35 results about "Uranyl nitrate" patented technology

The invention discloses an application of a molybdenum disulfide-based piezoelectric catalytic material in a seawater uranium extraction material, and relates to the technical field of seawater uranium extraction, thiourea and sodium molybdate are used as precursors, and a MoS2 nanoflower with a specific 1T / 2H composite crystal structure is constructed through a hydrothermal synthesis technology. The method is placed in a uranyl nitrate solution, ultrasonic treatment is combined, efficient extraction of uranium can be achieved, and the method has potential application to extraction of uranium from seawater. The method is simple in process and environment-friendly, almost does not generate harmful waste in the synthesis process, does not need complex operation, and has good amplification preparation potential.

The invention provides a recovery treatment method of fluoride slag in uranium hexafluoride preparation, belongs to the technical field of nuclear industry waste slag treatment, and solves the technical problems that a product obtained by an existing fluoride slag treatment method is high in metal impurity content and low in uranium recovery rate, secondary fluoride slag is easily generated, and the product cannot be returned to a uranium purification and conversion production line for reuse. The method comprises the following steps: at the temperature of 400-600 DEG C, contacting the fluorinated slag with excessive water vapor to carry out high-temperature hydrolysis reaction, so that UF4 and uranium intermediate fluoride in the fluorinated slag generate uranium oxides U3O8 and UO3 with the fluorine ion content of less than 1mg / g, and filtering to obtain uranium oxides; mixing uranium oxide with a nitric acid solution, dissolving and reacting to obtain a uranyl nitrate solution; the pH of the uranyl nitrate solution is adjusted to 3-4, hydrogen peroxide is added for a precipitation reaction, solid-liquid separation is conducted, and uranium peroxide solid precipitates are obtained. The method is high in uranium recovery rate, low in product impurity content and capable of returning to a uranium purification and conversion production line for reuse.

The present invention provides a method for fixing uranium and its application in a solid fuel cell, wherein UO2(NO3)2·6H2O, Ce(NO3)3·6H2O, Y(NO3)3·6H2O and a barium salt soluble in nitric acid and easily decomposed at high temperature are dissolved in deionized water, and a novel BaCe is prepared under the action of a complexing agent. 0.8‑x U x Y 0.2 O3 (BCUY) material, which can be used in solid fuel cells as a new proton conductor electrolyte material. The raw materials of uranyl nitrate used are prepared from discarded uranium salts, uranium minerals or secondary uranium minerals, which realizes the recycling of discarded uranium resources and provides a new way for the practical application of uranium in solid fuel cells. The present invention can not only immobilize uranium well, reduce the migration of uranium, and reduce its harm to the environment; but also the new BCUY material has good chemical stability and high application value, and can be used in solid fuel cells to extend the service life of the battery.

The invention relates to the technical field of nuclear fuel circulation, in particular to a system and method for automatically preparing uranium purification stock solution from diuranate. The method comprises the following steps: granulating diuranate; the method comprises the following steps: automatically adding nitric acid according to a set target value, heating water to a set temperature, automatically adding granular diuranate of which the theoretical demand quantity is 90%, carrying out a stirring reaction, automatically detecting the uranium concentration and the nitrate concentration, and carrying out repeated fine adjustment according to a measurement result until the error between the uranium concentration and the target concentration is within 2%; the dissolved slurry is sequentially subjected to two-layer filtration and then is subjected to belt type filtration, and uranium purification stock solution is obtained. As long as the uranyl nitrate solubility range is within the uranyl nitrate solubility range, the target concentration of the dissolving solution can be set at will, the deviation is within 2%, the turbidity of the dissolving solution after treatment is smaller than 40 ppm, and the silica gel concentration is smaller than 10 mg / L. According to the invention, the diuranate is continuously and automatically dissolved, the production efficiency is improved, and the labor intensity and the cost are reduced.

