909results about "Radioactive decontamination" patented technology

The embodiment of the invention relates to the field of radioactive waste treatment, in particular to a sealing device and curing equipment. According to the sealing device and the curing equipment provided by the embodiment of the invention, the blocking assembly and the sealing assembly jointly block the first opening formed by the material treatment device, so that the material treatment device is subjected to double sealing, and the internal material treatment environment is more stable; the locking assembly is driven to move in the circumferential direction so as to fix the plugging assembly to the sealing assembly, meanwhile, the locking assembly is arranged to be axially static so as to further reinforce fixation between the plugging assembly and the sealing assembly, and the material containing device can be more stable in the material treatment device; and good sealing performance and stability in the material treatment process are ensured.

The embodiment of the invention relates to the field of radioactive waste treatment, in particular to a material containing device and curing equipment. The first containing piece and the second containing piece of the material containing device are arranged to be in a detachable and separable form, so that powdery radioactive materials cannot escape from a seam, and meanwhile, after the powdery materials are pressed into solid materials, the first containing piece and the second containing piece can be separated; the solid materials contained in the first containing part can be independently transferred, the material containing device is prevented from being integrally transferred when the solid materials are transferred, and the transferring efficiency of the solid materials is improved. According to the curing equipment, the material containing device is adopted as a material conveying carrier, materials in different stages and states in the curing process can be contained, and leakage of the materials is prevented; meanwhile, in the whole curing process, form transformation of the materials can be achieved without replacing the containing device, the operation process is simplified, material spill caused by replacing the containing device is avoided, and therefore the curing efficiency is improved.

The embodiment of the invention relates to the field of radioactive waste treatment, in particular to material treatment equipment and curing equipment. According to the material treatment equipment provided by the embodiment of the invention, the powdery material in the material accommodating device is converted into a partially molten state through the material treatment device, and the partially molten material is converted into the solid material through the pressurizing device, so that the powdery material can undergo conversion of different states in the material accommodating device; therefore, powdery materials are converted into solid materials convenient to transfer and process, the operation of replacing a container is omitted, and material processing procedures are saved. According to the curing equipment provided by the embodiment of the invention, the sealing device is used for sealing the material treatment equipment, so that the state transition process of the powdery material has better sealing performance and stability, and the stable quality of the produced solid material is ensured.

The embodiment of the invention relates to the field of radioactive waste treatment, in particular to a solidification method. According to the method, a conveying and jacking device is used for conveying a material containing device to enable the material containing device to receive powdery materials, a clamping and transferring device and a lifting device are used for conveying the material containing device completing feeding into a material processing device to enable the powdery materials to be partially molten, and a pressurizing device is used for pressing and compacting the partially molten materials; and finally, ingot-shaped solid materials are formed in the material containing device, then the lifting device and the clamping and transferring device are utilized again to transfer the material containing device to the conveying and jacking device so as to transfer the ingot-shaped solid materials, in this way, the process from powder feeding to ingot forming of the materials is conducted in the material containing device, the situation that a material container is replaced in the midway is avoided, and the production efficiency is improved. Operation steps are simplified; and the clamping and transferring device, the lifting device and the conveying and jacking device all participate in the front process and the rear process of material treatment, so that reuse of equipment parts is achieved, and the equipment composition is simplified.

The invention relates to an adsorbent and a preparation method and application thereof, the adsorbent comprises a polymer matrix and a nanofiller uniformly embedded in the polymer matrix, the polymer matrix comprises a calcium / boron cross-linked polyvinyl alcohol-sodium alginate blend, and the nanofiller is sulfonic acid / carboxylic acid functionalized graphene oxide; the mass fraction of the nano filler in the adsorbent is 0.05%-0.60%; wrinkles and irregular hole structures are distributed on the surface of the adsorbent. The adsorbent is used for removing tritium in water and is high in adsorption capacity, stable in structure and good in recycling performance. The removal rate of tritium in the single-stage adsorption process is greater than 8.5%, and the removal rate of tritium in the third-stage adsorption process is 19.4-57.7%.

