Class / Patent application number | Description | Number of patent applications / Date published |
423002000 | Radioactive metal (At. No. 84+ or radioactive isotope of another metal) | 78 |
20100104489 | METHOD FOR THE PURIFICATION OF RADIUM FROM DIFFERENT SOURCES - The present application is directed to a method for the purification of Radium, in particular | 04-29-2010 |
20100202943 | Purification of Metals - A solid composition comprises: —MnO2; and—a compound represented by the general formula (I) wherein: R is a polymer; each Y is independently a hydrogen or a negative charge; Z is either hydrogen or is not present; each n is independently 1, 2, 3, 4, 5 or 6; wherein the MnO | 08-12-2010 |
20110142735 | METHOD FOR SEPARATING RADIOACTIVE COPPER USING CHELATING-ION EXCHANGE RESIN - Disclosed is a method for rapidly separating radioactive copper from nickel that contains radioactive copper and radioactive cobalt, which comprises dissolving nickel that contains radioactive copper and radioactive cobalt in an acid solution and leading it to pass through a chelating-ion exchange resin-filled column to thereby make nickel, radioactive copper and radioactive cobalt held by the chelating-ion exchange resin, and then leading an acid solution to pass through the chelating-ion exchange resin-filled column to elute nickel and radioactive cobalt, and thereafter leading an acid solution having a higher concentration than that of the previous acid solution to pass through the chelating-ion exchange resin-filled column after nickel and radioactive elution therefrom to thereby elute radioactive copper. | 06-16-2011 |
20110206579 | METHOD AND APPARATUS FOR THE EXTRACTION AND PROCESSING OF MOLYBDENUM-99 - A method for the extraction and purification of molybdenum, the method comprising the steps of: transferring an irradiated fuel solution to an extraction system, the irradiated fuel solution comprising iodine and molybdenum and other fission products, the extraction system comprising at least one sorbent column; passing the irradiated fuel solution upwards through the at least one sorbent-containing extraction column; directing the irradiated fuel solution towards a fuel management system by means of at least one discharge alignment valve; directing the extraction column eluate towards an iodine removal system; removing the iodine from the extraction column eluate; purifying the extraction column eluate; and collecting the purified eluate. Also disclosed is an apparatus for accomplishing the aforementioned method. | 08-25-2011 |
20110250107 | COLUMN GEOMETRY TO MAXIMIZE ELUTION EFFICIENCIES FOR MOLYBDENUM-99 - At least one system for eluting a radioactive material and a method of eluting a radioactive material is provided. The system for eluting a radioactive material may include an elution column configured to enclose an radioactive material, a first sealing member sealing a first end of the elution column, a second sealing member sealing a second end of the elution column, an elution supply source connected to the first end of the elution column via a first needle, a collection system connected to the second end of the elution column via a second needle, and a filter in the elution column, the filter being configured to support the radioactive material and prevent the radioactive material from contacting the second needle. | 10-13-2011 |
20120251415 | METHOD FOR THE PURIFICATION OF RADIUM FROM DIFFERENT SOURCES - The invention is directed to a method for the purification of Radium, in particular | 10-04-2012 |
20130039822 | PREPARATION OF CHITOSAN-BASED MICROPOROUS COMPOSITE MATERIAL AND ITS APPLICATIONS - Microporous glutaraldehyde-crosslinked chitosan sorbents, methods of making and using them, and a generator for the radioisotope | 02-14-2013 |
20130108525 | METHOD FOR SEPARATION OF CHEMICALLY PURE OS FROM METAL MIXTURES | 05-02-2013 |
20130171046 | Purification of Metals - A solid composition comprises:
| 07-04-2013 |
20130302224 | Selective Regeneration of Isotope-Specific Media Resins in Systems for Separation of Radioactive Isotopes from Liquid Waste Materials - Processes, systems, and methods for selectively regenerating an ion exchange resin generally comprises washing the ion exchange resin with an elution agent that encourages only selected contaminants, and especially selected radioactive isotopes, to disengage or decouple from the resin and enter solution in the elution agent, which thereafter is identified as the elution agent solution. The elution agent solution is then passed through a column of isotope-specific media (ISM). When the selected radioactive isotopes within the elution agent solution come into contact with the constituent media isotopes of the ISM, the selected radioactive isotopes are retained on the reactive surface areas of the ISM or within the interstitial spaces of the porous structures of the constituent media isotopes of the ISM. In some embodiments, the constituent media isotopes of the ISM are embedded, impregnated, or coated with the specific radioactive isotope that the particular ISM are adapted to separate. | 11-14-2013 |
20140234186 | Extraction Process - A process for extracting Cs-137 from i) an acidic solution obtained by dissolving an irradiated solid target comprising uranium, ii) an acidic solution comprising uranium which has previously been irradiated in a nuclear reactor, or iii) an acidic solution comprising uranium which has been used as reactor fuel in a homogeneous reactor, the acidic solution i), ii) or iii) having been treated to harvest Mo-99, wherein the process comprises contacting the treated acidic solution with an adsorbent comprising ammonium molybdophosphate (AMP). In an embodiment, the AMP is combined with an organic or inorganic polymeric support, for example AMP synthesised within hollow aluminosilicate microspheres (AMP-C). | 08-21-2014 |
