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This page is updated frequently with new Uranium-related patent applications. Subscribe to the Uranium RSS feed to automatically get the update: related Uranium RSS feeds. RSS updates for this page: Uranium RSS RSS


Uranium dioxide nuclear fuel pellet having metallic microcells and fabricating method thereof

Uranium dioxide nuclear fuel pellet having ceramic microcells and fabricating method thereof

Date/App# patent app List of recent Uranium-related patents
07/17/14
20140197557
 Method for preparing a porous nuclear fuel patent thumbnailnew patent Method for preparing a porous nuclear fuel
A method for producing a porous fuel including uranium, optionally plutonium, and optionally at least one minor actinide, the method including: a) compacting a mixture including a first type of agglomerate including uranium oxide in a form of uranium dioxide uo2, optionally plutonium oxide, and optionally at least one minor actinide oxide, and a second type of agglomerate including uranium oxide in a form of triuranium octaoxide u3o8, optionally plutonium oxide, and optionally at least one minor actinide oxide; b) reducing the compacted mixture in a reducing medium, to reduce all or part of the triuranium octaoxide u3o8 into uranium dioxide uo2, the second type of agglomerate being prepared prior to the compacting by a series of specific operations.. .
07/03/14
20140185731
 Uranium dioxide nuclear fuel pellet having metallic microcells and fabricating method thereof patent thumbnailUranium dioxide nuclear fuel pellet having metallic microcells and fabricating method thereof
A uranium dioxide nuclear fuel pellet includes metallic microcells having a high protection capacity for fission products and a high thermal conductivity simultaneously arranged in the nuclear fuel pellet to trap fission products, such that extraction of fission products may be restrained in a normal operation condition and that the temperature of a nuclear fuel may be lowered to enhance the performance of the nuclear fuel, only to restrain extraction of radioactive fission products toward the environment in an accident condition to enhance a stability of the nuclear fuel pellet, and a fabricating method thereof.. .
07/03/14
20140185730
 Uranium dioxide nuclear fuel pellet having ceramic microcells and fabricating method thereof patent thumbnailUranium dioxide nuclear fuel pellet having ceramic microcells and fabricating method thereof
A uranium dioxide nuclear fuel pellet has about 50 to about 400 μm (with respect to a 3-dimentional size) microcells formed of a ceramic material having a chemical attraction with fission products generated in the nuclear fuel pellet to absorb and trap the fission products, such that the extraction of the fission product may be retrained in a normal operation condition and that the performance of the nuclear fuel may be enhanced by mitigating pci. In addition, highly radioactive fission products including cs and i having a large generation amount or a long half-life enough to affect the environments can be trapped in the pellet in an accident condition, without being released outside..
06/19/14
20140171724
 Method for chemically stabilizing uranium carbide compounds, and device implementing the method patent thumbnailMethod for chemically stabilizing uranium carbide compounds, and device implementing the method
A process for chemical stabilization of a uranium carbide compound having formula: ucx+yc with x≧1 or 2 and y>0, x and y being true numbers, placed in a stabilization chamber, comprises: a rise in chamber internal temperature for “oxidation” of the compound based on uranium carbide between approximately 380° c. And 550° c., the chamber being fed with a neutral gas; isothermal oxidative treatment at the oxidation temperature, the chamber being placed under o2 partial pressure; controlling completion of stabilization of the compound, comprising monitoring the amount of molecular oxygen consumed and/or carbon dioxide or carbon dioxide and carbon monoxide given off, until achievement of an input set-point value for the amount of molecular oxygen, of a minimum threshold value for the amount of carbon dioxide or minimum threshold values for the carbon dioxide and carbon monoxide.
06/05/14
20140154727
 Geobacter strains that use alternate organic compounds, methods of making, and methods of use thereof patent thumbnailGeobacter strains that use alternate organic compounds, methods of making, and methods of use thereof
In preferred embodiments, the present invention provides new isolated strains of a geobacter species that are capable of using a carbon source that is selected from c3 to c12 organic compounds selected from pyruvate or metabolic precursors of pyruvate as an electron donor in metabolism and in subsequent energy production. The wild type strain of the microorganisms has been shown to be unable to use these c3 to c12 organic compounds as electron donors.
