Method for preparation of spent nuclear fuel reprocessing solutions containing complexing agents for extraction of multivalent actinides
FIELD: power industry.
SUBSTANCE: method for preparation of spent nuclear fuel reprocessing solutions containing complexing agents for extraction of multivalent actinides at suppression of action of complexing agents consists in introduction to a solution of nitric-acid solutions of transient metals that fix complexing impurities better than plutonium does. As complexing agents, the solution can contain ethanedioic acid, mellitic acid and other polybasic acids and oxygen acids, DTPA and EDTA. As added binding agents, there used are nitric-acid solutions of molybdenum and/or zirconium, including spent nuclear fuel solution based on uranium-molybdenum alloys introduced in equimolar amounts or amounts close to them as to metal: complexing agent ratio.
EFFECT: invention allows extracting multivalent actinides from spent nuclear fuel solutions containing complexing agents applying non-destructive methods and without strong change of reagent medium.
The invention relates to the field of processing of spent nuclear fuel (SNF) and can be used in preparing solutions reprocessing of spent fuel, containing complexing substances for extraction extraction multivalent actinides.
When the SNF is a sufficiently large number of industrial solutions containing complexing agents, which are either formed by dissolving the spent nuclear fuel, for example, carbide SNF, or were introduced in the process for separation of components, in particular, diethylenetriaminepentaacetic acid (DTPA) or ethylenediaminetetraacetic acid (EDTA), or formed after the deposition of the elements, for example, after oxalate deposition of plutonium.
These substances form strong complexes with plutonium, which makes it impossible for the handling of these solutions without processing or preparation, which eliminates complexes the effect of these substances.
The usual method of preparation of these solutions prior to the extraction of actinides using diluted tributyl phosphate (TBP) (PUREX process) is the oxidation treatment for deep destruction of the complexing agents, which is typically in the presence of catalysts. In particular, there is a method of destruction of the oxalate ion in the mother solution after precipitation of plutonium put the m prolonged heating (boiling) solution in the presence of manganese(P) as a catalyst (Koltunov B.C. Kinetics and mechanism of oxidation of oxalic acid by nitric acid in the presence of ions of Mn2+. Kinetics and Catalysis, 1968, v.9, No. 5, s-1041).
For the destruction of DTPA or EDTA virtually no satisfactory simple ways, and their harmful effect is suppressed by the strong acidification of the solution (Even S.I., Nedelina L.V., Cockroaches V.M. and others About the specifics of the separation of plutonium and neptunium in terms of reprocessing spent nuclear fuel complex RT-1. / Radiochemistry. 1998. t, No. 4. s-335).
Closest to the present invention is a method of preparing a solution vysokobarievogo carbide mixed uranium-plutonium spent fuel ozonation for extraction extraction multivalent actinides (Natarajan R. Reprocessing of FBTR mixed carbide fuel - some process chemistry aspects. Proc. 16thAnn. Conf. Ind. Nucl. Soc. INSAC-2005. Mumbai, 2005. Paper IT_21.). In this process, which is taken as a prototype, plays the role of a catalyst dissolved shrapnel cerium.
The disadvantage of this method is not only the necessity of profound destruction Melitopol, oxalic and other polybasic carboxylic acids and hydroxy acids, but also the transition of PU(IV) in less extractable state of PU(VI), which requires additional redox processing.
The invention solves the problem of preparing solutions of reprocessing spent fuel, containing the comp is EcoObraz substances, for extraction extraction multivalent actinides, using non-destructive methods and without dramatic changes in the chemical environment.
To achieve the mentioned technical result in the above method is proposed to conduct training solutions reprocessing of spent fuel, containing complexing substances for extraction extraction multivalent actinides, including from solutions after dissolution of SNF-based metal carbides, by introducing into the extraction process nitric acid solutions of transition metals, linking interfering impurities more heavily than plutonium. The additives are used nitric acid solutions of molybdenum and/or zirconium, including the solution of spent fuel based on uranium-molybdenum alloys. "Gasim" complexing agents are oxalic acid, malletova and other polybasic acids and oxyacids, DTPA and EDTA. There is also the mother liquor from oxalate precipitation of plutonium without or after processing. The order of mixing of solutions with reactive components is arbitrary, i.e. binders are introduced in the source solution and/or in any area of the brain extractor, in equimolar or related quantities with respect to the complexing agents.
