Abstract: A process for reducing a level of radionuclei in a phosphogypsum material having a radioactive contaminant is provided, wherein the process includes the steps of: (a) mixing the phosphogypsum material with an acid decontamination solution to form a mixture; (b) separating the mixture into a liquid stream having a large portion of the radioactive contamination and a solid stream having a residual component of the acid decontamination solution; and (c) removing the residual component of the acid decontamination solution from the solid stream, thus providing a gypsum material that meets or exceeds federal standards for level of radioactive nuclei and providing a process that is environmentally friendly and capable of avoiding the generation of phosphogypsum stacks, as well as eliminating existing or abandoned phosphogypsum stacks, further providing for recovery of the radioactive nuclei if desired.

Abstract: A process for purifying a stock Sr-90 solution containing stable and radioactive impurities, holding the purified Sr-90 solution for Y-90 ingrowth, and subsequently extracting the Y-90 from the Sr-90/Y-90 solution. The stock solution is sequentially passed through two thermoxide-type sorbents (T-3 and T-5), which hold the impurities while passing the Sr-90 solution. After ingrowth of Y-90, the Sr-90/Y-90 solution is passed through sorbent T-3, which preferentially sorbs the Y-90 while passing the Sr-90 solution. The Y-90 is then eluted from the T-3 sorbent. The T-3 and T-5 sorbents are specially prepared compounds of zirconium dioxide and titanium dioxide, respectively, that preferentially sorb Y-90 under predetermined conditions of solution pH and NaCl concentration.

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: Inorganic ion exchange materials for the separation of 90Y from 90Sr include clinoptilolite, potassium titanosilicate pharmacosiderite, sodium titanosilicate and sodium nonatitanate. These materials are suitable for making a 90Y generator that contains 90Sr immobilized on an ion exchange column of the materials. The materials have a very high selectivity for 90Sr, a very low selectivity for 90Y, good radiation and thermal stability, low toxicity, fast reaction kinetics, and can be readily and reproducibly synthesized. A method is thus provided for eluting 90Y from the ion exchange material with an aqueous solution to obtain a carrier-free solution of 90Y.

Abstract: A method and apparatus for isolating and purifying a 213Bi radioactive isotope from an 225Ac source using a primary column and a primary sorbent which preferentially retains 225Ac over 213Bi when exposed to a compatible solvent in combination with a secondary column having a secondary sorbent which retains 213Bi when exposed to a mixture of the compatible solvent and 213Bi. A “compatible solvent” is a solvent which will preferentially remove 213Bi radioactive isotopes from a primary sorbent without removing 225Ac radioactive isotopes, and then allow 213Bi radioactive isotopes removed from the primary sorbent to be retained on a secondary sorbent, without having to dilute or otherwise chemically or physically modify the compatible solvent in between exposure to the primary and secondary sorbents.

Abstract: The invention includes a method of separating isotopes from a mixture containing at least two isotopes in a solution. 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 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 precipitate. The 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. The invention also includes a system for producing an actinium-225/bismuth-213 product.

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: A multicolumn selectivity inversion generator has been developed in which bismuth-213 is selectively extracted from an HCl solution of the actinium-225 parent (and its radiogenic descendents) by a primary separation column containing a separation medium containing a neutral oxygenated organophosphorus extractant. After rinsing with dilute HCl, the bismuth-213 is stripped with a buffered NaCl solution. The stripped solution is passed through a cation-exchange resin guard column that retains the actinium-225 and radium-225 contaminants while the bismuth-213 elutes. This generator method minimizes the unpredictable effects of radiation damage to the support material and permits the reliable production of bismuth-213 of high chemical and radionuclidic purity.

Abstract: A gas-free system for separating a solution of substantially impurity-free daughter products from an associated parent load solution includes a pump, a plurality of multi-port valves, separation medium and a processor. An uncoiled conduit extends between a third port of a second multi-port valve and a first multi-port valve. A processor is operably coupled to a pump, and the plurality of multi-port valves. A method for separating a solution of substantially impurity-free daughter product from an associated parent load solution is also disclosed.

