95 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 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 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 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 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 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 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 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 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 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.

The application discloses a high-strength, high-corrosion-resistance and low-cost titanium alloy suitable for spent fuel reprocessing and a preparation method thereof, and belongs to the field of structural materials for spent fuel reprocessing. The chemical composition of the titanium alloy comprises the following components in percentage by mass: Al 4%, Zr 3%, V 0.5%-5.0%, H <0.008%, O <0.08%, N <0.04%, C <0.08%, and the balance Ti. The preparation method comprises the following steps: raw material drying and preheating, vacuum / argon protection smelting, overturning remelting, casting and hot forging. The forging breakdown temperature is 1000 DEG C, the final forging temperature is greater than or equal to 880 DEG C, the cumulative deformation is 40+ / -5%, the strain rate is 3 s ‑1 ~8 s ‑1 . The alloy has excellent corrosion rate and high yield strength in 6 mol / L HNO3 containing oxidizing ions at 100 DEG C, and is suitable for corrosion-resistant load-bearing components of spent fuel reprocessing equipment.

The application discloses an I-129 and Kr-85 online continuous monitoring system and method, relates to the technical field of spent fuel reprocessing radiation monitoring, and comprises a Kr-85 sampling detection device, an input of a first I-129 sampling detection device, and an output of a Kr-85 sampling detection device; a second I-129 sampling detection device, an output of a Kr-85 sampling detection device, and an input of a flushing interface; and a flushing interface, an output of a first I-129 sampling detection device and a second I-129 sampling detection device, and an input of an electrical control box. The application controls the on-off state of solenoid valves YV2-YV5 through the electrical control box. The application can remove Kr-85 in sampling gas through air flushing and differential measurement, and can realize I-129 and Kr-85 online continuous measurement when the concentration difference between Kr-85 and I-129 is more than 10 9 times.

The utility model belongs to the technical field of equipment heating, and particularly relates to a device suitable for pulse extraction column plate section heating. The device comprises a plurality of heating pipelines arranged in parallel; the heating pipeline is arranged around the plate section of the 2BX column and is parallel to the plate section; a preset distance is formed between the heating pipeline and the outer wall of the plate section; and hot water with a preset temperature is introduced into the heating pipeline. Heat transfer is used in a heating mode, water is used as a medium, cost is low, recycling can be achieved, damage is not prone to occurring, maintenance is avoided, radiation injury to personnel is greatly reduced, and the problem that the temperature of 2BX pulse column plate section feed liquid of a spent fuel reprocessing plant cannot meet the design requirement is solved.

The invention relates to the technical field of spent fuel aftertreatment process flow monitoring, and discloses a spent fuel dissolution degree monitoring system and method.The system comprises a radiation collection module, an automatic control module, a signal processing module and a process master control module, and the process master control module is used for collecting state signals of a loader; the signal processing module is used for controlling the automatic control module to switch the shielding thickness in the radiation acquisition module on the basis of the state signal of the loader and in combination with external preset measurement parameters, and then calculating the solubility of the spent fuel through energy spectrum analysis on the basis of the electric signals output by the radiation acquisition module under different shielding thicknesses. The invention provides a measuring system for monitoring the content of undissolved substances in the technical field of spent fuel post-treatment process equipment operation monitoring. The measuring system can specifically detect the content of the undissolved substances in spent fuel.

The application provides a spent fuel reprocessing high-temperature nitric acid full stress corrosion simulation experiment device, and relates to the technical field of spent fuel reprocessing. The device comprises an electric control system and a load loading system, a corrosion reaction system, an electrochemical test system, a temperature control system and a tail gas treatment system connected with the electric control system. The load loading system is used for loading various loads. The corrosion reaction system is composed of a glass reaction tank and a reaction tank cover. The reaction tank contains nitric acid solution for corrosion reaction of the sample. The sample penetrates the glass reaction tank and is sealed by a sealing kit. The electrochemical test system is arranged in the reaction tank cover and used for electrochemical test of the sample corrosion reaction. The temperature control system is used for adjusting and controlling the temperature of the sample corrosion reaction of the load loading system. The tail gas treatment system realizes treatment of the tail gas generated in the sample corrosion process. The device solves the problem that the stress corrosion test under high-temperature nitric acid cannot be carried out at present.

