202results about "Specific isotope recovery" patented technology

Methods and apparatus are provided for producing and extracting Mo-99 and other radioisotopes from fission products that overcome the drawbacks of previously-known systems, especially the excessive generation of radioactive wastes, by providing gas-phase extraction of fission product radioisotopes from a nuclear fuel target using a mixture including halide and an oxygen-containing species with heat to convert the fission product radioisotopes to gas (e.g., Mo-99 to MoO₂Cl₂ gas). The gaseous species are evacuated to a recovery chamber where the radioisotopes solidify for subsequent processing, while the substantially intact uranium target made available for further irradiation and extraction cycles.

A process for producing technetium-99m from a molybdenum-100 metal powder, comprising the steps of:

• • (i) irradiating in a substantially oxygen-free environment, a hardened sintered target plate coated with a Mo-100 metal, with protons produced by a cyclotron;
  • (ii) dissolving molybdenum ions and technetium ions from the irradiated target plate with an H₂O₂ solution to form an oxide solution;
  • (iv) raising the pH of the oxide solution to about 14;
  • (v) flowing the pH-adjusted oxide solution through a resin column to immobilize K[TcO₄] ions thereon and to elute K₂[MoO₄] ions therefrom;
  • (vi) eluting the bound K[TcO₄] ions from the resin column;
  • (vii) flowing the eluted K[TcO₄] ions through an alumina column to immobilize K[TcO₄] ions thereon;
  • (viii) washing the immobilized K[TcO₄] ions with water;
  • (ix) eluting the immobilized K[TcO₄] ions with a saline solution; and
  • (x) recovering the eluted Na[TcO₄] ions.

A system for radioisotope production uses fast-neutron-caused fission of depleted or naturally occurring uranium targets in an irradiation chamber. Fast fission can be enhanced by having neutrons encountering the target undergo scattering or reflection to increase each neutron's probability of causing fission (m, f) reactions in U-238. The U-238 can be deployed as layers sandwiched between layers of neutron-reflecting material, or as rods surrounded by neutron-reflecting material.

A method for separating germanium-68 from gallium-68 using an anion exchange resin and a chelating/complexing agent containing a plurality of carboxylic acid groups and at least three carbon atoms to the first solution is disclosed together with a generator apparatus for providing a source of gallium-68.

Methods for producing Tc-99m radioisotope by proton irradiation of a fluid target matrix. A method of producing Tc-99m includes irradiating a fluid target matrix comprising Mo-100 with a proton beam to transform at least a portion of Mo-100 to Tc-99m. Optionally, the fluid target matrix further includes at least one of O-18, O-16, or N-14, which upon exposure to the proton beam concurrently transform at least a portion of O-18 to F-18, at least a portion of O-16 to N-13, at least a portion of the O-16 to O-15, or at least a portion of N-14 to C-11. The method further includes isolating Tc-99m and optionally at least one of F-18, N-13, O-15, or C-11 from the irradiated fluid target matrix. An additional source of Tc-99m is available from the decay of Mo-99 that is co-produced from the Mo-100 during irradiation with the proton beam.

An improved biomarker generator and a method suitable for efficiently producing short lived radiopharmaceuticals in quantities on the order of a unit dose. The improved biomarker generator includes a particle accelerator and a radiopharmaceutical micro-synthesis system. The micro-accelerator of the improved biomarker generator is optimized for producing radioisotopes useful in synthesizing radiopharmaceuticals in quantities on the order of one unit dose allowing for significant reductions in size, power requirements, and weight when compared to conventional radiopharmaceutical cyclotrons. The radiopharmaceutical micro-synthesis system of the improved biomarker generator is a small volume chemical synthesis system comprising a microreactor and/or a microfluidic chip and optimized for synthesizing the radiopharmaceutical in quantities on the order of one unit dose allowing for significant reductions in the quantity of radioisotope required and the processing time when compared to conventional radiopharmaceutical processing systems.

To provide a method of producing radioactive molybdenum solution suitable for extracting ^99mTc to be used as radioactive diagnostic drug by way of establishing a production process for high-density MoO₃ pellets with a lower amount of insoluble content when dissolving the pellets. The method has the steps of: preparing MoO₃ powder, fabricating a MoO₃ pellet by filling said MoO₃ powder in a heated die and sintering in an air, oxidizing said MoO₃ pellet, producing neutron-irradiated MoO₃ pellets by irradiating a neutron on said oxidized MoO₃ pellet, and obtaining radioactive molybdenum solution by dissolving said neutron-irradiated MoO₃ pellet.

