH of the said solution; and inducing a reaction by adding polymerization initiator of a high molecule while injecting carbon dioxide to the said mixture.

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: 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: This invention relates to the integration of ammonium carbonate leach processes with established acid and alkaline uranium leach processes as multifunctional industrial processes for the extraction, high degree purification and conversion of processed or semi-processed uranium as U3O8, UO2, or most tetra or hexa-valent forms of uranium, and where applicable, for the recovery of uranium from uranium ores, using advanced multiple stage membrane based technologies for the separation and concentration of uranium in solution from heavy metals and lighter elements that may be present in the solution, and the selective leach and precipitation properties of an ammonium carbonate leach.

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: 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: The present invention is directed to a new more environmentally friendly method for the recovery of uranium from pregnant liquor solutions that comprise high concentration of sulfate by using an amino phosphonic functionalized resin.

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: Disclosed herein is a uranium ion exchange adsorption method using ultrasound. The method includes placing a slurry obtained by mixing uranium ions, sulfuric acid and an ion exchange resin into a reaction bath, and stirring the slurry in the reaction bath while simultaneously applying ultrasound to the reaction bath to allow the uranium ions to be adsorbed to the ion exchange resin through ion exchange adsorption. The method has an improved ion exchange adsorption rate of the uranium ions.

Abstract: This disclosure relates to articles that comprise polymeric winged fibers. The winged fibers have a high surface area because of their structure, which includes a core surrounded by a plurality of lobes. Channels of one micron or less in width are formed between adjacent lobes to form paths for the capture and/or transport of gases, liquids or particles. The winged fibers are assembled in woven or non-woven fabrics for use in wipes, absorbent pads, composite structures, apparel, outdoor wear, bedding, filtration systems, purification/separation systems, thermal and acoustic insulation, cell scaffolding, and battery membranes.

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: This invention relates to the integration of ammonium carbonate leach processes with established acid and alkaline uranium leach processes as multifunctional industrial processes for the extraction, high degree purification and conversion of processed or semi-processed uranium as U3O8, UO2, or most tetra or hexa-valent forms of uranium, and where applicable, for the recovery of uranium from uranium ores, using advanced multiple stage membrane based technologies for the separation and concentration of uranium in solution from heavy metals and lighter elements that may be present in the solution, and the selective leach and precipitation properties of an ammonium carbonate leach.

Abstract: Interlayer insulating films 5,7 (insulating films) provided in a memory capacitor cell 8 are formed between a gate electrode 3 and a counter electrode 8C formed on a silicon wafer 1. The interlayer insulating films 5,7 comprise a borazine-based resin, having a specific dielectric constant of no greater than 2.6, a Young's modulus of 5 GPa or greater and a leak current of no greater than 1×10?8 A/cm2.

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 oxine ligands 5-chloro-8-hydroxyquiniline and 5-sulfoxyl-8-hydroxyquinoline are covalently bound, using, for example, the Mannich reaction, to a silica gel polyamine composite made from a silanized amorphous silica xerogel and polyallylamine. The resulting modified composites, termed CB-1 (X?Cl) and SB-1 (X?SO3H), respectively, show a clear selectivity for trivalent over divalent ions and selectivity for gallium over aluminum. The compounds of the invention can be applied for the sequestration of metals, such as heavy metals, from contaminated mine tailing leachates.

Abstract: The invention covers the combination of utilizing the selectivity of an adsorbent to remove species from a liquid containing mixtures of ions and then subjecting the loaded resin to a chromatographic displacement utilizing the most selectively adsorbed species to displace the undesired co-adsorbing impurities. The technique can be used even when the most selectively adsorbed species is present as a minor constituent in the feed solution.

Abstract: Disclosed herein is a uranium ion exchange adsorption method using ultrasound. The method includes placing a slurry obtained by mixing uranium ions, sulfuric acid and an ion exchange resin into a reaction bath, and stirring the slurry in the reaction bath while simultaneously applying ultrasound to the reaction bath to allow the uranium ions to be adsorbed to the ion exchange resin through ion exchange adsorption. The method has an improved ion exchange adsorption rate of the uranium ions.

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: A process for selectively removing contaminant ions from a solution includes: a) contacting the solution with magnetic particles coupled to selectively chelating ion exchange functionality containing moieties prepared by: i) activating carboxyl groups on the selectively chelating ion exchange functionality containing moieties by the formation of an acyl fluoride, and ii) reacting the acyl fluoride with the magnetic particles, the magnetic particles having a particle size of less than 10 microns; b) allowing the chelating functionality coupled magnetic particles to selectively bind one or more of the contaminant ions; and, c) extracting the chelating functionality coupled magnetic particles and contaminant ions from the solution by magnetic filtration.