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.

Abstract: Process for the purification and concentration of isotopes which isotopes are dissolved in a dilute acidic solution and adsorbed on an (optionally) activated surface of a d10-metal whereby the isotopes are selectively desorbed by elution whereby in situ reduction of oxidation products is achieved and an apparatus for performing said process.

Abstract: A method and apparatus are provided for automatically separating radionuclides using a chromatographic separation process. The method includes the steps of displaying a first flow diagram on a display depicting flow of the radionuclides through a first set of separation elements of the plurality of separation processing elements, but only during a first step of the chromatographic separation process and displaying a second flow diagram on the display depicting flow of the radionuclides through a second set of separation elements of the plurality of separation processing elements, but only during a second step of the chromatographic separation process.

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 method for removing a wide variety of radioactive contaminants from a contaminated surface sufficient for the surface to be classified as a low-level waste or as free release. Contaminated surfaces may be classified as Class C, Class B, Class A, or high-level radioactive wastes prior to treatment. An aqueous solution having a wetting agent and a complex substituted keto-amine is provided. The aqueous solution is left on the surface for a time sufficient to remove the radioactive contaminants into the aqueous solution which is then removed. Depending on the type and condition of the surface, an acid may be added to the aqueous solution to aid in the contaminant removal process. However, typical metals surfaces may often be treated effectively without the use of concentrated acids or with dilute concentrations of such acids. The method of the present invention has the effect of removing substantially all of the radioactive contaminants from a variety of previously contaminated surfaces.

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: Novel stable forms of technetium and methods for their production are provided. These methods have particular application for the removal of technetium from industrial waste compositions.

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.

Abstract: A method for removing radioactive contaminants from a given surface. An aqueous solution having a wetting agent and a complex substituted keto-amine is provided. The solution is left on the surface for a time sufficient to remove the radioactive contaminants into the aqueous solution which is then removed. Depending on the type and condition of the surface, a concentrated acid may be added to the aqueous solution to aid in the contaminant removal process. In such a case, a pH of less than 3.0, and preferably less than 1.5 is maintained. If a concentrated acid is used, the acidic solution containing radioactive contaminants is preferably neutralized by an alkaline material to a pH of between 5.5 and 9.0. Removal of thorium contamination from railcars is one useful application of the invention. The method of the present invention has the effect of removing substantially all of the radioactive contaminants from a previously contaminated surface.

Abstract: The present invention provides a generator capable of producing therapeutic Bismuth-213 doses. Also disclosed are methods of preparing Bismuth-213-labeled compounds using such generator and applications of the labeled compounds.

Abstract: An extractant medium for extracting alkaline earth cations from an aqueous acidic sample solution is described as are a method and apparatus for using the same. The separation medium is free of diluent, free-flowing and particulate, and comprises a Crown ether that is a 4,4′(5′)[C4-C8-alkylcyclohexano]18-Crown-6 dispersed on an inert substrate material.

Abstract: A method for removing radioactive contaminants from a given surface. An acidic solution having a pH of less than 3.0, and preferably less than 1.5, of a complex substituted keto-amine, and a mixture of a saturated and unsaturated lower alcohol (e.g. isopropyl alcohol and propargyl alcohol), and concentrated acid is applied to a contaminated surface. The solution is left on said surface for a time sufficient to remove the radioactive contaminants into the acidic solution and then removed. The acidic solution containing the radioactive contaminants is preferably neuralized by an alkaline material to a pH of between 5.5 and 9.0. Removal of thorium contamination from railcars is one useful application of the invention. The method of the present invention has the effect of removing substantially all of the radioactive contaminants from a previously contaminated surface.

Abstract: This invention relates to a method and an apparatus for preparing Bi-213 to be integrated in a radioimmunocojugate for human therapeutic use. According to the invention the method comprises the sequence of steps as follows: a) an ampoule 5 containing colloid-free actinium-225, obtained from drying an actinium nitrate solution, is loaded into a container 20 provided with radiation panels 21; b) a dissolving medium is poured into the ampoule 5; c) the solution obtained in the ampoule is transferred into an ion exchange column 6; d) an elution medium is circulated through the column 6; e) the eluate containing eluted Bi-213 is pumped towards a vial 10 for quantification and quality control.

