| Class / Patent application number | Description | Number of patent applications / Date published |
|---|
| 075393000 | Utilizing Radioactive material, producing or treating Radioactive metal | 24 |
| 20100064853 | METHODS FOR CHEMICAL RECOVERY OF NON-CARRIER-ADDED RADIOACTIVE TIN FROM IRRADIATED INTERMETALLIC Ti-Sb TARGETS - The invention provides a method of chemical recovery of no-carrier-added radioactive tin (NCA radiotin) from intermetallide TiSb irradiated with accelerated charged particles. An irradiated sample of TiSb can be dissolved in acidic solutions. Antimony can be removed from the solution by extraction with dibutyl ether. Titanium in the form of peroxide can be separated from tin using chromatography on strong anion-exchange resin. In another embodiment NCA radiotin can be separated from iodide solution containing titanium by extraction with benzene, toluene or chloroform. NCA radiotin can be finally purified from the remaining antimony and other impurities using chromatography on silica gel. NCA tin-117 | 03-18-2010 |
| 20110083531 | SELECTIVE GOLD EXTRACTION FROM COPPER ANODE SLIME WITH AN ALCOHOL - The invention relates to a method for recovering gold from an acid digest of a gold-containing copper anode slime. The acid digest is selectively extracted with an alcohol having low miscibility with water. Gold is then recovered from the resulting alcoholic extract. | 04-14-2011 |
| 20110277592 | SEPARATION METHOD FOR CARRIER-FREE RADIOLANTHANIDES - A method for separating a lanthanide from a mixture containing at least one other lanthanide is provided. In particular, an HPLC and liquid separation method using a chromatographic column for separating a lanthanide from a mixture containing at least one other lanthanide is provided. | 11-17-2011 |
| 20120011965 | GALLIUM-68 RADIOISOTOPE GENERATOR AND GENERATING METHOD THEREOF - A Gallium-68 (Ga-68) radioisotope generator includes a generating column and a citrate eluent. The generating column is at least partially filled with an ion-exchange resin with glucamine groups to absorb germanium-68 (Ge-68) and Ga-68 radioisotopes. The citrate eluent is added into the generating column to desorb the Ga-68 radioisotope and form an eluent containing the Ga-68 radioisotope in the form of Ga-68 citrate. A method for generating Ga-68 radioisotope is also disclosed. | 01-19-2012 |
| 20120090431 | METHOD OF RECOVERING ENRICHED RADIOACTIVE TECHNETIUM AND SYSTEM THEREFOR | 04-19-2012 |
| 20120144957 | POLYMER COMPOSITE FOR EXTRACTING CESIUM FROM NUCLEAR WASTE AND/OR OTHER INORGANIC WASTES/SOLUTIONS - A polymer composite with superior granulometric properties for the extraction of active and non-active cesimn from high level acidic radioactive nuclear waste and/or other inorganic wastes/solutions that is particularly useful to nuclear industry. The void volumes of the said polymer composite is varied to obtain the desired Cs ion exchange kinetics wherein the composite material is radiation resistant and stable in highly acidic and alkaline medium. | 06-14-2012 |
| 20120152059 | METHOD FOR SELECTIVELY RECOVERING AMERICIUM FROM A NITRIC AQUEOUS PHASE - A method with which americium may be selectively recovered from a nitric aqueous phase containing americium, curium and fission products including lanthanides and yttrium, but which is free of uranium, plutonium and neptunium or which only contains these three last elements in trace amounts. The method is applicable for treatment and recycling of irradiated nuclear fuels, in particular for removing americium from raffinates stemming from methods for extracting and purifying uranium and plutonium such as the PUREX and COEX™ methods. | 06-21-2012 |
