| Class / Patent application number | Description | Number of patent applications / Date published |
|---|
| 588002000 | By fixation in stable solid media | 61 |
| 20110301399 | AQUEOUS WASTE DISPOSAL USING SUPERABSORBENT - A method for disposing of liquid aqueous laboratory waste is described, which involves solidifying the liquid waste with a suitable isovolumic, space-filling superabsorbent polymer within a disposable impermeable film-type container held in an open top, reusable rigid outer container such as a laboratory beaker, and closing and removing the disposable container containing the solidified waste from the rigid outer container. The waste held in the film-type container can be finally disposed of through incineration or deposit in a suitable landfill. | 12-08-2011 |
| 20120071703 | Method of immobilizing nuclear waste - Method of immobilizing nuclear waste comprising:
| 03-22-2012 |
| 20130023713 | METHOD FOR THE PRE-CALCINING TREATMENT OF AN AQUEOUS NITRIC SOLUTION COMPRISING AT LEAST ONE RADIONUCLIDE AND OPTIONALLY RUTHENIUM - A method for treating before calcination a nitric aqueous solution comprising at least one radionuclide and ruthenium is provided. The method comprises a step for adding to the solution a compound selected from lignins, lignocelluloses, optionally as salts and mixtures thereof. | 01-24-2013 |
| 20140005461 | METHOD FOR STABILIZING WASTE AND HAZARDOUS WASTE | 01-02-2014 |
| 20150038760 | METHOD FOR PROCESSING RADIOACTIVELY-CONTAMINATED WATER - The present invention provides an efficient and low cost method for processing radioactively-contaminated water. The method for processing radioactively-contaminated water comprising a freeze concentration step of generating ice having lowered concentration of radioactive substance from radioactive substance containing contaminated water and concentrating the radioactive substances in the residual contaminated water by the interface progressive freeze concentration process. Preferably, the method further comprises a nitrogen substitution step of reducing dissolved oxygen in the contaminated water and adding nitrogen gas to the contaminated water, as a previous step of the freeze concentration step. Preferably, the radioactive substance is radioactive cesium. | 02-05-2015 |
| 20150065775 | SYSTEM AND METHOD TO CONTROL MIGRATION OF CONTAMINATES WITHIN A WATER TABLE - System and Method is described that controls the release of contaminated water by rapidly freezing the ground water, including salt water, which permeates the area underneath the a contamination source, so that the resulting ice lens mitigates the extent to which radioactive water is released into the environment. An aperture in the containment area allows the dispersal and dilution of the contaminates by allowing in ground water from outside, and/or removing water from the containment area. The variable aperture may be a physical valve or preferably an opening in the ice shield which size may be controlled by freezing or thawing portions of the ice shield. | 03-05-2015 |
| 20220139586 | ORGANIC IODINE REMOVER - As an organic iodine remover that removes organic iodine in a containment vessel of a nuclear reactor, an organic agent (for example, an ionic liquid, an interfacial active agent, a quaternary salt, or a phase transfer catalyst) having a function of dissolving and decomposing the organic iodine and retaining iodine is used. The organic iodine remover is a substance composed of a cation and an anion. The organic iodine remover is, in particular, an organic iodine remover in which, in a structure of the cation of the organic agent, carbon or oxygen is bonded to, via a single bond, to a phosphorus element, a sulfur element or a nitrogen element, the number of carbon chains is 2 or more, and an anionic structure is configured with a substance with high nucleophilicity. By using such an organic agent, the organic iodine is removed with an efficiency of 99% or more. | 05-05-2022 |
| 588003000 | Cement, concrete, or hydraulic setting | 12 |
| 20080207977 | 300-Year disposal solution for spent nuclear fuel - A method including a combination of intermediate storage and reprocessing is utilized to process spent nuclear fuel (SNF) and thereby effect a disposition of that SNF within a period of 300 years. The method includes five or more years of pool water storage wherein ninety-nine percent (%) of the fission wastes energy decays. The waste material is then stored in an air convention storage facility, before processing to separate Cesium and Strontium from the waste is effected. This air convection cooling may be done in convection air-cooled concrete casks. During 50 years of convection air-cooled storage the energy contained in the waste material declines another one half %. Thereafter, at any point the SNF is processed to sufficiently separate 99.999% of the 97% of actinides (approximately 95% U238 uranium, 1% U235 uranium, and 1% Pu239 plutonium) from the 3% fission wastes. Again, it is only necessary to provide approximately 99.999% separation of the TRU's (transuranic waste) from the fps (fission products)—more specifically, sufficient separation so that the residual fps are contaminated with less than 100 nCi/g TRU's, as defined in the Class C regulations—10CFR61. The separated actinides and transuranics are thereafter utilized in the manufacture of MOX (mixed oxide) or fast burner reactor fuel pellets for future reactor fuel. The remaining fission wastes, containing Cesium and Strontium, are then placed into containers and subsequently put into dry storage for the remainder of around 300 years, where most of the remaining half % of its radiation energy material, i.e., Cesium and Strontium decays. Thereafter this fission waste is put into a low level Class-C nuclear waste repository, which may include leaving them in the intermediate storage facility that is also designed to accommodate and dispose Class C waste. | 08-28-2008 |