20140294700 | Method of Manufacturing Non-carrier-added high-purity 177Lu Compounds as well as Non-carrier-added 177 Lu Compounds - The present invention relates to a column chromatographic method of manufacturing non-carrier-added high-purity | 10-02-2014 |
20140328736 | METHOD FOR SEPARATION OF CHEMICALLY PURE OS FROM METAL MIXTURES - A method for separating an amount of osmium from a mixture containing the osmium and at least one other additional metal is provided. In particular, method for forming and trapping OsO | 11-06-2014 |
20140369903 | Method and system for purifying charged radioisotopes - The present invention provides a simple and inexpensive method for removing metal and other impurities in a radioisotope solution. The invention further includes the development of a new parent/daughter generator system for collecting the daughter isotope in a concentrated solution. | 12-18-2014 |
20150139870 | PREPARATION OF CHITOSAN-BASED MICROPOROUS COMPOSITE MATERIAL AND ITS APPLICATIONS - Microporous glutaraldehyde-crosslinked chitosan sorbents, methods of making and using them, and a generator for the radioisotope | 05-21-2015 |
20160177421 | TECHNETIUM RECOVERY FROM HIGH ALKALINE SOLUTION | 06-23-2016 |
423003000 | Actinide series metal (At. No. 89+) | 62 |
20080219904 | PROCESS FOR ENRICHING URANIUM CONTAINING FEED MATERIAL - The invention provides a series of techniques for processing uranium containing feed materials such as uranium ores, reprocessed uranium, uranium containing residues and uranium containing spent fuel. The processes described involve fluorination of uranium containing material, separation of the uranium containing material from other materials based on ionization thereof with the non-ionized fluorine containing material being recycled. Metallic uranium and/or plutonium and/or fission products may result. The technique offers advantages in terms of the range of materials which can be reprocessed and a reduction in the number of complexity of stages which are involved in the process. | 09-11-2008 |
20100040522 | PROCESS AND APPARATUS FOR THE DRYING OF YELLOWCAKE - A process for the drying of yellowcake, the yellowcake initially being in the form of a low solids content, uranium rich feed slurry, the process including the stages of: a. dewatering the feed slurry to produce a first dewatered solids cake with a solids content higher than the feed slurry; b. re-slurrying the first dewatered solids cake with sufficient water to dissolve at least some impurities and to produce an intermediate slurry with a solids content lower than the first dewatered solids cake; c. dewatering the intermediate slurry to produce a second dewatered solids cake with a solids content higher than the feed slurry; and d. drying the second dewatered solids cake to produce dried yellowcake. | 02-18-2010 |
20110293492 | IN SITU PROCESS AND METHOD FOR GROUND WATER REMEDIATION AND RECOVERY OF MINERALS AND HYDROCARBONS - Devices, systems, and methods relating to advanced, high pressure oxidation are described. The devices, systems, and methods can be used to decontaminate ground water in a well or opening in a ground water table, and to recover minerals and hydrocarbons from subterranean deposits. | 12-01-2011 |
20120156115 | Systems and Methods for Treating Material - Systems for treating material are provided that can include a vessel defining a volume, at least one conduit coupled to the vessel and in fluid communication with the vessel, material within the vessel, and NF | 06-21-2012 |
20120301374 | PROCESS FOR THE PRODUCTION OF A URANIUM TRIOXIDE YELLOWCAKE FROM A URANIUM PEROXIDE PRECIPITATE - The present invention provides a process for the production of a uranium trioxide yellowcake from a uranium peroxide precipitate, the peroxide precipitate being in the form of a low solids content, uranium rich feed slurry, the process including the stages of: a. thickening the feed slurry to produce a thickener underflow with a solids content in the range of 15 to 50% w/w and a thickener overflow; b. dewatering the thickener underflow to produce a solids cake with a solids content of at least 50% w/w and a dewater overflow; and c. calcining the solids cake at a temperature in the range of 450° C. to 480° C. to produce a calcined uranium trioxide yellowcake. | 11-29-2012 |
20130216455 | TREATMENT OF TANTALUM- AND/OR NIOBIUM-CONTAINING COMPOUNDS - A process for treating a feedstock comprising tantalum- and/or niobium-containing compounds is provided. The process includes contacting the feedstock with a gaseous fluorinating agent, thereby to fluorinate tantalum and/or niobium present in the feedstock compounds. The resultant fluorinated tantalum and/or niobium compounds are recovered. | 08-22-2013 |
20140112846 | USE OF A KMGF3 COMPOUND FOR TRAPPING METALS IN THE FORM OF FLUORIDES AND/OR OXYFLUORIDES IN A GASEOUS OR A LIQUID PHASE - The invention relates to the use of a compound of formula KMgF | 04-24-2014 |
20150337412 | OXIDATION OF AMERICIUM IN ACIDIC SOLUTION - A process is for oxidizing americium(III) to americium(VI) includes providing a aqueous acidic composition comprising americium(III) and a mineral acid and exposing the composition to ozone and silver ion under conditions suitable for oxidation of the americium(III) to americium(VI). Nitric, acid is a suitable mineral acid for the process. Extraction of the americium from the silver is possible using organic phosphonate extractant. | 11-26-2015 |
