Method for recovering radioactive sludge and bottom deposits
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 the field of environmental protection, and more specifically to the field of radioactive waste management. Most effectively the inventive method can be used in the processing of radioactive silt and sediment accumulated in sprinkling pools or other bodies of water in nuclear power plants (NPP), with the subsequent inclusion of products of processing in a stable solid matrix.
There is a method of curing radioactive silt deposits NPP (RF Patent No. 2106704; G 21 F 9/16; publ. 10.03.98), which consists in mixing sludge, representing a suspension with humidity around 90%, with the crushed powder of granulated blast furnace slag, defending and sharing the resulting solution to decanter and dense sludge, mixed sludge, clay powder and sodium alkali and curing the resulting mixture. The method provides high strength of the cured block and low leaching of radionuclides.
The disadvantages of this method are the presence of the organic phase in the cured block that the long-term storage should lead to a significant decrease of the strength properties of the compound, and a low degree of filling of its radioactive waste, which reduces the efficiency of the method.
Also known is a method of processing radioactive soils (silts), containing organic com is ananti (RF Patent No. 2106705; G 21 F 9/28; publ. 10.03.98), according to which radioactive soil (silt) after firing (temperature 800-1000° (C) with mineralizer and other additives to grind specific surface 2500-4500 cm2/g, shut water and stand before the formation of the monolith. As a result of processing receive water resistant and mechanically strong product, characterized by a small amount.
The disadvantages of this method are the use of ballast additives, leading to the decrease of the degree of filling of the final product of the processing of radioactive waste, and high temperature firing, increasing the intensity and leading to the release of cesium from firing material.
The closest to the technical nature of the claimed method is a method of joint cementing radioactive soils containing organic components, and liquid radioactive waste (LRW) NPP (RF Patent No. 2124243; G 21 F 9/16, 9/32; publ. 27.12.98). In addition to the soil, this method allows you to process and silt ponds.
The essence of this method lies in the fact that recyclable solid radioactive waste (SRW), containing organic components, is heated up to 500-800°S, the product of heating ground to a specific surface area not lower than 1000 cm2/g and mixed with cement and LRW, and the weight ratio between the cement and the product of heating TRO is not less than 0.1, and the weight ratio between the liquid and the mixture of the product of heating and cement is 0.25 to 0.8. Prepared cement paste can stand up to the education suitable for long-term storage of cement stone.
The disadvantages of the known technical solutions are:
- high radiation hazard to the implementation of the method;
- high energy intensity of way;
- low efficiency of the method;
- a large amount of cement stone.
Increased radiation hazard the implementation of the known method is associated with significant entrainment of radioactive caesium by heating the solid to a temperature of 800°C. the firing temperature exceeds the boiling point of cesium (637±10° (C) (Chirkin B.C. Thermophysical properties of materials in nuclear technology. The Handbook. M: Atomizdat, 1968, s). In the cesium vapor must be evaporated in the form of aerosols to go with gas emissions.
High energy consumption known way connected with the necessity of grinding the calcined solid radwaste to the value of specific surface area not lower than 1000 cm2/g and heating TRO to 800°C.
Increased volume of cement and reduced efficiency of the method are defined so that to obtain the final product processing with satisfactory regulated parameters (in accordance with GOST R 51883-2002. Waste radioactive is eontroversy. General technical requirements) should be used when cementing burnt TRO significant amount of cement.
The technical result, which aims invention is:
- reducing the amount of cement in comparison with the volume of the original waste;
- improving radiation safety way;
- reducing the intensity of way;
- increase in the efficiency of the method.
To achieve this result, a method of processing radioactive sludge and bottom sediments containing organic matter of different origin, which consists in heating of solid radioactive waste (SRW), the dispersion of the product heating, cementing and endurance, while pre radioactive silt and sediment is separated into the solid and the liquid filtrate on the filter centrifuge, heating TRO is carried out at a temperature of 500-600°S, product heating is crushed to pieces no larger than 30 mm, the crushed pieces of cement cement mortar, which is a mixture of vysokopronitsaemogo cement with a specific surface area of not less than 8000 cm2/g and a liquid phase at a weight ratio of the liquid phase of the cement 0.6 to 1.4.