| 20080249347 | Waste Stabilization and Packaging System for Fissile Isotope-Laden Wastes - A method for shredding, blending, and packaging wastes so that the shipment of waste is acceptable for transportation and disposal. Containerized wastes, laden with fissile isotopes, such as that from sodium fluoride traps and granulated carbon collection assemblies, if not already conditioned, are conditioned for blending by shredding them to particle size and then blending the particles with a grout mix either in a reusable mixing vessel or in a disposable mixing vessel. The blend of waste and grout is selected to meet governmental and disposal site requirements for an acceptable waste shipment. If blended in a reusable vessel, the blended waste is transferred to shipping containers. The shipping containers are placed in shielded shipping casks for transportation to a disposal facility. The grout mix is a combination of grout and neutron poisons such as borated sand to prevent criticality. | 10-09-2008 |
| 20080255400 | Hydrotalcite-Like Substance, Process for Producing the Same and Method of Immobilizing Hazardous Substance - A hydrotalcite-like substance that is capable of ion exchange with target anions, being of small crystal size and that exhibits large basal spacing, excelling in anion exchange performance; a process for producing the same; and a method of immobilizing hazardous substances. A hydrotalcite-like substance is produced by mixing an acidic solution containing aluminum ions and magnesium ions with an alkaline solution containing an alkali to thereby synthesize a hydrotalcite-like substance, followed by, without ageing, water removal or neutralization. The molar ratio of aluminum ions and magnesium ions is preferably in the range of 1:5 to 1:2. Hazardous substances can be immobilized by pulverizing the hydrotalcite-like substance after synthesis and adding the powder to a subject, or adding a hydrotalcite-like substance in slurry form to a subject, or carrying out addition so as to cause the synthesis directly at the position of the subject. Furthermore, anion adsorption can be performed by a filter containing the hydrotalcite-like substance. | 10-16-2008 |
| 20090149689 | SYSTEM FOR AND METHOD OF FILLING A CONTAINER WITH HAZARDOUS WASTE - A system for filling a container with hazardous waste includes a primary confinement chamber that houses a lid handling mechanism and a filling head. The lid handling mechanism may be used to remove and/or recouple the lid to the container as part of the process of filling the container in such a way to ensure the exterior of the container is not contaminated by the hazardous waste. The filling head may be configured to add the hazardous waste to the container, mix the contents of the container, and/or vent air from the container. The system may include additional mechanisms in the primary confinement chamber to add dry cementitious material to the container, add premixed wet cementitious materials to the container, add cementitious material to seal off the top of the lid, and/or measure the level and test whether the contents of the container meet quality assurance requirements. The hazardous waste may include radioactive hazardous waste. In particular, the radioactive waste may be transuranic waste that emits alpha particles. | 06-11-2009 |
| 20090156878 | Cement-Based Composition for the Embedding of a Boron-Containing Aqueous Solution, Embedding Process and Cement Grout Composition - Cement-based composition for the embedding of a boron-containing aqueous solution, said composition being composed of a sulphoaluminate cement optionally comprising gypsum, and of a sand. | 06-18-2009 |
| 20110218377 | Method for solidifying and stabilizing waste acid - The present invention discloses a method for solidifying and stabilizing waste acid including steps of condensing waste acid containing phosphoric acid to reduce the volume; mixing the condensed waste acid with waste acid containing fluoroboric acid to solidify and stabilize the mixed waste acid. The pH of the mixed acid is adjusted by adding barium hydroxide as a neutralizer. The efficiency of solidifying waste acid can be improved by partially granulating and by adding solidifying agent indirectly. The method of the present invention can prevent intensely exothermic reaction caused by adding solidifying agents. Furthermore, the method of the present invention is controlled in a temperature of 30 to 45° C. to improve the polymerization of the mixed waste acid so that the efficiency of solidification thereof can be also improved. | 09-08-2011 |
