Cement-polymeric composition for preserving radioactive wastes of medium reactivity
FIELD: nuclear engineering; preservation of dry, wet, and liquid radioactive wastes.
SUBSTANCE: proposed composition has resin portion of cold-cured compound ATOMIK and filler. Resin portion ingredients are as follows, parts by weight: epoxy oligomer, 100;, curing agent (aromatic amines), 38-50 furfural, 9-11. Used as filler is composition incorporating following ingredients, parts by weight: cement, 50-100; marshalite, 50-100 or bentonite, 50-100, or when they are jointly used: marshalite, 90-100 and bentonite, 90-100. Such composition provides for desired radiation resistance in absence of leaching of alpha-, beta-, and gamma-active radionuclides from preserved specimens of reactor graphite.
EFFECT: enhanced radiation stability of preserved specimens free from pits and voids, and adequate lifetime of preserved wastes; ability of their depreservation.
1 cl, 2 dwg, 2 tbl, 2 ex
The invention relates to polymeric compositions of cold hardening used in nuclear engineering for the preservation of intermediate level radioactive waste (RW), in dry, wet and liquid state.
Known polymer composition is cold-curing, resistant to radiation, which includes a resin part (100 parts by weight) (patent RF №2239643, IPC 08 L 63/00, 2001). The resin portion further comprises an aliphatic epoxy resin and a low volatile ester of phthalic acid and aliphatic alcohol with a mass ratio of 90:5:5 to 30:25:45, and as the amine hardener is a product of the interaction of aromatic di - or polyamine (K), salicylic acid (L), benzyl alcohol (M) and furfural (N) in the ratio K:L:M:N 88:2:8:2 44:10:30:16. If this composition contains, in parts by weight:
|The resin part||100|
The disadvantage of this polymer composition is that it is not possible at room temperature monolithic pour contained in the vessel shattered fragments of graphite reactor of the rings. When sawing these samples were found shells and cavities, and to reduce the viscosity of the composition by heating below the consequently, as this increases the rate of the curing reaction and shortens the time of her life. When the curing of the composition may occur in the tank mixing of the resin part with hardener. In addition, this method of preservation is to provide a monolithic casting RAO composition it is necessary that the temperature of the tank waste was not lower than the temperature of the composition, otherwise the heated compound in contact with preserved material has cooled and the viscosity will decrease. In real conditions subject to conservation materials can be in the wet state. In this case, as well as for liquid radwaste using this resin composition is not suitable.
Also known polymer composition to isolate the solid radioactive waste (patent RU №2251561, 10.05.2005,). The composition comprises, in parts by weight: 100 epoxy Dianova resin as a resin part, 40-95 polyamide resin as a hardener, 25-45 furfural, 0-380 filler. In use as a filler bentonite, cement and Marsala.
Its disadvantages include insufficient radiation resistance evolution, complexity and duration of the technical process of making composition (the necessity of heating the mixture, the daily exposure of the mixture), lack of life of the mixture. When using such a composition when the room is Noah temperature in preserved samples were observed unfilled cavity, shell. The composition was aterials, not able to impregnate conserved graphite blocks, which led to poor sealing and consequently not sufficiently reduced the leaching of radionuclides.
The technical objective of the claimed invention to provide a resin composition for a simple and reliable conservation intermediate level radioactive waste in dry, wet and liquid state, in the absence of a conserved RAO shells and cavities.
The technical result of the present invention is the increased radiation resistance, in the absence of leaching of alpha-, beta - and gamma-active radionuclides from preserved samples of reactor graphite, in the absence of shells and cavities in canned RAO with sufficient lifetime and raskonservirovana.
