386 results about "Radioactive waste" 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 lifting feeding equipment and curing equipment. According to the lifting feeding equipment provided by the embodiment of the invention, the conveying jacking device drives the material containing device to move in different directions, so that the material containing device abuts against the vacuum feeding device, the powdery materials can be conveyed to the material containing device from the vacuum feeding device, the powdery materials can be prevented from escaping through jacking communication, and the working efficiency is improved. And material waste and operation space pollution are avoided. According to the curing equipment provided by the embodiment of the invention, the material containing device is moved through the lifting feeding equipment, and the powdery materials are conveyed to the material containing device, so that the powdery materials can be prevented from being scattered; and meanwhile, the material containing device containing the powdery materials can reach other processing stations by moving the material containing device into or out of the containing space so as to carry out the next processing procedure.

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 utility model relates to a nuclear power high-radiation object containing device and containing and transferring equipment. The containing and transferring equipment comprises the nuclear power high-radiation object containing device. The nuclear power high-radioactivity storage device is specially designed for underwater operation, and can safely store and seal high-radioactivity waste under the condition that normal operation of a reactor is not affected. The sealing mechanism ensures that the radioactive substance storage space is completely sealed in the transferring and long-time storage process, radioactive substances are prevented from leaking, and the safety of the treatment process is greatly improved. And secondly, the torque input end allows torque to be input through an external tool to achieve rotation of the lock cylinder, so that the closing or opening action is rapidly and effectively completed, the operation process of containing and closing the radioactive substances is simplified, and the containing and closing efficiency of the radioactive wastes is improved.

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.

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.

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 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.

The invention discloses a preparation method and application of a melamine modified collagenous fiber iodine adsorbent, and relates to the technical field of nuclear medical radioactive waste treatment. The preparation method comprises the following steps: activating collagen fibers and modifying melamine; wherein the melamine modification comprises the following steps: taking the activated collagenous fiber, adding water, melamine and a dialdehyde cross-linking agent, and stirring at 35-45 DEG C for 6-8 hours to obtain the melamine modified collagenous fiber iodine adsorbent. According to the application, the prepared melamine modified collagenous fiber iodine adsorbent is used for adsorbing radioactive iodine. After melamine grafting, the nitrogen content of the collagenous fiber is remarkably increased, the collagen fiber has more prominent adsorption capacity on iodine steam and iodine in a solution, and an effective strategy is provided for removal and collection of radioactive iodine such as medical wastewater and radioactive waste liquid.

The invention relates to the technical field of radiation protection and environment monitoring, and particularly discloses a radioactive waste gas emission radiation dose real-time monitoring system which comprises a fixed monitoring base station network, an autonomous mobile monitoring cluster, a data fusion and cooperative control center and a radiation dose field three-dimensional reconstruction and evaluation terminal. Through air-ground cooperative monitoring, plume prediction and intelligent path planning based on an atmospheric diffusion model, and a physically constrained spatial interpolation algorithm, efficient tracking of a radioactive plume three-dimensional dynamic diffusion process and accurate reconstruction and visual evaluation of a three-dimensional dose field are realized. According to the system, an air-ground cooperative monitoring system composed of a fixed base station network and an autonomous mobile cluster is constructed, so that a passive and static monitoring normal form is fundamentally changed.

The invention discloses a multi-stage leaching purification method of a targeted therapeutic drug. The multi-stage leaching purification method comprises the following steps: sequentially filling an adsorption column with a porous organic polymer layer and an adsorption ball layer; controlling the flow rate of the leacheate and the column temperature for leaching, and setting an adsorption saturation threshold value; carrier gas is adopted to heat and desorb the porous polymer layer, and carrier gas components are switched to heat and desorb the adsorption ball layer; the method comprises the following steps: introducing carrier gas carrying iodine vapor into an alkaline absorption tower to treat waste gas, eluting impurities of adsorption balls by adopting an acidic mixed solution, and performing catalytic degradation by combining ultraviolet light; a chelating agent solution is used for regenerating adsorption balls, a silver salt solution is used for heavily loading silver ions, an organic solvent is used for eluting and regenerating a porous polymer, regular calibration and part replacement are conducted on a system, and organic pollutants and radionuclides in a matrix are difficult to treat synchronously through a traditional single-layer adsorption material; according to the leaching purification method, the generation amount of radioactive waste liquid is remarkably reduced, and meanwhile the shielding protection requirement of the follow-up purification step is reduced.

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.

