Method of neutralization of the low-mineralized and medium- mineralized low-active liquid wastes in the field conditions

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

 

The invention relates to the field of decontamination of liquid radioactive waste (LRW) membrane-sorption methods and can be used for water purification from radionuclides on mobile installations processing of LRW in the field.

In the operation of nuclear power plants and other nuclear facilities in addition to education reagent LRW (decontamination, cleaning, recycling solutions, and others), characterized by a high salinity and radioactivity, there is a pollution of significant amounts of low and medium - mineralized natural (including marine) waters radionuclides to concentrations exceeding permissible only by 3-4 orders of magnitude. Such wastes are often formed on the objects do not have their own water treatment plants, i.e. requiring the use of mobile (transportable) installations in the field.

There is a method of disposal of low mineralized water in field conditions at the facility, including the clearance to mechanical and ultrafilter, desalination by reverse osmosis filters and purification on a regenerated ion exchange filters with curing the formed radioactive concentrates included in Portland cement [1]. This method in its technical essence and the achieved effect closest to the proposed and selected as a prototype.

OS is ESD a disadvantage of this method is its low efficiency in the processing of moderately mineralized liquid waste (for example, radioactively contaminated sea water salinity of 35 g/l). In fact, when the salt content of processed LRW more than 3-4 g/l reverse osmosis does not provide desalination solutions below 0.2 g/l (odds purification by monovalent salts 15-20 [2]). At the same time when the purification on ion-exchange filters, fluids with a salt content of more than 0.2 g/l (upper limit of the optimal use of ion-exchange filters [3]) there is a rapid saturation of the ion exchange filter that requires regeneration (handle H2SO4and NaOH) or frequent replacement. In any case, there is a significant increase in the volume of disposed waste by cementation of reverse osmosis concentrates and regenerates (during curing concentrate to 200 g/l the degree of inclusion in the cement to 7.5% salts [4]), and for qualitative regeneration requires 4-5 fold excess of reagents, or by cementing waste ion-exchange resin (degree of inclusion in the cement, not more than 10-12% by dry weight [5]). In addition, if clearing from90Sr solutions, the salinity of which is determined by sodium salts, ion-exchange filters up to 104again, the wipe from the137Cs is determined by the degree of demineralization, i.e. removing all sodium salts [6].

Remove137Cs effectively produced by precipitation from solutions of Fe is rociando of such metals, as Fe, Ni, Cu, Zn. The most commonly used Nickel ferrocyanide, as it is less sensitive to salt composition of waste and saves sorption properties in a wide range of pH [7]. However, the bleaching solutions after chemical treatment of the ferrocyanides Nickel-potassium is very time consuming, and the resulting precipitates are high (95-99%) humidity, resulting in their subsequent cementing to a significant increase in the volume of buried waste.

The problem solved by this invention is the extension of the scope of mineralization processed LRW, increased degree of purification from radioactive caesium and the reduction of the volume of solidified radioactive waste without compromising their quality.

The invention consists in that in the method of disposal of low and medium - mineralized low-level liquid wastes in the field, including clearance to mechanical and ultrafilter, desalination by reverse osmosis filters and purification on ion-exchange filters with chemical treatment of spent ion-exchange resins and curing the formed secondary radioactive waste included in Portland cement, chemical treatment of waste ion-exchange resins are not acid (H2SO4) and alkali (NaOH)and potassium ferrocyanide (K4[Fe(CN)6]) and salts Koba is it, and then the treated resin is used as the sorption prefilter before ionoobmennym filter.

The method is as follows.

Moderately mineralized (up to 35 g/l dry residue) low (up to 10-5CI/l), mainly chloride-sulphate-bicarbonate, liquid waste is sent for mechanical and ultrafilter to remove sediment and oil products. The waste is fed to the desalination (reduction of salt content in 15-20 times, that is, to values not more than 1.8-2.3 g/l) on the reverse osmosis filter and purification on sorption prefilter and ion-exchange filter (salt content of the effluent not exceeding 1.0 g/l and a specific activity of not more than 10-9CI/l). Regeneration of ion-exchange filter is not performed, and download prefilter is exhausted ion exchange resins (ion exchange filter), treated with ferrocyanide of potassium (K4[Fe(CN)6]) and a salt of cobalt (e.g., CoCl2•6N2About). At prefilter is the purification of the principal amount of radioactive caesium, and ion-exchange filters from Radiostantsiya and other radionuclides. Formed during the neutralization of secondary radioactive waste utverjdayut included in Portland cement. And with the stage of treatment in cementing send only spent ion-exchange resin of th the development of a resource prefilter. In the process of neutralization is achieved total removal of radiocesium not less than 104time. The degree of inclusion in the Portland cement waste ion-exchange resin after chemical treatment and use as prefilters reaches 20% by dry weight, and BaselCement of radiocaesium from hardened product after 90 days is less than 1•10-4g/cm2•d.

