Method for reconditioning reusable extractant
FIELD: recovery of irradiated nuclear fuel.
SUBSTANCE: proposed method for reconditioning reusable extractant includes treatment of the latter with aqueous alkali solution. Extractant containing uranium in amount of minimum 5 g/l is treated with alkali solution whose concentration is over 10 mole/l followed by sediment separation.
EFFECT: reduced radionuclide content of reusable extractant including difficult-to-remove radioactive ruthenium.
5 cl, 2 tbl, 2 ex
The invention relates to methods of regeneration recycling of the extractant and can be used in the technology for reprocessing of irradiated nuclear fuel.
When the extraction processing of irradiated standard uranium blocks in the circulating extractant (here under the extractant is understood as the solvent, for example, tributyl phosphate, in a hydrocarbon diluent to accumulate the products of its degradation, which have the ability to retain radionuclides. Part radionuclides laundered at inter-cycle regeneration, but some, such as radiolucency accumulate due to the formation of complexes with the products of nitration and oxidation of the extractant. This form of ruthenium strongly retained in the organic phase in the regeneration operation. In aqueous solutions there are various forms nitrosobutane that can follow one another and having different extractibility. The metal complexes with the products of destruction and nitration of extractant are not destroyed either by acid or alkaline treatment of irradiated organic solutions, which leads to their accumulation in the recycled extractant. In the process of Stripping the ruthenium held nitrated organic phase, partially enters reextract that, ultimately, affects the quality of the recycled uranium. P is this increase cleaning recycling of the extractant from the ruthenium is one of the major problems at the radiochemical production.
The wide industrial application have found ways regeneration of the extractant by water soda-alkaline treatments of the extractant.
Known methods of regeneration recycling of the extractant, whereby the spent solvent before returning to extraction washed with 0.04 M solution of nitric acid and 0.1 M solution of soda; 0.5 M solution of soda, demineralized water and 0.1 M sodium hydroxide solution. (Processing of nuclear fuel. Edited Stoler, M.: Atomizdat, 1964, c.139, 257, 266).
Known methods of regeneration extractants by chemical processing of various solutions, including concentrated (2-10 mol/l) solution of alkali (Gfedorov. Radiation chemistry of extraction systems. M.: Energoatomizdat, 1986, pp.182-187) prototype.
A disadvantage of the known methods used in industry, is the low efficiency of the washing of the extractant from radionuclides, especially from radiolucent.
The objective of the invention is to increase the degree of purification of radionuclides, especially from radiolucent.
The set task is solved by the fact that in the regeneration process of recycling of the extractant, including the processing of the aqueous solution of alkali, the extractant with a uranium content of not less than 5 g/l is treated with alkali solution with a concentration of more than 10 mol/l, followed by the separation of the precipitate.
The treatment machine e is strigent with a uranium content of 10-20 g/L.
Treatment is carried out with a solution with a concentration of NaOH from of 13.75 to 15 mol/L.
The alkaline solution is added in an amount to provide a residual uranium content in the extractant after the precipitate ≤0.05 g/l
Processing of the extractant is carried out at a temperature of 60-70°C.
Because the technology of reprocessing of irradiated nuclear fuel is not difficult to get working extractant containing a certain amount of uranium (for example, when incomplete Stripping), it is advisable to make use of the cation, which is available in the technology, i.e. hexavalent uranium, and not to enter for deposition of a foreign element, from which then the extractant must be released.
Example 1. Spend 4 series of experiments with recycling the extractant (30% TBP in the hydrocarbon diluent). All experiments use the same portion of the extractant. In working extractant administered hexavalent uranium. In the experiments of the first series change the concentration of hexavalent uranium extractant coming on treatment with alkali, from 5 to 30 g/L. In the second series change the concentration of the alkali solution from 8.75 to 15 mol/l In experiments 3 series range residual uranium content in the circulating solvent from 0.01 to 10 g/l after the treatment with alkali extractant and separating the resulting precipitate. In experiments 4-series temperature deposition ur is on alkali change from 20 to 100° C. In all series of experiments to determine the content of ruthenium-106 in the circulating extractant before and after precipitation and separation of uranium and counting efficiencies from ruthenium-106.
The results of the experiments are shown in table 1.
|the number of series||Contents [U] in the extractant before treatment with alkali, g/l||Contents [U] in the extractant after precipitation and separation of uranium, g/l||Contents [NaOH] in the processing solution, mol/l||t °||The purification coefficient KPTSRu-106|
|20||0,01||of 13.75||90||the 4.7|
|20||0,01||of 13.75||100||a 4.9|
From the results of table 1 shows that the optimal conditions for purification of circulating extractant from radiolucent-106 (purification factor of 4.5) are the following: the content of uranium in the extractant coming to an alkaline treatment, 10-20 g/l (30 g/l impractical), the residual uranium content in the circulating extractant ≤0.05 g/l, the concentration of alkali in the processing solution of 13.75 from up to 15 mol/l, temperature of processing at least 60°C. the purification Coefficient of the extractant from radiolucent-106 increases with increasing temperature, but in the technology of processing of uranium blocks temperature limit 80°in accordance with requirements of fire-fighting and requirements for chemical resistance extraction solution, therefore, the best track is tons to read the temperature of 60-70° C.
