Method of treatment of radioactive solutions and unit for its implementation
FIELD: physics, nuclear.
SUBSTANCE: offered group of the inventions relates to the devices for processing of radioactive solutions. In the offered method of processing of radioactive solutions before filling of the vessel with the solution in its bottom place an additional vessel from a thin dielectric film is placed. Then the radioactive solution is poured into the vessel, with adding of the substances for treatment process control. After that the solution is exposed to irradiation by unipolar electromagnetic impulses with the power more than 1 MW and with duration less than 1 ns, minimum frequency 1 kHz. The solution is treated within 10-30 minutes, held in the vessel during 1-4 days, then the treated solution is drained and additional vessel is removed for utilization. The offered unit contains a current-conducting housing (1), the central part of which contains electrode (2), designed as a horizontal plate. The plate repeats the housing cross-section shape, but have the sizes 20-30% from the cross-section area of the housing. The unit also contains the generator of unipolar electromagnetic impulses, located outside the housing (3). During the treatment at the bottom of the housing the additional vessel (4) is placed.
EFFECT: simplification of devices for treatment of radioactive solutions with keeping of high quality of cleaning.
2 cl, 1 dwg
The invention relates to the field of processing of materials containing radioactive substances, namely, the means of processing of radioactive solutions, and aqueous solutions by electrochemical methods, and can find application in the nuclear industry and waste water treatment.
A known method of purification of liquid waste from ions of heavy metals and radioactive isotopes that are described in the same patent RF №2127459 on CL G21F 9/06, Appl. 25.12.97, publ. 10.03.99.
The known method includes the electrochemical treatment of waste by oxidation of complex compounds of heavy metals and radioactive isotopes on the electrodes using as the cathode gas diffusion electrode, and the anode - narashimha electrode, followed by deposition of isotopes of heavy metals and radioactive ions with collectors and separating the slurry phase from the solution.
The disadvantage of this method is the need for a gas diffusion electrode complex structures.
A method of processing ammonium sulfate radioactive solutions described in the patent of Russian Federation №2271587 on CL G21F 9/16, 9/20, Appl. 06.11.2003, publ. 10.03.2006.
The known method includes galvanomagnetic processing, in which the radioactive solution is passed through vibramicina galanopoulou iron-coke, is the orrection pH of radioactive solutions in two stages with separation of the precipitate after each stage, when this precipitate first-stage adjustment of the pH is mixed with montmorillonite clay, pressed pellets, spend their drying and subsequent calcination to obtain glass ceramics in which the immobilization of sludge containing radionuclides, and the filtrate is first treated in the froth layer for Stripping ammonia and then passed through a natural ion exchanger.
The disadvantage of this method is the complexity of its implementation because of its multi-stage system and a large number of operations.
The known method and device for processing liquid radioactive waste described in the patent of Russian Federation №2116680 "Installation for decontamination of liquid radioactive waste" CL G21F 9/06, Appl. 24.06.94, publ. 27.07.98.
The known method is that in the cavity of the processing chamber with the capacity of the filter loaded polymeric polyelectrolyte hydrogel creates between the chamber walls and capacity filter electric field of 0,001-0,0001/m, using them as electrodes serving of liquid radioactive waste (LRW) for processing by irradiating them with light radiation in the infrared range and spraying their sprinkler on polymer polyelectrolyte hydrogel, before it is controlled by weight saturation to 1:50-1:100, where 1 is the weight of the hydrogel, then subjected to the processed amount of liquid radwaste radiation of the acoustic wave is in the range of 25-45 kHz at a temperature of from 0 to 10°C for intensification of the process of separation LRW water and radioactive waste, that when they settle in the hopper is subjected to a local area heating to reduce the content of moisture and is directed to a container for further processing.
Known device includes a housing in which the camera is to be processed, having a perforated wall, the electrodes to create an electric field, the cavity of the chamber connected to the tank for draining purified water, with the camera placed perforated the capacity of the filter, and the installation is also equipped with emitters of the acoustic waves, the emitters of light waves and heat exchanger, and the chamber is filled polyelectrolyte polymer of the hydrogel, and perforated the capacity of the filter is located in the environment of the hydrogel, also connected by railway with the capacity for discharge of treated water in the upper part of the body is a tube, equipped with a perforated nozzle for spraying the liquid raw on the surface of the processing chamber, and the cavity of the processing chamber and the body cavity is connected to the vacuum unit by means of a branch pipe. At the same time as the acoustic emitters use the emitters of ultrasonic waves, and as electrodes for radiation electric fields use the chamber wall and the perforated wall of the vessel filter.
