Method of processing fluorination cinder
FIELD: physics, atomic power.
SUBSTANCE: invention relates to methods of processing uranium-containing solutions obtained from dissolving fluorination cinder when producing uranium hexafluoride. The method includes dissolving cinder in nitric acid solution, extracting uranium from the fluorine-containing nitric acid solution via reduction thereof with hydrazine on a platinum catalyst, with constant removal of uranium tetrafluoride precipitate from the catalyst surface, separating the catalyst from the nitric acid solution and the uranium tetrafluoride precipitate, providing an equimolar radio of fluoride ions to uranium (IV) in the obtained solution and separating the uranium tetrafluoride precipitate from the nitric acid solution, wherein the nitric acid solution is reused to resolve fluorination cinder with preliminary fortification with nitric acid.
EFFECT: invention provides a high degree of reducing uranium and reduces the amount of nitrate- and fluorine-containing wastes.
3 cl, 2 dwg
The invention relates to a method of refining uranium-fluorine-containing solutions obtained from dissolution Ogarkov fluoridation in the production of uranium hexafluoride.
Processing Ogarkov fluoridation containing oxychoride uranium and fluoride impurity elements, aims at the regeneration of uranium, which can be carried out, for example, by dissolving Ogarkov fluoridation in nitric acid and subsequent processing of uranium-fluorine speed by separating its uranium deposition, or extraction.
A method of refining uranium-fluorine turnovers extraction-precipitation scheme with pre-concentrated uranium by precipitation in the form of polyuranates sodium or ammonium [Tinin centuries, Balakhonov Century, Dorda F. A., Lazarchuk centuries, Ledovskikh A. K., matyukha C. A., Portnyagin E. O. Processing of uranium-fluorine-containing sublimation speed of production. Innovation in the nuclear industry: problems and solutions: Materials of scientific-practical conference of students, postgraduates and young scientists 27-30 November 2006, , Seversk: Ed. SGT, 2006, S. 24]. The disadvantage of this method is that the fluorine-containing mother liquor cannot be reused by dissolving the uranium-fluorine momentum due aciclovi fluorine.
There is a method of recovery of uranium (VI) with hydrazine nitrate solutions using latinovich catalysts [B. C. Turowski, Century, Balakhonov, Y. C. Boer, C. A. Matyukha. Influence of method of preparation of platinum catalysts on ionite media on the efficiency of recovery of uranium (VI) with hydrazine nitrate solutions // Bulletin of the Tomsk Polytechnic University. 2010. So 316. No. 3, S. 39-43]. This method includes the introduction in the nitric acid solution of uranyl nitrate containing hydrazine, platinum catalyst. The catalyst provides for the recovery of uranium (VI) with hydrazine nitrate solutions to the tetravalent state. As a carrier of platinum used anion-exchange resin. The optimum content of platinum is 4-6 wt.%. This method allows for 2 hours in a solution of nitric acid at 42-60% to recover the uranium (VI) with hydrazine on the platinum catalyst.
The disadvantages of the method:
- low degree of recovery of uranium (VI);
when the fluorine-containing processing solutions, the catalyst loses its catalytic activity, because its surface is closed from solution by precipitation.
There is a method of processing highly enriched uranium [RF patent №2112744, IPC C01G 43/00, publ. 10.06.1998] (prototype), in which the stubs fluoridation is dissolved in a solution of nitric acid, uranium is extracted by the extraction and after denitration of uranyl nitrate obtained octaoxide truran return on fluoridation. In the refined about the sought fluoride and nitrate ions.
The disadvantages of the method:
- large volume of liquid radioactive waste generated during the processing of fluoride and nitrosoureas refined. Neutralization of refined lime slurry [Harrington H, Ruelle A. production Technology of uranium. - M.: gosatomizdat, 1961, S. 184] with the separation of the neutral slurry of solid particles of calcium fluoride and the clear liquor into the open drainage system leads to the pollution of waters by nitrates.
The objective of the invention is to reduce the number of nitrate-containing fluorine waste generated during reprocessing uranium-fluorine-containing solutions obtained from dissolution Ogarkov fluoridation in the production of uranium hexafluoride.
