146 results about "Fuel reprocessing" patented technology

The invention relates to a crown ether covalent organic framework material as well as a preparation method and application thereof. The preparation method comprises the following steps: dissolving a crown ether monomer and an amino monomer in an o-dichlorobenzene solvent, adding a catalyst acetic acid aqueous solution, and uniformly dispersing to form a reaction solution; the method comprises the following steps: freezing a reaction solution through a liquid nitrogen bath, vacuumizing, filling nitrogen, carrying out a heating reaction, and carrying out post-treatment operations such as washing and drying on a reaction product, so as to obtain the crown ether covalent organic framework material BADT-DB18C6-COF. The crown ether covalent organic framework material is formed by condensation of a crown ether aldehyde monomer and an amino monomer, the crown ether monomer is 4, 4 ', 4' ', 4 '''-(6, 7, 9, 10, 17, 18, 20, 21-octahydrodibenzo [B, K] [1, 4, 7, 10, 13, 16] hexaoxacyclo octadecene-2, 3, 13, 14-tetrayl) tetrabenzaldehyde, and the crown ether monomer is 4, 4 ', 4' ', 4'''-(6, 7, 9, 10, 17, 18, 20, 21-octahydrodibenzo [B, K] [1, 4, 7, 10, 13, 16] hexaoxacyclo octadecene-2, 3, 13, 14-tetrayl) tetrabenzaldehyde. The amino monomer is 4 ', 5'-bis (4-aminophenyl)-[1, 1 ': 2', 1 ''-terphenyl]-4, 4''-diamine. The crown ether covalent organic framework material BADT-DB18C6-COF prepared by the invention is mainly used as an adsorbent, is used for adsorbing and recovering uranium, thorium and americium ions from a solution, shows good adsorption capacity on radionuclides U-238, Th-232 and Am-243, and is particularly suitable for the field of spent fuel post-treatment in the nuclear industry.

The invention relates to corrosion-resistant austenitic stainless steel for spent fuel aftertreatment and a preparation method of the corrosion-resistant austenitic stainless steel, belongs to the field of nuclear materials and technologies, and solves the key problem that existing domestic commercial austenitic stainless steel is prone to intergranular corrosion in a nitric acid medium in a special spent fuel aftertreatment environment. The stainless steel comprises the following components in percentage by weight: 15.0 to 20.0 percent of Cr, 12.0 to 18.0 percent of Ni, 3.0 to 6.0 percent of Si, 0.05 to 2.0 percent of Mn, 0.05 to 2.0 percent of W and Mo, 0.20 to 0.20 percent of N, 0.20 to 0.20 percent of Nb and V, 0.05 to 0.05 percent of rare earth element RE and the balance of Fe and inevitable impurity elements: less than or equal to 0.010 percent of C, less than or equal to 0.005 percent of P, less than or equal to 0.005 percent of S, less than or equal to 0.005 percent of O and less than or equal to 0.005 percent of B. The preparation method of the stainless steel comprises the steps of vacuum induction melting, vacuum consumable remelting, thermal machining and solution treatment.

The invention relates to the technical field of high polymer materials, and particularly discloses a nuclear-grade cable for spent fuel aftertreatment and a preparation method of the nuclear-grade cable. Comprising the following steps: S1, taking hydroxyl-containing cyclotriphosphazene, tetra (4-hydroxybenzene) ethylene, oleic acid acyl chloride, guar gum hydroxypropyl trimethyl ammonium chloride, triethylamine and a mixed solvent, cooling, dropwise adding triethylamine in a nitrogen atmosphere, dropwise adding a phenoxyphosphoryl dichloride solution, stirring for reaction, filtering, and performing rotary evaporation to obtain a composite flame retardant; s2, polyethylene, aluminum hydroxide, a composite flame retardant and an anti-radiation agent are added into an internal mixer for mixing, a cross-linking agent is added, uniform mixing, extrusion and granulation are performed, and a cable material is obtained; the cable material is used for preparing an outer sheath layer of a nuclear-grade cable. S3, twisting a plurality of copper metal monofilaments to form a conductor, wrapping the conductor with an insulating layer, and wrapping the conductor with a polyimide film to form a wire core; a plurality of wire cores are twisted to form a cable core; a non-woven fabric wrapping tape layer and an outer sheath layer are wrapped outside the cable core; and obtaining the nuclear-grade cable.

