239 results about "Uranium" patented technology

The invention discloses a composite electrode material for removing uranium through photoelectrocatalysis and a preparation method and application of the composite electrode material, and belongs to the technical field of catalytic materials and wastewater treatment.The preparation method of the composite electrode material comprises the following steps that a ZIF-8 material, copper salt and ferric salt are mixed, and a first metal composite material is prepared; reacting the first metal composite material with a KOH solution to prepare a second metal composite material; carrying out carbonization treatment on the second metal composite material, and then carrying out acid treatment by using acid to obtain a bimetallic atom catalytic material; the preparation method comprises the following steps: dispersing a bimetallic atom catalytic material and a g-C3N4 material in a mixed solution containing phytic acid and a Nafion solution to obtain catalyst ink; and coating the catalyst ink on a substrate, and carrying out illumination drying to obtain the composite electrode material. According to the composite electrode material, under the synergistic effect of the bimetallic single atomic cluster catalyst and g-C3N4, respective advantages are fully exerted, and efficient deposition removal of uranium is achieved.

A uranium-based phosphor selected from (i) phosphors having formula I or II:where the phosphor having formula I or II is doped with an activator ion including Pr3+, Sm3+, or mixtures thereof, where 0≤a≤1, 0≤b≤1, 0.75≤x≤1.25, 0.75≤y≤1.25, 0.75≤z≤1.25, 2.5≤p≤3.5, 1.75≤q≤2.25, and 3.5≤r≤4.5; (ii) a phosphor having formula I or II, where the phosphor having formula I or II is doped with an activator ion selected from Eu3+, Pr3+, Sm3+, and mixtures thereof; and a counter ion comprising one or more alkali metal ions; and (iii) phosphors having formula IIIwhere the phosphors having formula III are doped with an activator ion selected from: Eu3+, Pr3+, Sm3+, and mixtures thereof, where A is Li, Na, K, Rb, Cs, or a combination thereof.

The invention discloses a production process of spherical alumina powder with high purity and low uranium content, which comprises the following steps: by taking angular alumina with the uranium content of 50-100ppb as a raw material, sequentially carrying out flame melting spheroidization, gradient pickling, hydrothermal strengthening uranium removal, wet grading coarse grain removal, low-temperature drying and airflow depolymerization to finally obtain the finished product spherical alumina powder with the uranium content of less than 30ppb. Through a multi-stage uranium removal system including incoming material management and control, gradient pickling, hydrothermal coupling and particle size optimization, the traditional single pickling efficiency bottleneck is broken through, and deep dissolution of lattice uranium and effective removal of coarse-grain uranium enrichment nuclei are achieved; the morphology and the specific surface are locked through whole-process low-temperature hydrothermal and low-shear drying, so that the powder is ensured to be used as a filler in resin sealing materials of high-integration integrated circuits, large-scale integrated circuits and super-large-scale integrated circuits, and the powder has the comprehensive performance of high filling, low thermal resistance and zero alpha-ray soft errors; and the limit requirements of node packaging materials of 3nm and below are met.

The invention discloses a preparation method and application of an S-type heterostructure material with a plasma resonance effect, and the preparation method comprises the following steps: grinding a precursor material, calcining, collecting solid powder, carrying out reflux purification in deionized water, and collecting white powder; dispersing the white powder in a hydrazine hydrate solution, carrying out solvothermal reaction to obtain a faint yellow solution, repeatedly dissolving and precipitating by respectively using acidic and alkaline reagents, washing and drying to obtain a faint yellow solid product; uniformly grinding the faint yellow solid product, calcining the faint yellow solid product in an atmosphere furnace, and cooling to obtain orange solid powder; and adding the orange solid powder into an organic solvent of a precursor containing a tungsten source, and centrifuging, washing and drying after the solvothermal reaction is finished to obtain the S-type heterostructure material with the plasma resonance effect. The S-type heterostructure material with the plasma resonance effect has relatively high efficiency for simulating separation and removal of uranium species in post-treatment wastewater and shows relatively good anti-radiation stability.