The set task is solved by the fact that in the method of processing Ogarkov fluoridation, including the dissolution of Ogarkov in the nitric acid solution, the extraction of uranium from nitric acid solution, the uranium restore hydrazine in the fluorinated nitric acid solution on a platinum catalyst at a constant cleaning of the surface of the catalyst precipitation of uranium tetrafluoride. The catalyst is separated from the nitric acid solution and the precipitate of uranium tetrafluoride. Provide at least equimolar ratio of fluoride ions to uranium (IV) in the resulting solution, separate the precipitate of uranium tetrafluoride and the nitric acid solution, which, after elodokere by nitric acid re-used for dissolving Ogarkov fluoridation. Cleaning the surface of the catalyst precipitation of the uranium tetrafluoride is carried out by mechanical mixing with a catalyst or transmission ornitologijas nitrate solution through the catalyst bed in the catalytic column.
In Fig.1 presents the dependence of the growth in the number of sediment tetrafluoride uranium as the duration of the recovery process; Fig.2 - changes in the concentration of uranium (VI) in solution in the recovery process.
The method is as follows.
Spent catalytic reduction of uranium (VI) with hydrazine on the platinum catalyst (6% Pt on anion-exchange resin A-100 particle size of 0.25-0.50 mm) in nitric acid solution obtained from dissolution Ogarkov fluoridation containing U(VI) - 0.25 mol/l HNO3to 2.0 mol/l, HF - 1.5 mol/l and N2H4- 1.0 mol/l Recovery was conducted at 60°C. the Volume ratio of the catalyst to the solution was 1:10. In the process of recovering the catalyst is mechanically mixed with a solution. After two hours the solution with the precipitate and the catalyst was moved to the strainer. After filtering the grid remained the catalyst, and the nitric acid solution to precipitate the uranium tetrafluoride (in the form of double salts with hydroinform) left in the filtrate. The precipitate of uranium tetrafluoride and the nitric acid solution was separated by filtration. Isotonik the second solution is returned to the dissolution Ogarkov fluoridation, pre darkrai him on the concentration of nitric acid, closing its loss due to the formation of nitrogen oxides in the process of dissolution Ogarkov fluoridation.
It should be noted that the recovery of uranium (VI) in solutions of HF was accompanied by the growth of the precipitate of uranium tetrafluoride (see Fig.1), and the appearance in the sludge solution did not reduce the catalytic activity of the catalyst. Preservation of catalytic activity of the catalyst was ensured effective cleaning of its surface from the precipitate by mixing the catalyst solution (in additional experiments it was found that the catalyst surface is cleared of sediment UF4and the flow of the solution, if the recovery of uranium is carried out in a catalytic column by passing the solution through a layer of catalyst).
It is evident from Fig.2 shows that the content of uranium (VI) in solution during the first hour decreased from 62,9 g/l to 6.2 g/l, i.e. the degree of recovery of uranium (VI) to tetravalent state was 90.1%; after two hours the content of uranium (VI) in the solution was equal to 2.2 g/l is the degree of recovery of 96.5%. For comparison conducted the recovery of uranium (VI) under the same conditions, but used the nitric acid solution of uranyl nitrate without fluoride ions. The result, shown in Fig.2 shows that the recovery of uranium (VI) in the absence of fluoride ionospheric a much lower rate: two hours recovered 42.2% of uranium (VI).
The results presented in Fig.2, allow us to conclude that the presence of fluoride ions can be almost fully recovered uranium (VI) and put it in the precipitate of uranium tetrafluoride. In the case of fluoride ions in the solution obtained from dissolution Ogarkov fluoridation is not sufficient for full recovery of uranium (VI) in the solution after separation from the catalyst is further added fluoride ions, at least equimolar with respect to uranium (IV) number.
Thus, the catalytic reduction of uranium by hydrazine in the fluorinated nitric acid solution provides a high degree of recovery of uranium (VI) and its transfer in the precipitate of uranium tetrafluoride and allows you to reuse the filtered nitrate solution by dissolving operation Ogarkov fluoridation, thereby reducing the amount of nitrate-containing fluorine waste.