To achieve the specified result of the proposed cement-polymer composition for conservation intermediate level radioactive wastes containing resin part of the compound of cold hardening "atomic"consisting of epoxy oligomer, hardener, and aromatic amines, and furfural, and fillers cement Marsala and/or bentonite, in the following ratio of components, parts by weight:
|epoxy oligomer||100||the above curing agent||38-50|
|Marsala or bentonite||50-100|
As the resin and hardener were used compound of cold hardening "atomic", manufactured by CJSC "ENPC APICAL" beyond 2257-998-18826195-01. It contains as a resin part, for example, Dianov a type resin ED-20, aliphatic epoxy resin deg-1 (100 parts by weight)as a hardener, such as a hardener based on polyamide resin L-19 (38-50 parts by weight), (see figure 1). Cement was used stamps 500.
As RAO used samples of reactor grade graphite GR-280, manufactured from a block of graphite columns of the 3rd unit of Leningrad NPP after 18 years of operation. In the study of radiation gas reactor graphite was discovered that it, in comparison with the source material that contains significantly more gaseous products, which include radioactive gases. Consequently, when developing the conservation of graphite blocks and rings, extracted from the reactor, it was suggested not to use the method in chumney impregnation with preservative, in order not to pollute the environment with radioactive gases, takuminokami from the pores of reactor graphite, and to carry out the impregnation of a low viscosity compositions.
For the manufacture of cement-polymer composition (CPC) in the capacity of the injected resin part of the compound of cold hardening "atomic", add the furfural 11 parts by weight, the filler (cement, Marsala or cement, bentonite), stirred for getting ready CTC. For execution of works on preservation of the crushed fragments of the reactor graphite blocks, rings, graphite spillages and other radioactive wastes generated during the dismantling of the reactor (both dry and wet), as well as sorbents for the purification of liquids containing radionuclides, these materials are loaded into metal barrels for conservation and fill them ready CTC. After curing the CPC formed monolithic blocks with high radiation resistance of the binder and the absence of leaching of radionuclides.
Table 1 presents examples of CTC (compounds No. 1, 2), and table 2 summarizes the main properties of the CTC.
|Cement-polymer composition for conservation intermediate level radioactive waste|
|Name of the component||Composition 1 (for dry RAO), parts by weight||Composition 2 (for wet and liquid waste), parts by weight|
|The resin part of the compound "atomic"||100||100|
|The hardener compound "atomic"||40||50|
|and / or bentonite||90||50|
|Index||The composition according to the patent of Russian Federation №2251561||The proposed composition.|
|The limit of compressive strength, MPa||75-80||60-89|
|Radiation resistance, Mrad||11·103||About 30·103|
|Radiation-chemical yield of gaseous products of radiolysis, cm3/g·happy||7·10-10||10-10|
|(at doses up to 11·103Mrad)||(at doses up to 30·103Mrad)|
|The curing time under water a day.||5|
|RMSE is the awn leaching Cs
|1·10-7||Was not observed|
|The rate of leaching Cs137(γ-activity), g/cm3·d||1·10-7||1·10-8|
|Time life hours||0.5 to 3||Around 12|
When creating a center that could be used for conservation splintered fragments of graphite reactor rings, spillages and other small radioactive fragments, located in the tank, by impregnation their compound without mixing, it is necessary that the viscosity of the compound was minimal. To reduce the viscosity of the compound was proposed to be part of an active diluent - furfural, as it has high wettability, the ability of chemical combination with the epoxy resin and contains heterocycles with radiation resistance. The addition of furfural (9-11 parts by weight per 100 parts by weight of resin) significantly reduces the viscosity and slows the curing of the compound (see figure 2).
The introduction of cement into the polymer composition contributes to a significant increase of tensile strength, yield strength and elastic modulus. The magnitude of this increase depends on the number of input marshalite and / or bentonite. Limit flowed the honor, tensile strength CTC equal to 500 and 600 kg/cm2. Introduction to CTC 50 parts by weight of marshalite leads to a significant increase in the elastic modulus (from 1200 to 1500 kg/cm2). Supplement CTC more cement and maralita (on 100 parts by weight, respectively) leads to an even more significant increase in yield strength (830 kg/cm2), tensile strength (930 kg/cm2) and elastic modulus (33000 kg/cm2). For CTC with increasing doses there is a continuous growth of the above physico-mechanical parameters to a dose of 3600 Mrad.