423005000 | Fusing | 2 |
20140161691 | PROCESS AND DEVICE FOR BRINGING TWO IMMISCIBLE LIQUIDS INTO CONTACT, WITHOUT MIXING AND AT HIGH TEMPERATURE, WITH HEATING AND KNEADING BY INDUCTION - The invention relates to a process and a device for bringing two immiscible liquids into contact, without mixing and at high temperature, with heating and kneading by induction. In particular, the invention relates to a process and a device for bringing into contact metals and salts which are molten at high temperatures, for example as high as approximately 1,100 K. | 06-12-2014 |
20150340109 | PROCESS FOR SEPARATING AT LEAST ONE FIRST CHEMICAL ELEMENT E1 FROM AT LEAST ONE SECOND CHEMICAL ELEMENT E2, INVOLVING THE USE OF A MEDIUM COMPRISING A SPECIFIC MOLTEN SALT - The invention pertains to a process for separating at least one first chemical element E | 11-26-2015 |
423006000 | Ion exchanging or sorbing | 18 |
20100028226 | EXTRACTION OF URANIUM FROM WET-PROCESS PHOSPHORIC ACID - A process for the extraction of uranium compounds from wet-process phosphoric acid includes lowering the iron concentration of the wet-process phosphoric acid and reducing the valency of any remaining ferric iron in the wet-process phosphoric acid to ferrous iron, and then extracting uranium compounds from the wet-process phosphoric acid. The process can include separating a side stream from a feed stream of wet-process phosphoric acid, wherein the side stream has a greater concentration of the uranium compounds than the feed stream by filtration. Extracting uranium compounds from the wet-process phosphoric acid can be by ion exchange process or by solvent extraction. | 02-04-2010 |
20120134900 | URANIUM ION EXCHANGE ADSORPTION METHOD USING ULTRASOUND - Disclosed herein is a uranium ion exchange adsorption method using ultrasound. The method includes placing a slurry obtained by mixing uranium ions, sulfuric acid and an ion exchange resin into a reaction bath, and stirring the slurry in the reaction bath while simultaneously applying ultrasound to the reaction bath to allow the uranium ions to be adsorbed to the ion exchange resin through ion exchange adsorption. The method has an improved ion exchange adsorption rate of the uranium ions. | 05-31-2012 |
20120189513 | IONIC IMPURITIES REJECTION AND CHROMATOGRAPHIC PURIFICATION USING ION EXCHANGE - The invention covers the combination of utilizing the selectivity of an adsorbent to remove species from a liquid containing mixtures of ions and then subjecting the loaded resin to a chromatographic displacement utilizing the most selectively adsorbed species to displace the undesired co-adsorbing impurities. The technique can be used even when the most selectively adsorbed species is present as a minor constituent in the feed solution. | 07-26-2012 |
20130022519 | Extraction of Uranium from Wet-Process Phosphoric Acid - A process for the extraction of uranium compounds from wet-process phosphoric acid includes lowering the iron concentration of the wet-process phosphoric acid and reducing the valency of any remaining ferric iron in the wet-process phosphoric acid to ferrous iron, and then extracting uranium compounds from the wet-process phosphoric acid. The process can include separating a side stream from a feed stream of wet-process phosphoric acid, wherein the side stream has a greater concentration of the uranium compounds than the feed stream by filtration. Extracting uranium compounds from the wet-process phosphoric acid can be by ion exchange process or by solvent extraction. | 01-24-2013 |
20130336854 | COMPOSITIONS AND METHODS FOR TREATING NUCLEAR FUEL - Compositions are provided that include nuclear fuel. Methods for treating nuclear fuel are provided which can include exposing the fuel to a carbonate-peroxide solution. Methods can also include exposing the fuel to an ammonium solution. Methods for acquiring molybdenum from a uranium comprising material are provided. | 12-19-2013 |
20160016150 | ORGANIC-INORGANIC HYBRID MATERIAL OF USE FOR EXTRACTING URANIUM (VI) FROM AQUEOUS MEDIA CONTAINING PHOSPHORIC ACID, PROCESSES FOR PREPARING SAME AND USES THEREOF - The invention relates to an organic-inorganic hybrid material which comprises an inorganic solid support on which are grafted organic molecules of the general formula (I) hereafter: | 01-21-2016 |
423007000 | Organic synthetic resin | 12 |
20090022638 | Ion exchanger for winning metals of value - The present invention relates to the use of monodisperse, macroporous anion exchangers of type I or type II in hydrometallurgical processes for winning metals of value. | 01-22-2009 |
20100034713 | PROCESS FOR SEPARATING A CHEMICAL ELEMENT FROM URANIUM (VI) STARTING FROM A NITRIC AQUEOUS PHASE, IN AN EXTRACTION CYCLE FOR THE URANIUM - A chemical element to be very efficiently separated from uranium starting from an acid aqueous phase, in an extraction cycle for the uranium, when this chemical element is present in said phase at a concentration less than that of the uranium, or even as a trace element, and when it is moreover less extractable by the extractant used in this extraction cycle than is the uranium. The chemical element can notably be neptunium(IV) or thorium 228. | 02-11-2010 |