This invention discloses a method for precipitation separation and ceramic solidification of divalent strontium and tetravalent uranium in radioactive waste liquid, relating to the field of radioactive waste liquid treatment and disposal. U in high-level radioactive waste liquid is in the form of hexavalent uranyl (UO2). 2+ U exists in the form of hydrogenation catalytic reduction, and the present invention uses hydrogenation catalytic reduction to prepare U. 4+ Sr 2+ and U 4+ ionic radius and Ln 3+ Due to the significant differences, CePO4·0.667H2O cerium phosphate, which has a stronger nuclide-containing capacity, was selected as the co-precipitation enrichment agent for Sr. 2+ and U 4+ Separation precipitant. Inexpensive raw materials such as UO2(NO3)2·6H2O, Sr(C2H3O2)2, NH4H2PO4, and Ce(NO3)3·6H2O are selected. By controlling the Sr... 2+ :U 4+ Ion ratio, reaction time, temperature, pH, and preparation of well-developed Sr crystals 2+ / U 4+ The precipitate of cerium phosphate solid solution with high solid content is avoided by using spark plasma solid-state sintering technology to prevent grain agglomeration and adhesion during high-temperature dehydration and crystal transformation of cerium phosphate. Furthermore, pressure sintering can promote further growth of nanocrystals and promote U... 4+ and Sr 2+ Homogenization within the monazite ceramic solidified body enables a one-step preparation of monazite ceramic solidified body from a lanthanite solid solution.

The utility model discloses a nuclear power plant radioactive waste barrel gamma scanning device and measuring equipment, and the nuclear power plant radioactive waste barrel gamma scanning device comprises a supporting seat, a sealing cover, a rotating disc, a control mechanism and a fixing assembly. The sealing cover is detachably mounted on the supporting seat, and a sealing cavity is formed between the sealing cover and the supporting seat; the turntable is detachably mounted on the supporting seat and rotates in the sealing cavity; the fixing assembly is installed on the rotating disc, the radioactive waste barrel is detachably fixed to the rotating disc, and the axis of the radioactive waste barrel is parallel to the plane of the rotating disc. Therefore, the total activity of the radioactive raw material in the radioactive waste barrel can be quickly obtained only through twice measurement, the measurement efficiency and the accuracy of activity reconstruction are improved, the layering frequency is reduced, and the measurement time is saved.

The utility model discloses a transport container for radioactive waste barrels, which comprises a top cover, a barrel body and a base, the barrel body is mounted above the base, the top cover is mounted above the barrel body, the top cover is used for sealing the barrel body, and the radioactive waste barrels are placed in the barrel body. The top cover comprises a cover body and a lifting part above the cover body, the lifting part is used for lifting the top cover, and a plurality of long slotted holes are formed in the cover body; the barrel body comprises a barrel body, a barrel body flange plate, a positioning rib plate, a wrap angle and a plurality of lifting lugs, the barrel body flange plate is fixed at the top end of the barrel body, the barrel body flange plate and the cover body are connected at any angle through a long slotted hole by using a bolt, the positioning rib plate is arranged at the lower end of the inner side wall of the barrel body, the wrap angle is arranged on the outer side of the barrel body flange plate, and the lifting lugs are fixed on the outer side of the barrel body; the base comprises a base plate and a plurality of forklift grooves, and the forklift grooves are fixed below the base plate. The transportation container is safe, stable, easy to load and transport, shockproof and capable of being integrally transferred.

The invention relates to a preparation method and application of a PAM / PB / PAO dual-network hydrogel adsorption material with flexibility and uranium adsorption selectivity. The invention relates to a preparation method and application of a PAM / PB / PAO dual-network hydrogel adsorption material. The technical problems that an existing hydrogel uranium adsorption material is poor in mechanical property, uneven in functional group distribution, low in adsorption selectivity in an actual complex water body and insufficient in cycling stability are solved. The method comprises the following steps: carrying out radiation crosslinking treatment to obtain PAM / PB / PAN composite hydrogel; and carrying out amidoximation reaction to obtain the PAM / PB / PAO dual-network hydrogel adsorption material. The material has excellent anti-fatigue property, high adsorption capacity and good cycle stability, and shows remarkable adsorption selectivity on uranyl ions in the presence of various competitive ions. The PAM / PB / PAO dual-network hydrogel adsorption material with flexibility and uranium adsorption selectivity prepared by the invention is applied to the field of uranium adsorption in radioactive waste liquid.