The present invention is directed to a generator apparatus for separating a daughter gallium-68 radioisotope substantially free of impurities from a parent gernanium-68 radioisotope, including a first resin-containing column containing parent gernanium-68 radioisotope and daughter gallium-68 radioisotope, a source of first eluent connected to said first resin-containing column for separating daughter gallium-68 radioisotope from the first resin-containing column, said first eluent including citrate whereby the separated gallium is in the form of gallium citrate, a mixing space connected to said first resin-containing column for admixing a source of hydrochloric acid with said separated gallium citrate whereby gallium citrate is converted to gallium tetrachloride, a second resin-containing column for retention of gallium-68 tetrachloride, and, a source of second eluent connected to said second resin-containing column for eluting the daughter gallium-68 radioisotope from said second resin-containing column.

An actinium-227 production device having a plurality of metallic or ceramic caplets, each enclosing a radium-226 compound in redundantly nested sealed cylinders. The radium-226 compound is compacted into a disk and diluted with heat transporting ceramic materials. A thermal neutron shield including spectrum shaping materials to protect actinium-227 produced from exposure to thermal neutrons is included along with a strong neutron absorber to shape the neutron spectrum such that radium-226 nuclei are exposed to neutrons in the higher epithermal energy groups upon entry into the target with an energy of between 20 eV and 1 KeV.

Recycling of isotopically enriched molybdenum metal targets that are suitable for the large scale cyclotron production of 99mTc or 94mTc includes the charged particle irradiation of an enriched molybdenum metal target to produce a technetium isotope, separation of the technetium isotope following irradiation of the molybdenum, re-claiming the molybdenum metal and reformation of the molybdenum target for a further irradiation step. This process may then be repeated. Separation of the technetium isotope preferably is achieved by oxidative dissolution of the molybdenum thereby removing it from a target support plate, and forming molybdate and pertechnetate. The technetium isotope is isolated by various means, such as the ABEC process. To reuse the molybdenum, additional steps of isolating the molybdate and reducing it back to molybdenum metal are required. The recovered molybdenum metal may then be reformed as a target for example by pressing or pressing and sintering, followed by bonding to a target support plate.

Systems, methods, and processes for a high throughput, low concentration processing of low activity tritiated light water include the electrolysis of at least some of the tritiated water to produce hydrogen and tritium gas. The hydrogen and tritium gas produced by electrolysis in some cases are combined with heated water vapor to increase throughput and passed through a liquid phase catalytic exchange column, which generally includes a catalyst that includes palladium coated with a hydrophobic polymer. As the hydrogen and tritium gas, along with heated water vapor, rise through the LPCE column, the tritium is retained on the catalyst. Deionized wash water passes down the column (i.e., in the opposite direction of the flow of the hydrogen gas and heated water vapor) and carries the retained tritium out of the LPCE column. Useful in separating tritium from radioactive waste materials and from the water from nuclear reactors.

Provided are methods to separate an isotope from a first solution including uranium. The methods may include (a) cleaning the first solution to form a second solution including the uranium and a third solution including the isotope; (b) oxidizing the third solution to form an oxidized isotope; and (c) separating the oxidized isotope.

This disclosure relates to methods for purifying and isolating astatine-211 from bismuth metal. Also disclosed are automated methods for purifying and isolating astatine-211 from bismuth metal.

A flowcell 2 is constituted of insulating substrates 2a and 2b. The two substrates 2a and 2b have been directly bonded to each other by a bonding method for attaining tenacious bonding, for example, anodic bonding or hydrofluoric acid bonding. A channel 6 has been formed at the interface between the substrates 2a and 2b. Part of the substrate 2a which faces the channel 6 has a carbon electrode 4a formed thereon by sintering a pasty carbon material, the electrode 4a extending along the channel 6. On the other hand, the substrate 2b has a groove 6a serving as the channel 6, and has an electrode 4b made of a metal film formed on a bottom surface of the groove 6a.

Methods of producing and isolating ⁶⁸Ga, ⁸⁹Zr, ⁶⁴Cu, ⁶³Zn, ⁸⁶Y, ⁶¹Cu, ^99mTc, ⁴⁵Ti, ¹³N, ⁵²Mn, or ⁴⁴Sc and solution targets for use in the methods are disclosed. The methods of producing ⁶⁸Ga, ⁸⁹Zr, ⁶⁴Cu, ⁶³Zn, ⁸⁶Y, ⁶¹Cu, ⁹⁹mTc, ⁴⁵Ti, ¹³N, ⁵²Mn, or ⁴⁴Sc include irradiating a closed target system with a proton beam. The closed target system can include a solution target. The methods of producing isolated ⁶⁸Ga, ⁸⁹Zr, ⁶⁴Cu, ⁶³Zn, ⁸⁶Y, ⁶¹CU, ^99mTC, ⁴⁵-Ti, ⁵²Mn, or ⁴⁴Sc by ion exchange chromatography. An example solution target includes a target body including a target cavity for receiving the target material; a housing defining a passageway for directing a particle beam at the target cavity; a target window for covering an opening of the target cavity; and a coolant gas flow path disposed in the passageway upstream of the target window.

