54 results about "Nuclear fuel cycle" patented technology

The invention provides a device and method for continuously measuring the diameter of a fuel rod, and belongs to the technical field of post-irradiation inspection of nuclear fuel circulation, and the method comprises the steps: enabling a support to move relative to a sliding rail assembly so as to adjust the detection point position of a diameter measuring instrument in the axial direction of the fuel rod, and enabling the diameter measuring instrument to move relative to the support so as to adjust the detection height of the diameter measuring instrument, the measurement of different axial positions of the fuel rod is realized; the chuck can clamp the fuel rod placed on the supporting piece, and the chuck drives the fuel rod to rotate so as to realize continuous measurement of different radial positions of the fuel rod, so that the diameter value and the ovality of a detection point are obtained, the detection continuity of the diameter and the ovality of the fuel rod is improved, and the detection efficiency of the diameter and the ovality of the fuel rod is improved; meanwhile, the driving device drives the chuck to rotate so as to drive the fuel rod to rotate, it can be ensured that the fuel rod rotates in place, the rotating accuracy of the fuel rod is improved, and the accuracy of data collection and the corresponding accuracy can be improved.

The invention provides a method for directly extracting neptunium based on a conjugate cavity structure. The conjugated cavity type molecule can be used for directly and efficiently extracting neptunium (Np). On one hand, the molecule has a cavity structure highly matched with metal ions, and on the other hand, the structure has high conjugacy and can be effectively combined with actinide elements with obvious electron delocalization. Based on the characteristics, direct and efficient extraction of Np is realized under the condition of not depending on an oxidation-reduction process. The method has a good application prospect in the field of advanced nuclear fuel circulation.

The invention provides an electrochemical selective extraction process and device, and relates to the technical field of material extraction in the nuclear fuel circulation process, and the electrochemical selective extraction process comprises the following steps: respectively carrying out corrosion potential tests on an uncontaminated stainless steel sample and a stainless steel sample contaminated with nuclide; obtaining the corrosion potential a of the uncontaminated sample relative to the cathode and the corrosion potential b of the contaminated nuclide sample relative to the cathode; putting a to-be-treated metal workpiece of which the surface is adhered with a pollutant material into an anode basket of an electrochemical extraction device, and connecting the anode basket to an anode of a direct-current power supply or a potentiostat; mixing nitric acid, sodium nitrate, a corrosion inhibitor and a slow release agent to prepare an electrolyte; a cathode basket of the extraction device is connected with a negative electrode of a direct-current power supply or a potentiostat, an anode basket is nested in the cathode basket, and the anode basket and the cathode basket are jointly immersed in electrolyte for extraction. According to the method, pollutants adhering to the irregular metal surface are effectively removed and recycled, the extraction rate of contaminated metal nuclides reaches 99% or above, and the introduction amount of matrix elements is lower than 3500 ppm / g matrix.

The invention discloses a method for catalytically reducing U (VI) under mild conditions, and belongs to the technical field of heavy metal pollution treatment and nuclear environment engineering. The spherical cobalt metal loaded nitrogen-doped carbon material (Co (at) MNCS) prepared by pyrolysis is used as a catalytic material, HCOOH is used as a reducing agent and an acidity regulator, strong acid does not need to be additionally added, and U (VI) can be efficiently reduced into U (IV) under the mild conditions of 25-60 DEG C and normal pressure by utilizing high catalytic activity and structural stability of the catalyst. Within the initial concentration range of 10-30 mg / L, the reduction efficiency exceeds 99%; almost complete reduction can be realized within 9 minutes at 25 DEG C; compared with a traditional hydrazine-containing and heavy metal-containing reducing agent or precious metal catalytic system, the method has the advantages that the environmental negative influence and the catalytic cost are remarkably reduced, the reduction efficiency of the catalyst still reaches 87% or above after 5 times of circulation, and the method conforms to the industrial development direction of nuclear fuel circulation greening and uranium-containing wastewater efficient treatment and has a wide practical application prospect.