However, increasing the amount of mineral filler compressive strength increases up to a certain limit. Supplement CTC cement and maralita (150 parts by weight) leads to a decrease of the yield strength and tensile strength up to 550 and 700 kg/cm2.
CTC, containing in its composition 100 parts of cement and 100 parts of bentonite, has a yield strength of 480 kg/cm2the tensile strength of 560 kg/cm2the modulus of elasticity under compression of 1600 kg/cm2.
Use as filler cheaper than maralita bentonite prevents the leaching of Cs137from canned RAO, and not reduced mechanical strength.
In the patent of Russian Federation №2239643 not specified about the use of bentonite as a filler, the filler is used quartz sand (Margalit 215 mA is C).
If necessary, the removal of solid raw (precious metals, stainless steel, zirconium, non-ferrous metals etc) canned products may be exempt from CTC by heating to temperatures above 300°C.
In addition, CTC has a relatively low cost (no more than 50 rubles per 1 kg). The cost of the compound "atomic" depends on the party purchased material and an average of about 100 rubles per 1 kg of CTC can be recommended:
for conservation of reactor graphite rings in wet or dry);
- as a protective coating for metal surfaces and sealing of objects in the water, including those exposed to gamma irradiation;
for conservation of the reactor compartments of nuclear submarines (including flooded);
for preservation of aqueous solutions (containing radionuclides), mixing them with CTC with the formation of solid material;
- for preparation of concrete, which in dry condition different from the usual concrete ductility and high strength;
- to eliminate defects and repair of concrete structures and products (fill cracks, holes, cavities).
Cement-polymer composition for conservation intermediate level radioactive wastes containing resin part of the compound of cold overiden what I "atomic", consisting of epoxy oligomer, hardener - aromatic amines and furfural, and fillers cement, Marsala and/or bentonite, in the following ratio of components, parts by weight:
|The above curing agent||38-50|
|Marsala or bentonite||50-100|
FIELD: recovery of spent fuel.
SUBSTANCE: proposed method for recovering plutonium-containing sorbents of alkali metal fluorides is characterized in that sorbents are treated with water vapor or vapor-air mixture at temperature of 300 to 1000 °C. Hydrogen fluoride produced in the process is removed. After that plutonium dioxide is extracted from reaction products.
EFFECT: facilitated procedure, reduced cost.
3 cl, 1 ex
FIELD: atomic engineering.
SUBSTANCE: proposed device has pipe with flanges at ends for connection to container flange and to flange of sleeve installed on filter flange, flexible hose one of whose ends is connected to pipe end, and manifold for accumulating radioactive sorbent extracted from filter; this manifold is made in the form of concave star whose arms are made in the form of rotary tubes. Hollow cylinder mounted in star center on concave part end accommodates spring-loaded rod with counterpoise; other end of flexible hose is connected to concave star by means of flexible coupling.
EFFECT: enhanced operating reliability and simplified design of device.
4 cl, 2 dwg
FIELD: systems for processing irradiated nuclear fuel of power generation reactors, possibly in application radiochemistry for separating ruthenium out of insoluble residues after processing irradiated nuclear fuel, recovery of ruthenium out of waste catalysts or other technical products.
SUBSTANCE: method comprises steps of placing insoluble residues in electrolyzer filled with nitric acid solution with added silver nitrate; supplying electric current for transmitting easily volatile formed RuO4 by action of air passing through electrolyzer to apparatus filled with absorbing agent.
EFFECT: possibility for separating ruthenium out of insoluble radioactive residues without burying secondary radioactive deposits due to complete dissolving of initial material.
FIELD: nuclear engineering.