Example 2. Conduct experiments with the selected optimal mode to determine the level of treatment, recycling of the extractant from all radionuclides present in it (zirconium-95, niobium-95, ruthenium-103, 106). In experiments were used recycled extractant 30% TBP in n-paraffin, the last pre-processing in the first extraction cycle soda solution.
|Radionuclide||Radiochemical composition of the extractant, Bq/l||KPTS|
|Before cleaning||After cleaning|
From the results of table 2 shows that the proposed method allows you to clear the current extractant not only from radionuclides ruthenium, but from zirconium and niobium.
After alkaline treatment of circulating extractant is separated from the pulp, clicks the functioning of the acid solution to neutralize the trapped alkali and return in the process.
The precipitate uranium containing radiolucency, dissolved in nitric acid, and the resulting solution was sent to the head of process for extraction. Radiolucency in return the uranium solution is mainly in extragonadal form, so when the extraction of radiolucency displayed in the raffinate and then goes to landfill. Thus provide the output radiolucency of technology.
The proposed method can significantly reduce the radionuclide content in the circulating extractant, including stubborn radiolucent.
1. The regeneration method of working extractant comprising processing the aqueous alkali solution, wherein the extractant with a uranium content of not less than 5 g/l, treated with alkali solution with a concentration of more than 10 mol/l, followed by the separation of the precipitate.
2. The method according to claim 1, characterized in that the processing is subjected to the extractant with a uranium content of 10-20 g/L.
3. The method according to claim 1, wherein the treatment is carried out with a solution with a concentration of NaOH from of 13.75 to 15 mol/L.
4. The method according to claim 1 or 3, characterized in that the alkaline solution is poured in an amount to provide a residual uranium content in the extractant after the precipitate ≤0.05 g/l
5. The method according to claim 1, characterized in that the treatment process is carried out at a temperature of 60-70°C.
FIELD: treatment of radioactive effluents and solid-phase saturated waters.
SUBSTANCE: some portion of organic fraction is reduced in first reactor by way of biological aerobic treatment. Filtrate/permeate taken from tangential filtering device is either directly used or supplied to first or next reactor. Solid phase is gravitationally extracted within tank of partial-flow filtering device and compacted in bottom region; concentrated effluents flowing from tangential filtering device are fed in next sedimentation region which is above first sedimentation region or above next one through intake channel; then effluents flowing above or from one side of sedimentation region are discharged through branch channel.
EFFECT: ability of selecting and technically optimizing separate modules.
34 cl, 5 dwg
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
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: 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: electrolytic apparatuses used in processes for extracting oxides.
SUBSTANCE: apparatus includes common cathode 12 and two types of anodes having different shapes and arrangement and placed in electrolytic bath 10. First anode 14 is arranged under cathode. Second anode 16 is arranged in parallel to cathode. First unit 18 for controlling electrolysis process is connected between cathode and first anode; second unit 20 for controlling electrolysis is connected between cathode and second anode. Combination of cathode and one anode is used for main electrolysis; combination of cathode and other anode is used for additional electrolysis for realizing electrolysis of material 22 in electrolytic bath.
EFFECT: prevention of non-uniformity of deposition, increased rate of processing, increased useful time period of crucible operation, processing of nuclear fuel elements in industrial processes with use of water-free processes.
3 cl, 7 dwg, 2 tbl
FIELD: nuclear industry; ceramic nuclear fuel production.
SUBSTANCE: the proposed method of ceramic nuclear fuel production includes fuel dispersion by means of oxidation with air flows, fed under the layer of heated fuel. The fuel oxidation is carried out when feeding two air flows, one of which is stopped during vibration release of obtained products of dispersion. The vibration release of dispersion products from the fuel layers is realized periodically with extraction of fine-dispersed powder. The device for ceramic nuclear fuel production includes vertical tubular multistage vibroreactor with the cascade of sieve grids and powder separator. The sieve grids are disposed one over another and equipped with poured tubes.
EFFECT: enhanced quality of ceramic nuclear fuel production.
8 cl, 1 dwg, 1 ex
FIELD: the invention refers to the field of treatment with spent nuclear fuel.
SUBSTANCE: the mode of processing irradiated nuclear fuel is in thermal oxidation of uranium dioxide on air and regeneration of received uranium protoxide in surroundings of hydrogen contents. At that oxidation-regeneration stages are conducted repeatedly. The oxidation is carried on at temperature 700-800°C and regeneration- at temperature 600-700°C. After that annealing of uranium dioxide is conducted at temperature 1000-1300°C with simultaneous vacuum distillation of volatile products of fission particular of caesium.
EFFECT: the advantages of the invention is in effective withdrawal of products of fission and in the reduction of the activity of fuel.
4 cl,1 dwg, 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