The present application relates to the technical field of nuclear fuel cycle, and particularly relates to a method for recycling uranium fluorination wastewater. The method is as follows: adjusting the concentration ratio of bicarbonate and carbonate in the uranium conversion leaching solution, carrying out decolorization and filtration treatment to obtain a filtrate; the filtrate is subjected to uranium adsorption through a resin, and the resin adsorbed with uranium is subjected to leaching; the adsorption tail liquid is added with magnesium hydroxide slurry to adjust the pH to above 11, and then filtered, and the filtrate is adsorbed by amidoamine-based adsorption fiber; the adsorption tail liquid is added with magnesium carbonate slurry to generate magnesium fluoride and carbonate ions, and then filtered to obtain a sediment and a filtrate respectively; the filtrate is returned as a leaching agent of fluorination tail gas or is discharged after treatment; the sediment is used together with new magnesium carbonate to participate in fluorine removal reaction again, and the unreacted magnesium carbonate is used again. The present application realizes efficient recovery of uranium in the leaching solution and recycling of the leaching solution, fully utilizes the residual alkali in the returned liquid, reduces the treatment cost of fluorination tail gas, controls the radioactivity of the waste residue, and realizes radioactivity exemption of the treated sediment.

The present application relates to the technical field of nuclear fuel cycle and radioactive waste treatment, in particular to a method and system for treating uranium-containing organic phase, the system comprises pretreatment washing tank, pyrolysis reactor, phase separation tank, uranium recovery device, vacuum distillation device, tail gas treatment device and solidification device connected in sequence; the method first removes acid by water washing pretreatment of uranium-containing organic phase at 40-50 DEG C, then adds phosphoric acid to catalyze pyrolysis to decompose TBP and transfer uranium to phosphoric acid phase, recovers uranium and n-dodecane after phase separation, simultaneously treats tail gas and solidifies residual waste, and further provides online monitoring and control subsystem to realize automatic control and safety interlocking; the present application has mild reaction condition, avoids strong corrosive substance generation, total uranium recovery rate reaches more than 90%, about 80% of n-dodecane can be recycled, greatly reduces radioactive waste production amount, is safe and reliable in operation, and is suitable for radioactive waste organic extractant reduction and resourceful treatment in nuclear fuel cycle.

The invention relates to the technical field of nuclear fuel circulation uranium conversion, in particular to a method for treating raffinate water-phase distillation raffinate. The treatment method comprises the following steps: adjusting the pH value of the distillation residual liquid to be greater than 5 by using sodium hydroxide, precipitating impurities and uranium together, and then filtering and separating to obtain filtrate and a filter cake; adding a mixed solution of sodium bicarbonate and sodium carbonate into the filtrate to obtain a leaching agent; a leaching agent is added into the filter cake, uranium in the filter cake is leached out, solid-liquid separation is conducted, and leaching residues and leaching liquid are obtained; mixing the leachate with acid and hydrogen peroxide, controlling the acidity pH of the final solution to be 1.5-4.0, and carrying out solid-liquid separation to obtain uranium peroxide solid and uranium peroxide precipitation mother liquor; the uranium peroxide solid is dissolved by nitric acid and then enters the extraction and purification process again. According to the method, the adverse effect of sodium nitrate in the raffinate water phase on the distillation and concentration process is avoided, and the recovery rate of the whole uranium purification and extraction process is increased to 99.99% or above.

The present invention relates to a uranium peroxide cluster with a core-shell structure and a preparation method and application thereof, wherein the crystal structure of the uranium peroxide cluster is a core-shell cluster structure; the core of the uranium peroxide cluster is formed by connecting 36 uranium polyhedra in a co-edge manner; the shell of the uranium peroxide cluster is composed of 6 uranium pentamers, each of which is composed of 5 uranium polyhedra and 5 oxalate ligands, and the 5 uranium polyhedra are connected in a co-edge manner to form a five-membered ring structure, and the negative charge of the uranium peroxide cluster is neutralized by potassium ions. The method of the present invention adopts oxalate with strong coordination ability with uranyl ions as ligands, selects potassium ions as counter ions, and prepares a uranium peroxide cluster with novel structure and properties by regulating the amount of reaction reagents, which is expected to be applied and promoted in nuclear fuel cycles.