In addition, as the liquid phase using water and/or liquid radioactive waste with TDS last not more than 100 g/l
The method requires that the tsya is when filtering radioactive silt and other sediments is their division into non-radioactive effluent and dewatered radioactive residue (solid residue)as a result of heating which the evaporation residue water and burnout of the organic component of the sediment with the formation of the removed non-radioactive volatile components. The burnt part of the sediment after grinding produces a material with a porosity sufficient to cementing its impregnation visokoprohodim cement mortar, and treatment does not increase the volume of sludge. The resulting cement stone is smaller than the original filter cake and its properties satisfy the requirements of GOST R 51883-2002.
The lower temperature range 500°provides evaporation of water from porous and capillary structures (in the temperature range of 120-160° (C) and the spontaneous combustion of any organic impurities (225-250° - for fuel oil and other petroleum products, peat, wood, agricultural waste and spinning production, shale, brown coal and 500° - for anthracite coal (see Transport technology: Textbook for universities W.-D. transport. /Aesome, Idirisov, Wedshare, Weerelt. M.: Transport, 1988, p.29), which provides a translation of them in mineral ash.
The upper temperature range W° To prevent boiling of cesium and its evaporation, because the pressure of the saturated cesium vapor at this temperature is not more than 5,1·104PA (Kalandarishvili A.G. Sources of a working body for thermionic energy converters. M.: Energoatomizdat, 1993, p.55).
If the sizes of the crushed sintered material will be more than 30 mm, to obtain a compound that meets the requirements of GOST R 51883-2002 is impossible.
If the specific surface of cement that is used as a cement binder, will be less than 8000 cm2/g and/or salinity LRW will be more than 100 g/l, it will be impossible to get a cement mortar with high penetrating power, consisting of the above cement and LRW, resulting in cementation, there will be no formation of solid monolithic final product.
When the weight ratio between the liquid phase and the cement is less than 0,6 cement mortar will have a high viscosity that will prevent its penetration into the voids formed by the parts and particles of the crushed sludge, and when the ratio is about 1.4 to get the final product will not meet the requirements of GOST R 51883-2002.
The proposed method of processing radioactive silt and sediment re what occurs in the following way. Radioactive silt filter method centrifigation. The resulting filter nonradioactive filtrate is poured into the storage location of the original sludge and radioactive dewatered sludge is heated at a temperature of 550°C. the Product heating (firing) crushed to a grain size of not more than 30 mm and impregnated with visokoprohodim cement mortar prepared on the basis vysokopronitsaemogo cement with a specific surface area of 9000 cm2/g and a liquid phase, in which water is used or LRW with a salt content of 50 g/L. the Weight ratio between the liquid phase and visokoprohodim cement is 1.0. Impregnated cement mortar product of roasting stand before the formation of the monolith.
Example No. 1 implementation of the proposed method for disposal of radioactive sludge deposits in the cards sprinkling basin NPP. Silt deposits are extracted from the bottom of the sprinkling basin NPP in the amount of 1000 g (833 cm3in a suspension with a humidity of about 80% and organic content 100 g Slurry is subjected to filtration in a centrifuge, resulting in 500 g of non-radioactive liquid filtrate is sent back to sprinkling pool, and 500 g (370 cm3) radioactive sludge. Radioactive solid precipitate is calcined at a temperature of 550°C. the Product of firing, received in the amount of 100 g, drobetz is on a roll crusher to the size of the pieces is not more than 30 mm (porosity of the crushed product is about 50%, bulk volume 129 cm3), after which the impregnated to 98.6 g (64,5 cm3) vysokopronitsaemogo cement weight ratio of LRW with a salt content of 50 g/l (43,8 g) and visokoprohodim cement with a specific surface area of 9000 cm2/g (54,8 g), is equal to 0.8. Impregnated cement mortar product of roasting stand before the formation of the monolith (129 cm3).
