Method of preparing technetium solutions
FIELD: chemistry.
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
The invention relates to the field of processing of liquid radioactive waste and can be used for localization of technetium from nitrate solutions followed by obtaining as a final form of solid metalloceramic matrix.
Known methods of coprecipitation and deposition of reduced forms of technetium on the hydroxides of metals [E. lamb, Sigrun X., E. Beauchamp in the book. "Proceedings of the Second international conference on peaceful uses of atomic energy (Geneva, 1958) 10. M: Atomizdat. S-126]. In these works we are talking mainly about the amounts of technetium during its extraction from irradiated blocks.
There are also known methods of coprecipitation of macroscopic quantities of technetium in the interaction in alkali (ammonia) environment with hydrazine and its derivatives, as well as with other reductants [Antoshkin, Slicom / study of the sorption of technetium on the precipitation of hydroxides of iron and aluminum in terms of conditions of lake Karachay. / Issues of radiation safety, 1999, No.1, p.12-17], [Swichkow and Wppirate "Behavior of technetium (IV)-(VII) in alkaline solutions in the presence of reducing agents, oxidizing agents, complexing agents and under the action of γ-irradiation". / Abstracts of the Second Russian conference on radiochemistry, Dmitrovgrad. 1997, s]. The disadvantages of the methods described here should be attributed to the instability of the effect of deposition due to oxide the Oia formed hydroxides of iron by oxygen or other oxidants, and the need to perform the deposition process in strongly alkaline (0.5 to 15 mol/l) and, therefore, strongly saline environment, which complicates the subsequent disposal of radioactive stock solution.
There is also known a method of deposition of technetium in the form of TCO2[patent RU 2201896, 10.04.2003]. Despite a fairly high rate deposition of technetium to 50-99,9% of the initial content in the solution, the main drawback of this method is the necessity of preliminary neutralization of oxidative air reductants such as U(IV) or Ti (III), the effect of which leads to the formation in solution of technetium dioxide and its coprecipitation with hydroxides mentioned tetravalent metals.
There are also other ways of curing solutions of technetium. As a matrix for technetium is used granular silica gel and rich metal sorbent further subjected to annealing [patent RU 2132094, 20.06.1999]. The main disadvantage of this invention consists in the necessity of processing technologo solution by complexing agents, which increases its adsorption on silica gel and then processing the saturated technetium sorbent reducers to obtain technetium dioxide, which leads to increased salinity mother liquor and complicates it further is pererabotka.
Also known is a method of obtaining metal technecally matrix recovery salt pertechnitat ammonium to metal [patent RU 2103403, 27.01.1998]. The main disadvantage of this invention is the extensive preparatory work to obtain salt pertechnitat ammonium, including sorption on cation exchange resin and neutralizing the resulting solution with ammonia, resulting in increased volumes of both liquid and solid radioactive wastes.
The prototype of the present invention is a method for the deposition of technetium from solution with a concentration of nitric acid and 1.5 mol/l and below with a solution of hydrazine or with a solution of iron (II) [patent RU 2199163, 20.02.2003].
Despite the good results of the deposition of the Cu: 95-99% of technetium in sediments of the original quantity in the solution, the main disadvantages of the prototype are the need for preliminary neutralization of the nitric acid solution with a concentration of nitric acid and 1.5 mol/l and lower solutions of salts of carboxylic acids or alkali to a pH of 5-11 and the lack of options for further processing of the received precipitation. Pre-treatment solution again leads to a significant salinization source solution that causes difficulties in subsequent processing.
The objective of the invention is to develop a method of localization of technetium deposition from ASOT the acidic solutions and then get on the basis of the precipitate formed solid suitable for further storage of the matrix.
The deposition is carried out on the initial concentrations of nitric acid or nitrate ion, not more than 3 mol/L.
For the deposition of technetium from nitrate solutions were used on-phenanthroline or α-bipyridine complexes of divalent transition metals such as Zn, Ni, Co, Cu, etc. mixed complexes of these organic compounds or mixed complexes containing o-phenanthrolin or α-bipyridyl with dibasic amines. The molar ratio of transition metal:organic ligand in the precipitator was 1:(1-12). The deposition is carried out at a molar ratio of metal precipitator:Cu = 1:(1÷2). The result is precipitation ORGANOMETALLIC pertechnetate that, after annealing in hydrogen flow at a temperature of 600-1200°C for 40 min in the presence of low-melting metal, such as Sn, Al, Zn, etc., or its oxide with a melting temperature of 200-800°With or without it allow to obtain metal-containing matrix.