1. A method of processing Ogarkov fluoridation, including the dissolution of Ogarkov in the nitric acid solution, the extraction of uranium from the fluorinated nitric acid solution, characterized in that the uranium restore hydrazine in the fluorinated nitric acid solution on a platinum catalyst at a constant cleaning of the surface of the catalyst precipitation of the uranium tetrafluoride, the catalyst is separated from the nitric acid solution and the precipitate of tetr the fluoride of uranium, provide at least equimolar ratio of fluoride ions to uranium (IV) in the resulting solution, separate the precipitate of uranium tetrafluoride and the nitric acid solution, which, after its doreplace by nitric acid re-used for dissolving Ogarkov fluoridation.
2. The method according to p. 1, characterized in that the cleaning of the surface of the catalyst precipitation of the uranium tetrafluoride is carried out by mechanical mixing with a catalyst.
3. The method according to p. 1, characterized in that the cleaning of the surface of the catalyst precipitation of the uranium tetrafluoride is carried out by passing ornitologijas nitrate solution through the catalyst bed in the catalytic colon.
SUBSTANCE: method includes neutralising complexing ligands contained in nitrate solution of carbide fuel via oxidation thereof with nitric acid in the presence of a catalyst in the form of a polyvalent metal which is located in the nitrate solution or is added before or after dissolving spent carbide nuclear fuel, said metal being selected from: cerium, iron, manganese, technetium, mercury. Further, the method includes heating the nitrate solution of the carbide fuel or performing oxidation directly in the process of dissolving the carbide fuel in nitric acid in the presence of a catalyst with subsequent dissolution in the oxidised solution of the carbide fuel of the oxide or metallic spent nuclear fuel, or performing simultaneous oxidation of complexing ligands and dissolving the oxide or metallic spent nuclear fuel in the solution of the carbide fuel. An alternative solution includes performing similar preparation of the spent carbide nuclear fuel for extraction processing, followed by mixing the oxidised solution of the carbide fuel with solutions of oxide or metallic spent nuclear fuel or direct addition of the required amount of a zirconium nitrate solution or a solution of another polyvalent metal-complexing agent.
EFFECT: avoiding the need to use powerful oxidising agent.
13 cl, 12 ex
FIELD: physics, atomic power.
SUBSTANCE: present invention relates to means of measuring burn-up of spent fuel assemblies of thermal neutron reactors. A diagnostic container is placed at the bottom of a cooling pond under water. The wall of the housing has an annular cavity with a liquid tracer substance surrounded by polypropylene layers and steel layers, as well as a cadmium layer. The housing has a centre cavity in which spent fuel assemblies are placed. The container is closed with a cover. Water is removed from the centre cavity with the spent fuel assemblies. The tracer substance is activated, drained into a laboratory vessel and stirred. A sample is then collected. The average specific activity of the tracer substance is measured and the neutron radiation intensity of the spent fuel assemblies and the associated burn-up of the spent fuel assemblies is determined. The annular cavity with the liquid tracer substance can consist of multiple rings insulated from each other. Burn-up is then determined for each ring and a burn-up profile for the spent fuel assemblies is then constructed.
EFFECT: enabling all-round encircling of the core region of the spent fuel assemblies with the tracer substance, excluding the effect of water on measurement accuracy, eliminating background effect when measuring specific activity of the tracer substance and high accuracy of determining burn-up of spent fuel assemblies of thermal neutron reactors without retrieving spent fuel assemblies from the cooling pond.
9 cl, 3 dwg
SUBSTANCE: oxides of transuranium elements are mixed with palladium metal powder in the following ratio, wt %: oxides of transuranium elements - 30-70, palladium metal - 70-30, and the obtained mixture is pressed. As a result, a composition for long-term storage of transuranium elements is obtained, which includes oxides of transuranium elements in palladium metal, which provides high chemical stability of the material, safety during indefinite storage, while preserving the capacity to extract transplutonium elements after dissolving the disclosed composition in nitric acid. The invention proposes the use of industrial ("reactor") palladium, which is a nuclear fuel fission product, to produce the disclosed composition.