| 20120041249 | METHOD FOR SOLIDIFYING AND STABILIZING WASTE ACID - The present invention discloses a method for solidifying and stabilizing waste acid including steps of condensing waste acid containing phosphoric acid to reduce the volume; mixing the condensed waste acid with waste acid containing fluoroboric acid to solidify and stabilize the mixed waste acid. The pH of the mixed acid is adjusted by adding barium hydroxide as a neutralizer. The efficiency of solidifying waste acid can be improved by partially granulating and by adding solidifying agent indirectly. The method of the present invention can prevent intensely exothermic reaction caused by adding solidifying agents. Furthermore, the method of the present invention is controlled in a temperature of 30 to 45° C. to improve the polymerization of the mixed waste acid so that the efficiency of solidification thereof can be also improved. | 02-16-2012 |
| 20140142365 | Method and apparatus for identification, stabilization and safe removal of radioactive waste and non hazardous waste contained in buried objects - A method and apparatus for the stabilization and safe removal of buried waste that is tested and classified as being transuranic or not transuranic waste and disposed accordingly. The buried waste (usually in vertical pipe units) is enclosed in a casing and ground and mixed with the surrounding soil. This process allows for chemical reactions to occur that stabilizes the mixture. The entire process is contained within the casing to avoid contamination. In situ or external testing is done for radio isotopes to classify the waste. If it is classified as transuranic the waste is removed in a controlled way into a retrieval enclosure and disposed off in drums. If the waste is not transuranic then grout is introduced into the mixture, allowed to set and the resulting monolith is removed and buried in trenches. | 05-22-2014 |
| 588004000 | With additional solid material to enhance fixation of radioactivity | 4 |
| 20100004498 | Reducing the profile of neutron-activated 60Co and removing in layers at the primary system of a permanently shut down nuclear power plant in order to accelerate its dismantling - The Decommissioning Phase SAFSTOR for Nuclear Power Plants, lasting for 50 to 60 years before dismantling begins, is to allow for natural decay of | 01-07-2010 |
| 20100160707 | Encapsulation of Waste for Storage - An apparatus for encapsulating waste material (e.g radioactive sludge from nuclear processing plant) in a container (e.g Nirex box) for long term storage, comprising: a first storage vessel, for holding sludge; a second storage vessel, for holding encapsulation medium (e.g. cement based grout); an inline mixer (e.g. a static inline mixer), coupled for receiving sludge, and coupled to the second storage vessel, and producing, in use, a mixture of the sludge and grout; wherein the inline mixer is arranged for filling the container. Preferably, a dewatering unit (e.g. HydroTrans based), coupled for receiving sludge and outputting dewatered sludge to be mixed by the inline mixer. An encapsulation system comprising the encapsulation apparatus, and corresponding encapsulation methods, are also disclosed. | 06-24-2010 |
| 20140336437 | CEMENT CURING FORMULATION AND METHOD FOR HIGH-LEVEL RADIOACTIVE BORON WASTE RESINS FROM NUCLEAR REACTOR - A cement curing formulation and curing method for high-level radioactive boron waste resins from a nuclear reactor. The curing formulation comprises the following raw materials: cement, lime, water, curing aids and additives. The curing method comprises: (1) weighing the raw materials and the high-level radioactive boron waste resins, and adding lime into a curing container; (2) then adding the high-level radioactive boron waste resins; (3) feeding other raw materials under stirring; (4) adding the cement and supplementing water depending on the moisture state of the cement, and stirring until uniform; and (5) standing and maintaining after stirring until uniform. The curing formulation has the features of a high curing containment rate, high strength of the cured body, better water resistance, better freeze-thaw resistance, and low radioactive leakage. | 11-13-2014 |
| 20150325320 | METHOD FOR STORING RADIOCONTAMINATED WASTE MATTER AND CONTAINER THEREFOR - A method for securely and safely storing radiocontaminated waste matter and a container therefor are provided. | 11-12-2015 |
| 588005000 | Bituminous | 1 |
| 20090069620 | RADIATION SHIELDS AND TECHNIQUES FOR RADIATION SHIELDING - Radiation shields and techniques for radiation shielding are provided. Bitumen substances, such as asphalt or tar, are mixed with radioactive waste, leaded glass, or a radioactive waste and leaded glass composite. In embodiments where the bitumen substance is mixed with leaded glass, the resulting mixture can be coated onto containers that house radioactive waste or the resulting mixture can be coated onto the outer surface of the radioactive waste. | 03-12-2009 |
| 588006000 | Resin or polymer; e.g., cellulose, polyethylene | 12 |
| 20090112042 | Decontamination method of metal surface contaminated by radioactive element - A decontamination gel is obtained to be spayed on a contaminated material. Places of contaminations of Co, Cs and Sr are shown by the gel. Then the gel is dried up in the air to form a film. Thus, the contaminations are cleaned by removing the film. | 04-30-2009 |
