Abstract: The present application relates to a process for the purification of waste materials or industrial by-products, the process comprising the steps of: a) Preparing a composition (C) by blending or mixing waste materials or industrial by-products comprising chlorine (B) with one or more materials comprising heavy metals (HM) b) Reacting (B) and (HM) by thermal treatment of (C) c) Separating evaporated heavy metal chloride compounds (HMCC) d) Obtaining a solid material after the thermal treatment step.

Abstract: The present application relates to a process for the purification of waste materials or industrial by-products, the process comprising the steps of: a) Preparing a composition (C) by blending or mixing waste materials or industrial by-products comprising chlorine (B) with one or more materials comprising heavy metals (HM) b) Reacting (B) and (HM) by thermal treatment of (C) c) Separating evaporated heavy metal chloride compounds (HMCC) d) Obtaining a solid material after the thermal treatment step.

Abstract: The present invention relates to a method of separating and removing a radioactive nuclide, particularly, C-14 and tritium from a radioactive waste resin, and an equipment therefor. The method of treating a radioactive waste resin of the present invention includes recycling of condensate water from which a C-14 radionuclide in the condensate water is removed, into a treatment part for a radioactive waste resin.

Abstract: The present invention provides solidified radioactive waste into which a titanium-containing adsorbent having a radioactive element adsorbed thereto is vitrified, the solidified radioactive waste being capable of confining a large amount of the titanium-containing adsorbent having a radioactive element adsorbed thereto, and furthermore elution of the radioactive element from the vitrified waste being suppressed. The method of the present application includes a step of heat-melting a mixture that includes a titanium-containing adsorbent having a radioactive element adsorbed thereto, a SiO2 source, and an M2O source (M represents an alkali metal element) to form vitrified waste. The titanium-containing adsorbent is preferably one or two or more kind such as silicotitanate, an alkali nonatitanate, and titanium hydroxide.

Abstract: Disclosed is a foam booster that is capable of producing or increasing the amount of foam in any given composition. The foam booster includes a saccharide blend having 30 wt. % to 50 wt. % of an aldohexose or mixture of aldohexoses, 20 wt. % to 55 wt. % of a ketohexose or mixture of ketohexoses; and 10 wt. % to 20 wt. % of a disaccharide or mixture of disaccharides.

Abstract: A reduced-temperature method for treatment of a fuel element is described. The method includes molten salt treatment of a fuel element with a nitrate salt. The nitrate salt can oxidize the outer graphite matrix of a fuel element. The method can also include reduced temperature degradation of the carbide layer of a fuel element and low temperature solubilization of the fuel in a kernel of a fuel element.

Abstract: A circulation pipe of a chemical decontamination apparatus including a malonic acid injection apparatus and an oxalic acid injection apparatus is connected to a purification system pipe, which is made of carbon steel, of a boiling water nuclear power plant. A malonic acid aqueous solution is injected from the malonic acid injection apparatus into the circulation pipe. An oxalic acid aqueous solution is injected from the oxalic acid injection apparatus into the circulation pipe. A reduction decontaminating solution including a malonic acid of 5200 ppm and an oxalic acid within a range of 50 to 400 ppm is supplied into the purification system pipe through the circulation pipe. Reduction decontamination for an inner surface of the purification system pipe is executed. After the reduction decontamination for the purification system pipe finishes, the malonic acid and oxalic acid included in the solution are decomposed and furthermore, the solution is purified.

Abstract: The present invention refers to a method and related plants for removing, by means of redox reactions, the 137Cs from polluted EAF dusts, with an initial average value of radioactivity concentration either higher or lower than 10,000 Bq/kg, the decontamination from the 137Cs initially present in the EAF dusts having a yield of 98%-100%; the present invention also refers to the use of chemical-physical destabilisation agents, by means of redox reactions, for obtaining EAF dusts decontaminated from 137Cs.