The hallmark of the proposed method is what I use multivalent transition metal, having the ability to bind the above complexing agents in connection stronger than their complexes with plutonium, and for these purposes may be a suitable industrial solutions.
In particular, the oxidation catalytic processing oxalate mother solutions destroys only free oxalic acid, whereas for the destruction of the complexes of plutonium system have strongly acidified with nitric acid. This is not required if you enter in a solution of 0.5-1 g/l Zr; then Pu easily extracted with the natural acidity of stock solution 2 mol/L.
A similar effect is observed in the case of extraction of macroscopic quantities of plutonium from its reextract the 1st cycle containing DTPA. Instead acidification of the solution to 3.5-4 mol/l HNO3in the process of the extraction of Pu at 1.5 mol/l HNO3after extracting the main mass of plutonium to enter into solution the required amount of Zr, and produce additional extraction of Pu. You can also directly enter Zr in the original solution.
When the reprocessing of spent fuel based on uranium carbide, the best solution is the introduction of compounds of molybdenum to link Melitopol and other polybasic acids, resulting from the dissolution of carbide SNF. For these purposes may be used a solution of spent fuel based on uranium-molybdenum alloy.>
Thus, it simplifies the processing of SNF solutions of complex composition and utilization of industrial tailings solutions.
The ability of the proposed technical solutions of the following examples.
When extraction disposal mother solutions from oxalate precipitation of plutonium containing 50-100 mg/l Pu and 10 g/l H2C2O4in 2 mol/l HNO3the distribution coefficient of Pu in 30% TBP equal to 0.2. By mixing this solution with a solution of SNF SBA based on uranium-molybdenum alloy containing 150 g/l U, 1 g/l Pu and 13.5 g/l Mo 5 mol/l HNO3based Mo:H2C2O4=1 (1.35 amount of oxalate stock solution to 1 volume of solution SNF SBA) receive a solution containing 64 g/l U, ~0.5 g/l Pu and 5.7 g/l Mo 3.3 mol/l HNO3. Last sent for recycling in PUREX process, and the residual Pu content in the raffinate is not more than 1 mg/l and U are not more than 10 mg/L. at the same time eliminates the formation of deposits on the basis of Mo.
The solution of spent fuel based on uranium carbide containing 50 g/l U 0.5 g/l of Pu and other elements in 5 mol/l HNO3cannot be processed without training due to the low distribution coefficient of Pu. This solution is mixed with a solution of SNF SBA based on uranium-metal alloy (see Example 1) at the rate of 3:1 by volume, after which the joint is attached, the solution is sent to PUREX-process moreover, the residual Pu content in the raffinate is not more than 1 mg/l and U are not more than 5 mg/L.
When extraction disposal of treated uterine fluids from oxalate precipitation of plutonium containing 50-100 mg/l Pu and 1.5 g/l DTPA and 0.5 g/l H2C2O4in 2 mol/l HNO3, unrecoverable balance is 25 mg/l Pu. By mixing this solution with a solution of the VVER spent fuel with a burnup of 40 GW*d/t, containing 300 g/l U, 2.7 g/l Pu and 1 g/l Zr, 0.6 g/l Mo and other elements in 3 mol/l HNO3on balance processing gain solution containing 250 g/l U and ~2.5 g/l Pu in 3 mol/l HNO3. Last sent for recycling in PUREX process, and the residual Pu content in the raffinate is not more than 1 mg/l and U are not more than 10 mg/L.
In refining the extraction comes reextract the plutonium from the 1st cycle PUREX process containing 5 g/l Pu, 1.5 mol/l HNO3and 1.5 g/l DTPA. After redox treatment solution the oxidation state of the elements of Pu+4 and Np+5. When extraction in 30% TBP the distribution coefficient of Pu in successive contacts is 0,77; 0,68; 0,58; 0,29; 0,11. Non-extractable residue Pu is 0.28 g/l After adding to this solution, 3 g/l Zr distribution coefficient of Pu increases to 5.3, which corresponds to the tabular value, and non-extractable residue is reduced to 3 mg/L.
2. The method according to claim 1, characterized in that the insertion of the binder components used nitric acid solutions of molybdenum and/or zirconium, including the solution of spent fuel based on uranium-molybdenum alloys.