The invention relates to the technical field of spent fuel post-processing, and provides an anode and an electrolyzer, the anode comprises a conductive mesh basket and a conductive wire, the conductive mesh basket is provided with a containing cavity and an overflowing hole, and the overflowing hole is communicated with the containing cavity; the conductive wires are connected with the conductive net basket and used for transmitting electric energy to the conductive net basket. According to the anode provided by the embodiment of the invention, the cost of the anode can be reduced, the process is simplified and the process difficulty is reduced.

The invention discloses a medium-strength high-plasticity Ti-Ta alloy resistant to boiling concentrated nitric acid corrosion. The medium-strength high-plasticity Ti-Ta alloy is composed of, by mass, 5.5%-6.5% of Ta, 0.16%-0.6% of O and the balance Ti and inevitable impurities. In addition, the invention further discloses a preparation method of the Ti-Ta alloy, sponge titanium, Ti-Ta intermediate alloy scraps and titanium dioxide powder wrapping titanium foil are mixed and then smelted, and a mixture is obtained; and then rolling and annealing to obtain the Ti-Ta alloy. The alpha-type titanium alloy with a uniform structure and a single phase is obtained through oxygen microalloying, so that the alloy has excellent mechanical properties and nitric acid corrosion resistance and also has excellent welding performance and machinability, the alloy is simple in component, the preparation method is short in period and high in efficiency, the strong plasticity and the nitric acid corrosion resistance are synergistically improved, and the preparation method is suitable for industrial production. Basic data support and candidate materials are provided for reinforcement and corrosion prevention of metal structure materials for spent fuel aftertreatment equipment and design of new materials.

The invention provides a remote communication method, device and equipment for a mixing and clarifying tank stirring device, belongs to the technical field of nuclear fuel post-processing, and solves the problem of control signal fluctuation caused by electromagnetic interference, cable aging and environment abrupt change in a complex industrial scene. The method comprises the following steps: acquiring signal quality data; preprocessing the signal quality data to obtain preprocessed signal quality data; determining predicted signal attenuation data according to the preprocessed signal quality data; comparing and analyzing the predicted signal attenuation data with a preset threshold value to obtain a control instruction; and controlling the mixing and clarifying tank to communicate according to the control instruction. According to the scheme, the anti-interference performance and the prediction precision can be improved, and the communication failure rate and the energy consumption are reduced.

The utility model discloses a secondary pre-diffraction high-precision X-ray fluorescence analysis system which comprises an X-ray tube used for generating an X-ray source, a detector and a measuring cell used for storing feed liquid to be measured, a first collimator is arranged between the X-ray tube and the measuring cell, a first diffraction crystal is arranged below the measuring cell, and a second collimator is arranged below the first diffraction crystal. A second diffraction crystal is arranged obliquely above the first diffraction crystal, a second collimator is arranged between the second diffraction crystal and the detector, and the signal output end of the detector is connected with a computer system used for X-ray tube power control and measurement result analysis and calculation. According to the utility model, the characteristic X-ray of Pu in the material liquid to be detected is subjected to monochromaticity through secondary diffraction of the first diffraction crystal and the second diffraction crystal, so that the detection limit of the concentration of Pu in the material liquid of the spent fuel post-treatment process can be delayed by one order of magnitude, high-precision analysis is realized, the problem of analysis of the concentration of Pu in the spent fuel post-treatment process is solved, the use effect is good, and the popularization and the use are convenient.

The present invention discloses a method for adsorbing radioactive iodide ions in radioactive wastewater, which includes the following process: adding silver copper oxide to the radioactive wastewater, and the iodide ions in the radioactive wastewater react with the silver copper oxide, so that the concentration of iodide ions in the radioactive wastewater is reduced. The adsorption of radioactive iodide ions in radioactive wastewater by the silver copper oxide of the present invention belongs to chemical adsorption, which has the characteristics of fast adsorption speed, high adsorption efficiency and adsorption capacity, the adsorbed product is not easy to desorb, is easy to solidify and dispose, is convenient for recycling, is convenient to use and is not easily affected by interfering ions and pH. The silver copper oxide has a strong selective adsorption effect on iodide ions in the wastewater environment of long half-life radioactive element iodine, and can effectively improve the radioactive pollution of iodine in the water environment, and solve the main problems of spent fuel reprocessing and radioactive water pollution.

