The way pyrochemical regeneration nuclear fuel
(57) Abstract:The invention relates to the field of processing of irradiated and defective nuclear fuel, in particular mononitrides uranium-plutonium fuel. The inventive container load nitride fuel and use the electrolyte, the composition of which in % by weight is not less than 15% of trichloride uranium, heat the electrolyte to a temperature of at least 600oC and create the container, the current density is not more than 0.3 A/cm2and on the electrode is not more than 0.4 A/cm2in addition , fusible electrode to create a current density of not more than 0.1 A/cm2and the melt is heated to a temperature not exceeding 700oC, and the container with nitride fuel is placed fusible metal or alloy, more electrophoretically than uranium nitride, and, in addition, is dissolved in a container of this metal or alloy and precipitate the uranium and plutonium on fusible electrode is the cathode. 4 C.p. f-crystals, 2 tab. The invention relates to the field of processing obolgannogo and defective nuclear fuel, in particular mononitrides uranium-plutonium fuel.Improvement of operational characteristics of fast reactors, increasing their betapace monocarbide. Fundamental properties mononitride fuel when used in the active zone a long way to ensure inherent safety, to reduce the sodium void reactivity effect (reactors, sodium cooled), to increase the reactivity margin on the fuel burnup, etc.The known method hydrometallurgical regeneration mononitride fuel (PUREX process), which consists in dissolving in nitric acid, the extract solution, extraction of actinides 30% solution of TBP in n-dodecane  the final product of which are oxides of uranium and plutonium.The disadvantages of this method are the large water volumes, which contain 15 to 20 percent fissile elements (U and Pu); the need for preliminary lengthy excerpts extracted from the reactor fuel to reduce activity to avoid decomposition of the extractants; a significant number of high-level waste solutions 150 700 l/tonne fuel and more (3 to 5 m3) solutions average activity; obtaining the final product in the form of oxides of uranium and plutonium, of which the synthesis and manufacture of cores (tablets) mononitride is a more complex technology, requiring for redlagaemoe method is a method of pyrochemical regeneration of nuclear fuel, namely, that in the molten eutectic mixture of salts of potassium chloride and lithium enter the uranium chloride and immersed container with nuclear fuel and two electrodes, one of which is made of fusible metal, then connect the container with the positive and negative electrodes with the poles of the current source and conduct electrolysis, after which subjected deposited on the electrodes (cathodes) products dividing the heat-emitting pure metals 
According to this method, the irradiated metal doped fuel is subjected to anodic dissolution in the electrolyte and the cathode to the recovery of metals on the cathode.The advantage of this method according to the authors is its compact size, small number of waste, reducing the cost of capital construction, the possibility of NPP allocation.The disadvantage of this method is its applicability only for the regeneration of the metal fuel, because other fuels, such as nitride, compounds of uranium, plutonium and other elements of the fission products are chemically stable compounds. In particular, the values of free energy of formation of mononitride uranium G298= -63,4 kcal/mol vikluchaetsya in the obtaining of metallic uranium and plutonium from irradiated and/or defective nuclear fuel, in particular, mononitrides uranium-plutonium fuel, for the production of nuclear fuel, suitable for the needs of the nuclear industry.The result of solving the aforementioned problem provides the decomposition mononitride fuel and the transfer of uranium, plutonium and actinides in the form of chlorides in the electrolyte, the recovery and allocation of metallic uranium and plutonium removal of fission products (PD) of nuclear fuel.According to the invention in the method of pyrochemical regeneration of nuclear fuel, which consists in the fact that in the molten eutectic mixture of salts of potassium chloride and lithium enter the uranium chloride, immerse the container with nuclear fuel and two electrodes, one of which is made of fusible metal, then connect the container with the positive and negative electrodes with the poles of the current source and conduct electrolysis, after which subjected deposited on the electrodes (cathodes) products dividing the heat-emitting pure metals, container load nitride fuel and use the electrolyte, which in the mass is not less than 15% trichloride uranium, heat the electrolyte to a temperature of at least 600oC and create on the container or road which case the electrode to create a current density of not more than 0.1 A/cm2and the melt is heated to a temperature not exceeding 700oC, and the container with nitride