The invention relates to the technical field of remote sensing prospecting, and discloses a uranium-copper multi-metal remote sensing prospecting method which comprises the following steps: S1, data acquisition and preprocessing; s2, alteration information is extracted; s3, multi-source information comprehensive analysis and prospecting prediction; s4, alteration abnormity verification is carried out; and S5, delineating the prospecting target area. According to the method, vegetation spectrum interference is stripped by carrying out principal component analysis on visible light-short wave infrared of a remote sensing image, weak iron staining and hydroxyl alteration signals are extracted, a geological and remote sensing coupling technology is carried out, alteration anomaly is superposed with a geological map, geochemical data and a construction interpretation result, a comprehensive remote sensing anomaly map is obtained, and the comprehensive remote sensing anomaly map is obtained. According to the method, the multi-parameter grading model GGS is established, and the target region is graded based on the multi-parameter grading model GGS, so that the probability of missed judgment of the target region is reduced.

The present application relates to the technical field of water treatment and environmental governance, and particularly relates to a nickel-strontium titanate electrode for removing uranium, a preparation method and application thereof, and the preparation method comprises the following steps: mixing a titanium source, an alcohol amine carrier and an alkaline solution to obtain a precursor solution; adding a nickel source and a strontium source into the precursor solution, performing ultrasonic treatment and hydrothermal reaction to obtain a nickel-strontium titanate material; and coating the nickel-strontium titanate material on the surface of a carbon fiber base material to obtain the nickel-strontium titanate electrode for removing uranium. Through the combination process of nickel doping, hydrothermal reaction and carbon cloth base coating, the prepared electrode forms a porous and uniformly dispersed composite structure, and meanwhile, the electrode can have optimal selectivity and removal efficiency for uranium in the range of pH 3-9. The nickel-strontium titanate electrode for removing uranium prepared by the present application is suitable for large-scale water treatment applications, and has a good application prospect in the fields of nuclear waste water and drinking water purification.

The application discloses a uranium-based metal organic framework material and a preparation method and application thereof, and the chemical formula of the uranium-based metal organic framework material is [(UO2 2+ )(H2L 2‑ )·H2O] n , [(UO2 2+ )3(HL 3‑ )2(H2O)3·3H2O] n or [(UO2 2+ )(H2L 2‑ )·2H2O] n , wherein n is an integer greater than 0, and L is an aromatic tetracarboxylic acid. A series of uranium-based MOF materials are synthesized by using a solvent-thermal method, the X-ray stability of the MOF structure is significantly improved by using the strategy of preparing luminescent uranium-based MOF materials by using flexible carboxylic acid aromatic ligands, the prepared uranium-based MOF material has the advantages of diversified structure, stable property, excellent performance and wide application range, and the uranium-based MOF material is applied to the imaging field for the first time.

The present disclosure relates to a method for monitoring failure of coated particles in fuel elements in a core of a pebble-bed high-temperature gas-cooled reactor, which is related to the technical field of nuclear reactor engineering and includes the following steps: S11, calculating an inventory of a short-lived noble gas fission nuclide; S12, obtaining a ratio of a release rate to a birth rate of the short-lived noble gas fission nuclide based on a temperature of the fuel elements using a Booth diffusion and release model; S13, deriving a theoretical expression for an activity concentration of the short-lived noble gas fission nuclide in a primary circuit using a migration model of the nuclide in the primary circuit; S14, obtaining an experimental measurement value of the activity concentration of the short-lived noble gas fission nuclide in the primary circuit at a sampling moment by gas sampling; S15, optimally calculating a failure fraction of the coated particles in the fuel elements and a share of uranium contamination in the matrix graphite in the core based on the theoretical expression and the experimental measurement value. The present disclosure can provide key parameters for the performance and status of the fuel elements in the core, which are required for radiation safety studies, source term calculations and accident analysis of the pebble-bed high-temperature gas-cooled reactor.

The invention provides an in-situ leaching uranium mining pipe cable lifting, releasing and bundling device and method, and belongs to the technical field of in-situ leaching uranium mining well site mining. A lifting pipe and a cable double-winch mechanism are arranged, the lifting pipe and a cable can be quickly wound / released, the cable is prevented from being wound and knotted, the cable does not need to be coiled on the ground, and the cable is prevented from being contaminated by soil or dirt; the pollution to the leaching liquid in the well caused by lifting and placing the cable for multiple times is avoided; when the lifting pipe and the cable need to be put down, the lifting pipe is led out of the first winch mechanism, the cable is led out of the second winch mechanism, and the lifting pipe and the cable penetrate out of the same end of the bundling device. The lifting pipe and the cable which are unwound continuously are bundled into a whole through the bundling device, the automation level of well site operation is improved, the labor cost is saved, the labor intensity is greatly reduced, the submersible pump lowering efficiency is improved, the contact time of an operator and a pipe cable assembly is greatly shortened, and the safety of submersible pump lowering operation is improved.