The application provides a design method of a medical isotope reactor fuel solution, which comprises: the fuel solution comprises uranyl nitrate and nitric acid; the concentration of F ‑ ≤1mg / L, Cl ‑ ≤1mg / L, total boron equivalent ≤4.0μg / gU, Ru ion ≤0.02g / L, Cr ion ≤0.05g / L, Al ion ≤11.12g / L, and other impurities meet the requirements of GB / T 10266 standard; the concentration of acid, the concentration of uranium and the temperature of the fuel solution are controlled to control the pH value of the fuel solution; the operation power density of the fuel solution is not greater than 2.0kW / L, and a catalyst is used to prevent the fuel solution from generating precipitates during operation. The design method of the medical isotope reactor fuel solution provided by the application can realize long-term stable operation of the fuel solution, and is safe and economical.

A method for recovering useful metals from a nitrate solution of a metal cation or a metal oxide cation is carried out by dropping the nitrate solution from the inlet of a production column. The production column contains a circulating solution containing a base selected from any one of ammonia, ammonium hydroxide, an alkali metal hydroxide, and an alkaline earth metal hydroxide. The nitrate reacts with the base in the circulating solution and precipitates as a metal oxide salt or a metal hydroxide salt. The precipitate and the circulating solution are discharged from the outlet nozzle of the production column, the precipitate is collected by a basket disposed below the production column, and the circulating solution is recovered in a recovery tank disposed below the basket. The recovered circulating solution is pumped back from the recovery tank to the production column. The nitrate of the metal cation may be, for example, the nitrate of a radioactive metal cation such as uranium or uranyl cation.

A large-flux uranyl nitrate solution backwashing filtering device comprises a filtering shell and a base, and the filtering shell comprises a lower end shell, a middle shell and an upper end shell; the barrel-shaped filter element is located in the filter shell, a hollow backwashing rotating shaft is arranged in the center of the interior of the barrel-shaped filter element, and the hollow backwashing rotating shaft penetrates through the barrel-shaped filter element upper end cover and the upper end shell and extends to the position above the upper end shell; a large-flux uranyl nitrate solution backwashing filtering method comprises the following steps that S1, a uranyl nitrate solution is guided into a filtering shell through a material inlet, and filtered materials are guided out from a liquid guiding-out pipe; s2, judging whether the barrel-shaped filter element is blocked or not through a differential pressure gauge, and backwashing the barrel-shaped filter element by connecting an inlet joint with an outlet of an external cleaning infusion pump; s3, the cleaning liquid obtained after back flushing of the barrel-shaped filter element is guided out from a waste outlet and a liquid guide-out pipe.

This invention discloses a purification method for producing molybdenum-99 in a reactor using uranyl nitrate solution as nuclear fuel, relating to the field of radioisotope production technology. The method involves first adding an oxidant to the crude molybdenum-99 product solution produced by the solution reactor, oxidizing the crude molybdenum-99 product to MoO4. 2‑ The solution is then evaporated to dryness; a digestion solution is added to the evaporation residue to dissolve the impurities; a dissolving solution is added to the digested solid to dissolve the solid, which is then evaporated to dryness; finally, an alkaline solution is added to the evaporated solid to dissolve it, yielding a molybdenum-99 product solution that meets pharmacopoeia requirements. Compared to traditional methods for purifying molybdenum, this invention does not introduce new impurities, has a low heating temperature, low energy loss, is easy to operate in a hot chamber, and has a high impurity removal rate.

The present invention provides a uranium-containing tellurium tungstate, a preparation method and an application thereof, belonging to the technical field of new polyacid chemistry materials. The chemical formula of the present invention is [H2N(CH3)2] 14 Na4[(UO2)2(H2O)2(Te3W 21 O 75 H)2]·18H2O, belonging to the monoclinic system, space group P21 / n, and the unit cell parameters are: #imgabs0# α = γ = 90°, β = 89.899°, #imgabs1# Z = 4. After stirring sodium tungstate, sodium perchlorate and sodium tellurite in an aqueous solution, dimethylamine hydrochloride is added and the pH is adjusted, and then uranyl nitrate is added for a water bath reaction followed by post-treatment to obtain the uranium-containing tellurium tungstate. The uranium-containing tellurium tungstate of the present invention has a clear structure and good stability, and has photochromism. Its preparation process is simple and easy to operate.

