Composite material for immobilisation of liquid radioactive wastes, and its application method

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

 

The invention relates to the field of nuclear technology and applies technologies for processing of radioactive waste. Currently, the problem of safe radioactive waste management is one of the main problems depend on the magnitude and dynamics of development of nuclear energy, as well as further implementation of radiation technology. The invention more specifically relates to the field of high salt processing liquid radioactive waste (LRW) low and medium level activity, containing up to 30% of organic substances. Such waste is generated, for example, as a result of low-salt decontamination of radioactively contaminated water at the enterprises for radioactive waste management and operating nuclear power plants thermal method (distillation or evaporation). This method of decontamination is implemented in a special evaporators (stills) with a supply of heat, water vapor through the wall of the apparatus. Final products of the working devices are distillate and a small volume of LRW called kovovymi residues (KO), in which the concentration of radionuclides and salts in the 60-300 times more than the original waste. Thus, the method of distillation or evaporation leads only to reduce the volume of LRW by concentration and does not solve the problem of providing long-term and safe clearway storage medium LRW.

Currently one of the most promising technologies for immobilization of liquid radioactive waste resulting from the process of distillation or evaporation, is monolithic, i.e. their inclusion in a matrix of solidified structures. These structures must meet the requirements of GOST R 51883-2002 (radioactive Waste cemented. General technical requirements). It should be noted that the technology of immobilization LRW through monolithic preferably should not require high energy costs and implemented for any positive temperatures on equipment used in conventional cementing.

Known (Hat and other Development and testing of matrix compositions for cementing LRW LCS "radon" // "Ecology and nuclear engineering", №2(17), s) matrix composition for cementing LRW special laundries, which includes: Portland cement grades M-400 M-500), calcium chloride, natural sorbent (bentonite or natural mixture of bentonite-zeolite). The disadvantages of this structure are as follows: the use of relatively expensive materials, which are widely used in construction, and not a very high degree of filling of the cement compounds dry radioactive salts.

Known (USSR author's certificate No. 1275560 And 1986) method of curing LRW by evaporation on the receiving sludge from crystallohydrates water and then turning sludge into bitumen at a temperature (50-90°C and the slurry to bitumen 1:4. This method has some significant drawbacks. So in the resulting composition slurry - bitumen content of the waste is small, and therefore, its efficiency is low. Furthermore, the method has a high energy consumption due to long-term evaporation of liquid and high fire risk due to Flammability of bitumen.

The closest to the essential features to the proposed technical solution is a composite material that is protected by RF patent No. 2378723 from 10.04.2010, therefore he adopted for the prototype. Known composite material for immobilization of radioactive and chemical toxic waste has the following composition, wt.%:

td align="right"> the rest of it.
powder caustic magnesite (PMC-87)30-50
filler30-50
catalytic additive in powder form
breed of shungite or white black0,01-0,5
aqueous solutions of magnesium chloride
density of 1.1-1.3 g/cm3or sulfate
magnesium with a density of 1.2-1.35 g/cm3

As filler composite material contains metallurgical slag with particle sizes up to 0,074 mm or ash from burning of organic and/or inorganic substances. Additionally, the material contains up to 0.5% bentonite clay.

The disadvantage of this composite material is the excess of one of the main, according to GOST R 51883-2002, indicators of the quality of the cement compounds, namely, the rate of leaching of cesium-137 (≤1·10-3g/cm2·day) - even at low filling degree (3-8 wt.%) compounds of radioactive salts. This is due to the fact that the matrix of magnesia cement (MC) is not a barrier for cesium-137, and efficiency introduced in the composite sorbents for retention of radionuclides with the required parameters is not enough.

Also known methods of immobilization of radioactive waste, including, in the mineral matrix blocks. For example, in accordance with the patent of the Russian Federation No. 2189652 method of immobilization of radioactive waste in the mineral matrix block, including the mixing of radioactive waste oxidizing agent, a reducing agent and mineral Supplement in a certain proportion, the gap between the outer and inner containers of powdered inorganic insulating material, loading the received masivo internal capacity, heating a mixture of radioactive waste oxidizing agent, a reducing agent and mineral Supplement by initiating it exothermic reaction between the oxidizing agent and reducing agent, obtaining a melt of the final product and its cooling. However, this method is technologically complex, energy intensive and requires the use of special devices.