Thus, the proposed method allows to besiege the technetium from the neutral and acid solutions of complexes of divalent transition metals with dibasic amines with getting technologo ORGANOMETALLIC sediment, subsequent annealing which allows to derive a metal matrix.
The localization of technetium method "direct wasp is Denia" using on-phenantroline transition metal complexes in solution remains in the middle of 2-20% of the technetium from the initial amount depending on the precipitator. When using the "reverse deposition" 98-100% TC goes into the sediment. BaselCement matrices obtained after precipitation annealing in hydrogen flow meets the standards for CAO.
Example 1
Direct deposition"
Neutral or nitrate technetium solution with a concentration of technetium 1 g/l and the concentration of nitrate-ion 0,5, 1, 1,5, 2, 3 mol/l at room temperature, add a solution of precipitator, representing nitrate o-phenantroline complex of copper (II) with a ratio of transition metal:organic ligand is 1:9. The deposition is carried out at a molar ratio of metal precipitator:technetium, which is 1:1,1.
Table 1 provides data on the deposition of technetium concentrated solutions of organic complexes of divalent transition metals from neutral and acid solutions.
| Table 1 | |||
| A solution of precipitator | The concentration of nitric acid in the original solution, mol/l | The metal content in the sediment after deposition of the initial content in the solution, % | |
| Metal precipitant | TC | ||
| 0 | 91 | 98 | |
| Cu(phen)9(NO3)2, | |||
| 0,5 | 93 | 98 | |
| Fe(bipy)3(NO3)2, | |||
| 1 | 95 | 98 | |
| Fe(bipy)2(phen)(NO3)2, | |||
| 1,5 | 95 | 95 | |
| Co(phen)2(en)(NO3)2, | |||
| 2 | 94 | 93 | |
| Co(bipy)2(en)(NO3)2, | |||
| 3 | 88 | 91 | |
| where | |||
| phen - o-phenanthrolin, | 4 | the precipitate is not formed | the precipitate is not formed |
| bipy-bipyridyl, | |||
| en-ethyldiamine |
The residual concentration of technetium in solution after precipitation was 0.02-0.1 mg/l
Example 2
Direct deposition"
Neutral or nitrate technetium solution with a concentration of technetium 1 g/l and the concentration of nitrate-ion 0,5, 1, 1,5, 2 mol/l at room temperature, add a solution of precipitator, representing nitrate o-phenantroline complex of iron (II) with a ratio of transition metal:organic ligand is 1:3. The deposition is carried out at a molar ratio of metal precipitator:technetium equal to 1:1,7.
Table 2 shows data on the deposition of technetium concentrated aqueous solutions of an organic complex of iron (II) from neutral and acid solutions.
| Table 2 | |||
| A solution of precipitator | The concentration of nitric acid in the original solution, mol/l | The metal content in the sediment after deposition of the initial content in the solution, % | |
| Fe | TC | ||
| Fe(phen)3(NO3)2, | 48 | 90 | |
| where | 0,5 | 52 | 86 |
| phen - o-phenanthrolin | 1 | 61 | 85 |
| 1,5 | 64 | 80 | |
| 2 | 68 | 80 | |
| 3 | 70 | 80 | |
| 4 | The precipitate is not formed | The precipitate is not formed |
The residual concentration of technetium in solution after precipitation amounted to 0.1-0.2 mg/l
Example 3
"Reverse deposition"
To the concentrated solution of precipitator poured original neutral or nitrate technicaly solution with a concentration on metal 1 g/l and the concentration of nitrate-ion up to 3 mol/l depending on the precipitator. When used as a precipitator of concentrated solution of an organic complex of iron up to 95% of the technetium is tons of original content in the solution passes into the sediment.
Example 4
Localization of technetium carried out by the method of "reverse" deposition, i.e. the gradual addition of an initial solution to the precipitant solution with constant stirring. The deposition is conducted as in neutral solutions and acid solutions with concentrations of nitric acid 0,5, 1, 1,5, 2 and 3 mol/l and the concentration of technetium 1 g/l at room temperature.