Abstract: Embodiments of this disclosure relate to tissue digestion and, more particularly, to methods and apparatuses for varying the number, size, and/or location of one or more tissue compartments within a digestive fluid vessel. Some examples include partitions that may be selectively positioned within a vessel (and may be selectively removable from the vessel) to create one or more tissue compartments. The partitions may be positioned, repositioned and rearranged within the vessel to form one or more tissue compartments. The partitions may be solid or include apertures, and may be oriented in horizontally and/or vertically. Alternate embodiments include one or more selectively closeable apertures that permit digestive fluid to circulate along alternate pathways, which can permit tissue digestion with reduced digestive fluid levels. Still further embodiments include baskets that are selectively positionable within the tissue digester.

Abstract: A decontamination method of solid-state material contaminated by radiocesium comprising bringing the solid-state material containing radiocesium in contact with a first processing solution and preferably eluting cesium ion from the solid-state material to the liquid phase under the presence of potassium ion or ammonium ion.

Abstract: The invention relates to the treatment of carbonaceous radioactive waste, comprising the delivery of waste to one or more radioactive isotope separation stations isotopes, said isotopes being among at least carbon 14, chlorine 36, and tritium. Advantageously, the delivery to each of the stations occurs in wet form, with water being a common medium for conveying the waste to each of the separation stations.

Abstract: Disclosed is a method for treating a radioactive organic waste, the radioactive organic waste including a cation exchange resin adsorbing radionuclide ions, the method including the step of bringing the radioactive organic waste into contact with an organic acid salt aqueous solution containing an organic acid salt and whereby desorbing the radionuclide ions from the cation exchange resin, in which the organic acid salt contained in the organic acid salt aqueous solution includes a cation that is more readily adsorbable by the cation exchange resin than hydrogen ion is. This enables reduction in concentration of a radioactive substance in the radioactive organic waste and reduction in amount of a high-dose radioactive waste.

Abstract: Gas treatment systems and methods are provided. A system includes at least one device defining a space and having a gas inlet and a gas outlet. The device also includes an electrode assembly, where the electrode assembly includes a dielectric plate, at least one first electrode, at least one second electrode, and a conductive layer. The electrodes are elongate electrodes disposed on a first major surface of the dielectric plate and arranged substantially in parallel. Further, the conductive layer extends over a second major surface of the dielectric plate, is electrically coupled to the one of the electrodes, and is electrically isolated from the other electrode. The system includes a circuit configured for generating a pulsed electric field between the electrodes.

Abstract: A method for treating a cladding within which there is a sintered material composed wholly or partly of sintered calcium hydride, the method comprising a step during which the sintered material is contacted with a reaction mixture comprising steam, carbon dioxide, and a chemically inert gas, the contacting being carried out for a duration allowing the sintered calcium hydride to be converted into a calcium carbonate powder. The treatment method provided by the invention results in a chemically inert waste, a limitation on the volume of wastes obtained, while allowing the removal or even recovery of these wastes by appropriate processing streams.

Abstract: A chemical decontamination reagent containing a reducing agent, a reductive metal ion, and an inorganic acid is provided to remove a radioactive oxide layer on a metal surface. The reagent can dissolve the radioactive oxide layer on the metal surface effectively at a relatively low temperature and enables a simple process of contacting the reagent to the radioactive oxide, thus economically effective in terms of cost and time required for the process. Since the decontamination does not use a conventional organic chelating agent such as oxalic acid, but the reducing agent as a main substance, the residuals of the reducing agent remained after decontamination can be decomposed and removed with an oxidizing agent. Due to the easy decomposition with the chemical decontamination reagent, secondary wastes can be minimized and the radionuclides remained in the decontamination reagent solution can be removed effectively.

Abstract: A radioactive waste acid is recycled. The waste acid is a fluoboric waste acid. The waste acid contains a lot of oxides and radioactive nuclei. The waste fluoboric acid is processed to obtain a purified fluoboric acid. The amount of radioactive nuclei is greatly reduced. Thus, the present invention has a simple procedure with low cost and reduced power consumption.

Abstract: It is an object of the present invention to provide a beneficial method for detoxifying a harmful compound to detoxify the harmful compound containing arsenic etc., effectively. The method of detoxifying a harmful compound according to the present invention is characterized in that a methyl radical and/or a carboxymethyl radical is (are) contacted with a harmful compound comprising at least one element selected from the group comprising arsenic, antimony and selenium to detoxify the harmful compound. Furthermore, in a preferred embodiment of the method of detoxifying a harmful compound according to the present invention, the method is characterized in that the radical is generated by the exposure to light.