The closest to the essential features to offer in the invention is a method for the immobilization of liquid radioactive waste in accordance with the patent of the Russian Federation No. 2214011 (prototype), including their concentration and curing with curing the mixture to form a strong solid monolithic block, locking in its structure components of radioactive waste, characterized in that the curing is carried out by mixing radioactive waste with a solution of magnesium chloride density of 1.2-1.35 g/cm3, magnesia binder and fine mineral filler with a particle size of 0,005-0,015 mm Disadvantages of this method is the excess of the normative rate for leaching of cesium-137 and insufficient occupancy of the end product of radioactive waste containing significant amounts of organic and surface-active substances.

The technical result, which is opravleno the invention, is the development of such a composite material and method of its application, which allow you to immobilized LRW quality that meets GOST R 51883-2002, namely the rate of leaching of cesium-137 is provided ≤1·10-3g/cm2·day upon reaching the degree of filling compounds dry radioactive salts 35-37%.

With this purpose in composite material for immobilization of high salt liquid radioactive waste including hardener, powder caustic magnesite and catalytic carbon additive, as a hardener primenenii solid salt and the composition of the material added to the solution of ferrocyanide of potassium and Nickel nitrate, and calcium chloride in the following ratio, wt%:

solid salt5-6
powder caustic magnesite27-28
catalytic carbon additive0,1-0,2
a solution of ferrocyanide of potassiumthe 0.05-0.1
a solution of Nickel nitratethe 0.05-0.1
calcium chloride (CaCl2) 0,1-6
liquid radioactive wasterest

Additional differences of the proposed composite material is used as a solid salt setevogo of magnesium chloride (MgCl2·6H2O) or semimodule of magnesium sulfate (MgSO4·7H2O), and as a catalytic additive white soot, shungite or pyrocarbon.

To achieve the technical result of the invention is proposed in the method of applying composite material for immobilization of high salt liquid radioactive waste, consisting in their concentration and curing using a composite material comprising a hardener, powder caustic magnesite and catalytic carbon additive, curing the mixture to form a strong solid monolithic block, locking in its structure components of radioactive waste, as a hardener is proposed to use solid salt, and the composite material further introduce solutions of ferrocyanide of potassium and Nickel nitrate, and calcium chloride in the following ratio, wt%:

solid salt5-6
powder caustic magnesite27-28
catalytic carbon additive0,1-0,2
a solution of ferrocyanide of potassiumthe 0.05-0.1
a solution of Nickel nitratethe 0.05-0.1
calcium chloride (CaCl2)0,1-6
liquid radioactive waste with a pH≤a 9.7-9,9,rest

The cure suggested to the introduction of liquid radioactive waste of the ingredients of the composite material in the following sequence: 1 - solid salt, 2 - a solution of ferrocyanide of potassium, 3 - solution of Nickel nitrate, 4 - calcium chloride, 5 catalytic carbon-containing additive and 6 - caustic magnesite powder with continuous stirring of the mixture.

Additional differences between the proposed method is used as a solid salt of six aqueous magnesium chloride (MgCl2·6H2O) or semimodule of magnesium sulfate (MgSO4·7H2O), and as a catalytic additive white soot, shungite or pyrocarbon.

The application of the proposed composite material as hardeners powder is as caustic magnesite instead unsaturated solutions of sodium chloride or of magnesium sulfate solid salts allows the use of water, contained in the composition LRW for the formation of saturated solutions of sodium chloride or of magnesium sulfate, and enhance the utilization of the final product (hardened compound) radioactive waste.

Introduction solutions hexacyanoferrate potassium (yellow blood salt) and Nickel nitrate, forming by mixing insoluble compound - Nickel ferrocyanide, potassium (FCNC), which is an effective selective sorbent for cesium, provides a significant reduction in the rate of leaching of the solidified compound of cesium-137.

