Abstract: A generator system for providing a carrier-free radioisotope in the form of an acid comprises a chromatography column in tandem fluid connection with an ion exchange column, the chromatography column containing a charge of a radioactive parent isotope. The chromatography column, charged with a parent isotope, is eluted with an alkali metal salt solution to generate the radioisotope in the form of an intermediate solution, which is passed through the ion-exchange column to convert the radioisotope to a carrier-free acid form.

Abstract: A method of treatment of a high-level radioactive waste comprising heating the radioactive waste at a high temperature of about 500.degree. to 3000.degree. C. to vaporize part of the elements contained in the radioactive waste, and cooling the resultant vapor to separately collect the elements. In one embodiment, the heating step is replaced by a reduction-heating step wherein heating is carried out in the presence of a reducing agent, e.g. hydrogen. In another embodiment, the heating step may be followed by the reduction-heating step.

Abstract: A process and an apparatus are provided which separate the lower molecular weight isotope fraction from the higher molecular weight isotope fraction of an element which forms a water soluble acid at temperatures amenable to ion exchange. The separation is accomplished by shifting the equilibrium constants for absorption and desorption by temperature variations of the acidic solution. The process and apparatus are preferably employed to enrich a boric acid solution in its boron-10 isotope fraction by employing the use of ion exchange resins which are capable of thermally storing and releasing boron isotopes.

Abstract: The thermal neutron capture cross-section of zirconium may be altered by altering its natural isotope distribution through a steady state chromatographic separation of these isotopes using an anion exchange resin as the stationary phase of the chromatographic column. Zirconium is dissolved in a very strong acid which causes the formation of a zirconium anion, such as the ZrOCl.sub.4.sup.-2 anion formed in six normal hydrochloric acid, and eluted off the column with a weaker acid. Distinct elution volumes representative of each isotope are collected. In a preferred embodiment, the process also separates the zirconium from hafnium and the other impurities normally present in the product obtained by chlorinating zircon sand and utilizes a continuous annular chromatograph.

Abstract: A process for selective separation of strontium-82 and strontium-85 from proton irradiated molybdenum targets comprises dissolving the molybdenum target in a hydrogen peroxide solution to form a first solution containing ions selected from a group consisting of molybdenum, niobium, technetium, selenium, vanadium, arsenic, germanium, zirconium, rubidium, zinc, beryllium, cobalt, iron, manganese, chromium, strontium, and yttrium; passing the solution through a first cationic resin whereby ions selected from a group consisting of zinc, beryllium, cobalt, iron, manganese, chromium, strontium, yttrium a portion of zirconium and a portion of rubidium are selectively absorbed by the first resin; contacting the first resin with an acid solution to strip and remove the absorbed ions from the first cationic exchange resin to form a second solution; evaporating the second solution for a time sufficient to remove substantially all of the acid and water from the solution whereby a residue remains; dissolving the residue i

Abstract: Novel cesium hydroxytriphenylborates are described. The addition of hydroxytriarylborate ions to aqueous media containing cesium ions to form such compounds as a precipitate is useful, for example, in the removal of radioactive cesium from nuclear fission plant waste streams and from cesium ion containing solutions resulting from the digestion of cesium ores.

Abstract: A method and apparatus for extracting strontium and technetium values from biological, industrial and environmental sample solutions using a chromatographic column is described. An extractant medium for the column is prepared by generating a solution of a diluent containing a Crown ether and dispersing the solution on a resin substrate material. The sample solution is highly acidic and is introduced directed to the chromatographic column and strontium or technetium is eluted using deionized water.

Abstract: The isotopes of zirconium can be partially or completely separated by loading an aqueous solution of an ionic compound of zirconium onto a cationic exchange resin with pentavalent phosphorus derived active groups which serves as the stationary phase of a chromatograph, eluting the compound with an aqueous acid and collecting distinct elution volumes representative of each isotope. In a preferred embodiment, the eluant is a strong mineral acid, such as hydrochloric acid and the chromatograph is of a type, such as a continuous annular chromatograph, that it can be operated in a continuous, steady state manner.

