1599results about "Reactor fuel elements" patented technology

The invention discloses a method for processing foreign matters in the lower reactor internals of a reactor, which comprises the following steps of: completing global video inspection of the lower reactor internals of the reactor through a video inspection mechanism, determining whether foreign matters exist or not, and determining the positions, the sizes and the types of the foreign matters according to global images of the lower reactor internals of the reactor acquired by the video inspection mechanism. A data processing module of the position navigation mechanism is used for determining a corresponding drain hole area number of the target foreign body, a mechanical arm body type of the mechanical arm mechanism, a gripping apparatus type of the gripping mechanism and an initial movement posture of the mechanical arm mechanism; and the remote control operation mechanism is used for operating the mechanical arm mechanism to move and operating the grabbing mechanism to accurately grab the target foreign matter. According to the method, a high-efficiency full-area inspection and foreign matter grabbing method capable of covering the reactor internals at the lower part of the reactor is formed, the traditional mode that the reactor internals at the lower part of the reactor are firstly hoisted out of the pressure vessel and then the foreign matter is grabbed is changed, and the risk of damage to the reactor internals and the pressure vessel body is reduced.

A molten salt reactor system may produce fission products, such as molybdenum-99, which may be extracted by an extraction system. The extraction system may be configured to sparge the molten fuel salt with an inert gas to dislodge gaseous fission products from the molten fuel salt, so that the gaseous fission products flow through an off-gas outlet into a gas transfer assembly. The example extraction system may also include introducing a gaseous halogenating agent to prevent the fission products from depositing on a pipe surface of the gas transfer assembly. The gaseous fission products may flow from the gas transfer assembly to a carboy containing an aqueous solution, to process and / or filter the fission products. The resulting solution containing processed fission products in the carboy may be further distilled into final, usable products.

The invention discloses a pressurized water reactor rod changing method and system, electronic equipment and a computer readable storage medium, and the method comprises the steps: obtaining a plurality of rod pre-changing strategies corresponding to a first burnup point according to a plurality of rod changing interpolation relationships of a plurality of control rods when it is monitored that the current burnup point is the first burnup point, each control rod corresponds to a rod changing interpolation relation, the rod changing interpolation relation comprises a plurality of burnup points and a rod pre-changing strategy corresponding to each burnup point, and the rod pre-changing strategy is used for indicating the pre-insertion step number of the corresponding control rod; adjusting the pre-insertion step number of each control rod according to a rod pre-changing strategy corresponding to each control rod at the first burn-up point; and rod replacement processing is carried out based on the multiple adjusted control rods. According to the invention, disturbance of the pressurized water reactor can be reduced when rod sequence exchange is carried out on the control rod.

The invention relates to the technical field of fused salt purification, and provides a fused salt purification device and a fused salt purification method.The fused salt purification device comprises a heating furnace, heat insulation layers, a reactor, a sleeve, a gas inlet and outlet device and a crucible, the heating furnace is provided with multiple sections of heating areas in the height direction, and the heating areas are separated by the heat insulation layers and can independently control the temperature; the reactor is partially arranged in the heating furnace and extends into the multiple sections of heating areas, and the reactor is provided with a reaction cavity and a cavity opening communicated with the reaction cavity; the sleeve is placed in the reaction cavity from the cavity opening and covers the cavity opening, a placement cavity is formed in the sleeve, and the gas inlet and outlet device is communicated with the placement cavity and used for providing inert gas for the placement cavity; the crucible is placed in the containing cavity and used for containing the fused salt to be purified. After the to-be-purified fused salt is heated and melted, each section of heating area can be controlled from top to bottom to slowly cool, so that the fused salt is solidified from top to bottom, purification is completed, impurities flow to the bottom of the crucible, the upper-layer purified fused salt is obtained, and the structure is simple.