| 20100094073 | METHOD FOR TREATING WASTE PETROLEUM - The present invention relates to a method for treating various waste petroleum into eco-friendly solid that leaching of oil would not occur. The method of the present invention can treat radioactive waste petroleum used in a nuclear power station as well as various waste petroleum, thereby stabilizing waste petroleum chemically and physically, wherein the method comprises mixing waste petroleum with a concentrated sulfuric acid and a concentrated nitric acid; adding sodium hydroxide, thereby carrying out a second polymerization reaction to produce solid particles; colloidizing a mixture obtained by uniformly stirring the solid particles; adding a diisocyanate compound in reactor, thereby carrying out a third polymerization reaction to obtain a new compound in the form of powder; discharging a generated gas into the atmosphere; and filling the compound into a resin as a filling material and compression molding and reclaiming the filled compound. | 04-15-2010 |
| 20120184796 | CAPTURE, REMOVAL, AND STORAGE OF RADIOACTIVE SPECIES IN AN AQUEOUS SOLUTION - The present invention relates to a method of capturing radioactive species in an aqueous solution and removing the radioactive species for disposal. The method includes the steps of providing a macroporous bead form sequestration resin, subjecting the bead form sequestration resin to radioactive species contained in the aqueous solution to allow the bead form sequestration resin to capture the radioactive species; and disposing the radioactive species in a radioactive storage facility. | 07-19-2012 |
| 20130109902 | RADIOACTIVE-SUBSTANCE-ABSORBENT, RADIOACTIVE-SUBSTANCE-ABSORBENT PRODUCTION DEVICE, DECONTAMINATING METHOD, AND BAG UNIT | 05-02-2013 |
| 20140194665 | RADIOACTIVE CESIUM ADSORBENT, METHOD FOR PRODUCING THE SAME, AND METHOD FOR REMOVING RADIOACTIVE CESIUM IN ENVIRONMENT WITH SAID ADSORBENT - The present invention relates to a radioactive cesium adsorbent, a method for producing the same, and a method for decontaminating the environment from radioactive cesium with the adsorbent. The radioactive cesium adsorbent of the present invention includes a hydrophilic fiber substrate supporting a Prussian blue analogue, in particular, Prussian blue, and the Prussian blue analogue is immobilized in the inside of the fibers. | 07-10-2014 |
| 20140235916 | METHOD FOR REMOVING RADIOACTIVE CESIUM, HYDROPHILIC RESIN COMPOSITION FOR REMOVAL OF RADIOACTIVE CESIUM, METHOD FOR REMOVING RADIOACTIVE IODINE AND RADIOACTIVE CESIUM, AND HYDROPHILIC RESIN COMPOSITION FOR REMOVAL OF RADIOACTIVE IODINE AND RADIOACTIVE CESIUM - A method for removing radioactive cesium and/or iodine from a radioactive substance in liquid and/or a solid matter using a hydrophilic resin composition comprising a hydrophilic resin and a metal ferrocyanide compound, wherein the hydrophilic resin includes at least one hydrophilic resin selected from the group consisting of a hydrophilic polyurethane resin, a hydrophilic polyurea resin, and a hydrophilic polyurethane-polyurea resin each having at least a hydrophilic segment, and a metal ferrocyanide compound is dispersed in the hydrophilic resin composition in a ratio of at least 1 to 200 mass parts relative to 100 mass parts of the hydrophilic resin. | 08-21-2014 |
| 20140288346 | METHOD FOR REMOVING RADIOACTIVE CESIUM, HYDROPHILIC RESIN COMPOSITION FOR REMOVING RADIOACTIVE CESIUM, METHOD FOR REMOVING RADIOACTIVE IODINE AND RADIOACTIVE CESIUM, AND HYDROPHILIC RESIN COMPOSITION FOR REMOVING RADIOACTIVE IODINE AND RADIOACTIVE CESIUM - The present invention intends to provide a method for removing radioactive cesium, or radioactive iodine and radioactive cesium that is simple and low-cost, further does not require an energy source such as electricity, moreover can take in and stably immobilize the removed radioactive substances within a solid, and can reduce the volume of radioactive waste as necessary, and to provide a hydrophilic resin composition using for the method for removing radioactive cesium, or radioactive iodine and radioactive cesium, and the object of the present invention is achieved by using a hydrophilic resin composition containing: at least one hydrophilic resin selected from the group consisting of a hydrophilic polyurethane resin, a hydrophilic polyurea resin, and a hydrophilic polyurethane-polyurea resin each having at least a hydrophilic segment; and a zeolite dispersed therein in a ratio of at least 1 to 200 mass parts relative to 100 mass parts of the hydrophilic resin. | 09-25-2014 |