Abstract: An electrochemical ion exchange cell comprises a metal carbollide as ion exchange material. The carbollide is preferably a polynuclear cobalt dicarbollide and is typically substituted by chlorine. Also novel are polynuclear metal carbollides comprising a substituted carbollide cage, as well as metal carbollides comprising a carbollide cage substituted by a moiety having a —COOH or —SH group.

Abstract: A high radiation resistant inorganic sorbent is used for the extraction of Mo-99 from the fission byproducts of a uranyl sulfate nuclear-fueled homogeneous-solution nuclear reactor. The sorbent is a composition of hydrated titanium dioxide combined with 5 to 40 mole percent zirconium hydroxide formed into pellets and saturated with an isotope composition of uranium.

Abstract: The invention relates to a process for separating and purifying multi-curie quantities 90Y of sufficient chemical and radiochemical purity suitable for use in medical applications without a series of 90Sr selective extraction chromatographic columns while minimizing loss of radioactive 90Sr parent and waste stream.

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

Abstract: The present invention is an improvement to the method of separating and purifying gadolinium from a mixture of gadolinium and europium having the steps of (a) dissolving the mixture in an acid; (b) reducing europium+3 to europium+2; and (c) precipitating the europium+2 with a sulfate ion in a superstoichiometric amount; wherein the improvement is achieved by using one or more of the following: (i) the acid is an anoic acid; (ii) the reducing is with zinc metal in the absence of a second metal or with an amount of the second metal that is ineffective in the reducing; (iii) adding a group IIA element after step (c) for precipitating the excess sulfate prior to repeating step (c); (iv) the sulfate is a sulfate salt with a monovalent cation; (v) adding cold europium+3 prior to repeating step (c).

Abstract: Hafnium is recovered from irradiated tantalum by: (a) contacting the irradiated tantalum with at least one acid to obtain a solution of dissolved tantalum; (b) combining an aqueous solution of a calcium compound with the solution of dissolved tantalum to obtain a third combined solution; (c) precipitating hafnium, lanthanide, and insoluble calcium complexes from the third combined solution to obtain a first precipitate; (d) contacting the first precipitate of hafnium, lanthanide and calcium complexes with at least one fluoride ion complexing agent to form a fourth solution; (e) selectively adsorbing lanthanides and calcium from the fourth solution by cationic exchange; (f) separating fluoride ion complexing agent product from hafnium in the fourth solution by adding an aqueous solution of ferric chloride to obtain a second precipitate containing the hafnium and iron; (g) dissolving the second precipitate containing the hafnium and iron in acid to obtain an acid solution of hafnium and iron; (h) selectively ad

Abstract: A system and method for eluting one or more radioisotope processing elements, such as generators and/or concentration and purification components, with a gas over eluent delivery mechanism is disclosed. The system is adapted to an elution/concentration system using two such mechanisms, one for an initial elution of one or more generators, and a second for "re-elution" of concentrated and purified radioisotope contained in a concentration subsystem. The operation of such a dual mechanism system may also be completely or partially automated. The invention also discloses a concentration subsystem containing all single-use components needed for concentration and/or purification, as a single use, self sealed, cartridge that can safely store waste eluate for safe and simple disposal.

Abstract: The present invention is a method of separating a short-lived daughter isotope from a longer lived parent isotope, with recovery of the parent isotope for further use. Using a system with a bi-directional pump and one or more valves, a solution of the parent isotope is processed to generate two separate solutions, one of which contains the daughter isotope, from which the parent has been removed with a high decontamination factor, and the other solution contains the recovered parent isotope. The process can be repeated on this solution of the parent isotope. The system with the fluid drive and one or more valves is controlled by a program on a microprocessor executing a series of steps to accomplish the operation. In one approach, the cow solution is passed through a separation medium that selectively retains the desired daughter isotope, while the parent isotope and the matrix pass through the medium. After washing this medium, the daughter is released from the separation medium using another solution.