Disclosed in the present invention is a tritium production module system in a reactor. The tritium production module system has a multi-layer structure distributed in the radial direction of the reactor, and the tritium production module system comprises a tritium production area and a circulation area, wherein the tritium production area comprises a tritium breeding material, the tritium breeding material interacting with neutrons to produce tritium elements; the tritium production module system is in fluid communication with the outside of the tritium production module system via the circulation area; the circulation area comprises a tritium pipeline, the tritium pipeline being used for circulation of a reactor coolant and a tritium-containing medium, and the tritium pipeline extending in the axial direction of the reactor; a plurality of holes are provided in a pipe wall of the tritium pipeline, and the tritium pipeline is in fluid communication with the tritium production area via the holes, thereby improving the ability and efficiency of tritium transfer in the tritium production module system.

The invention discloses an in-situ global foreign matter checking and grabbing system for reactor internals on the lower portion of a reactor. The in-situ global foreign matter checking and grabbing system for the reactor internals on the lower portion of the reactor comprises a base device, an adjusting device and a foreign matter grabbing device. The foreign matter grabbing device is arranged on the base device through the adjusting device. The foreign matter grabbing device comprises a flexible arm assembly, a flexible arm driving assembly, a quick change assembly, at least one clamping jaw used for grabbing foreign matter and a video tracking and checking assembly used for searching the foreign matter and monitoring the grabbing condition. The flexible arm assembly is arranged on the adjusting device; the at least one clamping jaw is detachably arranged on the flexible arm assembly through the quick-change assembly; the video tracking and checking assembly is arranged on the adjusting device; according to the invention, the flexible arm assembly carries various types of clamping jaws to carry out automatic grabbing operation, so that the in-service nuclear power plant reactor internals can accurately, efficiently and safely grab and clean foreign matters in the in-situ interior of a pressure vessel in a whole region.

The invention belongs to the technical field of nuclear power, and particularly relates to a device and method for remote installation of a heat insulation sleeve relieving cushion block. Comprising a gripping apparatus and a screwing tool, the gripping apparatus is used for gripping the cushion block, the bolt and the locking cap to mount and dismount the cushion block, and the screwing tool is used for positioning and mounting the relieving cushion block on the upper cover plate of the control rod guide cylinder and fastening the bolt. The clamping device has the beneficial effects that the air cylinder is adopted to drive the grabhooks to be opened and closed, clamping of the cushion block is achieved, meanwhile, it is guaranteed that the grabhooks are in an open state during air loss through the built-in springs, and the cushion block is prevented from falling off. A grabbing structure that the bolts and the locking caps are lifted along with the cushion blocks is adopted, meanwhile, the anti-falling function is achieved, and integrated operation of clamping, conveying, positioning and screwing of the cushion blocks is achieved.

The invention relates to a preparation method of an irradiation-resistant and accident-resistant SiC / SiC nuclear fuel cladding. The preparation method comprises the following steps: (1) alternately carrying out the following step (A) and step (B) from step (A) to step (A) to obtain a tubular SiC fiber preform with at least two PyC interface layers and at least one SiC nanowire interface layer: (A) depositing the PyC interface layers on the surface of the tubular SiC fiber preform through a chemical vapor deposition process; (B) loading a catalyst on the PyC interface layer through vacuum impregnation, and generating a SiC nanowire interface layer on the surface of the PyC interface layer in situ through a chemical vapor deposition process; (2) carrying out densification treatment on the tubular SiC fiber preform by adopting a chemical vapor infiltration process; and (3) depositing a SiC coating on the surface of the tubular SiC fiber preform by adopting a chemical vapor deposition process to obtain the SiC / SiC nuclear fuel cladding tube.

