Abstract: Iodine-125 is produced by neutron irradiation of .sup.124 Xe gas to form .sup.125 Xe and permitting decay of .sup.125 Xe to form .sup.125 I. Irradiation of the xenon-124 is effected in a first chamber within an enclosure and decay is effected in a second chamber within the enclosure and free from neutron flux. The apparatus is submersible in a nuclear reactor pool so as to absorb any radiation escaping the apparatus during the process. Xenon can be caused to move between the chambers remotely, underwater. The second chamber is removable from said enclosure and is transported to a suitable location to recover the .sup.125 I from its interior. Such recovery is effected by admitting an aqueous wash solution into the second chamber, whereupon it is heated, causing water from the wash solution to reflux and cleanse the interior surfaces of the second chamber, thus creating an aqueous solution of .sup.125 I, which then is caused to drain into a suitable container.

Abstract: The current invention involves a means for the production and extraction of the isotope molybdenum-99 for medical purposes in a waste free, simple, and economical process. Mo-99 is generated in the uranyl sulphate nuclear fuel of a homogeneous solution nuclear reactor and extracted from the fuel by a solid polymer sorbent with a greater than 90% purity. The sorbent is composed of a composite ether of a maleic anhydride copolymer and .alpha.-benzoin-oxime.

Abstract: The present invention is a method of removing an impurity of plutonium, lead or a combination thereof from a mixture of radionuclides that contains the impurity and at least one parent radionuclide. The method has the steps of (a) insuring that the mixture is a hydrochloric acid mixture; (b) oxidizing the acidic mixture and specifically oxidizing the impurity to its highest oxidation state; and (c) passing the oxidized mixture through a chloride form anion exchange column whereupon the oxidized impurity absorbs to the chloride form anion exchange column and the 22.sup.9 Th or 2.sup.27 Ac "cow" radionuclide passes through the chloride form anion exchange column. The plutonium is removed for the purpose of obtaining other alpha emitting radionuclides in a highly purified form suitable for medical therapy.

Abstract: A process for the production of radiostrontium consists in that a target of metallic rubidium is bombarded by a flow of accelerating charged particles. The target of irradiated rubidium is melted, whereas the extraction of radiostrontium is carried out by sorption on the surface of a sorbing material immersed into the irradiated molten metallic rubidium. As the sorbing material, use is made of materials selected from the group consisting of heat-resistant metals or metallic oxides or silicon which are inert with respect to rubidium. The resultant radiostrontium is extracted from the irradiated rubidium. The temperature of the sorbing material is selected to be close to the optimum one for the sorption of radiostrontium which is within the range of from the melting point of metallic rubidium to 220.degree. C. And the temperature of molten rubidium is selected to be close to the optimum one for the desorption of radiostrontium within the range of from 220.degree. C. to 270.degree. C.

Abstract: Methods for the production of radionuclides suitable for use in radiopharmaceuticals for diagnostic and therapeutic applications, and specifically, to the production of .sup.186 Re, .sup.188 Re and other radionuclides such as .sup.195m Pt and .sup.198 Au using an inorganic Szilard-Chalmers reaction. Thin-film and powdered .sup.185 or 187 Re.degree. metal targets, and .sup.185 or 187 Re oxide/metal oxide target compositions with rhenium in a lower, relatively reduced oxidation state are prepared. The thin-film rhenium targets are aged for at least about 24 hours and then irradiated with neutrons in the present of an oxidizing medium sufficient to form a product nuclide in the higher oxidized state of perrhenate, ReO.sub.4.sup.-. Significantly, the rate and/or extent of oxidation of target nuclides which do not react with a neutron is controlled.

Abstract: The present invention is a method of obtaining a radionuclide product selected from the group consisting of .sup.223 Ra and .sup.225 Ac, from a radionuclide "cow" of .sup.227 Ac or .sup.229 Th respectively. The method comprises the steps of a) permitting ingrowth of at least one radionuclide daughter from said radionuclide "cow" forming an ingrown mixture; b) insuring that the ingrown mixture is a nitric acid ingrown mixture; c) passing the nitric acid ingrown mixture through a first nitrate form ion exchange column which permits separating the "cow" from at least one radionuclide daughter; d) insuring that the at least one radionuclide daughter contains the radionuclide product; e) passing the at least one radionuclide daughter through a second ion exchange column and separating the at least one radionuclide daughter from the radionuclide product and f) recycling the at least one radionuclide daughter by adding it to the "cow". In one embodiment the radionuclide "cow" is the .sup.

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 heavy ion generator is used with a plasma desorption mass spectrometer to provide an appropriate neutron flux in the direction of a fissionable material in order to desorb and ionize large molecules from the material for mass analysis.The heavy ion generator comprises a fissionable material having a high n,f reaction cross section. The heavy ion generator also comprises a pulsed neutron generator that is used to bombard the fissionable material with pulses of neutrons, thereby causing heavy ions to be emitted from the fissionable material. These heavy ions impinge on a material, thereby causing ions to desorb off that material. The ions desorbed off the material pass through a time-of-flight mass analyzer, wherein ions can be measured with masses greater than 25,000 amu.

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: Iodine-125 is produced by neutron irradiation of .sup.124 Xe gas to form .sup.125 Xe and permitting decay of .sup.125 Xe to form .sup.125 I. Irradiation of the xenon-124 is effected in a first chamber within an enclosure and decay is effected in a second chamber within the enclosure and free from neutron flux. The apparatus is submersible in a nuclear reactor pool so as to absorb any radiation escaping the apparatus during the process. Xenon can be caused to move between the chambers remotely, underwater. The second chamber is removable from said enclosure and is transported to a suitable location to recover the .sup.125 I from its interior. Such recovery is effected by admitting an aqueous wash solution into the second chamber, whereupon it is heated, causing water from the wash solution to reflux and cleanse the interior surfaces of the second chamber, thus creating an aqueous solution of .sup.125 I, which then is caused to drain into a suitable container.