Abstract: The isotopes of zirconium can be partially or completely separated by loading an aqueous solution of an ionic compound of zirconium onto a cationic exchange resin which serves as the stationary phase of a chromatograph, eluting the compound with an aqueous acid and collecting distinct elution volumes representative of each isotope. In a preferred embodiment, the eluant is a strong mineral acid, such as hydrochloric acid and the chromatograph is of a type, such as a continuous annular chromatograph that it can be operated in a continuous, steady state manner.

Abstract: A method of treatment of a high-level radioactive waste containing platinum group elements is provided in which boron and a boron compound is added to a calcined material of the high-level radioactive waste in an amount of 0.5 to 10% by weight in terms of boron as a simple substance, and the resultant mixture is heated at a temperature of about 1000.degree. C. or above under a reduction condition to melt the mixture and to alloy the platinum group elements present in the calcined material with boron. A layer of the resultant platinum group element alloys is then separated and recovered from a layer of residual oxides through sedimentation. The layer of the residual oxides is solidified to form a highly volume-reduced high-level radioactive solidified waste.

Abstract: Water soluble irradiation targets are disclosed for the production of .sup.186 Re and .sup.188 Re. The irradiation targets are selected for both water solubility and absence of elements which would produce contaminating isotopes for medical therapeutic and diagnostic use. In one embodiment, .sup.186 Re or .sup.188 Re is produced by the direct irradiation of a water soluble irradiation target comprising .sup.185 Re or .sup.187 Re, respectively. Preferred targets for this purpose include aluminum perrhenate, lithium perrhenate and magnesium perrhenate. In another embodiment, a zirconyl tungstate generator comprising .sup.188 W for the production of .sup.186 Re is obtained by irradiating a soluble irradiation target comprising .sup.186 W, dissolving the irradiated target in aqueous solution, reacting the dissolved target with an aqueous solution comprising zirconyl ion to form an insoluble zirconium tungstate precipitate and disposing the precipitate in an elutable container.

Abstract: The improved method and apparatus are capable of efficient removal of radioactive europium from solutions of radioactive gadolinium in a simple way. A mixture of a zinc and a graphite powder is packed into a column and both a conditioning solution corresponding to a liquid electrolyte and a sample solution containing radioactive gadolinium and europium are allowed to pass through the column.

Abstract: A method and generator for preparing the radioisotope of lead, .sup.212 Pb, whereby .sup.228 Th, in a closed chamber, is allowed to decay to gaseous .sup.220 Rn which is then readily separated from the thorium and other decay products by diffusing the .sup.220 Rn gas into a second chamber, where it decays to .sup.212 Pb which can then be collected from the second chamber. The .sup.228 Th preferably is amorphous, such as thorium stearate. Collection of .sup.212 Pb occurs in a medium of high and open porosity into which the .sup.220 Rn diffuses so that the decay products recoil into the medium. The .sup.212 Pb can be recovered from this medium by dissolving the medium or by reacting it with an antibody-chelating complex solution to entrap it in the antibody-chelating complex.

Abstract: The isotopes of zirconium can be partially or completely separated by loading an essentially anhydrous alcoholic solution of an ionic compound of zirconium onto a cationic exchange resin with pentavalent phosphorus derived active groups which serves as the stationary phase of a chromatograph, eluting the compound with an essentially anhydrous alcoholic solution of hydrogen chloride and collecting distinct elution volumes representative of each isotope. In a preferred embodiment, the active groups of the cation exchange resin are derived from di-2-ethylhexyl phosphoric acid, tri-n-octyl phosphine oxide or tributyl phosphate and the chromatograph is of a type, such as a continuous annular chromatograph that it can be operated in a continuous steady state manner.