The invention belongs to the field of gas technical purification, and particularly relates to a honeycomb-shaped high-concentration NO2 adsorption trapping material as well as a preparation method and application thereof. The preparation method comprises the following steps: (1) preparing adsorbent slurry by taking high-silicon molecular sieve raw powder as a raw material; (2) putting a honeycomb carrier into the adsorbent slurry for dipping and coating; and (3) naturally airing, drying and roasting the impregnated carrier to obtain the integral honeycomb-shaped high-concentration NO2 adsorbing and trapping material. On the basis of the honeycomb-shaped high-concentration NO2 adsorption trapping material (the molecular sieve silica-alumina ratio is 1000-5000), NO2 is efficiently trapped through a pressure swing adsorption technology, meanwhile, the adsorbent subjected to adsorption saturation is reversibly regenerated, high-concentration NO2 desorbed gas is obtained, NO2 recycling is achieved, NOx tail gas of a nuclear industry post-treatment plant is recycled, and the purpose of recycling the NOx tail gas of the nuclear industry post-treatment plant is achieved. A stable NO2 gas source is provided for a spent fuel post-treatment main process, the dual requirements of NO2 raw material preparation and tail gas denitration of a post-treatment plant are met, and purification and recycling of NO2 are achieved.

The present application relates to the technical field of spent fuel reprocessing, in particular to a 1AF metering tank gas damper adjusting device and method, the 1AF metering tank gas damper adjusting device comprising a fixed rack and a plurality of stirring mechanisms, the fixed rack being arranged directly above the 1AF metering tank, and the stirring mechanisms being movably arranged on one side of the fixed rack close to the 1AF metering tank. Through the adjustment of the fixed rack and the stirring mechanisms, the horizontal position and the vertical position of the stirring mechanisms can be quickly adjusted, and then the stirring effect of the gas damper in the 1AF metering tank with different number and arrangement position of the gas damper can be tested, so that the number and position arrangement of the gas damper in the 1AF metering tank can be confirmed and fixed, and the type, number and position arrangement of the stirring device in the 1AF metering tank with 1:1 scale of the commercial spent fuel reprocessing plant for the accurate metering of the 1AF liquid volume are provided with reference, the traditional fixed welding method is avoided, and the manpower, material resources and financial resources for the 1AF metering tank gas damper adjustment test are saved.

The invention provides a gas flow simulation method and device for a dissolver chute. The method comprises the steps that structural parameters of the dissolver chute and a connecting device of a target spent fuel aftertreatment system are obtained; determining operation parameters of a dissolver chute and a connecting device of the target spent fuel post-processing system during operation; according to the structure parameters and the operation parameters, simulating the gas flow at the chute of the dissolver chute and the connecting device in the operation state to obtain a gas flow simulation result; determining an error of a gas flow simulation result; when the error is within a preset range, obtaining a simulation model; and according to the simulation model, performing simulation analysis on gas flow of the dissolver chute and the connecting device to obtain a gas flow simulation result. According to the scheme provided by the invention, the gas flow condition at the dissolver chute of the spent fuel post-processing system can be simulated, so that the structural parameters and operating parameters of the chute are optimized, and the operating stability and economical efficiency of the spent fuel post-processing head end are improved.

The invention discloses an equivalent amplification design method of an electrolytic bath for uranium electrolytic refining in a dry post-treatment process, and belongs to the field of spent fuel post-treatment. According to the method, the influence of a stirring device on uranium ion distribution in a molten salt electrolytic bath and the uranium electrolytic refining rate and yield is obtained through a simulation result, the optimal rotating speed of the stirring device is determined, and then the stirring Reynolds number and the end face speed are further determined as amplification criteria of the electrolytic bath; after the amplification design is carried out, the change condition of each amplified key parameter is estimated by virtue of simulation software, and meanwhile, the influence on uranium ion distribution in the molten salt electrolytic bath and the uranium electrolytic refining rate and yield under different amplification criteria is compared and analyzed, so that an optimal amplification criterion is obtained; and finally completing equivalent amplification design of the electrolytic bath for uranium electrolytic refining in the dry post-treatment process, thereby realizing amplification design of the molten salt electrolytic bath. By adopting the method disclosed by the invention, the design and amplification of the electrolytic cell can realize precision and high efficiency, and the time cost and the capital investment are saved.

