Method of processing filter perlite powder
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.
The invention relates to the field of heterogeneous processing of liquid radioactive waste (LRW), in particular the reprocessing of spent fine abrasive filtering materials, and can be used in the reprocessing of spent filtrability (AF) systems specutacular.
In specutacular NPP are used upstream of the filters for the purification of aqueous media from oils and mists in which as alluvial filtering material use filtrability nuclear class. In the process of filtering surface layer filtrability polluted with sediment, hydroxides corrosion products of metals, oils and other, the resistance of the filter layer increases. When the excess pressure drop is higher than the permissible values are regeneration upstream of the filter, i.e. the removal of reclaimed filtrability. Spent filtering material hydrotransport is sent to storage tanks, where it is argued that transport water is decanted and sent for recycling. At many sites use separate storage of filtering materials of different nature, keep filtrability separately from the filtering materials of a different type. In recent years, the NPP begin to create the installation processing of the accumulated filtering materials, in particular, by cementing. However, when transporting filterprotocol certain difficulties, due to its abrasive action on the material equipment. Filtrability has a high porosity (up to 85÷90% of the volume) and low bulk density component of 0.1÷0.15 kg/DM3. In the technical specifications of the powder perlite filter (GOST 30566-98) indicated that the grain structure is determined by filtration on a sieve with a mesh size of 0.14 mm At nuclear power plants supply filtrability technical conditions with smaller grain structure. To obtain the AF nuclear class with a higher filtration capacity to colloidal impurities and high chemical stability, perform additional processing of the AF powder, consisting of ultrasonic grinding particles, and chemical processing, to create on the surface of particles of polycondensation and copolymer films. Filtrability is a natural aluminosilicate with a high silicon content. The resulting particles filtrability have needle configuration, together with the hardness of volcanic glass determines its abrasive action, leading to rapid wear of rubbing parts and decommissioning of equipment. The operating experience of the equipment cementing spent the filtering materials of the Ignalina NPP (Lithuania) showed that the failure of (the lack of pressure on the outlet) screw pump "Muan" eventough mixer cement compound occurs during the day. A known method of processing filtrability is bitumirovannaya. In practice, however, the inclusion of filtrability in bitumen practically do not produce, because it greatly reduces the fluidity of bitumen and its incorporation into the bitumen compound does not exceed 10% of the mass. on the dry filtering material.
The closest analogue of the claimed invention is a method of processing radioactive filteroperator described in the patent of Russian Federation №2435240. According to this method the pulp of filtrability is extracted from the storage tank, remove excess moisture and transported by the transport container, which is done by cementing. Filtrability chemical composition perfectly combined with cement, however, the degree of inclusion in the cement matrix is limited by its high porosity. In the Portland cement - 400 can include up to 12.5% of the mass. while maintaining sufficient mechanical strength of the cement blocks. When the mass ratio of the components is dry AF, PC-400 and water is 1:3:4, the volume of the Central Committee exceeds original OP only 15%, i.e. Kv=1,15.
The disadvantage of the nearest analogue of the method of processing radioactive waste filtrability is the complexity of the operation of the transportation hydrotransport of filtrability because of its high hardness and holchester configuration of particles filtrability that riodic to great wear and tear of transportation pipeline and pumps.
The problem solved by the invention is to reduce wear of the transport pipeline, pumps during transportation filtrability and simplifying the operation of the transport slurry.
The essence of this technical solution is that in the method of processing radioactive waste filtrability, which includes operations: extracting the pulp of filtrability of storage capacity, the removal of excess moisture, transportation hydrotransport and cementing proposed in the pulp prior to transportation from the storage tank to enter the waste ion-exchange resin in an amount of 10÷75% of the volume of filtrability at a density of 1÷1.5 g/cm3.
Conventional transportation suspension exhaust AF hydrotransport and further processing leads to abrasive wear of the equipment and pipelines. It is known that the negative impact of AF can be reduced by increasing the ratio of liquid: AF, which will increase the volume of LRW (water transport) and, therefore, increase the cost of treatment, and will also reduce the performance of the plant radioactive effluents and make it ineffective. To reduce wear of the equipment and pipelines proposed transportation and reprocessing of spent filtrability to produce together with spent IO is obmennymi resins (IOS) specutacular NPP, particles which have the correct a spherical shape with a density of 1÷1.5 g/cm3are elastic and do not have a negative impact on the equipment and pipelines. Grain of ion-exchange resins have a larger diameter and with a certain number of them, blocking the contact of grains filtrability with the surface of the equipment or pipeline, significantly reducing wear and tear.
Examples of the implementation of the proposed method with graphical demonstration of the values of the declared parameters are summarized in the table shown in figure 1, 2. The grain size of the ion is 0.35÷2.00 mm with a density of 1÷1.5 g/cm3, filtrability - 0,006÷0,030 mm Fine particles FP will occupy the free space between the grains of ion exchangers. The volumetric ratio of ion exchangers: AF is presented in figure 1, 2. Consider option 2 limit moisture stored pulp spent filtering materials in the storage containers. The moisture content of the pulp is 60÷65%, i.e. the volume ratio of T:W=1:1 (all the moisture from the pores and between the grains of the slurry). The first option (item 1, figure 1) ion - exchange resin is missing, depreciation of equipment and pipelines maximum. The second option (item 4, figure 2) is the most complete filling of the mixture of spent ion exchangers. The most dense packing of grains of ion-exchange resins (balls with the average size) is a face-centered or volume is but centered cubic packing. The volume occupied by the ion exchange resins in such packages will be 75%. When filling the entire space between the grains of resin wet filteroperator humidity 60÷65% volumetric ratio of T:W in the pulp will be 4:1. In this case, the minimum number of grains of AF will touch the surfaces of equipment and piping, any of them. It is possible to increase the content of ion exchange resins to correlation with filteroperator 90:10%, but in this case, a portion of the space between grains of ion exchangers will be used inefficiently and will remain filled with filteroperator. Usually the transportation of slurries OP and ion exchangers is carried out by hydrotransport, when the ratio of T:W compiled from 1:10 to 1:20 (item 3, figure 2). With the increase in the ratio T:W reduce abrasive action of FP on the equipment and pipelines, so the amount of insertion of spent ion-exchange resins in relation to the volume of filtrability can be reduced to 10% (item 2, figure 1). Joint transportation of spent filtrability and spent IOS is preferable also in view of their subsequent co-curing in the installation cementing heterogeneous waste.
This invention reduces the wear and tear of equipment and pipelines in the implementation of the method of processing radioactive waste filtrability in 80-100 times.
Spacepirate radioactive waste filtrability, includes extract of the pulp of filtrability from storage tank, remove excess moisture, transportation hydrotransport and cementing, characterized in that the pulp before transportation from storage tank to enter the waste ion-exchange resin in an amount of 10÷75% of the volume of filtrability at a density of 1÷1.5 g/cm3.
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
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
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