The application discloses an automatic air supplement method and device for an in-situ leaching uranium gas station and belongs to the technical field of in-situ leaching uranium. The automatic air supplement method comprises the following steps: according to the pressure of a gaseous gas sample storage tank and the injection pressure, automatically controlling the air supplement process of the gaseous gas sample storage tank, so that the pressure of the gaseous gas sample storage tank is greater than the injection pressure; and according to the injection pressure and the liquid injection pressure, automatically controlling the injection process, so that the injection pressure is greater than the liquid injection pressure. Through real-time detection of the pressure of each part of the in-situ leaching uranium gas station, the embodiment of the application realizes automatic and accurate regulation and control of the carbon dioxide system of the in-situ leaching uranium gas station, improves the efficiency and saves manpower.

This invention provides an optimized method and system for laser cleaning of product containers in uranium enrichment plants. It proposes a method for optimizing laser cleaning parameters. In the actual laser cleaning process, real-time data is input into the model to obtain the prediction results, i.e., the optimal laser cleaning parameter configuration, so as to achieve the best cleaning effect. A double wedge mirror spiral scanning method is used to achieve full coverage scanning of the inner wall of the container. The cleaning speed can reach up to 9000 mm / s, which improves the speed and efficiency of cleaning.

The invention discloses a magnesium-aluminum hydrotalcite / carbon nanotube composite material and a preparation method and application thereof, and the preparation method of the magnesium-aluminum hydrotalcite / carbon nanotube composite material comprises the following steps: S1, carrying out acid pretreatment on a carbon nanotube; and S2, preparing the magnesium-aluminum hydrotalcite / carbon nanotube composite material. Compared with the prior art, the preparation method of the magnesium-aluminum hydrotalcite / carbon nanotube composite material has the advantages that Mg (NO3) 2.6 H2O, Al (NO3) 3.9 H2O, CO (NH2) 2 and pretreated carbon nanotubes are adopted, and magnesium-aluminum hydrotalcite grows on the carbon nanotubes through a hydrothermal homogeneous precipitation method; through the synergistic effect of the carbon nanotubes and the magnesium-aluminum hydrotalcite, aggregation among nanoparticles is effectively prevented, the growth size of the particles is limited, and the magnesium-aluminum hydrotalcite / carbon nanotube composite material is good in crystallinity, small in particle size, high in specific surface area and high in adsorption performance. In a radioactive uranium element adsorption experiment, the adsorption efficiency is as high as 95.6%, and the adsorption capacity is as high as 17.32 mg / g.

The invention belongs to the nuclear industry and can be used for the production of fuel pellets from uranium-molybdenum metal powder enriched with uranium-235 to 7%, which can be used as fuel elements for nuclear reactors. The sintering of the pellets is carried out in an inert argon atmosphere at a temperature in the range of 1100°C to 1155°C, as the initial powder, a particle size of 160 μm is used, with a molybdenum content of 9.0 to 10.5 wt.%. The powder is preheated at a temperature of 500°C for 10-20 hours (in an argon medium), and then cold-pressed into pellets in a mold at a pressure of up to 950 MPa. According to another method, when preparing uranium-molybdenum pellets using a binder (plasticizer), the pellets are heated in argon from 300°C to 450°C for a duration of 2-4 hours before sintering, in order to remove the binder. The invention allows to increase the uranium content in the fuel, to reduce the heat accumulated in the core of the nuclear reactor, and to reduce the energy release in the event of a violation of the normal operating conditions of the nuclear reactor, which will improve its safety and emergency stability.