The invention discloses a spent fuel post-processing facility through-ground valve structure evaluation method, which is characterized in that through-ground valve modal analysis is carried out through a finite element analysis method, and an all-parametric modeling-multi-modal analysis-dynamic stress path evaluation integrated method is provided. According to the method, the dynamic coupling response of a plurality of key components is covered, the stress limit value is determined, accurate quantitative evaluation is achieved for high-risk areas such as reducing, welding seams and pressure-bearing interfaces, finally, the finite element calculation result is compared with the stress limit value according to the selected calculation evaluation path, and whether the valve structure meets the requirement or not can be judged. The method can effectively evaluate the structural characteristics of the through-ground valve, and compared with experimental research, numerical simulation greatly reduces the economic cost and shortens the structural evaluation period.

The invention provides a compound as well as a preparation method and application thereof. The compound is represented by the following formula (I) wherein R is a C1 to C14 alkyl group. The compound can efficiently extract tetravalent plutonium in a spent fuel post-treatment process, and is wide in applicable nitric acid concentration, fast in extraction kinetics and short in extraction equilibrium time.

The present application relates to the nuclear technology field, specifically to a method for separating thorium from a thorium and lanthanum mixture. The method comprises the following steps: in a three-electrode system, pulse constant current controlled potential electrolysis is performed on a molten salt system containing mixed ions of thorium and lanthanum series elements, and the deposition rate difference of thorium and lanthanum series element ions during pulse constant current electrolysis and the electrolytic reaction limitation of the control potential on the lanthanum series element ions are utilized to achieve efficient separation of thorium from the thorium and lanthanum mixture. The molten salt system containing mixed ions of thorium and lanthanum series elements is obtained by adding the thorium and lanthanum mixture into a chloride molten salt, mixing uniformly, and passing in chlorine gas mixture to chlorinate the mixture and dissolve it into the chloride molten salt. The method has the advantages of simple process, strong radiation resistance, no additional radioactive waste, and low cost, and is suitable for industrial application of molten salt reactor spent fuel reprocessing.

This invention provides a simulation method for the generation of organic iodine based on the Kolbe electrolysis reaction, belonging to the field of spent fuel reprocessing technology. This method introduces iodine into the Kolbe electrolysis reaction and compares the intrinsic ability of different forms of iodine to react with alkyl radicals to generate organic iodine. Compared with traditional simulation experiments under irradiation conditions, the method of this invention can conveniently obtain alkyl radicals through electrochemical reactions, eliminates the interference of water radiolysis products in the irradiation environment, and provides a clear understanding of the organic iodine conversion capabilities of different forms of iodine; moreover, it does not require specific irradiation equipment and shielding measures, facilitating real-time control and sampling analysis of the reaction, and has better simplicity and operability.

The invention discloses a method for solidifying radionuclide technetium in a solution and application of the method, and belongs to the technical field of spent fuel post-treatment. The curing method comprises the following steps: firstly, acidifying a technetium-containing solution by using nitric acid, then adding the acidified technetium-containing solution into a sodium metaaluminate solution, and reacting under a low-temperature condition to ensure that technetium is covalently bonded with high-activity mu < 1->-OH in a curing product in a form of pertechnetate (TcO4 <->) and is cured in pseudo-boehmite-like mixed crystals with stable crystal structures; the curing method is mild in operation condition and simple in technological process, overcomes the defects that in a traditional cement curing and glass curing method, the size of a cured product is increased, technetium is prone to volatilization and the like, has the advantages of being small in cured product size, high in curing efficiency, low in cost and environmentally friendly, the maximum curing amount can reach 100 mg / g, and the curing method is suitable for industrial production. The method is suitable for treating technetium-containing acidic nucleus waste liquid generated in the spent fuel post-treatment process.