3. The method according to claim 2, characterized in that the binder is introduced into the initial solution and/or in any area of the brain extractor in equimolar or related quantities with respect to the complexing agents.
4. The method according to claim 3, characterized in that as complexing agents in the solution containing oxalic acid, malletova and other polybasic acids and oxyacids, DTPA and EDTA.
5. The method according to claim 4, characterized in that the solution complexes with substances is the mother liquor from oxalate precipitation of plutonium.
6. The method according to claim 1 or 2, characterized in that the extraction of macroscopic quantities of plutonium from solution processing of spent nuclear fuel containing DTPA, the solution of zirconium is introduced into the middle C is by extraction after extraction of "free" plutonium.
FIELD: power engineering.
SUBSTANCE: calcination of a solution of radioactive wastes (RAW) is carried out in a microwave plasma reactor, then a homogeneous glass melt is produced in a frequency melter of direct induction heating. The method includes supply of the RAW solution into a zone of electrothermal processing, which comprises a zone of plasma microwave processing of the RAW solution in the water and vapour plasma and a zone of bath processing of the melt produced by direct induction heating of inorganic RAW ingredients, melting and electromagnetic mixing of inorganic RAW ingredients, continuous removal of the melt, cooling of the gas flow, cleaning of the gas flow from volatile products of RAW decomposition and from process dust. The device for realisation of the method comprises a plasma chamber, the upper part of which is made in the form of a truncated cone, equipped with an all-metal microwave plasmatron, which generates a flow of water and vapour plasma, a unit of RAW solution supply, a frequency melter of direct induction heating for melting and homogenisation of inorganic RAW ingredients, equipped with a pipeline for melt drainage, a collector - an accumulator of glass melt, a pipeline for gas flow transportation for cleaning.
EFFECT: solving the problem of complex environmentally and technical safe processing of RAW.
14 cl, 2 dwg
FIELD: power industry.
SUBSTANCE: invention refers to processing technology of high-salty liquid radioactive wastes of low and medium activity level, which contain up to 30% of organic substances by their being added to magnesite cement. Composite material has the following composition: loose dead-burned magnesite caustic powder - 27-28 wt %, hard salts - 5-6 wt %, calcium chloride (CaCl2) - 0.1-6 wt %, catalytic carbon-bearing additive - 0.1-0.2 wt %; potassium ferrocyanide solution - 0.05-0.1 wt %; and nickel nitrate solution - 0.05-0.1 wt %, and liquid radioactive wastes are the rest. The following sequence of ingredients is added to liquid radioactive wastes: hard salts, potassium ferrocyanide solution, nickel nitrate solution, calcium chloride, catalytic carbon-bearing additive, and loose dead-burned magnesite caustic powder. The invention allows obtaining compounds meeting the main requirements of their quality as per GOST R 51883-2002 (cesium leaching rate -137 ≤1-10-3, achieved - 2-10-5g/cm2·day, and compressive mechanical strength ≥5 MPa), with filling of dry radioactive layers of up to 37 wt %.
EFFECT: compliance with the main requirements.
FIELD: process engineering.
SUBSTANCE: invention relates to treatment of radioactive fluid and pulpy wastes formed in recovery of radiated nuclear fuel. Proposed method comprises destructing oxalate ions in mother waters by nitric acid in the presence of variable-valency metal ions. Processing of oxalate mother solution and pulpy wastes comprises mixing mother solution with solid phase of hydroxide pulp.
EFFECT: power savings, decreased amount of radioactive wastes.
3 cl, 3 tbl
FIELD: process engineering.
SUBSTANCE: installation for removal of liquid radioactive wastes (LRW) from temporary storage reservoirs comprises floating platform arranged there inside and composed of a tank equipped with system of ultrasound radiators connected with ultrasound oscillation generator and remote control system. Said ultrasound radiators are regularly arranged on floating platform walls and bottom to disperse and dissolve the sediments and to displace the platform in preset direction. Installation is equipped with LRW lifting and discharging system comprising pump and pipelined and remote control and observation system. Besides, said installation is integrated with LRS treatment unit.
EFFECT: higher efficiency and safety.
11 cl, 2 dwg
FIELD: process engineering.
