Method of recycling waste solutions in production of uranium tetrafluoride
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
The invention relates to hydrometallurgy uranium and can be used for the disposal of waste solutions in obtaining uranium tetrafluoride (TFU) from nitrate leaching solutions concentrates using the processes of extraction, Stripping and heat treatment of uranium compounds derived from reextractors obtaining uranium dioxide and its subsequent fluoride-chloride solutions.
In the scheme of production TFU (with the extraction refining nitrate leaching solutions by tributyl phosphate (TBP) and subsequent solid-phase by re-extraction solution of ammonium carbonate) uranium is precipitated from solution in the form of crystals of tricarbocyanine ammonium (AUTKA). The precipitate is subjected to calcination to obtain UO2directed at getting TFU by treating it with a solution of a mixture of hydrochloric and hydrofluoric acids.
In this scheme there are three types of waste solutions containing uranium and require special processing:
- refined - nitric acid solutions with concentrations of nitric acid, from 50 to 100 g/DM3on HNO3and uranium up to 100 mg/DM3;
fluoride - chloride uterine solutions (FHM) after receiving TFU with HF concentration up to 5 g/DM3HCl up to 50 g/DM3and uranium up to 8 g/DM3;
- carbonate uterine solutions (KM)formed in achiev is Tate solid-phase uranium Stripping of a saturated TBP and separation of crystals AUTKA containing ammonium carbonate to 60 g/DM 3, uranium up to 8 g/DM3and excess ammonia to 20 g/DM3.
Known methods of utilization of parent uranium production by neutralizing alkaline reagents, followed by the separation of sludge and discharge the clarified part on the site.
The closest analogue to the claimed technical solution is known a method of disposal of circulating solutions of uranium production, including treatment with alkali or lime "milk", Department of uranium precipitation from solution and their return to the stage leaching original products, bleaching waste solutions tailings pond [Sabrekova OG, raw HE Methods of preparation of working solutions the production of uranium tetrafluoride to reset. Materials of scientific-practical conference dedicated to the 50th anniversary of Seversk state technological Academy, may 18-22, 2009, "Technology and automation of atomic energy and industry of TAAP-2009", Seversk, 2009. P.26].
According to the method:
- refined (nitric acid solution) to neutralize the lime "milk" and dumped on the site;
KM is subjected to heating for the destruction of ammonium carbonate and distillation of ammonia and carbon dioxide, VAT residue is treated with lime "milk" to deposition of uranium and filtered, after which the precipitation of the TP is to participate in the process cycle, and the clarified portion of the discharge at dam;
- FHM neutralize lime "milk" or a sodium hydroxide solution and filtered, then the precipitation is returned to the process cycle, the clarified part of the shed on the site.
This processing method allows to prevent waste solutions in surface waters and translate the most toxic and radioactive polyvalent metals from the liquid phase into a solid. After filling the tailings resulting precipitation will be securely isolated from contact with the environment.
The main disadvantages of this method are as follows:
- each type of uterine fluids are processed separately, which leads to unjustified increase in the number of units of equipment, including reactors, clarifiers, filters, etc.;
substantial consumption of reagents, in particular lime, sodium hydroxide and flocculants, electricity, steam;
- high residual uranium content after processing KM (>100 mg/l);
- increased liquid waste on the site due to the dilution of the mother liquor lime "milk";
- a large number of impurities, is returned to the process cycle with selected precipitation (including calcium carbonate coming from lime "milk"), which leads to things is to promote increased consumption of nitric acid when dissolved and lower quality TFU.
The task of the invention is to reduce costs nitric acid, lime and sodium hydroxide, the reduction of energy consumption for processing solutions, the reduction of the discharge of effluents on the site.
The technical result is achieved by the utilization of working mother solutions for the production of uranium tetrafluoride, including the operation of the precipitation of uranium from solution by neutralization, characterized in that the acidic chloride-fluoride uterine fluids from the production of uranium tetrafluoride and/or refined, resulting from the extraction of uranium from nitric acid solutions of organophosphorus compounds, jointly processed carbonate mother liquor produced during uranium Stripping and deposition of tricarbocyanine ammonium, blowing solutions air and dosedanjem uranium sodium hydroxide at a pH value of 8.0 to 12.0.
