Method of platinum group metals extracting from voloxidized snf acid dissolving product

FIELD: fuel.

SUBSTANCE: invention relates to spent nuclear fuel (SNF) processing. Method of platinum group metals extracting from product of voloxidized SNF acid dissolving involves, that obtained after voloxidiation SNF is dissolved in nitric acid in the range of temperatures of 83–86 °C for 4–5 hours to obtain residual nitric acid content in product in range of 1.42–2.3 mol/l,uranium in range of 480–600 g/l, obtained product is thermostated in range of temperatures of 69–80°C for 2–48 hours, adding flocculant and dispersing reaction mixture, performing sediment accumulation in device bottom part by sedimentation deposition in range of temperatures of 35–57 °C for 6–24 hours. Product clarified part is separated by decantation, averaging sedimented deposit in remaining clarified solution amount in device.

EFFECT: invention enables to extract into deposit of more 78,7 % of platinoids and separate 98,1 % of suspensions formed during SNF dissolving.

15 cl, 2 ex

 



 

Same patents:

FIELD: physics.

SUBSTANCE: re-extraction step using water-soluble complexions - monoamides of diglycollamic acids - is provided.

EFFECT: enabling precipitation of plutonium or a mixture of uranium and plutonium directly from the obtained re-extract.

3 cl, 4 dwg

FIELD: physics.

SUBSTANCE: method comprises separating nuclear fuel material containing thorium metal and processing an oxide of the nuclear fuel material in a reactor (1), containing thorium oxide in spent fuel, placed in a bin (3a). The disclosed method includes a first step for electrolytic reduction of thorium oxide, which includes applying electrode potential across an anode (2) and a cathode (3) in a first melt of alkali-earth metal halides, a first step of washing the reduction product and a main step for electrolytic separation of the reduction product. The first melt of halides further includes an alkali metal halide and at least one of the substances: calcium chloride, magnesium chloride, calcium fluoride and magnesium fluoride. The disclosed method can further include a second for electrolytic reduction of uranium oxide, plutonium oxide and light actinide oxides in a second melt of alkali metal halides.

EFFECT: selective separation of thorium metal from spent nuclear fuel material containing thorium oxide.

16 cl, 9 dwg

FIELD: physics, atomic power.

SUBSTANCE: invention relates to nuclear engineering, particularly thermal-neutron reactor fuel. The fuel composition for water-cooled thermal-neutron reactors includes a mixture of regenerated plutonium and enriched uranium in the form of oxides, wherein the enriched uranium used is enriched regenerated uranium, with the ratio of components defined by energy potential equal to the potential of freshly prepared fuel from enriched natural uranium which provides 100% reactor core loading.

EFFECT: invention enables to fully and simultaneously recycle regenerated uranium and plutonium extracted from spent nuclear fuel.

6 cl, 4 ex

FIELD: physics, atomic power.

SUBSTANCE: invention relates to pyrochemical technologies of processing irradiated nuclear fuel, particularly oxide fuel. The method for non-current production of uranium (V) in molten alkali metal chlorides (NaCl-2CsCl, NaCl-KCl, LiCl-KCl), containing uranium (VI) ions, involves holding zirconium metal in the atmosphere over the melt as a getter at temperature of 550-750°C for 180-250 minutes Uranium (V) is formed via thermal decomposition of uranyl chloride, which is accelerated by zirconium metal as shown by recorded absorption spectra of the melt.

EFFECT: non-current production chloride melts with high content of uranium (V) without introducing foreign components into the melt.

1 dwg

FIELD: physics.

SUBSTANCE: separation of electropositive fission products from fission of alkali metal chlorides is carried out by chemically reducing the electropositive fission products on molybdenum metal. Separation of a portion of the electropositive fission products (zirconium and niobium) is carried out based on an exchange mechanism to form zirconium and niobium dioxides. Molten molybdenum is removed in form of molybdenum pentachloride by bubbling chlorine gas through the salt melt.

EFFECT: invention provides high percentage of extraction of electropositive fission products, a simple apparatus scheme, cheap process of treating nuclear fuel.