| 20150318064 | METHOD FOR REMOVING RADIOACTIVE CESIUM, HYDROPHILIC RESIN COMPOSITION FOR REMOVAL OF RADIOACTIVE CESIUM, METHOD FOR REMOVING RADIOACTIVE IODINE AND RADIOACTIVE CESIUM, AND HYDROPHILIC RESIN COMPOSITION FOR REMOVAL OF RADIOACTIVE IODINE AND RADIOACTIVE CESIUM - The present invention intends to provide a method for removing radioactive cesium, or radioactive iodine and radioactive cesium that is simple and low-cost, further does not require an energy source such as electricity, moreover can take in and stably immobilize the removed radioactive substances within a solid, and can reduce the volume of radioactive waste as necessary, and to provide a hydrophilic resin composition using for the method for removing radioactive cesium, or radioactive iodine and radioactive cesium, and the object of the present invention is achieved by using a hydrophilic resin composition containing: at least one hydrophilic resin selected from the group consisting of a hydrophilic polyurethane resin, a hydrophilic polyurea resin, and a hydrophilic polyurethane-polyurea resin each having at least a hydrophilic segment; and a clay mineral dispersed therein in a ratio of at least 1 to 180 mass parts relative to 100 mass parts of the hydrophilic resin. | 11-05-2015 |
| 20160118153 | CAPTURE, REMOVAL, AND STORAGE OF RADIOACTIVE SPECIES - A method of capturing radioactive species from an aqueous solution and removing the radioactive species for disposal, includes: contacting the aqueous solution with a first sequestration resin comprising a sequestration ligand coupled to a sulfonic acid based polymer resin backbone, to allow the first sequestration resin to capture the radioactive species; removing the first sequestration resin with the captured radioactive species from the aqueous solution; and using an acid to lower a pH of the first sequestration resin to release the radioactive species from the first sequestration resin. | 04-28-2016 |
| 588007000 | Ion exchange resin | 2 |
| 20090012343 | WASTE DISPOSAL METHOD - The invention provides a method for the production of a stable monolith, the method comprising the encapsulation of a waste material in the monolith by means of chemical bond formation within the monolith, and a method for the disposal and storage of waste materials, which comprises the production of a stable monolith by such method. Waste materials which are particularly suited to treatment according to the invention include various geopolymer precursors, most particularly ion exchange materials such as aluminosilicate materials, and the invention is particularly suited to the disposal and long term storage of radioactive waste materials. | 01-08-2009 |
| 20130090512 | RESIN VOLUME REDUCTION PROCESSING SYSTEM AND RESIN VOLUME REDUCTION PROCESSING METHOD - The cost relating to a reduction in volume and storage of a waste resin including a radioactive nuclide is reduced. In an aspect of the invention, a volume reduction processing system | 04-11-2013 |
| 588008000 | Polymer derived from ethylenically unsaturated monomer | 1 |
| 20140121439 | COMPOSITION AND PROCESS FOR PROCESSING RADIOACTIVE WASTE FOR SHIPMENT AND STORAGE - A process for encapsulating a radioactive object to render the object suitable for shipment and/or storage, and including the steps of preparing a plastic material, causing the plastic material to react with a foaming agent, generating a foaming plastic, encapsulating the radioactive object in the foaming plastic, and allowing the foaming plastic to solidify around the radioactive object to form an impervious coating. | 05-01-2014 |
| 588009000 | Clay or claylike | 5 |
| 20100069697 | Radioactive Material Sequestration - A radioactive material sequestration system may include a radionuclide containment composition dispenser and a sorption based media container. The radionuclide containment composition dispenser may be configured for holding a radionuclide containment composition and be capable of dispensing the radionuclide containment composition to remove radionuclides from a radioactive material. The radionuclide containment composition is a mixture of a clay mineral and water. The sorption based media container may be configured for holding a sorption based media; receiving dispensed radionuclide containment composition; and sequestering the radionuclides. The radioactive material sequestration system may also include a probe. | 03-18-2010 |
| 20100099937 | Secondary Process for Radioactive Chloride Deweaponization and Storage - A radioactive containment composition may be created for containing radionuclides from a radioactive material by mixing a clay mineral with water. This mixture may form an aqueous clay suspension. The mixture can be refined by filtering to remove coarse material. The aqueous clay suspension may be applied to a radioactive material, allowing the radionuclides to be exchanged with cations in the aqueous clay suspension. The resulting aqueous slurry, a silver-based solution may be added to produce a suspension. The suspension may be collected, heated and analyzed. | 04-22-2010 |
| 20100217061 | Counter Weapon Containment - A radioactive containment composition may be created for containing radionuclides from a radioactive material by mixing a clay mineral with water. This mixture may form an aqueous clay suspension, which in turn can be refined by filtering to remove coarse material. The aqueous clay suspension may be applied to a radioactive material, allowing the radionuclides to be exchanged with cations in the aqueous clay suspension. The resulting aqueous slurry may be collected, heated and analyzed. | 08-26-2010 |
