Battelle Memorial Institute

Isoray Medical, Inc.

Other

Exxon Research and Engineering Company

Exxon Nuclear Company, Inc.

Exxon Nuclear, Inc.

Lane A Bray

Allyn L Boldt

Garrett N Brown

Jack L Ryan

David J Swanberg

Laura J Silva

Jay W Grate

Oleg B Egorov

Dennis W Wester

Leland L Burger

Jay Draper

Matthew J Bales

Karen Thompson, Legal Representative

Todd M Corneillie

Harold H Van Tuyl

James L Madsen

Jaquetta R DesChane

Jay C Draper

Donald C Di Lawrence

The present invention provides a method of separating and purifying Cesium-131 (Cs-131) from Barium (Ba). Uses of the Cs-131 purified by the method include cancer research and treatment, such as for the use in brachytherapy. Cs-131 is particularly useful in the treatment of faster growing tumors.

Method of separating and purifying cesium-131 from barium carbonate

The present invention provides a method of separating and purifying Yttrium-90 (Y-90) from Strontium-90 (Sr-90). In addition, a zirconium (Zr) clean-up step for the Y-90 is provided. Uses of the Y-90 purified by the method include cancer research and treatment. Y-90 is particularly useful in cell directed therapy, e.g., where the Y-90 is attached directly or indirectly to a targeting molecule such as an antibody.

Method of separating and purifying Yttrium-90 from Strontium-90

Method of separating and purifying Cesium-131 from Barium nitrate

This invention includes methods of fabricating brachytherapy implant seeds, methods of fabricating brachytherapy implant seed cores, and brachytherapy implant seeds independent of method of fabrication. In one implementation, a brachytherapy implant seed includes a sealed inorganic metallic cylinder having a radioactive core received therein. The radioactive core includes an inorganic amorphous silicate glass tube having an exterior surface extending axially along the tube. An inorganic crystalline ceramic coating is received on at least a portion of the inorganic amorphous glass tube exterior surface. The coating includes a therapeutic dose of radioactive material. A radiographic marker is received within the sealed inorganic metallic cylinder. Other aspects and implementations are contemplated.

Brachytherapy implant seeds

The present invention provides a method of preparing Cesium-131 (Cs-131) as a dispersed radioisotope. Uses of the dispersed Cs-131 prepared by the method include cancer research and treatment, such as for the use in brachytherapy. Cs-131 is particularly useful in the treatment of faster growing tumors.

Method for preparing particles of radioactive powder containing Cesium-131 for use in brachytherapy sources

The present invention is an improvement to the method of separating and purifying gadolinium from a mixture of gadolinium and europium having the steps of (a) dissolving the mixture in an acid; (b) reducing europium&plus;3 to europium&plus;2; and (c) precipitating the europium&plus;2 with a sulfate ion in a superstoichiometric amount; wherein the improvement is achieved by using one or more of the following: (i) the acid is an anoic acid; (ii) the reducing is with zinc metal in the absence of a second metal or with an amount of the second metal that is ineffective in the reducing; (iii) adding a group IIA element after step (c) for precipitating the excess sulfate prior to repeating step (c); (iv) the sulfate is a sulfate salt with a monovalent cation; (v) adding cold europium&plus;3 prior to repeating step (c).

Method of separating and purifying gadolinium-153

The present invention is a method of separating a short-lived daughter isotope from a longer lived parent isotope, with recovery of the parent isotope for further use. Using a system with a bi-directional pump and one or more valves, a solution of the parent isotope is processed to generate two separate solutions, one of which contains the daughter isotope, from which the parent has been removed with a high decontamination factor, and the other solution contains the recovered parent isotope. The process can be repeated on this solution of the parent isotope. The system with the fluid drive and one or more valves is controlled by a program on a microprocessor executing a series of steps to accomplish the operation. In one approach, the cow solution is passed through a separation medium that selectively retains the desired daughter isotope, while the parent isotope and the matrix pass through the medium. After washing this medium, the daughter is released from the separation medium using another solution. With the automated generator of the present invention, all solution handling steps necessary to perform a daughter/parent radionuclide separation, e.g. Bi-213 from Ac-225 'cow' solution, are performed in a consistent, enclosed, and remotely operated format. Operator exposure and spread of contamination are greatly minimized compared to the manual generator procedure described in U.S. patent application Ser. No. 08/789,973, now U.S. Pat. No. 5,749,042, herein incorporated by reference. Using 16 mCi of Ac-225 there was no detectable external contamination of the instrument components.

