Sorbent for entrapping radioactive iodine from gas phase
FIELD: nuclear-industry radiochemical enterprises for recovering and immobilizing gaseous radioactive wastes.
SUBSTANCE: sorbent used for entrapping radioactive iodine from gas-aerosol stream running from irradiated nuclear fuel cutting-and-dissolving unit has porous base impregnated with nitric acid silver salt (AgNO3); used as sorbent base is porous silicon carbide of 30 to 60% porosity.
EFFECT: enhanced corrosion and mechanical resistance of material in chemically active media.
1 cl, 2 dwg, 3 tbl
The invention relates to the field of processing and immobilization of gaseous radioactive waste radiochemical enterprises of the nuclear industry, in particular to the field trapping of iodine (in particular, iodine-129) from the aerosol stream with a host of cutting-dissolution of irradiated nuclear fuel.
Processes for processing of irradiated nuclear fuel (SNF) power reactors and transport facilities are inextricably linked with cleaning aerosol waste from long-lived iodine-129 (T1/2=1,57·107years). As a sorbent for his capture uses porous material impregnated nitrate of silver.
Analogue of the invention can serve as a silver-containing sorbent Selexid" in accordance with the technical conditions "Siloxy-sorbent for the capture of radioactive iodine from the gas environments". LCVS 94.373.00.000, Pine forest, THREAD them. A.P. Alexandrov, 1995.
Sorbent Selexid" does not meet the requirements of iodine purification in the processing of spent fuel, since it has a small size pellets (2 mm) and low specific surface area, which contributes to a significant increase in aerodynamic resistance of the gas-cleaning apparatus and reflected on the absorptive capacity of the sorbent. Known sorbent to adsorb iodine [S.I. Smooth, I.E. Pyatin, I.A. Istomin. Capture Yoda in the processing of irradiated nuclear fuel power plants. Article in the journal "Nuclear energy", I. 92, issue 6, June 2002], which is taken as a prototype of the invention, where as the basis saturated salt of silver (AgNO3), is used alumina grade a in the form of white granules of cylindrical shape with a base diameter of 3 mm to 4 mm and a height of 10 mm to 15 mm according to GOST 8136-85 "active aluminium Oxide. Technical conditions".
The disadvantages of aluminum oxide are:
- prone to mechanical failure and attrition in the processes of exploitation and regeneration;
low chemical stability in alkaline and acidic media;
- dusting in the process of repackaging.
The objective of the invention is to increase the mechanical strength in operational processes, regeneration and repackaging, and chemical resistance of the sorbent in acidic and alkaline environments while maintaining the basic performance requirements of the sorbent and its basis (effective saturation of the base impregnator, high dynamic capacity for iodine, the possibility of regeneration of the sorbent and silver extraction for reuse). It is solved through the application as a sorbent for the capture of iodine from the gas phase porous silicon carbide, impregnated with a salt of silver nitrate (AgNO3).
Silicon carbide is a material, which is a chemical compound of silicon is carbon (SiC); the Mohs hardness of 9.1; microhardness 3300-3600 kgf/mm2. Get it in electric furnaces of the resistance sililirovanie carbon particles pairs of silicic acid. Raw materials are materials that are rich in silica: vein quartz, quartz sand and quartzite, containing not less than 99,0-99,5%, SiO2and the carbonaceous material is petroleum coke. Silicon carbide is a promising material modern instrumentation due to the high radiation resistance and thermal stability.
The proposed material different from aluminum oxide to the fact that silicon carbide has a high corrosion resistance in aggressive environments, so it does not react with inorganic acids and alkalis, even at the temperature of boiling. Laboratory studies have shown that the corrosion of silicon carbide in groundwater with pH˜8 at a temperature of 170°0.15 mm per year. The mechanical strength of the silicon carbide substantially exceeds the strength of aluminum oxide, accordingly, this material is more resistant in the process of repackaging and transportation.