The invention relates to the technical field of cranes, in particular to a bridge crane capable of realizing secondary stacking, which comprises a crane cart, a crane trolley, a vertical frame and a secondary stacking mechanism, the vertical frame is fixedly connected to the side face of the crane trolley and extends in the Z-axis direction. The secondary stacking mechanism is installed on the side, away from the crane trolley, of the vertical frame and used for stacking goods to the area above the bridge frame. Wherein the secondary stacking mechanism comprises a supporting plate, a transverse frame and a clamp; through cooperation of the supporting plate, the transverse frame and the clamp in the secondary stacking mechanism, a plurality of cargoes (such as nine low-medium-level radioactive waste bags arranged in a 3 * 3 matrix) arranged in a matrix mode can be clamped at a time, the cargoes do not need to be hoisted and transferred one by one like in the prior art, the stacking efficiency is greatly improved, and then the operation period is shortened.

A shielding module for a radioactive waste container (C) comprising a perimeter wall (1) and a closing lid (2), which define a cavity (3) suitable for housing the container (C), and a number of metal elements (4) attached along the perimeter wall (1) and protruding into said cavity (3), said metal elements (4) forming means for limiting lateral movement, for centring and for dampening impacts against the side wall of the container (C) with respect to the shielding module (10), in the event of potential external impacts or the sideways fall of the assembly formed by the container (C) and the shielding module (10).

Radioactive Waste Container

The utility model provides a plasma torch and a plasma furnace for disposing radioactive wastes, and aims to solve the technical problem of radioactive leakage when a cathode is replaced midway. The plasma torch comprises an anode tube which is internally provided with an anode material, one end of the anode tube is a nozzle, the other end of the anode tube is a connecting part, and the connecting part is provided with a working gas channel; one end of the cathode rod is movably arranged in the connecting part, the cathode rod is made of a cathode material, and a gap exists between the end part of the cathode rod and the anode material; the driving end of the driving part is in power connection with the cathode rod, and the driving part is used for driving the cathode rod to move into the anode tube; the cathode of the plasma power supply is electrically connected with the cathode rod, and the anode is electrically connected with the anode material; and the output end of the working gas supply device is communicated with the working gas channel. According to the plasma torch, the cathode rod is made to move in the connecting part through the driving piece, the distance between the cathode rod and the anode material is reduced, and therefore radioactive leakage caused by cathode material replacement is avoided.

The invention discloses a treatment method of magnesium-containing high-radioactivity waste liquid, and belongs to the technical field of high-radioactivity waste treatment.The treatment method comprises the steps that the magnesium-containing high-radioactivity waste liquid is sequentially filled with first-stage, second-stage, third-stage and fourth-stage chromatographic columns with different column fillers, high-heat-release fission nuclides in the magnesium-containing high-radioactivity waste liquid are removed, the radioactivity intensity of the magnesium-containing high-radioactivity waste liquid is reduced, and the high-radioactivity waste liquid is obtained; the method comprises the following steps: adding anhydrous oxalic acid into magnesium-containing high-level radioactive waste liquid to obtain magnesium-containing low-level radioactive waste liquid, further adjusting the magnesium-containing low-level radioactive waste liquid to be alkalescent, adding anhydrous oxalic acid, removing magnesium element in the waste liquid by adopting a precipitation method, reducing the salinity of the low-level radioactive waste liquid, and converting the magnesium-containing high-level radioactive waste liquid into low-salinity and low-radioactivity waste liquid which can be safely stored or further treated. According to the method, the high-radioactivity waste liquid can be subjected to rapid degradation treatment in situ, the radioactivity intensity and salinity of the high-radioactivity waste liquid are effectively reduced, the technological process is simple, the heat release nuclide removal efficiency is high, the secondary waste amount is small, and the method is suitable for implementation and easy to popularize.

The invention discloses a compression system for treating a radioactive waste filter element, which is characterized by comprising a lifting machine, a filter element receiving device, a crusher, a conveyor, a charging basket transfer box, a local discharge filter, a high-efficiency filter, an induced draft fan and a control cabinet, and can completely realize receiving, crushing, conveying and compression processes of the radioactive filter element. Due to the specific crushing-compressing step-by-step process, the device can be used for treating filter elements of various specifications, and the compatibility is relatively good. According to the device, through a technical route of compression in the barrel after crushing and by utilizing the characteristic of small volume rebound after compression of a metal frame, rebound after compression of a filter material is inhibited, and a better compression volume reduction effect is realized.

The invention relates to the technical field of digital twinning technology and radioactive waste treatment, and discloses a digital twinning-based radioactive large tank dismantling scheme optimization method and system. The method comprises the steps of obtaining multi-working-condition operation data and operator behavior data to generate a digital twinborn model basic data set, generating a system risk state quantitative index by using a system critical state evaluation model, identifying a potential operation error type based on an error pattern library, generating a risk coupling index by using a system-human factor risk coupling model, and establishing a twinborn model based on the risk coupling index. And generating a risk pre-control intervention strategy based on the risk coupling index. According to the method, the technical problem that system risks and human risks are separated in a traditional method is solved, and cooperative prevention of composite risks is achieved.