And, finally, the introduction of the compound calcium chloride can increase the degree of filling of the radioactive compound salts and associate included in the waste phosphates, oxalates, silicates insoluble in the connection.

One of the embodiments of the developed method concreting LRW described below. As LRW used distillation residues Laundry containing phosphates, silicates, sulfates, oxalates, and organic surfactants. The salt content of the bottoms was 400-600 g/l, and organic and surfactants to 30 wt.%.

A. LRW served by the pump from the supply tank, equipped with a metering device in the mixer setup cementing (if pH LRW more of 9.7 to 9.9, enter the necessary is the volume of hydrochloric acid), then the dispenser fed into the mixer crystalline magnesium sulfate or magnesium sulfate and the mixture is stirred for 3-5 minutes After such contact water of crystallization of salts enters LRW and his little thins.

B. In the mixture "A"that are in the mixer, using the dispenser introduce a measured amount of yellow blood salt and the mixture is stirred for 3-5 min, and then there serves a measured volume of Nickel nitrate and the mixture is stirred for 3-5 min, and then incubated for 2-3 h with intermittent mixing every 0.5 h For the reaction:

K4[Ni(CN)6]+4Cs+→CS4[Ni(CN6]+4K+

get FCNC, which has a high cleaning efficiency of cesium-137 (coefficient of cleaning up to 103). It should be borne in mind that the formed insoluble compound at pH≥10 starts effectively to dissolve.

C. In a mixture of "B" dispenser dropping a measured sample of the powder of calcium chloride at a rate of 0.1-0.2 g per 1 g of dry radioactive salts. Contained in the composition LRW phosphates, oxalates, silicates, carbonates, sodium, sodium salts of fatty acids form insoluble compounds. The volume of the mixture clots (sludge), which after stirring for 3-5 min attain stiff state average density. As calcium phosphate little dissolve the ima, when sediment is cocrystallization, coprecipitation and sorption of radionuclides, i.e. additional cleaning. At this pH environment of the mixture should be in the range of 9.7 to 9.9. [Precipitation methods LRW purification: HOWTO / LTI them. Lensoviet. - L., 1989. - c.6-7].

, The mixture out of the container-dispenser with metered sample is injected catalytic additive in the form of powders "white soot", shungite or pyroxylin, then a portion of bentonite clay, which is 0.5-1% of the mass MC, and carry out stirring for 3-5 minutes Catalytic additive in amounts of 0.2-0.6 wt.% by weight of dry radioactive salts promotes the combination of components of liquid and MC with more stable compounds. Bentonite clay is a sorbent for strontium-90 activity in which LRW commensurate with the activity of cesium-137, and an additional sorbent for cesium. Then in the mixer portions pour the caustic magnesite powder (brand PMK-87) with constant stirring magnesium test. When LRW becomes thick, add water or LRW, then again a portion of the PMK, the water or liquid and so come to a complete emptying of the dispenser. The obtained cement paste of the desired consistency, pour in a regular 200L iron drum or sunk in concrete protective containers NSC for the storage, Tran is porterhouse and disposal of low and medium activity. Capacity stand in the air before turning the compound into the monolith.

The positive properties of the developed composite material and method of its use for processing of liquid raw monolithic (creation of magnesia compound) are confirmed experimentally.

Example 1. Determination in accordance with the invention according to the mechanical stability of the solidified compounds from a ratio of the curing residual ingredients - chloride (sulfate) magnesium powder and caustic magnesite (PMC-87).

In the experiments as LRW used real distillation residues from the installation distillation LRW with the salinity of 400-600 g/l, sulfate and magnesium chloride brand "H" (GOST 7759-73) and caustic magnesite powder (PMC-87), made according to GOST 1216-87. This powder obtained by dust when firing natural magnesite. He 82-83% of MgO, contains up to 2.5% Cao and 2.5% SiO2and has a density of 3.1 to 3.4 g/cm3.

The dependence of the strength of solidified LRW Laundry on the ratio PMC/MgCl2shown in table 1. The table shows that when creating a cementitious compounds the ratio of PLA/MgCl2(MgSO4must be not less than 4:1.