The invention belongs to the technical field of reactor pool water turbidity control before nuclear power plant refueling overhaul loading / unloading, and particularly relates to a reactor pool turbidity control method before nuclear power plant overhaul loading / unloading. Front-end control is carried out on work related to reactor pool water quality from a unit refueling overhaul preparation stage, a descending stage, a low-low water level stage and an ascending stage; in the preparation stage, a water filling source refueling water tank before reactor pool unloading is purified before unit refueling overhaul, it is expected that purification is conducted till the turbidity of the refueling water tank is smaller than 0.1 NTU, and it is guaranteed that the water filling source of the reactor pool is clear before unloading; in the descending stage, the unit overhaul first loop 80 DEG C platform purification is completed, and discharging is completed; in the low-water-level stage, a primary loop coolant is discharged to a refueling water tank after unloading is completed, and a primary loop system is filled with water after low-water-level maintenance is completed; and the ascending stage is from primary loop water filling to charging completion. Therefore, the purpose of controlling the influence of turbid water in the reactor pool before loading / unloading on normal loading / unloading work is achieved.

The invention discloses a high-temperature-irradiation-resistant zirconium alloy and a manufacturing method and a workpiece thereof, and belongs to the field of zirconium alloys. Wherein the zirconium alloy resistant to high temperature irradiation comprises the following components in percentage by weight: not less than 96% of Zr, 0.61%-1.4% of Sn, 0.41%-1.4% of Nb, 0.1%-0.5% of Fe, 0.11%-0.2% of O and 0.001%-0.3% of Cr; and from 0.015% to 0.03% of Si and / or from 0.001% to 0.06% of Ge; wherein Sn + Nb + O is greater than or equal to 1.5%, Fe / Cr is greater than or equal to 1.5%, Si + Ge is greater than or equal to 0.015%, and (Sn + Nb + O) / (Fe + Cr + Si + Ge) = 1.5-25.0. The alloy has good mechanical properties, corrosion resistance, creep resistance and radiation growth resistance, and is suitable for manufacturing of pressurized water reactor core structural members.

The invention belongs to the technical field of sodium-cooled fast reactor loading and reloading, and particularly relates to an argon heating and control system and method for a sodium-cooled fast reactor assembly. Comprising heaters and draught fans, each heater is connected with one draught fan through a pipeline, the two heaters are connected in parallel through a pipeline and connected with exhaust valves connected in series, the two heaters connected in parallel are further connected with a control cabinet through a pipeline, one draught fan connected with the heater is further connected with the two draught fans connected in series through a pipeline, and a valve is arranged on the pipeline. The serially connected fans are connected with serially connected air inlet valves through pipelines provided with valves, and the air inlet valves are connected with a control cabinet through pipelines. The system has the beneficial effects that gas is heated through the heater, then gas flowing is accelerated through the draught fan, and the heated gas is subjected to rapid circulation heat exchange in the container so that the temperature rising function can be achieved.

The embodiment of the invention relates to the technical field of nuclear reactors, in particular to a cock assembly for a reactor. The reactor comprises a reactor container and a reactor core assembly arranged in the reactor container, the reactor container comprises a container opening, and the reactor further comprises a refueling machine. The cock assembly comprises a first cock, a second cock and a third cock. The first cock is rotatably arranged at a container opening of a reactor container of a reactor, the first cock is provided with an eccentric first through hole, and the size of the first through hole is larger than the radius of the first cock. The second cock is arranged in the first through hole of the first cock and can rotate relative to the first cock, and the second cock is provided with an eccentric second through hole. The third cock is arranged in the second through hole of the second cock and can rotate relative to the second cock, and the third cock is provided with an eccentric third through hole. And the refueling machine is arranged in a third through hole of the third cock. According to the cock assembly, the moving range of the refueling machine can be increased, and the size of the first cock is reduced.

A reactor body structure and a reactor system. The reactor body structure (100) comprises: a reactor pressure vessel (1), the reactor pressure vessel (1) comprising a bottom closure head (3), and the bottom closure head (3) having an arc-shaped inner wall (32); a reactor internal (2), the reactor internal (2) comprising a reactor core supporting lower plate (22); and a reactor core (21), the reactor core (21) being arranged on the reactor core supporting lower plate (22); wherein the reactor core supporting lower plate (22) is located on the lower side of the center of sphere of the arc-shaped inner wall (32), and the outer edge of the reactor core supporting lower plate (22) is provided with a flow guide corner (221); the reactor core supporting lower plate (22) is provided with through holes (222), the through holes (222) are stepped holes and each comprise a top hole (222a) in the upper side and a bottom hole (222b) in the lower side, and the diameter of the bottom hole (222b) is smaller than that of the top hole (222a).

