Abstract: A method and a generator for separating daughter radioisotope from a stock solution containing the daughter radioisotope and its parent are provided. The generator is provided with a glass adsorbent which preferentially adsorbs daughter radioisotope relative to the parent radioisotope and circulation means for withdrawing the stock solution from a reservoir, contacting the solution with the glass adsorbent to adsorb the daughter radioisotope and returning the parent radioisotope-enriched solution to the reservoir. In the process stock solution containing daughter radioisotope having a predetermined pH is contacted with the glass adsorbent to selectively adsorb the daughter radioisotope thereon and the daughter radioisotope is eluted with an eluant at a second predetermined pH. The process also includes a step of washing the glass adsorbent with a wash solution at the first predetermined pH to remove traces of the parent radioisotope from the adsorbent prior to eluting the daughter radioiotope.

Abstract: The invention provides a new high temperature isotopic exchange fuel processing loop for tritium recovery in which tritium is exchanged out of impurities such as tritiated ammonia, methane and water by swamping with H.sub.2 and isotopically equilibrating the mixture in a high temperature reactor. Downstream of the reactor is a Pd/Ag permeator for separation of the hydrogen from the impurities, the separated hydrogen being sent to an isotope separation system for tritium recovery and the residual impurities being recycled. The process eliminates the need for impurity oxidation and electrolysis of DTO and does not rely on complicated catalytic decomposition reactions.

Abstract: A process of preparing an actinide compound of the formula An.sub.x Z.sub.y wherein An is an actinide metal atom selected from the group consisting of thorium, uranium, plutonium, neptunium, and americium, x is selected from the group consisting of one, two or three, Z is a main group element atom selected from the group consisting of nitrogen, phosphorus, oxygen and sulfur and y is selected from the group consisting of one, two, three or four, by admixing an actinide organometallic precursor wherein said actinide is selected from the group consisting of thorium, uranium, plutonium, neptunium, and americium, a suitable solvent and a protic Lewis base selected from the group consisting of ammonia, phosphine, hydrogen sulfide and water, at temperatures and for time sufficient to form an intermediate actinide complex, heating said intermediate actinide complex at temperatures and for time sufficient to form the actinide compound, and a process of depositing a thin film of such an actinide compound, e.g.

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: A process for recovering At-211 from a target and simultaneously producing an At-211-labeled radiopharmaceutical agent involves introducing volatilized At-211, produced by dry distillation of an irradiated Bi-209 target, into a solution containing a compound that binds the At-211 to form an At-211-labeled radiopharmaceutical agent. The radiopharmaceutical agent may have therapeutic use as is, or may be attached to a targeting protein such as an antibody prior to administation to a patient.

Abstract: Layered divalent metal pentavalent metallate, M(HAO.sub.4).sub.x (RAO.sub.3).sub.y (Z).sub.z wherein M is divalent metal, A is a pentavalent metal, e.g. phosphorus, R is a substituent group other than H or OH, e.g., phenyl, covalently bonded to A, x+y=1, y is greater than 0, z ranges from 0 to 2, inclusive, and Z is an intercalated moiety is prepared by contacting a source of divalent metal with a substituted phosphonic acid-type compound under hydrothermal conditions.

Abstract: A diagnostric device, composition and method of diagnosing airway dysfunction in a patient is disclosed. The apparatus and method require the subjecting of a pharmaceutical acceptable radionuclide, that is the composition, to an elevated temperature in an enclosed space in the presence of either an inert gas or oxygen to produce an inhalable product. The product is inhaled by a patient. A film is located adjacent the airways enabling mapping of the deposition of the radionuclide in the airways of the patient's lungs.

Abstract: A container for radioactive material includes therein a catalyst device for catalyzing the recombination of gaseous hydrogen and oxygen in the container to give water, e.g. in a storage vessel for radioactive material. The device comprises a substrate in the form of at least one metal member (e.g. stainless steel) carrying a thermally sprayed low surface area ceramic coating (e.g. plasma sprayed alumina) which carries a high surface area ceramic coating (e.g. alumina from a sol), which latter coating carries catalytically active material (e.g. Pt metal), the catalyst having an outer, hydrophobic gas permeable coating (e.g. polysiloxane) for preventing the catalytically active material from contacting water.