The disadvantage of this method is the complexity of the process, due to the use is the use of light irradiation, spray, acoustic treatment, heating and so on, which complicates the construction of the installation.
The known method and installation for processing of radioactive solutions, described in the patent of the Russian Federation No. 2319237, Appl. 13.06.2006, publ. 10.03.2008 and selected as a prototype.
The known method is that in the radioactive solution when applying to the processing type of the chemical elements and substances to control the process, the solution was heated and sprayed, then put it under the influence of an electric field in the processing chamber simultaneously with the irradiation of a solution of a unipolar electromagnetic pulses with a capacity of more than 1 MW and a duration less than 1 NS, repetition rate of not less than 1 kHz.
The known apparatus comprises a body placed in it by the electrodes to create an electric field, one of which is the case, the emitter of electromagnetic radiation, a heating device, a device for dispersion of radioactive solution, located in the upper part of the body located outside the casing and connected with it the unipolar generator of electromagnetic pulses with a capacity of over 1 MW, pulse duration of less than 1 NS and a repetition frequency of not less than 1 kHz, the heating device is connected through the pump and nozzle with atomizing device constituting the nozzle block, the second is electrode made in the form of a rod with the needle, located in the Central part of the body that serves as the emitter and connected to the generator.
Known tools provide high quality treatment, however, are very complex, due to the way the need for heating, spraying with careful filtering solution, and device - appropriate devices: device heating, and a spraying device, requiring the use of a filter.
The goal is to simplify processing tools providing high-quality cleaning.
The problem is solved by the fact that:
in the method of processing radioactive solutions, which consists in the fact that the radioactive solution is poured into the container, add the chemical elements and substances to control the process, and then subjecting it to irradiation unipolar electromagnetic pulses with a capacity of more than 1 MW and a duration less than 1 NS, repetition rate of not less than 1 kHz, according to the invention, the irradiation of the solution is carried out using an electrode made in the form of a horizontal plate, before lling solution in its lower part placed additional capacity of thin dielectric films, then pour the solution with the chemical elements and substances for process control processing, handle it in the course the e 10-30 minutes, maintain in capacity within 1-4 days, after which the treated solution is drained and removed the extra capacity, which is subjected to burial;
in an apparatus for handling radioactive solutions containing the body placed in it by the electrodes, one of which is the case, the emitter of electromagnetic radiation located outside the casing and connected with it the unipolar generator of electromagnetic pulses with a capacity of over 1 MW, pulse duration of less than 1 NS and a repetition frequency of not less than 1 kHz, while the second electrode is made in the form of a rod located in the Central part of the body that serves as the emitter and connected to the generator, according to the invention the casing is made of conductive material, the electrode is made in the form of a horizontal plate, the shape of the cross section of the casing and having a size of 20-30% from the sectional area of the casing, which is made with possibility of accommodation in its lower part at the time of treatment and disposal after processing additional capacity of thin dielectric films.
Use in planographic emitter, the shape of the cross section of the hull, but the smaller size allows for more efficient handling of the feed solution, which together with the use in the method of processing the additional capacity of the thin dielectric plait the key at the bottom of the housing, carrying out processing for 10-30 minutes and subsequent storage of the solution in the tank within 1-4 days provides the deposition of additional capacity generated in the process of precipitation, which is then removed together with the additional capacity and landfilled, and enables you to simplify the process and high quality cleaning.
Running in the electrode is in the form of a plate, the shape of the cross section of the body having a size of 20-30% of the sectional area of the housing, makes it possible to increase the size and efficiency of processing, in conjunction with the execution of the body with the possibility of accommodation in its lower part on the processing time and its subsequent removal after the additional processing capacity of the thin dielectric film allows a simpler design of the installation to obtain a high quality.
The technical result - the simplification of the processing of radioactive solutions while providing high quality cleaning.
The inventive method has the novelty in comparison with the prototype, differing from it in such essential characteristics as the electrode is in the form of a flat plate in the shape of the cross section of the casing and having a size of 20-30% of the sectional area of the body, placing at the bottom of the casing installation per the d filling it with a solution of additional capacity of thin dielectric films, the next Bay solution with chemical elements and substances for process control processing, processing it within 10-30 minutes, keeping the body within 1-4 days, discharge of the treated solution and the removal of additional capacity, which is subjected to disposal, providing collectively achieve the specified result.