Example No. 2 (prototype) differs from example 1 in that the product of firing, received in the amount of 100 g, are ground in a ball mill to a specific surface of 1000 cm2/g, and then mixed with cement (1000 g) in a weight ratio of cement-product of milling 10. To the resulting dry mixture is added to 550 g of LRW with a salt content of 50 g/l (the weight ratio between the liquid and the dry mixture is 0.5) to obtain a cement paste, which is then aged to the formation of the monolith. When the density of the compound 1,776 g/cm3its volume is 929 cm3.
Characteristics of the hardened products are shown in the table. Analysis of the final products showed that unlike the prototype, the use of which leads to an increase of the final product, by the present method achieved a significant reduction in volume of the final product compared to the volumes of the raw sludge suspended solids and the filter cake (Kvof 0.9 and 0.4 for the prototype and Kv =6,46 are 2.87 for the proposed method). The consumption of cement for receiving cement that meets the requirements of GOST R 51883-2002, by the present method, is much less than when using the prototype.
The prototype of the increase in the number of burnt sludge in cement stone reduces its quality. The consequence of this is that when cementing the same quantities of baked mud cement stone prototype inferior quality stone, obtained by the present method, and does not meet the requirements of GOST R 51883-2002.
1. A method of processing radioactive silt and sediment, including the production of dehydrated radioactive sludge and a filtrate, heating the dehydrated radioactive sludge, grinding product heating, cementing and shutter speed, characterized in that the dehydrated radioactive residue get on the filtration centrifuge, the heating is carried out at a temperature of 500-600°S, product heating is crushed to pieces no larger than 30 mm, the crushed pieces of cement visokoprohodim cement mortar, which is a mixture of cement with a specific surface area of not less than 8000 cm2/g and a liquid phase at a weight ratio of the liquid phase of the cement 0.6 to 1.4.
2. The method according to claim 1, characterized in that the quality of the liquid phase using water and/or liquid radioactive waste with TDS last not more than 100 g/l
SUBSTANCE: method involves use of alkali solutions containing excess of oxidant, namely alkali metal metaperiodates, at temperature 70-80є.
EFFECT: enabled dissolution of alloy.
FIELD: decontaminating metal wastes by way of their remelting.]
SUBSTANCE: proposed method includes delivery of metal to be decontaminated to water-cooled ingot-forming equipment and decontamination of melt using refining slag. Refining slag in the form of melt is first to be fed to ingot-forming equipment. Then pre-melted radioactive metal wastes are fed at speed affording maintenance of permanent level of molten refining slag within current-conducting section of ingot-forming equipment at which metal ingot decontaminated from radionuclides in the course of remelting can be drawn out.
EFFECT: enhanced economic efficiency of method.
3 cl, 1 dwg
FIELD: decontamination engineering.
SUBSTANCE: proposed pump has housing, pulse line, inlet ball-and-socket valve with ball lift limiter, delivery pipeline with outlet ball-and-socket valve, and control system. Housing communicates with bottom nozzles through pipe and bottom-nozzles chamber that accommodates shaft provided with flap. Shaft is coupled through movable bearing assembly, gear wheel, and toothed rack with turn and immersion depth control actuator of bottom nozzles. Bottom end of inlet ball-and-socket valve seat has slots and mounts in addition spring with movable perforated rack. In addition housing may accommodate top pipe for its communication through ball-and-socket check valve with washing head that has nozzle and pipe union. Stop is mounted in bottom end of pipe union coaxially with respect to hole in check-valve ball lift limiter. Top part of washing head is joined with aid of actuating shaft through movable bearing assembly, gear wheel, and toothed rack with turn and angle-of-tilt control actuators of top nozzle.
EFFECT: enhanced reliability and safety in operation.
8 cl, 6 dwg
FIELD: decontamination engineering.