Method for sequential injection of liquid samples for radioisotope

A method of producing and purifying Cs-131 comprising the steps of dissolving irradiated Ba comprised of natural or enriched Ba including Ba-130, Ba-131, and Cs-131 from the decay of Ba-131, in an acid, precipitating the Ba, separating the Cs-131 using an ion exchange media, and eluting the Cs-131 from the exchanger to recover the purified Cs-131.

Method of separation of Cesium-131 from Barium

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. In addition to plutonium; lead, iron, cobalt, copper, uranium, and other metallic cations that form chloride anionic complexes that may be present in the mixture; are removed from the mixture on the chloride form anion exchange column.

Method of separating short half-life radionuclides from a mixture of

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.227 Ac, the at least one daughter radionuclide is a .sup.227 Th and the product radionuclide is the .sup.223 Ra and the first nitrate form ion exchange column passes the .sup.227 Ac and retains the .sup.227 Th. In another embodiment the radionuclide 'cow'is the .sup.229 Th, the at least one daughter radionuclide is a .sup.225 Ra and said product radionuclide is the .sup.225 Ac and the .sup.225 Ac and nitrate form ion exchange column retains the .sup.229 Th and passes the .sup.225 Ra/Ac.

Methods of separating short half-life radionuclides from a mixture of

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. A preferred stripping solution for purification of .sup.213 Bi for use in medical applications includes sodium acetate, pH 5.5. A protein conjugated with bifunctional chelating agents in vivo with the NaOAc, to receive the .sup.213 Bi as it is being released from the anion exchange resin.

Bismuth generator method

Disclosed is a method for removing radioactive contaminants from metal surfaces by applying steam containing an inorganic acid and cerium IV. Cerium IV is applied to contaminated metal surfaces by introducing cerium IV in solution into a steam spray directed at contaminated metal surfaces. Cerium IV solution is converted to an essentially atomized or vapor phase by the steam.

Method for decontamination of radioactive metal surfaces

A method for purifying Y-90 from a Sr-90/Y-90 'cow' wherein raw Sr-90/Y-90 source containing impurities is obtained from nuclear material reprocessing. Raw Sr-90/Y-90 source is purified to a fresh Sr-90/Y-90 source 'cow' by removing impurities by addition of sodium hydroxide and by removing Cs-137 by further addition of sodium carbonate. The 'cow' is set aside to allow ingrowth. An HDEHP organic extractant is obtained from a commercial supplier and further purified by saturation with Cu(II), precipitation with acetone, and washing with nitric acid. The 'cow' is then dissolved in nitric acid and the purified HDEHP is washed with nitric acid and scrubbed with either nitric or hydrochloric acid. The dissolved 'cow' and scrubbed HDEHP are combined in an organic extraction, separating Y-90 from Sr-90, resulting in a Sr-90/Y-90 concentration ratio of not more than 10(E-7), and a metal impurity concentration of not more than 10 ppm per curie of Y-90. The separated Y-90 may then be prepared for delivery.

Method of separation of yttrium-90 from strontium-90

A method of recovering catalyst material from latent catalyst material solids includes: a) combining latent catalyst material solids with a liquid acid anolyte solution and a redox material which is soluble in the acid anolyte solution to form a mixture; b) electrochemically oxidizing the redox material within the mixture into a dissolved oxidant, the oxidant having a potential for oxidation which is effectively higher than that of the latent catalyst material; c) reacting the oxidant with the latent catalyst material to oxidize the latent catalyst material into at least one oxidized catalyst species which is soluble within the mixture and to reduce the oxidant back into dissolved redox material; and d) recovering catalyst material from the oxidized catalyst species of the mixture. The invention is expected to be particularly useful in recovering spent catalyst material from petroleum hydroprocessing reaction waste products having adhered sulfides, carbon, hydrocarbons, and undesired metals, and as well as in other industrial applications.

Electrochemical catalyst recovery method

Plutonium, strontium, and cesium found in aqueous waste solutions resulting from nuclear fuel processing are removed by contacting the waste solutions with synthetic zeolite incorporating up to about 5 wt % titanium as sodium titanate in an ion exchange system. More than 99.9% of the plutonium, strontium, and cesium are removed from the waste solutions.

Growing community of inventors

Richland, WA, United States of America

Lane A Bray