The use of porous silicon carbide for the manufacture of iodine sorbent may be in the form of granules of various shapes, and in the form of a filter cartridge (AF) in the case of small recycled modular filter capable OS is out cleaning the gas phase from iodine.
For the manufacture of iodine sorbent porous silicon carbide integriruetsa salt of silver nitrate (AgNO3). For this porous sorbent substrate (silicon carbide with a porosity of from 30% to 60% in the form of granules or in the form of a filter cartridge) is impregnated with a solution of silver salts with the desired concentration of silver and dried at a temperature of from 100 to 150°C. the Operation of impregnation and drying is repeated until the complete absorption of the solution.
After saturation of the sorbent radioactive iodine from the gas phase may conduct its regeneration with the aim of re-use. If necessary, you can extract the silver from the proposed sorbent for re-use.
For experiments to study the possibility of manufacturing a sorbent based on porous silicon carbide (porosity 30%) was used, the sample material (hereinafter OP - filter cartridge) in the form of a hollow cylinder with a base diameter of 70 mm, a height of 105 mm and a wall thickness of 5 mm. the Original mass of the OP was 184574 mg Impregnation OP nitrate of silver was carried out by nitrate solution with a concentration of silver of 2.6 g/l and nitric acid 5 mol/l, for several stages, each of which represents a process of uniform wetting of the filter and subsequent drying. Thus, achieved full of pohlad is the solution.
The OP was weighed before feeding and after him. The results of saturation AF silver presented in figure 1.
After the sixth stage, there was a complete absorption of the solution containing 2.2 g of silver, and the concentration of silver in the filter cartridge was 11.9 mg/g
Experiments on the trapping of iodine OP conducted in a laboratory setup (Figure 2).
Figure 2: 1 - tight tank, 2 - heat resistant glass, 3 - iodine-127, 4 - oven, 5 - MP, 6 - bubbler, 7 - NaOH solution.
At the bottom of the heat-resistant glass 2 was poured by the addition of iodine-127 in excess of the stoichiometric value relative to the formation of silver iodide. FP 5 was placed on a metal grid attached above the level of the filling of iodine crystals. A glass of FP was installed in the hermetic container 1, provided with a socket for removal of the gas phase in the bubbler 6, filled with a solution of sodium hydroxide with a concentration of 2 mol/l, the Saturation of the sorbent with iodine was carried out at a temperature of 200°in a shaft furnace 4. The beginning and the end of the evaporation process of iodine was controlled by changing the colour of the solution in the bubbler and inlet ducts.
The captured amount of iodine was determined by weighing the OP before and after saturation. The results of the experiment are presented in table 1.
Rez is ltati saturation OP iodine-127
|The mass of the filter cartridge with AgNO3mg||The mass of the filter cartridge after saturation mg||The mass of iodine mg||The concentration of iodine, mg/g|
The concentration of iodine-127 in the absorber was 18.1 mg/g, which is greater than the number of iodine in the form of silver iodide formed in accordance with the stoichiometry of the reaction
This can be explained by the adsorption of molecular iodine on the surface of pores of AF.
The regeneration of the OP spent alkaline solution of hydrazine-nitrate with a concentration of alkali 30 g/l and hydrazine 15 g/L. treatment of the sorbent solution was 30 minutes at a temperature of 80°C. After carrying out the regeneration of the filter cartridge was washed with hot distilled water until pH=5-7, dried and weighed. From solutions formed during regeneration of the PF (regenerated and rinsed) iodine is concentrated in the form of copper iodide. Results regeneration of AF and deposition of iodine in iodide of copper are shown in table 2.