Table 1
The influence of correlation PMC/MgCl2from 2:1 to 4:1 on the mechanical stability of the compounds in water
The cement composition g*The ratio of PMK-8
MgCl2(MgSO4)
View samples after soaking in water**
PMC-87MgCl2(MgSO4)
8,04,02:1Fell apart after 1 day
8,63,42,5:1Fell apart after 3 days
9,03,03:1Many cracks
9,42,7a 3.5:1Less than expert. 3
9,62,44:1No cracks
*) - content LRW all compounds were equally - 12 g (600 g/l), as catalytic additives used a sample of shungite or white carbon black, the weight of which was 0.2-0.3 wt.% from m the ssy reagents of the Central Committee.
**) - drying time samples after manufacturing in the room at 18-20°C was 10 days, and extracts them in water - 20 day.

Example 2. Confirmation of the effectiveness of the recommended the invention of the sorbent for the retention of caesium-137 in magnesium compounds.

To compare the behavior of cesium-137 in compounds without the inclusion of selective sorbents and efficiency of sorbents used for its retention, were produced compounds of magnesia cement and Portland cement-400. In the experiments used natural sorbents: zeolite (C), vermiculite (), bentonite (B), a natural mixture of bentonite (40-45%) with zeolite (33-35%), occurring on the territory of Belgorod region, as well as sintetizirane in the proposed ferrocyanide compound of Nickel potassium. The amount of sorbent in the compounds was 10% by weight of cement, using FCNC other sorbents in them was not.

According to GOST 29114-91 leaching in distilled water at a temperature of 25°C recommended for 1, 3, 7, 10, 14, 21, 28 days, and the next 10-14 days. Testing stops when the rate of leaching becomes almost constant (limit of measurement accuracy ±10%).

To determine the activity of cesium-137 videlity passed in cones in gamma spectrometric measurements to the e was performed on the analyzer with the semiconductor germanium-lithium detector.

Each of the contact solution was determined by the activity of cesium and calculated the percentage of activity transferred from samples in solution, and the rate of leaching (g/cm2·day) by the formula:

Requilibrium=a·m/A0·s·ν=K·a/A0·ν, g/cm2·day

whereandradioactivity in an aliquot of cesium-137, leached for a period of time;

And0specific radioactivity nuclide in the original sample, was (3-6)·105Bq/sample by introducing the compound "strong" solution of cesium-137;

s - open "geometric" surface of the sample, cm2;

ν is the length of the n-th period of leaching, day;

m is the mass of the sample;

K=m/s.

When conducting laboratory experiments, the coefficient K=m/s was equal to approximately 1/2. From an analysis of the formula of the rate of leaching, it follows that to obtain it with a numerical value of not more than 1·10-3it is necessary that the ratioand/A0do not exceed 1/20. That is, during the test (100 days) of the samples should vydeliajutsia not more than 20% of the initial activity: Requilibrium=1·10-3=1/2·20/100·1/100.

The results of the leaching of cesium-137 from compounds of different composition, containing and not containing selective sorbents for cesium, are given in table 2, from the analysis we can draw the following conclusions:

matrix about what their cements are not a barrier for cesium-137 (see the results for samples 1.1 and 3.1), and lose 85-98% activity;

matrix cements containing CaCl2and shungite (see clause 1.2, 3.2 and 3.3)have the same shortcoming: 40 days of samples goes 85-82% activity;

- when using Portland cement most effectively Sorb cesium B-C-clay, the second efficiency is bentonite clay

- when using MC most effectively Sorb cesium synthesized in the matrix ferrocyanide Nickel-potassium (FCNC), the second efficiency is B-C-clay, then bentonite clay; the rate of leaching was 2·10-5, 4,5·10-4and 7.5·10-4g/cm2·day, respectively.

In the manufacture of cement samples with synthesis of Nickel ferrocyanide besieged no less than 95-97% cesium. But then with the first 6-7 Pixelate happened leaching of cesium-137 80 percent or more by dissolving the most part formed connection Cs4[Ni(CN)6]as it steadfastly only up to a pH of not more than 10.