EFFECT: prolonged and reliable isolation of transuranium elements and preserving the capacity for extraction and use thereof in future, or for further processing using a transmutation process.
2 cl, 2 tbl
SUBSTANCE: invention relates to materials with neutron-absorbing properties for protection against neutron radiation. Claimed is a fire-resistant neutron-protective material, consisting of a magnesium phosphate binding agent (24-33 wt %) and a powder part (76-67 wt %, with the powder part containing titanium hydride TiH2 (90.3-95.5 wt %), magnesium oxide MgO (2.7-4.5 wt %) and boron carbide B4C (1.8-5.2 wt %). Components are mixed to a homogeneous state and poured into a special cavity, and after hardening are subjected to thermal processing.
EFFECT: obtained material possesses long-term mechanical strength, heat resistance to ≈300°C, high heat conductivity, a temperature coefficient of linear expansion, close to the coefficient of construction steels, and high specific density of hydrogen and boron, contained in it, which ensures high coefficients of neutron radiation attenuation.
SUBSTANCE: invention relates to radiochemical technology and can be used in production of "reactor" 99Mo as a generator of 99mTc of a biomedical purpose, as well as in an analysis of technological solutions for preliminary separation of Mo or Mo and Zr in extraction reprocessing of solutions of technology of spent nuclear fuel of nuclear power plants (NPP SNF). Described are versions of methods of selective extractive separation of a considerable part of molybdenum or together molybdenum and zirconium from radioactive solutions with obtaining an extract. A reprocessed radioactive solution is processed with an extractant, which represents poorly soluble in a water phase alcohol, in the presence of an extracted complexing agent. As the complexing agent, hydroxamic acids with a number of carbon atoms 6-12 can be used, which ensures sufficiently complete extraction of molybdenum and zirconium in an organic phase. Molybdenum or molybdenum and zirconium are separated from the extract in the compact form by sublimation or re-extraction.
EFFECT: obtaining the extract, purified from alpha- and gamma-radioactive admixtures more than by 100 times, and further separate extraction of radionuclides from the extract, combined in the final stage of the process with the extractant regeneration.
17 cl, 2 tbl, 12 ex
SUBSTANCE: invention relates to method of determining optimal parameters of dissolution of oxides of transition metals in solutions, which contain complexing agent, and can be applied in atomic energy. As parameters, applied are volume coefficients of distribution of radioactive isotopes of transition metals, which determine composition of oxides, between solutions, which contain complexing agent, and strong-base anionites in form of said complexing agent (complexits) and balanced values of solution pH. Radioactive isotopes of transition metals are introduced into fixed volumes of analysed solutions, after which, fixed volumes of complexits are introduced into solutions. Initial and final activity of solutions is measured. Also measured are balanced values of pH of solutions. Ranges, satisfying optimal parameters, which must be supported in contours of NEP directly in the process of dissolution of oxides in the process of washing and deactivation of NEP contours, are determined by the results of measurements.
EFFECT: increased reliability of determination of optimal parameters of dissolution of oxides of transition metals in complexing agent solutions and absence of necessity of carrying out complex analyses of metals to determine concentration of metal cations in solutions of complexons.
2 cl, 2 dwg, 1 tbl
FIELD: power engineering.
SUBSTANCE: machine for cutting of pipelines, preferably caissons from a storage tank, comprises a body and a rotor covering the cut pipe, with a metal cutting device fixed on it. The machine body is made in the form of a circular ball step-bearing, the fixed part of which is attached to the rotary board of the storage tank together with the drive for rotation of the movable part, coaxially and elastically fixed to the rotor flange. The rotor is made in the form of a cylindrical thin-walled hollow column that covers a caisson, with a rotary stem fixed on it, at the end of which a small-size driving cutting machine is attached with an abrasive disc.
EFFECT: creation of a reliable and comparatively cheap machine for cutting and trimming of pipelines, in particular, caissons in unavailable zones of radiation-hazardous facilities at the depth of at least 3 metres from the outer surface of the facility to provide for nuclear and radiation safety.
4 cl, 7 dwg
FIELD: power engineering.