Abstract: Isotopes of either germanium or silicon are separated by a chemical exchange reaction. Preferably the compound is a hydro halide or halide composition and the donor can be a wide variety of compounds such as a lower molecular weight alcohol.

Abstract: A process for producing substantially radio-impurity-free Bi-212 is disclosed. An acidic feed solution containing Pb-212 is contacted with an extraction medium to bind the Pb-212 thereto. The extraction medium is rinsed with a second acid solution to remove impurities therefrom, leaving a substantially impurity-free Pb-212-laden extraction medium The Pb-212 grows on the extraction medium to form Bi-212 by radioactive decay. The Bi-212 is then eluted from the extraction medium with an acid solution to form a substantially radio-impurity-free Bi-212 acid solution. An apparatus for carving out this process and a process for the therapeutic use thereof are also disclosed.

Abstract: The present invention relates to a method for separating metals by micellar ultrafiltration, that can be used to process radioactive waste.This process consists of adding to the aqueous solution a complexing anionic surfactant having the formula: ##STR1## M.sup.+ =cation, for example NA.sup.+m=1 or 2R=aliphatic group of at least eight carbon atoms, and optionally a nonionic surfactant such as Triton X-100 or Brij 35 to produce micelles (10) in the solution, and subsequently of filtering the solution through an ultrafiltration membrane (3) to retain the previously formed micelles.

Abstract: A method of producing and purifying Cs-131 comprising the steps of dissolving irradiated Ba comprised of natural or enriched Ba including Ba-130, Ba-131, and Cs-131 from the decay of Ba-131, in an acid, precipitating the Ba, separating the Cs-131 using an ion exchange media, and eluting the Cs-131 from the exchanger to recover the purified Cs-131.

Abstract: A method and apparatus which enables a complete multi-stepped chemical treatment process to occur within a single, sealed-bladdered vessel 31. The entire chemical process occurs without interruption of the sealed-bladdered vessel 31 such as opening the sealed-bladdered vessel 31 between various steps of the process. The sealed-bladdered vessel 31 is loaded with a batch to be dissolved, treated, decanted, rinsed and/or dried. A pressure filtration step may also occur. The self-contained chemical processing apparatus 32 contains a sealed-bladder 32, a fluid pump 34, a reservoir 20, a compressed gas inlet, a vacuum pump 24, and a cold trap 23 as well as the associated piping 33, numerous valves 21,22,25,26,29,30,35,36 and other controls associated with such an apparatus. The claimed invention allows for dissolution and/or chemical treatment without the operator of the self-contained chemical processing apparatus 38 coming into contact with any of the process materials.

Abstract: Trivalent actinides can be separated from trivalent lanthanides in aqueous solutions at H.sup.+ concentrations of 2 mol/l to 0.001 mol/liter by extraction with bis(aryl)dithiophosphinic acid and a synergist such as TBP, TOPO and TBPO in an organic solvent. A high separation efficiency results. The method is applicable to high-level liquid waste from reactors and nuclear-material processing plants.

Abstract: Strontium-82 is recovered from spent strontium-82/rubidium-82 generators to provide a source of strontium-82 for additional strontium-82/rubidium-82 generators. The process involves stripping of the strontium-82 from used strontium-82/rubidium-82 generators followed by purification of the strontium-82 material to remove additional metal contaminants to desired levels.

Abstract: A process for treating oils and solvents contaminated by radioactive substances includes a step of subjecting the oils and solvents to the action of pre-selected micro-organisms in the presence of air and a very large volume of water, relative to the volume of oils and solvents to be treated. The micro-organisms destroy organic molecules, and in particular transform them into CO.sub.2 and H.sub.2 O. Further steps of the process include preparing a predetermined volume of water having predetermined characteristics of dissolved oxygen concentration, pH and redox potential and adding a predetermined charge of oils and solvents contaminated by radioactive substances to this volume of water, the charge corresponding to a volume of oils and solvents which is a predetermined fraction of the predetermined volume of water. The charge is subjected to the action of micro-organisms at a predetermined temperature and for a predetermined time period.