Abstract: The present invention relates to a vacuumable gel and the gel use to decontaminate surfaces, for example, radioactive decontamination. The gel is composed of a colloidal solution comprising: from 5 to 25 wt % of an inorganic viscosity modifier; from 0.01 to 0.2 wt % of a surfactant, preferably, a surfactant in an amount strictly below 0.1 wt %, wt % relative to the total weight of the gel; from 0.5 to 7 mol, per liter of gel, of an inorganic acid or base; and optionally from 0.05 to 1 mol, per liter of gel, of an oxidizer having a standard redox potential Eo greater than 1.4 V in a strong acid medium or of the reduced form of this oxidizer; the remainder being water. The gel may be applied, by spraying, to a surface to be decontaminated, and removed in the form of dry residues by suction or brushing after drying.

Abstract: The present disclosure relates to a decontamination method and apparatus for cladding hull wastes of spent nuclear fuels, capable of decontaminating a small quantity of spent nuclear fuels remaining on surfaces of the cladding hull wastes and radioactive fission products penetrated into the cladding hulls through an electrochemical dissolution. The method includes inserting the cladding hull waste into an anodic basket, immersing a reference electrode and a cathodic electrode as well as the anodic basket in a molten salt, dissolving a surface of the cladding hull waste by applying a voltage or current to the anodic basket with respect to the cathodic electrode or the reference electrode, removing the anodic basket, and removing a salt remaining on the surface of the cladding hull waste.

Abstract: The present invention relates to a process for recycling SPL from primary aluminum production, comprising the steps of grinding the separated fractions or cuts, separating materials by means of mechanical treatment, mixing salt slags with SPL, dissolving in water the product obtained in the previous step, carrying out a chemical reaction between water and the materials to be made inert, sedimenting, removing cyanides, filtering to obtain a soluble fraction and another insoluble fraction, washing the insoluble fraction, crystallizing salts of the soluble fraction and aging or conditioning the insoluble fraction. The present invention further relates to the product obtained by means of said process and the use thereof in different fields of the art.

Abstract: A method for precipitating uranium from an aqueous solution and/or sediment comprising uranium and/or vanadium is presented. The method includes precipitating uranium as a uranyl vanadate through mixing an aqueous solution and/or sediment comprising uranium and/or vanadium and a solution comprising a monovalent or divalent cation to form the corresponding cation uranyl vanadate precipitate. The method also provides a pathway for extraction of uranium and vanadium from an aqueous solution and/or sediment.

Abstract: Methods and compositions to extract radionuclides such as various actinides and lanthanides from organic and/or aqueous solutions by utilizing extractant functionalized carbon nanotubes are disclosed. More particularly, phosphorous-containing (such as phosphine oxides, phosphoric acids or phosphates) organic extractants and other predesigned extractants (such as crown ethers, calncrown derivatives, malonamide and diglycolamide derivatives, polyethylene glycol derivatives, cobalt dicarbollide derivatives, and N-donating heterocyclic ligands) can be covalently and/or non-covalently employed on the surfaces and/or ends (tips) of carbon nanotubes for the purpose of removal radionuclides such as various actinides and lanthanides from organic and/or aqueous solutions. Extractant functionalized carbon nanotubes can be used for extracting radioactive nuclides from nuclear waste or spent nuclear fuel, which are produced and/or reprocessed from the nuclear power generation or other nuclear application.

Abstract: The method of dissolving the solids formed in the apparatus and pipework of a nuclear plant, in which the solids are brought into contact with an aqueous dissolving solution chosen from aqueous solutions of carbonate ions having a concentration of greater than or equal to 0.3M, aqueous solutions of bicarbonate ions, and solutions of a mixture of nitric acid and of a polycarboxylic acid chosen from oxalic acid and triacids.