SUBSTANCE: method includes construction of inspection wells in a reservoir bed, their equipment with facilities of water lift, facilities to measure level and pressure of subsurface water in them, to pump subsurface water from them, performance of physical and chemical tests of pumped subsurface waters. During pumping a volume of subsurface water is extracted from the well, and this volume is less than the one contained in its borehole, density of the extracted subsurface water is measured, afterwards it is supplied back into the well in the interval, from which it was pumped, measurement of subsurface water pressure is carried out along piezometric tubes lowered into a well filter and filled by water with available density, afterwards water density is determined in the reservoir bed.
EFFECT: improved validity of produced data and elimination of hydrodynamic mode disturbances in wells in process of monitoring.
1 cl, 1 dwg
FIELD: process engineering.
SUBSTANCE: invention relates to separation of fluids by evaporation. Proposed method comprises evaporation of solvent from solution film formed on rotary drum inner surface. Solution is fed inside drum to allow displacement of solution toward unloading side. Solvent vapors are collected and condensed. Heat power consumed for solvent evaporation, drum rpm and level of solution inside the drum are selected provided concentrated product is discharged in the form of fluid. Flow rate of solution fed into drum is to be calculated by mathematical tools.
EFFECT: sufficient quality of concentrated product and solvent condensed vapors provided in continuous operation without drop in efficiency in wide range of concentrations of solutions.
11 cl, 1 dwg
SUBSTANCE: invention relates to the technology of recycling nuclear energy materials and specifically to methods of cleaning uranium hexafluoride from ruthenium fluorides, and can be used in returning uranium, extracted from spent nuclear fuel, into the fuel cycle of light water reactors. The method of cleaning uranium hexafluoride from ruthenium fluorides involves reaction of gaseous uranium hexafluoride with a sorbent, the sorbent used being porous granules of sintered metal powder of nickel or nickel which contains up to 10 wt % copper.
EFFECT: invention provides efficient cleaning of uranium hexafluoride from ruthenium fluorides at low temperatures, safety and simplification of the process, and also enables to reuse the sorbent after regeneration.
5 cl, 2 tbl, 5 ex
SUBSTANCE: invention relates to hydraulic metallurgy, particularly to extraction of uranium from used phosphate solutions. This process consists in adding the solvent to initial solution, said solvent being selected from the series: KMnO4, K2Cr2O7, HNO3, H2O2, KClO3. Then, uranium-bearing sediment is precipitated by acidity correction by ammonia to pH 2.8-4.0 at 20-35°C. Filtered precipitate is treated by 20-35% solution of NaOH at 80-85°C for 1.5-2.0 hours.
EFFECT: higher yield of uranium, return of high-enriched uranium to fuel cycle, lower costs of higher safety at long-term storage, accounting and control.
2 cl, 1 tbl, 4 ex
SUBSTANCE: method involves leaching a concentrate with aqueous nitric acid solution at high temperature to obtain a pulp which consists of a solid phase and an aqueous phase, filtering off the aqueous phase in form of uranyl nitrate solution, extraction refining uranium using tributyl phosphate in a hydrocarbon diluent. The filtered uranyl nitrate solution, which contains uranium in concentration of 200-400 g/l, dissolved silicon in concentration of 1.0-3.2 g/l and nitric acid in concentration of 1-2 mol/l, is held until stabilisation of viscosity before being fed for extraction.
EFFECT: preventing escape of the aqueous phase with the uranium extract, which improves efficiency of the extraction stage, lowers content of impurities in the uranium extract and enables to obtain a product which meets ASTM C 788-03 requirements.
1 dwg, 1 tbl
SUBSTANCE: invention relates to method of uranium extraction from mother liquors. Method includes obtaining resin, modified by aminophosphonic groups, and obtaining mother liquor, which contains from 25 to 278 g/l of sulphate and uranium. After that, mother liquor is passed through resin, modified by aminophosphonic groups, in acid form to separate uranium from mother liquor. Then, elution of uranium from resin is realised.
EFFECT: possibility of sorption extraction of uranium from solutions, which contain high concentrations of sulfate.