A radioactive fluoride anion concentrating device capable of concentrating ¹⁸F⁻ ions speedily and efficiently. A flow cell (11) is composed of a metal plate electrode (21), an insulating sheet (23) and a carbon plate electrode (25) located so that the sides of electrodes may be opposed to each other with the insulating sheet (23) inserted between them. An example of the plate metal plate electrode (21) is obtained by forming a film of metallic material on an insulation plate, and an example of the insulating sheet (23) is a PDMS from which a groove being a channel (26) having a thickness of ≦500 μm is cut out. The thickness of the sheet is desirably about 100 μm. The upper and lower sides of the flow cell (11) are fixed by fixing jigs (27) and (29).

A method of isolating ⁹⁹Mo produced using a (n,γ) reaction according to example embodiments may include vaporizing a source compound containing ⁹⁸Mo and ⁹⁹Mo. The vaporized source compound may be ionized to form ions containing ⁹⁸Mo and ⁹⁹Mo. The ions may be separated to isolate the ions containing ⁹⁹Mo. The isolated ions containing ⁹⁹Mo may be collected with a collector. Accordingly, the isolated ⁹⁹Mo may have a relatively high specific radioactivity and, in turn, may be used to produce the diagnostic radioisotope, ^99mTc, through radioactive decay.

Disclosed is a novel adsorbent for use in a ⁹⁹Mo/^99mTc generator, which is a medical diagnostic radioisotope generator, and in a ¹⁸⁸W/¹⁸⁸Re generator, which is a therapeutic radioisotope generator. The adsorbent composed of sulfated alumina or alumina-sulfated zirconia exhibits adsorption capacity superior to that of conventional adsorbents, and is stable and is thus loaded in a dry state in an adsorption column so that the radioisotope ⁹⁹Mo or ¹⁸⁸W can be adsorbed. Thus, it is possible to miniaturize the column, and such a miniaturized column is small, convenient to use, and highly efficient, and extracts a radioisotope satisfying the requirements for pharmaceuticals, and thus can be useful for radioisotope generators extracting ^99mTc or ¹⁸⁸Re.

A method for the separation of the rare-earth fission product poisons comprising providing a spent nuclear fuel. The spent nuclear fuel comprises UO₂ and rare-earth oxides, preferably Sm, Gd, Nd, Eu oxides, with other elements depending on the fuel composition. Preferably, the provided nuclear fuel is a powder, preferably formed by crushing the nuclear fuel or using one or more oxidation-reduction cycles. A compound comprising Th or Zr, preferably metal, is provided. The provided nuclear fuel is mixed with the Th or Zr, thereby creating a mixture. The mixture is then heated to a temperature sufficient to reduce the UO₂ in the nuclear fuel, preferably to at least to 850° C. for Th and up to 600° C. for Zr. Rare-earth metals are then extracted to form the heated mixture thereby producing a treated nuclear fuel. The treated nuclear fuel comprises the provided nuclear fuel having a significant reduction in rare-earths.

The invention belongs to the field of medical radioactive diagnosis nuclides, and relates to a <99>Mo production method and <99>Mo production equipment based on a bremsstrahlung and photonuclear reaction bifunctional target. In the prior art, the introduction of a conversion target can cause the low bremsstrahlung efficiency of a method for producing <99>Mo by using electron beams. A purpose of the present invention is to solve the problem in the prior art. According to the present invention, a high-energy electron beam emitted from an electron accelerator directly bombards an integrated bremsstrahlung and photonuclear reaction <100>Mo target; the production equipment comprises an electron accelerator, a target box, a target plate arranged in the target box, and a cooling system, wherein the target plate is a bremsstrahlung and photonuclear reaction bifunctional target plate, the bremsstrahlung and photonuclear reaction bifunctional target plate is a <100>Mo target, the target box hasan opening for the entering of an electron beam, and the electron beam enters the target box and directly bombards the bremsstrahlung and photonuclear reaction bifunctional target to produce <99>Mo. According to the present invention, by using the <100>Mo integrated bremsstrahlung and photonuclear reaction target, the strong self-attenuation on the bremsstrahlung twice in the traditional conversion target methods of high atomic number tungsten (W) and the like can be well overcome so as to greatly improve the production efficiency of <99>mo.

A liquid lithium-cooled fission reactor optimized for producing radioactive materials. The reactor is designed to enhance the availability of rare radioactive materials that have significant value for national defense, industrial research, and medical care. This invention has tangible design attributes that can be tailored to create one or more scarce and valuable radioactive materials. In particular, the reactor design is optimized for low-cost production of large quantities of radioactive tritium needed in national-defense and fusion-breeder programs. There are four core designs applied to this invention, all of which produce tritium and surplus heat that can generate byproduct electricity, thereby reducing the cost of radioactive-material production. Three of the embodiments furnish radioactive fission products, such as molybdenum-99, that can be extracted with high efficiency and rapid processing, thus fulfilling a critical supply and price shortfall in radioisotopes used for medical diagnosis and treatment.