Abstract: A process for the production of yttrium-90, generated by radioactive decay of strontium-90, by means of a generator including a column containing strontium-90 adsorbed on a cation-exchange resin, wherein yttrium-90, the decay product of strontium-90, is obtained by the flow of an eluent through the resin column of the generator, an acetate buffer solution being used as the eluent. The process enables a solution to be obtained which is usable directly for labelling protein substrates, particularly monoclonal antibodies, with yttrium-90.

Abstract: A crown ether cesium ion extractant is disclosed as is its synthesis. The crown ether cesium ion extractant is useful for the selective purification of cesium ions from aqueous acidic media, and more particularly useful for the isolation of radioactive cesium-137 from nuclear waste streams. Processes for isolating cesium ions from aqueous acidic media using the crown ether cesium extractant are disclosed as are processes for recycling the crown ether cesium extractant and processes for recovering cesium from a crown ether cesium extractant solution.

Abstract: A process for producing substantially impurity-free Bi-213 cations is disclosed. An aqueous acid feed solution containing Ac-225 cations is contacted with an ion exchange medium to bind the Ac-225 cations and form an Ac-225-laden ion exchange medium. The bound Ac-225 incubates on the ion exchange medium to form Bi-213 cations by radioactive decay. The Bi-213 cations are then recovered from the Ac-225-laden ion exchange medium to form a substantially impurity-free aqueous Bi-213 cation acid solution. An apparatus for carrying out this process is also disclosed.

Abstract: An improved method for producing .sup.99m Tc compositions. .sup.100 Mo metal is irradiated with photons in a particle (electron) accelerator to produce .sup.99 Mo metal which is dissolved in a solvent. A solvated .sup.99 Mo product is then dried to generate a supply of .sup.99 MoO.sub.3 crystals. The crystals are thereafter heated at a temperature which will sublimate the crystals and form a gaseous mixture containing vaporized .sup.99m TcO.sub.3 and vaporized .sup.99m TcO.sub.2 but will not cause the production of vaporized .sup.99 MoO.sub.3. The mixture is then combined with an oxidizing gas to generate a gaseous stream containing vaporized .sup.99m Tc.sub.2 O.sub.7. Next, the gaseous stream is cooled to a temperature sufficient to convert the vaporized .sup.99m Tc.sub.2 O.sub.7 into a condensed .sup.99m Tc-containing product. The product has high purity levels resulting from the use of reduced temperature conditions and ultrafine crystalline .sup.99 MoO.sub.3 starting materials with segregated .sup.

Abstract: An improved method for producing .sup.99m Tc compositions from .sup.99 Mo compounds. .sup.100 Mo metal or .sup.100 MoO.sub.3 is irradiated with photons in a particle (electron) accelerator to ultimately produce .sup.99 MoO.sub.3. This composition is then heated in a reaction chamber to form a pool of molten .sup.99 MoO.sub.3 with an optimum depth of 0.5-5 mm. A gaseous mixture thereafter evolves from the molten .sup.99 MoO.sub.3 which contains vaporized .sup.99 MoO.sub.3, vaporized .sup.99m TcO.sub.3, and vaporized .sup.99m TcO.sub.2. This mixture is then combined with an oxidizing gas (O.sub.2(g)) to generate a gaseous stream containing vaporized .sup.99m Tc.sub.2 O.sub.7 and vaporized .sup.99 MoO.sub.3. Next, the gaseous stream is cooled in a primary condensation stage in the reaction chamber to remove vaporized .sup.99 MoO.sub.3. Cooling is undertaken at a specially-controlled rate to achieve maximum separation efficiency.