Removing iodine-127 of the silver-containing filter cartridge
|The mass of the filter-holder mg||The mass of th is Yes in the sediment, mg||The efficiency of extraction of iodine, %|
|before regeneration||after regeneration|
With the aim of obtaining silver from the spent sorbent on the basis of porous silicon carbide OP after removing from it the iodine was treated with nitric acid with a concentration of 5 mol/l for 30 minutes at a temperature of 80°With, then the OP was dried and weighed. Uterine and washing solutions were analyzed for their content of silver (the original content of silver in the OP - 2.2 g) in two ways. In the first case, the mass concentration of silver in solution was determined by x-ray fluorescence energy dispersive analyzer ERA-03. The second silver besieged from solution in the form of chloride. The results of the analyses of the solutions presented in table 3.
Extraction of silver from the regenerated filter cartridge
|The mass of the filter-holder mg||Weight of silver mg||The efficiency of extraction of silver, %|
|To retrieve||after extraction, the||in solution||in draft|
As can be seen from the data presented in table 3, the efficiency of extraction of silver from the filter cartridge was (99,4±0,3)%, which is higher than the degree of extraction of silver from the sorbent on the basis of aluminum oxide, where she is on the level (97,4±0,6)%.
Distinctive features of the proposed material are significantly higher corrosion and mechanical resistance in aggressive environments. This can significantly prolong the life of iodine sorbent, to increase efficiency in the use of expensive silver.
Sorbent for the capture of radioactive iodine, consisting of a porous base, impregnated with salt of silver nitrate (AgNO3), characterized in that as the basis of sorbent used porous silicon carbide with a porosity of from 30%to 60%.
FIELD: recovery of radioactive wastes.
SUBSTANCE: proposed method for matrix immobilization of industrial wastes includes preparation of source solution of industrial wastes and impregnation of ceramic matrix with this solution followed by roasting this matrix; source liquid radioactive wastes used for the purpose are first treated with promoter crystallization solution doped with oxide-forming admixtures whereupon radioactive wastes are introduced in ceramic matrix and roasted using microwave energy at temperature of 900 - 1 000 °C. Such procedure provides for recovering great amount of radioactive wastes included and chemically bonded in ceramic matrix which makes it possible to reduce leaching and to enhance matrix strength and life.
EFFECT: facilitated procedure, reduced cost, enhanced quality of radioactive waste immobilization and environmental friendliness.
FIELD: environment control.
SUBSTANCE: dry or wet, granulated, powdered, or milled spent ion-exchange resins are included in matrix in N- or H-form. Used as matrix base is mixture of blast-furnace slag milled to fraction of 0.075 mm and chrysotile-asbestos in the amount of 5 mass percent. Sodium hydroxide solution is added to mixture in the amount of 100 -150 g/l.
EFFECT: enhanced degree of filling the compound and enhanced reliability of further storage.
1 cl, 1 dwg, 2 tbl, 8 ex
FIELD: recovery of liquid radioactive wastes.
SUBSTANCE: proposed method for treatment of sulfuric ammonium radioactive solutions includes electrotype coagulation treatment involving passage of radioactive solutions through vibroliquefied iron-coke galvanic pair. Radioactive solution pH is corrected and radionuclide-containing deposit separated. In the process pH of radioactive solutions is corrected in two stages with deposit being separated after each stage. First-stage deposit is mixed up with montmorillonite clay and granules are molded. Then they are dried out and agglomerated to produce glass-ceramics wherein radionuclide-containing deposits are immobilized. Filtrate is treated with foam layer to blow off ammonia and then passed through natural ion-exchanger. Method for immobilizing radionuclide-containing deposit in glass-ceramics includes molding of granules of radionuclide-containing deposit mixed up with montmorillonite clay. Granules are molded in two stages. Core of granule of radionuclide-containing deposit mixed up with montmorillonite clay is produced during first stage and covered with shell of glass-ceramics based on homogeneous mixture of montmorillonite clay and quartz sand, during second stage, quartz sand content of mixture being 10 - 30 mass percent.
EFFECT: enhanced reliability of radioactive waste immobilization.
6 cl, 2 dwg, 1 tbl, 3 ex
FIELD: recovery of liquid radioactive wastes.