fuel is placed fusible metal or alloy, more electrophoretically than uranium nitride, and, in addition, after running out of fuel is dissolved placed in the container more electrophoretically than uranium nitride, a low-melting metal or alloy and precipitate the uranium and plutonium on fusible electrode the cathode.Irradiated or defective mononitride fuel is subjected to electrochemical decomposition (at the anode). During anodic dissolution occurs the separation of PD in the form of gases (nitrogen, xenon, krypton, partially iodine) and in the form of insoluble sludge residue (zirconium nitride, molybdenum, technetium, and noble metals). In the electrolyte together with the uranium, plutonium and actinides are transferred and accumulated chlorides of alkali, alkaline earth and rare earth metals.The restoration of the main mass of the uranium is held on the electrode located at the operating temperatures in the solid state, with the formation of metallic crystals dendritic structure, and plutonium, actinides and the remnants of the uranium restore on fusible electrode (cadmium) with the formation of the alloy.Received on lit.Separation of plutonium and other actinides from the fusible alloy is carried out by distillation of the metal cathode.To exclude the presence in the electrolyte of the oxygen containing compounds and obtain the given content trichloride uranium latter is introduced into the electrolyte by anodic dissolution of metallic uranium.Example 1.The following examples of implementation was conducted in an electrolyzer coupled with a glove sealed camera service with dry inert atmosphere (argon).In the crucible of steel HIT load the mixture of potassium chloride and lithium medictions composition (for example, 55 and 45% by weight, respectively), melt it and saturate the melt trichloride uranium (for example, by anodic dissolution of uranium in the electrolyte). Then the melt was placed a container in the form of, for example, perforated molybdenum basket loaded with pills from mixed (uranium-plutonium) nitride fuel with a diameter of 6.9 mm, a length of 11 mm, a density of 83 95% of TP and two electrodes, one of which is made of fusible metal. Connect the container with a positive and "solid" electrode with the negative poles of the power source. By passing direct current produced by the kind of the cathode in the form of dendritic powder. Because in the process of electrolysis of spent fuel also stand out DD is cleaned metallic uranium and plutonium from harmful impurities.For the extraction of metals from the cathode sludge spend dividing melting at a temperature of 1250oC, after which the released electrolyte is returned to the electrolysis, and an ingot of an alloy of uranium with plutonium, purified from the AP, sent on receipt of the fuel alloy.The way pyrochemical regeneration nuclear fuel spent in different modes and the electrolyte composition.Obtained empirically the results are shown in tables 1 and 2.Based on these data, it was determined that to obtain the claimed technical result is needed in the eutectic mixture of salts of potassium chloride and lithium chloride to input at least 15% by weight of uranium chloride, heated to a temperature of at least 600oC while maintaining the current density at the anode is not more than 0.3 A/cm2and on "hard" the cathode is not more than 0.4 A/cm2because when the temperature is less than 600oC and the content in the melt of trichloride uranium in an amount not less than 15% by weight yield of uranium reaches industrial values and uneconomical, and at current densities on Antioch processes: dissolution of the material of the container and/or salt decomposition of the electrolyte, the formation of fine crystals of uranium and plutonium.When these values were processed 600 g will smeet of mononitrides uranium and plutonium for 15 hours of continuous operation. The dissolution rate was 0.4 0.45 g/cm2o'clock In the resulting cathode product contained 564 g of uranium and plutonium, and 245 g (30%) salt electrolyte. The extraction of uranium and plutonium during electrolysis was 95 97%
From the obtained uranium and plutonium method of hydrogenation and nitriding was again obtained nitride to produce fuel pellets. Electrochemical processing smelting and separation did not practically affect the quality of remanufactured cores (for example, the increase in oxygen occurred from 0.1% to 0.15 wt%).Example 2.When carrying out the regeneration process of example 1 PU stands for "solid" electrode is the cathode in the form of fine powder, which leads to significant losses of plutonium. For more complete extraction of uranium and plutonium second electrode, made of a fusible metal, such as cadmium, are also connected to the negative pole of the current source and create it a current density of 0.1 A/cm2. At the anode the decomposition of mononitride is and fusible plutonium with the formation of an alloy with the metal electrode.It was found that increasing the current density on the fusible electrode
