Abstract: Applicant's invention is a radionuclide generator resin material for radiochemical separation of daughter radionuclides, particularly 213Bi, from a solution of parental radionuclides, the resin material capable of providing clinical quantities of 213Bi of at least 20-mCi, wherein the resin material comprises a silica-based structure having at least one bifunctional ligand covalently attached to the surface of the silica-based structure. The bifunctional ligand comprises a chemical group having desirable surface functionality to enable the covalent attachment of the bifunctional ligand thereon the surface of the structure and the bifunctional ligand further comprises a second chemical group capable of binding and holding the parental radionuclides on the resin material while allowing the daughter radionuclides to elute off the resin material. The bifunctional ligand has a carbon chain with a limited number of carbons to maintain radiation stability of the resin material.

Abstract: Applicant's invention is a radionuclide generator resin material for radiochemical separation of daughter radionuclides, particularly 213Bi, from a solution of parental radionuclides, the resin material capable of providing clinical quantities of 213Bi of at least 20-mCi, wherein the resin material comprises a silica-based structure having at least one bifunctional ligand covalently attached to the surface of the silica-based structure. The bifunctional ligand comprises a chemical group having desirable surface functionality to enable the covalent attachment of the bifunctional ligand thereon the surface of the structure and the bifunctional ligand further comprises a second chemical group capable of binding and holding the parental radionuclides on the resin material while allowing the daughter radionuclides to elute off the resin material. The bifunctional ligand has a carbon chain with a limited number of carbons to maintain radiation stability of the resin material.

Abstract: A method for producing 229Th includes the steps of providing 226Ra as a target material, and bombarding the target material with alpha particles, helium-3, or neutrons to form 229Th. When neutrons are used, the neutrons preferably include an epithermal neutron flux of at least 1×1013 n s?1·cm?2. 228Ra can also be bombarded with thermal and/or energetic neutrons to result in a neutron capture reaction to form 229Th. Using 230Th as a target material, 229Th can be formed using neutron, gamma ray, proton or deuteron bombardment.

Abstract: A method for producing 229Th includes the steps of providing 226Ra as a target material, and bombarding the target material with alpha particles, helium-3, or neutrons to form 229Th. When neutrons are used, the neutrons preferably include an epithermal neutron flux of at least 1×1013 n s?1·cm?2. 228Ra can also be bombarded with thermal and/or energetic neutrons to result in a neutron capture reaction to form 229Th. Using 230Th as a target material, 229Th can be formed using neutron, gamma ray, proton or deuteron bombardment.

Abstract: A method of producing and purifying promethium-147 including the steps of: irradiating a target material including neodymium-146 with neutrons to produce promethium-147 within the irradiated target material; dissolving the irradiated target material to form an acidic solution; loading the acidic solution onto a chromatographic separation apparatus containing HDEHP; and eluting the apparatus to chromatographically separate the promethium-147 from the neodymium-146.

Abstract: A method of producing and purifying promethium-147 including the steps of: irradiating a target material including neodymium-146 with neutrons to produce promethium-147 within the irradiated target material; dissolving the irradiated target material to form an acidic solution; loading the acidic solution onto a chromatographic separation apparatus containing HDEHP; and eluting the apparatus to chromatographically separate the promethium-147 from the neodymium-146.

Abstract: A method for producing 229Th uses 230Th as a target material. 229Th can be formed from 230Th using energetic particles, including neutron, gamma ray, proton or deuteron bombardment.

Abstract: A method for producing 229Th includes the steps of providing 226Ra as a target material, and bombarding the target material with alpha particles, helium-3, or neutrons to form 229Th. When neutrons are used, the neutrons preferably include an epithermal neutron flux of at least 1×1013 n s?1·cm?2. 228Ra can also be bombarded with thermal and/or energetic neutrons to result in a neutron capture reaction to form 229Th. Using 230Th as a target material, 229Th can be formed using neutron, gamma ray, proton or deuteron bombardment.

Abstract: A new composition of matter includes 195mPt characterized by a specific activity of at least 30 mCi/mg Pt, generally made by method that includes the steps of: exposing 193Ir to a flux of neutrons sufficient to convert a portion of the 193Ir to 195mPt to form an irradiated material; dissolving the irradiated material to form an intermediate solution comprising Ir and Pt; and separating the Pt from the Ir by cation exchange chromatography to produce 195mPt.