20110091365 | PICOLYLAMINE RESINS - The present application relates to novel gel-type or macroporous picolylamine resins which are based on at least one monovinylaromatic compound and at least one polyvinylaromatic compound and/or a (meth)acrylic compound and contain tertiary nitrogen atoms in structures of the general formula (I) | 04-21-2011 |
20120189514 | IONIC IMPURITIES REJECTION AND CHROMATOGRAPHIC PURIFICATION USING ION EXCHANGE - The invention covers the combination of utilizing the selectivity of an adsorbent to remove species from a liquid containing mixtures of ions and then subjecting the loaded resin to a chromatographic displacement utilizing the most selectively adsorbed species to displace the undesired co-adsorbing impurities. The technique can be used even when the most selectively adsorbed species is present as a minor constituent in the feed solution. | 07-26-2012 |
20130022520 | Extraction of Uranium from Wet-Process Phosphoric Acid - A process for the extraction of uranium compounds from wet-process phosphoric acid includes lowering the iron concentration of the wet-process phosphoric acid and reducing the valency of any remaining ferric iron in the wet-process phosphoric acid to ferrous iron, and then extracting uranium compounds from the wet-process phosphoric acid. The process can include separating a side stream from a feed stream of wet-process phosphoric acid, wherein the side stream has a greater concentration of the uranium compounds than the feed stream by filtration. Extracting uranium compounds from the wet-process phosphoric acid can be by ion exchange process or by solvent extraction. | 01-24-2013 |
20130039823 | Industrial Extraction of Uranium Using Ammonium Carbonate and Membrane Separation - This invention relates to the integration of ammonium carbonate leach processes with established acid and alkaline uranium leach processes as multifunctional industrial processes for the extraction, high degree purification and conversion of processed or semi-processed uranium as U3O8, UO2, or most tetra or hexa-valent forms of uranium, and where applicable, for the recovery of uranium from uranium ores, using advanced multiple stage membrane based technologies for the separation and concentration of uranium in solution from heavy metals and lighter elements that may be present in the solution, and the selective leach and precipitation properties of an ammonium carbonate leach. | 02-14-2013 |
20130164197 | EXTRACTION OF URANIUM FROM WET-PROCESS PHOSPHORIC ACID - In a preferred embodiment, a process for extracting uranium from wet-process phosphoric acid (WPA), comprises separating uranium from WPA to produce a loaded uranium solution stream and a uranium depleted WPA stream. The loaded uranium solution stream is then contacted by with an ion exchange resin. Uranium species bound to the ion exchange resin are eluted by contacting the resin with a solution comprising anions to produce a loaded uranium eluant stream. The loaded uranium eluant stream is treated to provide a uranium containing product. | 06-27-2013 |
20130343969 | Particulate Materials for Uranium Extraction and Related Processes - Extraction method for recovering metals. Phosphoric acid is contacted with an extractant suspension of solid particulate material comprising a para- or ferromagnetic material core surrounded by an outer shell of a chelating polymer whereby a metal is the solution is adsorbed on the chelating polymer, thereby removing it from the phosphoric acid solution. The metal-containing solid particulate material is magnetically separated from the solution and the metal is stripped from the solid particulate material for reuse. | 12-26-2013 |
20140044615 | METHOD AND SYSTEM FOR EXTRACTION OF URANIUM USING AN ION-EXCHANGE RESIN - The invention discloses a method for recovering uranium from an acidic leach solution or leach pulp in salt water using an amino-phosphorus resin, wherein the liquid phase of the leach solution or leach pulp contains greater than 3 g/L chloride ion in solution. The resin may comprise a functional group comprising an amino phosphonic group, an amino-phosphinic group, an amino phosphoric functional group and/or a combination thereof. According to the invention the leach solution or leach pulp may be generated by in-situ leaching, vat leaching, heap leaching and/or agitated leaching at ambient, elevated temperature and/or elevated pressure conditions in saline or hyper-saline water. | 02-13-2014 |
20140294701 | SURFACE-FUNCTIONALIZED MESOPOROUS CARBON MATERIALS - A functionalized mesoporous carbon composition comprising a mesoporous carbon scaffold having mesopores in which polyvinyl polymer grafts are covalently attached, wherein said mesopores have a size of at least 2 nm and up to 50 nm. Also described is a method for producing the functionalized mesoporous composition, wherein a reaction medium comprising a precursor mesoporous carbon, vinyl monomer, initiator, and solvent is subjected to sonication of sufficient power to result in grafting and polymerization of the vinyl monomer into mesopores of the precursor mesoporous carbon. Also described are methods for using the functionalized mesoporous carbon, particularly in extracting metal ions from metal-containing solutions. | 10-02-2014 |