06/05/14
20140153684
 Dry phase reactor for generating medical isotopes patent thumbnailDry phase reactor for generating medical isotopes
An apparatus for generating medical isotopes provides for the irradiation of dry-phase, granular uranium compounds which are then dissolved in a solvent for separation of the medical isotope from the irradiated compound. Once the medical isotope is removed, the dissolved compound may be reconstituted in dry granular form for repeated irradiation..
05/29/14
20140147363
 Graphite article patent thumbnailGraphite article
One use for irradiated graphite after remediation processing is to recycle it into a new graphite artifact. Examples of such artifacts include an electrode to be used for vitrification of radionucleotides, graphite or carbon articles for uranium processing, a moderator for a htgr, in particularly a gen iv htgr, other types of graphite products for nuclear facilities, charcoal filters, silicon carbide applications, etc.
05/29/14
20140145097
 Radiation shields and methods of making the same patent thumbnailRadiation shields and methods of making the same
Various non-limiting embodiments disclosed herein generally relate to metallurgically dense radiation shields formed from bismuth alloys comprising from 10 weight percent to 60 weight percent tin that are essentially free of toxic heavy metals chosen from lead, cadmium, and uranium; and radiation attenuation devices comprising the same. Other non-limiting embodiments disclosed herein relate to methods of making metallurgically dense radiation shields comprising bismuth alloys comprising from 10 weight percent to 60 weight percent tin.
05/22/14
20140140905
 Separation and recovery of molybdenum values from uranium process distillate patent thumbnailSeparation 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, uf6, and molybdenum oxytetrafluoride, moof4.
05/22/14
20140137703
 Process for removing uranium in copper concentrate via magnetic separation patent thumbnailProcess for removing uranium in copper concentrate via magnetic separation
The present invention describes a process for removing uranium from a copper concentrate by magnetic separation (low and high field) to reduce the uranium content to commercially acceptable levels.. .
05/08/14
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..
04/24/14
20140112858
Fuel preparation for use in the production of medical isotopes
The present invention relates generally to the field of medical isotope production by fission of uranium-235 and the fuel utilized therein (e.g., the production of suitable low enriched uranium (leu is uranium having 20 weight percent or less uranium-235) fuel for medical isotope production) and, in particular to a method for producing leu fuel and a leu fuel product that is suitable for use in the production of medical isotopes. In one embodiment, the leu fuel of the present invention is designed to be utilized in an aqueous homogeneous reactor (ahr) for the production of various medical isotopes including, but not limited to, molybdenum-99, cesium-137, iodine-131, strontium-89, xenon-133 and yttrium-90..
04/24/14
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 notably finds application in the nuclear industry, in which it can advantageously be used to purify uranium hexafluoride (uf6) present in a gaseous or liquid stream, with regard to metal impurities which are also present in this stream.. .
04/10/14
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.. .
03/27/14
20140083570
Hot rolling of thick uranium molybdenum alloys
Disclosed herein are processes for hot rolling billets of uranium that have been alloyed with about ten weight percent molybdenum to produce cold-rollable sheets that are about one hundred mils thick. In certain embodiments, the billets have a thickness of about 7/8 inch or greater.
03/20/14
20140080217
Uranium analysis using luminescence enhancing oxidant and oxidant composition
According to the present invention, there is provided a method of determining a concentration of uranium including: a) a primary measuring step of measuring luminescence intensity or luminescence attenuation of uranium (vi) of an oxidant added sample obtained by adding an oxidant composition to a detection target sample; b) a secondary measuring step of adding different volumes of standard solution containing uranium (vi) having a predetermined concentration to a plurality of oxidant added samples, respectively, and then measuring luminescence intensity or luminescence attenuation of uranium (vi) contained in each standard solution added sample; and c) a calculating step of calculating a concentration of uranium (vi) contained in the detection target sample by a standard addition method based on the primary and secondary measurements. With the method for determining a concentration of uranium according to the present invention, the concentration of uranium may be further rapidly and accurately analyzed..