The invention belongs to the field of structural materials for spent fuel aftertreatment, and particularly relates to a high-temperature-nitric-acid-corrosion-resistant titanium alloy for spent fuel aftertreatment and a preparation method of the high-temperature-nitric-acid-corrosion-resistant titanium alloy. The titanium alloy comprises, by mass, 1.0%-5.5% of Al, 1.5%-6.0% of Zr and the balance Ti and inevitable impurities, and the mass ratio of Al to Zr is larger than or equal to 0.8 and smaller than or equal to 1.8. The preparation method comprises the steps of vacuum melting and casting, forging and two-stage annealing treatment (in the first stage, the temperature of 750 + / -10 DEG C is kept for 50-70 minutes; and in the second stage, heat preservation is conducted for 4-6 hours at the temperature of 480 + / -20 DEG C. The prepared titanium alloy shows excellent corrosion resistance in a high-temperature nitric acid environment (for example, boiling 3 mol / L HNO3 contains multiple ions), and the average corrosion rate can be as low as 0.0023 mm / y. The alloy is remarkably reduced in cost and suitable for resisting high-temperature nitric acid corrosion in the field of spent fuel post-treatment.

The invention belongs to the technical field of spent fuel post-processing, and particularly relates to an irradiation-resistant control system and delay reduction control method for a pipe wall cleaning soft robot. Comprising a man-machine interface layer, a communication layer and a function layer, the man-machine interface layer provides a final man-machine interaction interface, and final man-machine interaction control is achieved; the communication layer comprises main control equipment and various data acquisition modules, is connected with various functional modules of the upper computer and the robot, and realizes real-time communication and control between the upper computer and the robot; the functional layer comprises various motion states and working modes of the robot body and is directly controlled by the main control equipment. The method has the beneficial effect that stable motion control over the robot can be achieved in the high-radiation complex pipeline environment. The problem that the pneumatic soft robot is difficult to realize stable control or even cannot be controlled in an extreme environment is solved.

The invention provides a filtering device for spent fuel post-treatment and a dismounting and mounting method. The filtering device comprises a cylindrical filter, the filter cup is arranged in the cylindrical filter; the sealing assembly is arranged between the cylindrical filter and the filter cup; the seal assembly includes: an annular body; the first boss is arranged on the first side of the annular body; the second boss is arranged on the second side of the annular body; the diameter of the first boss is smaller than that of the annular body, a convex face is formed between the annular body and the first boss, and the annular body, the first boss, the second boss and the convex face are integrally formed. The diameter of the second boss is smaller than that of the first boss. The sealing assembly of the filtering device is long in service life, an operator can conveniently disassemble and assemble the sealing assembly at a long distance through an auxiliary tool when the sealing assembly is damaged, the frequency that the operator enters a radiation area for operation is reduced, and the exposure dose level of the operator is reduced.

The invention discloses a method and a system for on-site volume reduction of spent fuel of a high-temperature gas cooled reactor. The method comprises the following steps: heating spent fuel graphite nodules under oxygen-containing flow to remove a graphite layer to obtain TRSIO particles and generated tail gas; the TRSIO particles are discharged and then loaded into a container with a shield to achieve spent fuel volume reduction; generated tail gas is cooled and then is discharged after being subjected to filtering, molecular sieve adsorption and efficient filtering in sequence. According to the volume reduction method, the volume of the high-temperature reactor spent fuel is reduced by about 95%, the transportation cost of the high-temperature reactor spent fuel is reduced by about 50% or above, the volume reduction method can be used for removing a high-temperature reactor spent fuel post-treatment graphite layer and treating high-temperature gas cooled reactor decommissioning structure graphite, and radioactive substance minimization is achieved.