By neutralizing FHM and/or refined at the first stage of processing to pH 4,0-5,2 spent the greatest amount of neutralizing agent from the total amount required for complete neutralization, so the process is continued until the specified pH range. At lower pH values sharply reduced consumption, pH, increasing the pH above these values leads to incomplete precipitation of uranium from solution. The same is observed for the neutralisation of acidic RA the solutions carbonate liquor to pH values below 4.0. The purge air is required for the destruction of carbonate and bicarbonate ions and remove them from solution in the form of CO2. If the solution is not to expose the air purge, the uranium bound in carbonate complex, deposited will not. At the second stage neutralizing solution (slurry) neutralized with sodium hydroxide solution until a pH value of 8.0 to 12.0. Use sodium hydroxide instead of lime "milk" is due to the fact that with the introduction of NaOH does not happen deposition of fluoride (fluorite), together with uranium. Fluorine almost all (90-95%) remains in solution. Therefore, the return obtained uranium precipitation leaching will not lead to an increase in fluoride concentration in the solution and deterioration of extraction. While maintaining the second stage pH values below 8.0 for decreasing the degree of deposition of uranium above 12.0 - is unreasonable consumption of alkali. By neutralizing with sodium hydroxide pulp air is not blown, as in this case, the effect of increasing the degree of deposition of uranium does not occur.
In the patent and scientific literature there are no data on co-processing waste solutions production TFU. Therefore, the proposed technical solution is characterized by novelty and has significant differences.
Thus, the analysis of the proposed technical re the texts shows between the distinctive features of the proposed method and achieve this result there is a new cause-and-effect relationship: the presence of these signs in the claimed method provides a positive effect, but the absence of these signs does not have effect, the aim of the invention.
Comparison of the performance of the proposed and previously known method (prototype) is shown in the examples.
Example 1. In this example the results, showing the range of values of pH at which there is the most complete precipitation of uranium from the uterine fluid during their joint processing and the necessity of purging air from acidic solutions after neutralization KM.
- refined and KM;
- FHM and KM;
- mixtures refined with FHM and KM.
With the aim of studying the dependence of the residual concentration of uranium from the pH values (table 1) raised a number of experiments which were conducted neutralization refined solution of ammonium carbonate (imitator carbonate stock solution) to a certain pH values (pHRef), after that the solution was barotiwala the air to stabilize the pH values (pHCN). In separated from the precipitate solution was determined residual concentration of uranium.
|The dependence of the residual concentration of uranium (Cu) in the treated solutions from pHRefby neutralizing refined KM, followed by blowing air|
|№ p/p||The deposition parameters|
|pHRef||pHCN||The amount of uranium in the sediment, g||Withandin solution, mg/DM3|
|17||5,20||the 7.65||is 0.135||13,9|
|19||of 5.40||to 7.84||0,115||232,2|
For the experiment was taken in 100 ml of solution with a content of nitric acid 70 g/DM3and uranium 2 g/DM3that was poured into the beakers with a volume of 200 ml In a beaker with the solution was placed in the bubbler. From the burette into the beaker submitted a solution of ammonium carbonate and simultaneously introduced air. Stirring the solution ASU is actulaly with a magnetic stirrer.
The results of the experiments showed that the fully refined uranium from sediments in the area of pHReffrom 4.0 to 5.2. Unbalance in the amount of precipitated uranium and its residual concentration in solution is associated with the dilution due to the introduction of a solution of ammonium carbonate.
It should be noted that no air blowing neutralized solutions of uranium practically not deposited, and only in the region of pH 5, you receive a small suspension with the transition of the uranium precipitate from the solution within 15% of its total number. Further neutralization to a pH of 11.5 with sodium hydroxide does not dosageno uranium.