The present invention relates to a method for separating trace plutonium from a large amount of uranium by using a solid phase leaching resin, and belongs to the field of spent fuel post-processing. A spent fuel post-processing simulation liquid containing uranium and plutonium is prepared, and the spent fuel post-processing simulation liquid is pumped into a leaching resin column at a first preset flow rate to perform a dynamic test, and the uranium and plutonium concentration in the aqueous phase at the outlet is analyzed at intervals to draw a penetration curve; an eluent is pumped into the leaching resin column after adsorption equilibrium at a second preset flow rate, and the eluent is used to elute plutonium. The method disclosed in the present invention is used to efficiently separate and purify trace plutonium in a large amount of uranium by using an extraction chromatography method having the characteristics of selectivity of solvent extraction and high efficiency of chromatographic separation, and a solid phase leaching resin with a wide range of acidity applications and high selectivity for Pu is used to achieve efficient separation of trace plutonium in a large amount of uranium, and the separation process does not require acid adjustment, and the operation is simple, the method is simple, the operation is simple, and the practicability is strong, which has a strong promotion advantage.

The present application belongs to the technical field of spent fuel post-processing, and specifically relates to a control method and device for the organic phase circulation of spent fuel post-processing. The method includes: obtaining the organic phase supply and demand relationship between various systems in the post-processing process; based on the organic phase supply and demand relationship, obtaining the organic phase flow rate and organic phase usage of each extraction supply and demand; based on the total volume of the input and output organic phase of each extraction process being equal, and the total conservation relationship of TBP and KE, determining the input and output calculation formula of each organic phase between each extraction process and the contaminated solvent distillation process; according to the input and output calculation formula, controlling the organic phase flow rate between various systems in the post-processing process. The present application can be directly used for spent fuel post-processing commissioning and production operations, realizing the quantitative replacement of organic phase in the extraction process and the inventory monitoring of the process system, and ensuring the continuous and stable operation of the process system under the organic phase circulation system.

The present invention belongs to the technical field of spent fuel reprocessing, and specifically relates to a pulse backwash filtration device suitable for radioactive liquid, comprising a radioactive liquid feed section provided with a radioactive liquid storage tank (1) and a backwash water feed section provided with a backwash water storage tank (2), the radioactive liquid feed section and the backwash water feed section being respectively connected to a flocculation tank (3), the radioactive liquid feed section being able to filter the radioactive liquid sent from the flocculation tank (3) through the radioactive liquid storage tank (1) and then send it out, and the backwash water feed section being able to backwash the radioactive liquid storage tank (1) through the backwash water in the backwash water storage tank (2). The entire set of equipment of the present invention has no mechanical moving parts, is not easily damaged, has high reliability, is convenient to clean and maintain, and is suitable for installation in a radioactive hot cell.

The utility model discloses an electrochemical test experimental device for boiling nitric acid corrosion in a high-temperature reduced-pressure environment, which relates to the technical field of electrochemical test and consists of a reduced-pressure boiling corrosion reaction system, an electrochemical test system, a heating system, an acid gas absorbing and drying system, a vacuum degree control system and an automatic integrated control system. The decompression boiling corrosion reaction system is connected with the acid gas absorption and drying system, the vacuum degree control system and the heating system are respectively connected with the acid gas absorption and drying system and are in contact with the decompression boiling corrosion reaction system, and an automatic integrated control system is adopted to control the experimental device. The experimental device disclosed by the utility model can test and simulate the electrochemical corrosion behavior of high-temperature reduced-pressure boiling nitric acid in the environment of the evaporator for spent fuel post-treatment, and can realize weight loss and electrochemical test of various materials under the condition of reduced-pressure boiling strong corrosive solution at the same time.

The invention relates to a method for improving hardness of corrosion-resistant austenitic stainless steel for post-treatment key equipment, belongs to the field of nuclear energy materials and technologies, and solves the key problem that the austenitic stainless steel has excellent intergranular corrosion resistance but insufficient hardness in a nitric acid solution in a spent fuel post-treatment environment. The preparation process of the austenitic stainless steel comprises the steps of burdening, vacuum induction smelting, vacuum consumable arc smelting, hot forging machining, hot rolling machining, solid solution heat treatment and cold rolling machining, after 10-40% of cold deformation machining is adopted for solid solution state steel, the hardness is increased to 230 HV0.2 or above, and the service index requirements of some key components are met; and intergranular corrosion does not occur in a 6mol / L after-treatment environment nitric acid solution at the temperature not higher than 95 DEG C, and the annual corrosion rate after soaking until the corrosion rate is stable is less than or equal to 0.30 mm / a. The method is mainly applied to the field of spent fuel post-treatment and can also be applied to the field of nitric acid chemical industry.