The invention belongs to the technical field of uranium compounds, and particularly relates to a preparation method of high-purity uranyl nitrate. According to the method, the high-purity uranyl nitrate hexahydrate and uranyl nitrate trihydrate are continuously prepared by taking a ternary system phase diagram as guidance and combining a pH controlled micro-disturbance impurity removal technology and a mother liquor closed-loop circulation technology. In-situ selective removal of amphoteric impurities is achieved through accurate pH oscillation impurity removal, the impurity removal effect is enhanced in cooperation with thermal-state filtration, a target crystal phase is directionally separated out through regulation and control of a phase diagram, mother liquor after centrifugation can be directly reused for a dissolution procedure through split-flow compounding, and therefore full utilization of materials is achieved; and environmental pressure and resource waste caused by discharge of high-acidity waste liquid are avoided. The method provided by the invention ensures high purity of the product and crystal phase stability, realizes flow simplification, high-efficiency utilization of materials and zero wastewater discharge, reduces the comprehensive production cost, has good industrial suitability, and can meet the preparation requirements of high-purity raw materials such as nuclear fuel grade and the like.

The present application relates to a kind of large-scale industrial grade natural uranium extraction column modeling simulation method.Establish the mass transfer diffusion model of pulse baffle extraction column, establish the thermodynamic model of uranyl nitrate extraction process, establish the hydraulics model of pulse baffle extraction column, the thermodynamic model of uranyl nitrate extraction process and the hydraulics model of pulse baffle extraction column are combined, mass transfer diffusion model is solved using Runge-Kutta method, i.e., the uranyl nitrate concentration result of outlet water phase and organic phase of pulse baffle extraction column can be calculated;The calculation result is checked, the final calculation result is obtained for the output meeting the requirements, and the calculation is returned again for the output not meeting the requirements.The present application can simulate the extraction process of uranyl nitrate with concentration up to 450g / L in industrial pulse baffle column, the simulation result has high precision, and the error between simulation result and measured value of uranyl nitrate extracted by real pulse baffle extraction column is within 8%.

A method for separating molybdenum-99 and iodine-131 from a nuclear reactor fuel solution that uses a uranyl nitrate aqueous solution as fuel, comprising the following steps: S1, causing a fuel solution containing fission products to flow through a pretreated extraction column and, after elution and desorption, obtaining a 99Mo and 131I solution, the fuel solution containing the fission products being a uranyl nitrate solution obtained at the end of the operation of a nuclear reactor; S2, causing the 99M and 131I solution obtained in step S1 to flow through a separation column for 99Mo adsorption, a filler material of the separation column being ST-150 resin to which α-benzoin oxime is adsorbed; and S3, an adsorption effluent of the separation column being a 131I crude product, and successively cleaning and desorbing the separation column, so as to obtain a 99Mo crude product.

The invention discloses a high-activity uranium trioxide production method and equipment for treating large-flux uranyl nitrate, and the method comprises the following steps: introducing fuel gas into a combustion chamber through a gas supply system, and enabling the fuel gas to be subjected to combustion reaction to generate high-temperature gas; high-temperature gas is introduced into the reaction chamber, a uranyl nitrate solution is conveyed into the reaction chamber, the uranyl nitrate solution makes contact with the high-temperature gas, then a thermal decomposition reaction is conducted, and powder mixed gas is obtained; powder mixed gas in the reaction chamber is conveyed to a powder receiving hopper through an L discharging valve and filtered through a filtering device, then the gas is introduced into a tail gas purification system to recycle nitric acid, nitric oxide in tail gas is converted into nitrogen and then discharged, the powder is led out to a collecting device after being settled, and a uranium trioxide finished product is obtained. The uranium trioxide product prepared by the method is large in specific surface area, high in activity, large in treatment flux, high in tail gas treatment efficiency and good in effect, and can solve the problems that high-flux uranyl nitrate cannot be treated and nitric oxide severely exceeds the standard in the prior art.