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 present invention provides an operational plan generation device, etc., that generates operational plans that take into account the uncertainty of future conditions regarding nuclear fuel cycle facilities. [Solution] The operation plan generation device 10 includes a constraint acquisition unit 121 that acquires combinations of multiple patterns of constraints as operational constraints for the nuclear fuel cycle facility, an operation plan search unit 122 that searches for an operation plan for the nuclear fuel cycle facility toward the optimization of a predetermined objective function under the constraints of each of the aforementioned combinations, and an output control unit 126 that outputs the search results for the operation plan for each of the aforementioned combinations.

The invention discloses a heating system and a heating method for simulating a water loss accident, and belongs to the technical field of nuclear fuel circulation and irradiation effect research, the heating system comprises a fuel rod, the fuel rod comprises a fuel cladding and a fuel core body, the fuel rod is arranged in a quartz tube sample cavity, the whole fuel rod is arranged in a reaction furnace body, and an infrared temperature detector monitors the temperature in real time through a peep window; and data is fed back to the processing box to dynamically adjust power output. The fuel rod is used as a resistance material, low-voltage alternating current is applied, self-heating is achieved through the Joule effect, heat transfer lag of a traditional heating mode is eliminated, the infrared temperature detector monitors the temperature in a non-contact mode through the squint window and feeds back the temperature to the processing box in real time to dynamically adjust power output, and closed-loop control is formed. And the ultrafast temperature rise requirement under the water loss accident working condition is accurately simulated.

The present disclosure relates to nuclear chemical, particularly radiochemical, technologies at different stages of the nuclear fuel cycle, such as the production of purified nuclear materials (uranium, zirconium) or the reprocessing of spent nuclear fuel from nuclear power stations, in which extraction processes and operations for purifying nuclear materials are used. An example method, which includes the partial decomposition of nitric acid during continuous evaporation while a solution containing a reducing agent is fed into the bottom part of an evaporator having a circulating bottoms solution, consists in that the process is carried out such that the solution is kept in the bottom part of the evaporator for more than 2 hours under the addition of an aqueous solution of formaldehyde and formic acid (hereinafter “the mixture”) or a solution of formic acid after the process has been started using the mixture.

The invention relates to the technical field of nuclear fuel cycle post-treatment, and provides a method for separating caesium-137. The method specifically comprises the following steps: digesting the caesium-137-containing waste by using an acid solution to obtain a raw material solution, carrying out acid removal treatment on the raw material solution to obtain a to-be-separated solution, and carrying out chromatographic separation treatment on the to-be-separated solution to separate out the caesium-137. The cesium-137 separation method can effectively realize cesium-137 separation, has the advantages of simplicity and convenience in operation, low energy consumption, low cost input and promotion of resource recycling, and is suitable for large-scale application.

The invention belongs to the technical field of nuclear fuel circulation and waste liquid treatment, and particularly relates to a method for screening and separating uranyl ions from a strong nitric acid water phase containing uranium, niobium and zirconium. The preparation method comprises the following steps: step 1, preparing a graphene oxide film by adopting an improved Hummers oxidation method; step 2, preparing a permeation device; step 3, carrying out an ion screening experiment through the permeation device; and 4, sample determination and result calculation. According to the method, the graphene oxide film is developed to serve as a screening and separating material, the screening and separating rule of the graphene oxide film on uranium in a U-Nb-Zr system in a high-acid solution is explored, the related screening and separating technological process is established, and an early-stage technical foundation can be laid for scientific research and production of the related material development process in the future.

