A system for manufacturing radionuclide generators includes an enclosure defining a radioactive environment. The enclosure includes radiation shielding to prevent radiation within the radioactive environment from moving to an exterior of the enclosure. The system also includes a conveyance system having a forward track and first carriages positioned on and movable along the forward track for conveying racks in a first direction. The conveyance system also includes a first walking beam mechanism magnetically coupled to the first carriages to move the first carriages. The conveyance system further includes a return track and second carriages positioned on and movable along the return track for conveying racks in a second direction opposite the first direction. The forward track and the return track form a loop.

An initial introduction control part introduces an aqueous solution containing molybdenum 99 and technetium 99m, and an organic solvent being capable of dissolving the technetium 99m into an extraction tank. A micro-mixing control part micro-mixes the aqueous solution and the organic solvent by heating and stirring a mixed solution of the aqueous solution and the organic solvent introduced into the extraction tank with a heater, while applying ultrasonic to the mixed solution. A separation control part separates the mixed solution micro-mixed into two phases of aqueous solution and an organic solvent. A taking-out introduction control part passes the organic solvent separated into two phases through an adsorption column be capable of adsorbing molybdenum 99 and introduces the organic solvent into an evaporation elution tank. An evaporation control part evaporates the organic solvent and leaves residue by reducing pressure inside the evaporation elution tank and heating the organic solvent introduced into the evaporation elution tank with a heater, while applying ultrasonic to the organic solvent. An elution control part introduces physiological saline solution into the residue and elutes technetium 99m into the physiological saline solution from the residue.

A nuclear fuel, the nuclear fuel comprising uranium aluminide grains, wherein the uranium aluminide grain properties are selected for good isotope extraction after irradiation and chemical digestion.

The invention provides an automated method for isolating a targeted isotope, the method having the steps of supplying a dissolved uranium targets into a first reaction environment; precipitating non-targeted isotope within the first reaction environment transferring liquid phase targeted isotope to a second reaction environment; precipitating the liquid phase targeted isotope in the second reaction environment; dissolving the precipitated targeted isotope; transferring the dissolved targeted isotope to a third reaction environment; and precipitating non-targeted isotope (i.e., iodine), such that the targeted isotope remains in the solution. Also provided is an automated system for isolating isotopes, the system having a plurality of reaction environments adapted to pneumatically receive and disgorge reactants and products via remotely actuated valves positioned between each of the reaction environments.

An initial introduction control part introduces an aqueous solution containing molybdenum 99 and technetium 99m, and an organic solvent being capable of dissolving the technetium 99m into an extraction tank. A micro-mixing control part micro-mixes the aqueous solution and the organic solvent by heating and stirring a mixed solution of the aqueous solution and the organic solvent introduced into the extraction tank with a heater, while applying ultrasonic to the mixed solution. A separation control part separates the mixed solution micro-mixed into two phases of aqueous solution and an organic solvent. A taking-out introduction control part passes the organic solvent separated into two phases through an adsorption column be capable of adsorbing molybdenum 99 and introduces the organic solvent into an evaporation elution tank. An evaporation control part evaporates the organic solvent and leaves residue by reducing pressure inside the evaporation elution tank and heating the organic solvent introduced into the evaporation elution tank with a heater, while applying ultrasonic to the organic solvent. An elution control part introduces physiological saline solution into the residue and elutes technetium 99m into the physiological saline solution from the residue.

An object of the invention is to efficiently produce a radionuclide. While a fluid containing a raw material is circulated along a circulation passage, a first radionuclide is generated in the fluid from the raw material by irradiating the fluid with radiation rays midway along the circulation passage. Further, while the fluid is circulated along the circulation passage, a substance containing at least a part of the first radionuclide and a second radionuclide generated from the first radionuclide is taken out from the fluid, and the fluid containing the remaining raw material is returned to the circulation passage again for circulation.

Described are methods for preparing radionuclides, such as radionuclides having a high specific activity. The disclosed methods include irradiating target nuclide materials, in solution, with a neutron source. The radionuclides can be separated from the target nuclide material by providing a solid carbon nanostructured material, as a suspension of solids, proximal to the target nuclide material in solution and using the recoil to drive adsorption of the radionuclide onto the solid carbon nanostructured material to transfer the radionuclides from the liquid phase (in solution) to the solid phase (adsorbed to the suspended solid carbon nanostructured material). One or more surfactants can be incorporated into the solution to facilitate formation of a stable suspension of the solid carbon nanostructured material.

An irradiation target for the production of radioisotopes, comprising at least one plate defining a central opening and an elongated central member passing through the central opening of the at least one plate so that the at least one plate is retained thereon, wherein the at least one plate and the elongated central member are both formed of materials that produce molybdenum-99 (Mo-99) by way of neutron capture.

A fuel bundle surrogate for the irradiation of a target material, having a plurality of tube sheaths, each tube sheath being parallel to a longitudinal center axis of the fuel bundle surrogate, a plurality of end caps, a pair of end plates, wherein the end plates are disposed at opposing ends of the plurality of tube sheaths, and a first target comprised of a first target material suitable for producing the isotope by way of a neutron capture event, wherein the first target is disposed in a first tube sheath, and wherein the first tube sheath of the plurality of tube sheaths comprises an elongated thickened wall portion and a pair of annular end portions, each annular end portion being disposed on a corresponding end of the thickened wall portion and having a wall thickness that is less than a wall thickness of the thickened wall portion.

Methods for purifying the molybdenum-99 isotope are disclosed. Molybdenum-99 is loaded onto an anion exchange column and extracted. In some embodiments, the extraction solution may include nitric acid and nitrate salts. In other embodiments, a two stage elution is performed in which a nitric acid containing eluent and a hydroxide containing eluent are used in succession to extract molybdenum-99.