Feature magnesia cements is that when in prolonged contact with water (50-100 days) pH in them are about 6.5-6.8. In MC, the pH value is determined by the amount of alkali in the liquid and, as a rule, does not exceed the values 10-10,5.

Table 2
Composition curable mixtures when poured LRW and chemical resistance of the obtained sample
The composition of cement compoundRemote from the sample activity, %The average rate of leaching, g/cm2·day
1.1. LRW+cement985·10-3
1.2. LRW+CaCl2+cement854·10-3
1.3. LRW+CaCl2+cement+FCNC824·10-3
1.4. LRW+cement+CaCl2+C231,1·10-3
1.5. LRW+cement+CaCl2+Bof 5.42,7·10-4*
1.6. LRW+cement+CaCl2+B-C1,26·10-5
3.1. LRW+MC854·10-3
3.2. LRW+CaCl2+MC834·10-3
3.3. CaCl 2+shungite+MC824·10-3
3.3. LRW+C+CaCl2+shungite+MC402·10-3
3.4. LRW+In+CaCl2+shungite+MC493·10-3
3.4. LRW+B+CaCl2+shungite+MC157,5·10-4
3.5. LRW+CaCl2+shungite+MC+B241,2·10-3
3.6. LRW+B-C+CaCl2+shungite+MC94,5·10-4
3.7. LRW+FCNC+CaCl2+shungite+MC0,42·10-5

During the process of cementing after the introduction into the mixer setup LRW need to measure the pH and, if necessary, as prescribed by the proposed method, down to values of 9.7 to 9.9 concentrated hydrochloric acid solutions.

You should pay attention to the results of the experiments 3.4 and 3.5. They one and a half times differ in the amount of radioactivity released from the compounds, in spite of the same composition due to the fact that in clause 3.5) b is tonic was introduced in the prepared mixture of the last component, in 3.4 it had previously introduced in LRW. Conditions sorption of cesium-137 in the second experiment were much better. Therefore, to obtain the required quality parameters MC should strictly observe the sequence of input components and the temporal characteristics of the process.

In the literature indicates the optimum conditions for the co-precipitation of cesium-137 sediment FCNC, and the effect of organic complexing surfactants of this process is only for multiple solutions. For high salt LRW Laundry they are missing. Therefore, based on guidelines of LTI imlincomb, according to which for the efficient co-precipitation of cesium-137 concentration FCNC in LRW should be 0,002-0,005 mol/l ToPTScesium can be up to 103. But if LRW contain, for example, the oxalate ions in amounts of 10-100 mg/lcleaningreduced at least one order of magnitude (10 times). In such a complex mixture, which is LRW, only experimentally, one can determine the optimal concentration of ferrocyanide of potassium and nitrate of Nickel, which must be new to 95-97% Coosada trace amounts of cesium with sediment FRNK. In the above experiments with obtaining FCNC its concentration was 0.002 mol/L.

Example 3. Providing the maximum degree of filling the MC dry radioactive salts LRW Laundry.

In the former is erementar as LRW used real residual (CO) from the installation of distillation waste. Number of ingredients magnesium compounds used in the experiment, obtained the degree of filling and behavior of the samples are shown in table 3.

Table 3
The composition and properties of MK with a high content of dry salts
PMC-87, gMgCl2LRW, *gThe degree of filling compound salts, %The behavior of the MC input
731030collapsed
1135racks
1337racks
*) for accurate calculation of the degree of filling compounds dry salts simultaneously selected 2 parallel samples LRW, one was dried to constant weight, and the second was used for cementation; the mass of solids in the sample 3 was the equal of 5.6 g if the mass of the Central Committee after drying, 15, Shungite took in the amount of 0.2% by weight of the reactants MK; for synthesis FCNC took 0.5 ml of a 14% solution of potassium ferrocyanide and 0.75 ml of a 13% solution of Nickel chloride (molar ratio of these reactants was 1:1,5). Weight bentonite clay - 1% by weight of the reactants PMK+MgCl2.