Abstract: A gas cooled high temperature reactor with a core of preferably spherical fuel elements enclosed by a graphite roof and a side and bottom reflector. In order to minimize environmental contamination by fission products, even upon (hypothetical) core overheating accidents, the upper part of the roof reflector incorporates volatile iodine and cesium iodide, suitable metal atoms or molecules of the compounds of these metals bonded into the cracked binder material of the graphite. The dilution of the doping material must be high enough (from 1:100 to 1:10,000), so that the substances (for example alkaline earth metals or rare earths) are present as individual atoms or molecules. The iodine is chemically bound to the doping atoms, and iodide formed in this matter is retained in the cracked binder material. Molecules of high temperature oxides may also be incorporated.

Abstract: Contaminated surface layers are decontaminated by treatment with an aqueous fluorine base-containing decontamination solution. The aqueous decontamination solution contains 0.05 to 50 Mol of decontamination agent per liter, and the decontamination agent preferably comprises at least one substance from the group colon hexafluorosilicate acid, fluoroboric acid, and the salts of both of these. The decontamination solution produces the required high decontamination factors on metallic substances and brickworks as well. The used decontamination solution can, after regeneration, be recycled into the decontamination process.Release of decontaminated material by dissolution of the surface layer of the decontaminated objects provides decontamination of objects having complicated and hard-to-measure geometries.The decontamination agent (HBF.sub.4 -acid) is advantageously produced from contaminated boric acid from pressurized water reactor wastes by reaction with fluoride or hydrofluoric acid. The HBF.sub.

Abstract: Process for the fine purification of fission molybdenum, dissolved in ionic form together with ions of the fission products of I, Sn, Ce, Ru and Zr in aqueous mineral acid solution, in which the fission molybdenum is fixed by a metal oxide in a sorption step and is then released in a desorption step, comprising: (a) passing the aqueous solution over an amphoteric oxide to sorb the ions of Mo, I, Ce, Ru, Sn and Zr, (b) drying the resulting charged oxide and thereafter heating the dried oxide to a temperature of about 1200.degree. C. to about 1300.degree. C. to sublimate the Mo, and at the same time passing a carrier stream containing water vapor and oxygen over the charged oxide to take up the Mo, and (c) desublimating Mo out of the carrier gas stream by cooling to a temperature below 600.degree. C., and followed by further cooling to room temperature, dissolving the resulting Mo residue in an aqueous solution of a strong alkali to form a molybdate solution.

Abstract: A high-level radioactive waste liquid, produced in a reprocessing plant, is treated by a freeze-drying step. A low-level radioactive waste liquid containing water, nitric acid and nuclides is obtained from the sublimate of the freeze-drying step. Fission products, actinides, corrosion products (iron, chromium, nickel, etc.), sodium nitrate and sodium hydroxide, not sublimated, are separated as a residue. An alkaline solution such as sodium hydroxide solution is added to the residue to dissolve sodium nitrate and sodium hydroxide. The dissolved nitrate and hydroxide are then separated from the residue containing fission products and the corrosion products.

Abstract: An ion exchange material comprises an oxide of a first element selected from elements of Groups IVa, IVb, Va, Vb, VIa, VIb, VIIa, or lanthanide or actinide series of the Periodic Table, in combination with an oxide of at least one different element selected from elements of the afore-mentioned groups of the Periodic Table so as to form a composite material. For example, a composite ion exchange material may comprise silicon dioxide and manganese dioxide. By selecting the appropriate oxides a composite ion exchange material can be prepared which has greater mechanical and/or chemical stability than an individual oxide of the composite material.

Abstract: Ion exchange performed by contacting a solution containing ions to be absorbed with a hydrous oxide of uranium in a suitable form. The hydrous oxide may be uranium dioxide or uranium hydroxide, and may be prepared by reducing a uranium species, precipitating the reduced species, and drying the precipitate to give the required hydrous oxide.

Abstract: A sterile pyrogen free Re.sup.196 or Re.sup.188 radiopharmaceutical is prepared by reducing radioactive perrhenate in the presence of a bone seeking ligand to form a solution including complexed Re-ligand compound together with impurities such as uncomplexed perrhenate, ReO.sub.2 and free ligand. The radiopharmaceutical is obtained by a low pressure purification method. The solution is added to a separation column, preferably an ion exchange column, and a first portion is eluted and discarded. The radiopharmaceutical is collected in a second eluted portion defined as having 25-85% of the activity of the Re on the column, less than 10% unligated perrhenate and less than 1% ReO.sub.2. This can be used as is to image and treat metastatic bone leasions. A simple kit is also disclosed.