20150292061 | SEPARATION OF PROTACTINUM, ACTINIUM, AND OTHER RADIONUCLIDES FROM PROTON IRRADIATED THORIUM TARGET - Protactinium, actinium, radium, radiolanthanides and other radionuclide fission products were separated and recovered from a proton-irradiated thorium target. The target was dissolved in concentrated HCl, which formed anionic complexes of protactinium but not with thorium, actinium, radium, or radiolanthanides. Protactinium was separated from soluble thorium by loading a concentrated HCl solution of the target onto a column of strongly basic anion exchanger resin and eluting with concentrated HCl. Actinium, radium and radiolanthanides elute with thorium. The protactinium that is retained on the column, along with other radionuclides, is eluted may subsequently treated to remove radionuclide impurities to afford a fraction of substantially pure protactinium. The eluate with the soluble thorium, actinium, radium and radiolanthanides may be subjected to treatment with citric acid to form anionic thorium, loaded onto a cationic exchanger resin, and eluted. Actinium, radium and radiolanthanides that are retained can be subjected to extraction chromatography to separate the actinium from the radium and from the radio lanthanides. | 10-15-2015 |
20150354027 | CONTINUOUS ION EXCHANGE PROCESS INTEGRATED WITH MEMBRANE SEPARATION FOR RECOVERING URANIUM - A continuous ion exchange system and method for recovering uranium from a pregnant liquor solution wherein the method includes the steps of: (a) treating the pregnant liquor solution ( | 12-10-2015 |
423008000 | Liquid-liquid extracting | 17 |
20090068075 | Sodium Salt Recycling Process for Use in Wet Reprocessing Process of Spent Nuclear Fuel - The present invention is directed to a process for recycling of a sodium salt by decomposition of a sodium nitride liquid waste. The process comprises: a neutralization step in which a nitric acid liquid waste or an off-gas having nitric acid dissolved therein which is produced through a wet reprocessing process comprising a dissolution step for dissolving a spent nuclear fuel in nitric acid is neutralized by adding or contacting the nitrate liquid waste or the off-gas to or with at least one sodium salt selected from sodium hydroxide, sodium hydrogencarbonate and sodium carbonate, thereby yielding a sodium nitrate liquid waste; a sodium nitrate-decomposition step in which the sodium nitrate liquid waste is reductively decomposed with a reducing agent, thereby decomposing sodium nitrate into a nitrogen gas and the sodium salt; and a recycle step for recycling the sodium salt into the neutralization step or wet reprocessing process. The sodium nitrate liquid waste obtained in the neutralization step may be concentrated by evaporation to give a concentrated sodium nitrate liquid waste, and the condensed sodium nitrate liquid waste may be reductively decomposed in the sodium nitrate-decomposition step. | 03-12-2009 |
20090074639 | CONTROLLED COPPER LEACH RECOVERY CIRCUIT - The present invention relates generally to a process for controlled leaching and sequential recovery of two or more metals from metal-bearing materials. In one exemplary embodiment, recovery of metals from a leached metal-bearing material is controlled and improved by providing a high grade pregnant leach solution (“HGPLS”) and a low grade pregnant leach solution (“LGPLS”) to a single solution extraction plant comprising at least two solution extractor units, at least two stripping units, and, optionally, at least one wash stage. | 03-19-2009 |
20120128555 | METHOD FOR TREATING SPENT NUCLEAR FUEL - A method for treating spent nuclear fuel, which includes first decontaminating the uranium, plutonium and neptunium found in a nitric aqueous phase resulting from dissolving the nuclear fuel in HNO | 05-24-2012 |
20140030172 | 2,9-DIPYRIDYL-1,10-PHENANTHROLINE DERIVATIVES USEFUL AS ACTINIDE LIGANDS, METHOD FOR SYNTHESIZING SAME, AND USES THEREOF - The invention relates to novel compounds useful as ligands of actinides and which meet general formula (I) hereinafter: | 01-30-2014 |
20140072485 | METHODS FOR SEPARATING MEDICAL ISOTOPES USING IONIC LIQUIDS - A method for extracting a radioisotope from an aqueous solution, the method comprising: a) intimately mixing a non-chelating ionic liquid with the aqueous solution to transfer at least a portion of said radioisotope to said non-chelating ionic liquid; and b) separating the non-chelating ionic liquid from the aqueous solution. In preferred embodiments, the method achieves an extraction efficiency of at least 80%, or a separation factor of at least 1×10 | 03-13-2014 |
20140127095 | PROCESSES FOR RECOVERING METALS FROM AQUEOUS SOLUTIONS - Provided herein are processes for recovering molybdenum and/or other value metals (e.g., uranium) present in aqueous solutions from a large range of concentrations: from ppm to grams per liter via a solvent extraction process by extracting the molybdenum and/or other value metal from the aqueous solution by contacting it with an organic phase solution containing a phosphinic acid, stripping the molybdenum and/or other value metal from the organic phase solution by contacting it with an aqueous phase strip solution containing an inorganic compound and having a ≦1.0 M concentration of free ammonia, and recovering the molybdenum and/or other value metal by separating it from the aqueous phase strip solution. When the molybdenum and/or other value metal are present only in low concentration, the processes can include an organic phase recycle step and/or an aqueous phase strip recycle step in order to concentrate the metal prior to recover. | 05-08-2014 |
423009000 | Organo-nitrogen solvent | 6 |
20100124522 | Process for the extraction of technetium from uranium - A spent fuel reprocessing method contacts an aqueous solution containing Technetium(V) and uranyl with an acidic solution comprising hydroxylamine hydrochloride or acetohydroxamic acid to reduce Tc(V) to Tc(II, and then extracts the uranyl with an organic phase, leaving technetium(II) in aqueous solution. | 05-20-2010 |
