The present disclosure relates to systems and methods for producing tailored solutions or medicaments containing radioactive isotopes (e.g., alpha particle emitting radioactive isotopes). The solutions may be produced by appropriate aging and separation steps. Therapeutically effective amounts of Ac-225 and/or Bi-213 may thus be obtained.

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.

The present disclosure relates processes and systems for producing and/or purifying .sup.68Ga from an irradiated substrate of .sup.68Zn. In some embodiments, the process rely on the use two cation-exchange chromatography columns to separate .sup.68Ga from .sup.68Zn and other radionuclides and metallic impurities. The process achieves a high overall yield of .sup.68Ga and a high effective molar activity while being implementable in a time compatible with the short half-life of .sup.68Ga. In additional embodiments, the process is implemented by an automated system.

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.

The present invention provides a method for the generation of .sup.223Ra of pharmaceutically tolerable purity comprising i) preparing a generator mixture comprising .sup.227Ac, .sup.227Th and .sup.223Ra; ii) loading said generator mixture onto a strong base anion exchange resin; iii) eluting said .sup.223Ra from said strong base anion exchange resin using a first mineral acid in an alcoholic aqueous solution to give a first eluted .sup.223Ra solution; iv) loading the .sup.223Ra of the first eluted .sup.223Ra solution onto a strong acid cation exchange resin; and v) eluting the .sup.223Ra from said strong acid cation exchange resin using a second mineral acid in aqueous solution to provide a second eluted solution. The invention additionally provides products of corresponding purity and/or products obtained or obtainable by such a method.

A means for installing material, through a fuel assembly instrument thimble insert, into the existing instrument thimbles in nuclear fuel assemblies for the purpose of allowing the material to be converted to commercially valuable quantities of desired radioisotopes during reactor power operations during a remainder of a fuel cycle and removing the radioisotopes from the core through the reactor flange opening once the fuel assemblies have been removed for refueling. The invention also describes methods that can be used to harvest the irradiated material so it can be packaged for transportation from the reactor to a location where the desired radioisotope(s) can be extracted from the fuel assembly instrument thimble insert.

Methods for purifying .sup.89Zr are provided, .sup.89Zr compositions are provided, isotope compositions are provided that can include: a radio isotope and a nanoparticle, and methods for radio labeling monoclonal antibodies are provided.

Gallium-68 generators that are capable of producing gallium-68 from a germanium-68 source material are disclosed. The source material may be a matrix material (e.g., zeolite) in which germanium-68 is isomorphously substituted for central atoms in tetrahedra within the matrix material. Methods for forming gallium-68 generators are also disclosed.

An elution generator including an elution column having a container defining an interior volume and a septum, a radiation shield having an upper shield portion defining a central recess and a coaxial flow needle extending downwardly into the central recess, and a lower shield portion having body portion defining a central recess, wherein the elution column is disposed in the central recess of the lower shield portion, the body portion of the lower shield portion is disposed in the central recess of the upper shield portion, and the coaxial flow needle extends downwardly through the septum into the internal volume of the elution column.