The embodiment of the invention provides an anti-blocking device and an anti-blocking method for a fuel loading and unloading system of a high-temperature gas cooled reactor, and relates to the field of high-temperature gas cooled reactors. The anti-blocking device of the high-temperature gas cooled reactor fuel loading and unloading system comprises a driving module, a transmission module, an execution module and a control module, the transmission module is connected with the driving module, the execution module is connected with the transmission module, and the transmission module is driven by the driving module to drive the execution module to do linear reciprocating motion. And the control module is connected with the driving module and the execution module at the same time. A cooperative structure of the driving module, the transmission module and the execution module is arranged, a driving signal is converted into linear reciprocating motion of the execution module, mechanical disturbance is applied to a fuel ball stack in a blocked area, the lap joint balance or a bridge junction structure of the fuel ball stack is damaged, and active bridge breaking and jam releasing are achieved. And the jam releasing efficiency and success rate are obviously improved.

The invention provides a reactor pressure vessel flange sealing surface defect measuring method and a reactor pressure vessel flange sealing surface defect repairing method. The measuring method comprises the following steps: S1, measuring and calculating the minimum diameter and the maximum diameter of an annular band formed in a sealing ring indentation range; and S2, by taking the center of the container flange as a circle center, inspecting the defect, falling into the range of the annular belt between the minimum diameter and the maximum diameter, on the container flange, and measuring and recording the size and the orientation of the defect. According to the method, corresponding recommended repair methods are provided according to repair criteria for defects of different scales, the method can be used for guiding implementation of nuclear power plant RPV sealing surface maintenance, the repair workload is effectively reduced, the radiation dose of personnel is reduced, the operation and maintenance efficiency is improved, safe and stable operation of a nuclear power unit is ensured, and the economic benefit of power station operation is improved.

The invention relates to a disassembly device and method for a spent fuel rod bundle, and the disassembly device comprises a workbench, a fixed chuck which is fixedly installed on the workbench and is used for fixing a first end of a spent fuel rod, an outer sleeve which is arranged above the fixed chuck and is used for cutting the spent fuel rod bundle, and a cutting mechanism which is used for cutting the spent fuel rod bundle, the first chuck and the second chuck can move along the workbench and are used for carrying an outer sleeve of a spent fuel rod, so that the outer sleeve is separated from a spent fuel rod bundle; the supporting fixture is connected to the workbench in a lifting manner and is used for supporting the outer sleeve and the spent fuel rod bundle in preset time; the clamping mechanism is used for carrying the spent fuel rod bundles to the transfer platform, then carrying the spent fuel rod bundles to the conveying line in a layered mode and placing the spent fuel rod bundles in a single-layer rod arrangement form, the structure is simple, the automation degree is high, a traditional manual arrangement mode can be effectively replaced for treating the spent fuel rods, and the working efficiency is improved. The safety in the operation process is ensured; the spent fuel rod bundles which are complexly stacked can be efficiently disassembled, and the disassembling efficiency is remarkably improved; the spent fuel rods which are originally tightly stacked can be ingeniously placed in a uniform single-layer rod arrangement form, and the subsequent treatment process is simplified.

Disclosed herein are methods and systems for the recovery of uranium from used fuel salts using nonaqueous chemistry without electrometallurgy techniques. The used fuel salts are maintained in a molten state in a chlorination chamber, and are then chlorinated for a period of time sufficient to convert at least some UCl3 into a different uranium chloride, thereby generating a uranium-containing gas phase. The uranium-containing gas phase is separated from the resulting residue, and is condensed for collection.

