Method of producing oxyhalogenide and/or oxide of actinide(s) and/or of lanthanide(s) from medium comprising at least one molten salt

FIELD: chemistry.

SUBSTANCE: invention relates to a method of producing oxychloride and/or oxide of actinide(s), and/or lanthanide(s) from chloride of actinide(s), and/or lanthanide(s), present in a medium containing at least one molten salt of chloride type. Method involves a step for reacting chloride of actinide(s) and/or lanthanide(s) with wet inert gas.

EFFECT: invention provides efficient production of oxyhalogenide and/or oxide of actinide(s), and/or lanthanide(s), as well as formation with elements of actinides or lanthanides, products, different from oxyhalogenides or oxides, and excluding cation-contamination of medium containing molten salt, simplifying recirculation of molten salts.

11 cl, 3 ex

 



 

Same patents:

FIELD: physics, atomic power.

SUBSTANCE: invention relates to means of removing uranium dioxide used as nuclear fuel from coolant of primary and main loops of research and power reactors. In the disclosed method, the loops are treated with oxalate-peroxide solutions with pH=6.5-7.0. Treatment is carried out in a certain sequence of process steps which result in the complete dissolution of uranium dioxide.

EFFECT: tenfold increase in the rate of dissolution of uranium dioxide, which cuts reactor downtime, not without significant effect on structural materials of the loops and corrosive radioactive deposits on the inner surfaces of the equipment of the loops, which prevents disruption of steady-state heat-exchange between the coolant and the solid phase, and further increase in activity of liquid radioactive wastes.

1 dwg, 2 tbl, 2 ex

FIELD: chemistry.

SUBSTANCE: invention refers to a nanocomposite solid material of hexa- and octacyanometallates, methods for preparing them and using as mineral fixing agents. What is presented is the nanocomposite solid material containing metal-coordination polymer nanoparticles with ligands CN, containing cations Mn+, wherein M is a transient metal, and n is equal to 2 or 3; and anions [M'(CN)m]x-, wherein M' is a transient metal, x is equal to 3 or 4, m is equal to 6 or 8; the above cations Mn+ of the coordination polymer are bound through the metal-organic binding to the organic group of organic grafting chemically attached inside pores of a porous glass carrier, and the pores of the porous glass are produced by the selective chemical etching of a borate phase of solid borosilicate glass.

EFFECT: presented material has a reproducible and controlled composition and properties that is ensured by the reliability of the method for producing, and has the excellent binding properties.

36 cl, 4 dwg, 4 ex

FIELD: process engineering.

SUBSTANCE: invention relates to recovery of solid filter containing iodine as iodide and/or silver iodate and, possibly, physically sorbed molecular iodine in solid filter containing silver as nitrate. Proposed method comprises the following stages. 1. Removal of iodine from filter by its processing by base aqueous solution containing reducing agent and separation of filter from said base aqueous solution. 2. Extraction of silver from filter obtained at stage 1 by bringing said filter in contact with acid aqueous solution and separation of filter from said acid solution. 3. Impregnation of filter with silver produced at stage 2 by bringing said filter in contact with silver nitrate solution and filter drying.

EFFECT: reuse of used filters, reduced wastes.

10 cl, 3 ex

FIELD: process engineering.

SUBSTANCE: invention relates to electrokinetic decontamination of hard porous medium. Proposed method comprises extraction of contaminants from hard medium to electrolyte, mainly as inorganic gel. Note here that said extraction results from flashing between two electrodes arranged on hard medium surface and/or therein. Note also that contact between at least one of said electrodes and said hard medium is via ply of said gel. It comprises also gel drying. Said gel contains contaminants dried to fragile dry residue to be easily removed.

EFFECT: efficient decontamination of cement matrix materials at dismantling of military or industrial structures, particularly, nuclear plants etc, that can be contaminated with ecotoxic chemicals or radioactive substances.

15 cl, 9 dwg

FIELD: power engineering.

SUBSTANCE: subject of this invention is the method of treatment of carbon-containing wastes. In particular, the following is provided: the first type of treatment for production of carbon oxide, and the second type of treatment for production of solid residue of carbon oxide by means of reaction with the chosen element. The method in accordance with this invention includes: the first stage, during which the first and second types of treatment are applied simultaneously, and the second stage, during which only the first type of treatment is applied.

EFFECT: invention makes it possible to reduce quantity of solid residue.

31 cl, 2 dwg

FIELD: physics, nuclear.