Abstract: The process of the present invention includes preparing a first aqueous solution of rhenium in the form of radioactive perrhenate, wherein the concentration of rhenium in said solution is within the range of about 5.times.10.sup.-6 M to about 2.times.10.sup.-3 M, and then reducing and complexing the radioactive perrhenate by mixing a second solution or lyophilized solid with the first solution. The second solution or lyophilized solid comprises a ligand which complexes with the radioactive perrhenate and a reductant wherein the reductant is present in the second solution at a concentration in the range of about 0.005M to about 0.020M and the ligand is present in the second solution at a concentration in the range of about 0.01M and about 0.15M. The pH of the resultant solution is within the range of about 1.5 to about 5.5.

Abstract: A method for concentrating an isotope from an admixture containing the isotope by contacting the admixture with a catalyst capable of selectively catalyzing a reaction with the isotope, and recovering the concentrated isotope from the reaction products. In one embodiment, the method can be used to concentrate deuterium from sources such as hydrogen-rich gas.

Abstract: Tritium and deuterium are separated from a gaseous mixture thereof, derived from a nuclear fusion reactor or some other source, by providing a casing with a bulk getter therein for absorbing the gaseous mixture to produce an initial loading of the getter, partially desorbing the getter to produce a desorbed mixture which is tritium-enriched, pumping the desorbed mixture into a separate container, the remaining gaseous loading in the getter being deuterium-enriched, desorbing the getter to a substantially greater extent to produce a deuterium-enriched gaseous mixture, and removing the deuterium-enriched mixture into another container. The bulk getter may comprise a zirconium-aluminum alloy, or a zirconium-vanadium-iron alloy. The partial desorption may reduce the loading by approximately fifty percent. The basic procedure may be extended to produce a multistage isotope separator, including at least one additional bulk getter into which the tritium-enriched mixture is absorbed.

Abstract: Layered divalent metal pentavalent metallate, M(HAO.sub.4).sub.x (RAO.sub.3).sub.y (Z).sub.z wherein M is divalent metal, A is a pentavalent metal, e.g. phosphorus, R is a substituent group other than H or OH, e.g., phenyl, covalently bonded to A, x+y=1, y is greater than O, z ranges from 0 to 2, inclusive, and Z is an intercalated moiety is prepared by contacting a source of divalent metal with a substituted phosphonic acid-type compound under hydrothermal conditions.

Abstract: The present invention comprises a process for separating rare earth ions or actinide ions or mixtures thereof in solution by passing the solution through an ion exchange material to separate the rare earths or actinides or mixtures thereof. The ion exchange material has a surface area of about 5-1500 m.sup.2 /g. The ion exchange material is impregnated with a liquid containing alkali metal cations, Group Ib metal cations, ammonium cations, organic amines or mixtures thereof, at a pH range above about 9. A plurality of fractions of the solution is collected as the solution passes through the ion exchange material, preferably in a column. This process is particularly preferred for separating rare earth ions and especially lanthanum and neodymium. It is particularly preferred to purify lanthanum to contain less than 0.1 ppm, preferably less than 0.01 ppm, of neodymium.

Abstract: Desulphurization of saturated or near saturated gaseous hydrocarbon stream is effected using a bed of a particulate adsorbent comprising zinc oxide. Condensation of the higher boiling components is avoided by heating the raw gas to a temperature at least 10.degree. C. above its dew point. This heating is effected by heat exchange with the heated desulphurized gas and by heat exchange with the combustion products of a minor part stream taken from the raw gas and/or from the desulphurized product gas stream. The absorbent particles preferably have a high BET surface area and pore volume.

Abstract: The invention relates to a process and an installation for the treatment of tritium-contaminated, solid organic waste. The waste is contacted with the steam in enclosure (1) for extracting the tritium in the steam and the steam is then condensed at (11 and 13) to recover the tritium from the waste in the form of tritiated water.

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: .sup.1/3 Holmium in a carrier metallic hydroxide aggregate is disclosed for the treatment of arthritis and, in particular, rheumatoid arthritis. The compound disclosed preferably has a particle size of 1 to 40 microns, Beta energy emissions in the range of 1.76-1.84 MeV, low levels of gamma ray emissions and a radioactive half-life of 26.8 hours. The preferred metallic hydroxide is selected from the group consisting of Ferric Hydroxide, Aluminum Hydroxide, Bismuth Hydroxide, Chromium Hydroxide, Cupric Hydroxide, Manganese Hydroxide and Stannous Hydroxide. Methods are also disclosed for the preparation of the compound, as well as for the methods of its administration to a patient in need thereof.