Abstract: A desulfurizer for fuel gas for a fuel cell includes: a first adsorption tank including an adsorber having selective adsorption capacity for a thiophene-based compound and a second adsorption tank including an adsorber having selective adsorption capacity for a mercaptan-based compound. The desulfurizer uses separate adsorbers having selective adsorption capacity for a thiophene-based compound and a mercaptan-based compound, in multiple stages to perform a more efficient and economical desulfurizing of a fuel gas to remove various sulfur compounds, especially thiophene-based compounds and mercaptan-based compounds compared to a desulfurizer using a single adsorber.

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: The subject of the present disclosure is a multi-column sequenced separation process and a drive for implementing this process. The disclosure applies particularly to the separation of metal derivatives such as uranium, nickel, copper, cobalt and other precious metals present in leaching effluents in hydrometallurgical processes.

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: Toxic substances such as heavy metals are extracted from a medium using a sorbent composition. The sorbent composition is derived by sulfidation of red mud, which contains hydrated ferric oxides derived from the Bayer processing of bauxite ores. Exemplary sulfidizing compounds are H2S, Na2S, K2S, (NH4)2S, and CaSx. The sulfur content typically is from about 0.2 to about 10% above the residual sulfur in the red mud. Sulfidized red mud is an improved sorbent compared to red mud for most of the heavy metals tested (Hg, Cr, Pb, Cu, Zn, Cd, Se, Th, and U). Unlike red mud, sulfidized red mud does not leach naturally contained metals. Sulfidized red mud also prevents leaching of metals when mixed with red mud. Mixtures of sulfidized red mud and red mud are more effective for sorbing other ions, such as As, Co, Mn, and Sr, than sulfidized red mud alone.

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 relates to a process for the recovery of uranium in high silica environments comprising the use of a strong base macroreticular ion exchange resin.

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: 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: 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: 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 method for separation of actinide elements comprising feeding a solution containing actinide elements such as americium, curium, californium and the like, into a resin column in which a weakly basic primary, secondary or tertiary anion exchange resin obtained by resinifying pyridine, imidazole or alkylamine has been packed, and then feeding an eluent of a mixed solution of nitric acid and alkyl alcohol such as methanol, ethanol, propanol and the like into the resin column to chromatographically separate the actinide elements from each other. This method makes it possible to efficiently separate the actinide elements from each other by a unit operation at ordinary temperature and ordinary pressure while avoiding oxidation operation, and hence makes it possible to avoid generation of secondary wastes and operations difficult in terms of engineering, such as precipitation.

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: 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: 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 multicolumn selectivity inversion generator separation method has been developed in which actinium ions, a desired daughter radionuclide, are selectively extracted from a solution of the thorium parent and daughter radionuclides by a primary separation column, stripped, and passed through a second guard column that retains any parent or other daughter interferents, while the desired daughter actinium ions and radium ions elute. This separation method minimizes the effects of radiation damage to the separation material and permits the reliable production of radionuclides of high chemical and radionuclidic purity for use in diagnostic or therapeutic nuclear medicine.

Abstract: A multicolumn selectivity inversion generator separation method has been developed in which a desired daughter radionuclide is selectively extracted from a solution of the parent and daughter radionuclides by a primary separation column, stripped, and passed through a second guard column that retains any parent or other daughter impurities, while the desired daughter elutes. This separation method minimizes the effects of radiation damage to the separation material and permits the reliable production of radionuclides of high chemical and radionuclidic purity for use in diagnostic or therapeutic nuclear medicine.

Abstract: A decontamination system uses magnetic molecules having ferritin cores to selectively remove target contaminant ions from a solution. The magnetic molecules are based upon a ferritin protein structure and have a very small magnetic ferritin core and a selective ion exchange function attached to its surface. Various types of ion exchange functions can be attached to the magnetic molecules, each of which is designed to remove a specific contaminant such as radioactive ions. The ion exchange functions allow the magnetic molecules to selectively absorb the contaminant ions from a solution while being inert to other non-target ions. The magnetic properties of the magnetic molecule allow the magnetic molecules and the absorbed contaminant ions to be removed from solution by magnetic filtration.

Abstract: An object of the present invention is to provide a structure of a metal collecting material for use in seawater or river water or industrial wastewater to give the most excellent adsorption efficiency. Another object of the present invention is to provide a process for efficiently separating and recovering valuable metals or noxious metals from a metal collecting material having collected said valuable metals from seawater or noxious metals from river water or wastewater with the least possible waste generation. A metal collector according to an aspect of the present invention is formed by stacking at least partially alternate layers of a fibrous metal collecting material with a spacer for introducing a liquid to be treated into said collecting material, wherein the area of spacer side faces represents 25-75% of the total area of collecting material side faces and spacer side faces in the side faces of the stack.

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.