Abstract: A new composition of matter includes 195mPt characterized by a specific activity of at least 30 mCi/mg Pt, generally made by method that includes the steps of: exposing 193Ir to a flux of neutrons sufficient to convert a portion of the 193Ir to 195mPt to form an irradiated material; dissolving the irradiated material to form an intermediate solution comprising Ir and Pt; and separating the Pt from the Ir by cation exchange chromatography to produce 195mPt.

Abstract: A method of preparing high-specific-activity 195mPt includes the steps of: exposing 193Ir to a flux of neutrons sufficient to convert a portion of the 193Ir to 195mPt to form an irradiated material; dissolving the irradiated material to form an intermediate solution comprising Ir and Pt; and separating the Pt from the Ir by cation exchange chromatography to produce 195mPt.

Abstract: A method of separating lutetium from a solution containing Lu and Yb, particularly reactor-produced 177Lu and 177Yb, includes the steps of: providing a chromatographic separation apparatus containing LN resin; loading the apparatus with a solution containing Lu and Yb; and eluting the apparatus to chromatographically separate the Lu and the Yb in order to produce high-specific-activity 177Yb.

Abstract: A method of preparing high-specific-activity 195mPt includes the steps of: exposing 193Ir to a flux of neutrons sufficient to convert a portion of the 193Ir to 195mPt to form an irradiated material; dissolving the irradiated material to form an intermediate solution comprising Ir and Pt; and separating the Pt from the Ir by cation exchange chromatography to produce 195mPt.

Abstract: A .sup.99 Mo/.sup.99m Tc generator system includes a sorbent column loaded with a composition containing .sup.99 Mo. The sorbent column has an effluent end in fluid communication with an anion-exchange column for concentrating .sup.99m Tc eluted from the sorbent column.A method of preparing a concentrated solution of .sup.99m Tc includes the general steps of:a. providing a sorbent column loaded with a composition containing .sup.99 Mo, the sorbent column having an effluent end in fluid communication with an anion-exchange column;b. eluting the sorbent column with a salt solution to elute .sup.99m Tc from the sorbent and to trap and concentrate the eluted .sup.99m Tc on the ion-exchange column; andc. eluting the concentrated .sup.99m Tc from the ion-exchange column with a solution comprising a reductive complexing agent.

Abstract: A method of preparing a concentrated solution of a carrier-free radioisotope which includes the steps of:a. providing a generator column loaded with a composition containing a parent radioisotope;b. eluting the generator column with an eluent solution which includes a salt of a weak acid to elute a target daughter radioisotope from the generator column in a first eluate.c. eluting a cation-exchange column with the first eluate to exchange cations of the salt for hydrogen ions and to elute the target daughter radioisotope and a weak acid in a second eluate;d. eluting an anion-exchange column with the second eluate to trap and concentrate the target daughter radioisotope and to elute the weak acid solution therefrom; ande. eluting the concentrated target daughter radioisotope from the anion-exchange column with a saline solution.

Abstract: A generator system for providing a carrier-free radioisotope in the form of an acid comprises a chromatography column in tandem fluid connection with an ion exchange column, the chromatography column containing a charge of a radioactive parent isotope. The chromatography column, charged with a parent isotope, is eluted with an alkali metal salt solution to generate the radioisotope in the form of an intermediate solution, which is passed through the ion-exchange column to convert the radioisotope to a carrier-free acid form.

Abstract: A generator system for providing a carrier-free radioisotope in the form of an acid comprises a chromatography column in tandem fluid connection with an ion exchange column, the chromatography column containing a charge of a radioactive parent isotope. The chromatography column, charged with a parent isotope, is eluted with an alkali metal salt solution to generate the radioisotope in the form of an intermediate solution, which is passed through the ion-exchange column to convert the radioisotope to a carrier-free acid form.

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: Micro concentrations of .sup.68 Ga in secular equilibrium with .sup.68 Ge in strong aqueous HCl solution may readily be separated in ionic form from the .sup.68 Ge for biomedical use by evaporating the solution to dryness and then leaching the .sup.68 Ga from the container walls with dilute aqueous solutions of HCl or NaCl. The chloro-germanide produced during the evaporation may be quantitatively recovered to be used again as a source of .sup.68 Ga. If the solution is distilled to remove any oxidizing agents which may be present as impurities, the separation factor may easily exceed 10.sup.5. The separation is easily completed and the .sup.68 Ga made available in ionic form in 30 minutes or less.