Abstract: An apparatus and system for separating remaining powder of hulls includes a first remaining powder separating unit to be supplied with hulls of a spent nuclear fuel subjected to a high-temperature oxidation, and to include a first brush for separating remaining powder of the hulls; a hull alignment unit to be supplied with the hulls from the first remaining powder separating unit, and to align the hulls; a second remaining powder separating unit to be supplied with the aligned hulls from the hull alignment unit, and to include a second brush for separating remaining powder adhered on an inner peripheral surface of the hulls; and a third remaining powder separating unit to be supplied with the hulls from the second remaining powder separating unit, and to separate the remaining powder remaining on the inner/outer peripheral surface of the hulls by using air.

Abstract: The present invention relates to a method and apparatus for solidifying radioactive waste accompanying chlorine recycling or radioactive iodine removal. The present invention provide a method and apparatus for treating radioactive waste composed of volatile metal halide with high corrosiveness at a high temperature through a solidifying system including crushing, mixing/dehalogenating and sintering. In the inventive method and apparatus for solidifying radioactive waste, a dechlorination process by using a dehalogenation material is performed on highly corrosive metal salt under an oxidizing atmosphere by using a method of controlling a temperature on the surface of a reactor, and solidification is carried out by removing radioactive iodine while reusing chlorine contained in an exhaust gas. Accordingly, the generated amount of total waste can be reduced, and therefore the inventive method and apparatus can be usefully used in preparing a solidified body with high durability.

Abstract: Described are method of treating a radioactive organic waste stream comprising: (a) mixing a radioactive organic waste stream comprising organic compounds and radionuclides with phosphoric acid to form a reaction mixture; (b) heating the reaction mixture to a desired temperature in the presence of an oxidant to oxidize organic compounds present in the waste stream, and removing oxidized organic compounds from the reaction mixture; (c) optionally, adding a reducing agent to the reaction mixture to form insoluble radioactive metal phosphate compounds comprising one or more of the radionuclides, and separating the insoluble radioactive metal phosphate compounds from the reaction mixture; (d) optionally, adding a fluorine compound to the reaction mixture to react with uranium that may be present in the reaction mixture to form uranium hexafluoride, and removing uranium hexafluoride from the reaction mixture; (e) adding ammonia to the reaction mixture to neutralize phosphoric acid and to form ammonium phosphate

Abstract: A method of processing ion exchange resin radioactive waste, wherein the radioactive waste contains a plurality of fractions of radioactive waste based on density, which may include cation resin waste and anion resin waste, wherein at least one of the plurality of fractions comprises a total concentration of at least one radionuclide, including performing at least one of the following operations (i), (ii) or (iii): (i) separating the at least one of the plurality of fractions from the radioactive waste utilizing a moving freeboard; (ii) separating the radioactive waste into cation resin waste and anion resin waste; or (iii) removing at least a portion of the total concentration of the at least one radionuclide from: the radioactive waste; the at least one of the plurality of fractions separated by the moving freeboard; or at least one of the radioactive waste, the cation resin waste or the anion resin waste.

Abstract: Methods for removing nuclear waste from a component are provided. In one embodiment, the method includes the steps of supplying oxalic acid to the nuclear waste on the component to form an oxalic acid/waste solution, feeding the oxalic acid/waste solution from the component, feeding an oxidant to the oxalic acid/waste solution to form carbon-dioxide, water and a precipitate, and separating the precipitate from the water. A system for removing nuclear waste is also provided.

Abstract: An apparatus for synergistically combining a plasma with a comminution means such as a fluid kinetic energy mill (jet mill), preferably in a single reactor and/or in a single process step is provided by the present invention. Within the apparatus of the invention potential energy is converted into kinetic energy and subsequently into angular momentum by means of wave energy, for comminuting, reacting and separation of feed materials. Methods of use of the apparatus in the practice of various processes are also provided by the present invention.

Abstract: The invention relates to the treatment of carbon-containing radioactive waste. It particularly envisages: a first type of waste treatment to obtain a carbon oxide, and a second type of treatment to obtain a solid precipitate of the carbon oxide by reacting with a selected element. A method according to the invention comprises: a first phase during which both the first and second type of treatment are applied, and a second phase during which only the first type of treatment is applied.