The inventive installation is new in comparison with the prototype, differing from it in such essential characteristics as the execution of the electrode in the form of a flat plate in the shape of the cross section of the body having a size of 20-30% of the area of its cross section, running the body with locations in its lower part while processing the additional capacity of the thin dielectric film and its removal after processing, provides collectively achieve the specified result.
The applicant unknown solutions having the above salient features that collectively achieve the specified result, he felt, therefore, that the inventive method and installation for processing of radioactive solutions meet the criterion of "inventive step".
The proposed drug for the treatment of radioactive solutions can be widely used in the nuclear industry, and therefore is relevant to the comfort of the criterion of "industrial applicability".
The invention is illustrated in the drawing, which shows a functional diagram of the installation.
The inventive method of processing radioactive solutions is as follows.
Before lling solution in its lower part placed additional capacity of thin dielectric films. Then the radioactive solution is poured into the container, add the chemical elements and substances to control the process. After this, the solution is subjected to irradiation of a unipolar electromagnetic pulses with a capacity of more than 1 MW and a duration less than 1 NS, repetition rate of not less than 1 kHz. The solution process for 10-30 minutes, soak in capacity within 1-4 days, after which the treated solution is drained and removed the extra capacity, which is subjected to disposal.
The inventive system includes a conductive housing 1, in the Central part of which is placed an electrode 2 made in the form of a horizontal plate, the shape of the cross section of the body having a size of 20-30% of the sectional area of the housing, and located outside the housing of the generator 3 unipolar electromagnetic pulse of a power exceeding 1 MW, pulse duration of less than 1 NS and a repetition frequency of at least 1 kHz. While the electrode 2 is connected to one of the terminals of the generator 3 and the housing 1 to the second output of the generator 3. At the time of processing in the lower part of the housing 1 an additional capacity of 4 out of thin dielectric films.
The inventive method and installation are used for handling radioactive solutions as follows.
In case 1 place the additional capacity of 4 of the dielectric thin film, for example, of polyethylene. Pour in the housing 1 of the processed solution, add chemicals, such as acids or alkalis or salts. Then the solution is subjected to a through electrode 2 irradiation for 10-30 minutes unipolar electromagnetic pulses with a capacity of more than 1 MW and a duration less than 1 NS, repetition rate of not less than 1 kHz, obtained from the generator 3.
Irradiation of a solution powerful short pulses with a large area of the electrode leads to the radiolysis of water, which formed a very active reagent hydrated electrons (eaq. The interaction of hydrated electrons from radioactive chemical elements contained in the solution, and with the included chemical reagents causes a change in the chemical composition of the solution and precipitation. While keeping the solution in the installation within 1-4 days resulting from precipitation process are deposited in the additional capacity of 4, which is then removed from the housing 1 and are buried together with the fallout-waste.
As an example of the operation of the plant can p is ivesti the following experience. In the case of a cylindrical shape with a diameter of 120 and a height of 190 mm, made of foil-coated glass was placed additional vessel made of plastic with a diameter of 110 and a height of 160 mm was bathed in an aqueous solution of radioactive137Cs volume of 1 litre.
To the solution was added alkali NaOH in the amount necessary to obtain the pH of a solution is equal to 10.
As an electrode used in a copper plate with a diameter of 30 mm. To the body and the electrode connected to the pulse generator with the following parameters: peak power 2 MW, pulse duration of 1 NS, pulse repetition rate 1 kHz. The processing time is 15 minutes After treatment observed precipitation.
The measured activity value of the aqueous solution in the vessel before treatment was 125 kBq/l, after treatment, after 2 days of 80 kBq/L.
In comparison with the prototype of the proposed processing tools are simpler and reduce the water activity of the solution.
1. The method of processing radioactive solutions, which consists in the fact that the radioactive solution is poured into the container, add the chemical elements and substances to control the process, then subjected to solution irradiation unipolar electromagnetic pulses with a capacity of more than 1 MW and a duration less than 1 NS, repetition rate of not less than 1 kHz, characterized in that blucina solution is carried out using an electrode, made in the form of a horizontal plate, the shape of the hull, before lling solution in its lower part placed additional capacity of thin dielectric films, then pour the solution with the chemical elements and substances to control process, process it within 10-30 minutes, soak in capacity within 1-4 days, after which the treated solution is drained and removed the extra capacity, which is subjected to disposal.