After the manufacture of the compounds of their first withstood 30 days (for curing) in a dry environment, then tested for resistance in the aquatic environment for 50 days, and then dried to constant mass and then using a laboratory hydraulic press has identified the mechanical strength of the samples of series 2 and 3, which was at the level of 5.5-5.0 MPa, respectively. The degree of filling compounds dry salts for experiment 3 was reached 37% at the minimum acceptable strength of the specimen in compression.

This experiment will recalculate the actual concentrations of all ingredients included in the composition of the magnesium compound, based monolithic 1 kg TO real, with an average of from 440 to 460 g of dry radioactive salts.

In CO it is necessary to introduce an additional approximately 90 g MgCl2and 90 g CaCl2(10-20% by weight of dry salts). With regard to water of crystallization of the values of these masses should be multiplied by about 2. Therefore, the input mass of water with reagents will be about 180,

In the resulting compound to bind all the water you Budejovice 440-450 g PMC-87, containing up to 83% MgO. In terms of 100% substance is the mass of magnesium oxide will be approximately 367, Therefore, the mass of the components of the MC will be considered equal to 367+90=457 BC Recommended additives in the compound powder of shungite should be 0.3 to 0.5 wt.% by weight of component MC (1.4-2.3 g), and bentonite clay - 0,5-1% (2.3 to 4.6 g) per 1 kg of KOH. The mass of the synthesized selective sorbent FCNC in MC should be about 2.7 g 440 g of dry radioactive salts (0,006 wt.%). If you will be cemented to the salt content of not 600, but, for example, 150-200 g/l, for effective sorption of cesium-137 will be sufficient to create a concentration FCNC 0.002 wt.% from the mass of radioactive dry salts.

Thus, reliable monolithic LRW containing up to 30% of organic matter, with a degree of inclusion of dry radioactive salts 35-37%. Proposed in the invention is a composite material, the main components of which are common minerals and reagents manufactured by the domestic industry, allows you to be cured of LRW complex chemical composition, which are currently not subject to cementing. The method of application of the inventive composite material for immobilization of liquid radioactive waste does not require high energy costs and is carried out with any positive temperatures on the equipment used is ri conventional cementing.

1. Composite material for immobilization of liquid radioactive waste, including hardener, powder caustic magnesite and catalytic carbon additive, characterized in that as a hardener use solid salt, and further added a solution of ferrocyanide of potassium and Nickel nitrate, and calcium chloride in the following ratio, wt.%:

powder caustic magnesite27-28
solid salt5-6
calcium chloride (CaCl2)0,1-6
catalytic carbon additive0,1-0,2
a solution of ferrocyanide of potassiumthe 0.05-0.1
a solution of Nickel nitratethe 0.05-0.1
liquid radioactive wasterest

2. Composite material according to claim 1, characterized in that the solid salts used 6-water magnesium chloride (MgCl2·6H2O).

3. Composite material according to claim 1, characterized in that the solid salts use 7-waters of the initial magnesium sulfate (MgSO 4·7H2O).

4. Composite material according to any one of claim 2 and 3, characterized in that as catalytic additives used white soot.

5. Composite material according to any one of claim 2 and 3, characterized in that as catalytic additives use shungite.

6. Composite material according to any one of claim 2 and 3, characterized in that as a catalytic additive use pyrocarbon.

7. The method of applying composite material for immobilization of liquid radioactive waste, consisting in their concentration and curing using a composite material comprising a hardener, powder caustic magnesite and catalytic carbon additive, curing the mixture to form a strong solid monolithic block, locking in its structure components of radioactive waste, characterized in that as a hardener use solid salt, and the composite material further added a solution of ferrocyanide of potassium and Nickel nitrate, and calcium chloride,
in the following ratio, wt.%:

powder caustic magnesite27-28
solid salt5-6
chlorine is the ID of calcium (CaCl 2)0,1-6
catalytic carbon additive0,1-0,2
a solution of ferrocyanide of potassium and
of Nickel nitrate (FCNC)the 0.05-0.1
liquid radioactive wastethe rest,

moreover, the curing is carried out by introduction of liquid radioactive waste of the ingredients of the composite material in the following sequence: 1 - solid salt, 2 - a solution of ferrocyanide of potassium, 3 - solution of Nickel nitrate, 4 - calcium chloride, 5 catalytic carbon-containing additive and 6 - caustic magnesite powder with continuous stirring of the mixture.