The application belongs to the technical field of nuclear fuel cycle and nuclear waste liquid treatment. The application reports a kind of carboxyl-substituted o-phenanthroline water-soluble ligand based on selective masking of part of trivalent actinide under high acidity conditions, and under optimal conditions, the separation coefficient of trivalent europium can reach 120 in 1.5M nitric acid; the separation coefficient of trivalent cerium can reach 4.4. The single crystal structure of the complex is a double metal dimer structure, which guarantees the high acidity tolerance and high selectivity of the ligand. The synthesis of the ligand involved in the application is simple, the extraction efficiency is high, and the extraction mode is novel; the related content will greatly promote the design and performance optimization of water-soluble lanthanum, lanthanum, actinium and actinium separation ligand.

The invention relates to the field of nuclear technologies, in particular to an oxide-based uranium-plutonium nuclear fuel for an atomic power station with thermal neutron reactors. The technical result of the invention is the development of a REMIX-type fuel composition formulation making it possible, in the nuclear fuel cycle, to involve fissile nuclear materials regenerated from spent nuclear fuel with the simultaneous use of enriched natural uranium, and also moreover providing a solution to the problems of separately using previously isolated plutonium (including weapons-grade plutonium), of reducing the quantity of spent nuclear fuel by recycling fissile materials from spent nuclear fuel during the lifespan of a nuclear reactor (up to 5 recyclings), of economizing on natural uranium for manufacturing fuel and, as a result, of making the nuclear fuel cycle self-sustainable. The invention makes it possible to use REMIX fuel for solving an entire range of problems which a closed nuclear fuel cycle may face, taking into account the number of reactors, availability of natural uranium and the amount of stored spent nuclear fuel.

This invention provides an asymmetric o-phenanthroline amide pyrazole compound, its preparation method, and its applications, belonging to the field of nuclear fuel cycle and high-level radioactive waste treatment technology. The asymmetric o-phenanthroline amide pyrazole compound provided by this invention has the structure shown in Formula I. The asymmetric o-phenanthroline amide pyrazole compound provided by this invention exhibits good performance against Am under high acidity conditions. 3+ The extraction partition ratio is high, and the separation effect is excellent. Under 3 mol / L nitric acid conditions, the asymmetric o-phenanthroline amide pyrazole compound provided by this invention has a high separation efficiency for Am. 3+ and Eu 3+ Extraction partition ratio ( D The separation factors were 3.32 and 0.09, respectively. SF Am / Eu It is 36.9, for Am 3+ It has a good separation effect.

The invention belongs to the technical field of nuclear fuel cycle analysis, and relates to a pretreatment process for step-by-step separation of thorium, uranium, americium and plutonium in neptunium dioxide powder, and the process comprises seven steps of sample dissolution, sample valence adjustment, Th separation, U separation, Am separation, Pu separation and content determination. According to the method, the TEEVA, UTEEVA and TRU extraction chromatography resin coupling technology is applied, and step-by-step separation of Th, U, Am and Pu in neptunium dioxide powder is achieved; measuring the content of the separated trace actinide elements by using an inductively coupled plasma mass spectrometer (ICP-MS); through the step-by-step separation pretreatment process, the adding standard recovery rate of each element of Th, U, Am and Pu is greater than 90%, and the precision of the method is superior to 10%.

The invention provides a micron-sized spherical inorganic composite material, a preparation method and application thereof, and an adsorption method of Cs < + > in a solution, and belongs to the technical field of radioactive waste treatment in nuclear fuel circulation. The method comprises the following steps: mixing water, absolute ethyl alcohol, hydrochloric acid, a silicon source and ammonium phosphomolybdate, and concentrating to obtain a silicic acid solution; mixing benzyl alcohol, a methyl cellulose aqueous solution and a benzyl alcohol solution of an emulsifier to obtain a dispersion system; and adding the silicic acid solution into the dispersion system, and then heating to obtain the micron-sized spherical inorganic composite material. In the invention, absolute ethyl alcohol is used as a solvent to improve the solubility of a silicon source, hydrochloric acid is used as a catalyst to catalyze hydrolysis reaction of the silicon source, benzyl alcohol provides a hydrophobic environment, a methylcellulose aqueous solution provides a spherical template to induce formation of a spherical structure, and an emulsifier reduces surface tension to make a system form a stable emulsion. The micron-sized spherical inorganic composite material with high adsorption capacity, high selectivity and good stability is prepared.