20110002823 | Pooled Separation of Actinides froma Highly Acidic Aqueous Phase Using a Solvating Extractant in a Salting-out Medium - The invention relates to a process which makes it possible to separate together all the actinide(III), (IV), (V) and (VI) entities present in a highly acidic aqueous phase from fission products, in particular lanthanides, also present in this phase by using a solvating extractant in a salting-out medium. | 01-06-2011 |
20120219474 | Metal Solvent Extraction Reagents And Use Thereof - Reagent compositions, methods for their manufacture and methods of their use are described. In particular, provided are reagent compositions comprising an aldoxime and ketoxime with an alkyl substituent. Also provided are methods of metal recovery using these reagent compositions. | 08-30-2012 |
20120219475 | Compositions and Methods of Using a Ketoxime in a Metal Solvent Extraction Reagent - Provided are methods using a ketoxime in metal extraction. One aspect of the invention relates to a method for the recovery of metal from a metal-containing aqueous solution at an elevated temperature using a ketoxime. Another aspect relates to a method of separating iron/copper using a specific ketoxime. Aldoximes may also be added to the reagent compositions used in these methods. | 08-30-2012 |
20120219476 | Methods Of Metal Extraction Using Oximes - Provided are methods using ketoximes and/or aldoximes, including 3-methyl-5-alkylsalicylaldoxime and/or 3-methyl-5-alkyl-2-hydroxyacetophenone oxime, in reagent compositions for metal extraction/isolation. One such method is of extracting a metal from a nitrate-containing aqueous solution. Another such method is of extracting a metal from an aqueous ammoniacal solution. A third method is of multi-metal extraction based on a predetermined pH. | 08-30-2012 |
20130259776 | PROCESS FOR SEPARATING AMERICUM FROM OTHER METALLIC ELEMENTS PRESENT IN AN ACIDIC AQUEOUS OR ORGANIC PHASE AND APPLICATIONS THEREOF - A process which allows separation of americium present in an acid aqueous phase or in an organic phase from the other metal elements also found in this phase, by complexation of the americium with a water-soluble ethylenediamine derivative; and a process for selective recovery of americium from an acid aqueous phase containing, in addition to americium, other metal elements, which comprises the application of this separation process. | 10-03-2013 |
423010000 | Organo-phosphorus solvent | 5 |
20100150798 | ACTINIDE EXTRACTION METHODS AND ACTINIDE SEPARATION COMPOSITIONS - Methods of separating actinides from lanthanides are disclosed. A regio-specific/stereo-specific dithiophosphinic acid having organic moieties is provided in an organic solvent that is then contacted with an acidic medium containing an actinide and a lanthanide. The method can extend to separating actinides from one another. Actinides are extracted as a complex with the dithiophosphinic acid. Separation compositions include an aqueous phase, an organic phase, dithiophosphinic acid, and at least one actinide. The compositions may include additional actinides and/or lanthanides. A method of producing a dithiophosphinic acid comprising at least two organic moieties selected from aromatics and alkyls, each moiety having at least one functional group is also disclosed. A source of sulfur is reacted with a halophosphine. An ammonium salt of the dithiophosphinic acid product is precipitated out of the reaction mixture. The precipitated salt is dissolved in ether. The ether is removed to yield the dithiophosphinic acid. | 06-17-2010 |
20100310438 | Use of Butyraldehyde Oxime as an Anti-Nitrous Agent in an Operation for the Reductive Stripping of Plutonium - The invention relates to the use of butyraldehyde oxime as an anti-nitrous agent in a plutonium stripping operation based on a reduction of this element from oxidation state (IV) to oxidation state (III). | 12-09-2010 |
20130202501 | PROCESS FOR REPROCESSING SPENT NUCLEAR FUEL NOT REQUIRING A PLUTONIUM-REDUCING STRIPPING OPERATION - The invention relates to a process for reprocessing spent nuclear fuel which, among other advantages, does not require a plutonium-reducing stripping operation. | 08-08-2013 |
20140170039 | PROCESS OF SEPARATING AND PURIFYING THORIUM - The present invention relates to a process of separating and purifying thorium by solvent extraction, comprising: 1) mixing a thorium enrichment with an inorganic acid to produce a feed; 2) mixing a neutral phosphorus extractant with an organic solvent to obtain an organic phase; 3) extracting from the feed with the organic phase to obtain a loaded organic phase; 4) scrubbing the loaded organic phase with a scrubbing solution and then back-extracting thorium with a stripping solution to obtain a thorium solution; 5) mixing the thorium solution with an oxalate to obtain a precipitate, which is then sintered to obtain thorium oxide. The present process allows to increase the purity of thorium from 80%-99% to 99.99% or more with a yield of more than 98%. | 06-19-2014 |
20150010446 | EXTRACTION OF URANIUM FROM WET-PROCESS PHOSPHORIC ACID - A process for extracting uranium compounds from wet-process phosphoric acid (WPA) includes lowering iron content of WPA to produce a lowered iron WPA, reducing valency of any remaining iron in the lowered iron WPA to produce a reduced iron valency WPA. Uranium compounds are extracted from the reduced iron valency WPA via a solvent extraction process. | 01-08-2015 |