The application discloses a kind of nuclear material plate uranium uniformity detection devices, it includes the feeding roller way conveying device (1) sequentially arranged from left to right, support (2) and discharging roller way conveying device (3), support (2) is respectively provided with detector assembly (4) and calibration assembly (5) on front and rear ends, detector assembly (4) is set to the front side of feeding roller way conveying device (1), detector assembly (4) includes linear module (6), detector (7) and gamma ray radiation source (8), linear module (6) is set to the top surface of support (2), the slider (22) of linear module (6) is fixed with connecting frame (9), the rear end of connecting frame (9) is provided with L plate (10), and the horizontal plate of L plate (10) is provided with gamma ray radiation source (8).The beneficial effects of the present application are: compact structure, realize the automatic flow detection of the uranium surface density value of nuclear material plate, improve the detection efficiency, simple operation.

This invention provides an application of ion emitters based on covalent organic framework materials in thermal ionization mass spectrometry (TIEMS), specifically for improving the ionization efficiency of uranium during TIEMS measurements. Compared to conventional dual-band source modes in TIEMS, and ionization enhancement techniques such as graphite methods, benzene vapor carburizing methods, and graphene oxide methods, this invention significantly improves uranium ionization efficiency, facilitating the accurate measurement of ultra-trace and low-abundance uranium isotopes.

The utility model provides an underground negative pressure sand suction device for in-situ leaching uranium mining, which relates to the technical field of in-situ leaching uranium mining and comprises an air pipe joint connected to the tail end of an air pipe. The partial pressure column is connected to the lower end of the air pipe joint; the pressure release valve is arranged at the lower end of the pressure dividing column; one end of the discharge injection inclined pipe is connected to the side face of the partial pressure column; the three-phase mixing column is parallel to the partial pressure column, and the other end of the discharge injection inclined pipe is connected to the side face of the three-phase mixing column; the rotary sand suction head is arranged at the lower end of the three-phase mixing column, and a plurality of slits are obliquely formed in the side face of the rotary sand suction head; and the sand stirring assembly is connected to the pressure release valve. According to the device, sand-carrying liquid is conveyed through the grouting pipe, the wellhead liquid-solid two-phase flow velocity is increased, the underground sand removal efficiency is improved, the rotary sand suction head can control the proportion of sucked particles, and the problems that the grouting pipe is blocked, the gas lift negative pressure is insufficient, and the particles fall back due to straight-through type opening sand suction are solved.

The present application relates to a kind of natural conditions under the estimation method of superuranic nuclide resuspension release source term.This application obtains the environmental characteristic parameter of the place of contaminated area, calculates the related parameter of superuranic nuclide pollution condition, calculates the surface covering layer factor of contaminated area soil, calculates the resuspension release source term of superuranic nuclide, can start from the characteristics of soil environment, nuclide characteristics, natural condition three aspects, based on the mass load theory of particulate matter, using the release flux of particulate matter in contaminated area soil and the level of radioactivity concentration in soil, the resuspension release source term of superuranic nuclide is calculated, realizes the resuspension release amount of superuranic nuclide in soil environment pollution under natural conditions is scientifically and accurately evaluated.

The present application relates to the technical field of soil remediation, and particularly discloses a decontamination and separation device and method for uranium-containing contaminated soil based on chemical leaching, which comprises a separation device shell, an inlet cylinder for feeding is arranged in the separation device shell, a leaching liquid mixing box for leaching of the uranium-containing contaminated soil is arranged at the rear side of the inlet cylinder, a turnover plate for digging of the uranium-containing contaminated soil is installed at the front side of the inlet cylinder, a moving wheel for walking is installed at the bottom of the separation device shell, and a protective cover for shielding of the soil is fixed at the front side of the separation device shell. When the turnover plate moves counterclockwise, the soil is scattered by separation teeth, when the turnover plate rotates clockwise, the scattered soil is dug out and thrown to an inlet inclined plate and then enters the inlet cylinder, and the soil is fed by spiral blades in the inlet cylinder, so that the soil does not need to be manually dug, the digging and feeding can be simultaneously performed, and the work efficiency is improved.

The invention relates to the technical field of nuclear reactor cores, and provides a reactor core adopting different hydrides for combined application, the reactor core comprises a plurality of fuel assemblies which are sequentially arranged in a zoning manner along the radial direction from the center to the periphery; part of the fuel assemblies simultaneously comprise uranium zirconium hydride fuel rods filled with uranium zirconium hydride fuel and uranium yttrium hydride fuel rods filled with uranium yttrium hydride fuel. Metal hydride fuel rods are introduced into the moderator-free reactor core, the reactor core neutron energy spectrum is slowed down, the reactor core fission reaction rate is increased, the reactor core critical mass is reduced, meanwhile, the equivalent nuclear density of uranium and hydrogen in different areas of the reactor core is adjusted in the mode that different hydride types are used in a combined mode, and reactor core power distribution is further flattened.