The invention relates to the technical field of nuclear fuel circulation and irradiation effect, and provides a movable forcibly-placed sample transfer lead tank suitable for an LOCA shielding glove box, which comprises a transfer trolley, a lead tank is horizontally and slidably connected to the transfer trolley, a shielding cavity is formed in the lead tank, and a push-pull rod is axially and movably connected to the rear end of the lead tank. One end of the push-pull rod on the inner side of the shielding cavity is provided with a sample bin for loading a forced sample, the front end of the lead tank is connected with a first joint, and the end part of the first joint is detachably connected with a second joint for matching and butting a glove box; the replaceable second connector is matched with glove box flange connectors of different specifications, various glove boxes can be matched without overall transformation, the adaptability of the device is improved, the alignment difficulty of the lead tank and the glove box is reduced through optical alignment of the laser transmitter and the target and mechanical positioning of the positioning protrusion and the positioning hole, double guidance is achieved, and the production efficiency is improved. And the docking efficiency is improved.

The embodiment of the invention relates to the technical field of nuclear reactor design, in particular to an arrangement structure of a reactor nuclear power system and a reactor workshop.The arrangement structure comprises a purification and separation device which is arranged to purify and separate mixed gas in a reactor to obtain gaseous fuel and foreign gas; the fuel generating and conveying device is used for generating gaseous fuel, inputting the gaseous fuel into the reactor and inputting the gaseous fuel purified and separated by the purifying and separating device into the reactor; the thermoelectric conversion device is arranged to convert heat generated by the gaseous fuel in the reactor into electrical energy. According to the embodiment of the invention, the reactor, the purification and separation device, the fuel generating and conveying device and the thermoelectric conversion device are subjected to integrated same-site same-plant design, so that the circulation management of production, conveying, reaction and purification and separation of fuel can be realized, and the comprehensive utilization of energy and the whole-process management of nuclear fuel circulation are realized; the economical efficiency of reactor operation is improved, and leakage of radioactive substances can be reduced.

This invention discloses a method for operating a high-burnup spent fuel transfer container with graded pressure monitoring and depressurization, belonging to the field of nuclear fuel cycle technology, and includes three stages: loading, transfer, and unloading. In the loading stage, the container is connected to the loading well through-hole and filled with water to balance the water level before loading fuel. After completion, leakage testing and radiation monitoring are performed. In the transfer stage, a qualified container is transported as a whole to the target facility using a tractor. In the unloading stage, before opening the lid, the pressure in the lid chamber and inner chamber is detected via a quick-connect coupling. If overpressure occurs, the pressure is controlled by connecting to the facility's exhaust gas system. After fuel assembly damage monitoring, the container's inner cavity is cooled and cleaned. Subsequent steps include precise positioning, inner lid disassembly and cleaning for protection, and closed-loop recording. This method solves the problems of low positioning efficiency, high high-altitude risks, high personnel radiation dose, and equipment vulnerability in existing operations. It achieves graded pressure control and systematic pollution prevention and control, improving the safety, accuracy, and efficiency of operations, and has strong compatibility with existing facilities.