Abstract: Preparation of metal carbides, nitrides, borides, silicides and phosphides, also metal alloys and pure metals, by providing a precursor in which there are organic ligands bonded to the metal or metals, such precursor having the element X also bonded directly or indirectly to the metal or metals, the ligand-metal bonding being weaker than the X-metal bonding whereby on pyrolysis the product M.sub.a X.sub.b results in which M represents the metal or metals, X represents C, N, B, Si, P and a and b represent the atomic proportions of M and X. The subscript b may be zero if an alloy or pure metal is to be prepared. The product M.sub.a x.sub.b can be prepared by relatively low temperature pyrolysis and the precursor can be used as a solution or a low melting solid. This enables one to apply a surface coating or to shape the precursor into a fiber, rod or other shape and to pyrolyze the coating or shaped article. M is a transition, lanthanide or actinide metal or tin.

Abstract: A process for the enrichment and separation of oxides of heavy hydrogen isotopes from acid, aqueous solutions or other aqueous streams employs macrocyclic aminopolyether (APE) and organic cation exchange agents. The aqueous solution or the aqueous stream is mixed with at least one organic solvent which is miscible with water to form a liquid mixture. An organic acid cation exchange agent in the H-form is charged with a protonized APE to form a solid phase. The liquid mixture is brought into contact with the solid phase to bring about the enrichment of the heavy hydrogen isotope on the solid phase at a low temperature. The heavy hydrogen isotope from the solid phase is then released at a temperature that is higher by between 30.degree. K. and 150.degree. K. than the temperature at which the enrichment occurred.

Abstract: A process for dissolving PuO.sub.2, NpO.sub.2, or fuel containing PuO.sub.2 and/or NpO.sub.2 in acid, particularly nitric acid by electrolytically continuously maintaining an oxidizing regenerable reagent. The regenerable reagent may be added to the mixture of oxide(s) and nitric acid. Alternately, it may already be present as a fission product or other nuclear reaction product in irradiated nuclear reactor fuel, in which case it need not be added to the nitric acid.

Abstract: A process for reliably and consistently producing astatine-211 in small controlled volumes of a solution, which is selected from a choice of solvents that are useful in selected radiopharmaceutical procedures in which the At-211 activities are to be applied.

Abstract: A method for simultaneously preparing Radon-211, Astatine-211, Xenon-125, Xenon-123, Iodine-125 and Iodine-123 in a process that includes irradiating a fertile metal material then using a one-step chemical procedure to collect a first mixture of about equal amounts of Radon-211 and Xenon-125, and a separate second mixture of about equal amounts of Iodine-123 and Astatine-211.

Abstract: Beta double prime alumina is provided having a sensible amount of at least one polyvalent cationic species intercalated therein. In accordance with a preferred embodiment, such beta double prime alumina is provided having trivalent and/or tetravalent cationic species intercalated therein, especially species derived from the lanthanide and actinide series of elements. Certain of the foregoing materials exhibit phosphorescence or fluorescence, and some are believed to be capable of producing laser emission upon suitable irradiation.Methods for modifying beta double prime aluminas comprising contacting crystals of the aluminas with polyvalent cation-containing salts such as in the molten state or in the gaseous state are also provided. Laser and other optical devies are disclosed employing the modified beta double prime aluminas of this invention.

Abstract: A process and apparatus are described for separating heavy isotopes of hydrogen. The process uses multiple protium-deuterium, protium-tritium or deuterium-tritium exchange reactions between a metal protide and gaseous hydrogen alternately depleted and enriched in at least one heavy isotope of hydrogen. The apparatus for performing the process comprises two multistage reaction zones, one for forming a metal hydride enriched in at least one heavy isotope and one for converting the enriched hydride to a depleted hydride and enriched hydrogen gas. The apparatus is simple in design and the process is much less energy intensive than existing prior art processes.

Abstract: A method of recovering tritium from tritiated compounds comprises the steps of heating tritiated water and other co-injected tritiated compounds in a preheater to temperatures of about 600.degree. C. The mixture is injected into a reactor charged with a mixture of uranium and uranium dioxide. The injected mixture undergoes highly exothermic reactions with the uranium causing reaction temperatures to occur in excess of the melting point of uranium, and complete decomposition of the tritiated compounds to remove tritium therefrom. The uranium dioxide functions as an insulating material and heat sink preventing the reactor side walls from attaining reaction temperatures to thereby minimize tritium permeation rates. The uranium dioxide also functions as a diluent to allow for volumetric expansion of the uranium as it is converted to uranium dioxide.