| 20150011816 | SOLIDIFICATION METHOD OF RADIOACTIVE WASTE - A solidification method of radioactive waste is provided, including kneading a binder and an inorganic adsorbent to obtain a kneaded object, the in organic adsorbent included radionuclides; extruding the kneaded object to obtain an extruded material object; cutting the extruded material object to obtain at least one extruded material block; and firing the at least one extruded material block to solidify the at least one extruded material block. | 01-08-2015 |
| 20160027543 | METHOD FOR MANUFACTURING SOLIDIFIED BODY OF RADIOACTIVE WASTE AND MANUFACTURING APPARATUS FOR SOLIDIFIED BODY - A method for manufacturing a solidified body of a radioactive waste includes a kneading step (S | 01-28-2016 |
| 588010000 | Ceramic or ceramiclike | 15 |
| 20100191033 | ADSORBENT FOR RADIOELEMENT-CONTAINING WASTE AND METHOD FOR FIXING RADIOELEMENT - An adsorbent for radioelement-containing waste includes spherical layered double hydroxide (A) or spherical metal hydroxide (B). (A) is a nonstoichiometric compound represented by general formula (a) or (b): | 07-29-2010 |
| 20100317911 | METHOD FOR PREPARING CERAMIC WASTE FORM CONTAINING RADIOACTIVE RARE-EARTH AND TRANSURANIC OXIDE, AND CERAMIC WASTE FORM WITH ENHANCED DENSITY, HEAT-STABILITY, AND LEACH RESISTANCE PREPARED BY THE SAME - Disclosed herein is a method for preparing a ceramic waste form containing radioactive rare-earth and transuranic oxide, and the ceramic waste form with enhanced density, heat-stability, and leach resistance prepared by the same. | 12-16-2010 |
| 20140114112 | CERAMIC INGOT OF SPENT FILTER HAVING TRAPPED RADIOACTIVE CESIUM AND METHOD OF PREPARING THE SAME - A method of preparing a simple ceramic ingot of a spent filter having radioactive cesium trapped therein, and a ceramic ingot of a spent filter having improved properties such as leach resistance, thermal stability, and cesium content are provided. The method includes grinding and mixing a spent filter having cesium trapped therein, adding a solidifying agent, and sintering the spent filter. The method of preparing a ceramic ingot of a spent filter can be useful in preparing the ceramic ingot of the spent filter from only the spent filter by means of simple grinding and sintering, and in preparing the ceramic ingot of the spent filter by adding a small amount of a solidifying agent. The ceramic ingot of the spent filter has a high density and improved thermal stability, and shows improved leach resistance since a leach rate of a radioactive material is remarkably low. Therefore, the spent filter having radioactive cesium trapped therein can be effectively used to prepare a stable ceramic ingot. | 04-24-2014 |
| 588011000 | Glass, glasslike, vitreous | 12 |
| 20080281141 | Method For Confining a Substance by Vitrification - The present invention relates to a process for the manufacture of a glass frit for the containment by vitrification of a material comprising at least one oxidizable or reducible chemical species, and also to a process for the containment of said material by vitrification. The process for the manufacture of the glass frit comprises a step of incorporating into a raw glass frit at least one redox couple, the nature and the amount of which make it possible to maintain said at least one chemical species in its oxidized or reduced state. The containment process comprises mixing and hot melting the resulting glass frit and the material to be contained. The present invention makes it possible to optimize the containment of pollutants such as radionucleides, metals and metalloids. The material may be nuclear waste or a material derived from the incineration of household refuse. | 11-13-2008 |
| 20110224472 | Isotope-Specific Separation and Vitrification Using Ion-Specific Media - Apparatuses, processes and methods for the separation, isolation, or removal of specific radioactive isotopes from liquid radioactive waste, these processes and methods employing isotope-specific media (ISM). In some embodiments, the processes and methods further include the vitrification of the separated isotopes, generally with the ISM; this isotope-specific vitrification (ISV) is often a step in a larger scheme of preparing the radioactive isotopes for long-term storage or other disposition. A variety of ISM are disclosed. | 09-15-2011 |
| 20120022311 | PROCESS FOR PACKAGING RADIOACTIVE WASTES IN THE FORM OF SYNTHETIC ROCK - The present invention relates to a process for packaging radioactive wastes, in which the following successive steps are carried out:
| 01-26-2012 |
| 20130109903 | METHODS OF CONSOLIDATING RADIOACTIVE CONTAINING MATERIALS BY HOT ISOSTATIC PRESSING | 05-02-2013 |
| 20130303822 | SYSTEM AND METHOD FOR VITRIFICATION OF WASTE - A method for vitrifying waste to prevent the formation of molybdate secondary phases includes forming a feed mixture that includes the waste, a source of vanadium, and at least one of glass frit or glass forming chemicals and vitrifying the feed mixture in a melter to produce a glass product that includes the waste. | 11-14-2013 |
| 20140066684 | MITIGATION OF SECONDARY PHASE FORMATION DURING WASTE VITRIFICATION - A method for vitrification of waste to reduce the formation of persistent secondary phases comprising separating at least one glass frit constituent from an initial glass frit to form a modified glass frit. The waste, modified glass frit, and the at least one glass frit constituent are mixed together with the modified glass frit and the at least one glass frit constituent being added as separate components. The resulting mixture is vitrified. | 03-06-2014 |