The invention discloses a pebble bed dust migration filtering characteristic simulation method and system, a terminal and a medium, and relates to the field of fusion reactor nuclear fuel circulation. The simulation method comprises the steps of cladding multiplication pebble bed particle accumulation simulation, pebble bed structure data transmission, tritium extraction gas purging simulation in a cladding multiplication pebble bed, flow field data transmission, radioactive dust particle movement migration process simulation in the cladding multiplication pebble bed and radioactive dust particle capture deposition process simulation in the cladding multiplication pebble bed. And finally, obtaining data through result post-processing, and carrying out analog simulation analysis on the blowing migration filtering characteristics of the radioactive dust in the pebble bed of the fusion reactor cladding multiplication tritium production area under different multiplication cladding types, different blowing working conditions and different dust characteristics. According to the method, detailed dust migration filtering characteristic data can be obtained, and the purpose of improving the fineness and accuracy of radioactive dust migration filtering behavior simulation is achieved.

The embodiment of the invention relates to the field of data processing methods particularly suitable for matrix or vector calculation of specific functions, in particular to a method for determining a subcritical limit value of a target nuclear fuel cycle process scene. According to the method for determining the subcritical limit value of the target nuclear fuel cycle process scene provided by the embodiment of the invention, the reference experiment is screened according to the target nuclear fuel cycle process scene, so that the selected reference experiment data meeting the predetermined requirements are obtained, and the quality of the reference experiment data for verification is improved; correcting kernel data adopted by theoretical calculation of the selected reference experiment, so that the theoretical calculation is closer to a real physical scene; according to the corrected nuclear data, the offset and uncertainty of the calculation tool are determined, the uncertainty level caused by the calculation tool is reduced, unnecessary conservative assumptions are reduced, under the same safety level, the limiting conditions of the nuclear fuel circulation process are reduced, and the processing capacity of nuclear materials in nuclear fuel circulation process equipment is improved.

The invention discloses a nitrogen-oxygen atmosphere controllable high-temperature oxidation method for porous sponge zirconium, and relates to the technical field of nuclear fuel circulation and materials.The method comprises the following steps that S1, unqualified porous sponge zirconium materials are loaded into a high-temperature oxidation furnace, and then inert gas is introduced into the oxidation furnace to exhaust air; s2, the furnace temperature is increased to the target oxidation temperature at the set heating rate, and meanwhile mixed gas of nitrogen and oxygen is introduced into the furnace; s3, at the target oxidation temperature, keeping constant temperature and continuously introducing the mixed gas for oxidation reaction; and S4, after oxidation is finished, heating is stopped, and inert gas continues to be introduced for protective cooling to the room temperature. According to the method provided by the invention, the catalytic effect of nitrogen in the oxidation process is utilized to promote the porous sponge zirconium to be quickly and uniformly converted into zirconium oxide which is stable in phase and adaptive to subsequent ball milling in particle size and hardness, and the oxidation conversion rate is greater than or equal to 98%.

The invention relates to the professional field of nuclear fuel circulation and materials, and provides a Cr coating and zirconium alloy interface bonding strength analysis method, device, equipment and medium, the method comprises the following steps: establishing a Cr / Zr interface model, and carrying out sufficient structural relaxation to obtain a relaxed Cr / Zr interface model; calculating the interface binding energy corresponding to each relaxed Cr / Zr interface model, and screening out a target Cr / Zr interface configuration with the minimum interface binding energy from the relaxed Cr / Zr interface models; performing interface load deformation operation on the target Cr / Zr interface configuration according to a preset use condition, and performing structure optimization to obtain corresponding relation information between energy and strain; corresponding relation information between traction force and strain is determined; and according to the corresponding relation information between the traction force and the strain, the bonding strength of the Cr / Zr interface is obtained through calculation. According to the technical scheme, the theoretical strength of material interface bonding can be obtained, and theoretical support is provided for coating research and development and stability analysis.