SUBSTANCE: proposed method involves vitrification of liquid radioactive wastes using solution of phosphate or boron phosphate matrix in glass-making furnace by feeding liquid radioactive wastes and fluxing reagent solutions of phosphoric acid, sodium nitrate, sodium hydroxide, solution of boron acid or sodium tetraborate in sodium hydroxide to molten glass for melting glass mass. In the process fluxing reagent in the form of two solutions is dosed to glass-making furnace separately from radioactive wastes without their premixing with wastes.
EFFECT: optimized characteristics of melt.
2 cl, 2 ex
FIELD: nuclear engineering; methods of processing of radioactive waste.
SUBSTANCE: the invention is pertaining to the field of processing of radioactive waste. The method of melt-shutting of dangerous materials and-or products provides for arrangement of the melt-shutting material, the subjected to the melt-shutting dangerous materials and-or products in a metal container and their heating by UHF-energy. In the capacity of the melt-shutting material use a loose radio-transparent material, in which dip the subjected to the melt-shutting dangerous materials and-or products till their complete surrounding by the radio-transparent loose material. Then using UHF-energy heat up the subjected to the melt-shutting dangerous materials and-or products, at least, to the melting point of the loose radio-transparent material contacting with the subjected to the melt-shutting dangerous materials and-or products for formation around the subjected to the melt-shutting dangerous materials and-or products of a monolithic shell. The technical result of the invention is a reliable fixation of the hazardous substances.
EFFECT: the invention ensures a reliable fixation of the hazardous substances.
3 cl, 4 dwg
FIELD: processing of liquid A-wastes.
SUBSTANCE: the invention is pertaining to the field of processing of liquid A-wastes. The invention presents: a glass-forming boron phosphate compound for immobilization of the aluminum-containing liquid highly radioactive wastes by their vitrification. The compound contains: sodium oxide, aluminum oxide, boron oxide, phosphor oxide and natural impurities of oxides of multivalent chemical elements. At that it in addition contains lithium oxide at the following ratio of components (in mass %): Na2O - 22.0-26.0;Al2O3 - 13.0-28.0;B2O3 - 3.0-6.0;P2O5 - 38.0-55.0;Li2O - 0.5-1.0; the oxides of multivalent chemical elements - the rest. Advantages of the invention consist in production of a qualitative homogeneous glass.
EFFECT: the invention ensures production of a qualitative homogeneous glass.
FIELD: immobilization of liquid radioactive wastes.
SUBSTANCE: method for treatment of wastes formed by aqueous solution containing 3 - 10 mole/l of caustic soda includes introduction of powdered metakaolin in aqueous solution in the amount sufficient to obtain slurry capable of solidifying and forming crystal phase of zeolite A type. Then slurry is placed in mold and held therein for solidification to obtain zeolite A based solid molded product. After that molded product is dried out, and zeolite A phase is converted to nepheline type phase upon its heat treatment at 1000 to 1500 °C.
EFFECT: enhanced reliability of isolating radioactive wastes.
10 cl, 1 dwg, 2 ex
FIELD: reprocessing of radioactive wastes.
SUBSTANCE: the invention is pertaining to the methods of reprocessing of radioactive perlite suspensions. The method of reprocessing of radioactive perlite suspensions provides for their commixing with radioactive solutions, treatment of the glass-forming devices, stirring action for a uniform distribution of a solid phase and a vitrifying. At that the radioactive perlite suspensions containing in their compounds fluorine-ions and more than 0.05 g/l of alpha-active radionuclides are stirred with the homogeneous and-or heterogeneous nuclear waste, which contains sodium nitrite or sodium hydroxide, or aluminum nitrate or aluminum hydroxide, or their mixture. Before vitrification the perlite suspensions are treated with calcium compounds. The technical result of the invention is upgrade of the quality of the waste utilization.
EFFECT: the invention ensures upgrade of the quality of the waste utilization.
4 cl, 5 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: recovery of biologically hazardous wastes.