Abstract: A .sup.99 Mo/.sup.99m Tc generator system includes a sorbent column loaded with a composition containing .sup.99 Mo. The sorbent column has an effluent end in fluid communication with an anion-exchange column for concentrating .sup.99m Tc eluted from the sorbent column.A method of preparing a concentrated solution of .sup.99m Tc includes the general steps of:a. providing a sorbent column loaded with a composition containing .sup.99 Mo, the sorbent column having an effluent end in fluid communication with an anion-exchange column;b. eluting the sorbent column with a salt solution to elute .sup.99m Tc from the sorbent and to trap and concentrate the eluted .sup.99m Tc on the ion-exchange column; andc. eluting the concentrated .sup.99m Tc from the ion-exchange column with a solution comprising a reductive complexing agent.

Abstract: The invention provides a method for recovering metal beryllium from waste beryllium that has become radioactive in the reaction with neutrons, by converting the radioactive beryllium to halogenated beryllium with a halogen gas such as chlorine, bromine or iodine, and subsequently submitting the halogenated beryllium to thermal decomposition. In this way, the invention allows effective recovery of metal beryllium from waste, radioactive beryllium, thereby reducing production of radioactive wastes and ensuring an economical use of natural resources.

Abstract: A method for separating .sup.213 Bi from a solution of radionuclides wherein the solution contains a concentration of the chloride ions and hydrogen ions adjusted to allow the formation of a chloride complex. The solution is then brought into contact with an anion exchange resin, whereupon .sup.213 Bi is absorbed from the solution and adhered onto the anion exchange resin in the chloride complex. Other non-absorbing radionuclides such as .sup.225 Ra, .sup.225 Ac, and .sup.221 Fr, along with HCl are removed from the anion exchange resin with a scrub solution. The .sup.213 Bi is removed from the anion exchange resin by washing the anion exchange resin with a stripping solution free of chloride ions and with a reduced hydrogen ion concentration which breaks the chloride anionic complex, releasing the .sup.213 Bi as a cation. In a preferred embodiment of the present invention, the anion exchange resin is provided as a thin membrane, allowing for extremely rapid adherence and stripping of the .sup.213 Bi.

Abstract: A method of preparing a concentrated solution of a carrier-free radioisotope which includes the steps of:a. providing a generator column loaded with a composition containing a parent radioisotope;b. eluting the generator column with an eluent solution which includes a salt of a weak acid to elute a target daughter radioisotope from the generator column in a first eluate.c. eluting a cation-exchange column with the first eluate to exchange cations of the salt for hydrogen ions and to elute the target daughter radioisotope and a weak acid in a second eluate;d. eluting an anion-exchange column with the second eluate to trap and concentrate the target daughter radioisotope and to elute the weak acid solution therefrom; ande. eluting the concentrated target daughter radioisotope from the anion-exchange column with a saline solution.

Abstract: Methods and agents for extracting cesium and strontium ions from aqueous solutions, including aqueous fission product waste solutions, are disclosed using substituted metal dicarbollide and/or dicarbaextranide ions containing one or more chemical groups that increase solubility of the substituted metal dicarbollide or dicarbaextranide ions in non-nitrated, non-chlorinated solvents, or using metal dicarbollide or dicarbaextranide ion-substituted silicones.

Abstract: Preparation and use of polymeric materials containing hydrophobic anions and plasticizers for extraction of cesium and strontium. The use of polymeric materials containing plasticizers which are solvents for hydrophobic anions such as derivatives of cobalt dicarbollide or tetraphenylborate which are capable of extracting cesium and strontium ions from aqueous solutions in contact with the polymeric materials, is described. The polymeric material may also include a synergistic agent for a given ion like polyethylene glycol or a crown ether, for removal of radioactive isotopes of cesium and strontium from solutions of diverse composition and, in particular, for solutions containing large excess of sodium nitrate.

Abstract: Methods and agents for extracting cesium and strontium ions from aqueous solutions, including aqueous fission product waste solutions, are disclosed using substituted metal dicarbollide ions containing one or more chemical groups that increase solubility of the substituted metal dicarbollide ion in non-nitrated, non-chlorinated solvents, or using metal dicarbollide ion-substituted silicones.