The invention discloses a radiation-resistant high-entropy alloy magnetron sputtering coating and a preparation method thereof, and belongs to the technical field of radiation-resistant coating materials. On the basis of the multi-element synergistic effect and lamination interface design of the high-entropy alloy, FeCrNi and FeNiCrMn are used as target materials, and the coating with the nanometer lamination structure is prepared by accurately regulating and controlling the target current, the gas flow, the deposition time and other process parameters of magnetron sputtering; meanwhile, accurate control over the single-layer film thickness and the interface density is achieved by adjusting the coupling relation between the sputtering power and the deposition rate, and it is ensured that all unit layers are flat and compact, and interface combination is complete. And a multi-heterogeneous interface is formed through the design of a laminated interface, so that a defect capture network is formed, the hardness of a single-layer coating is improved, the hardening rate after irradiation can be lower than that of a traditional single-layer coating, and the problems of crystallization and brittleness failure induced by irradiation are effectively relieved.

The invention provides an integrated mounting system and method for a high-temperature gas cooled reactor core supporting structure, and relates to the technical field of nuclear power construction, the integrated mounting system for the high-temperature gas cooled reactor core supporting structure comprises a pre-assembly plant, and a bridge crane is arranged in the pre-assembly plant; the first operation platform is provided with pre-assembly cabins arranged in multiple rows and multiple columns, each pre-assembly cabin is correspondingly provided with two cantilever cranes, and the cantilever cranes are used for reshipment, namely installation, of equipment in the pre-assembly cabins. The integrated mounting system for the high-temperature gas cooled reactor core supporting structure solves the problem that the reactor core supporting structure is lack of an integrated pre-assembly environment outside a nuclear island plant, and provides guarantee for smoothly realizing integrated pre-assembly of the reactor core supporting structure outside the nuclear island plant; the construction difficulty is correspondingly reduced, the construction operation environment is improved, and the construction level of nuclear power engineering in China and the economical efficiency of high-temperature gas cooled reactor nuclear power are improved.

The invention relates to the field of corrosion-resistant nuclear materials, in particular to a liquid lead bismuth corrosion-resistant high-strength oxide dispersion-strengthened FeCrAl alloy and a preparation method thereof. The oxide dispersion strengthened FeCrAl alloy comprises the following components in percentage by mass: less than or equal to 0.01% of C; 11 to 25 percent of Cr; 5.0% to 10.0% of Al; 0 to 3.0% of W; ta + V < = 0.35%; zr: 0.1 to 0.5%; 0.1 to 0.5 percent of Y; 0.15 to 0.5 percent of O and the balance of Fe. According to the material, a proper amount of Zr element is added to promote Y-Zr-O phase nano-particles with small sizes to replace Y-Al-O phase nano-particles with slightly large sizes to be separated out preferentially, and the mechanical strength of the ODS FeCrAl alloy is improved. Aiming at the potential hazard that Zr element may hinder generation of an ODS FeCrAl alloy surface aluminum oxide film in a lead-bismuth environment, consumption of a large amount of Zr element in a Y-Zr-O strengthening phase form is promoted by regulating and controlling the content of ball-milled powder and alloy O, the Zr content of an alloy matrix is reduced, and adverse effects on generation of a protective aluminum oxide film are reduced and eliminated; and the ODS FeCrAl alloy with excellent lead and bismuth corrosion resistance and excellent high-temperature mechanical property is obtained.

The utility model provides a positioning device and an assembling platform comprising the same, the positioning device comprises a stacking bottom plate, stacking elements are installed on the stacking bottom plate layer by layer, at least two measurement marks are arranged on the stacking bottom plate and / or an installation platform outside the stacking bottom plate at intervals, and the measurement marks are used for being matched with measurement equipment arranged above the installation platform. Determining reference space coordinate information of the stacked bottom plate; the positioning device further comprises a guide rod, and the guide rod penetrates through the stacking bottom plate or the stacking element of the installed layer and the stacking element to be assembled so as to determine relative space coordinate information of the stacking element to be assembled relative to the stacking bottom plate. According to the positioning device and the assembly platform comprising the positioning device, under the condition that stacked elements are stacked layer by layer, the positioning accuracy, reliability and consistency of a multi-layer assembly structure are improved, and the assembly efficiency can be improved.

