Method for hardening liquid highly active wastes
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
The invention relates to a method of curing a liquid high-level waste resulting from reprocessing of nuclear fuel, with the aim of converting them into a compact material, suitable for long-term and safe storage.
Wastes in addition to fission products contain inactive elements included in the structural materials of fuel assemblies, the main of which are zirconium and iron.
A method of refining high-level waste nuclear reactors, including calcification HLW mixing calcinate HLW with TiO2, CaO, ZrO2, Al2About3heat the mixture in a reducing atmosphere to the working temperature from 1000°C to 1500°C under a pressure below atmospheric, the shutter speed at the operating temperature until crystals form a ceramic material, and cooled to obtain a solid material suitable for long term storage.
As a result of implementation of the known method is the formation of zirconolite ceramics with the calcinate HLW. [US patent N 4274976, G21F 9/34, publ. 23.06.81.]
Similar to the above is the method of processing high-level waste nuclear reactors containing radioactive elements actinides group (uranium, plutonium, neptunium, and so on), as well as zirconium and rare earth E. the elements, under which these wastes calicivirus, calcinate mixed with oxides of titanium, calcium and manganese in the following ratio, wt.%: TiO2- 50-60, Cao - 10-20, IGOs - 5-15, calcinate high-level waste - 15-25, the resulting mixture is heated in an oxidizing atmosphere up to operating temperature 1100-2000°C at a pressure below atmospheric, maintained at a working temperature until crystals form a ceramic material, after which the final product is cooled to obtain a solid material suitable for long term storage. The result is the formation of ceramic, which contains a crystalline material whose composition is characterized by the generic formula of (CA, Mn, REE)4(An, Zr, Ti)2Ti7O22, which by its physico-chemical nature, but also against the elements actinides groups, zirconium and rare earth elements similar to zirconolite, and the higher the pressure, the faster the process of crystal formation derived ceramic product. [Pat. Of the Russian Federation No. 2140106, G21F 9/16, publ. 10.10.1999.]
The disadvantage of this method is the low quality of the finished product due to its low resistance (applicable to HLW), the value of which is defined by total speed vymyvaemosti actinides from the of zirconolite at 90°C, is 10-4-10-5g/m2day.
Taken as a prototype method for processing waste generated during processing teplosistema assemblies of a nuclear reactor, characterized by moving them in a ceramic matrix. For this solution waste pre denitrify formaldehyde, which is then mixed mainly with titanium oxide, barium, calcium, and optionally aluminum, niobium. The resulting suspension is dried, annealed, at a temperature of 650-800°C, crushed to form powder. Next, the powder is pressed at 1000-1400°C and is sintered in a reducing atmosphere in the temperature range of 1000-1400°C.
When this is achieved the formation of the ceramic medium containing phase hollandite, perovskite and zirconolite containing 30-60 wt.% TiO2, 1-10 wt.% BaO, 1-10 wt.% CaO, in which it is assumed dissolution and fixation of fission products. (Journal of the Russian Federation No. 2002115623, G21F 9/00, publ. 20.12.2003,)
Despite the seeming simplicity of performance and chemical resistance of the obtained ceramics, this method is not without disadvantages. The method does not provide uniform distribution of fission products in the volume of the matrix, which negatively affects their long-term storage, a multi-stage process for the synthesis of a matrix, use of restorative environments and temperatures over 1000°C, significant costs oxides for turning rest the RA in the suspended state.
The objective of the invention is to simplify the recycling process - curing liquid waste while maintaining thermal and hydrolytic stability of the resulting monolithic material.
The problem is solved by a method of curing a liquid high-level waste and convert them into a gel-like state characterized by the introduction into the waste solutions of salts of zirconium, iron, and glycerol to a concentration of them in solutions, respectively, not less than 0.12, 0.6 and 0,23 M/l, maintaining the mixture for at least 2.5 hours, adding to the mixture of the solution of one-deputizing potassium phosphate in phosphoric acid to the molar ratio Zr:Fe:K:PO4=1:3:2:5-8, and subsequent drying, calcination of the obtained polymer gel zirconolite respectively, at 70-90°C and 300-400°C and melting the obtained granules at 980-1000°C.
Preferably, as salts of zirconium and iron to use respectively circinelloides and nitrate iron.
Usually use 2-2,1 M/l solution of potassium phosphate in 4 M phosphoric acid.
Thus, to simplify the process of solidification of liquid wastes is proposed instead of the oxide matrix using zirconium-iron-phosphate matrix, not inferior in its thermal and hydrolytic stability of the ceramic matrix specified in the prototype. When the eat the synthesis of the matrix occurs in the liquid phase upon mixing of the waste solution with the addition of metal salts with a solution of potassium phosphate in phosphoric acid. The method is based on the tendency of aqueous solutions of zirconium to the formation of polymeric compounds, educational opportunities, a significant number of simple and complex phosphates.
Previously in the liquid waste will be required in the amount of salts of zirconium and iron, as well as glycerin. The role of the latter is reduced to the formation of complex compounds with zirconium, and later to the denitration of the mixed solution. As the mixture is injected solution KH2PO4in phosphoric acid. After some time there is gelation of zirconolite, which is governed by the process of chelation circoncision with glycerin and process temperature. The resulting homogeneous gel zirconolite transparent and glassy, yellowish color. The gel is dried, calcined at t=300-400°C until the end of the selection oxides of nitrogen, if necessary, crushed and melted at a temperature of about 1000°C. the resulting melt the vitreous, has a color ranging from light brown to dark brown depending on the composition and quantity of waste and represents a solid solution of simple and complex phosphates of fission products in zelenotsvitna glass and partly in the form of independent phases, cemented with glass.