To increase the degree of extraction of uranium from solution to precipitate the use of the combined deposition: first, after neutralization of the raffinate KM and sparging air to stabilize the pH values (pH1) (the major part of the uranium turns into sludge)left in solution uranium thosedays further neutralization of the alkali solution (20%NaOH) to pH2(table 2).
|The dependence of the residual concentration of uranium (Cu) in the treated solutions from pHRefby neutralizing refined KM from blowing Vozduha treatment with sodium hydroxide solution|
|№ p/p||pHRef||pHCN||C0in a mixture of mg/DM3||The uranium concentration in solution (mg/DM3) after treatment with sodium hydroxide to pH2|
|pH 8||pH 9||pH 10||pH 11||the pH of 11.5|
|17.||of 5.40||to 7.84||232,2||104||77||63||51||42|
The results of the experiments show that the refined, neutralized KM to pH 4,0-5,2 with bubbling air after additional treatment with sodium hydroxide contain less uranium than without treatment with alkali.
By neutralizing FHM using KM and further passing air through the solution precipitate falls, which is associated with the formation of stable fluoride and chloride complexes of uranyl. At low pH values, these complexes inhibit the transition of uranium in the sediment. Therefore, for the deposition of uranium required neutralization with alkali to higher pH values than in the case of refined. Source FHM (the pH value was 1,81) was neutralized KM values to pHRefthen after the solution was allowed the air to stabilize the pH values (pH1), after which was added alkali (20% solution of NaOH) to pH2. After filtration the solution was analyzed for uranium content. The results of the experiments are presented in table 3.
|The dependence of the residual concentration of uranium (Cufrom pHRefby neutralizing FHM|
|rich||3,51||as 4.02||4,60||4,80||4,90||5,00||5,10||5,20||of 5.4|
|pH1||4,1||4,4||to 4.62||4,84||4,94||5,03||5,13||of 5.40||6,0|
|pH2||11,60||11,58||11,63||11,65||11,60||to 11.56||11,59||11,55||the 11.6|
|Cuin solution, mg/DM3||213,5||17,6||15,3||1,71||1,22||0,98||0,98||0,98||42,5|
The results of the experiments show that using this method, you can achieve a fairly complete precipitation of uranium by neutralizing FHM KM in the region of pH 4,0-5,2.
The following is an example of deposition of uranium from a mixture of all three types of uterine fluids. For this purpose were prepared mixture of nitrate raffinate FHM in volume ratios of 1:1 and 1:2, which then was kind of balanced out KM.
In experiments with mixtures of deposition of uranium simple transmission of air to achieve without success. As in the case with FHM, for the deposition had to use the alkali solution. The experiments were carried out in the same way as with one FHM. The results are presented in tables 4 and 5.
|The dependence of the residual concentration of uranium (Cufrom pHRefby neutralizing a mixture of refined and FHM in the ratio of 1:1 KM|
|pH1||of 3.77||4,25||4,87||to 4.98||5,09||5,18||5,28||6,1|
|pH2||11,7||are 11.62||11,66||11,58||11,59||the 11.6||11,5||11,44|
|Cuin solution, mg/DM3||64,6||5,52||2,20||1,96||1J1||1,47||1,71||48,6|
|The dependence of the residual concentration of uranium (Cufrom pHRefby neutralizing a mixture of refined and FHM in the ratio 1:2 KM|
|pH1||3,80||4,22||4,85||4,96||of 5.05||5,18||and 5.30||5,55|
|Cuin solution, mg/DM3||55,7||3,43||1,96||1,96||1,22||1,22||2,44||44,35|
Example 2. In glasses with a volume of 1 l was placed a stir bar and the electrodes of the pH meter. Three cups were filled with 0.5 l of a production solution raffinate composition major components: nitric acid 60 g/DM3, ammonium nitrate, about 100 g/DM2, uranium 0.11 g/DM2iron 10 g/DM2. In the other four glasses were injected the same amount FHM composition: 40 g/sup> 3hydrochloric acid, 2 g/DM2hydrofluoric acid and 6 g/DM2uranium. For the next experiment used a mixture FHM and the raffinate at a ratio of 1:1 and 2:1. Each of these mixtures was also poured into three glasses.
All of these party solutions subjected to the treatment according to the following schema:
1. Neutralization of 10% lime "milk".
2. Neutralization KM to the value of pH 5.0, followed by neutralization with 20% sodium hydroxide solution until a pH value of 10.5.
3. Neutralization KM to the value of pH 5.0, followed by blowing air for 20-30 min to stabilize the pH values of about 7 and neutralized with 20% sodium hydroxide solution until a pH value of 10.5.