The invention relates to a reactor core high-level detector pulling mechanism which provides effective radiation protection in the pulling process of a detector by introducing a shielding cylinder, so that the radiation dosage possibly received by an operator during operation is remarkably reduced, and the safety of a working environment is improved. Secondly, an alignment connecting device with an alignment camera is further arranged, the position of the lifting and moving assembly can be monitored and adjusted in real time, it is ensured that the lifting and moving assembly is accurately aligned and connected with the detector to be pulled out, and in the process that the shielding cylinder gradually goes deep into the reactor core water pool, the detection accuracy is improved. And position interference between the shielding cylinder and other structural members in the reactor core can be avoided, so that the problem that accurate butt joint is difficult in a traditional method is solved, and the detector pulling-out efficiency and the operation safety are improved.

The invention discloses a multifunctional fuel assembly inspection device for an inspection basket. The multifunctional fuel assembly inspection device comprises a hoisting and mounting module, a vertical moving module, a horizontal moving module, an inspection module and a foreign matter grabbing and video inspection module. According to the multifunctional fuel assembly inspection device for the inspection basket, the hoisting installation module is in butt joint with the inspection basket, relative positioning of the multifunctional inspection device and the inspection basket is achieved, different inspections are conducted on a fuel assembly placed in the inspection basket through the inspection module, and the inspection module can be connected with the horizontal moving module; and meanwhile, through adjustment of the vertical moving module and the horizontal moving module, inspection of different positions of the fuel assembly placed in the inspection basket by the inspection module can be completed, and the function of grabbing foreign matters in the fuel assembly is achieved through the foreign matter grabbing and video inspection module. The multifunctional inspection device is complete in module, convenient to replace and efficient to install, and has important significance on guaranteeing all-around inspection and safe operation of the fuel assembly.

The invention mainly provides a nuclear fuel cladding composite material and a preparation method thereof. The nuclear fuel cladding composite material comprises a zirconium alloy matrix, the niobium-rich diffusion barrier layer is formed on the surface of the zirconium alloy matrix; the Cr coating covers the niobium-rich diffusion barrier layer; and an Al-rich layer formed on the surface of the Cr coating. The preparation method comprises the following steps: firstly, carrying out surface pretreatment on the zirconium alloy matrix; a niobium-rich diffusion barrier layer is formed on the surface of the zirconium alloy matrix through ion implantation; then depositing a Cr coating on the niobium-rich layer; and finally, an Al element is injected into the surface of the Cr coating, and an Al-rich layer is formed so that an Al2O3 protective layer can be formed under the high-temperature working condition. According to the technical scheme, interface mutual diffusion can be effectively inhibited, the stability of the coating is improved, meanwhile, the Cr coating structure is optimized, the oxidation process is delayed, a gradient protection system is constructed, and the oxidation resistance of a high-temperature material is enhanced.

A nuclear reactor having an axially stratified fuel bed. The reactor features a reactor housing having a base, a top with an exhaust port, and a shaft. An axially stratified fuel bed within a reactor housing comprising: a first zone configured to operate at a first temperature T1, the first zone comprising a plurality of first fuel pellets, each first fuel pellet comprising a first radioactive ceramic core and a first ceramic seal coating; a second zone configured to operate at a second temperature T2, wherein T2gt; t1, the second zone comprising a plurality of second fuel particles, each second fuel particle comprising a second radioactive ceramic core and a second ceramic sealing coating. A coolant fluid flow path delivers coolant fluid from the reactor base to the vent, sequentially through the first and second zones along the flow path. The stability of the first ceramic sealing coating in T1 is greater than the stability in T2, and the stability of the second ceramic sealing coating in T2 is greater than the stability of the first ceramic sealing coating in T2.