The invention relates to a waste liquid treatment device for automatically separating radionuclides. The waste liquid treatment device comprises a first waste liquid collection passage and a second waste liquid collection passage, an upstream liquid supply channel shared by the two channels comprises an injection pump and a multi-way valve, and the injection pump inputs liquid into the multi-way valve. The first waste liquid collecting channel directly guides cleaning liquid, replacement liquid and the like which do not need to be reserved into a waste liquid barrel through a first output port of the multi-way valve. The second waste liquid collecting channel is sequentially connected with a plurality of resin columns, a waste liquid tank, a peristaltic pump and a waste liquid barrel through a second output port of the multi-way valve, waste liquid generated by the resin columns firstly enters the waste liquid tank for temporary storage, and when the liquid level of the waste liquid tank reaches a preset threshold value, the peristaltic pump automatically pumps and discharges the waste liquid into the waste liquid barrel for centralized collection. The device can distinguish reserved liquid from waste liquid according to steps, realizes automatic flow division, temporary storage and discharge of the waste liquid, avoids manual pouring of the waste liquid halfway, ensures continuous operation of a long-time and multi-channel separation process, and reduces the risk of personnel exposure in a radioactive waste liquid treatment process.

To provide a technique capable of contributing to environmental load reduction while securing performance as a water purification filter.SOLUTION: The water purification filter 11 satisfies at least one of the conditions that the increase in CO2 on the ground surface due to CO2 released at the time of disposal is not more than 240L per kilogram and that the use rate of petroleum-derived raw materials is not more than 15%.SELECTED DRAWING: Figure 3

The embodiment of the invention relates to the field of radioactive waste treatment, in particular to curing equipment. According to the curing equipment provided by the embodiment of the invention, the powdery materials are conveyed into the material containing device through the vacuum feeding device, and the material containing device is transferred to the lifting device through the clamping and transferring device, so that the material containing device is conveyed into the material processing device through the lifting device; the sealing device is used for sealing the material treatment device, so that the material reaction process is carried out in an environment with relatively good sealing performance; the pressurizing device is used for compacting the materials so as to obtain solid materials which are easier to transfer and process compared with powdery materials; the vacuum feeding device, the conveying and jacking device, the clamping and transferring device, the lifting device, the material treatment device and the pressurizing device are fixed to different positions of the body, enough space is provided for all treatment links of materials, the operation sequence of curing treatment is met, the process is simplified, and the treatment efficiency is improved.

The invention relates to the technical field of pipeline decontamination, in particular to an online pipeline ultrasonic decontamination device and method suitable for various pipe diameters, and the online pipeline ultrasonic decontamination device comprises a numerical control generator, a transducer, an amplitude-change pole, an adaptive clamp, an online monitoring device and other key components. And through the designed adaptive clamp, the device can act on pipelines with various pipe diameters and can move along the pipelines, and the decontamination requirements in different scenes are met. The online ultrasonic decontamination device is suitable for maintenance and cleaning work of pipelines with different pipe diameters in nuclear power facilities, and has important significance in improving the safety of the nuclear facilities and guaranteeing radiation protection for operators. The problem that an ultrasonic device is poor in applicability is effectively solved, meanwhile, the flexibility of the device is guaranteed, effective online decontamination of pipelines with various pipe diameters can be achieved through a single set of device, the decontamination cost of radioactive contamination of nuclear facility pipelines is effectively reduced, online efficient decontamination of radioactive contamination deposited in the pipelines is achieved, and the service life of the pipelines is prolonged. The safe and stable operation of the reactor is ensured, and the radiation risk of operators is reduced.