Abstract: Medical isotopes are produced using a lower power, low cost nuclear reactor which permits the use of all the fission products produced in the reactor. Medical isotopes such as Molybdenum-99 are produced in a reactor operating at a power of 100 to 500 kilowatts.

Abstract: Laser isotope separation is accomplished using at least two photoionization pathways of an isotope simultaneously, where each pathway comprises two or more transition steps. This separation method has been applied to the selective photoionization of erbium isotopes, particularly for the enrichment of .sup.167 Er. The hyperfine structure of .sup.167 Er was used to find two three-step photoionization pathways having a common upper energy level.

Abstract: The present invention provides a process for treating both cations and anions by using a self-regenerating, multi-ionic exchange resin column system which requires no separate regeneration steps. The process involves alternating ion-exchange chromatography for cations and anions in a multi-ionic exchange column packed with a mixture of cation and anion exchange resins. The multi-ionic mixed-charge resin column works as a multi-function column, capable of independently processing either cationic or anionic exchange, or simultaneously processing both cationic and anionic exchanges. The major advantage offered by the alternating multi-function ion exchange process is the self-regeneration of the resins.

Abstract: The invention relates to a method for producing actinium-225 and bismuth-213. According to the invention, radium-226 is irradiated in the thermal neutron flux of a nuclear reactor, the thorium fraction of the irradiation product is then chemically isolated and therefrom the actinium and radium mixture growing continuously by decay therein is chemically separated, this mixture serving as "cow" for the desired radionuclides which are growing continuously.

Abstract: By using AVLIS or other methods capable of providing a depleted isotopic mixture, troublesome isotopes such as Gd.sup.154, Gd.sup.156 and Er.sup.166 are selectively removed from naturally occurring isotopic mixtures, while avoiding the additional costs associated with complete fractionation of the mixture. Such mixtures can be used to provide a burnable nuclear fuel absorber having a selectively depleted isotope or isotopes. In particular, the invention concerns burnable absorbers containing erbium with a depleted 166 isotope, gadolinium with a depleted 156 isotope or with depleted 154 and 156 isotopes, and methods for making such absorbers.

Abstract: In a transuranium elements transmuting reactor core in which a reactor is charged with a plurality of fuel assemblies at a core and an amount of a transuranium element to be added is controlled so as to prevent a fuel element contained in the fuel assemblies from melting, the amount of the transuranium elements to be added to the fuel element is controlled so as to keep an excess reactivity of the reactor substantially zero through an operation of the reactor. A charging density of minor actinides is set to lessen outwards of a core central portion in a core area where a plutonium content is made even. The charging density of minor actinides is set high accordingly in an area where a plutonium is enriched high at the core of a plutonium enriched area where a plutonium content varies.

Abstract: By using AVLIS or other methods capable of providing a depleted isotopic mixture, troublesome isotopes such as Gd.sup.154, Gd.sup.156 and Er.sup.166 are selectively removed from naturally occurring isotopic mixtures, while avoiding the additional costs associated with complete fractionation of the mixture. Such mixtures can be used to provide a burnable nuclear fuel absorber having a selectively depleted isotope or isotopes. In particular, the invention concerns burnable absorbers containing erbium with a depleted 166 isotope, gadolinium with a depleted 156 isotope or with depleted 154 and 156 isotopes, and methods for making such absorbers.

Abstract: Apparatus for nuclear transmutation and power production using an intense accelerator-generated thermal neutron flux. High thermal neutron fluxes generated from the action of a high power proton accelerator on a spallation target allows the efficient burn-up of higher actinide nuclear waste by a two-step process. Additionally, rapid burn-up of fission product waste for nuclides having small thermal neutron cross sections, and the practicality of small material inventories while achieving significant throughput derive from employment of such high fluxes. Several nuclear technology problems are addressed including 1. nuclear energy production without a waste stream requiring storage on a geological timescale, 2. the burn-up of defense and commercial nuclear waste, and 3. the production of defense nuclear material. The apparatus includes an accelerator, a target for neutron production surrounded by a blanket region for transmutation, a turbine for electric power production, and a chemical processing facility.

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 188Re 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.188 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: 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: 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: 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: A method for decreasing the amount of hazardous radioactive reactor waste materials by separation from the waste of materials having long-term risk potential and exposing these materials to a thermal neutron flux. The utilization of thermal neutrons enhances the natural decay rates of the hazardous materials while the separation for recycling of the hazardous materials prevents further transmutation of stable and short-lived nuclides.

Abstract: A means for and method of producing and retaining tritium utilizing lithium bismuth and nickel is disclosed herein. The lithium bismuth serves to produce the tritium when exposed to neutrons. The nickel serves to dissolve and thereby retain the tritium which is produced by the lithium bismuth. As disclosed herein, both of these materials are contained within a common tubular housing which is pervious to the neutrons but impervious to any tritium which is produced therein.

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 combination of passive and active neutron measurements which yields quantitative information about the isotopic composition of transuranic wastes from nuclear power or weapons material manufacture reactors is described. From the measurement of prompt and delayed neutron emission and the incidence of two coincidentally emitted neutrons from induced fission of fissile material in the sample, one can quantify .sup.233 U, .sup.235 U and .sup.239 Pu isotopes in waste samples. Passive coincidence counting, including neutron multiplicity measurement and determination of the overall passive neutron flux additionally enables the separate quantitative evaluation of spontaneous fission isotopes such as .sup.240 Pu, .sup.244 Cm and .sup.252 Cf, and the spontaneous alpha particle emitter .sup.241 Am.