SUBSTANCE: proposed method that can be used for decontaminating radionuclide-contaminated metal surfaces of various nuclear power installations, manufacturing and other pieces of equipment, including those having complex configuration and subject to disposal and burial, involves their scanning by contact-arc discharges from current supply having flat current-voltage characteristic of at least 280 A/V at voltage of 7 - 4 V across working electrodes; granules of contaminated metal brought out of decontamination zone are sized to minimum 0.05 mm.
EFFECT: enhanced productivity and personnel safety.
FIELD: nuclear engineering.
SUBSTANCE: proposed method for removing spent nuclear fuel cladding includes following steps: cutting of spent fuel elements into fragments and separation of these fragments into part of fuel rods and part of spent nuclear fuel. Fuel elements are made of austenitic stainless steel. Cutting step lasts until main part of fragments measuring below 2 mm is obtained; separation step involves magnetic separation. Device implementing proposed method has cutting unit that functions to cut spent fuel rods into fragments. Mentioned cutting unit has rotary cutting tool incorporating parallel knives and shield that holds fragments until they are finely cut. Provision is made for magnetic separator that functions to magnetically separate fragments produced by cutting unit.
EFFECT: enhanced fragment separation efficiency.
5 cl, 5 dwg
FIELD: decontaminating solid iodine filters used in nuclear industry.
SUBSTANCE: proposed method includes bringing filters in contact with aqueous solution of reducing agent chosen from hydroxylamine, hydroxylamine salts, ascorbic acid, ascorbic acid salts, mixed ascorbic acid anhydrides, sodium boron hydride, sodium hypophosphate, formaldehyde, urea, formic acid, and their mixtures so as to extract iodine from filter and to dissolve it in aqueous solution. Silver can be also dissolved at the same time or sequentially in reducing agent or in any other suitable aqueous solution.
EFFECT: enhanced degree of decontamination, facilitated procedure using aqueous solution and simple vat.
FIELD: atomic industry; methods of extraction of silver from the waste sorbents containing iodine-129.
SUBSTANCE: the invention is pertaining to the field of reprocessing and utilization of the solid A-waste of the radiochemical enterprises of the atomic industry, in particular, to the method of the immobilization of iodine-129 and extraction of silver from the waste sorbents, which may be used for manufacture of the iodine absorber. The silver-containing sorbent is treated with the heated up to 75-80°C an alkali solution of hydrazine-nitrate with concentration of alkali from 30 up to 100 g/l and of hydrazine - from 15 up to 50 g/l. The solution is held during no less than 60 minutes and drained off in the separate tank for iodine-129 concentrating from it. Then conduct the sorbent water flushing. After that the sorbent is for 30 minutes treated with the nitric acid having concentration from 3 up to 10 mole/l and heated up to 80°C. The technical result is the repeated usage of the silver extracted from the waste sorbent and the increase of the operational life of the iodine purification assemblage at reprocessing of the irradiated nuclear fuel with the minimum costs.
EFFECT: the invention ensures the repeated usage of the silver extracted from the waste sorbent, the increased operational life of the iodine purification assemblage at reprocessing of the irradiated nuclear fuel with the minimum costs.
1 dwg, 2 tbl, 2 ex
FIELD: remote decontamination of surfaces, primarily those polluted by radioactive materials.
SUBSTANCE: pollutants are removed by means of grip in the form of frame that mounts flexible screen. Grip is first conveyed to working place and brought in contact with surface to be decontaminated. Then grip frame is dosed with adhesive compound in the form of aqueous solution of polyvinyl alcohol with plasticizer of 120 - 165 s viscosity and applied directly to flexible nonmetal screen. Adhesive is held to complete hardening then grip is removed together with pollutants.
EFFECT: enhanced degree of surface decontamination.
3 cl, 4 dwg, 2 tbl
FIELD: environment protection.