423011000 | Forming insoluble substance in liquid | 11 |
20100316543 | Method for separating and recycling uranium and fluorine form solution - A separation and recycling method for recycling uranium and fluoride from a waste liquid sequentially and separately is disclosed. The method comprises a uranium-recycling process and a fluoride-recycling process. In the uranium-recycling process, an alkali metal compound or monovalent cation and a coagulant aid are added into the waste liquid to promote the precipitation of uranium. In the fluoride-recycling process, an alkaline earth metal compound, a strong acid and a coagulant aid are added into the uranium-removed waste liquid to precipitate fluoride. By means of the method of the present invention, the uranium and fluoride contents of the uranium-removed and fluoride-removed waste liquid are compliant with the effluent standards of the environmental laws. | 12-16-2010 |
20110212005 | METHOD FOR PREPARING URANIUM CONCENTRATES BY FLUIDIZED BED PRECIPITATION, AND PREPARATION OF UO3 AND U3O8 BY DRYING/CALCINING SAID CONCENTRATES - Method for producing a uranium concentrate in the form of solid particles, by precipitation from a uranium-containing solution using a precipitating agent, in a vertical reactor comprising a base, a top, a central part, an upper part, and a lower part, the solid particles of the uranium concentrate forming a fluidized bed under the action of a rising liquid current which circulates from the base towards the top of the reactor successively passing through the lower part, the central part and the upper part of the reactor, and which is created by introducing a liquid recycling current (flow) at the base of the reactor, said liquid recycling current being tapped at a first determined level (A) in the upper part of the reactor and sent back without settling to the base of the reactor, excess liquid being also evacuated via an overflow located at a second determined level (B) in the upper part of the reactor; a method in which the upper limit (C) of the fluidized bed of solid particles is controlled so that it is positioned at a level below the first and second determined levels. | 09-01-2011 |
20140044616 | METHOD FOR PRECIPITATING ONE OR MORE SOLUTES - The invention deals with a method for precipitating at least one solute in a reactor comprising: | 02-13-2014 |
20150361527 | SYSTEM AND METHOD FOR PARALLEL SOLUTION EXTRACTION OF ONE OR MORE METAL VALUES FROM METAL-BEARING MATERIALS - The present disclosure relates to a process and system for recovery of one or more metal values using solution extraction techniques and to a system for metal value recovery. In an exemplary embodiment, the solution extraction system comprises a first solution extraction circuit and a second solution extraction circuit. A first metal-bearing solution is provided to the first and second circuit, and a second metal-bearing solution is provided to the first circuit. The first circuit produces a first rich electrolyte solution, which can be forwarded to primary metal value recovery, and a low-grade raffinate, which is forwarded to secondary metal value recovery. The second circuit produces a second rich electrolyte solution, which is also forwarded to primary metal value recovery. The first and second solution extraction circuits have independent organic phases and each circuit can operate independently of the other circuit. | 12-17-2015 |
423012000 | By coprecipitating with carrier | 1 |
20190144968 | METHOD FOR REMOVING RADIOACTIVE ELEMENT THORIUM IN RARE EARTH MINERAL | 05-16-2019 |
423015000 | Forming compound containing plural metals or metal and ammonium | 2 |
20140140905 | SEPARATION AND RECOVERY OF MOLYBDENUM VALUES FROM URANIUM PROCESS DISTILLATE - A method for treating process distillate heavies produced during uranium fluoride purification is described. The heavies contain primarily uranium hexafluoride, UF | 05-22-2014 |
20160002751 | SELECTIVE EXTRACTION OF CERIUM FROM OTHER METALS - Methods for the extraction of cerium and/or thorium from metal compounds and solutions. A single step or two-step extraction method may be applied to selectively precipitate thorium and/or cerium as hydroxides under controlled pH conditions such that a substantially thorium-free and/or cerium-free rare earth element (REE) solution may be formed, such as for the subsequent separation of individual rare earth elements. | 01-07-2016 |
423016000 | Forming peroxide (e.g., U04, etc.) | 2 |
20090269261 | Process for Recovering Isolated Uranium From Spent Nuclear Fuel Using a Highly Alkaline Carbonate Solution - Disclosed is a process for recovery of uranium from a spent nuclear fuel using a carbonate solution, characterized by excellent proliferation resistance of preventing leaching of transuranium element (TRU) nuclides such as Pu, Np, Am, Cm, etc. from the spent nuclear fuel as well as environmental friendliness of minimizing waste generation, wherein a highly alkaline carbonate solution is used to separate uranium alone from the spent nuclear fuel. | 10-29-2009 |