03/13/14
20140070434
Ionic liquids as templating agents in formation of uranium-containing nanomaterials
A method for forming nanoparticles containing uranium oxide is described. The method includes combining a uranium-containing feedstock with an ionic liquid to form a mixture and holding the mixture at an elevated temperature for a period of time to form the product nanoparticles.
03/13/14
20140070137
Method for the preparation of anhydrous hydrogen halides, inorganic substances and/or inorganic hydrides by using as reactants inorganic halides and reducing agents
A method for completely reducing an inorganic halide to obtain a non-halogen inorganic substance and/or hydride thereof and preferably anhydrous hydrogen halide fluid using inorganic halide substances, such as sulfur hexafluoride, nitrogen trifluoride, tungsten hexafluoride, uranium hexafluoride and others by reduction with a reducing agent at a proper temperature. The reducing agents may be molecular hydrogen, inorganic hydrides and inorganic metallic elements; molecular hydrogen is preferable, but in certain instances the inorganic hydrides are used, as well as inorganic metallic elements such as calcium and magnesium..
02/20/14
20140051150
Isolation of a protein responsible for uranium (vi) reduction
The present invention relates to the isolation and characterization of a protein responsible for the reduction of uranium (vi) to uranium (iv). The present invention extends to the use of the isolated protein in the reduction of uranium (vi) to uranium (iv) and further extends to a process for the bioremediation, or at least partial remediation, of a site contaminated with a source of u (vi).
02/13/14
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.
02/06/14
20140037518
Method of recycling spent nuclear fuel
This method provides for a two-phase voloxidation of a reaction mass using gas-air mixture, the reaction mass including fragmented uranium dioxide snf elements with containers. The first phase is carried out at 400÷650° c.
01/30/14
20140027315
Dual containment pressure vessel for storage and transport of uranium hexafluoride
A cylinder for storage and transport of uranium hexafluoride includes a generally tubular main body with a distally arranged end member defining an interior region. An interior tubular member is received in the interior region.
01/09/14
20140011668
Method of producing a catalyst body containing uranium oxide as active component
A method for producing a uranium oxide catalyst body includes the following steps: uo2+x powder, with x≦0.7, having a purity of at least 50% is sintered in a first sintering process to obtain a uo2+y intermediate, where y≦0.25. Then the uo2+y intermediate is oxidized with oxygen and converted in the process into a u3o8−z powder, with z≦1.
01/02/14
20140001381
System for storage and transport of uranium hexafluoride
An overpack for receiving a steel cylinder, such as a stainless steel cylinder, containing uranium hexafluoride includes a semi-cylindrical top portion having an arcuate main body and opposing first and second semi-circular end members, and includes a semi-cylindrical bottom portion having an arcuate main body and opposing first and second semi-circular end members, with the first end members of the top and bottom portion being aligned, and the second end members of the top and bottom portion being aligned, and the overpack is disposed in a cradle.. .
12/26/13
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.
12/19/13
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.
12/19/13
20130336833
Powder of an alloy based on uranium and on molybdenum useful for manufacturing nuclear fuels and targets intended for producing radioisotopes
Applications: manufacturing of nuclear fuels, notably for experimental nuclear reactors; manufacturing of targets for producing radioisotopes, notably for the medical industry.. .
12/19/13
20130333519
Method for preparing a powder of an alloy based on uranium and molybdenum
Applications: manufacturing of nuclear fuels, notably for mtrs.. .
12/12/13
20130329849
Metal nuclear-fuel pin including a shell having threads or fibers made of silicon carbide (sic)
A nuclear-fuel pin including a linear element made of a metal nuclear-fuel material consisting of uranium and/or plutonium, and cladding including fe and cr or an alloy including at least both of said elements, comprises a main shell provided around the linear nuclear-fuel element, said shell including threads or fibers made of sic. A method for producing a nuclear-fuel pin is also provided..