The residual concentration of technetium in solution after precipitation when using a concentrated solution of an organic complex of copper was 20 mg/L.
Example 5
The obtained precipitation ORGANOMETALLIC pertechnetate dried for 20 minutes at a temperature of 200-250°C., and then calcined from 20 minutes to 1 hour in a stream of hydrogen at a temperature of 600-1200°C.
Example 6
To the obtained precipitation ORGANOMETALLIC pertechnetate to reduce the porosity of the samples generated during the calcination, add a low-melting metal such as Al, Sn, Zn, etc. with a melting temperature of 200-800°C or its oxide. The samples are dried for 20 minutes at a temperature of 200-250°C., and then calcined from 20 minutes to 1 hour in a stream of hydrogen at a temperature of 600-1200°C.
These examples show that the proposed solution has the following advantages: no need for pre-oxidation-reduction processing is similar solution, and its neutralization; the method is extremely simple in its execution and is either adding a concentrated solution of precipitator directly in the source solution, or Vice versa, you can also get technetium in final form, suitable for further storage and processing.
1. A method of processing technicaly solutions, including deposition of technetium from nitric acid solutions of a compound containing divalent transition metal, wherein the precipitation is carried out on the initial concentrations of nitric acid or nitrate ion, not more than 3 mol/l, concentrated aqueous solutions of o-phenantroline or α-bipyridinium complexes of divalent transition metals, or mixed complexes of these organic compounds, or mixed complexes containing o-phenanthrolin or α-bipyridyl with dibasic amines, and the resulting precipitation ORGANOMETALLIC pertechnetate translated into a form suitable for long term storage.
2. The method according to claim 1, characterized in that the ratio of the transition metal organic ligand in solution precipitator is 1:(1-12), and the deposition is carried out at a molar ratio of metal precipitator:TC equal to 1:(1÷2).
3. The method according to claim 1, characterized in that the process of deposition of technetium exercise as a method of "direct the th deposition", the addition of the precipitant solution in the original solution and the method of "reverse deposition", gradual addition of the starting solution in the solution precipitator.
4. The method according to claim 1, characterized in that the precipitation is calcined in a stream of hydrogen at a temperature of 600-1200°C.
5. The method according to claim 1, characterized in that the calcination of the obtained precipitation is carried out in the presence of low-melting metal, such as Sn, Al, Zn, etc. or its oxide with a melting temperature of 200-800°C in a stream of hydrogen at a temperature of 600-1200°C.
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.
1 dwg
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.
5 cl
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.
2 cl
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: chemistry.
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
FIELD: chemistry.
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: chemistry.
SUBSTANCE: method of processing a radioactive solution involves the following. First, an iron (III) compound in form of chloride or sulphate is added to the solution in amount of 0.04-0.05 mol/l to form an iron-containing precipitate. At the first step, a minimum amount of mineral acid - hydrochloric or sulphuric acid - is added, and at the second step 0.18-0.24 g-eq/l of the corresponding acid is added to the solution. The solution is held for not less than 120 hours at room temperature or not less than 18 hours at 70-95°C and sodium sulphide is added to the solution in a molar amount which is 1.5 times greater than the amount of the added iron (III) compound to form a basic collective precipitate of radionuclides of cobalt and caesium and a mother solution containing an organic complexing agent and a residual amount of radionuclides of cobalt and caesium. The mother solution is subjected to a post-treatment cycle by adding an iron (III) compound in amount of 0.02-0.04 mol/l with respect to iron (III) and mineral acid in an amount which is equivalent to content of sodium in the added sodium sulphide, holding the obtained mother solution and adding additional sodium sulphide in molar amount which is 1.5 times greater than the amount of the additionally added iron (III) to form an additional collective precipitate of the post-treated mother solution.
EFFECT: invention enables to increase processability of the method by replacing oxidation of the organic complexing agent with cationic substitution of the cobalt radionuclide therein, reduce the amount of reagents used while ensuring high degree of purification of solutions.
6 cl, 4 ex
FIELD: 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.
3 ex
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.
1 cl
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.
1 cl
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