Abstract: A method is described and claimed for the replenishment of deuterium or tritium concentration in a laser isotope separation process for the production of D.sub.2 O, or for the recovery of tritium from contaminated D.sub.2 O or H.sub.2 O. A working compound is selectively photodissociated by a laser beam and is replenished by contacting an exchange liquid in a countercurrent fashion in a first contacting column. The exchange liquid is replenished with isotope by in turn being contacted with a feed stream in a second contacting apparatus. This second contacting apparatus may be a gas/liquid or liquid/liquid contacting apparatus and the countercurrent flows therein may be about equal or unequal.

Abstract: The multi-stage process separates at least one heavy isotope from a hydrogen-containing compound or a hydrogen containing mixture, using ammonia synthesis and a mixture of hydrogen and nitrogen. The main product is water at least substantially free of deuterium and tritium, additional products being compounds enriched in deuterium and tritium, and nitrogen enriched in .sup.15 N.

Abstract: A process for the incremental enrichment of deuterium and/or tritium in a material which is suitable for the isotope exchange of deuterium and tritium with hydrogen, and an arrangement for the implementation of the process. The process and arrangement for the enrichment of deuterium and/or tritium in water which, in addition to a high transport speed for the molecules which participate in the isotope exchange, evidences a high enrichment factor for each enrichment stage and a high yield, so that at a relatively small number of stages and low energy consumption there is attainable an overall high degree of enrichment. For each enrichment stage, water containing deuterium and/or tritium is introduced into a carrier gas flow, reduced and set to a hydrogen (H.sub.2) partial pressure of maximally 100 mbar.

Abstract: A method and apparatus are described for the joint production and separation of tritium. Tritium is produced in an aqueous homogenous reactor and heat from the nuclear reaction is used to distill tritium from the lower isotopes of hydrogen.

Abstract: A process for enriching and separating heavy hydrogen isotopes having a heavy hydrogen cation (deuterium and/or tritium) from substance streams containing them, wherein the respectively present hydrogen isotopes are exchanged in chemical equilibria. A protic, acid solution containing deuterium and/or tritium is brought into contact with a value material from the group of open-chained polyethers or aminopolyethers, macro-monocyclic or macro-polycyclic polyethers, macro-monocyclic or macro-polycyclic amino polyethers, and mixtures of these values, in their free or proton salt form to form a reaction product of the heavy hydrogen cation with the value or value salt and bring about enrichment of deuterium and/or tritium in the reaction product. The reaction product containing the value or value salt is separated from the solution.

Abstract: Hydrogen isotope (e.g. deuterium) exchange from liquid water to a gaseous halohydrocarbon (e.g. fluoroform, CF.sub.3 H--CF.sub.3 D) is obtained at an operating temperature in the range 0.degree. to 100.degree. C. using a catalytically active mass comprising a porous anion exchange resin in the hydroxide ion form and enriched gaseous halohydrocarbon stream is decomposed by isotope selective photo-decomposition into a first, gaseous stream enriched in the hydrogen isotope, which is removed as a product, and a depleted gaseous halohydrocarbon stream, which is recirculated for enrichment again. The catalytically active mass may, for example, be in the form of resin particles suspended in a fluidized bed or packed as resin particles between sheets wound into a roll. One of the sheets may be corrugated and have open interstices to form a packing in a column which permits countercurrent gas and liquid flow past the resin. Preferably the wound sheets are hydrophilic to retard flooding by the liquid water.

Abstract: The invention is a method for detecting the presence of or a predisposition to cancer in mammalian individuals. A trace element, specifically tin, is administered to the individual in the form of a radioactive isotope or a compound containing a radioactive tracer. The tin-bearing compound thereafter accumulates in the various bodily organs and is excreted. Indicative of the presence of cancer or a predisposition is the lack of accumulation of the radioactive tin compound in the thymus and lymphatic system of the mammal. By comparison of the radioisotope concentration in the urine and blood, the early detection of a precancerous or leukemic condition can be made. The method is applicable to a wide variety of mammals, including human beings.

Abstract: The invention relates to a process and to an apparatus for treating solutions containing tritiated water.According to this process, an electrolyte such as soda is added to the solution, the solution obtained then undergoes electrolysis in order to release tritium in the gaseous state. Working takes place in an electrolytic cell having a cathode made from a metal which aids the diffusion of tritium, e.g. a Pd-Ag alloy coated with porous palladium black. The cathode forms a separating wall between the solution to be electrolyzed and a tritium reception compartment. The tritium which has diffused through the wall of the cathode is recovered in the compartment.Use for the recovery of tritium from effluents resulting from the reprocessing of irradiated nuclear fuels.