12/05/13
20130322591
Fuel assembly
Nuclear fuel assemblies include fuel elements that are sintered or cast into billets and co-extruded into a spiral, multi-lobed shape. The fuel kernel may be a metal alloy of metal fuel material and a metal-non-fuel material, or ceramic fuel in a metal non-fuel matrix.
12/05/13
20130322590
Extension of methods to utilize fully ceramic micro-encapsulated fuel in light water reactors
A 12×12 fully ceramic micro-encapsulated fuel assembly for a light water nuclear reactor includes a set of fcm fuel rods bundled in a square matrix arrangement. The fully ceramic micro-encapsulated fuel is comprised of tristructural-isotropic particles.
12/05/13
20130322588
Nuclear reactor and power generation facility
A nuclear reactor provided with a core which is provided with a new fuel part which contains uranium and a burning part in which fuel burns, wherein power is generated by fission of plutonium and wherein the burning part moves in a direction toward the new fuel part from the beginning to end of the operation cycle. The nuclear reactor is provided with a reactivity applying mechanism to apply the reactivity which can change the power of the core when the temperature of the coolant which flows through the inside of the core changes and performs control to change the temperature of the coolant which flows through the inside of the core in accordance with the change of power which is demanded for the core..
11/28/13
20130312570
Purification process
A process for purifying mo-99 from an acidic solution obtained by dissolving an irradiated solid target comprising uranium in an acidic medium, or from an acidic solution comprising uranium and which has previously been irradiated in a nuclear reactor, or from an acidic solution comprising uranium and which has been used as reactor fuel in a homogeneous reactor, the process comprising contacting the acidic solution with an adsorbent comprising a zirconium oxide, zirconium hydroxide, zirconium alkoxide, zirconium halide and/or zirconium oxide halide, and eluting the mo-99 from the adsorbent using a solution of a strong base, the eluate then being subjected to a subsequent purification process involving an alkaline-based mo-99 chromato-graphic recovery step on an anion exchange material. Also provided is apparatus for carrying out the process..
11/21/13
20130308741
Nuclear fuel containing recycled and depleted uranium, and nuclear fuel bundle and nuclear reactor comprising same
Fuel bundles for a nuclear reactor are described and illustrated, and in some cases includes fuel elements each having a first fuel component of recycled uranium; and a second fuel component of at least one of depleted uranium and natural uranium blended with the first fuel component, wherein the blended first and second fuel components have a first fissile content of less than 1.2 wt % of 235u. Other fuel bundles are also described and illustrated, and include a first fuel element including recycled uranium, the first fuel element having a first fissile content of no less than 0.72 wt % of 235u; and a second fuel element including at least one of depleted uranium and natural uranium, the second fuel element having a second fissile content of no greater than 0.71 wt % of 235u..
11/14/13
20130301780
Nuclear fuel containing a neutron absorber
Fuel bundles for a nuclear reactor are described and illustrated, and in some cases include fuel elements each having a fissile content of 235u between about 0.9 wt % 235u and 5.0 wt % 235u, and wherein at least one of the fuel elements is a poisoned low-enriched uranium fuel element including a neutron poison in a concentration greater than about 5.0 vol %.. .
11/07/13
20130297229
Method of measuring radioactive material of ingot using hpge gamma scintillator
Disclosed is a method of measuring radioactive material of an ingot, in which an ingot having a volume produced by subjecting metal waste generated in nuclear fuel processing or production facilities to melting decontamination undergoes gamma spectroscopy using a hpge detector to measure gamma rays of u-235 (185.72 key, 57.2%) among uranium isotopes, followed by performing detector calibration using a certified reference material and self-absorption correction depending on the density of a medium using mcnp computer code, and which enables rapid determination of whether the ingot subjected to melting decontamination may be self-disposed of or not.. .