SUBSTANCE: invention relates to nuclear power and particularly issues of handling liquid radioactive wastes. The method of cleaning still residues of liquid radioactive wastes from radioactive cobalt and caesium by oxidising the still residue and extracting activated corrosion products by filtering, characterised by that hydrogen peroxide is added to the still residue and repeatedly passed in circulation mode through a tubular reactor, while exposing the still residue to hard UV radiation; the process is carried out at solution pH 7-10 and temperature 45-98C; after separating the sludge by microfiltration, said sludge containing radioactive cobalt, iron and manganese, solutions of diethyldithiocarbamates of alkali metals and transition metal salts are simultaneously added to the filtrate at the same pH and temperature values, followed by separation by microfiltration of the formed precipitate of diethyldithiocarbamates of transition metals with radioactive cobalt; and radioactive caesium is extracted on ion-selective sorbents in steps using a counterflow method with removal of spent sorbents by microfiltration together with cobalt diethyldithiocarbamate.

EFFECT: invention enables to avoid use of an ozonator station, makes the process safer and considerably increases efficiency of processing liquid radioactive wastes.

5 cl, 3 ex, 4 tbl

FIELD: power engineering.

SUBSTANCE: proposed method consists in thermal oxidative of irradiated fuel from uranium dioxide and comprises separating fuel elements into fragments, and oxidative treatment of said fragments by gas-air mix in two steps: first, by mix of air and carbon dioxide at 400-650C, and, second, by steam-air mix at 350-450C. Both steps are implemented at continuous or intermittent mechanical activation of reaction stock.

EFFECT: higher yield of tritium, reduced sublimation of cesium compounds.

3 cl, 1 ex

FIELD: power industry.

SUBSTANCE: treatment method of radioactive contaminated metal and graphite wastes of uranium-graphite nuclear reactors involves loading of radioactive contaminated metal wastes and flux to oven, melting of metal wastes, and removal of molten metal and formed slag flux from the oven. Before radioactive contaminated metal wastes are loaded to the oven there loaded is layer of radioactive contaminated graphite and it is ignited in oxidising medium with plasma generated by plasmatron of the oven; after that, plasmatron is switched off and loading of fragmented radioactive contaminated metal wastes and flux are loaded to the furnace downwards and layer by layer.

EFFECT: invention allows minimising the volume of secondary radioactive wastes, reducing energy consumption and excluding the possibility of occurrence of emergency situation.

1 dwg

FIELD: nuclear physics.

SUBSTANCE: invention relates to environmental protection, specifically to devices for treating highly active sources of ionising radiation by enclosing them in a metallic matrix directly in the storage, and can be used at centralised nuclear waste disposal points. The proposed device has an electrofusion device, a buffer chamber with a flexible metal conductor inside it. The electrofusion device is placed in the buffer chamber, the input of which is mated with the flexible metal conductor through a flange connection with a locking mechanism. The buffer chamber is fitted with an outlet pipe and is connected to a system for cleaning and pumping air, which consists of a filter and a ventilator. The flexible metal conductor is made in form of a spiral metal hose made from stainless steel with four heat insulating layers made from glass or basalt, or organosilicon fibre. Between the first and second heat insulating layers there is a thermal converter which is connected to an electrical circuit in the control console of the device, and between the second and the third layers there is a copper current conducting bus which is connected at the lower end of the flexible metal conductor with the spiral metal hose using a collar clamp. The upper end of the flexible metal conductor is mated with above mentioned flange connection with a locking mechanism.

EFFECT: proposed device prevents crystallisation of molten matrix material in the flexible metal conductor and breaking of the metal conductor in the loading channel of the storage.

1 cl, 2 dwg

FIELD: nuclear power production industry.

SUBSTANCE: radioactive ion-exchanging resins preparation for immobilisation into bulk structures. Radioactive ion-exchanging resin is mixed with hard non-organic inert bulk material. After that, during mixing the above mixture is exposed to thermal treatment at the temperature no less than 250C but not higher than 300C.

EFFECT: reduction of power consumption, prevention of radioactive ion-exchanging resins products agglomeration after thermal treatment, prevention of radioactive ion-exchanging resins inflammation risk during thermal treatment, prevention of volatile resin and oil compositions formation, simplification of process, increased compatibility of radioactive ion-exchanging resins with cement-like materials after their thermal treatment.

FIELD: metallurgy.

SUBSTANCE: proposed method comprises immersion of alloy into salt melt to change rare-earth element from liquid alloy into melt by oxidation. Note here that said oxidation us performed in zinc chloride melt at 420-550C while melt zinc ions are used as oxidiser.

EFFECT: higher yield.

2 tbl, 2 ex

FIELD: fuel systems.