Abstract: A waste gas generated in the reprocessing of used nuclear fuel is at first subjected to removal of explosive, gaseous substances such as hydrocarbons and nitrogen oxides, and materials capable of forming the explosive gaseous substances, such as oxygen, and condensible substances such as carbon dioxide, water and ammonia from the waste gas, and then to cryogenic distillation, thereby separating and recovering Kr-85 from the waste gas. As the separation and recovery of Kr-85 is carried out by cryogenic separation after the removal of the substances having a possibility to explode in a cryogenic distillation apparatus and also the removal of condensible substances having a possibility to clog a piping system of the apparatus, the operation of the apparatus is ensured. It is also disclosed that the oxygen and nitrogen oxides can be completely removed by catalytic hydrogen reduction.

Abstract: A method of enhancing the thermal desorption of a first isotope of a diatomic gas from a metal comprises the steps of (a) establishing a partial pressure of a second isotope of the diatomic gas in vicinity of the metal; heating the metal to a temperature such that the first isotope is desorbed from the metal; and reducing the partial pressure of the desorbed first isotope while maintaining the partial pressure of the second isotope substantially constant. The method is especially useful for enhancing the desorption of tritium from the Zr-Al getter in a plasma confinement device.

Abstract: Hydrogen isotope is separated and recovered with remarkable reduction in energy consumption in a compact apparatus using a cold reactor and a hot reactor based on dual temperature process by a combination of a first step of at least one isotope exchange reaction stage comprising contacting mists of a feed liquid cocurrently together with a reacting gas with an isotope exchange reaction catalyst at a low temperature, and a second step of a plurality of stages each comprising a gas-liquid contact of the liquid from the first step with vapor of the liquid and the reacting gas and an isotope exchange reaction between the vapor of the liquid and the reacting gas at a high temperature.

Abstract: A process for desorbing fission iodine from a solution in a dissolver containing nitric acid and nuclear fuel. At least part of the solution is distilled by boiling creating an iodine-containing vapor, and the condensate vapor is conducted in an ascending condenser in countercurrent with its condensate. The resulting condensate is the returned to the dissolver. The desorption of the iodine in the condenser is effected by means of a gas.

Abstract: An apparatus for isotope exchange reaction is used in an apparatus for removing tritium in heavy water-moderated nuclear reactors or heavy water production plants, and comprises a plurality of mist generators and a plurality of reactor columns, the mist generators and the reactor columns being alternately arranged and connected to one another successively in the alternate order. An ultra-sonic generator is provided each in the mist generators. A hydrophobic catalyst bed and a mist separator are provided each in the reactor columns. Water containing hydrogen isotopes to be removed is led into the mist generators to form mists. Mists of the water are supplied into the reactor columns together with hydrogen gas and isotope exchange reaction is carried out in the hydrophobic catalyst beds.

Abstract: Process for the selective extraction of iodine from aqueous solutions containing the same, comprising utilizing as the iodine solvent a hydrophobic compound comprising ether groups, having a non-ionic character, of the type ##STR1## in which R is a methyl and/or ethyl group, n is a number from 1 to 200, and X is the remainder of an hydropholic compound which comprises at least one reactive hydrogen atom.

Abstract: A process for the removal of tritium from the product solutions obtained in the reprocessing of irradiated nuclear fuels by the Purex process comprising a plurality of series-connected extraction cycles having an organic solvent.

Abstract: The invention relates to a method of separating isotopes from a gas mixture, of once ionized uranium the isotopes being separated by the combined effect of photon irradiation by laser light followed by Penning ionization of ions selectively excited by the laser light.According to the invention a gas of once ionized uranium isotopes is irradiated with laser light made up of photons having an energy which resonates with an excitation level of an isotope of the gas mixture. Metastable atoms or molecules are simultaneously introduced into the mixture and have an excitation energy which is less than the energy for a further ionization of the constituents of the mixture and greater than the energy gap between the energy for a further ionization of the isotope excited by the laser light and the energy of the excitation level of the same isotope.

Abstract: Mercuric nitrate-nitric acid scrub solutions containing radioiodine may be reduced in volume without excessive loss of volatile iodine. The use of concentrated nitric acid during an evaporation process oxidizes the mercury-iodide complex to a less volatile mercuric iodate precipitate.

Abstract: Disclosed is a method for treating substantially homogeneous monocalcium phosphate (MCP) solutions that contain undesirable amounts of radium which comprises centrifuging these MCP solutions sufficiently to separate at least a portion of the radium from the solution. Such treated MCP solutions may then be reacted with sulfuric acid to simultaneously produce a phosphoric acid solution and a calcium sulfate precipitate such as gypsum having low levels of radium content.