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.

This invention provides a self-driven monitoring-type uranium adsorption and recovery system, belonging to the field of radioactive wastewater treatment and resource recovery technology. The system uses a proton exchange membrane (MFC) as its basic architecture, separating the anode and cathode chambers. Its core innovation lies in the biomimetic anode in the anode chamber, a three-dimensional porous conductive composite material formed by the self-assembly and freeze-drying of graphene oxide and uranium-reducing microorganisms, possessing both high-efficiency adsorption, biocatalysis, and rapid electron transfer functions. An external circuit monitoring system collects loop current / voltage signals in real time, using current decay as an early warning indicator of adsorption saturation. The uranium recovery unit removes UO2 particles from the biomimetic anode using physical or chemical methods, achieving resource recovery, and the anode is regenerable and reusable. This system integrates adsorption, in-situ reduction, self-driven monitoring, and resource recovery, solving the problems of difficult adsorbent regeneration, lagging monitoring, reliance on external energy, and complex recovery processes in existing technologies.

The application discloses a Ce2S3@Fe composite material and a preparation method and application thereof, and belongs to the technical field of radioactive wastewater treatment, and the preparation of the Ce2S3@Fe composite material comprises the following steps: under the protection of inert gas, cerium sulfide is added into a ferric iron solution, and then a sodium borohydride solution is slowly added dropwise, so that the ferric iron is reduced to zero-valent iron by using the reducing property of the sodium borohydride solution and is uniformly loaded on the surface of the cerium sulfide, thereby forming the Ce2S3@Fe composite material; and the prepared Ce2S3@Fe composite material is applied to purification of uranium-containing wastewater. 2+ The application adopts the above Ce2S3@Fe composite material, the preparation method and the application thereof, and has the advantages of convenient operation, low cost, environmental friendliness, fast reduction rate, long UO2 stabilization time, all-weather operation, simultaneous consideration of radioactive wastewater treatment and uranium resource recovery.

This application discloses a method for preparing a porous few-layer carbon nitride / mercapto-oxime-based seawater uranium extraction composite material with a multi-level structure, relating to the field of seawater uranium extraction materials. The composite material is prepared by repeatedly calcining melamine at high temperatures to obtain a porous few-layer carbon nitride material with high photocatalytic performance. Polyacrylonitrile is functionalized with mescapto-oxime to obtain a polymescapto-oxime solution. The porous few-layer carbon nitride is then uniformly dispersed in the polymescapto-oxime solution, and a sacrificial pore-forming agent, polyvinylpyrrolidone, is added to obtain a spinning solution. Finally, a one-step electrospinning technique is used to prepare a porous few-layer carbon nitride / mercapto-oxime-based seawater uranium extraction composite material with high adsorption performance. The resulting composite material exhibits excellent uranium adsorption performance and adsorption rate, is recyclable, and the preparation method is simple and environmentally friendly. The prepared composite material has excellent uranium adsorption performance and adsorption rate, is recyclable, and has broad application prospects in the field of seawater uranium extraction.

The present application belongs to control method, and specifically relates to a kind of in-situ leaching mine uranium hydrometallurgy qualified liquid liquid separation control method.A kind of in-situ leaching mine uranium hydrometallurgy qualified liquid liquid separation control method, wherein, including the following contents: step one: set liquid separation uranium concentration;Step two: set liquid separation time;Step three: exception handling.The present application has the following remarkable effects: the uranium concentration of the precipitated qualified liquid is maintained at 36-39g / L, the uranium concentration of the qualified liquid is increased by 90-95%, the effect is good, and the operator can set the relevant parameters according to the specific situation at any time to ensure that the precipitated qualified liquid has an ideal uranium concentration.