8. The method according to claim 7, characterized in that the solid salts used 6-water magnesium chloride (MgCl2·H2O).

9. The method according to claim 7, characterized in that as a solid salt use 7-water magnesium sulfate (MgSO4·H2O).

10. The method according to any of PP and 9, characterized in that as catalytic additives used white soot.

11. The method according to any of PP and 9, characterized in that as catalytic additives use shungite.

12. The method according to any of PP and 9, characterized in that the AC is este catalytic additives used pyrocarbon.



 

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4 cl, 2 dwg, 6 ex

FIELD: power engineering.

SUBSTANCE: invention relates to processing of liquid radioactive wastes (LRW), mostly, nitrate ones, containing alkaline and alkali-earth elements, also salts of sodium, radio isotopes 137Cs and 90Sr. The method to immobilise radioactive wastes in a mineral-like matrix includes synthesis of mineral with high content of radioactive elements and mixing particles of the radioactive mineral with a filler in a glass-like condition, consisting of stable materials, with subsequent thermal treatment of produced charge at high temperatures. Synthesis of the mineral and mixing with the filler are carried out simultaneously, and the mineral producing agent and the filler are cenospheres from fly ashes from coal burning, which are mixed with solution of radioactive wastes in the specified ratio, afterwards the cenosphere-water mixture is evaporated to produce a cenosphere-salt composition, which is exposed to calcination at the temperature of 750-1100°C within the specified time to produce a glass crystalline material that contains mineral-like phases-fixators of radio nuclides.

EFFECT: simplification and reduction of energy intensity of hardening process of liquid LRW in the form of mineral-like ceramics.

4 cl, 2 dwg, 4 ex

FIELD: power industry.

SUBSTANCE: immobilisation method of liquid radioactive waste to ceramics involves concentration of radioactive solutiona, its mixing with phosphate matrix and further heat treatment. Waste concentrated to the level of high-active waste is calcinated after having been mixed with bone phosphate till ceramic sinter is obtained. Sinter is capsulated into glass. All the process stages are performed in one and the same reaction vessel. Vitrification is performed at temperatures not exceeding 1000°C. Invention allows obtaining stable mineral-like structural forms: kosnarite [(Na, Cs, Sr, Ln)(Zr, An, Fe)2(PO4)3], monazite [(Ln, An)PO4], zirconium oxide [(Zr, Ln, An)O2], which have essentially large capacity approximately in relation to all radionuclides.

EFFECT: during one process both fixing of radionuclides to stable polycrystalline matrix, and creation of additional physical barrier in the form of glass capsule takes place.

2 cl, 3 ex

FIELD: power industry.

SUBSTANCE: preparation method of liquid high-active wastes for vitrification so that borosilicate glass is obtained consists in mixing of waste with liquid glass-forming element. First, nitric acid is added to liquid waste. Hydrous kremnezol is used as liquid glass-forming element. Mixing of liquid waste with hydrous kremnezol is performed on the basis of borosilicate glass with content of silicon oxide of 35-55 wt %.

EFFECT: improving mixing quality of waste with glass-forming element, providing continuous supply of mixture to the melting device without formation of any suspension plugs in pipelines, and achieving homogeneity of glass mass and final product, achieving high chemical stability of the obtained glass, and improving safety and reliability of vitrification of waste.

3 cl, 3 ex, 4 tbl

FIELD: chemistry.

SUBSTANCE: method for low-temperature immobilisation of liquid radioactive wastes (LRW) involves converting LRW components into a solid glass-like phase via treatment thereof with chloride water-alcohol solutions of silicon- and boron-containing compounds and then holding the reaction mixture at temperature 5-60°C in a sol-gel process. Powdered cement is added to products of chloride water-alcohol hydrolysis of LRW and alkyl silicates. The reaction mixture is stirred until a solid gel-like matrix forms.

EFFECT: shorter duration of the hydrolysis step followed by formation of a self-packing glass-like solid mass.