2. Installation for processing of radioactive solutions containing casing is placed in the Central part of the electrode is located outside the housing and connected with it one of the conclusions, and the second output from the generator electrode of a unipolar electromagnetic pulse of a power exceeding 1 MW, pulse duration of less than 1 NS and a repetition frequency of not less than 1 kHz, characterized in that the casing is made of conductive material, the electrode is made in the form of a horizontal plate, the shape of the cross section of the casing and having a size of 20-30% of the sectional area of the housing, which is made with possibility of accommodation in its lower part at the time of treatment and disposal after processing additional proportionate to the capacity of the flexible dielectric film.
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
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: method involves converting wastes to a gel-like state and is characterised by that solutions of highly active wastes are mixed with zirconium and iron salts and glycerine to concentration of said salts of not less than 0.12, 0.6 and 0.25 M/l respectively, holding the obtained mixture for not less than 2.5 hours, followed by adding to the mixture a solution of mono-substituted potassium phosphate in phosphoric acid to molar ratio of components Zr:Fe:K:PO4=1:3:2:5-8, drying, calcining the obtained polymer gel of zirconyl phosphate at 70-90°C and 300-400°C, respectively, and melting the obtained granules at 980-1000°C.
EFFECT: converting wastes into compact material which is suitable for long-term and safe storage.
3 cl, 2 tbl, 1 ex
SUBSTANCE: invention relates to hydrometallurgy of uranium and can be used to recycle mother solutions formed when producing uranium tetrafluoride from nitrate solutions via extraction, re-extraction and heat treatment of uranium compounds obtained from re-extracts to obtain uranium dioxide and further treatment thereof with chloride and fluoride solutions. The method of recycling mother solutions from production of uranium tetrafluoride involves mixing said solutions at pH 4.0-5.2 by bubbling air until pH stabilises and treating with sodium hydroxide at pH 10.5-11.0, separating the uranium-containing residues from the solutions and return thereof to the step of leaching raw products, settling the waste solutions in a tailing pond and pumping the remaining part of the solutions into the ground.
EFFECT: low consumption of nitric acid, sodium hydroxide and lime, reduced discharge of liquid wastes in the tailing pond.
3 cl, 6 tbl
FIELD: process engineering.
SUBSTANCE: invention relates to processing of heterogeneous liquid radioactive wastes, particularly, to processing of used fine abrasive filter materials and can be used for processing of waste filter perlite powder of special water treatment systems. Proposed method consists in extraction of filter perlite powder pump from storage tank, removal of excess moisture, transfer by hydrotransport, cementation, and adding ion exchange resins in amount of 10÷75% of filter perlite powder volume at density of 1÷1.5 g/cm3 to said pulp before transfer from storage tank.
EFFECT: 80-100 times decreased wear of equipment and pipelines.
SUBSTANCE: invention relates to processing liquid radioactive wastes formed when processing spent nuclear fuel. Described is a method of processing technetium solutions, which involves precipitation of technetium from nitrate solutions with concentration of nitric acid or the nitrate ion of not more than 3 mol/l, with concentrated aqueous solutions of o-phenanthroline or α-bipyridyl complexes of divalent transition metals, or mixed complexes of said organic compounds or mixed complexes containing o-phenanthroline or α-bipyridyl with dibasic amines. The obtained precipitates of organometallic pertechnetates are calcined in a hydrogen current at temperature of 600-1200°C with or without a low-melting metal or oxide thereof with melting point of 200-800°C to obtain stable matrices that are suitable for further storage and processing.
EFFECT: obtaining technetium in the final form which is suitable for further storage and processing.
5 cl, 2 tbl, 6 ex
FIELD: power industry.
SUBSTANCE: method provides for sedimentation of waste in an initial tank with draining of contaminants from surface to an oil product sump, pre-cleaning on mechanical bulk filters with modified nitrogen-containing coals and coarse and fine cleaning microfilters, softening and demineralisation on a reverse-osmosis filter with deposition of wastes in two intermediate tanks. Filtrate of reverse-osmosis filters is supplied for additional cleaning on ion-exchange filters, and concentrate is returned to the first intermediate tank before microfilters as an alkalising reagent prior to saturation as to salts with curing of formed radioactive concentrates by introduction to Portland cement. Coals saturated with oil products are replaced with new ones, and waste ones are burnt with oil products drained from the initial tank, including ash residue in Portland cement together with waste concentrates.
EFFECT: improving strength of cement stone by 1,5-2 times and reliable fixation of radionuclides in it.
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.
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: 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