The invention relates to a nuclear reactor detector assembly recovery system and method, and the system comprises a shielding transfer device which is used for pulling out a detector from a nuclear reactor and transferring the detector to a preset position; the coiling equipment is arranged on the guide rail assembly and driven by the lifting assembly to move below the liquid level of the component pool, and the coiling equipment is used for coiling part of the detector; and the dismantling platform is arranged at the pool opening of the component pool and used for bearing the hanging frame assembly, the shearing assembly and the clamping assembly. The working process of the shielding transfer equipment and the working process of the coiling equipment can be independently carried out, so that the shielding transfer equipment can work continuously, and the recovery working efficiency and the recovery convenience of the detector are improved; the corresponding detector is pulled out, released and coiled under the liquid level, shielding is carried out through the shielding transfer equipment in the transfer process, the problem that the radioactivity of the lower section of the detector is high is effectively solved, and the safety of the site construction environment is improved.

The invention relates to a transportation container for micro-reactor new fuel elements and a loading and unloading method thereof.The transportation container comprises an outer barrel, an inner barrel and an outer cover, the outer barrel and the inner barrel are coaxially arranged to form a double-layer heat insulation structure, a first interlayer of the outer barrel is filled with heat insulation materials, a second interlayer of the inner barrel is filled with air, and double fireproof heat insulation protection is constructed; due to the three-layer sealing design of the outer cover, the first inner cover and the second inner cover, the air tightness is further improved, leakage of the fuel element is prevented, and critical safety is ensured; the container is of a multi-rib-plate and shock absorber structure, the first rib plate, the third rib plate and the second rib plate improve the strength of the barrel and the strength of the outer cover respectively, the shock absorber further enhances the anti-seismic property, and transportation safety is guaranteed. The base is combined with the symmetrical check blocks, the positioning pins and the reinforcers, so that the stability and the bearing capacity of the container are enhanced. The invention further provides a simple, convenient, efficient and safe loading and unloading method which can be widely applied to transportation and storage of micro-reactor fuel elements.

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

The utility model provides a guiding device and an assembling platform comprising the guiding device, the guiding device comprises a lifting driver, a lifting platform and at least one guiding rod vertically arranged on the lifting platform, and the lifting driver is connected with the lifting platform and drives the lifting platform to drive the guiding rod to vertically lift. The guide rod is used for rising from the lower portion and sequentially and freely penetrating through the stacking bottom plate and the stacking elements arranged on the stacking bottom plate so as to limit the stacking elements. According to the guiding device and the assembling platform comprising the guiding device, automatic lifting guiding is achieved, the proper position of the guiding rod can be automatically adjusted according to the height of the multi-layer stacked element, and the guiding efficiency is high. And lifting is stable, and guiding stability is facilitated. The guide rod improves the accuracy and reliability of positioning in a penetrating mode, and the situation that the assembly quality is affected due to the fact that the guide rod collides with the stacked elements can be avoided. Upper and lower stacked elements are guided through the same guide rod, so that the positioning consistency of the upper and lower layers is facilitated.

The invention discloses an online detection device and method for a lead-bismuth reactor oxygen sensor. The online detection device for the lead-bismuth reactor oxygen sensor comprises a lead-bismuth container (10), an electrochemical workstation (20) and a data processing system (30), wherein the lead-bismuth container (10) is filled with liquid lead-bismuth, and an oxygen sensor (40) to be detected is inserted into the lead-bismuth container (10); a reference electrode and a counter electrode of the electrochemical workstation (20) are respectively connected with a reference electrode of the oxygen sensor (40) to be detected, and a sensing electrode and a working electrode of the electrochemical workstation (20) are respectively connected to the liquid lead bismuth through leads (23 and 24). According to the on-line detection device for the oxygen sensor of the lead-bismuth reactor, the running state of the oxygen sensor (40) to be detected is monitored on line through electrochemical impedance spectroscopy according to the response of the oxygen sensor (40) to be detected to alternating voltage applied within a specific frequency range.