SUBSTANCE: invention relates to environmental protection against liquid radioactive wastes (LRW) that make byproducts of used nuclear fuel treatment or other industrial activities. Proposed method comprises converting LRW components into solid phase by processing them with silicon-bearing compounds of geothermal origin at 5-60°C. Dispersions of silicon dioxide spherulites are used as a hardener and produced by membrane concentration of natural geothermal solution, spherulite diameter making 4-150 nm at silicon dioxide concentration not lower than 105 g/kg, using microfibers from inorganic oxides, for example, basalt used in amounts of 0.5-5 wt % silicon dioxide dispersion weight.
EFFECT: higher safety.
4 dwg, 8 ex, 2 tbl
SUBSTANCE: disclosed is material which contains polyazacycloalkane which is grafted on polypropylene fibre, a method of producing said material and a method of removing metal cations from a liquid by bringing said liquid into contact with said material.
EFFECT: disclosed material combines excellent selectivity of binding heavy metals, lanthanides or actinides with excellent operational characteristics.
55 cl, 6 dwg, 9 tbl, 8 ex
SUBSTANCE: method of processing spent ion-exchange resins contaminated with radioactive elements involves wet grinding of resin grains to particle size 1-45 mcm, adding alkali to the obtained suspension to pH 10.5-11.0, liquid-phase oxidation of the suspension while feeding air into the oxidation zone under conditions of supercritical state of water at temperature 450-550°C and pressure 230-250 atm, removing gaseous oxidation products in form of CO2 and N2, separating the solid and liquid phases by filtering and subsequent deactivation of the liquid phase.
EFFECT: invention enables to reduce the volume of radioactive wastes for permanent storage, is characterised by absence of secondary gaseous wastes and does not require use of aggressive chemicals.
5 cl, 1 ex, 1 tbl
SUBSTANCE: method of processing a radioactive solution involves the following. First, an iron (III) compound in form of chloride or sulphate is added to the solution in amount of 0.04-0.05 mol/l to form an iron-containing precipitate. At the first step, a minimum amount of mineral acid - hydrochloric or sulphuric acid - is added, and at the second step 0.18-0.24 g-eq/l of the corresponding acid is added to the solution. The solution is held for not less than 120 hours at room temperature or not less than 18 hours at 70-95°C and sodium sulphide is added to the solution in a molar amount which is 1.5 times greater than the amount of the added iron (III) compound to form a basic collective precipitate of radionuclides of cobalt and caesium and a mother solution containing an organic complexing agent and a residual amount of radionuclides of cobalt and caesium. The mother solution is subjected to a post-treatment cycle by adding an iron (III) compound in amount of 0.02-0.04 mol/l with respect to iron (III) and mineral acid in an amount which is equivalent to content of sodium in the added sodium sulphide, holding the obtained mother solution and adding additional sodium sulphide in molar amount which is 1.5 times greater than the amount of the additionally added iron (III) to form an additional collective precipitate of the post-treated mother solution.
EFFECT: invention enables to increase processability of the method by replacing oxidation of the organic complexing agent with cationic substitution of the cobalt radionuclide therein, reduce the amount of reagents used while ensuring high degree of purification of solutions.
6 cl, 4 ex
FIELD: power industry.
SUBSTANCE: water treatment method of natural or artificial water reservoir from radioactive isotopes and harmful chemical substances involves the intake of source water, its pre-treatment and the main treatment by two-stage reverse osmosis so that filtrate is obtained, which is supplied to consumer as treated water and concentrate returned to water reservoir. Combined concentrate obtained at the stage of pre-treatment and the first stage of reverse osmosis is returned to water reservoir, and concentrate obtained at the second stage of reverse osmosis is supplied to the first stage of reverse osmosis.
EFFECT: invention allows improving the efficiency of treatment procedure and reducing the amount of secondary wastes.
6 cl, 1 dwg, 1 ex, 1 tbl
FIELD: water treatment.
SUBSTANCE: innovation relates to the chemical treatment of industrial and domestic wastewaters containing lubricating and cooling fluids, radioactive contaminating substances, washing solutions, ions of heavy metals; the method includes the decontamination and demulsification of wastewater by the mixture of ferrous chloride (FeCl3), calcium chloride (СаСl2) and permanganate of alkalic or the mixture of alkaline-earth metals at the mass ratio of (1/2):(0.15/0.2):(0.01/0.02) on conversion to arid salts and volume concentration of 1-3% of the initial solution, рН counts to 5-6.
EFFECT: innovation enhances the efficiency of wastewater purification.