SUBSTANCE: proposed method includes introduction of toxic and radioactive material neutralization wastes into reactor together with charge, heating them with aid of superhigh-frequency energy, and hardening; toxic material neutralization wastes are decomposed in the course of heating. Reactor charge composed of river sand and cullet doped with boron acid and red lead in 5 : 1 proportion is used for hardening wastes by vitrification.
EFFECT: enhanced quality of waste recovery.
1 cl, 1 dwg, 1 tbl
FIELD: technology of handling of the liquid nuclear wastes of the nuclear fuel and power cycle; methods of reprocessing of the liquid nuclear wastes.
SUBSTANCE: the invention is pertaining to the procedure of the liquid nuclear wastes handling of the nuclear fuel and power cycle and may be used during reprocessing of the liquid nuclear wastes (LNW). The method includes the preliminary concentration, ozonization, microfiltration of the vat residue with fractionation of the permeate and the concentrate and the ion-selective purification of the permeate using the ion-selective a sorbent. At that the microfiltration is conducted at least in two stages: the permeate of each previous stage of the microfiltration is directed to the microfiltration as the source solution for the subsequent stage of the microfiltration, and at the final stage of the permeate from the microfiltration is sent to the utilization. The concentrate produced at each next stage of the microfiltration is mixed with the source solution of the previous stage of the microfiltration. The concentrate produced at the first stage of the microfiltration is directed to the conditioning and dumping. The ion-selective sorbent is added in the permeate of the previous stage of the microfiltration before the final stage of the microfiltration. The invention ensures: reduction of the volume of the liquid nuclear wastes due to the deep purification of the LNW with the high saline share from the radionuclides and extraction of the radionuclides in the compact form of the sparingly soluble compounds at the corresponding increase of the factor of purification of the salts extracted at the stage of the vat residue treatment; reduction and optimization of the consumption of the permeate and concentrate interacting with the source solution as well as produced at the further stages.
EFFECT: the invention ensures: reduction of the volume of the liquid nuclear wastes; the deep purification of the LNW with the high saline share from the radionuclides and extraction of the radionuclides in the compact form of the sparingly soluble compounds at the corresponding increase of the factor of purification of the salts extracted at the stage of the vat resudue treatment; reduction and optimization of the consumption of the permeate and concentrate interacting with the source solution ands produced at the further stages.
FIELD: environment control including environment protection in atomic industry.
SUBSTANCE: proposed method for decontaminating radioactive aqueous solutions from radionuclides includes at least one contact of solution with complexing sorbent that has solid-medium immobilized active polymeric layer condensed with chelates. Medium is chosen from following group: activated cellulose; synthetic copolymers with divinyl benzene, activated chloromethyl or hydroxymethyl, or chlorosulfonic groups. Active sorbing layer has ethylenediamine or diethylene tridiamine, or tetraethylene pentaamine, or polyethylene polyamine with copolymers; chelates are chosen from group incorporating carboxyl-containing chelates, phosphonic-group chelates, and hydroxyl-containing chelates. Proposed method enables extraction of radionuclides both in ionic and colloidal condition from solutions doped with highly concentrated impurities; sorbent used for the purpose retains its sorbing properties upon repeated regenerations and is capable of decontaminating solutions both in dynamic and static modes with different pH of solutions being decontaminated.
EFFECT: enlarged functional capabilities.
11 cl, 3 tbl
FIELD: methods of the sorption decontamination of waters from the radioactive impurities.
SUBSTANCE: the invention is pertaining to the method of the sorption decontamination of waters from the radioactive impurities. The method of decontamination of the water from radiostrontium includes the treatment of the water with a sorbent based on the burned bauxite ore. At that the ore is burned together with calcium -magnesium lime - CaCO3·MgCO3 and sodium soda salt - Na2CO3 at the temperature of no less than 1200°C and flushed with the water to remove the solvable sodium compounds. It is preferable, that the mixture of the bauxite ore is subjected to burning with calcium-magnesium lime and sodium soda salt in the mass ratio of 1 : 0.55-0.60 : 0.055-0.060. The method ensures an increased effectiveness of removal of the radiostrontium at usage of the initial bauxite ore without lowering of effectiveness of the water decontamination from radiocesium, and also allows to reduce considerably the amount of the spent sorbents, which are subjected to disposal.