Abstract: A method for restraining a chemical discharge comprising (a) deploying a binding agent into a receptacle containing a hazardous material in a liquid state upon the occurrence of at least one predetermined event that increases the risk of accidentally discharging or leaking the hazardous material from the receptacle; and (b) contacting the hazardous material with the binding agent to form a composition comprising at least a portion of the hazardous material and the binding agent and having at least one property selected from a solid or semisolid state, a viscosity greater than the viscosity of the hazardous material at ambient conditions, a vapor pressure lower than the vapor pressure of the hazardous material at ambient conditions, and a surface tension greater that the surface tension of the hazardous material.

Abstract: The invention concerns a process enabling the complete combustion and oxidation of the mineral fraction of combustible waste contained in an apparatus intended to treat waste by direct incineration-vitrification, said process comprising the following steps: a step to add said waste to the apparatus for its depositing on the surface of a molten glass bath contained in the apparatus, an incineration and oxidation step of the waste on the surface of the glass bath, an incorporation step to incorporate combustion products in the glass during which the glass bath, the combustion products and any vitrification additives added to the glass bath are heated until a paste-like, liquid mass is obtained, a step during which said mass is removed from the apparatus and left to cool to obtain finally what is called a confinement matrix, said process being characterized in that the complete combustion and oxidation of the waste is achieved partly during the waste adding step and partly during the waste incineration and ox

Abstract: A reuse apparatus of eutectic salt waste produced in an electro refining process and a method thereof is a technology that in order to collect the eutectic salt of the eutectic salt waste, oxidizes/precipitates nuclides (rare earth and TRU) within the eutectic salt waste, an oxygen dispersing method is used to perform a layer separation into the eutectic salt layer and the precipitate layer. Then, the precipitate layer and eutectic salt layer are separated and collected, so that the eutectic salt layer is directly reused and the eutectic salt within the precipitates is reused by separating and collecting it using distillation/condensation processes. The reuse apparatus of the eutectic salt waste and a method thereof thereby increases the collecting efficiency of the eutectic salt and allows the compositions of the collected eutectic salt to have the same compositions as the eutectic salt used in the electro refining process.

Abstract: The invention relates to a process for converting an alkaline-earth metal chloride to at least one salt chosen from the tungstates and molybdates of this metal, which comprises the reaction of the alkaline-earth metal chloride with at least one precursor of tungsten or of molybdenum chosen from tungsten oxides, molybdenum oxides, tungstates and molybdates, this reaction being carried out in a solvent constituted by KCl or by an LiCl/KCl mixture and at a temperature at least equal to the melting point of this solvent. Applications: synthesis of alkaline-earth metal tungstates and molybdates, extraction of alkaline-earth metals from media in which they are found in the form of chlorides, recovery of alkaline-earth fission products from a salt flux in the context of the reprocessing of a spent nuclear fuel in molten chloride media, etc.

Abstract: The invention relates to the treatment of carbon-containing radioactive waste. It particularly envisages: a first type of waste treatment to obtain a carbon oxide, and a second type of treatment to obtain a solid precipitate of the carbon oxide by reacting with a selected element. A method according to the invention comprises: a first step during which both the first and second type of treatment are applied, and a second step during which only the first type of treatment is applied.

Abstract: The current methods of removing the surface-deposited layer contaminated by radioisotopes has the disadvantage of rigorous applicability conditions in such terms as the substance to be treated, the area and the shape, so the scope of their applicability is quite limited; in addition, 100% decontamination is not guaranteed and in spite of more than 50% that can be decontaminated, a greater part of the deposited radioisotopes remain unremoved, limiting the decontamination performance of the methods. Parts, structural components, etc. that have their surfaces contaminated by radioisotopes in the accelerators, nuclear reactors, RI product manufacturing factories, nuclear fuel factories, nuclear fuel reprocessing factories, etc.

Abstract: Disclosed herein is provided an evaporator/calciner in which hazardous materials, such as radioactive liquids, are converted into chemically stable, solid forms by evaporating, drying and calcination within a single vessel, that can then be sealed and used for long term storage.