20160122199 | Uranium Recovery From UF6 Cylinders - A process for recovering residual uranium from emptied uranium hexafluoride shipping cylinder during cleaning, including rinsing a uranium hexafluoride shipping cylinder with hydrofluoric acid to dissolve a heel of uranium hexafluoride therein to form a mixture of sediment, precipitates and a uranium solution; separating the uranium solution from the sediment and precipitates; mixing sodium hydroxide with the uranium solution to precipitate sodium diuranate; separating the solid sodium diuranate from the sodium fluoride solution formed; re-dissolving the sodium diuranate in sodium carbonate solution to form uranyl carbonate complex solution; and adjusting the pH of uranyl carbonate complex solution further to precipitate uranyl peroxide with the addition of hydrogen peroxide. Sodium fluoride solution produced is further treated to remove fluoride by percolating it through a calcite limestone bed to form calcium fluoride solid. | 05-05-2016 |
423018000 | Acid leaching | 2 |
20110250108 | ADVANCED DRY HEAD-END REPROCESSING OF LIGHT WATER REACTOR SPENT NUCLEAR FUEL - A method for reprocessing spent nuclear fuel from a light water reactor includes the step of reacting spent nuclear fuel in a voloxidation vessel with an oxidizing gas having nitrogen dioxide and oxygen for a period sufficient to generate a solid oxidation product of the spent nuclear fuel. The reacting step includes the step of reacting, in a first zone of the voloxidation vessel, spent nuclear fuel with the oxidizing gas at a temperature ranging from 200-450° C. to form an oxidized reaction product, and regenerating nitrogen dioxide, in a second zone of the voloxidation vessel, by reacting oxidizing gas comprising nitrogen monoxide and oxygen at a temperature ranging from 0-80° C. The first zone and the second zone can be separate. A voloxidation system is also disclosed. | 10-13-2011 |
20140341790 | EXTRACTION OF METALS FROM METALLIC COMPOUNDS - Methods for the extraction of metals such as rare earth metals and thorium from metal compounds and solutions. The methods may include the selective precipitation of rare earth elements from pregnant liquor solutions as rare earth oxalates. The rare earth oxalates are converted to rare earth carbonates in a metathesis reaction before being digested in an acid and treated for the extraction of thorium. A two-step extraction method may be applied to precipitate thorium as thorium hydroxide under controlled pH conditions such that pure thorium precipitate is recovered from a first step and a thorium-free rare earth solution is recovered at the subsequent step. The resulting rare earth solutions are of extremely high purity and may be processed directly in a solvent extraction circuit for the separation of rare earth elements, or may be processed for the direct production of a 99.9% bulk rare earth hydroxide/oxide concentrate. | 11-20-2014 |
423019000 | Volatizing | 2 |
20130336855 | ADVANCED DRY HEAD-END REPROCESSING OF LIGHT WATER REACTOR SPENT NUCLEAR FUEL - A method for reprocessing spent nuclear fuel from a light water reactor includes the step of reacting spent nuclear fuel in a voloxidation vessel with an oxidizing gas having nitrogen dioxide and oxygen for a period sufficient to generate a solid oxidation product of the spent nuclear fuel. The reacting step includes the step of reacting, in a first zone of the voloxidation vessel, spent nuclear fuel with the oxidizing gas at a temperature ranging from 200-450° C. to form an oxidized reaction product, and regenerating nitrogen dioxide, in a second zone of the voloxidation vessel, by reacting oxidizing gas comprising nitrogen monoxide and oxygen at a temperature ranging from 0-80° C. The first zone and the second zone can be separate. A voloxidation system is also disclosed. | 12-19-2013 |
20140037518 | METHOD OF RECYCLING SPENT NUCLEAR FUEL - The method concerns processing irradiated (spent) nuclear fuel (SNF), it is primarily aimed at isolating and trapping tritium, and can be used in nuclear power industry for treating SNF. | 02-06-2014 |
423020000 | Acid leaching | 4 |
20090202405 | Chemical Beneficiation of Raw Material Containing Tantalum-Niobium - The invention relates to a process for the chemical beneficiation of raw material containing tantalum-niobium such as wastes, scoria, concentrates and ores. | 08-13-2009 |
20110182786 | HYDROMETALLURGICAL PROCESS AND METHOD FOR RECOVERING METALS - A mineral processing facility is provided that includes a cogen plant to provide electrical energy and waste heat to the facility and an electrochemical acid generation plant to generate, from a salt, a mineral acid for use in recovering valuable metals. | 07-28-2011 |
20120134901 | HIGHLY EFFICIENT URANIUM LEACHING METHOD USING ULTRASOUND - A highly efficient uranium leaching method using ultrasound is disclosed. The uranium leaching method includes preparing black slate powder containing uranium by pulverizing black slate containing uranium, placing the black slate powder and water in a reaction bath, and performing uranium leaching by adding and mixing sulfuric acid and an oxidant with the black slate powder and water to prepare a mixture in the reaction bath while applying ultrasound to the reaction bath. In this method, uranium leaching efficiency can be maximized by adding sulfuric acid to the uranium ore while applying ultrasound thereto. | 05-31-2012 |
20120328492 | USE OF CERTAIN CHEMICAL ELEMENTS FOR INHIBITING THE FORMATION OF PRECIPITATES CONTAINING ZIRCONIUM MOLYBDATE IN AN AQUEOUS SOLUTION CONTAINING THE ELEMENT MOLYBDENUM AND THE ELEMENT ZIRCONIUM - A method inhibits the formation of zirconium molybdate precipitate in an aqueous solution containing the element molybdenum and the element zirconium by adding a chemical element selected from plutonium, tellurium, antimony and mixtures thereof with the aqueous solution. The method can be used for reprocessing used fuels with the element molybdenum and the element zirconium. | 12-27-2012 |