SUBSTANCE: invention is related to recycling of return nuclear fuel (RNF) and materials of blanket region (BR) of fast breeder reactors (FBR) for their multiple use with the possibility to adjust content in creation of a new fuel composition. Initial chemical state of processed material may vary: oxides, nitrides, metals and alloys. Method represents a combination of serial processes of chemical transformation of radiated nuclear fuel (RNF): fluoridation with gaseous fluorine and extraction of main uranium mass; transition of fluoridation remains into oxides (pyrohydrolysis); - chlorination of oxides in recovery conditions with group separation of plutonium chlorides, uranium (left in process of fluoridation) and fission products. Further "plutonium" and "uranium" fraction, and also fraction containing fission products (and, possibly, minor-actinides), are used each separately in various processes according to available methods. Earlier produced uranium hexafluoride, with low boiling fluorides of fission products, is cleaned from the latter and used, according to objectives of processing, also by available methods. Using waterless processes with application of salt melts, suggested version makes it possible to realise continuous highly efficient processes of fuel components production, moreover, it is stipulated to carry out preparation stages in continuous mode. Plant for processing of spent nuclear fuel containing uranium and plutonium includes three serially installed devices: fluoridiser; pyrohydrolysis device; chlorinator-condensator-granulator device. Two last devices are of flame type. The last of devices represents a pipe with a central element, in which lines of inlet product and reagents supply are installed. In lower part there is an expansion in the form of pear with a flame burner along its axis, and medium part has a row of conical shelves inside, between which there are nozzles with pipelines for chlorides outlet. Nozzle for chlorides outlet is also arranged in lower point of pear-like part. Nozzle for exhaust of non-condensed gases is provided in upper part. Granulator is arranged as reservoir with low boiling incombustible liquid, and to produce drops, capillaries are provided at pipeline ends.

EFFECT: highly efficient method for processing of spent nuclear fuel of practically any composition from thermal reactors and fast breeder reactors, blanket region of fast breeder reactors and some other types of reactors with the possibility to produce several other types of fuel compositions.

19 cl, 3 dwg

FIELD: chemistry.

SUBSTANCE: group of inventions concerns application of polymer-containing solution or water suspension paste and a device of ruthenium collection in gaseous discharge. The solution or water suspension paste contains one alkylene glycol polymer and/or one alkylene glycol co-polymer. The alkylene(s) contains 2-6 carbon atoms for ruthenium collection in gaseous discharge. The device includes a ruthenium collection cartridge with a substrate bearing alkylene glycol polymer or co-polymer. The alkylene(s) contains 2-6 carbon atoms.

EFFECT: improved ruthenium collection and chemical recovery of ruthenium oxide.

22 cl, 8 dwg

FIELD: nuclear engineering.

SUBSTANCE: proposed method for volume crystallization of plutonium dioxide includes treatment of molten alkali-metal chlorides with plutonium compound dissolved therein, as well as treatment of melt obtained in the process by oxygen-containing gas mixture and precipitation of large-crystal plutonium dioxide on bath bottom. In the process closed-porosity graphite granules are disposed on melt surface, their contact with melt being afforded as they are consumed. Apparatus for volume crystallization of plutonium dioxide from molten alkali-metal chlorides with plutonium compound dissolved therein has bath, cover, melt mixing system, and device for feeding soluble plutonium compounds and gas mixture to melt. Bath, parts and assemblies contacting the melt are made of ceramic material shielded at melt boundary level with pyrographite parts. Gas mixture feeding devices have ceramic and pyrographite tubes.

EFFECT: enhanced durability of equipment.

3 cl, 4 dwg

The invention relates to methods of non-aqueous dissolution of uranium and uranium-containing materials and can be used to extract uranium from spent nuclear fuel, metallurgical wastes of uranium and its alloys and products

The invention relates to the field of production and processing of nuclear fuel

The invention relates to nuclear energy, in particular to the production of plutonium metal and mixed uranium-plutonium oxide fuel

The invention relates to the field of processing of irradiated and defective nuclear fuel, in particular mononitrides uranium-plutonium fuel

FIELD: chemistry.

SUBSTANCE: invention relates to the radiochemical industry and nuclear power engineering, is aimed at obtaining a mixed dioxide (U,Pu)O2 and can be used to produce mixed uranium-plutonium MOX fuel for VVER-1000 reactors and fast neutron reactors (BN-600, BN-800) at nuclear power plants. The method of producing a solid plutonium dioxide solution in a uranium dioxide matrix includes reacting nitrate complexes of uranium and plutonium with relative content thereof in the solution of 95-70 wt % and 5-30 wt %, respectively, with hydrazine hydrate to obtain a mixed amorphous compound of uranium and plutonium, holding the mixed amorphous compound of uranium and plutonium in the mother solution at 80-90C for not less than 3.5 hours to obtain a precipitate of a fine powder of homogeneously mixed hydrated plutonium and uranium dioxide, separating the precipitate from the mother solution and heating to 280-300C to form the end product.

EFFECT: invention provides a cost-effective, simple and less energy consuming method of producing a solid plutonium dioxide solution in a uranium dioxide matrix.

2 cl, 6 dwg, 2 ex

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