11/07/13
20130297228
Method of measuring radioactive material of ingot using nai gamma scintillator
Disclosed is a method of measuring radioactive material of an ingot, in which an ingot having a volume produced by subjecting metal waste generated in nuclear fuel processing or production facilities to melting decontamination undergoes gamma spectroscopy using a nai detector to measure gamma rays of u-235 (185.72 kev, 57.2%) among uranium isotopes, followed by performing detector calibration using a certified reference material and self-absorption correction depending on the density of a medium using mcnp computer code, and which enables rapid determination of whether the ingot subjected to melting decontamination may be self-disposed of or not.. .
11/07/13
20130296628
Method of disposing of radioactive metal waste using melting decontamination
Disclosed is a method of disposing of radioactive metal waste using melting decontamination, including sorting radioactive metal waste generated in nuclear fuel processing or production facilities by predetermined sorting criteria, and charging sorted metal waste into a melting furnace so as to be melted; adding a impurity remover to the melt of the melting furnace to remove generated slag; pouring the melt having no slag into a mold to form an ingot; subjecting the ingot to gamma spectroscopy using a gamma spectrometer to measure gamma rays of u-235 (185.72 kev, 57.2%) among uranium isotopes, performing detector calibration using a certified reference material and self-absorption correction depending on the density of a medium using mcnp computer code, and calculating total radioactivity of the ingot from the quantified radioactivity and mass of u-235; and efficiently and rapidly determining whether the ingot subjected to radioactivity measurement satisfies a clearance limit.. .
10/10/13
20130266112
Nuclear fuel containing recycled and depleted uranium, and nuclear fuel bundle and nuclear reactor comprising same
Nuclear fuels for nuclear reactors are described, and include nuclear fuels having a first fuel component of recycled uranium, and a second fuel component of depleted uranium blended with the first fuel component, wherein the blended first and second fuel components have a fissile content of less than 1.2 wt % of 235u. Also described are nuclear fuels having a first fuel component of recycled uranium, and a second fuel component of natural uranium blended with the first fuel component, wherein the blended first and second fuel components have a fissile content of less than 1.2 wt % of 235u..
09/19/13
20130240805
Uranium dioxide nuclear fuel containing mn and al as additives and method of fabricating the same
Uo2 nuclear fuel pellets are fabricated by adding additive powder comprising mn compound and al compound into uo2 powder.. .
09/12/13
20130233716
Room temperature electrodeposition of actinides from ionic solutions
Uranic and transuranic metals and metal oxides are first dissolved in ozone compositions. The resulting solution in ozone can be further dissolved in ionic liquids to form a second solution.
08/29/13
20130223582
Fabrication method of burnable absorber nuclear fuel pellets and burnable absorber nuclear fuel pellets fabricated by the same
A fabrication method of burnable absorber nuclear fuel pellets and burnable absorber nuclear fuel pellets fabricated by the same are provided, in which the fabrication method includes adding boron compound and manganese compound to one or more type of nuclear fuel powders selected from the group consisting of uranium dioxide (uo2), plutonium dioxide (puo2) and thorium dioxide (tho2) and mixing the same (step 1), compacting the mixed powder of step 1 into compacts (step 2), and sintering the compacts of step 2 under hydrogen atmosphere (step 3). According to the fabrication method, sintering can be performed under hydrogen atmosphere at a temperature lower than the hydrogen atmosphere sintering that is conventionally applied in the nuclear fuel sintered pellet mass production, by adding sintering additives such as manganese oxide or the like..
08/22/13
20130217135
Compound for forming fluorescent uranium complex, method for synthesizing thereof, fluorescent probe for detecting uranium and method for analyzing uranium
Disclosed is a compound suitable for use as a fluorescent probe for the detection of uranium. The compound enables the qualitative and quantitative analysis of uranium present in waste samples using less expensive apparatuses.
08/15/13
20130206599
Method for measuring the uranium concentration of an aqueous solution by spectrophotometry
A method for measuring the uranium concentration of an aqueous solution including the following successive steps: a) electrochemical reduction towards valence iv, of the uranium present in the aqueous solution with a valence greater than iv, this reduction being implemented at ph<2 and by passing an electrical current in the solution; b) measurement of the absorbance of the solution obtained on completion of step a) at a chosen wavelength between 640 and 660 nm, and preferably 652 nm; and c) determination of the uranium concentration of the aqueous solution by deduction of the uranium concentration of valence (iv) present in the aqueous solution obtained on completion of step a) from measurement of the absorbance obtained in step b).. .