Abstract: Contaminated surface layers are decontaminated by treatment with an aqueous fluorine base-containing decontamination solution. The aqueous decontamination solution contains 0.05 to 50 Mol of decontamination agent per liter, and the decontamination agent preferably at least one substance from the group: hexafluorosilicate acid, fluoroboric acid, and the salts of both of these. The decontamination solution produces the required high decontamination factors on pressurized water reactors, boiling water reactors, metallic substances, high temperature alloys and brickworks as well. The used decontamination solution can, after regeneration, by recycled into the decontamination process. Release of decontaminated material by dissolution of the surface layer of the decontaminated objects provides decontamination of objects having complicated and hard-to-measure geometries. The decontamination agent (NBF.sub.

Abstract: A solvent extraction composition for radium including a high molecular wet organophilic carboxylic acid and an organophilic macrocycle dissolved in a suitable solvent.

Abstract: A formulation is provided for the treatment of synovial inflammation, the formulation consisting of dysprosium-165 hydroxide macroaggregates or yttrium-90 hydroxide macroaggregates, wherein the formulation is substantially devoid of other co-precipitating agents in particular, ferric hydroxide. A method of treatment of synovial inflammation is also provided in which the method comprises adminstering an intra-articular injection of the above formulation. Preferably the above formulation is in the form of a isotonic apyrogenic injectable solution.

Abstract: A system for extracting sodium, particularly radioactive .sup.22 Na, from aluminum utilizes a monel exhaust system for exhausting sodium vapor emitted from a molten aluminum target, in a draft of helium gas. The aluminum target is heated until it is melted in a graphite support cup. The graphite support does not react with the sodium. Moreover, the graphite is understood to be permeable to the helium gas, its porosity being believed to provide the significant advantage that the graphite holder will not absorb much of the sodium vapor. The use of graphite avoids a disadvantageous monel-aluminum reaction whereby an alloy from which sodium cannot be distilled is formed. The resulting sodium vapor is precipitated in a monel exhaust tube which is subjected to temperature control. Sodium can be rinsed from the monel exhaust tube with water.

Abstract: Ruthenium in aqueous solution in a first, oxidizable oxidation state (e.g. as RuNO(NO.sub.3).sub.3) is converted to an insoluble form in a second, different oxidation state (e.g. as RuO.sub.2.nH.sub.2 O) by establishing an electrochemical cell wherein the solution is the electrolyte and electrochemically oxidizing and reducing the ruthenium in the cell. The insoluble form may be filtered from the liquid. The ruthenium treatment may be a stage in the removal of radioactive species from liquids such is in the treatment of medium and low level activity liquid waste streams, wherein actinides are precipitated and filtered off either before or after ruthenium treatment. Subsequently, residual activity may be removed from the stream by either or both of (a) absorption, followed by filtration and electro-osmotic dewatering and (b) electrochemical ion exchange. Filtration fluxes may be maintained by direct electrochemical membrane cleaning.

Abstract: A process for the extraction of cations from an aqueous effluent. An effective quantity of at least one polymer or copolymer, which is physically or chemically crosslinked, is introduced into the aqueous effluent. The polymer or copolymer is based on one or more units selected from salified acrylic acid, quarternized acrylic acid, and a copolymer comprising from 40 to 60 mole % of acrylic acid and from 60 to 40 mole % of at least one dialkylaminoalkyl acrylate in which each alkyl group contains from 1 to 4 carbon atoms. This process is used to treat aqueous effluents.

Abstract: Compositions comprising carbon-based supports and containing heteropolycyclic compounds such as bipyridyl, phenanthroline or derivatives thereof are used in separation of dissolved late transition metals such as Fe, Co, Ni and Cu from solutions, in particular when such solutions contain high concentrations of other dissolved metals such as Zr, Hf, La, Al and Ba. Such compositions are also used to separate late transition metals such as Pt group metals from each other.