Abstract: A method of reducing the pertechnetate in TcO.sub.4.sup.- comprises mixing together an aqueous solution of pertechnetate, e.g. the eluent from a technetium generator, metallic tin or an alloy thereof as a reducing agent for the pertechnetate, and a soluble salt of a metal below tin in the electrochemical series, e.g. copper, as an activator for the tin metal reducing agent. A complexing agent for the reduced technetium or a colloid stabilizer may also be included. The pH is preferably 3 to 12. Also claimed is a closed container containing the tin reducing agent, the activator, and the complexant or colloid stabilizer if used, preferably in a freeze-dried sterile state, to which the pertechnetate solution may be added.

Abstract: A method is disclosed for controlling the release of radioiodine produced by the fission of uranium 235 during the process for the production of molybdenum 99. In the process for producing molybdenum 99, an aluminum-uranium alloyed target is irradiated with a neutron flux to bring about fission of the Uranium 235 producing molybdenum 99, radioiodine and other fission by-products. The method herein comprises reacting the irradiated target with a caustic solution. A reactive silver is then mixed with the caustic solution in an amount sufficient to react with substantially all of the radioiodine present in the caustic solution thereupon forming a silver precipitate of radioiodine. The precipitate formed is then separated from the caustic solution resulting in a substantially radioiodine-free caustic solution containing molybdenum 99.

Abstract: Disclosed is a process for producing a calcium sulfate product containing low amounts of radium comprising the steps of:(a) adding at least one sequestering agent to a substantially homogeneous radium-containing monocalcium phosphate (MCP) solution, the amount of the sequestering agent added being at least the stoichiometric amount necessary to complex with the radium in the MCP solution;(b) reacting the treated MCP solution with H.sub.2 SO.sub.4 to form a monocalcium phosphate/phosphoric acid solution, to precipitate a first calcium sulfate product, and to precipitate at least a portion of the radium, the amount of said H.sub.2 SO.sub.4 being up to about 50% of the stoichiometric amount necessary to react with the total CA.sup.++ present in the MCP solution;(c) separating the monocalcium phosphate/phosphoric acid solution from the radium-containing calcium sulfate precipitate;(d) then reacting the monocalcium phosphate/phosphoric acid solution with H.sub.2 SO.sub.

Abstract: The invention discloses diagnostic compositions for use in obtaining images of a patient's lungs. The basic components of the composition of the invention are sodium pertechnetate which is radioactive and ethanol. This composition may be combusted and the resulting products cooled or alternatively the composition may be inserted into a pressure vessel with an aerosol. In both cases a gas like mixture results. A particular advantage is that a patient is able to breath the mixture of the invention in a normal way and does not need to undergo any training in inhalation.

Abstract: An isotope separation apparatus comprises a plurality of independent developing units, each comprising 2 to 20 adsorbent-packed columns forming a continuous developing circuit or passageway, and the developing units are connected to at least one common main pipe for supplying an isotope mixture solution, a regenerating agent solution, or an eluent solution. Also, in a further embodiment the developing units are connected to common liquid-discharge main pipes.The separation or concentration of isotopes such as uranium isotopes, nitrogen isotopes, boron isotopes, etc., is performed by continuously developing the isotope mixture solution passed through the individual adsorbent-packed columns successively in each developing units.

Abstract: The process for oxidizing irradiated fuel pellets using as the oxidant NO.sub.2 or a mixture of NO.sub.2 with its dissociation products, O.sub.2 and NO, in order to release the Volatile Fission Products (iodine, zenon, krypton and tritium) from the pellets.

Abstract: A Technetium-99m generator has a matrix having a compound of molybdenum-99 bound into or forming the matrix, the compound of the molybdenum-99 being substantially insoluble in an eluant which can be used in a radiopharmaceutical and the molybdenum compound permitting diffusion of technetium-99m therethrough and elution therefrom. The molybdenum compound can be a monomolybdate, an isopolymolybdate or a heteropolymolybdate and zirconium molybdate is preferred although other cation molybdates may be used. Methods of preparation of the generator include dissolving irradiated molybdenum trioxide in alkaline solution and precipitating the molybdate at a selected pH and packing the precipitate in finely divided form into a column. Alternatively the generator may be formed from a non-radioactive, molybdenum compound and activated later when desired by passing a solution of a compound of molybdenum-99 through the column.