The method for uranium purification and removal from dissolving slag based on modification and acid leaching includes the following steps: 1. Modification treatment: The dissolving slag is mixed with an alkaline solution to form a mixed solution. The alkaline solution dissolves silicon oxides, fully exposing the uranium-containing phase. The mixed solution after reaction is subjected to solid-liquid separation, washing, and drying to obtain pretreated slag. 2. Acid leaching for uranium removal: The pretreated slag is mixed with nitric acid to form a slurry, which is then acid-leached. The uranium-containing phase in the pretreated slag reacts with the nitric acid to generate uranyl nitrate solution. The slurry after acid leaching is subjected to solid-liquid separation, washing, and drying to obtain acid-leached slag. The advantages of this invention are: the uranium content of the acid-leached slag obtained at the end of the process is reduced to below 0.05%, which can be disposed of as ordinary solid waste in landfills without the need for packaging in raw material containers, thus fundamentally solving the problem of high storage costs for dissolving slag.

The invention belongs to the technical field of nuclear fuel development, and particularly relates to a preparation method of a large-grain UO2 fuel pellet. The preparation method comprises the following steps: (1) uniformly mixing a grain growth substance and an acid-deficient uranyl nitrate solution, adding saturated ammonium carbonate or introducing reaction gas for precipitation reaction to generate an AUC precipitate; the reaction gas is a mixture of NH3 and CO2; (2) filtering the AUC precipitate to obtain a filter cake, and washing, filtering, drying and calcining to obtain powder; (3) granulating the powder, and forming to obtain a green fuel pellet; and (4) sintering the green fuel pellet to obtain the large-grain UO2 fuel pellet. The average grain size of the prepared large-grain UO2 fuel pellet is larger than or equal to 40 micrometers, and the large-grain UO2 fuel pellet has the high melting point, good water corrosion performance and irradiation stability of traditional UO2 fuel and further has the low fission gas release rate and the high heat conductivity.

The present application relates to the technical field of uranium waste treatment, and particularly relates to a method for recovering uranium from sodium lime adsorbed with uranium. The method for recovering uranium from sodium lime adsorbed with uranium comprises the following steps: S1, adding a nitric acid solution to the sodium lime adsorbed with uranium to perform uranium elution, and obtaining a uranyl nitrate solution; S2, separating the eluted sodium lime from the uranyl nitrate solution; and S3, adjusting the pH value of the uranyl nitrate solution, so that the uranium is recovered in the form of sodium diuranate precipitate. The present application is simple in operation, high in uranium recovery rate, and can recycle the uranium in a recyclable form while reducing the solid waste treatment cost.

The present invention provides a method for evaporative concentration of uranyl nitrate solution. The method for evaporative concentration of uranyl nitrate solution in the present invention uses an MVR evaporation concentration system to carry out the evaporation concentration of uranyl nitrate solution. The MVR evaporation concentration system includes an evaporator and a separator. A ceramic packing layer and a stainless steel packing layer are respectively provided at the lower and upper sections of the packing section of the separator. The uranyl nitrate solution is evaporated and concentrated in the evaporator to form a concentrated solution and secondary steam. The secondary steam is purified by the separator and then recycled to the evaporator. The method for evaporative concentration of uranyl nitrate solution in the present invention can reduce the nitric acid concentration in the secondary steam and the impurity content in the concentrated solution, ensure the quality of the final product uranium trioxide in the uranium purification production line, reduce the production cost and equipment failure rate of the production line, and realize the long-term stable operation of the uranium purification production line.