SUBSTANCE: proposed method includes treatment of ground with aqueous solution incorporating mineral acids used as decontaminating chemical agents. For ground decontamination use is made of mixture of sulfuric and phosphoric acids, proportion of their concentrations, as expressed in moles, being 1-3. Phosphoric acid concentration amounts to 0.5-2M. Procedure is conducted at temperature of 50-100 °C followed by separation of aqueous solution containing decontaminating chemical agents from decontaminated ground. After that cesium radionuclides are extracted from separated solution. For the purpose alkali metal or ammonia is supplied to mentioned solution until concentration of 6 · 10-4 to 7 · 10-4M is attained and cesium radionuclide containing sediment is produced. Sulfuric or phosphoric acid is added to aqueous solution with residual content of chemical agent up to working concentration and mixed up with new portion of contaminated ground.
EFFECT: enhanced safety of method and rate of ground decontamination.
1 cl, 1 tbl, 1 ex
FIELD: metal recovery processes.
SUBSTANCE: process comprises keeping metal-containing matrix in high-pressure chamber within solvent medium in presence of water and fluorine-containing organic acid followed by accumulation of extracted metal in solution. Organic acid utilized is selected from di(octafluoroformyl)phosphoric acid, di(dodecafluoroheptyl)phosphoric acid, and mixture of acid with trialkyl phosphate. Solvent is used in its liquid state.
EFFECT: enabled recovery of radionuclides from solid phase.
2 cl, 4 tbl, 4 ex
FIELD: environment protection against radioactive pollutants; immobilization of nuclear radioactive wastes.
SUBSTANCE: proposed method for solidifying liquid radioactive wastes includes their spray drying and calcination, calcination product vitrification using flux dopes, melt draining to tank, and annealing of solid block. Liquid radioactive wastes are heated in advance in continuous flow to evaporate 30-80 percentage ratio of water contained in them, and steam-water mixture produced in the process is sprayed under its own pressure in chamber heated to temperature of 600-800 °C. Solid particles of calcination product are separated from steam-gas mixture by sedimentation at temperature below 300 °C and filtration.
EFFECT: extended service life of equipment, enhanced productivity and radiation safety for personnel handling highly radioactive products.
FIELD: 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: 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: methods of liquid radioactive wastes processing.
SUBSTANCE: the invention is pertaining to the field of liquid radioactive wastes processing. The invention presents a method of neutralization of the low-mineralized and medium-mineralized low-active liquid wastes in the field conditions, which includes the liquid wastes purification by mechanical filters and ultrafilters. The subsequent desalination is conducted by reverse-osmotic filters and an after-purification - by ion-exchange filters with a reactant treatment of the spent ion-exchange resins using potassium ferrocyanide and cobalt salts. Then the treated resin is used as a sorption prefilter, in which they use purification of the wastes before their feeding to the ion-exchange filter. The formed secondary A-wastes are fixed in the stable medium. Advantages of the invention consist is an improved purification efficiency and reduction of the secondary wastes volume.
EFFECT: the invention ensures improved purification efficiency and reduction of the secondary wastes volume.
FIELD: 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
FIELD: extraction processes for recovery of nuclear fuel, uranium concentrates, and uranium-containing reusable parts.
SUBSTANCE: proposed process for uranium extraction affinage includes dissolution of uranium concentrate at nitric acid excess of 0.75 - 1.0 mole/l and temperature of 80 - 95 °C; prior to extraction uranyl nitrate solution is doped with urea nitrate; post-extraction raffinate and alkali decanting product produced as result of re-extract treatment are separately subjected to carbamide denitration with solution being cooled down and urea nitrate sediment separated; decanting products produced in the process are mixed up and subjected to electrochemical treatment.
EFFECT: reduced nitric acid consumption and escape of raffinate-containing nitrate ions, escape of nitric oxides in uranium concentrate dissolution, and uranium loss with effluents.
7 cl, 5 dwg