To provide a radioactive waste discharge plan preparation support system capable of efficiently processing and managing radioactive waste.SOLUTION: A radioactive waste discharge planning support system includes a storage unit that stores information on an amount of radioactivity of radioactive waste and a measurement date of the amount of radioactivity, and a discharge planning unit that predicts a time when the amount of radioactivity becomes equal to or less than a specified value based on the information on the amount of radioactivity and the measurement date of the amount of radioactivity stored in the storage unit for the radioactive waste having the amount of radioactivity equal to or more than the specified value, and creates a discharge plan of the radioactive waste based on the predicted time.SELECTED DRAWING: Figure 1

The invention discloses a mixing and clarifying tank, a liquid level adjusting method thereof and a nuclide recovery and extraction system, which can adjust the liquid level of a two-phase interface in an overflow chamber by adjusting the air pressure difference between the overflow chamber and a clarifying chamber through an air pressure adjusting assembly, thereby avoiding the increase of maintenance workload of the mixing and clarifying tank due to the increase of mechanical moving parts. The mixing and clarifying tank comprises a clarifying tank body and a liquid level adjusting device. A sealing top cover is arranged at the top of the clarifying tank body, and the clarifying tank body comprises a mixing chamber, a clarifying chamber and an overflow chamber which are sequentially communicated from front to back; the overflow chamber is defined by a front heavy-phase flow baffle, a rear sealing baffle and a sealing top cover. The liquid level adjusting device comprises an air pressure adjusting assembly. The air pressure adjusting assembly is arranged in the overflow chamber and used for adjusting the air pressure difference between the overflow chamber and the clarifying chamber so as to adjust the liquid level difference between the highest liquid level in the overflow chamber and the liquid level of the two-phase interface in the overflow chamber, and then the liquid level of the two-phase interface in the overflow chamber is adjusted to the target liquid level.

The invention relates to the technical field of radionuclide separation, in particular to an Eu (III) / An (III) separation method based on a zirconium phosphonate-based metal organic framework adsorbent, which comprises the following steps: adding the zirconium phosphonate-based metal organic framework adsorbent into an Eu (III) and / or An (III) nuclide-containing aqueous solution with the acid concentration of 0.001-3M to carry out adsorption reaction, and then carrying out solid-liquid separation, wherein the zirconium phosphonate-based metal organic framework adsorbent is a zirconium phosphonate-based metal organic framework material exchanged by metal cations M; the metal cation M is selected from one of alkali metal ions, alkaline earth metal ions or divalent transition metal ions. According to the invention, the interlayer chemical properties of the material are regulated and controlled through ion exchange, so that the adsorbent has ultrahigh selectivity, rapid adsorption kinetics, high adsorption capacity and excellent structural stability on trivalent actinide / simulated nuclides in extremely high-acid, high-salt and intense-radiation environments; the key problems of low separation efficiency and poor stability in a high-acid complex system in the prior art are solved, and the method is suitable for treating high-level liquid waste.

The invention discloses a pyrochlore structure high-entropy ceramic solidified body material as well as a preparation method and application thereof, and belongs to the technical field of nuclear waste solidification. The preparation method of the high-entropy ceramic solidified body material with the pyrochlore structure comprises the following steps: weighing raw material powder of La2O3, Nd2O3, Sm2O3, Gd2O3, Eu2O3 and ZrO2 according to the proportion in (La0. 2Nd0. 2Sm0. 2Gd0. 2Eu0. 2) 2Zr2O7; the weighed raw material powder is subjected to ball milling treatment, and ball-milled powder is obtained; under the inert gas atmosphere, the ball-milled powder is subjected to heat treatment, and heat-treated powder is obtained; and carrying out spark plasma sintering on the heat treatment powder to obtain the pyrochlore structure high-entropy ceramic solidified body material. The powder is treated by using a heat treatment technology, so that the powder is homogenized, a pyrochlore phase is formed, grain growth is more favorably inhibited, a fine grain structure is obtained, and the irradiation stability and the mechanical property of the pyrochlore material are improved.