The invention belongs to the technical field of nuclear power, and particularly relates to a device and method for preparing UF4. The invention provides a device and a process method for directly preparing uranium tetrafluoride from a uranyl nitrate solution. The uranium tetrafluoride preparation device combining the characteristics of centrifugal spray drying and a horizontal fluorination reactor is designed for the first time. The uranium tetrafluoride is directly prepared from the uranyl nitrate solution in the same reactor through denitration, reduction and fluorination. Therefore, the technical problems that in an existing small-amount high-abundance UF4 preparation process, the material input amount of a dry process is large, the nuclear critical safety pressure of a wet process is large, and the efficiency is low can be effectively solved, and a safe and efficient preparation device and an operation process method are provided for hundreds of grams to several kilograms of high-abundance materials in scientific research and production.

The invention provides a gas compounding system and method for catalytic dehydrogenation of a uranyl nitrate solution reactor, and belongs to the technical field of medical isotope reactors, the system comprises a hydrogen-oxygen recombiner communicated with a uranyl nitrate solution reactor container, and an injection pump communicated with the hydrogen-oxygen recombiner; a cooling water tank in which a gas-liquid separator and a heat exchanger are arranged is arranged below the injection pump; the gas-liquid separator is connected with the injection pump through a pipeline; the heat exchanger is connected with an equipment cooling water system pipeline; the top of the cooling water tank is communicated with the uranyl nitrate solution pile through an exhaust pipeline; the side part of the cooling water tank is communicated with the uranyl nitrate solution pile through an overflow pipe; the bottom of the cooling water tank is communicated with the injection pump through a pipeline, and a circulating pump is arranged on the pipeline where the cooling water tank is communicated with the injection pump. According to the gas compounding system for catalytic dehydrogenation of the uranyl nitrate solution reactor and the gas compounding method of the gas compounding system, the number of equipment is small, the structure is simple, safety is high, the integration level is high, and the operation mode is simple.

The invention belongs to the technical field of UC powder preparation, solves the problem that only one uranium source and one carbon source can be used, and discloses a UC powder preparation method which comprises the following steps: carrying out acid dissolution on uranium oxide by using nitric acid to obtain a uranyl acid solution, or directly dissolving uranyl nitrate or uranyl fluoride by using deionized water to obtain a uranyl solution, adding a carbon source into the uranyl acid solution, and stirring to obtain a UC powder solution; the method comprises the following steps: dissolving uranyl acid into an acid solution, dispersing a carbon source in the acid solution, adding ammonia water into the uranyl acid solution in which the carbon source is dispersed under a stirring condition to obtain a carbon-containing ADU precipitate, filtering the carbon-containing ADU slurry obtained by precipitation, drying the filtered carbon-containing ADU filter cake, and heating and calcining the dried carbon-containing ADU powder in an inert atmosphere to obtain UC powder. According to the method, the carbon source and uranium are mixed by adopting a wet method-dry method combined process, so that the mixing uniformity is ensured, and the uranium oxide is not only limited to uranium dioxide.

Useful metals can be recovered from a nitrate solution of metal cations or metal oxygen-containing cations by adding the nitrate solution into a forming column having inlet and outlet nozzles, the nitrate solution being added in a drop-by-drop manner through the inlet. The shaped column comprises a recycle solution comprising a base selected from the group consisting of ammonia, ammonium hydroxide, alkali metal hydroxide, and alkaline earth metal hydroxide. The nitrate is reacted with a base in the recycle solution to produce a precipitate of a metal oxide salt or a metal hydroxide salt. The sediment and the recycle solution exit the forming column through an outlet nozzle, the sediment being collected in a basket below the forming column, while the recycle solution being collected in a collection tank below the basket.

The application discloses a purification method for producing iodine-131 by a reactor with a uranyl nitrate solution as nuclear fuel, and comprises the following steps: adding a reducing agent into a crude iodine-131 product solution to reduce sodium iodine-131 into iodine-131 element; distilling the iodine-131 element to obtain iodine-131 element vapor; and absorbing the iodine-131 element vapor by using an alkaline solution to obtain a sodium iodine-131 solution product. The purification method can effectively remove radioactive impurities in the iodine-131 extracted from the reactor with the uranyl nitrate solution as the nuclear fuel, and can obtain the sodium iodine-131 solution product meeting the medical requirements. The purification method for the iodine-131 isotope is simple, and can effectively remove the impurities in the iodine-131 isotope, so that the iodine-131 isotope meets the medical requirements, and the average recovery rate of iodine can reach more than 90%.