In comparison with the known membrane-sorbtsionnymi ways neutralization LRW this method not only ensures the removal of radiocesium not less than 104even when the salt content of up to 35 g/l, but leads to a reduction in the volume of solidified with Portland cement ion exchange resins at higher fixation strength of them radiocesium, which is not obvious from the prior art, i.e. meets the criterion of inventive step.

Examples of specific performance.

Example 1. As a moderately mineralized low-level liquid waste used dirty sea (salinity 35 g/l) water containing 24,0 g/l NaCl, 5.0 g/l MgCl2, 3,9 g/l Na2SO4, 1.1 g/l CaCl2, 0.7 g/l KCl, 0.2 g/l NaHCO3, 0.1 g/l KBr, 0.05 g/l of iron and 0.03 g/l of petroleum products (suspended solids 0.1 g/l). Specific activity was 5·10-6CI/l for cesium-137 and 5·10-6CI/l for strontium-90.

Neutralization of PR is led by cleaning on the mechanical and Ultrafiltered from radionuclides adsorbed on suspended solids and colloids, followed by desalting by reverse osmosis filter and purification on ion-exchange filters (KU-2 in the N+form and AV-17 in the HE-form) from radionuclides that are part of complexes and salts. Reverse osmosis filter worked at a pressure of 7 MPa with desalination to 2 g/l and the concentration of LRW to 200 g/liter of Ion-exchange filter 1 volume of resin provided desalting up to 40 mg/l to about 20 volumes of solution (without reverse osmosis desalination, i.e. at 35 g/l is only about 1 volume).

Formed during the neutralization of secondary radioactive solutions were utverjdali the Portland cement GOST 10178-85 (Portland cement or slag Portland cement grade 400). Reverse osmosis concentrate (200 g/l) was mixed with cement in the ratio of 0.6:1.0 and spent ion-exchange resin with water and cement in the ratio of 0.17 to 0.2:0,43-0,46:1.0, and utverjdali within 28 days.

The specific activity of purified water was 5·10-9CI/l for cesium-137 and 5·10-10CI/l for strontium-90. Caulk cement compounds with salt concentrate had a strength of more than 10 MPa, and with ion exchange resins for more than 5 MPa at leachability of cesium-137 (after 90 days) about 1·10-3g/cm2·d that meets the technical requirements of RD 95 10497-93 [8]. The volume of the solidified waste sent for Zaho is onania, were, respectively, 26% and 8-9% (for a total of up to 34-35%) of the volume of the original waste.

Example 2. Differs from example 1 in that the spent ion-exchange resin was not utverjdali and regenerates 5-fold excess of N2SO4and NaOH. Exhaust alignment is sent to the reverse osmosis filters for purification and concentration, and the resulting concentrates were cemented. The volume of solidified concentrate (200 g/l) sea salt and regenerates were respectively 26% and 8-9% (for a total of up to 34-35%) of the volume of the original waste.

Example 3. Differs from example 1 in that the spent ion-exchange resin was not utverjdali, and was treated with a 0.5 M solution of ferrocyanide of potassium (K4[Fe(CN)6]), and then a 0.5 M solution of cobalt chloride (CoCl2•6H2O). The treated resin was used as the sorption prefilter before freshly loaded ion exchange filter and utverjdali only after the development of the resource for radiocaesium (breakthrough of cesium in the filtrate). In this case, fresh ion exchange filter worked up to its saturation salts of rigidity, that is, one volume of resin was purified up to 200 volumes of solution after the reverse osmosis filter (TDS up to 2 g/l). Approximately the same volume stood and sorption prefilter, providing the final content of cesium-137 in the filtrate of not more than 1·10-10CI/L. After Virab the TCI resource resin from the ion exchange filter was subjected to chemical treatment and prefilter, and resin from prefilter sent to curing. Spent resin from prefilter was mixed with water and cement in the ratio of 0.43:0,65:1,0. Despite more (2 times) filling the resin and 12-13% less consumption of cement in comparison with example 1, the strength of the cured product was more than 10 MPa, and BaselCement cesium-137 (after 90 days) less than 1·10-4g/cm2·d that meets the security requirements for radioactive cement compounds in the open ground [9]. The volume of solidified concentrate (200 g/l sea salts and resins from the sorption prefilter were, respectively, 26% and 0,7-0,8% (for a total of not more than 27%) of the volume of the original waste.