The utility model discloses a kind of for uranium removal device of cobalt hydrometallurgy, for in the process of cobalt hydrometallurgy, the high salt solution containing metal ion is carried out uranium removal.The device includes solution module, steady flow module and drop liquid module, solution module is used to prepare the reaction solution of preset mass concentration;Steady flow module is communicated with solution module, and steady flow module is used to control the uniformity of reaction solution and the stability of flow;Drop liquid module is communicated with steady flow module, and drop liquid module is used to quantitatively drop reaction solution into the uranium removal solution to carry out uranium removal reaction.The utility model uses modular design to optimize the reaction condition of phosphoric acid precipitation method, by making the reaction solution of concentration uniform quantitatively drop into the uranium removal solution with stable flow and rate to carry out reaction, can preferentially precipitate uranium element, improve uranium removal effect, and can reduce cobalt and other metal and reaction solution to generate coprecipitate, improve cobalt recovery rate, reduce the consumption of reaction solution.

The invention belongs to the technical field of wastewater treatment and reuse, and particularly relates to a comprehensive reuse method for uranium purification rectification waste liquid. The method comprises the following steps: obtaining a rectification waste liquid diluent and alkali adjusting slurry; removing iron and aluminum impurities; complex extraction and recovery of molybdenum; purifying and recycling uranium; and recovering nitrate. By adopting the technical method disclosed by the invention, uranium, molybdenum, calcium, magnesium and the like in the uranium purification rectification waste liquid are effectively recovered. The uranium recovery rate reaches 97%, the molybdenum recovery rate reaches 98%, the calcium recovery rate reaches 95%, the magnesium recovery rate reaches 95%, almost no wastewater is discharged after treatment, and the uranium purification rectification waste liquid is turned into wealth.

The application discloses a kind of in-situ leaching uranium process in leaching agent injection control method and system, it is related to in-situ leaching uranium technical field, main purpose is to solve the matching precision of leaching agent injection parameter and injection hole ore bed characteristics is low, leading to low problem of leaching efficiency.Mainly including obtaining the real-time operation data of each branch well in target ore bed;Through the individualized liquid injection behavior model of each branch well, according to the real-time operation data, the expected leaching agent concentration space-time distribution and expected uranium leaching efficiency of corresponding branch well are simulated;According to the expected leaching agent concentration space-time distribution and the expected uranium leaching efficiency, the leaching agent injection control instruction of each branch well is generated according to concentration flow control strategy;The leaching agent injection control instruction is sent to the edge execution mechanism of corresponding branch well, to inject leaching agent into corresponding branch well by the edge execution mechanism according to the leaching agent injection control instruction.It is mainly used for controlling leaching agent injection process.

The application discloses a marine organism fossil U-Pb dating method based on LA-ICP-MS, and comprises the following steps: S1, according to stratigraphic division, marine organism fossils and iron manganese oxides of each layer are selected from a deep sea sediment column, and the rest is taken as a sediment residue; S2, the marine organism fossils are cleaned, and marine organism fossil target samples are prepared; S3, a laser ablation inductively coupled plasma mass spectrometer is used to perform laser ablation, point marking, sequencing and testing on the marine organism fossil target samples, and uranium lead isotope data is obtained; S4, baseline deduction and mapping are performed on the uranium lead isotope data, the lead isotope ratio of the sediment residue is anchored as an initial value to perform ordinary lead correction, and the age data of the marine organism fossils is obtained; the laser ablation mode is a line scanning mode. The marine organism fossil U-Pb dating method realizes dating test of marine organism fossils, and has high determination accuracy.

This invention discloses a method for constructing a physical field-enhanced electrochemical uranium deposition system, comprising: constructing an electrolytic cell: employing a bottom-in, top-out electrolytic cell with dimensions of 70 mm × 70 mm × 45 mm, the solution inlet located at the bottom left, and the outlet located at the bottom right, 20-43 mm from the bottom; and circulating the solution via a peristaltic pump connected to an external hose, using Co3O4@FeO x -IF is the cathode, NiO x -NF is the anode; a physical field coupling reaction is constructed: fluoride- and uranium-containing wastewater is added to the electrolytic cell, and the reaction is carried out under the conditions of a constructed physical field; the physical field is one of parallel coupling of a flow field and an electric field, or parallel coupling of a flow field and an electric field with an external magnetic field. This invention achieves effective product deposition, and a high removal rate can be achieved for fluoride- and uranium-containing wastewater systems when constructing a physical field, proving the feasibility and superiority of constructing a physical field.