SUBSTANCE: proposed device incorporates provision for admission to inner space of container through hole. This facility is, essentially, vehicle moved by drive. Vehicle traveling gear is free to move from pulled-in quiescent position to working position having large track width (B).
EFFECT: enhanced reliability and safety in operation.
12 cl, 3 dwg
FIELD: chemical technology; deactivation and decontamination of radioactive industrial products and/or wastes.
SUBSTANCE: proposed method designed for deactivation and decontamination of radioactive industrial products and/or production wastes incorporating Th-232 and its daughter decay products (Ra-228, Ra-224), as well as rare-earth elements, Fe, Cr, Mn, Al, Ti, Zr, Nb, Ta, Ca, Mg, Na, K, and the like and that ensures high degree of coprecipitation of natural radionuclides of filtrates, confining of radioactive metals, and their conversion to environmentally safe form (non-dusting water-insoluble solid state) includes dissolution of wastes, their treatment 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 that stoichiometrically required for precipitation of total content of metal oxyhydrate; then pulp is filtered and barium chloride is injected in filtrate in the amount of 0.4 - 1.8 kg of BaCl2 per 1 kg of CaCl2 contained in source solution or in pulp and pre-dissolved in sulfuric acid of chlorine compressors spent 5-20 times in the amount of 0.5 - 2.5 kg of H2SO4 per 1 kg of BaCl2. Then lime milk is added up to pH = 11 - 12 and acid chloride wash effluents of equipment and production floors are alternately introduced in sulfate pulp formed in the process at pulp-to-effluents ratio of 1 : (2-3) to pH = 6.5 - 8.5. Filtrate pulp produced in this way is filtered, decontaminated solution is discharged to sewerage system, sediment of barium and calcium sulfates and iron oxysulfate are mixed up with oxyhydrate sediment formed in source pulp neutralization, inert filler and 0.5 - 2 parts by weight of calcium sulfate are introduced in pasty mixture while continuously stirring them. Compound obtained in the process is placed in molds, held therein at temperature of 20 - 50 oC for 12 - 36 h, and compacted in blocks whose surfaces are treated with water-repelling material.
EFFECT: reduced radioactivity of filtrates upon separation of radioactive cakes.
8 cl, 1 dwg, 1 ex
FIELD: rare, dispersed and radioactive metal metallurgy, in particular hydrometallurgy.
SUBSTANCE: invention relates to method for reprocessing of polymetal, multicomponent, thorium-containing radwastes, formed when reprocessing of various mineral, containing rare-earth elements, Nb, Ta, To, V, Zr, Hf, W, U, etc. Method includes treatment of solution and/or slurry with alkaline agent; introducing of sulfate-containing inorganic compound solution and barium chloride; treatment of obtained hydrate-sulfate slurry with iron chloride-containing solution, and separation of radioactive precipitate from solution by filtration. As alkali agent magnesia milk containing 50-200 g/dm2 of MgO is used; treatment is carried out up to pH 8-10; sodium sulfate in amount of 6-9 g Na2SO4/dm2 is introduced as solution of sulfate-containing inorganic compound; barium chloride solution is introduced in slurry in amount of 1.5-3 g BaCl2/dm2. Hydrate-sulfate slurry is treated with solution and/or slurry containing 0.8-16 Fe3+/dm2 (as referred to startingsolution) of iron chloride, followed by treatment with high molecular flocculating agent and holding without agitation for 0.5-2 h. Radioactive precipitate is separated from mother liquor, washed with water in volume ratio of 0.5-2:1; then washed with sodium chloride-containing solution and/or slurry in volume ratio of 0.5-2:1; radioactive precipitate is removed from filter and mixed with mineral oxides in amount of 0.5-0.8 kg MgO to 1 kg of precipitate. Formed pasty composition is fed in forms and/or lingots and presses with simultaneous heating up to 80-1200C.