08/08/13
20130202501
Process for reprocessing spent nuclear fuel not requiring a plutonium-reducing stripping operation
This process finds particular application in the processing of uranium oxide fuels and uranium and plutonium mixed oxide fuels.. .
08/08/13
20130202076
Nuclear fuel bundle containing thorium and nuclear reactor comprising same
Fuel bundles for a nuclear reactor are disclosed, and in some embodiments include a first fuel element including thorium dioxide; a second fuel element including uranium having a first fissile content; and a third fuel element including uranium having a second fissile content different from the first fissile content. Nuclear reactors using such fuel bundles are also disclosed, including pressurized heavy water nuclear reactors.
07/25/13
20130188248
Reflective optical element and method of manufacturing the same
A reflective optical element e.g. For use in euv lithography, configured for an operating wavelength in the range from 5 nm to 12 nm, includes a multilayer system with respective layers of at least two alternating materials having differing real parts of the refractive index at the operating wavelength.
07/18/13
20130180914
Long-term sequestration of uranium in iron-rich and iron-enriched sediment
In situ formation of u(vi)-fe(iii) oxides and hydroxides can provide effective uranium remediation. The reason for this is that such compounds can effectively sequester uranium, even in the (vi) oxidation state.
07/11/13
20130175719
Method for fabricating porous uo2 sintered pellet for electrolytic reduction process for recovering metallic nuclear fuel using continuous process of atmospheric sintering and reduction, and porous uo2 sintered pellet fabricated by the same
A method for fabricating porous uo2 sintered pellets to be fed into an electrolytic reduction process for the purpose of metallic nuclear fuel recovery is provided, which includes forming a powder containing u3o8 by oxidizing a spent nuclear fuel containing uranium dioxide (uo2) (step 1), fabricating u3o8 green pellets by compacting the powder formed in step 1 (step 2), and fabricating uo2 sintered pellets by sintering the u3o8 green pellets fabricated in step 2 at 1000 to 1600° c., in an atmospheric gas, and cooling the same for reduction, by changing the atmosphere to a reducing atmospheric gas (step 3). The porous uo2 sintered pellets can be fabricated, which do not have any defects.
07/04/13
20130168322
Amidoxime-modified polyacrylonitrile porous body
An amidoxime-modified pan porous body obtained by reacting with hydroxylamine a polyacrylonitrile porous body that is monolithic, has a thickness of 1 mm or more and contains polyacrylonitrile (pan) as the main component to convert a nitrile group of the polyacrylonitrile porous body into an amidoxime group. This porous body is a porous body for adsorbing a metal ion, for example, an ion of metal such as copper, iron, nickel, vanadium, indium, gallium, silver, mercury, lead, uranium, plutonium, cesium, barium, lanthanum, thallium and strontium..
06/27/13
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.
06/27/13
20130160797
Cathode scraper system and method of using the same for removing uranium
Embodiments include a cathode scraper system and/or method of using the same for removing uranium. The cathode scraper system includes a plurality of cathode assemblies.
06/13/13
20130148774
Porous uo2 sintered pellets and method for fabricating porous uo2 sintered pellets and electrolytic reduction using same
A method for fabricating porous uo2 sintered pellets to be fed into the electrolytic reduction process for the purpose of metallic nuclear fuel recovery is provided, which includes forming a powder containing u3o8 by oxidizing spent nuclear fuel containing uranium dioxide (uo2) (step 1), fabricating green pellets by compacting the powder formed in step 1 (step 2), fabricating uo2+x sintered pellets by sintering the porous u3o8 green pellets fabricated in step 2 at 1200 to 1600° c., in an atmospheric gas (step 3), and forming uo2 sintered pellets by cooling the uo2+x sintered pellets to room temperature, and reduction the same at 1000 to 1400° c., in a reducing atmosphere (step 4).. .