| 20120160061 | INCREASE IN THE SEPARATION FACTOR BETWEEN AMERICIUM AND CURIUM AND/OR BETWEEN LANTHANIDES IN A LIQUID-LIQUID EXTRACTION OPERATION - A method using diglycolamide for increasing the separation factor between americium and curium and/or between lanthanides during an extraction operation. The operation comprising putting an acid aqueous phase, in which are found the americium, curium and/or lanthanides, in contact with an organic phase non-miscible with water, containing at least one extractant in an organic diluent. The aqueous and organic phases are then separated, and the diglycolamide is added to the aqueous phase. This method can be used for processing and recycling irradiated nuclear fuels, in particular for selectively recovering americium from high activity aqueous solutions such as raffinates stemming from the processing of irradiated nuclear fuels with a PUREX or COEX™ method; processing of rare earth ores of the monazite, xenotime or bastnaesite type, in order to facilitate separation of <> rare earths from <> rare earths and of yttrium, or that of two rare earths with adjacent or close atomic numbers. | 06-28-2012 |
| 20120186396 | COMPOUNDS USEFUL AS LIGANDS OF ACTINIDES, THEIR SYNTHESIS AND THEIR USES - The invention relates to novel compounds which are useful as ligands of actinides, to the synthesis of these compounds and to their uses. | 07-26-2012 |
| 20120285294 | MULTIPLE GENERATOR ELUTION SYSTEM - A multiple generator elution system for selectively eluting from a plurality of parent-daughter generators according to an elution schedule it calculates taking into account supply data, demand data, and available activity in each of the generators. | 11-15-2012 |
| 20120325052 | Method and Device for Producing a Radionuclide - The present relates to a method and a device for producing a radionuclide in which an absorption column containing the radionuclide is eluted by means of an eluent in a first flow direction and subsequently in a second, opposite flow direction. | 12-27-2012 |
| 20150348662 | PROCESS AND APPARATUS FOR SEPARATION OF TECHNETIUM-99M FROM MOLYBDATE - Systems and methods for separation or isolation of technetium radioisotopes from aqueous solutions of radioactive or non-radioactive molybdate salts using a polyalkyl glycol-based cross-linked polyether polymer. Some embodiments can be used for the effective purification of radio-active technetium-99m produced from low specific activity | 12-03-2015 |
| 20160040267 | PRODUCTION OF COPPER-67 FROM AN ENRICHED ZINC-68 TARGET - An apparatus including a heating element and a sublimation vessel disposed adjacent the heating element such that the heating element heats a portion thereof. A collection vessel is removably disposed within the sublimation vessel and is open on an end thereof. A crucible is configured to sealingly position a solid mixture against the collection vessel. | 02-11-2016 |
| 075394000 | Thorium(Th) | 1 |
| 20100018347 | SEPARATION OF RADIUM AND RARE EARTH ELEMENTS FROM MONAZITE - A method of chemically extracting radium-228, rare earth metals, thorium, the decay products of thorium, and phosphates from thorium-containing ores. The method involves breaking thorium-containing ore into fragments, wetting the fragments with a concentrated strong acid to make a slurry, heating the slurry, passing the heated solution through a first anion exchange column, retaining metals and radium-228 captured on the resin, allowing the radium-228 ions to decay to actinium-228, purifying the actinium-228 fraction, sending the actinium-228 fraction through a capture column, eluting the captured thorium-228 with acid, removing radium from the solution, retaining the radium-228 fraction for isomer in-growth, retaining decay products from the radium-228, separating the REEs from the process stream; and eluting and retaining the REEs. | 01-28-2010 |
| 075396000 | Plutonium(Pu) | 2 |
| 20090320646 | METHOD FOR SELECTIVE SEPARATION OF TRIVALENT AND TETRAVALENT ACTINOIDS FROM TRIVALENT LANTHANOIDS USING HYBRID DONOR TYPE EXTRACTANT HAVING FUNCTIONAL GROUP CARRYING ACTIVE OXYGEN AND NITROGEN ATOMS - A method for separating and recovering trivalent and tetravalent actinoids in a simple and less costly manner without using an organophosphorus compound is provided. This method selectively separates and recovers the tetravalent actinoid plutonium Pu (IV) and the trivalent actinoids americium Am (III) and curium Cm (III) from trivalent lanthanoids Ln (III), etc. with the use of an extractant having a functional group with neutral multidentate ligand activity which is a hybrid donor type organic compound having both of donor atoms, i.e., an oxygen atom and a nitrogen atom. | 12-31-2009 |