7 cl, 1 tbl, 3 ex
SUBSTANCE: proposed process comprises leaching of uranium by nitric acid and separation of water phase from undissolved precipitate. Then, undissolved precipitate is mixed with fluorine-bearing agent, dissolution of produced charge and/or charge as a suspension in nitric acid solution. Produced solution is returned to production process for extraction of uranium. Nitric acid concentration in solution makes at least 2 mol/l. Dilution is carried out at fluorine-ion concentration at, at least, 15 g/l. Dilution is performed at 60-100°C.
EFFECT: decreased losses of uranium, minimised wastes.
4 cl, 1 dwg, 1 tbl
SUBSTANCE: method involves dissolving wastes in concentrated nitric acid, oxalate precipitation from the solution, drying and calcining the americium oxalate to americium dioxide. The solution obtained by dissolving wastes with high concentration of impurity cations, one of which is ferric iron, is mixed with a reducing agent for reducing ferric iron to ferrous iron. After reduction, the solution with acidity by nitric acid of 1-2.5 mol/l is taken for extraction of americium with a solid extractant based on different-radical phosphine oxide, followed by washing and re-extraction of americium. Oxalate precipitation is carried out from the re-extract with americium concentration of not less than 3 g/l and nitric acid concentration of not less than 3 mol/l, said precipitation being carried out in two steps: adding an oxalate ion to the americium-containing solution in weight ratio to americium of (2-7):1 and then adding water to the separated precipitate in volume ratio to the precipitate of (3-8):1 and the oxalate ion in weight ratio to americium of (1-4):1. The obtained reaction mixture is boiled and taken for separation of americium oxalate from the solution.
EFFECT: high output of the product and degree of purity thereof.
SUBSTANCE: metallic uranium obtaining method involves electrolysis of uranium dioxide in the melt of lithium and potassium chlorides in an electrolysis unit with a graphite anode and a metal cathode and release of metallic uranium on the cathode and carbon dioxide on the anode. First, mixtures of uranium dioxide and carbon are prepared in molar ratio of 6:1 and 1:1 by crushing the corresponding powders; the obtained powders are briquetted into pellets. To the anode space of the electrolysis unit, which is formed with a vessel with porous walls, which is arranged in a ceramic melting pot, there loaded are pellets obtained from mixture of uranium dioxide and carbon, and melt of lithium and potassium chlorides. To the cathode space of the electrolysis unit, which is formed with the vessel walls with porous walls and the ceramic melting pot, there loaded is melt of lithium and potassium chlorides and uranium tetrachloride in the quantity of 5-15 wt % of lithium and potassium chlorides. Electrolysis is performed at the electrolyte temperature of 500-600°C, cathode density of current of 0.5-1.5 A/cm2, anode density of current of 0.05-1.5 A/cm2, in argon atmosphere with periodic loading to anode space of pellets of mixture of uranium dioxide and carbon.
EFFECT: current yield of metallic uranium is 80-90% of theoretical.
SUBSTANCE: method involves dissolving a chemical concentrate of natural uranium in nitric acid solution, extracting and re-extracting uranium. The dissolved concentrate contains 1.2-3.7 wt % iron to uranium, 1.4-4.0 wt % sulphur to uranuim and 0-0.7 wt % phosphorus to uranium in nitric acid solution. Nitric acid and water are taken in an amount which provides the following concentration in the solution fed for extraction: uranium 450-480 g/l, iron (III) ions 0.1-0.3 mol/l, sulphate ions 0.2-0.6 mol/l, phosphate ions 0-0.10 mol/l, and free nitric acid 0.8-2.4 mol/l, and saturation of extractant with uranium during extraction is maintained in accordance with the ratio: Y ≤90.691-34.316·[SO4]+7.611·([Fe]-[PO4])+5.887·[HNO3]-9.921·[SO4]·[HNO3]+19.841·[SO4]2+7.481·([Fe]-[PO4])·[HNO3]-64.728·([Fe]-[PO4])·[SO4]+92.701·[SO4]·[HNO3]·([Fe]-[PO4])-185.402·[SO4]2·([Fe]-[PO4]), where Y is saturation of the extractant with uranium, %, and concentration in the solution fed for extraction, mol/l: [SO4] - sulphate ions, [PO4] - phosphate ions, [HNO3] - nitric acid, [Fe] - iron (III) ions.