Abstract: There is provided a generator for short-lived radionuclides. The generator comprises a support, an ion-exchange agent and a parent radionuclide in a steady-state equilibrium with a daughter nuclide, which daughter nuclide can be selectively eluted from said column. A suitable parent radionuclide is .sup.191 OS in equilibrium with .sup.191m Ir. There is also provided a specific Os(VI) complex which has certain advantages.

Abstract: A process for preparing a tungsten-188/rhenium-188 generator having a tungstate matrix containing W-188 produced by irradiating tungsten-186 in the tungstate compound. High activity, carrier-free rhenium-188 may be obtained by elution. Substrates for further purifying the rhenium-188 eluate are also described.

Abstract: Compounds of the formula I:R+[.sup.99m TcNX.sub.4 ]- Iwherein R+ represents a cation and X represents a halo group, are prepared by reaction with an azide compound in the presence of a hydrohalic acid. The compounds are useful in production of .sup.99m Tc-labelled radio- pharmaceuticals by reaction with ligands and monoclonal antibodies or antibody fractions.

Abstract: Silica gels modified with amino groups or magnesium silicates are suitable carrier materials for technetium-99m generators since they retain copper(II) ions well and thus produce a copper-free eluate.

Abstract: Contaminated surface layers are decontaminated by treatment with an aqueous fluorine base-containing decontamination solution. The aqueous decontamination solution contains 0.05 to 50 Mol of decontamination agent per liter, and the decontamination agent preferably comprises at least one substance from the group, hexafluorosilicate acid, fluoroboric acid, and the salts of both these. The decontamination solution produces the required high decontamination factors on metallic substances and brickworks as well. The used decontamination solution can, after regeneration, be recycled into the decontamination process.Release of decontaminated material by dissolution of the surface layer of the decontaminated objects provides decontamination of objects having complicated and hard-to-measure geometries.The decontamination agent (HBF.sub.4 -acid) is advantageously produced from contaminated boric acid from pressurized water reactor wastes by reaction with fluoride or hydrofluoric acid. The HBF.sub.

Abstract: The invention relates to a method of and apparatus for the regregation of radioactive iodine isotopes from fluid samples. The method consists in leading the degassed and in certain cases also otherwise prepared sample into a column filled with an amorphous solid medium for binding quantitatively the cations, fluoride anions and contaminations of colloidal state, then the effluent flow continuously through an energy-selective gamma detector and continuously recording the signals generated in the detector by the radioactive iodine isotopes. The proposed apparatus comprises a sampling unit (MV), a degassing vessel (G), fluid transfer pumps (P1, P2), pipings, cocks (CS), a bubble removing cell (BC), a column (K) with amorphous material filling, an energy-selective detector system (GD) for measuring gamma radiation, and a signal processing and recording unit (JR).

Abstract: A process for continuously separating uranium isotopes by an oxidation-reduction reaction using an anion exchanger comprising the steps of feeding an eluant containing a deactivated oxidizing agent and a deactivated reducing agent to an anion exchanger to regenerate them and to form an oxidizing agent zone in the anion exchanger, and reusing the regenerate reducing agent and the oxidizing agent zone for further uranium isotope separation.

Abstract: Method for producing isotopically enriched material by vibration-vibration excitation of gaseous molecules wherein a middle mass isotope of an isotopic mixture including lighter and heavier mass isotopes preferentially populates a higher vibrational mode and chemically reacts to provide a product in which it is enriched. The method can be used for vibration-vibration enrichment of .sup.17 O in a CO reactant mixture.