The application relates to a high-throughput testing device and method for simulating the crevice corrosion of spent fuel reprocessing boiling nitric acid, and belongs to the field of spent fuel reprocessing boiling nitric acid corrosion testing devices. The device combines continuously changing crevice widths with wire beam electrodes, so that each rod-shaped sample corresponds to different crevice widths, and the crevice corrosion electrochemical signals under different crevice widths are monitored at one time. Therefore, the device can capture the corrosion signals of a micro area, obtain the corrosion rules under different crevice widths, complete high-throughput electrochemical testing, integrate the signals captured by each micro electrode, and obtain the corrosion condition of the whole material. Meanwhile, polytetrafluoroethylene is used as raw material to manufacture a top wire, a bolt, a nut and a bottom support in the device, the device has good electric insulation performance and strong corrosion resistance, and can meet the electrochemical experiment under a high-temperature concentrated nitric acid environment for a long time. The simulation and research of the crevice corrosion under the boiling concentrated nitric acid environment are realized.

The application discloses a check valve and a pipeline system of a nuclear fuel reprocessing plant. The check valve comprises a valve body, a valve clack, an elastic rod and a control assembly. The valve body is installed in a pipeline and is internally provided with a fluid cavity. One end of the valve body is provided with an inlet, and the other end is provided with an outlet. The inlet faces the upstream of the valve body, and the outlet faces the downstream. The inlet and the outlet are both communicated with the fluid cavity. The valve clack is located in the fluid cavity of the valve body, and the upper end surface of the valve clack is attached to the inner wall of the upper end of the fluid cavity to block the inlet. One end of the elastic rod is rotationally connected with the valve body, and the other end is rotationally connected with the valve clack. The elastic rod drives the valve clack to move in the vertical direction through swinging. The control assembly is electrically connected with the elastic rod, is used for monitoring the rotation angle of the elastic rod, and judges the vertical position of the valve clack according to the rotation angle information of the elastic rod to obtain the opening degree information of the valve. The check valve has the function of remotely monitoring the opening degree of the valve.

The invention provides a blade point type small pipeline desilting centrifugal drill bit which comprises a drill bit body, a centrifugal rotating body and a torsion spring, the centrifugal rotating body is rotationally installed on a rotating shaft of the drill bit body, the torsion spring is installed on the drill bit body to provide torsion Ft to enable the centrifugal rotating body to be folded, and when the centrifugal drill bit rotates, the torsion Ft is driven by the torsion Ft to rotate. And under the action of centrifugal force, the centrifugal rotating body overcomes the torsion Ft to be opened. According to the small pipeline dredging device, different dredging diameters can be expanded according to the rotating speed after the small pipeline dredging device is shrunk and enters the inside through a variable-diameter bent pipeline, small pipeline dredging of pipelines with different diameters can be achieved, a power source and a driver are arranged outside the pipeline, and high-dose radiation can be resisted. The spent fuel post-treatment pipeline dredging device solves the problem of spent fuel post-treatment pipeline dredging, and guarantees passing ability and dredging thoroughness in narrow, variable-diameter, multi-turning and high-radiation pipelines.

The present disclosure relates to an analysis and detection apparatus and a spent fuel reprocessing system. The analysis and detection apparatus comprises: a base; a carrying platform movably arranged on the base in a first direction and a second direction perpendicular to each other in a horizontal plane; an accelerator assembly and a radiation source, wherein the radiation source is located on one side of the accelerator assembly in the second direction, and the radiation source is used for receiving radiations emitted by the accelerator assembly to generate particles; a measuring assembly fixedly mounted relative to the base and used for accommodating an object under detection to implement a detection mode; a calibration assembly detachably arranged and used for accommodating a calibration object to implement a calibration mode; a shielding body used for providing radiation protection for the radiation source, the measuring assembly, and the calibration assembly; and a detector assembly arranged on one side of the shielding body in the first direction; wherein the accelerator assembly, the shielding body, and the detector assembly are all arranged on the carrying platform, and the carrying platform is configured to drive the shielding body to reach or leave a protection position in the first direction.