Drying of the gel may be done in two ways. First option: mixed solution before heliopaths what I poured into plastic trays and dried in a stream of hot air (70-90°C). Second option: mixed solution was dispersed into droplets in the heated mineral oil obtaining in the end a spherical particle gel matrix, which speeds up the drying process and eliminates the need for grinding of the product.
Prepare a solution of the waste to obtain a zirconium-iron-phosphate matrix. For this purpose, we first define the content of zirconium and iron in the waste and adjust them in accordance with a molar ratio by adding salts circinelloides - ZrOCl28H2O and nitrate iron - Fe(NO3)39H2O to their concentration in the solution, respectively, not less than 0.12 and 0.6 M/l and glycerol - 0,23 M/l (molar ratio of zirconium:glycerol=1:1,1). These concentrations provide a sufficiently strong and durable gel zirconolite. The prepared solution is incubated to equilibrium for 2.5 hours
Next, prepare a solution of one-deputizing phosphate potassium - KN2RHO4by dissolving 4-molar phosphoric acid to the concentration in the solution of 2 M/L.
The behavior of zirconium in solution is difficultly predicted for many reasons, including the history of its origin and the storage time of its salts. So the pre-test process of gelation at low volumes. Record the time the I beginning of gel formation in the mixed solution to determine the time of his life. Time correction "life" performed by adding glycerol or cooling solutions. Mix the bulk of the waste with a 2 M/l solution of potassium phosphate 4 M/l phosphoric acid, poured into plastic trays or dispersed into droplets in mineral oil. The drying is conducted by a stream of heated air (70-90°C), then calcined at 300-400°C before the termination of the allocation of nitrogen oxides and melt obtained granules at a temperature of about 1000°C.
The resulting polymer gel zirconolite characterized by a molar ratio Zr:Fe:K:PO4=1:3:2:6. The gel effectively absorbs microwave radiation, which allows for the annealing and melting in a microwave oven.
Test hydrolytic stability of the matrix on the leaching of transuranic elements was carried out as follows.
Prepared 120 ml of nitric acid solution of simulators fission products, the corresponding salt and radionuclide composition of waste reprocessing plants. The solution was divided in two, each part of which is introduced aliquots of nitric acid solutions of Np-239 and Pu-239 Am-241 in another part.
Dissolved in each part of the solution sample ZrOCl28H2Oh and Fe(MO3)3N2On to a final concentration of 0.2 M/l and 0.6 M/l, respectively. Introduced and mixed in a solution of glycerin to its content of about 23 M/l and has withstood 2.5 h to establish equilibrium. At the same time prepared under moderate heat 2 M/l solution KN2RHO44 M phosphoric acid.
Mixed the resulting solution with a solution of potassium phosphate (2 M/l KN2RHO44 M/l MN3RHO4). After the formation of the gel zirconolite, which occurred on 36 minutes from start of mixing at a temperature of 22°C (molar ratio Zr:Fe:K:PO4=1:3:2:5-8), the gels were dried, calcined and melted in landowych crucible at a temperature of 980-1000°C.
Data on the samples are presented in table 1.
|Nucle-dy||The mass of the nuclide in the volume of the prepared matrix mg||The amount of BC in the volume of the prepared matrix||The weight of the sample||The surface of the sample, cm2||Mass fraction of actinide in the original sample, %|
|Am-241||0,1||of 1.27×107||9,9||of 17.5||0,001|
Hydrolytic leaching was carried out at a temperature of 90°With a radiometric determination of the content of radionuclides in pixelated.
The obtained results data hydrolytic stability at a temperature of 90°C and the ratio of STB./VH2O≤10, R is the rate of leaching are presented in table 2.
|Time, day||R, g/(cm2·day)|
Thus, the proposed method allows to process acidic solutions (up to 2.5 M/l in nitric acid) and get compact product with an apparent density of from 2.2 to 2.6 g/cm3that order, and reduces the volume of waste. Hydrolytic stability matrix limits the rate of leaching of incorporated radionuclides in the range from 10-6up to 10-8g/cm2day (90°C), which are accepted as predicted during long-term storage.
1. The method of curing the liquid high-level waste by converting waste into a gel-like state, characterized by the fact that in solutions of high-level waste is injected salt of zirconium, iron, and glycerol to a concentration of them in solutions, respectively, not less than 0.12, 0.6 and 0,23 M/l, incubated the mixture for at least 2.5 h and then adding to the mixture of the solution of one-deputizing phosphate potassium phosphate, cyclotide molar ratio Zr:Fe:K:PO 4=1:3:2:5-8, drying, calcining, the resulting polymer gel zirconolite respectively, at 70-90°C and 300-400°C and melting the obtained granules at 980-1000°C.
2. The method according to claim 1, characterized in that salts of zirconium and iron using, respectively, circinelloides and nitrate iron.
3. The method according to claim 1, characterized in that use 2-2,1 M/l solution of potassium phosphate in 4 M phosphoric acid.
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: 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
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