The proposed method extends the range of salt content of processed LRW up to 35 g/l, which is particularly important when the decontamination of radioactively contaminated marine waters. When this is achieved the total removal of cesium-137 and strontium-90 to 104time that provides for low-level LRW can reset neutralized water into the environment. The volume of the solidified waste ion-exchange resin is reduced 10 times, and increasing 10 times the strength of fixation of radiocaesium allows you to dispose of such cement blocks in the simplest ground burial.

The proposed method can be implemented on the same Fatherland, the public equipment, as the prototype (for sorption pretreatment uses the same filter as for the ion exchange treatment), i.e. industrially applicable. Reuse of waste ion-exchange resin as a selective sorbent not only reduces the total volume of buried solidified waste by 30%, but also increases their quality and environmental safety.

Sources of information

1. Yepimakhov V.N., Oleinik MS Method for the disposal of low-mineralized low-level radioactive waste in the field. - RF patent №2144708, bull. No. 2, 2000.

2. Dynarski SCI, Pushkov A.A., Switze A.A. and other Purification and concentration of liquid wastes with low levels of radioactivity reverse osmosis. - Atomic energy, 1973, t.35, issue 6, s-408.

3. Janicevic A.A. Purification of contaminated water. - M, Atomizdat, 1974, s.

4. Sobolev, I.A. and other Disposal of radioactive waste at centralized points. - M, Energoatomizdat, 1983, p.40-45.

5. Bonnevie-Svendsen M. Tallberg K., Aittola P., e.a. Studies on the incorporation of spent ion-exchange resins from nuclear power plants into bitum and cement. - In: Symposium on the ion-site management of power reactor wastes, Zurich, 26-30 Marh, 1979, Paris, 1979, p.155-174.

6. Rausen F. W., Solovieva SA Removal of radioactive isotopes from wastewater. - Atomic energy, 1965, Vol.18, issue 6, s.623-626.

7. Nikiforov A.S., Aulchenko CENTURIES, Zhikharev M.I. Disposal of liquid radioactive waste. -M, Energet is misdal, 1985, p.36.

8. The quality of the compounds formed by the cementation of liquid radioactive waste of low and intermediate level waste. - Technical requirements. - RD 95 10497-93. - M.: Minatom of the Russian Federation, 1993.

9. Bazhenov, Y.M., Volkov, I., Dukhovich FS and other safety Conditions during storage of radioactive cements. - Isotopes in the USSR, 1970, V.17, p.17-22.

The method of disposal of low and medium - mineralized low-level liquid wastes in the field, including clearance to mechanical and ultrafilter, desalination by reverse osmosis filters and purification on ion-exchange filters with chemical treatment of spent ion-exchange resins and curing the formed secondary radioactive waste is included in the Portland cement, characterized in that the chemical treatment of waste ion-exchange resin is carried ferrocyanide and potassium salts of cobalt, and then the treated resin is used as the sorption prefilter, which produce waste treatment prior to the ion exchange filter.



 

Same patents:

FIELD: radioactive waste processing by the burning method.

SUBSTANCE: the proposed furnace for radioactive waste burning has a case, within of which non less three burning chambers are coaxially located. The chambers have a general chamber for ash reburning. The general chamber is equipped with a fire grate. Each of burning chambers is autonomous, has the device for supply of fuel and oxidizer and is equipped with a fire grate and gate. The chamber for reburning the waste gases is coaxially located in the central part of the furnace. This chamber consists of external and internal cases, forming a labyrinth gas duct. The burning chambers are connected with the chamber for reburning the waste gases by means of gas-escape channels, having unequally-height and unequally-directional levels of location in each two neighboring burning chambers. A pressure-tight lock chamber of cylindrical shape is located over the burning chambers and chamber for waste gas reburning. The lock chamber is general for all burning chambers and is equipped with a cover, charging branch, pipe of exhaust ventilation, heat exchanger, transport system and control unit. The unit for control of the drives of transport system, fire grates, gates and strips is located on the lock chamber cover. The transport system consists of drive and container with a tray and limit switch. The container is fulfilled in the form of horizontally located cylinder. It is foreseen the circular motion of the cylinder. The drive of the transport system is connected with the container by means of an open gearing and is equipped with a gear, located on the output shaft. The output and input pipes for acceptance and unloading of wastes are located with good alignment one under other on the cylindrical surface of the container. The bushings are located on the side end surfaces of the container. The tray has possibility of rotation and has the shape of a bed, repeating the curvature of the internal surface of container case. Besides, the semi-axes are located on the semicircular side end surfaces of the container. These semi-axes together with the container bushings form a movable joint of the type of axis-bushing.

EFFECT: increased economical efficiency, processing efficiency, operation reliability and environment safety.

1 cl, 7 dwg

FIELD: recovery of liquid radioactive wastes.