2 cl

FIELD: power engineering.

SUBSTANCE: method of isotopic recovery of regenerated uranium includes increase of content of isotope U-235 in hexafluoride of regenerated uranium to the specified value in the range of 2.0÷5.0 wt %, reduction of relative concentration of isotope U-232 in the mixture of uranium isotopes and direct enrichment of hexafluoride of regenerated uranium with isotope U-235 on the double-cascade plant from separation stages of gas centrifuges. At the same time in the first cascade the regenerated uranium is enriched with isotope U-235 up to 5.0÷10.0 wt % with maintenance of ratio of mass flows of the dump flow and bleed flow of the cascade in the range of (6.9÷18.4):1. Flows of the dump and bleed of the first cascade are sent for supply of the second cascade. The regenerated uranium is bled from the separating stage of the central part of the second cascade.

EFFECT: complete treatment of a burnt mixture of uranium isotopes from the most radiation-hazardous nuclide U-232 and production of commercial low-enriched uranium hexafluoride at minimum retuning of industrial cascades of centrifuges.

2 cl, 1 dwg, 7 tbl

FIELD: power engineering.

SUBSTANCE: invention may be used in radiochemical production for regeneration of radiated nuclear fuel, and also for processing of manufactured, but rejected nuclear fuel. A drum mill for processing of radiated or rejected nuclear fuel comprises a vessel, inside of which there is a perforated drum and a sieve drum installed concentrically as joined with a drive of rotation and closed at the ends. The latter is filled with grinding solids in the form of rods. The drum comprises a nozzle for loading of nuclear fuel into the sieve drum for processing and a nozzle to unload the processed nuclear fuel. The drum mill is equipped with a nozzle to supply a gaseous oxidiser into the vessel. The sieve drum is fixed on the perforated drum from inside, and the vessel bottom is arranged as flat and inclined, is equipped with a container attached to it. The nozzle for unloading of the processed nuclear fuel is installed in the container and is communicated with a vacuumising system.

EFFECT: combination of two technological processes in one device: grinding of initial nuclear fuel and its recrystallisation, makes it possible to exclude additional operations related to transportation of ground fuel.

2 dwg

FIELD: chemistry.

SUBSTANCE: invention relates to methods of dissolving fuel which is a mixture of uranium and plutonium oxides. The method involves dissolving MOX fuel in a nitric acid solution which also contains fluoride and gadolinium ions with the following concentrations: nitric acid (6-9) mol/l, sodium fluoride (0.05-0.08) mol/l and gadolinium nitrate in terms of gadolinium (1.3-1.5) g/l.

EFFECT: invention enables complete dissolution of mixed uranium-plutonium fuel without formation of residues and nuclear safety of the dissolution process.

3 cl, 1 tbl

FIELD: physics.

SUBSTANCE: mass separation of a multicomponent plasma stream is carried out using crossed radial electric and azimuthal magnetic fields as the plasma stream moves through a separating volume. The value of the azimuthal magnetic field is selected such that it ensures magnetisation of ions and electric drift of the plasma. The magnetic field generating system does not have an azimuthator. The combination of electromagnetic fields of the mass-separator ensures maximum dispersion at the focal point of central mass ions. The accompanying electron gun is linear, which ensures injection of electrons along the azimuthal magnetic field.

EFFECT: broader functional capabilities of mass-separators and simple design thereof.

7 dwg

FIELD: power industry.

SUBSTANCE: method and device are implemented when quasineutral axial symmetric multiple-component flow of plasma is obtained by means of plasma accelerator, during transportation of flow through azimuth device with transverse radial magnetic field, flow of plasma flow divided into masses through separating volume with stationary radial electric and homogeneous longitudinal constant magnetic fields, collection of ions of two groups of spent nuclear fuel (SNF) to ion receivers located on cylindrical surfaces. Ions of the third group of SNF are collected to annular end receiver.

EFFECT: group of inventions allows enlarging functional capabilities of plasma optic mass-separator owing to minimising negative impact of power and angular spreads of ions of various chemical elements in plasma flow and proper selection of the shape, quantity and location of receivers of groups of ions.