EFFECT: the invention ensures an increased effectiveness of removal of radiostrontium from the water at usage of the initial bauxite ore without lowering of effectiveness of the water decontamination from radiocesium, allows to reduce considerably the amount of the spent sorbents, which are subjected to disposal.
2 cl, 10 ex
SUBSTANCE: invention relates to application of pectin solution as detergent for skin and hair in radioactive and environmentally hazardous regions. Pectin containing in solution due to chelating action bonds to heavy metal ions to form stable compounds (micelles) having very large size and prevents transferring thereof trough transdermal barrier. Formed micells may be easily removed from human body.
EFFECT: new detergent for skin and hair useful in radioactive and environmentally hazardous regions.
2 ex, 5 tbl
FIELD: processing of liquid radioactive wastes.
SUBSTANCE: the proposed method for cleaning of liquid radioactive wastes includes their processing with absorber-sorbent. Ash wastes are used as the absorber-sorbent. The ash wastes are introduced into liquid radioactive wastes in the stage of their neutralization at pH = 0.5-2 at the flow rate, ensuring the ratio of S : L = 1 : (15-50). Then the obtained suspension is separated into the liquid and solid phases.
EFFECT: increased degree of cleaning from radionuclides and enhanced separation rate of produced pulps.
4 cl, 1 tbl, 11 ex
FIELD: reactive sorbents.
SUBSTANCE: liquid is conditioned at pH 4-6 and then is brought into contact with chelating ion-exchange resin composed of grafted polyazocycloalkanes at temperature above or equal to 60°C.
EFFECT: enhanced metal removal efficiency.
35 ex, 9 dwg, 9 tbl, 5 ex
FIELD: recovery of liquid radioactive wastes.
SUBSTANCE: proposed method includes treatment of radioactive wastes by heavy-metal ferrocyanide and separation of sediment obtained. In the process liquid radioactive wastes are treated with heavy-metal ferrocyanide formed upon introduction of potassium ferrocyanide and bivalent nickel and/or copper and iron salts taken in amount abundant with respect to stoichiometric quantity. Clarified solution is treated with oxidant upon separation of sediment and filtered through catalytic material containing manganese dioxide. Then filtered-off solution is passed through highly acid cationite in Na-form and highly basic anionite in Cl-form.
EFFECT: enhanced decontamination quality.
5 cl, 1 tbl, 7 ex
FIELD: recovery of liquid radioactive wastes.
SUBSTANCE: proposed method includes bringing liquid radioactive wastes in contact with matrix saturated with selective ion-exchange material (solid extracting agent). Glass-crystal material with open porous structure is used as matrix for the purpose. Matrix material is produced from hollow glass-crystal cene spheres formed from mineral particles of volatile ash produced as result of black coal combustion and saturated with selective ion-exchange material.
EFFECT: facilitated procedure of radionuclide extraction.
5 cl, 1 tbl, 5 ex
FIELD: environment protection from radionuclides.
SUBSTANCE: proposed method for producing chemical sorbent to absorb nuclear fuel fission products (radionuclides of iodine, ruthenium, and their volatiles) includes impregnation of activated carbon in triethylene diamine followed by its drying at temperature of 110-130 °C. For the purpose use is made of activated carbon produced from raw bituminous coal having micropore volume vmp = 0.28-0.33 cm3/g and total pore volume vΣ = 0.85-1.0 cm3/g. Activated carbon is impregnated until triethylene diamine content is reduced to 1-2% of base mass. Then chemical sorbent obtained in the process is dried out and sifted.
EFFECT: enhanced quality of chemical sorbent obtained.
1 cl, 1 tbl, 9 ex