3 tbl, 2 ex

FIELD: ecology.

SUBSTANCE: method of solidifying radio active wastes and other dangerous wastes involves heating, saturation with a solution until achieving the required degree of filling, binding and immobilisation of radionuclies and other dangerous wastes inside a block. The block is made from crystalline hydrates of synthetic mineral salts by pouring molten crystalline hydrates into a mould containing liquid wastes. The melt is cooled until formation of solid crystalline hydrates to obtain a solid block.

EFFECT: low power consumption and labour input, high safety of moving, loading and storing solidified wastes.

54 cl, 5 dwg

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: power industry.

SUBSTANCE: water treatment method of natural or artificial water reservoir from radioactive isotopes and harmful chemical substances involves the intake of source water, its pre-treatment and the main treatment by two-stage reverse osmosis so that filtrate is obtained, which is supplied to consumer as treated water and concentrate returned to water reservoir. Combined concentrate obtained at the stage of pre-treatment and the first stage of reverse osmosis is returned to water reservoir, and concentrate obtained at the second stage of reverse osmosis is supplied to the first stage of reverse osmosis.

EFFECT: invention allows improving the efficiency of treatment procedure and reducing the amount of secondary wastes.

6 cl, 1 dwg, 1 ex, 1 tbl

FIELD: water treatment.

SUBSTANCE: innovation relates to the chemical treatment of industrial and domestic wastewaters containing lubricating and cooling fluids, radioactive contaminating substances, washing solutions, ions of heavy metals; the method includes the decontamination and demulsification of wastewater by the mixture of ferrous chloride (FeCl3), calcium chloride (СаСl2) and permanganate of alkalic or the mixture of alkaline-earth metals at the mass ratio of (1/2):(0.15/0.2):(0.01/0.02) on conversion to arid salts and volume concentration of 1-3% of the initial solution, рН counts to 5-6.

EFFECT: innovation enhances the efficiency of wastewater purification.

1 cl, 5 tbl

FIELD: power industry.

SUBSTANCE: processing method of low-mineralised liquid radioactive wastes involves filtration through selective ferrocyanide absorbent, and then through ion exchange absorbents. Absorbent is obtained by sequential processing of organic carrier with solutions of potassium ferrocyanide and nickel salts with excess content of each reagent. As organic carrier for obtaining selective ferrocyanide absorbent there used is saw dust pre-dried at temperature of 105-110°C with sizes of 1-4 mm. At that, nickel-potassium ferrocyanide is synthesised immediately in structure of saw dust. Used radioactive ferrocyanide absorbent is burnt. Ash residue is added to cement compound as binding agent.

EFFECT: invention allows increasing by more than 10 times the capacities of selective ferrocyanide absorbent on the basis of pre-dried saw dust in comparison to application of nickel-potassium ferrocyanide to ion exchange resins; increasing operational life of sorption filters; cheapening of preparation technology of selective absorbent and reducing the volume of secondary disposed radioactive wastes.

3 ex

FIELD: chemistry.

SUBSTANCE: invention relates to the technology of decontaminating liquid radioactive wastes using a membrane-sorption method in field conditions. The method of decontaminating low-mineralised low-activity liquid radioactive wastes is realised through cleaning on mechanical and ultra-filters, desalination on reverse-osmosis filters and post-cleaning on ionite filters with post-evaporation of the formed radioactive concentrates until salt saturation at temperature lower than 100°C in a container and subsequent addition of the saturated salt concentrates to portland cement. Also, wastes are cleaned from radio nuclides on filters using selective sorbents with protection from ionising radiation. Post-evaporation of concentrates from reverse-osmosis filters is carried out at atmospheric pressure in a container. The condensate from the evaporation apparatus is returned to the ultra-filters. The concentrate from the ultra-filters is returned to the mechanical filters. Initial radioactive concentrates with beta-radiating nuclides specific activity greater than 0.1 MBq/kg and/or radioactive concentrates with specific activity greater than 0.1 MBq/kg before evaporation are directed to the selective sorbent filters.

EFFECT: more efficient removal of radionuclides even without using selective sorbent filters.

1 dwg, 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

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