2 cl, 7 dwg

FIELD: recovery of spent nuclear fuel.

SUBSTANCE: proposed method involves enhancement of uranium-235 isotope content in recovered uranium to 2.0-5.0 mass percent while reducing absolute and/or relative concentration of uranium even isotopes. Method includes division of isotope mixture of raw uranium reagent in gas-centrifuge isotope-division cascade and mixing of separated commercial isotope mixture with uranium thinner. Isotope mixture division is effected in two-cascade arrangement. Raw uranium reagent is enriched with uranium-235 fissionable isotope in first single cascade up to content over 90 mass percent. Second single cascade is used for cleaning isotope mixture from uranium-232 and uranium-234 isotopes. Select flow of second cascade enriched with uranium-235 isotope is conveyed as commercial isotope mixture for mixing up with uranium-thinner.

EFFECT: enhanced quality of reducing recovered uranium and minimized uranium-thinner requirement.

12 cl, 1 dwg, 6 tbl

FIELD: extraction processes for recovery of nuclear fuel, uranium concentrates, and uranium-containing reusable parts.

SUBSTANCE: proposed process for uranium extraction affinage includes dissolution of uranium concentrate at nitric acid excess of 0.75 - 1.0 mole/l and temperature of 80 - 95 °C; prior to extraction uranyl nitrate solution is doped with urea nitrate; post-extraction raffinate and alkali decanting product produced as result of re-extract treatment are separately subjected to carbamide denitration with solution being cooled down and urea nitrate sediment separated; decanting products produced in the process are mixed up and subjected to electrochemical treatment.

EFFECT: reduced nitric acid consumption and escape of raffinate-containing nitrate ions, escape of nitric oxides in uranium concentrate dissolution, and uranium loss with effluents.

7 cl, 5 dwg

FIELD: recovery of irradiated nuclear fuel.

SUBSTANCE: proposed method for reconditioning reusable extractant includes treatment of the latter with aqueous alkali solution. Extractant containing uranium in amount of minimum 5 g/l is treated with alkali solution whose concentration is over 10 mole/l followed by sediment separation.

EFFECT: reduced radionuclide content of reusable extractant including difficult-to-remove radioactive ruthenium.

5 cl, 2 tbl, 2 ex

FIELD: recovering solid irradiated nuclear fuels.

SUBSTANCE: proposed method for recovering solid irradiated nuclear fuel in the form of various uranium-containing composites (metal, carbide, oxide, and the like) for its further reuse in nuclear-fuel cycle includes dispersion of mentioned composites by way of thermal oxidation followed by vacuum baking at the same time distilling in vacuum volatile products of fission, such as cesium-137. Dispersion is conducted in controlled oxygen-containing medium while cycling them at temperatures ranging between 400 and 1000 °C; baking is conducted for 1 h at residual pressure of maximum 10-2 Pa and temperature of minimum 1300 °C.

EFFECT: ability of reducing gamma-activity of irradiated nuclear fuel, mainly cesium, to desired level for its reuse in nuclear fuel cycle.

5 cl, 1 tbl, 3 ex

FIELD: recycling technology for power-generating nuclear materials.

SUBSTANCE: concentration of uranium-235 fissionable isotope is raised above source content to desired value of 2-5 mass percent by direct enrichment in cascade of gas centrifuges. Uranium-235 concentration in cascade dump is 0.1-0.3 mass percent. At the same time uranium isotope mixture having lower concentration of uranium-232 and uranium-236 isotopes than burnt-up source mixture is thinned with hexafluoride. To this end uranium-thinner hexafluoride is introduced to interstage stream of cascade at same or almost same concentration of uranium-235 fissionable isotope. For thinning use is made of hexafluoride of natural mixture of uranium isotope, hexafluoride of uranium isotope mixture separated from burnt-up nuclear fuel, as well as mixture of source uranium isotope burnt-up mixture and hexafluoride of natural mixture of uranium isotopes or hexafluoride of uranium isotope mixture separated from burnt-up nuclear fuel.