4. In addition to the listed methods of neutralization FHM processed only 20% solution of sodium hydroxide to a pH of 11.5.
MILES had the following composition: ammonium carbonate 79 g/DM3the uranium - 3.2 g/DM3.
As can be seen from the presented results (table 6), with the joint treatment of uterine solutions residual uranium concentration is slightly higher than in the processing of lime or sodium hydroxide.
|The results of processing waste solutions of different methods.|
|Perera is atively solution||The composition of the solution||The processing circuitry solutions||The specific reagent consumption, ml/DM3||The residual uranium content, mg/DM3||The volume of solution after neutralization, ml|
|The raffinate||HNO3- 60 g/DM3, NH4NO395 g/DM3U (VI) 0.11 g/DM3, Fe (III) 10 g/DM3||1 scheme||380||3||690|
|2 scheme||368 CMR to pH of 4.95+6,6 NaOH||1960||874,6|
|scheme 3||368 CMR to pH 4.9 purge air to a pH of 6.9+4,2 ml of 20% NaOH to pH 11||the 9.7||872,2|
|FHM||HCl - 40 g/DM3, HF - 2 g/DM3U (VI) - 6 g/DM3||1 scheme||410||About 1||705|
|2 scheme||385 CMR to pH of 4.9+6,8 NaOH||4150||scheme 3||385 CMR to pH 4,85 air blowing to pH 6,85+4,5 ml of 20% NaOH to pH 11||of 5.4||589,5|
|4 scheme||a 20% solution of NaOH||About 1||530|
|The mixture FHM and the raffinate at a ratio of 1:1||HNO3- 30 g/DM3, NH4NO347,5 g/DM3HCl - 20 g/DM3, HF - 1 g/DM3, Fe (III) 5 g/DM3U (VI) was 3.05 g/DM3||1 scheme||400||10||700|
|2 scheme||394 CMR to pH of 4.95+NaOH to 7.2||2910||901,2|
|scheme 3||372 CMR to pH 4.9 air purge to pH 7.0+4,6 ml of 20% NaOH to pH 11||8,5||876,6|
|A mixture of chloride and fluoride of the mother liquor and the raffinate in the ratio 2:1||HNO3- 20 g/DM3, NH4NO331,7 g/DM3HCl - 26,6 g/DM, HF - 1.3 g/DM3, Fe (III) and 3.3 g/DM3U (VI) - a 4.03 g/DM3||1 scheme406||4,5||703|
|2 scheme||394 CMR to pH of 4.95+NaOH to 7.2||3000|
|scheme 3||372 CMR to pH 4.9 air purge to pH 7.0+4,6 ml of 20% NaOH to pH 11||8,1|
However, it should be borne in mind that KM after the distillation of ammonia at elevated temperature and subsequent treatment with lime VAT residue contain 60-80 mg/DM3uranium. When mixing all mother solutions after neutralization, the average content of uranium in waste solutions is significantly less than 30-40 mg/DM3.
In the proposed method, in addition to reducing the residual concentration of uranium decreases the total amount of wastewater is about 1.5 times due to the fact that they do not enter the lime milk for neutralisation are used primarily KM.
Thus, by reducing the volume of waste solutions and the concentration of uranium in a significant reduction of the total uranium sent to the tailings pond.
1. The method of disposal of waste fluids production of uranium tetrafluoride, including the operation of the precipitation of uranium from solution by neutralization notable is the acid chloride-fluoride uterine fluids from the production of uranium tetrafluoride and/or refined, resulting from the extraction of uranium from nitric acid solutions of organophosphorus compounds, jointly processed carbonate mother liquor produced during uranium Stripping and deposition of tricarbocyanine ammonium with simultaneous blowing of the neutralized solution with air to stabilize the pH, followed by processing the resulting pulp with sodium hydroxide without air blowing.
2. The method according to claim 1, characterized in that the joint processing of the chloride-fluoride uterine fluids and/or refined spend carbonate mother liquor to pH 4,0-5,2.
3. The method according to claim 2, characterized in that the obtained solution is further processed with sodium hydroxide to a pH of 8.0 to 12.0.
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
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: 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