SUBSTANCE: proposed method for recovering alpha-active nitric acid solution containing trivalent iron includes pre-concentration of solution being recovered by its evaporation to produce regenerated nitric acid in still bottoms, whereupon still bottoms are neutralized to pH of 1 - 2 and trivalent iron is partially recovered by sodium sulfate until valence forms ratio Fe3+ : Fe2 = 2 : 1 is attained. This is followed by next neutralization with alkali to pH = 10 - 11. Magnetite sediment obtained after settling down is conveyed for curing. Then solution is decanted, clarified solution is magnetically separated and additionally cleaned.

EFFECT: reduced volume of secondary wastes, reduced consumption of chemical agents used for the purpose.

2 cl, 1 tbl, 1 ex

FIELD: decontaminating natural water reservoirs from radionuclides.

SUBSTANCE: proposed method includes introduction of sorbent in contaminated water body. Used as sorbent is zeolite powder or man-produced substance, such as red slime or slime production waste. Mirabilite is used as precipitant and in addition water is decontaminated by using water-plant, such as reed and pondweed planted out in water reservoir in advance.

EFFECT: enhanced efficiency of water reservoir decontamination from radionuclides.

1 cl, 1 ex

FIELD: recovery of liquid radioactive wastes.

SUBSTANCE: proposed method is used for localizing spent granular, powdered, or milled ion-exchange resins in Na- or H-form in dry or wet condition by including them in solid matrix. Matrix base is made of blast-furnace slag milled to fractions below 0.075 mm and tempered with sodium hydroxide solution of 100 - 150 g/l concentration.

EFFECT: enhanced degree of filling compound with wastes and extended range of its application.

1 cl, 1 dwg, 8 ex

FIELD: treatment of heterogeneous liquid radioactive wastes.

SUBSTANCE: proposed method includes extraction of radium from radioactive oil slimes by means of hot water, acid or alkali solutions. Before doing so radioactive oil slime is subjected to recovery annealing with lack of hydrogen in atmosphere of incomplete combustion of carbon and hydrocarbons produced by using oil products. Recovery annealing temperature is maintained between 700 and 900 °C for 1 to 3 h. Annealed oil slime is treated with hot steam and once more with heat steam and sulfuric acid at concentration of the latter between 5 and 10% relative to mass of extracting solution.

EFFECT: enhanced quality of radioactive waste treatment.

4 cl, 1 dwg, 1 tbl, 1 ex

FIELD: immobilization of heterogeneous radioactive wastes.

SUBSTANCE: proposed method includes production of dehydrated radioactive sediment and filtrate on filtering centrifuge; heating of dehydrated radioactive sediment at 500 - 600 °C; crushing of products of heating into fragments measuring maximum 30 mm; case-hardening of crushed fragments with high-penetration cement solution which is, essentially, mixture of cement having specific surfaced area of minimum 8000 cm2/g and liquid phase at liquid phase-to-cement mass proportion of 0.6 - 1.4; for the final procedure mixture obtained is cooled down.

EFFECT: reduced amount of radioactive wastes, enhanced radiation safety, and reduced power requirement.

2 cl, 1 tbl, 2 ex

The invention relates to a method for protection of groundwater from pollution by radionuclides overflow of polluted pond during the rainy season, floods and similar events

The invention relates to the protection of the environment and can be used for decontamination of radioactively contaminated territories

The invention relates to containers-storage of liquid industrial waste containing radionuclides of strontium and cesium in natural soils, and can be used for building system-induced geochemical barriers that purify water from radionuclides

The invention relates to containers storage of soil or plant residues, contaminated with radioactive isotopes of strontium and cesium in natural soils, and can be used for the construction of a system of man-made barriers limiting distribution in surface and ground waters, soils radioactive isotopes

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: 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: 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

The invention relates to the field of processing of liquid radioactive waste, in particular, to methods of extraction of precious metals
The invention relates to the field of processing of liquid radioactive waste

The invention relates to the field of processing of liquid radioactive waste

The invention relates to the field of chemical technology and can be used in the purification and concentration of toxic solutes, including radioactive high level of activity
The invention relates to the field of disposal of liquid waste, in particular to the disposal of liquid nitric acid waste containing radioactive substances
The invention relates to the field of chemical technology and can be used for neutralization and decontamination of radioactive solutions and waste waters containing Th-232 and child products of disintegration (Ra-228, Ra-224) in excess of established NRB and OSPB, and REE, Sc, Fe, Cr, Mn, Al, Ti, Zr, Nb, Ta, Ca, Mg, Na, K, etc

The invention relates to the field of radiochemical technology, namely the processing of water-tail nitric acid solutions resulting from the reprocessing of irradiated nuclear fuel (SNF) and containing technetium

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

Up!