the cathode of more than 0.1 A/cm2is to restore it along with plutonium, uranium, resulting in solidification of the alloy and the selection of the metal uranium and plutonium in the form of fine powder that prevents a more complete extraction of uranium and plutonium.Example 3.In the melt mixture of salts (see example 1), heated over 700oC, electrolysis was performed when indicated in example 1 modes and determined that the output of uranium and plutonium current has not changed. But increased consumption and corrosion of the equipment and require special cooling system sealed assemblies of equipment.Example 4.In order to improve the contact of the fuel with the container, the retention of fission products and localization slurry the residue in the container has napravila layer of fusible metal (cadmium, tin or lead) or alloy, more electropolishing than uranium nitride.The pyrochemical process of regeneration was carried out at current density at the anode is not more than 0.3 A/cm2on hard the cathode is not more than 0.4 A/cm2and fusible (cadmium) the cathode is not more than 0.1 is, for example by transferring them to a container of fuel pellets, resulting in improved electrical contact of the tablet container and stopped the removal of the talus with tablets in the electrolyte, which contributed to more complete fuel processing and removal of PD.Direct extraction of uranium and plutonium amounted to 93.7% of the Rest of the nitrides in the container, and the electrolyte is not found.Example 5.To increase the degree of regeneration of the fuel after being in the container-anode tablets nonconforming mixed nitride fuel dissolved in the electrolyte under the same conditions described in example 4, to produce a partial dissolution at the anode weld on him fusible metal, saturate it with salt electrolyte and thereby provide further separation of the uranium and plutonium on fusible electrode is the cathode. 1. The pyrochemical method of regeneration of nuclear fuel, which consists in the fact that in the molten eutectic mixture of salts of potassium chloride and lithium enter the uranium chloride, immerse the container with nuclear fuel and two electrodes, one of which is made of fusible metal, then connect the container with the positive and negative electrodes with the poles of the current source and spend Elke emitting pure metals, characterized in that the container load nitride fuel and use the electrolyte, which includes at least 15 wt. trichloride uranium, heat the electrolyte to a temperature of at least 600oWith and build on the container-anode current density of not more than 0.3 A/cm2and on the electrode-the cathode is not more than 0.4 A/cm2.2. The method according to p. 1, characterized in that the fusible electrode-cathode create current density is 0.1 A/cm2.3. The method according to PP. 1 and 2, characterized in that the melt is heated to a temperature not exceeding 700oC.4. The method according to PP. 1 to 3, characterized in that in the container with nitride fuel is placed fusible metal or alloy, more electrophoretically than uranium nitride.5. The method according to PP. 1 to 4, characterized in that after running out of fuel in the container dissolve placed in it more electrophoretically than uranium nitride, a low-melting metal or alloy and then precipitated onto fusible electrode-cathode uranium and plutonium.
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
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: 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
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 420-550°C while melt zinc ions are used as oxidiser.
EFFECT: higher yield.
2 tbl, 2 ex
FIELD: technological processes.
SUBSTANCE: invention relates to a method and a device for bringing two immiscible fluids into contact. Method of binging into contact without mixing of the first substance consisting of metal or alloy of metals in liquid state and the second substance consisting of salt or salt mixture in liquid state, in which: the first substance in solid state is placed in the first container, the first container is put into contact with the second substance in solid state placed in the second container, the first and the second containers are exposed to electromagnetic field effect, the first substance in liquid state is brought into motion, the second substance in solid state starts melting under the effect of heat flow from the first container, the second substance in liquid state is brought into motion, the first substance in liquid state stays in contact with the second substance in liquid state for a certain period of time, the first container is removed from the second substance in liquid state, the first container is cooled until the first substance returns into solid state.
EFFECT: improved mass transfer kinetics.
35 cl, 10 dwg, 2 tbl
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