EFFECT: obtaining raffinates with low uranium content.
SUBSTANCE: method includes sorption of rich components from production solutions by ion-exchange material counterflow under controlled pH of environment and oxidation-reduction potential Eh. Sorption is performed by ion-exchange materials in stages from production solutions containing uranium, molybdenum, vanadium and rare earth elements. At the first stage uranium and molybdenum are extracted by anion-exchange material sorption. At the second stage vanadium is extracted by anion-exchange material sorption with hydrogen dioxide available at Eh of 750-800 mV, pH of 1.8-2.0 and temperature of 60°C, at that vanadium sorption is performed till complete destruction of hydrogen dioxide and till Eh is below 400 mV. Then barren solutions are transferred to cationite at pH of 2.0-2.5 and Eh of 300-350 mV for extraction of rare earth elements.
EFFECT: sorption concentration and selective separation of uranium and molybdenum from vanadium, and vanadium from rare earth elements, and rare earth elements from iron and aluminium, intensification of sorption process, reduction of flow diagram and possibility of environmentally sound oxidants use.
1 dwg, 4 tbl, 1 ex
SUBSTANCE: processing method of black-shale ores includes crushing, counterflow two-stage leaching by sulfuric acid solution upon heating, separation of pulps formed after leaching at both stages by filtration. Then valuable soluble materials are washed from deposit at the second stage with strengthened and washing solutions being produced, marketable filtrate is clarified at the first stage for its further processing. Ore is crushed till the size of 0.2 mm, leaching at the first stage is performed by cycling acid solution with vanadium under atmospheric pressure, temperature of 65-95°C during 2-3 hours, till residual content of free sulphuric acid is equal to 5-15 g/l. Leaching at the second stage is performed at sulphuric acid rate of 9-12% from the quantity of initial hard material under pressure of 10-15 atm and temperature of 140-160°C during 2-3 hours. Cake filtered after the first stage is unpulped by part of strengthened solution which content is specified within 35-45% of total quantity.
EFFECT: high-efficiency extraction of rich components, possibility of pulps separation by filtration after leaching with high properties thus reducing costs for separation processes.
3 cl, 1 dwg, 1 tbl
SUBSTANCE: processing method of black-shale ores with rare metals extracting includes leaching of ore by sulphuric acid solution with dilution of rare metals. Leaching is performed in autoclave by sulphuric acid solution consisting of free and combined sulphuric acid with ratio of H2SO4(free):H2SO4(comb)=2:1, and containing 25-45 g/l of iron sulphate, 70-90 g/l of aluminium sulphate and 0.5 g/l of nitric acid. At that the process is performed under pressure in autoclave equal to 10-15 atm with mixing at temperature of 140-160°C in concentration range of general H2SO4(gen) equal to 350-450 g/l under pulp density S: L=1:0.7-0.9, preferably 1:0.8, under constant oxidation-reduction potential Eh in the system equal to 350-450 mV during 2-3 hours till residual concentration of free H2SO4(free) is within 45-75 g/l.
EFFECT: increasing break-down of ore and extraction of rare metals: vanadium, uranium, molybdenum and rare-earth elements, reducing consumption of acid and improving efficiency of autoclave volume usage.
1 tbl, 1 ex
SUBSTANCE: invention relates to chemical engineering of inorganic substances and can be used to process depleted uranium hexafluoride. The method of processing uranium hexafluoride involves dissolving uranium hexafluoride in water to obtain a uranyl fluoride solution; the obtained solution is treated with ammonia water to obtain solid ammonium polyuranate and ammonium fluoride solution; the solid fraction is filtered out and annealed at temperature of 450-600°C to thiuram octaoxide and after filtration, the solution is evaporated to obtain solid ammonium fluoride.
EFFECT: invention provides an efficient industrial method of processing uranium hexafluoride to obtain commercial-grade versions of uranium and a fluorine-containing product.