EFFECT: ability of producing marketable uranium hexafluoride at minimal quantity of uranium-separating factory facilities used for thinner production.

16 cl, 3 dwg, 3 tbl

FIELD: physics.

SUBSTANCE: invention relates to the technology of recycling nuclear energy material. The said method of isotopic recycling uranium involves correction of the composition of burnt uranium isotopic mixtures in a double centrifugal cascade with selection of hexafluoride of uranium isotopic mixtures, purified from dangerous radioactive isotopes U-232 and U-234, through selection of the heavy fraction of second ordinary cascade. Separation of uranium isotopic mixtures in the second ordinary cascade is done in the presence of a carrier gas. The carrier gas has average molecular weight ranging from 346 to 348 amu.

EFFECT: invention reduces the mass of dangerous radioactive wastes and reduces loss of fissile U-235 isotope.

7 cl, 1 dwg, 2 tbl

FIELD: nuclear physics.

SUBSTANCE: invention relates to a nuclear fuel cycle, and specifically to methods of treating contaminated hazardous 232U, 234U, 236U isotopes of uranium material on a gas centrifuge cascade. The method involves treating contaminated uranium material fed into a gas centrifuge cascade, obtaining low enriched uranium from cascade selection using natural uranium hexafluoride in an intermediate cascade selection into which natural uranium hexafluoride is fed, producing a product with low concentration of at least one of the hazardous 232U, 234U, 236U isotopes compared to contaminated material with mass ratio of contaminated uranium material to natural uranium hexafluoride taken for treatment equal to (1÷25):100.

EFFECT: treating uranium material contaminated with hazardous impurities, obtaining quality material with permissible content of limiting hazardous isotopes, widening of the raw material base for fission plants, less separation work for processing material.

7 cl, 4 dwg, 7 ex

FIELD: power industry.

SUBSTANCE: separation of uranium isotopes in the form of hexafluoride of isotopic mixture of uranic regenerate on the installation consisting of two subsequent gas centrifugal stages; at that, hexafluoride of uranium recovered as per isotopic composition is obtained in flow of the second stage bleeding powered with heavy fraction of the first stage, which has been obtained during enrichment of light fraction of the first stage with 235U isotope to the concentration not exceeding 20 wt %. Isotopes are separated at the first stage so that there obtained is concentration of isotopes 232U and 234U in heavy fraction of the first stage, which provide required concentrations 232U and 234U at the second stage bleeding, and isotopes are separated at the second stage till the second stage bleeding flow is saturated with isotope 235U, which ensures the required concentration of isotope 236U in the second stage bleeding flow.

EFFECT: eliminating hazardous high concentrations of fissionable isotope 235U.

FIELD: power industry.

SUBSTANCE: method and device are implemented when quasineutral axial symmetric multiple-component flow of plasma is obtained by means of plasma accelerator, during transportation of flow through azimuth device with transverse radial magnetic field, flow of plasma flow divided into masses through separating volume with stationary radial electric and homogeneous longitudinal constant magnetic fields, collection of ions of two groups of spent nuclear fuel (SNF) to ion receivers located on cylindrical surfaces. Ions of the third group of SNF are collected to annular end receiver.

EFFECT: group of inventions allows enlarging functional capabilities of plasma optic mass-separator owing to minimising negative impact of power and angular spreads of ions of various chemical elements in plasma flow and proper selection of the shape, quantity and location of receivers of groups of ions.

2 cl, 7 dwg

FIELD: physics.

SUBSTANCE: mass separation of a multicomponent plasma stream is carried out using crossed radial electric and azimuthal magnetic fields as the plasma stream moves through a separating volume. The value of the azimuthal magnetic field is selected such that it ensures magnetisation of ions and electric drift of the plasma. The magnetic field generating system does not have an azimuthator. The combination of electromagnetic fields of the mass-separator ensures maximum dispersion at the focal point of central mass ions. The accompanying electron gun is linear, which ensures injection of electrons along the azimuthal magnetic field.

EFFECT: broader functional capabilities of mass-separators and simple design thereof.

7 dwg

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