The application provides a radioactive separation resin life prediction method and system based on machine learning, relates to the technical field of radioactive waste liquid treatment, and the method comprises the following steps: acquiring real-time online monitoring data of a radioactive separation resin in a running process, wherein the real-time online monitoring data comprises local adsorption capacity, local temperature and local radiation dose collected from three point positions, namely, upper, middle and lower positions, of the axial distribution of a resin column; based on the real-time online monitoring data, a microstructure deterioration proxy model is constructed, and a microstructure deterioration quantitative index of the resin under a current working condition is calculated; and the real-time online monitoring data and the microstructure deterioration quantitative index are subjected to feature fusion, and a fusion feature set is constructed. The application guarantees the continuous, stable and safe operation of a radioactive waste liquid treatment and nuclide separation system.

The application provides a palladium ion extractant material and a preparation method thereof, and the pyridine-oxadiazole ligand is used as the extractant, most Pd 2+ in the aqueous phase can be extracted into the organic phase in a short time, and has a very high extraction rate; selective extraction separation of Pd 2+ under acidic conditions can be realized, the applicable range of acidity is wide, and excellent extraction capacity is achieved in the range of 0.2-4M nitric acid concentration; the extractant has excellent selectivity to Pd 2+ in the aqueous phase containing various metal ions; the pyridine-oxadiazole extractant has a good application prospect in the separation and recovery of Pd 2+ in high-level radioactive waste; the m-nitro-trifluorotoluene is used as a diluent of the extraction system, has good solubility to the organic ligand and the extractant, and the 2-bromohexanoic acid can increase the lipophilicity of the extractant.

This invention relates to the field of radioactive waste management technology, specifically disclosing a trolley device for removing residual tritium from heavy water reactor waste filter cartridges. The device comprises a trolley body, a molecular sieve absorption chamber, and a tail gas treatment filter, connected in sequence. The trolley body has a built-in heating element, a weight monitor at the bottom, and a gas outlet at the right end. The trolley body transports tritium-containing vapor to the molecular sieve absorption chamber through the outlet. The molecular sieve absorption chamber contains several molecular sieves that adsorb the tritium-containing vapor, forming tail gas. The tail gas is transmitted through a pipeline to the tail gas treatment filter, which contains several filter elements for supplementary absorption. A tail gas monitor is located at the outlet of the tail gas treatment filter. This device allows the waste filter cartridges to be treated to be lifted and loaded into the trolley for tritium removal during transport and storage. After treatment, the tritium water content of the waste filter cartridges can be significantly reduced, making them meet relevant disposal requirements.

The utility model relates to the technical field of radioactive waste treatment, and provides a semi-volatile substance trapping device. The semi-volatile substance trapping device comprises: a heater having a first heating section and a second heating section; the heating pipe penetrates through the heater, the heating pipe is provided with a volatilization area and a trapping area, the volatilization area is arranged corresponding to the first heating section, and the trapping area is arranged corresponding to the second heating section; the first heating section is used for heating a solid-state semi-volatile substance in the volatilization area, so that the solid-state semi-volatile substance is converted into a gaseous-state semi-volatile substance; the trapping material support is located in the trapping area, the trapping material support is used for containing trapping materials, and the gaseous semi-volatile substances can flow to the trapping area from the volatilization area and are trapped by the trapping materials. According to the semi-volatile substance trapping device, the trapping effect of the semi-volatile substances is improved, and deep exploration and research of volatilization and trapping characteristics of various semi-volatile substances in nuclear waste are promoted.

The invention discloses a glass bead feeding device and method for radioactive waste liquid glass solidification, the glass bead feeding device comprises a drying primary screening device, a fine screening device, a storage bin and a discharging device which are sequentially arranged according to the flowing sequence of glass beads, the drying primary screening device comprises a baking box, and a heater is arranged at the top in the baking box; a feeding hopper and a discharging hopper are arranged on the two sides of the baking box respectively, the fine screening device is arranged below the discharging hopper, a primary screening mechanism located between the feeding hopper and the discharging hopper is arranged in the baking box, and the primary screening mechanism is used for primarily screening out glass beads meeting the particle size and conveying the glass beads to the discharging hopper. The screening device has the advantages that the follow-up screening pressure is reduced, the number of times of cleaning the screen is reduced, and the working efficiency and the screening precision are improved.