| 20140221720 | Filling Devices, Systems And Methods For Transferring Hazardous Waste Material Into A Sealable Container - The present invention provides systems, methods and devices for storing and/or disposing of hazardous waste material such as calcined material. In certain embodiments, the system comprises a filling nozzle having a valve body having a distal end and an outer surface, the outer surface proximate the distal end being configured to sealingly and removeably couple to an inner surface of a filling port of the container. In certain embodiments, the method comprises (a) coupling an outer surface of a filling nozzle with an inner surface of a filling port of a container to form a first seal (b) adding hazardous waste material into the container (c) decoupling the filling port from the filling nozzle and (d) inserting a fill plug into the filling port, the fill plug forming a second seal with the inner surface of the filling port, the second seal being distally spaced from at least a portion of the first seal with respect to the container. | 08-07-2014 |
| 20150126795 | VITRIFICATION PROCESS METHOD OF ALUMINUM AND FILTER RADIOACTIVE WASTES - Provided is a vitrification process method of aluminum and filter radioactive wastes to produce high quality of glass solidification fit for legislations and rules as vitrification final product, comprising developing frit composition needed in vitrifying the aluminum and filter radioactive wastes, suitably mixing the aluminum and filter radioactive wastes with the frit and producing glass solidification having composition range of oxides of aluminum and filter radioactive wastes to maintain lower than 100 poise viscosity which is operating parameter of a melting furnace. The vitrification process method of aluminum and filter radioactive wastes comprise mixing the aluminum and filter radioactive wastes with the frit in an induction heating low temperature melting furnace and meting it at the temperature of 1,100˜1,200° C. to produce glass solidification. | 05-07-2015 |
| 20160012927 | SYSTEM AND METHOD FOR THE CAPTURE AND STORAGE OF WASTE | 01-14-2016 |
| 20160027544 | Radioactive Waste Solidification Method - A radioactive waste (zeolite to which Cs-137 was adsorbed) in a waste tank and a glass raw material (soda lime glass) in a glass raw material tank are supplied into a solidifying vessel. Graphite in a graphite tank is also supplied into the solidifying vessel. The solidifying vessel is filled with a mixture of the radioactive waste, glass raw material, and graphite and is then disposed in an adiabatic vessel. The radioactive waste and glass raw material in the adiabatic vessel are heated by thermal energy generated due to radiation emitted from Cs-137. The heat is transferred to the peripheral portion of the solidifying vessel through the graphite, raising the temperature of the peripheral portion. The glass raw material is melted and enters clearances among the radioactive waste, producing a vitrified radioactive waste. This radioactive waste solidification method can shorten a time taken to produce a vitrified radioactive waste. | 01-28-2016 |
| 20160141060 | SYSTEM AND METHOD FOR VITRIFICATION OF WASTE - A method for vitrifying waste to prevent the formation of molybdate secondary phases includes forming a feed mixture that includes the waste, a source of vanadium, and at least one of glass frit or glass forming chemicals and vitrifying the feed mixture in a melter to produce a glass product that includes the waste. | 05-19-2016 |
| 588012000 | Boron containing | 1 |
| 20090326312 | METHOD FOR VITRIFICATION OF FISSION PRODUCTS - The mass to be vitrified undergoes a reduction operation in order to have the ruthenium pass from an oxidized state to a metal state in order to reduce the viscosity, the electric conductivity and to obtain good chemical kinetics. | 12-31-2009 |
| 588013000 | Ion exchange material | 2 |
| 20140081067 | SORPTION AND SEPARATION OF VARIOUS MATERIALS BY GRAPHENE OXIDES - Various aspects of the present invention pertain to methods of sorption of various materials from an environment, including radioactive elements, chlorates, perchlorates, organohalogens, and combinations thereof. Such methods generally include associating graphene oxides with the environment. This in turn leads to the sorption of the materials to the graphene oxides. In some embodiments, the methods of the present invention also include a step of separating the graphene oxides from the environment after the sorption of the materials to the graphene oxides. More specific aspects of the present invention pertain to methods of sorption of radionuclides (such as actinides) from a solution by associating graphene oxides with the solution and optionally separating the graphene oxides from the solution after the sorption. | 03-20-2014 |