The invention relates to an explosion-proof bag type dust collector for a nuclear fuel circulation facility. Comprising a shell, a dust removal bag, a feeding pipe, a discharging pipe and an adjusting assembly. The dust removal bag is arranged inside the shell; an explosion-proof structure is arranged on the shell; the other end of the feeding pipe is provided with an adjusting assembly; one end of the discharge pipe is communicated with the bottom of the shell, and the other end is a discharge port; the adjusting assembly is used for controlling the feeding pipe to communicate with or not communicate with the outside and comprises a first sleeve, a communicating piece and a driving assembly. One end of the first sleeve communicates with the feeding pipe, and the other end of the first sleeve communicates with the outside and comprises a communicating section and a closed section; the communicating part is arranged in the first sleeve, a communicating channel communicating with the radial side face and the axial end face facing the feeding pipe is formed in the communicating part, the communicating part is driven by the driving assembly to reciprocate between the communicating section and the closed section, and when the communicating part is located in the communicating section, the outside and the feeding pipe are communicated through the communicating channel; and the radial communication port of the communication channel is blocked by the inner wall of the closed section when in the closed section.

The invention relates to the technical field of nuclear fuel circulation uranium conversion, in particular to a method for preparing exempt calcium fluoride slag from a uranium-containing and fluorine-containing solution. The method comprises the following steps: 1, adding calcium carbonate into the uranium peroxide precipitation mother liquor, reacting, and carrying out solid-liquid separation to obtain a solid mixture of calcium fluoride and calcium carbonate and calcium fluoride precipitation mother liquor; 2, deeply purifying one part of the calcium fluoride precipitation mother liquor to remove uranium and fluorine, then discharging, and using the other part as an absorbent; a calcium carbonate precipitation replacement method is used for replacing a lime milk direct precipitation method, the precipitation environment of calcium fluoride is changed from strong basicity to a carbonate buffer system, the characteristic that carbonate and uranyl ions generate a stable complex is utilized, uranium is prevented from being attached to the precipitate, the concentration of uranium in the mother liquor is kept basically stable, and the method is suitable for industrial production. The specific radioactivity intensity of the calcium fluoride slag reaches the exempt level, and compared with the mode that the calcium fluoride slag serves as the emissive slag to be stored, the storage cost can be greatly reduced, and the influence on the environment is reduced.

This invention discloses a modular, controllable atmosphere high-temperature oxidation device for the recovery of substandard sponge zirconium, relating to the fields of nuclear fuel cycle and materials technology. The device includes: a furnace body system, comprising, from the outside to the inside, a steel outer shell, a refractory insulation layer, and a detachable furnace liner; multiple independently controlled heating zones arranged along the furnace body axis, each equipped with an independent heating element and thermocouple; an atmosphere control system including nitrogen and oxygen sources, an inert gas source, a mass flow controller, and a mixer; a sealing and cooling system employing a high-temperature sealing ring and a double-layer furnace shell design, with circulating cooling water flowing through the interlayer; and an automatic control system equipped with a temperature-atmosphere-time automated control program for fully automated operation and remote monitoring of the oxidation process. The device provided by this invention can achieve efficient and uniform oxidation treatment of irregular porous sponge zirconium within the range of 500-1000℃, transforming it into zirconium oxide with consistent physical properties.

The present invention relates to the field of nuclear chemistry, in particular radiochemistry, technologies for different stages of the nuclear fuel cycle, such as production of purified nuclear materials (uranium, zirconium) or reprocessing of spent nuclear fuel from nuclear power plants, where extraction processes and operations are used for purification of nuclear materials. The essence of the method, which includes partial decomposition of nitric acid during continuous evaporation while feeding a solution containing a reducing agent to the bottom part of an evaporator with a circulating bottom solution, lies in conducting the process so that after starting the process with an aqueous solution of formaldehyde and formic acid (hereinafter referred to as "mixture"), the solution remains in the bottom part of the evaporator for more than 2 hours with addition of the mixture or a solution of formic acid. This makes it possible to select a scenario where during evaporation of a model highly active raffinate at a given concentration level (specific volume of the bottom solution does not exceed 0.4 m 3 / ton of radioactive waste), barium nitrate does not precipitate and nitrous oxide is not released. In addition, due to a significant decrease (up to one tenth) in the formaldehyde concentration in the reducing agent and the possibility of continuing the process with a formic acid solution in the complete absence of formaldehyde, the fire and explosion safety of the process increases.