Abstract: The invention relates to a composition which makes it possible to obtain a gelled aqueous foam capable of decontaminating, stripping or degreasing a surface. The composition of the present invention comprises one or more surfactants, one or more acidic or basic reactants and a gelling agent. The decontamination foam obtained from this composition exhibits long lifetimes, generally of between 1 and 10 hours, guaranteeing a prolonged time of action on the surface and a high decontamination effectiveness. These foams can be used to remove the radioactivity from an inaccessible plant, which is large in size and complex in design, by simple filling or by simple spraying over an accessible surface.

Abstract: The invention concerns uranium-chelating peptides as well as their uses for decontaminating soils and water, and for detecting and treating people contaminated by uranium. Said peptides have a helix-loop-helix type structure comprising the sequence of a calmodulin loop including at least one mutation of neutral residues selected from the group consisting of S, T, C, H, Y, N and Q, of one, two or three residues of at least one of the four calmodulin calcium binding sites: site I: residues selected among D20, D22 and D24 residues; site II: residues selected among D56, D58 and N60 residues; site III: residues selected among D93, D95 and N97 residues; site IV: residues selected among D129, D131 and D133 residues; said positions being indicated with reference to the human calmodulin sequence.

Abstract: A system for chemically decontaminating radioactive material.

Abstract: The invention relates to a method for separating uranium(VI) from one or more actinides selected from actinides(IV) and actinides(VI) other than uranium(VI), characterized in that it comprises the following steps: a) bringing an organic phase, which is immiscible with water and contains the said uranium and the said actinide or actinides, in contact with an aqueous acidic solution containing at least one lacunary heteropolyanion and, if the said actinide or at least one of the said actinides is an actinide(VI), a reducing agent capable of selectively reducing this actinide(VI); and b) separating the said organic phase from the said aqueous solution. Applications: reprocessing irradiated nuclear fuels, processing rare-earth, thorium and/or uranium ores.

Abstract: The present invention relates to the use of a mixed carbonate of formula AB(CO3)2, in which A and B are different and chosen from alkali metals, alkaline-earth metals and rare earths, for the containment of radioactive carbon. This use may for example involve a process comprising: mixing CO2 having a radioactive carbon to be contained, or a simple carbonate of an alkali, alkaline-earth or rare-earth metal having a radioactive carbon to be contained, with an aqueous solution of a mixture of ACln and BClm or with an aqueous solution of a mixture of A(OH)n, and B(OH)m in order to obtain a precipitate of AB(CO3)2, where n and m are integers sufficient to compensate for the charge of A and B respectively; recovery of the AB(CO3)2 precipitate in powder form; and then pressing and sintering of the powder at a 20 temperature below the decarbonation temperature of the mixed carbonate manufactured in order to obtain sintered pellets of mixed carbonates for the containment of the radioactive carbon.

Abstract: Methods and compositions to extract radionuclides such as various actinides and lanthanides from organic and/or aqueous solutions by utilizing extractant functionalized carbon nanotubes are disclosed. More particularly, phosphorous-containing (such as phosphine oxides, phosphoric acids or phosphates) organic extractants and other predesigned extractants (such as crown ethers, calncrown derivatives, malonamide and diglycolamide derivatives, polyethylene glycol derivatives, cobalt dicarbollite derivatives, and N-donating heterocyclic ligands) can be covalently and/or non-covalently employed on the surfaces and/or ends (tips) of carbon nanotubes for the purpose of removal radionuclides such as various actinides and lanthanides from organic and/or aqueous solutions. Extractant functionalized carbon nanotubes can be used for extracting radioactive nuclides from nuclear waste or spent nuclear fuel, which are produced and/or reprocessed from the nuclear power generation or other nuclear application.

Abstract: A method of suppressing deposition of radionuclides on components of a nuclear power plant comprises forming a ferrite film by contacting a first chemical including iron (II) ions, a second chemical for oxidizing the iron (II) ions to iron (III) ions, and a third chemical for adjusting the pH of a processing solution containing a mixture of the first and second chemicals to be 5.5 to 9.0 with the metal member surface in a time period from a finishing stage in decontamination step of removing contaminants formed on the surface of metal member composing the nuclear power plant, and suppressing deposition of radionuclides on the metal member by the ferrite film.