The invention provides a rubidium and cesium adsorbent and a preparation method and application thereof. The preparation method comprises the following steps: providing a doped and modified Prussian blue analogue, wherein the Prussian blue analogue comprises a Prussian blue matrix material and a transition metal doping element doped in the Prussian blue matrix material; granulating a mixed material containing the doped and modified prussian blue analogue and a granulating agent to obtain a rubidium and cesium adsorbent precursor; a mixed reaction system containing the rubidium and cesium adsorbent precursor and a surface modifier are subjected to a reaction, the surface modifier contains one or more functional structures of an amino group, a sulfonic group, a carboxyl group, a sulfydryl group, an acryloyloxy group, an epoxy group, a crown ether structure and a calixarene structure, and therefore the rubidium and cesium adsorbent is obtained. Through a multi-aspect synergistic strategy of transition metal doping, granulation and surface modification, the prepared rubidium and cesium adsorbent has good stability and high adsorption capacity, and has strong selectivity to Cs < + > / Rb < + >.

The invention discloses functionalized graphene for removing tritium from radioactive waste liquid and a preparation method and application of a graphene film, the functionalized graphene is a flaky substance formed by stacking lamellar structures, and the distance between the stacked lamellar structures is 7.5-8.7. The preparation method comprises the following steps: preparing graphene oxide powder; performing sulfonation on the graphene oxide powder to obtain sulfonated graphene aerogel; performing carboxylation on the graphene oxide powder to obtain carboxylated graphene; and mixing and stirring the sulfonated graphene aerogel and carboxylated graphene to obtain the functionalized graphene. The tritium in the tritium-containing liquid is separated by utilizing the membrane formed by the functionalized graphene, and the tritium removal technology has the advantages of mild reaction conditions, low energy consumption, easiness in large-scale production and the like.

A waste extraction system that includes a column set positioned between and fluidly coupled to an upstream segment of a main waste pathway and a column effluent tank, the column set having a uranium adsorption column fluidly coupled to a supporting adsorption column along the main waste pathway. The uranium adsorption column is upstream the supporting adsorption column. The waste extraction system also includes a strip waste pathway extending from the uranium adsorption column to the column effluent tank bypassing the supporting adsorption column, an anion exchange resin housed in the uranium adsorption column, and a cation exchange resin housed in the supporting adsorption column.

The invention discloses a radioactive waste graphite incineration treatment and tail gas purification and sealing treatment process, and relates to the field of radioactive waste treatment. Flue gas generated by incineration treatment is cooled and purified; according to the present invention, the carbon dioxide in the cooled and purified flue gas is concentrated and compressed, and the compressed carbon dioxide is subjected to geological storage so as to achieve the permanent fixation of the 14C, such that the problem that the existing radioactive waste graphite consumes the excessive stock can be solved, the generated secondary radioactive waste is less, and the emission problem of the 14C is avoided through the utilization of the new process.

A remote restart system for a vitrification facility and an operating method thereof are provided. The operating method of the remote restart system for a vitrification facility includes a step of operating the vitrification facility; and a step of causing a power loss to the vitrification facility, the power loss occurring during or before the operation of the vitrification facility, and the remote restart system assists in restarting the vitrification facility when the power loss occurs to the vitrification facility, and the remote restart system includes a low-temperature melting furnace, a glass frit provided inside the low-temperature melting furnace, a titanium ring for heating the glass frit inside the low-temperature melting furnace, and a bar-shaped handling unit that is inserted inside the low-temperature melting furnace and handles the glass frit in a state suitable for reigniting the glass frit.

The present application relates to a kind of tritium separation method of tritium-containing wastewater and its reactor system, application, the tritium separation method of tritium-containing wastewater of the present application is ultraviolet irradiation promotes separation method, comprising: step 1) tritium-containing wastewater is introduced into the reaction containment of reactor system and is contacted with adsorption material to form liquid-solid two-phase contact system, step 2) after ultraviolet light is irradiated to liquid-solid two-phase contact system, make tritium enrich in adsorption material;Step 3) after the liquid-solid two-phase contact system of irradiation is completed, solid-liquid separation is obtained to low concentration tritium treatment water and tritium-rich adsorption material.The method of the present application is irradiated by ultraviolet light, accelerates hydrogen / tritium isotope exchange and adsorption process, can obtain higher tritium removal rate and adsorption amount under the same processing time.The present application belongs to energy-saving and environmental protection industry, especially in wastewater environmental control and nuclear pollution wastewater treatment in the environmental protection application of the recovery treatment of nuclear radioactive pollution elements plays a key role.

