Abstract: 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.

Abstract: 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.

Abstract: 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.

Abstract: 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.

Abstract: 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.

Abstract: 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.

Abstract: 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+3 to europium+2; and (c) precipitating the europium+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+3 prior to repeating step (c).

Abstract: 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.

Abstract: 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.

Abstract: 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.

Abstract: 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.

Abstract: 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.

Abstract: 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.

Abstract: 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.

Abstract: 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.

Abstract: A process for dissolving PuO.sub.2, NpO.sub.2, or fuel containing PuO.sub.2 and/or NpO.sub.2 in acid, particularly nitric acid by electrolytically continuously maintaining an oxidizing regenerable reagent. The regenerable reagent may be added to the mixture of oxide(s) and nitric acid. Alternately, it may already be present as a fission product or other nuclear reaction product in irradiated nuclear reactor fuel, in which case it need not be added to the nitric acid.

Abstract: A method for removing volatile fission products from irradiated fuel which comprises passing a hydrogen-containing inert gas by said fuel which is at an elevated temperature of at least about 1000.degree. C. for a sufficient time to remove a significant portion of said fission products and a second step wherein inert gas alone is swept past the fuel while at said elevated temperature.

Abstract: The process for oxidizing irradiated fuel pellets using as the oxidant NO.sub.2 or a mixture of NO.sub.2 with its dissociation products, O.sub.2 and NO, in order to release the Volatile Fission Products (iodine, zenon, krypton and tritium) from the pellets.