1 cl, 5 tbl
FIELD: chemical technology; recovery of deactivated and decontaminated radioactive industrial wastes.
SUBSTANCE: proposed method that can be used for deactivating and decontaminating industrial radioactive wastes incorporating Tb-232 and their daughter decay products (Ra-228, Ra-224), as well as rare-earth elements, Fe, Cr, Mn, Sl, Ti, Zr, Nb, Ta, Ca, Mg, Na, K, and the like includes dissolution of wastes, treatment of solutions or pulps with barium chloride, sulfuric acid, and lime milk, and separation of sediment from solution. Lime milk treatment is conducted to pH = 9 - 10 in the amount of 120-150% of total content of metal oxyhydrates stoichiometrically required for precipitation, pulp is filtered, and barium chloride in the amount of 0.4 - 1.8 kg of BaCl2 per 1 kg of CaCl2 contained in source solution or in pulp, as well as pre-diluted sulfuric acid spent 5 - 20 times in chlorine compressors in the amount of 0.5 - 2.5 kg of H2SO4 per 1 kg of BaCl2 are introduced in filtrate. Alternately introduced in sulfate pulp formed in the process are lime milk to pH = 11 - 12, then acid chloride wash effluents from equipment and industrial flats at pulp-to-effluents ratio of 1 : (2 - 3) to pH = 6.5 - 8.5, and pulp obtained is filtered. Decontaminated solution is discharged to sewerage system and sediment of barium and calcium sulfates and iron oxysulfate are mixed up with oxyhydrate sediment formed in source pulp neutralization process; then 35 - 45 mass percent of inert filler, 10 - 20 mass percent of magnesium oxide, and 15 -m 25 mass percent of magnesium chloride are introduced in pasty mixture formed in the process while continuously stirring ingredients. Compound obtained is subjected to heat treatment at temperature of 80 - 120 oC and compressed by applying pressure of 60 to 80 at.
EFFECT: reduced radioactivity of filtrates upon separation of radioactive cakes due to enhanced coprecipitation of natural radionuclides.
7 c, 1 ex
FIELD: recovery of liquid radioactive wastes.
SUBSTANCE: method for extracting nitric acid from solution includes bringing solution in contact with nitrogen-containing agent and separating the phases. For the process use is made of organic nitrogen-containing material forming poorly soluble sediment together with nitric acid. Urea nitrate sediment deactivating method includes treatment of inert nozzle in fluidized bed at temperature of 750 - 800 °C with fuel combustion products having residual oxygen content of 2 - 3 volume percent.
EFFECT: reduced cost.
7 cl, 5 ex
FIELD: radioactive waste treatment.
SUBSTANCE: suspension of magnetite obtained via electroerosion dispersing of iron-containing raw material in distilled water is added to solution to be processed. Adjusting pH of solution to 11-12 leads to precipitation. Decanted solution is subjected to magnetic separation followed by ion-exchange purification.
EFFECT: enhanced purification efficiency.
3 cl, 1 tbl
FIELD: methods of liquid radioactive wastes processing.
SUBSTANCE: the invention is pertaining to the field of liquid radioactive wastes processing. The invention presents a method of neutralization of the low-mineralized and medium-mineralized low-active liquid wastes in the field conditions, which includes the liquid wastes purification by mechanical filters and ultrafilters. The subsequent desalination is conducted by reverse-osmotic filters and an after-purification - by ion-exchange filters with a reactant treatment of the spent ion-exchange resins using potassium ferrocyanide and cobalt salts. Then the treated resin is used as a sorption prefilter, in which they use purification of the wastes before their feeding to the ion-exchange filter. The formed secondary A-wastes are fixed in the stable medium. Advantages of the invention consist is an improved purification efficiency and reduction of the secondary wastes volume.
EFFECT: the invention ensures improved purification efficiency and reduction of the secondary wastes volume.
FIELD: recovery of radioactive wastes.
SUBSTANCE: proposed method for matrix immobilization of industrial wastes includes preparation of source solution of industrial wastes and impregnation of ceramic matrix with this solution followed by roasting this matrix; source liquid radioactive wastes used for the purpose are first treated with promoter crystallization solution doped with oxide-forming admixtures whereupon radioactive wastes are introduced in ceramic matrix and roasted using microwave energy at temperature of 900 - 1 000 °C. Such procedure provides for recovering great amount of radioactive wastes included and chemically bonded in ceramic matrix which makes it possible to reduce leaching and to enhance matrix strength and life.