This invention discloses a method for preparing a detergency-removing composite solution and its gel membrane, as well as its applications. Based on the host-guest inclusion mechanism of cyclodextrin, its hydrophobic cavity is achieved by "encapsulating" hydrophobic molecules or their hydrophobic portions. The volume of the hydrophobic portion / functional group of the complexing agent molecule is smaller than the volume of the cyclodextrin cavity. The complexing agent forms a complex with cyclodextrin, thus dispersing well in the solution. Furthermore, complexing agents with good coordination ability for radioactive ions can almost always form inclusion complexes with cyclodextrin. The polysaccharide molecular chains used in this invention are rich in hydrophilic functional groups such as hydroxyl and amino groups, giving the gel membrane good hydrophilicity, low ion transport resistance, and short equilibrium time. The hydroxyl and amino functional groups also participate in the coordination process of radioactive ions, further improving the adsorption capacity of the gel membrane. The method and application of this invention have strong universality, can simultaneously remove multiple radioactive ions, and can be used on a large scale.

Methods for processing metal components from the nuclear industry are described. In some embodiments, a method includes processing a component having a surface contaminated with radionuclides by dissolving at least a portion of the surface with an acid solution to produce a waste solution including the radionuclides, and separating the radionuclides from the waste solution. In some embodiments, a method includes processing a component including a target radionuclide by immersing at least a portion of the component in an electrolyte bath, applying an electric current to cause electrodissolution, and electroplating the target radionuclide from the solution onto a cathode. In some embodiments, a method includes processing a metal composite component by dissolving it to produce a solution including dissolved metal and particulate material, and separating the particulate material.

This invention discloses a storage well for storing containers of high-temperature products, comprising a well shaft. The well shaft includes a support member protruding from the inner wall of the well shaft to support the product container and separate the product container from the inner wall of the well shaft. This allows the support member, the inner wall, and the outer wall of the product container to form a ventilation channel for heat exchange. The storage well of this invention not only provides stable support but also excellent ventilation and heat dissipation, effectively improving storage safety. This invention also provides a construction method for the storage well and a method for storing the high-temperature product containers.

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 present application relates to a kind of radioactive feed liquid filtering device and method suitable for hot cell environment, the device includes lower barrel, filter core and upper cover, by the cooperation of compression bolt and threaded hole fastening, combined with integrated U-shaped sealing gasket can complete the compression and sealing of upper cover, filter core and lower barrel three components, reduce the difficulty of operation of manipulator, improve assembly sealing reliability, avoid radioactive material leakage;Filter core-U-shaped sealing gasket preloading structure reduces the fine assembly operation in radioactive environment, can realize filter core quick disassembly, adapt to the application scenarios of frequently replacing filter core, support flexible adjustment filtering condition.Moreover, by setting guide column-guide hole, bolt limiting-anti-falling structure, both help the hot cell operation of manipulator.The device and method provided by the present application are suitable for nuclear fuel reprocessing, radioactive analysis experiment, hot cell solid-liquid separation, radioactive filter residue acquisition and research, high sealing, high reliable filtering scene.

The invention provides a method and a device for preparing an adsorbent for adsorbing rubidium and cesium from lepidolite leachate. The preparation method of the adsorbent for adsorbing rubidium and cesium from the lepidolite leachate comprises the following steps: S1, weighing straw powder, and carrying out soda boiling treatment on the straw powder; s2, filtering while hot, collecting filter residues, washing with deionized water, and drying; s3, transferring the filter residues into a high-pressure kettle, and adding deionized water for hydrothermal reaction to obtain a reaction product. According to the preparation method and device of the adsorbent for adsorbing rubidium and cesium from the lepidolite leaching solution, the adsorbent has the effects of high adsorption capacity, good selectivity and stability in the aspect of rubidium and cesium adsorption, and the material can adsorb a large number of rubidium and cesium ions due to the rich pore structure and ion exchange sites; and the enrichment capacity on low-concentration rubidium and cesium is improved.