| 20110265605 | METHODS OF PRODUCING AND RECOVERING PLUTONIUM-238 - Methods of producing plutonium-238 are disclosed. One method includes dissolving neptunium-237 in a nitric acid solution to produce a neptunium target solution, subjecting the neptunium target solution to neutron radiation to produce plutonium-238, and removing the plutonium-238 from the neptunium target solution. A second method includes exposing a solution of neptunium-237 to neutron radiation to produce plutonium-238, complexing the plutonium-238 with an organophosphorus compound, and separating the plutonium-238/organophosphorus compound complex from the solution of neptunium-237. A third method includes dissolving neptunium-237 to form a neptunium-237 target solution, exposing the neptunium-237 to thermal neutrons to produce plutonium-238, utilizing an organophosphorus compound to complex the plutonium-238 and the organophosphorus compound, extracting the plutonium-238/organophosphorus compound complex from the irradiated neptunium target solution, and recovering the plutonium-238. | 11-03-2011 |
| 075398000 | Uranium(U) | 8 |
| 20110303051 | RECOVERY OF RESIDUAL COPPER FROM HEAP LEACH RESIDUES - A process for recovering copper from heap leach residues, the process comprising treating heap leach residues to provide treated heap leach residues providing improved permeability of a heap of the treated heap leach residues, and leaching the heap of the treated heap leach residues with a leaching solution. Treating the heap leach residues includes: a) blending the heap leach residues with additional material to provide a blend; or b) agglomerating the heap leach residues; or c) both blending the heap leach residues with additional material and agglomerating. | 12-15-2011 |
| 20120125158 | Method for the Recovery of Uranium from Pregnant Liquor Solutions - The present invention is directed to a new more environmentally friendly method for the recovery of uranium from pregnant liquor solutions that comprise high levels of chloride by using an amino phosphonic functionalized resin. | 05-24-2012 |
| 20120174712 | Metal Abstraction Peptide (MAP) Tag and Associated Methods - Compositions comprising a tripeptide having the sequence XC | 07-12-2012 |
| 20120247276 | METHOD FOR PURIFYING THE URANIUM FROM A NATURAL URANIUM CONCENTRATE - A method with which uranium from a natural uranium concentrate may be purified, including | 10-04-2012 |
| 20120297929 | EXTRACTION PROCESS - A method for the selective recovery of uranium from a sulphate-based acidic aqueous solution of uranium containing iron and other metals by means of solvent extraction, in which the extractant used in the organic extraction solution is bis(2-ethylhexyl) phosphate and a liquid branched trialkyl phosphine oxide is the modifying agent. It is typical of the method that the uranium concentration in the feed solution is less than 50 mg/l and a reducing agent is introduced into the aqueous and/or extraction solution to prevent the permanent oxidation of iron to trivalent. In the method the majority of the extraction solution is circulated in a circuit consisting of the extraction stage and the storage tank and only a small part of the uranium-loaded extraction solution is routed to scrubbing and stripping. | 11-29-2012 |
| 20130340571 | DISSOLUTION AND RECOVERY OF AT LEAST ONE ELEMENT NB OR TA AND OF AT LEAST ONE OTHER ELEMENT U OR RARE EARTH ELEMENTS FROM ORES AND CONCENTRATES - The main subject-matter of the present invention is a process for the dissolution of at least one element chosen from niobium and tantalum and at least one other element chosen from uranium and the rare earth elements, advantageously for the dissolution of niobium, tantalum, uranium and rare earth elements, present in an ore or an ore concentrate. Said process comprises:
| 12-26-2013 |
| 20140096646 | TREATMENT METHOD OF SPENT URANIUM CATALYST - The present invention relates to a treatment method of spent uranium catalyst, and more specifically, to a method which can considerably reduce the volume of the spent uranium catalyst to be disposed of and simultaneously minimize secondary wastes that can be generated during the process of treating the spent uranium catalyst. | 04-10-2014 |
| 075399000 | Reduction | 1 |
| 20130104698 | Method of catalytic oxidation of U4+ to U6+ using a catalyst Muhamedzhan-1 | 05-02-2013 |