EFFECT: filtrate with reduced radioactivity due to increased codeposition coefficient of natural Th-232-group radioactive nuclide, in particular Ra-224 and Ra-228, with radioactive precipitates.
10 cl, 1 ex
FIELD: decontamination engineering.
SUBSTANCE: proposed method includes treatment of circuit coolant with acid solutions and washing. In the process treatment with acid solutions is made by chemical loosening for 2-10 h. Dynamic loosening is effected prior to chemical loosening and then coolant temperature is periodically raised in reactor core to 150-200 °C.
EFFECT: reduced time and enhanced effectiveness of decontamination treatment process.
2 cl, 5 dwg, 1 tbl
FIELD: nuclear power engineering.
SUBSTANCE: compaction involves cutting members into fragments using electroerosive destruction of member wall by pulse spark-arch discharges emerging between member and electrode. In addition, high-temperature treatment in oxidizing medium, in particular vapor formed, is carried out. Cutting and heat treatment are accomplished in water.
EFFECT: simplified procedure and increased safety.
FIELD: decontaminating liquid radioactive wastes.
SUBSTANCE: proposed method for recovering and solidifying radioactive coagulation pulps includes their mixing with liquid nitric acid solutions for dissolving, fluxing with glass-forming elements, and vitrifying them.
EFFECT: enhanced reliability of radioactive waste immobilization.
3 cl, 6 ex
FIELD: electrical engineering; water-cooled and induction heated crucibles.
SUBSTANCE: proposed crucible that can be used for melting minerals, mineral-like materials, ceramic materials, glasses and other glass-like materials characterized in high melting point and also for embedding radioactive and nonradioactive wastes into glass and/or glass-like materials compatible with them has water-cooled cover, water-cooled drain valve, water-cooled drain assembly, water-cooled drain assembly cover, charging assembly, internal feeder, gas outlet pipe, guide bar, upper header, cylindrical case, lower header, flat water-cooled bottom , and inductor. Crucible incorporates provision for regulating melt drainage capacity and for maintaining constant chemical composition of drained melt compared with that of material fed to water-cooled crucible.
EFFECT: reduced heat loss and power requirement, simplified design, reduced accident probability, enhanced quality of drained melt.
1 cl, 8 dwg
FIELD: immobilizing radioactive wastes.
SUBSTANCE: proposed method includes filling melting crucible with charge and melting the latter. Molten charge is drained through melting crucible die hole to container wherein glass is additionally melted. Charge is melted, charge melt is poured into container, and glass is additionally melted in common heating zone. Device for manufacturing high-level glass filled container has melting crucible incorporating die hole and bottom part, and container. Melting crucible is disposed above container on spacer ring in common heating zone. Spacer ring height is greater than total height of container and die hole. Ratio of die hole diameter to its height is chosen to ensure formation of thermal plug therein from solidified glass produced in previous melting process.
EFFECT: enhanced reliability of immobilizing radioactive wastes.
3 cl, 1 dwg, 2 ex
FIELD: immobilization of radioactive wastes.
SUBSTANCE: proposed method includes precipitation of tribasic phosphates of transplutonium and rare-earth elements from nitric acid solution, their dehydration and compacting. During precipitation stage aqueous polyacrylamide solution is added in the amount of minimum 1 mass percent of tribasic phosphates, and molding powder is produced in thin-film rotary concentrator.
EFFECT: enhanced reliability of radioactive waste immobilization.
1 cl, 1 tbl, 3 ex
FIELD: radioactive waste disposal.
SUBSTANCE: immobilization of liquid radioactive waste comprises preparing solution of radioactive waste with liquid glass-forming additive and organic reducer solution followed by introducing the two solutions into melter to melt glass. Liquid boron-containing glass-forming additive is prepared from solid sodium tetraborate Na2B4O7*xH2O (x=0-10) and boric acid or from their mixture and solution of organic reducer, in particular 1,20ethanediol or 1,2,3-propanetriol.
EFFECT: increased radioactive waste fixation reliability.
6 cl, 2 tbl, 3 ex