The invention provides a method for synthesizing semi-amine ether by using air as an oxidizing agent, and relates to the field of organic synthesis. In a synthesis reaction system, a compound containing a C-H bond of ether and a nucleophilic reagent containing an N-H bond are used as substrates, uranyl nitrate is used as a photocatalyst, diphenyl diselenide is used as a cocatalyst, and synthesis of the semi-amine ether is carried out at room temperature under air atmosphere and blue LED illumination. The method does not need chemical oxidant or inert gas protection, the yield is up to 99%, the catalyst cost is low, and the method is suitable for synthesis of drugs and functional molecules. The method is efficient, green and low in cost, construction of C-N bonds is achieved under the mild condition, and the problems that a traditional method is high in pollution and cost and harsh in condition are solved.

A method of manufacturing particles of UO2 for a porous matrix of a target for use in the manufacture of 99Mo, comprising: infiltrating a solution of uranyl nitrate into a polymer template (62); either (i) introducing an alkali chemical to the uranyl nitrate infiltrated polymer template (64), causing precipitation of uranium oxide / hydroxide, and converting the uranium oxide / hydroxide to U3O8 (68) and concurrently removing the polymer template (70), by heating the infiltrated polymer template; or (ii) converting the uranyl nitrate to U3O8 (66) and concurrently removing the polymer template (74), by heating the infiltrated polymer template; and reducing the U3O8 to UO2 via heating in a reducing atmosphere (72).

This invention discloses a purification apparatus and method for iodine-131 produced by a nuclear reactor using uranyl nitrate aqueous solution as fuel, relating to the field of radioactive isotope production technology. The apparatus includes a carrier gas cylinder, an evaporator, a waste liquid tank, a primary absorption tank, a secondary absorption tank, a tertiary absorption tank, and a product tank. The carrier gas cylinder is connected to the evaporator via an inlet pipe. A liquid inlet is located at the top of the evaporator. An outlet pipe is connected to the top of the evaporator, connecting to the primary absorption tank. The primary absorption tank is connected to the secondary absorption tank via a pipe, and the secondary absorption tank is connected to the tertiary absorption tank via a pipe. Discharge pipes are connected to the bottom of the primary, secondary, and tertiary absorption tanks, and these pipes are all connected to the product tank. The bottom of the evaporator is connected to the waste liquid tank via a drain pipe. This apparatus and method achieve high product recovery and high product purity after iodine-131 purification. 131 The sodium hydroxide solution product meets the requirements of medical technical specifications.

A purification device and method for production of iodine-131 using a nuclear reactor with a uranyl nitrate aqueous solution as fuel, relating to the technical field of production of radioisotopes. The device comprises a carrier gas cylinder (01), an evaporator (03), a waste liquid tank (10), a first-stage absorption tank (05), a second-stage absorption tank (06), a third-stage absorption tank (07) and a product tank (09). The carrier gas cylinder (01) is communicated with the evaporator (03) by means of a gas inlet pipe. A liquid feeding opening (04) is provided at the top of the evaporator (03). A gas outlet pipe is connected to the top of the evaporator (03), the gas outlet pipe being communicated with the first-stage absorption tank (05); the first-stage absorption tank (05) is communicated with the second-stage absorption tank (06) by means of a pipe, and the second-stage absorption tank (06) is communicated with the third-stage absorption tank (07) by means of a pipe. The bottom of each of the first-stage absorption tank (05), the second-stage absorption tank (06) and the third-stage absorption tank (07) is connected to a liquid outlet pipe, the liquid outlet pipes being all communicated with the product tank (09). The bottom of the evaporator (03) is connected to the waste liquid tank (10) by means of a liquid discharge pipe. Using the device and method for purifying iodine-131 achieves high product recovery rate and high product purity, ensuring that an obtained sodium iodide [131I] solution product can satisfy the requirements of medical technical specifications.