The application discloses a method for reducing the content of plutonium and neptunium in a uranium product solution in a spent fuel reprocessing PUREX process, which comprises the following steps: adjusting the 2DF material liquid, washing the extraction organic phase by using a washing agent with complexing and reducing double effects in an extraction device, and reducing the content of neptunium and plutonium in the uranium-containing organic phase, so as to obtain a uranium product solution with greatly reduced content of plutonium and neptunium through back extraction. 2 The method has the advantages of simple process, and the decontamination factor of plutonium and neptunium is 10 4 orders of magnitude after 2D tank extraction and washing, and the content of plutonium and neptunium in the uranium product solution is greatly reduced after only one extraction cycle of the uranium line, so that the requirements for the operation control precision and the solvent quality of the process are reduced, the probability of unqualified uranium product solution due to the excessive content of plutonium and neptunium is greatly reduced, the quality of the uranium product is improved, the method has a remarkable removal effect on the fission products such as zirconium, niobium and ruthenium in the uranium product, and can be effectively applied to the PUREX process.

The embodiment of the present application relates to the technical field of spent fuel reprocessing, and specifically discloses a shearing device for spent fuel assemblies. The shearing device comprises a box body, a fixed blade assembly, and a pressing assembly. The box body is provided with a feeding port through which the spent fuel assemblies are pushed into the box body. The fixed blade assembly is fixed inside the box body and is matched with the spent fuel assemblies to bear the spent fuel assemblies. The pressing assembly is movably arranged inside the box body and is correspondingly arranged on the two sides of the feeding port with the fixed blade assembly. The pressing assembly is movable relative to the fixed blade assembly and is used to press the spent fuel assemblies against the fixed blade assembly. The movable blade assembly is movably arranged inside the box body and is parallel to the fixed blade assembly. When the movable blade assembly moves towards the fixed blade assembly, the movable blade assembly cooperates with the fixed blade assembly to shear the spent fuel assemblies.

The invention discloses a liquid temperature control cyclic irradiation test device and method, belongs to the technical field of nuclear power spent fuel post-treatment, and aims to solve the problems of how to improve irradiation uniformity and effectively control the temperature of a sample in the irradiation process. The liquid temperature control circulating irradiation test device comprises an irradiation container, a thermostat, a buffer bottle and liquid conveying equipment, the buffer bottle is arranged in a temperature control groove of the thermostat; and the irradiation container is circularly connected with the buffer bottle through a pipeline to form a liquid circulation system. Liquid can be driven to circularly flow in the liquid circulating system through the liquid conveying equipment, and the liquid circulating system is used in the irradiation test process of an exogenous method research on a sample, the circularly flowing sample can be fully and mechanically mixed uniformly, the irradiation uniformity is improved, and the temperature difference of each part of the sample can be ensured to be relatively small. By arranging the thermostat and the buffer bottle, heat exchange can be carried out between the thermostat and liquid in the buffer bottle, so that the temperature of a sample in the liquid circulation system is effectively controlled.

The application discloses a kind of nuclear spent fuel reprocessing equipment magnetic powder sealing device and method, including shell, first pole shoe, second pole shoe, magnetic component and magnetic conducting component, two end surfaces of magnetic conducting component in the axial direction of inner cavity form sealing gap between first pole shoe and second pole shoe respectively.When injecting magnetic powder into sealing gap, make the axial direction of inner cavity parallel with the direction of gravity, and inner cavity opening is upward.Assemble, first pole shoe is assembled in the inner cavity of shell first, the pole tooth of first pole shoe is upward, magnetic powder is injected to the pole tooth of first pole shoe, then magnetic component and magnetic conducting component are assembled in inner cavity, and magnetic powder is injected above magnetic conducting component.After second pole shoe and end cover are assembled in the inner cavity of shell, magnetic powder is uniformly distributed by rotating magnetic conducting component driven by rotating shaft.In the process of assembling, magnetic component is not magnetized or weakly magnetized, finally, magnetic component is energized or magnetized, so that the uniformity of magnetic powder injection and the sealing performance of device can be improved.