Abstract: A method of reprocessing neutron-irradiated contaminated boron carbide, including the steps of:dissolving the contaminated boron carbide in a mixture of sulfuric and nitric acids at an elevated temperature, to convert the boron carbide to boric acid, and to form an exhaust gas of nitric oxide and carbon dioxide;introducing water or tritiated, boric acid-containing water in liquid form under the surface of the hot acid mixture of the dissolving step, to produce steam;steam distilling the boric acid developed in the dissolving step with the aid of the steam generated in the introducing step, to separate the boric acid from nonvolatile radioactive and nonradioactive corrosion and irradiation products;condensing the boric acid-containing steam from the steam distilling s tep to yield boric acid and a condensate containing water or .sup.3 H-containing water;evaporating the condensate from the condensing step to dry and calcine the boric acid to form boron oxide and water vapor of .sup.

Abstract: A process for the stripping of cesium ions from an aqueous solution in which a precipitation agent is added to the aqueous solution and the resulting precipitate, containing the CS.sup.+ ions is stripped from the solution. Sodium or lithium tetraphenylborates, carrying electron-attracting substituents on the phenyl rings are employed as precipitation agent.

Abstract: A method is provided for removing sulphur-containing gases from fossil fuel-fired combustors comprising the step of introducing sorbent particles modified with lignosulfonates into the exhaust gases.

Abstract: This invention relates to a standard component .sup.99m TC elution generator useful for medical purposes and consisting of prefabricated component parts. The main generator column of the device may be used both as an irradiation container and an elution container, enabling the user to supply activated or nonactivated parts. The main generator column made from by neutrns little activable materials serves first as reactor irradiation ampoule and after having been activated in the reactor by neutrons and after a simple adjustment serves directly as the generator column. It is filled with water insoluable molybdates or polymolybdates (with the molybdenum content in the range 10-40%), easily releasing .sup.99m Tc generated by radioactive decay of the mother .sup.99 Mo formed in it by neutron activation. This column filling serves originally as target material for reactor irradiations and afterwards it is directly used as the generator elution matrix.

Abstract: The process of the present invention comprises adding an alkaline earth metal hydroxide such as barium hydroxide to a radioactive liquid waste containing sodium sulfate as the main component to convert the latter into an insoluble alkaline earth metal salt such as barium sulfate, adding silicic acid to by-product sodium hydroxide to prepare water glass and solidifying the radioactive insoluble alkaline earth metal salt with the water glass. According to this process, exudation of radioactive substances from the solid can be prevented and the solid having a high durability can be obtained at a low cost.

Abstract: A process for the recovery of an organic phase and an aqueous phase from three phase emulsions such as those formed during the extraction of uranium from wet phosphoric acid is described. Such emulsions contain a light-density liquid phase and a heavy-density liquid phase and are stabilized by a solid phase of intermediate density. The emulsion is first heated and then separated into its three components using a centrifugal filter with an appropriate filter aid. At high filter feed rates the dense liquid phase passes freely through the filter media while the light phase liquid begins to accumulate in the centrifuge and is withdrawn from the centrifuge via a properly positioned skimmer tube. The apparatus containing the skimmer tube is also described.

Abstract: A process for the separation of substances hindering the recovery of the fissionable materials uranium and plutonium and for the separation of the fissionable materials to be recovered in a reprocessing process for spent, irradiated nuclear fuel- and/or fertile materials. A second and a third wash of the organic phase is performed for residual ruthenium separation and residual zirconium separation, and there is a repetition of the Pu stripping step with simultaneous electrolytic reduction of Pu(IV) to Pu(III). The second and third wash solutions or the aqueous phase employed for the repetition of plutonium stripping, respectively, each contains a high concentration of product uranyl nitrate. The aqueous run-off from the second and third washes and from the repetition of the Pu stripping step is indirectly fed back into the aqueous fuel solution employed in the first extraction step of the reprocessing method by first feeding these run-offs to an intermediate treatment.

Abstract: Cesium is extracted from acidified nuclear waste by contacting the waste with a bis 4,4'(5) [1-hydroxy-2-ethylhexyl]benzo 18-crown-6 compound and a cation exchanger in a matrix solution. Strontium is extracted from acidified nuclear waste by contacting the waste with a bis 4,4'(5') [1-hydroxyheptyl]cyclohexo 18-crown-6 compound, and a cation exchanger in a matrix solution.