A method for crystallizing, separating, and purifying uranyl nitrate hexahydrate comprises crystallizing uranyl nitrate hexahydrate crystals from a concentrated uranyl nitrate nitric acid solution, countercurrently washing the crystals, collecting the uranyl nitrate hexahydrate crystals into a collection container while squeezing washing liquid from a container into a washing area, and unloading the washed product. The uranyl nitrate hexahydrate crystals are crystallized from a feed stream under isothermal conditions, washed, and collected. The technical achievement simplifies the process flow, achieving high and independent recovery and product purity indicators. This method can be used in nuclear fuel cycle water metallurgy processes during the refining stage of natural uranium or recycled uranium.

A useful metal may be recovered from a solution of a nitrate salt of a metal cation or a metal oxycation, by adding the solution of the nitrate salt to a formation column having an inlet and an outlet nozzle, the solution of the nitrate salt being added in a dropwise fashion through the inlet. The formation column contains a recirculating solution containing a base selected from the group consisting of ammonia, ammonium hydroxide, an alkali metal hydroxide, and an alkaline earth metal hydroxide. The nitrate salt reacts with the base in the recirculating solution to produce a metal oxide salt or a metal hydroxide salt as a precipitate. The precipitate and the recirculating solution exit the formation column through the outlet nozzle and are captured the precipitate in a basket beneath the formation column while recovering the recirculating solution in a catch tank under the basket. The recovered recirculating solution is pumped from the catch tank to the formation column. The nitrate salt of the metal cation may be a nitrate salt of a radioactive metal cation, e.g., uranium or a uranyl cation.

The present invention belongs to the technical field of uranium purification and conversion, and specifically relates to a modeling method for a large-scale industrial-grade sieve plate uranyl nitrate stripping column. Input the parameters of the pulse sieve plate stripping column, which are divided into process operation parameters and structural parameters; establish a thermodynamic model of the uranyl nitrate stripping process, establish a pulse sieve plate stripping column hydraulic model, combine the thermodynamic model of the uranyl nitrate stripping process with the pulse sieve plate stripping column hydraulic model, and solve it using the Jung Kutta method; under the premise of the process operation parameters and structural parameters of the pulse sieve plate stripping column, calculate the uranyl nitrate concentration results at the outlet of the aqueous phase and the organic phase of the pulse sieve plate stripping column; verify the calculation results, obtain the final calculation results for the output that meets the requirements, and return to calculate again for those that do not meet the requirements. The present invention is suitable for simulating the natural uranium stripping process of the pulse sieve plate stripping column. The method has high accuracy, and the simulated data are all within the error range of 8% of the actual data.

The invention relates to the technical field of UN nuclear fuel, and particularly discloses a high-density single-phase UN fuel microsphere and a preparation method thereof in order to solve the problem that existing uranium nitride fuel microsphere crystal structure phase composition is relatively complex, the preparation method comprises the following steps: S1, taking a uranyl nitrate solution, adding urea, urotropine and a carbon source, and mixing to prepare carbon-containing sol; s2, dispersing the carbon-containing sol into an organic solvent through a dispersion pore plate to form carbon-containing gel balls, and then washing, drying and calcining to obtain UO3-C dry calcined balls; s3, the UO3-C dry calcined pellets are sequentially subjected to carbon thermal reduction and nitriding sintering treatment, and the single-phase UN fuel microspheres are obtained. In the stage of preparing the uranyl nitrate solution, a special carbon source is added, so that the distribution uniformity of the carbon source in the UO3-C dry calcined ball is ensured; and in the later conversion processes of carbon thermal reduction, nitridation and the like, reducing gases with different proportions are introduced, and finally, the UN fuel microspheres composed of high-density and single-phase UN components are prepared.