EFFECT: facilitated procedure, reduced cost, enhanced quality of radioactive waste immobilization and environmental friendliness.
FIELD: technology of handling of the liquid nuclear wastes of the nuclear fuel and power cycle; methods of reprocessing of the liquid nuclear wastes.
SUBSTANCE: the invention is pertaining to the procedure of the liquid nuclear wastes handling of the nuclear fuel and power cycle and may be used during reprocessing of the liquid nuclear wastes (LNW). The method includes the preliminary concentration, ozonization, microfiltration of the vat residue with fractionation of the permeate and the concentrate and the ion-selective purification of the permeate using the ion-selective a sorbent. At that the microfiltration is conducted at least in two stages: the permeate of each previous stage of the microfiltration is directed to the microfiltration as the source solution for the subsequent stage of the microfiltration, and at the final stage of the permeate from the microfiltration is sent to the utilization. The concentrate produced at each next stage of the microfiltration is mixed with the source solution of the previous stage of the microfiltration. The concentrate produced at the first stage of the microfiltration is directed to the conditioning and dumping. The ion-selective sorbent is added in the permeate of the previous stage of the microfiltration before the final stage of the microfiltration. The invention ensures: reduction of the volume of the liquid nuclear wastes due to the deep purification of the LNW with the high saline share from the radionuclides and extraction of the radionuclides in the compact form of the sparingly soluble compounds at the corresponding increase of the factor of purification of the salts extracted at the stage of the vat residue treatment; reduction and optimization of the consumption of the permeate and concentrate interacting with the source solution as well as produced at the further stages.
EFFECT: the invention ensures: reduction of the volume of the liquid nuclear wastes; the deep purification of the LNW with the high saline share from the radionuclides and extraction of the radionuclides in the compact form of the sparingly soluble compounds at the corresponding increase of the factor of purification of the salts extracted at the stage of the vat resudue treatment; reduction and optimization of the consumption of the permeate and concentrate interacting with the source solution ands produced at the further stages.
FIELD: environment protection against radioactive pollutants; immobilization of nuclear radioactive wastes.
SUBSTANCE: proposed method for solidifying liquid radioactive wastes includes their spray drying and calcination, calcination product vitrification using flux dopes, melt draining to tank, and annealing of solid block. Liquid radioactive wastes are heated in advance in continuous flow to evaporate 30-80 percentage ratio of water contained in them, and steam-water mixture produced in the process is sprayed under its own pressure in chamber heated to temperature of 600-800 °C. Solid particles of calcination product are separated from steam-gas mixture by sedimentation at temperature below 300 °C and filtration.
EFFECT: extended service life of equipment, enhanced productivity and radiation safety for personnel handling highly radioactive products.
FIELD: nuclear engineering; preservation of dry, wet, and liquid radioactive wastes.
SUBSTANCE: proposed composition has resin portion of cold-cured compound ATOMIK and filler. Resin portion ingredients are as follows, parts by weight: epoxy oligomer, 100;, curing agent (aromatic amines), 38-50 furfural, 9-11. Used as filler is composition incorporating following ingredients, parts by weight: cement, 50-100; marshalite, 50-100 or bentonite, 50-100, or when they are jointly used: marshalite, 90-100 and bentonite, 90-100. Such composition provides for desired radiation resistance in absence of leaching of alpha-, beta-, and gamma-active radionuclides from preserved specimens of reactor graphite.
EFFECT: enhanced radiation stability of preserved specimens free from pits and voids, and adequate lifetime of preserved wastes; ability of their depreservation.
1 cl, 2 dwg, 2 tbl, 2 ex
FIELD: treatment of radioactive effluents and solid-phase saturated waters.
SUBSTANCE: some portion of organic fraction is reduced in first reactor by way of biological aerobic treatment. Filtrate/permeate taken from tangential filtering device is either directly used or supplied to first or next reactor. Solid phase is gravitationally extracted within tank of partial-flow filtering device and compacted in bottom region; concentrated effluents flowing from tangential filtering device are fed in next sedimentation region which is above first sedimentation region or above next one through intake channel; then effluents flowing above or from one side of sedimentation region are discharged through branch channel.
EFFECT: ability of selecting and technically optimizing separate modules.
34 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