05/02/13
20130104698
Method of catalytic oxidation of u4+ to u6+ using a catalyst muhamedzhan-1
The proposed methods are exemplarily utilized in uranium hydrometallurgy for selective extraction of uranium out of ore by in situ or heap leaching. According to the disclosure, the methods encompass catalytic oxidation of u4+ to u6+ using a proposed oxidizing catalyst “muhamedzhan-1”, filtration of this solution through ore, transferring hexavalent uranium, trivalent iron, and other metal ions into a production solution, extraction of uranium yielding a barren solution and re-circulation of this solution back for ore leaching.
05/02/13
20130104577
Defrosting
An apparatus comprising a cooling chamber comprising a cooling unit for cooling uranium hexafluoride in the chamber; and a heater comprising a first region inside the cooling chamber arranged to defrost the cooling unit; and a second region outside the cooling chamber arranged to receive heating fluid cooled in the first region and to heat the received fluid. A method of defrosting a cooling unit is also described..
04/25/13
20130101470
Reaction chamber for exothermic material
A chamber for reacting exothermic material, comprising a multilevel structure including at least: a receptacle for storing said material, corresponding to a lower level; a median level comprising a reactive load containing at least one alkaline-earth carbonate, so as to absorb heat emitted during the oxidation reaction of said material, said alkaline-earth carbonate decomposing under the effect of the heat in an endothermic reaction; an upper level comprising a cover. According to one variant, the material stored is a carbide of plutonium and/or uranium..
04/11/13
20130087464
Room temperature electrodeposition of actinides from ionic solutions
Uranium metal can be electrochemical deposited from room temperature ionic liquid (rtil), tri-methyl-n-butyl ammonium n-bis(trifluoromethansulfonylimide) [me3nnbu][tfsi] providing an alternative non-aqueous system for the extraction and reclamation of actinides from reprocessed fuel materials. Deposition of u metal is achieved using tfsi complexes of u(iii) and u(iv) containing the anion common to the rtil.
03/21/13
20130072741
Removal equipment and method for the storage facility for transuranium compounds
A nuclear chemistry laboratory adopts in-site cutting technique to remove the large-scale glove box contaminated by transuranium compounds. During removal operation, to prevent further spreading of contamination, it is necessary to build an alpha airtight quarantine tent around the glove box that is ready to be cut.
03/21/13
20130071659
Fiber-based adsorbents having high adsorption capacities for recovering dissolved metals and methods thereof
A fiber-based adsorbent and a related method of manufacture are provided. The fiber-based adsorbent includes polymer fibers with grafted side chains and an increased surface area per unit weight over known fibers to increase the adsorption of dissolved metals, for example uranium, from aqueous solutions.
02/14/13
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.. .
01/24/13
20130022686
Combinations of liquid filtration media and methods for enhanced filtration of selected water contaminants
By sequentially aligning various filtration media and delivery systems, enhanced synergistic reduction of water contaminants is obtained compared to the prior art or separate use of the individual media/filters. Specific filtration media are formulated with proper proportioning and sequencing to enhance the ability to reduce metals that cause staining, odors and bad taste such as iron, copper and manganese.
01/24/13
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.
01/24/13
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.
01/10/13
20130010914
Composite materials, bodies and nuclear fuels including metal oxide and silicon carbide and methods of forming same
Methods of forming composite bodies and materials including a metal oxide, such as, uranium dioxide, and silicon carbide are disclosed. The composite materials may be formed from a metal oxide powder, a silicon carbide powder and, optionally, a carbon powder.


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Uranium topics: Thorium Dioxide, Solubility, Plutonium Dioxide, Fluorescent Probe, Qualitative, Quantitative, Absorbance, Aqueous Solution, Spectrophotometry, Boron Carbide, Euv Lithography, Phosphoric Acid, Electrolyte, Hydroxylamine, Uranium Hexafluoride

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