| 20160141058 | APPARATUS AND METHOD FOR REMOVAL OF NUCLIDES FROM HIGH LEVEL LIQUID WASTES - A method for treating a liquid waste is provided. The method includes supplying the liquid waste to a plurality of cross flow filters from at least one high level waste tank; filtering the liquid waste via the plurality of cross flow filters to form a clarified salt solution; removing at least one radionuclide from the clarified salt solution via a plurality of elutable ion exchange columns filled with an ion exchange media to form an eluate and a decontaminated salt solution; and removing at least one radionuclide from the eluate via a first non-elutable adsorption component to form a dewatered radionuclide sorbent and a decontaminated eluate solution. | 05-19-2016 |
| 588014000 | Silicon containing | 2 |
| 20110224473 | Microwave-Enhanced System for Pyrolysis and Vitrification of Radioactive Waste - Systems and processes for reducing the volume of radioactive waste materials through pyrolysis and vitrification carried out by microwave heating and, in some instances, a combination of microwave heating and inductive heating. In some embodiments, the microwave-enhanced vitrification system comprises a microwave system for treating waste material combined with a modular vitrification system that uses inductive heating to vitrify waste material. The final product of the microwave-enhanced vitrification system is a denser, compacted radioactive waste product. | 09-15-2011 |
| 20130197293 | NANO FLEX HLW/SPENT FUEL RODS RECYCLING AND PERMANENT DISPOSAL - Methods for converting toxic waste, including nuclear waste, to quasi-natural or artificial feldspar minerals are disclosed. The disclosed methods may include converting, chemically binding, sequestering and incorporating the toxic waste into quasi-natural or artificial Feldspar minerals. The quasi-natural or artificial feldspar minerals may be configured to match naturally occurring materials at a selected disposal site. Methods for the immediate, long term, quasi-permanent disposal or storage of quasi natural or artificial feldspar materials are also disclosed. | 08-01-2013 |
| 588015000 | Metal containing | 5 |
| 20110144408 | PROCESS FOR WASTE CONFINEMENT BY VITRIFICATION IN METAL CANS - Process for confinement of waste containing at least one chemical species to be confined, by in-can vitrification in a hot metal can into which waste and a vitrification additive are added, the waste and the vitrification additive are melted to obtain a glass melt which is then cooled, characterised in that at least one oxidising agent is also added into the metal can and in that the concentration of oxidising agent(s) expressed as oxide(s) in the glass melt is between 0.1 and 20% by mass, preferably 4 and 20% by mass, even more preferably 5 and 15% by mass, and even more preferably 10 and 13% by mass of the glass melt mass. | 06-16-2011 |
| 20110306817 | Method For Processing A Nitrous Aqueous Liquid Effluent By Calcination And Vitrification - A method for treating a nitric aqueous liquid effluent containing nitrates of metals or metalloids, comprising a step for calcination of the effluent in order to convert the nitrates of metals or metalloids into oxides of said metals or metalloids, at least one compound selected from the nitrates of metals or metalloids and the other compounds of the effluent leading upon calcination to a tacky oxide, and a dilution adjuvant leading upon calcination to a non-tacky oxide being added to the effluent prior to the calcination step, a method wherein the dilution adjuvant comprises aluminium nitrate and at least one nitrate selected from iron nitrate and rare earth nitrates. | 12-15-2011 |
| 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. | 11-07-2013 |
| 20130296629 | METHOD OF TREATING RADIOACTIVE METAL WASTE USING MELT DECONTAMINATION - Disclosed herein is a method of treating radioactive metal waste using melt decontamination, wherein radioactive metal waste, which is generated from nuclear fuel processing facilities or nuclear fuel production facilities, and which cannot be easily treated by surface decontamination because it has a complicated geometric shape, and the surface contamination of which cannot be measured, can be treated by melt decontamination. The method is advantageous in that radioactive metal waste, which cannot be treated by conventional surface decontamination, can be treated, so that radioactive metal waste can be recycled, thereby obtaining economic profits, and further in that a large storage space necessary for cutting and then storing radioactive metal waste is not required, and in that excessive manpower and cost are not required. | 11-07-2013 |
| 20150332798 | DENSIFIED WASTE FORM AND METHOD FOR FORMING - Materials and methods of making densified waste forms for temperature sensitive waste material, such as nuclear waste, formed with low temperature processing using metallic powder that forms the matrix that encapsulates the temperature sensitive waste material. The densified waste form includes a temperature sensitive waste material in a physically densified matrix, the matrix is a compacted metallic powder. The method for forming the densified waste form includes mixing a metallic powder and a temperature sensitive waste material to form a waste form precursor. The waste form precursor is compacted with sufficient pressure to densify the waste precursor and encapsulate the temperature sensitive waste material in a physically densified matrix. | 11-19-2015 |