The present invention provides a molten material discharge device and discharge method for a glass melting furnace in a nuclear power plant. The molten material discharge device for a glass melting furnace in a nuclear power plant includes a target body containing molten material positioned at the top and a discharge assembly for continuously discharging the molten material to the bottom, the discharge assembly including a main body module, a drive module for providing driving force, a nozzle module for discharging the molten material from the target body, a heating module for heating the nozzle module by induction heating, a first temperature measurement module for acquiring first temperature information due to heating the nozzle module, and a second temperature measurement module for acquiring second temperature information, which is temperature information due to heating the nozzle module, separately from the first temperature information, the heating module maintains the nozzle module at a set temperature based on the first and second temperature information.

A reactor containment vessel vent system continuously releases steam generated in a reactor containment vessel to the atmosphere even when a power supply is lost. In the reactor containment vessel vent system, the noble gas filter that allows steam to pass through but does not allow radioactive noble gases to pass through among vent gas discharged from the reactor containment vessel is provided at a most downstream portion of the vent line. An immediate upstream portion of the noble gas filter and the reactor containment vessel are connected to each other by the return pipe via the intermediate vessel. When the radioactive noble gases having pressure equal to or higher than predetermined pressure stay in the immediate upstream portion of the noble gas filter, the staying radioactive noble gases flows into the intermediate vessel by the relief valve. Thus, steam permeability is maintained and the system continuously releases the steam.

The invention discloses an adsorption material for high-humidity and high-temperature iodine-containing aerosol and a preparation method thereof.According to the prepared adsorption material, polymers such as melamino-formaldehyde resin serve as a three-dimensional porous framework, and the adsorption material is prepared through a mild one-step solution processing method; an ionic polyimine network based on a lysine salt ionic liquid and carbonyl-rich polyvinylpyrrolidone are compounded on a skeleton in situ to form a porous adsorbent, and the porous adsorbent realizes efficient capture of elemental iodine (I2), methyl iodide (CH3I) and iodine aerosol through a synergistic mechanism of electrostatic induced adsorption and in-situ packaging and curing, so that the adsorption efficiency is improved, and the adsorption efficiency is improved. Under the harsh industrial conditions that the temperature is 100 DEG C or above and the relative humidity is 90% or above, the dynamic adsorption capacity to iodine reaches up to 160.7 wt%, the filtering efficiency is kept 99.8% or above, meanwhile, excellent iodine retention stability, recoverability and large-scale preparation potential are achieved, and an efficient, low-cost and practical solution is provided for treatment of radioactive iodine in nuclear waste gas.

The application provides a method and system for controlling air-borne radioactivity in a control area cabin of a water surface nuclear power ship. A gradient negative pressure is formed in the control area cabin of the nuclear power ship, such as a nuclear power plant containment cabin, an auxiliary cabin, a pollution monitoring and decontamination cabin, and an access passage cabin, by using facilities such as an exhaust fan and a negative pressure compressor, so that the air-borne radioactivity generated during normal operation of the nuclear power plant is controlled. The application can better limit the range of air-borne radioactivity in the control area cabin, prevent the air-borne radioactivity from spreading and leaking to other cabins, and provide strong protection for the radiation safety of personnel on the nuclear power ship.

The invention discloses a three-component covalent organic framework electrostatic spinning composite fiber material, which is prepared by using 1, 3, 6, 8-tetra-(p-aminophenyl)-pyrene, 3, 3 '-dihydroxy-[1, 1'-biphenyl]-4, 4 '-dicarboxaldehyde and benzo [1, 2-B: 4, 5-B'] dithiophene-2, 6-dicarboxaldehyde as monomers. The obtained composite fiber material is simple and convenient to synthesize, has excellent chemical stability, has dual functions of adsorption and photocatalysis, can convert soluble uranium into insoluble minerals, has the removal efficiency exceeding 97%, and has outstanding advantages in uranium-containing wastewater treatment.