Abstract: In a process for continuously separating uranium isotopes, .sup.235 U and .sup.238 U, by oxidation-reduction chromatography using a column of an anion exchanger, the separation of uranium isotopes and the regeneration of the oxidizing agent and reducing agent both deactivated during the separation of uranium isotopes can be simultaneously effected in a single column by returning to and passing through the column of an anion exchanger having a uranium adsorption zone an eluate containing a deactivated oxidizing agent and a deactivated reducing agent to regenerate the deactivated oxidizing agent and the deactivated reducing agent and to form an oxidizing agent zone, and returning to and passing through the column the regenerated reducing agent for the reduction of uranium.

Abstract: An organic extraction phase, containing fission materials uranium and plutonium, as well as a minor amount of fission- and corrosion-products, from an extraction step or from a purification step of a nuclear fuel reprocessing process is treated with an aqueous, basic, carbonate ion containing solution, whereby the fission materials and at least a part of the fission- and corrosion-products are stripped into the aqueous phase and the uranium is present in the form of uranium-carbonato-complex. The organic phase is then separated from the aqueous phase. Plutonium is then removed from the separated aqueous phase. The remaining aqueous solution containing the uranium-carbonato-complex and a small residual amount (C) of fission products is adjusted to a ratio of uranyl ion concentration to carbonate ion- or CO.sub.3.sup.-- /HCO.sub.3.sup.-- concentration of 1(UO.sub.2.sup.++) to 4.5(CO.sub.3.sup.-- or CO.sub.3.sup.-- /HNO.sub.3.sup.-), respectively, or more, at a maximum U-concentration of not more than 6- g/l.

Abstract: An initial aqueous pertechnetate solution which has a low electrolyte concentration compared with physiological saline is treated by passage through a bed of insoluble ion exchange material for adsorbing the pertechnetate. In a second step an ionic eluant having a smaller volume than the pertechnetate solution is passed through the bed to remove the pertechnetate to provide a relatively concentrated and pure eluate solution, the ion exchange material in the bed and the ionic eluant having respective properties for providing the eluate with a pH suitable for radiopharmaceutical purposes. The bed retains thereon radionuclidic impurities which may have been in the initial solution. The method can be valuable for treating the eluate from a portable technetium generator which is eluted with water and advantageously the insoluble material in the bed is zirconium oxide.

Abstract: Potassium osmate, of the formula K.sub.2 Os O.sub.2 (OH).sub.4), used to make a column for the generation of Ir-191 m, which is used in first pass angiography to detect cardiac defects in patients.

Abstract: An efficient and economical separation of the gadolinium isotopes Gd.sup.155, Gd.sup.156 and/or Gd.sup.157 from mixtures containing the same is effected by a liquid chromatographic column system using an ion exchange resin and an eluant solution. A portion of the eluant solution is used to saturate the columns and a feed solution containing the mixture of gadolinium isotopes is charged to the columns. Further eluant solution is passed through the saturated columns containing the feed solution to form a fraction of eluant solution rich in the desired gadolinium isotopes which is separated from the remainder of the eluant upon discharge from the columns.

Abstract: The invention relates to a process for the recovery of molybdenum-99 from an irradiated uranium alloy target.This process comprises the following stages:(a) dissolving the irradiated uranium alloy target in sulphuric acid,(b) separating the iodine and tellurium present in the thus obtained solution,(c) oxidizing the molybdenum-99 present in the Mo (VI) solution, and(d) extracting the thus oxidized molybdenum-99 with a hydroxamic acid of formula: ##STR1## in which R is a radical chosen from the group including straight or branched-chain alkyl radicals, the phenyl radical, phenyl radicals substituted by at least one alkyl radical, arylalkyl radicals and the cyclohexyl radical and R' is a hydrogen atom or an alkyl radical.The hydroxamic acid can be tri-n-butylacetohydroxamic acid and the molybdenum recovered can be used in a technetium generator for medical purposes.