Abstract: A reuse apparatus of eutectic salt waste produced in an electro refining process and a method thereof is a technology that in order to collect the eutectic salt of the eutectic salt waste, oxidizes/precipitates nuclides (rare earth and TRU) within the eutectic salt waste, an oxygen dispersing method is used to perform a layer separation into the eutectic salt layer and the precipitate layer. Then, the precipitate layer and eutectic salt layer are separated and collected, so that the eutectic salt layer is directly reused and the eutectic salt within the precipitates is reused by separating and collecting it using distillation/condensation processes. The reuse apparatus of the eutectic salt waste and a method thereof thereby increases the collecting efficiency of the eutectic salt and allows the compositions of the collected eutectic salt to have the same compositions as the eutectic salt used in the electro refining process.

Abstract: A recycling system for a waste converting apparatus collects residues from a post-processing means and re-introduces the residues into the apparatus such that the residues are exposed to the high temperature zone thereof.

Abstract: The present invention relates to a vacuumable gel that can be used for decontaminating surfaces, and also to the use of this gel. The decontamination may be, for example, a radioactive decontamination. The gel of the present invention is composed of a colloidal solution. It comprises from 5 to 25 wt % of an inorganic viscosity modifier relative to the total weight of the gel; from 0.01 to 0.2 wt % of a surfactant relative to the total weight of the gel, and, particularly preferably, a surfactant in an amount strictly below 0.1 wt % relative to the total weight of the gel; from 0.5 to 7 mol, per litre of gel, of an inorganic acid or base; and optionally from 0.05 to 1 mol, per litre of gel, of an oxidizer having a standard redox potential E0 greater than 1.4 V in a strong acid medium or of the reduced form of this oxidizer; the remainder being water. It may be applied, by spraying, to a surface to be decontaminated, and removed in the form of dry residues by suction or brushing after drying.

Abstract: An improved radioactive waste (RW) processing is usable, e.g., for reducing radioactivity of wastes produced by nuclear plants. The processing comprises an entry and a base stages. The entry stage is exemplified either by a “standard” option or an “own” option encompassing the grinding RW and obtaining a pulp. Both the options provide technological mixture for the base processing, which encompasses intermingling the mixture with predetermined substances, kneading and filtering the mixture, removing wastes, processing the mixture in either a regular or in a special dissolver, or in both. The resulted product is subjected to rotation in a special centrifuge utilizing the Searl effect that allows further reducing the RW radioactivity. The so treated RW is then packed into specially constructed glass containers, or compressed and covered by special glass-mass. Composition and processing instructions for the substances and glass-mass, and the constructions of the special dissolver and centrifuge are disclosed.

Abstract: A method (and concomitant kit) for treating a surface to reduce subsequent 137Cs nuclide desorption comprising contacting the surface with a first cation-containing solution, the cation being one or more of Cs+, Rb+, Ag+, Tl+, K+, and NH4+, and contacting the surface with a second cation-containing solution, the cation being one or more of Cs+, Rb+, Ag+, Tl+, K+, and NH4+, thereby reducing amounts of radioactive cesium embedded in clays found on the surface.

Abstract: The invention relates to a process for the preparation of a product based on a phosphate of at least one element M(IV), for example of thorium and/or of actinide(IV)(s). This process comprises the following stages: a) mixing a solution of thorium(IV) and/or of at least one actinide(IV) with a phosphoric acid solution in amounts such that the molar ratio PO 4 M ? ? ( IV ) ?is from 1.4 to 2, b) heating the mixture of the solutions in a closed container at a temperature of 50 to 250° C. in order to precipitate a product comprising a phosphate of at least one element M chosen from thorium(IV) and actinide(IV)s having a P/M molar ratio of 1.5, and c) separating the precipitated product from the solution. The precipitate can be converted to phosphate/diphosphate of thorium and of actinide(s). The process also applies to the separation of uranyl ions from other cations.