The invention discloses a method for denitration of a kettle type evaporator in spent fuel post-treatment. The method comprises the following steps: S1, medium radioactive waste liquid is conveyed into the kettle type evaporator through a medium radioactive waste liquid feeding pipeline; s2, heating through a heating coil outside the kettle type evaporator; s3, medium radioactive waste liquid is continuously conveyed into the kettle type evaporator through a medium radioactive waste liquid feeding pipeline; s4, observing the change of a thermometer and the change of a densimeter, and when the density of the lower part of the solution reaches 1.112 g / cm < 3 > or the temperature reaches 105 DEG C, adding a formaldehyde solution into the kettle type evaporator through a formaldehyde pipeline; s5, when the density of the lower part of the solution in the kettle type evaporator reaches 1.125 g / cm < 3 > or the temperature reaches 110 DEG C, stopping adding the radioactive waste liquid, and only adding the formaldehyde solution; s6, when the density of the lower part of the solution in the kettle type evaporator is reduced to 1.112-1.125 g / cm < 3 >, or the temperature is reduced to 103-105 DEG C, repeating the step S4; and when the density of the lower part of the solution is reduced to 1.08-1.112 g / cm < 3 >, or the temperature is reduced to 97-99, repeating the step S3.

The invention relates to the technical field of spent fuel post-treatment, and discloses a method for adsorbing and separating Th (IV) from an acidic aqueous solution and an adsorbent, a composite material of a nitrogen-containing soft ligand derivative R-POPhen and macroporous resin is used as the adsorbent, and the Th (IV) is adsorbed and separated from the acidic aqueous solution; the structure of R-POPhen is shown in the specification, the adsorbent is formed by compounding a nitrogen-containing ligand of phenanthroline phosphorus oxide and macroporous resin, adsorption separation of thorium can be achieved under high acidity, and the adsorption efficiency can still reach 75% or above under the acid concentration of 4 mol / L. # imgabs0 #

The utility model relates to the field of spent fuel post-treatment corrosion electrochemistry, in particular to an electrochemical corrosion sample used in a severe environment. The sample is a cylinder with one end being a hemisphere, a hole is formed in the center of the flat end face of the sample, an internal thread I is arranged in the hole, the side, where the internal thread I is located, of the sample is fastened to one end of the connecting screw through threads, and a gasket I for enhancing sealing performance is arranged at the joint of the sample and the connecting screw. The other end of the connecting screw rod is fastened with the bolt through threads, and a gasket II for enhancing the sealing performance is arranged at the joint of the connecting screw rod and the bolt; and a glass tube is sleeved outside the connecting screw rod, one end of the glass tube is tightly pressed on the gasket I, and the other end of the glass tube is tightly pressed on the gasket II. The device can be suitable for preparing an electrochemical sample which needs to work in a severe environment, the electrochemical corrosion behavior of a material in the severe environment is researched, and the situation that the sample prepared by the existing electrochemical sample preparation method is damaged by an electrolyte solution can be effectively avoided.

The application discloses a method for determining the concentration of hydroxylamine nitrate, preparing verification set and correction set samples in the concentration range of spent fuel reprocessing process, collecting near-infrared absorption spectra of the verification set and the correction set, establishing a quantitative correction model and verifying to obtain a quantitative model, performing near-infrared spectrum detection on a solution of the spent fuel reprocessing process, not needing to pretreat the solution, and simultaneously obtaining the concentration of hydroxylamine nitrate by processing the absorption spectra of the corresponding wavelengths by the quantitative model; the method is simple, improves the analysis efficiency, and is high in accuracy; and the method provides effective data support for online detection of the concentration of hydroxylamine nitrate.

The embodiment of the invention discloses a uranium concentration detection system in a nuclear fuel post-treatment process. The uranium concentration detection system in the nuclear fuel post-treatment process comprises a buffer device and a detection device. The buffer device comprises a buffer container and a liquid inlet pipe, the buffer container is communicated to the process pipeline through the liquid inlet pipe, and the buffer container is used for buffering a to-be-detected solution from the process pipeline. The detection device comprises a sample container, a first connecting pipe and a detector, the sample container is communicated with the buffer container through the first connecting pipe, the sample container is used for containing a to-be-detected solution from the buffer container, and the detector is used for detecting the uranium concentration in the to-be-detected solution in the sample container. According to the detection system provided by the embodiment of the invention, the liquid flow stability of the to-be-detected solution flowing into the sample container is improved, so that the problem of unstable liquid flow is solved, then the uranium concentration of the to-be-detected solution in the sample container is detected through the detector, and the accuracy of the detection result of the uranium concentration can be improved to a certain extent.