1111results about "Reactor fuel elements" patented technology

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 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 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.

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 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.

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.

The invention belongs to the technical field of maintenance of upper reactor internals of a reactor, and particularly relates to a process and a device for maintaining a bolt locking cap of a control rod guide cylinder of an upper reactor internals. Comprising a bolt torque reinspection tool and a locking cap automatic bending tool. The bolt torque reinspection tool is matched with a torque wrench to realize reinspection of bolt torque; the automatic bending tool for the locking cap is used for controlling the bulging claw to correspond to a groove of a locking bolt head and driving the bulging claw to press downwards to bend the locking cap, so that looseness prevention of the bolt is achieved. The device has the beneficial effects that a bolt torque reinspection tool and an automatic locking cap bending tool are conveyed to a working area through a ring crane, and the field environment is not changed; the bolt torque reinspection tool and the locking cap automatic bending tool are both of a long-rod-shaped structure as a whole, an operator can perform remote operation on an operation platform, and whether the torque of the guide cylinder bolt meets the locking requirement or not can be inspected in cooperation with a torque spanner.

The invention discloses a reactor lower internals reactor core supporting plate global foreign matter video inspection grabbing device which comprises a rack, a video inspection device and a foreign matter grabbing device, a positioning pin is arranged at the bottom of the rack, and a containing groove used for storing foreign matter is further formed in the rack; the video inspection device is movably installed on the rack in the height direction and comprises a video inspection mechanism used for conducting foreign matter video inspection. The foreign matter grabbing device is movably installed on the rack in the height direction and comprises a foreign matter grabbing mechanism used for grabbing foreign matter. By applying the video inspection and grabbing equipment for the global foreign matters on the reactor core supporting plate of the reactor lower reactor internals, the foreign matters on the reactor core supporting plate can be inspected and grabbed when the reactor internals are in situ, the damage to the components and a pressure vessel body caused by hoisting of the reactor internals can be effectively avoided, the steps of hoisting of the reactor internals and the like are reduced, and the production efficiency is improved. And the overall foreign matter checking and grabbing operation efficiency can be effectively improved.

The invention discloses a burnable poison loading method for reducing tritium emission of a nuclear power plant and belongs to the field of nuclear power. The method comprises the following steps: providing a first type fuel assembly adopted by a reactor, providing a tritium-reduced burnable poison rod, and loading discrete water-through annular burnable poison or Gd-containing burnable poison in the tritium-reduced burnable poison rod; at least part of the integral burnable poison fuel rods in the first type of fuel assemblies are replaced with tritium-reduced burnable poison rods to obtain second type of fuel assemblies, iterative calculation is carried out to obtain second type of fuel assembly structures meeting safety design requirements, and the second type of fuel assemblies are used for fuel replacement. According to the method, the tritium-reducing burnable poison rod with longer burnup time is utilized, so that the requirement for soluble boron in a primary loop is reduced, and the effect of reducing tritium emission is achieved.

A covering for reinforcing a base layer of a nuclear fuel cladding is provided. The covering comprises a first layer configured to cover a first portion of the outer surface of the base layer of the nuclear fuel cladding, a second layer surrounding the first layer and the base layer and a third layer surrounding the second layer. The first layer comprises a fiber based material, the second layer comprises an interfacing material and the third layer comprises Chromium. A reinforced cladding for nuclear fuel and a method for producing a reinforced nuclear fuel cladding are also provided.

The embodiment of the invention provides a high-temperature gas cooled reactor fuel ball conveying pipeline cutting and jam releasing device, and relates to the technical field of high-temperature gas cooled reactors. The high-temperature gas cooled reactor fuel ball conveying pipeline cutting jam releasing device comprises a cutting function block, the cutting function block is provided with a first end and a second end which are opposite, a first mounting hole penetrating through the first end and the second end is formed in the cutting function block, and the first end is used for abutting against a blocked fuel ball; the cutting assembly is arranged in the first mounting hole and connected with the cutting function block, and the cutting assembly is used for cutting the fuel ball abutting against the first end; and the driving assembly is detachably arranged at the second end, and the driving assembly is used for driving the cutting function block to move. The problem that the blocked fuel balls in the conveying pipeline are inconvenient to release at present is solved, and continuous and stable operation of the high-temperature gas cooled reactor is guaranteed.

The nuclear power station pool foreign matter collecting and transferring device comprises a shielding container and a collecting basket, the shielding container comprises a first containing cavity and is provided with shielding materials capable of shielding rays, and the collecting basket comprises a second containing cavity and can be arranged in the first containing cavity; and drainage structures are arranged on the collecting basket and the shielding container. According to the nuclear power station pool foreign matter collecting and transferring device, after the foreign matter is collected, the foreign matter does not need to be transferred underwater again, the collecting basket can be placed in the shielding container or the foreign matter is directly placed in the shielding container, and the shielding container can shield the foreign matter with high radiation dose; radiation of foreign matters to human bodies and the environment can be reduced, the operation risk of personnel is reduced, and use is simpler and more convenient; the shielding container and the collecting basket can be used in cooperation and can also be used independently, and on-site flexible use is facilitated; the arrangement of the drainage structure facilitates the discharge of radioactive residual water in the foreign matters, so as to reduce the foreign matter collection and transfer weight.

The embodiment of the invention provides a ball path dredging device for a high-temperature gas cooled reactor fuel loading and unloading system, and relates to the technical field of high-temperature gas cooled reactor fuel dredging. The ball path dredging device for the high-temperature gas cooled reactor fuel loading and unloading system comprises a box body, a coiling machine, a gas cylinder set, a tool head set, a filtering piece and a negative pressure piece, the box body is provided with a first installation cavity, a second installation cavity, a third installation cavity and a fourth installation cavity, the coiling machine is installed in the first installation cavity, and the gas cylinder set is installed in the second installation cavity; the tool head set is installed in the second installation cavity, the filtering piece is installed in the third installation cavity, and the negative pressure piece is installed in the fourth installation cavity. Key components such as a coiling machine, a gas cylinder group, a tool head group, a filtering piece and a negative pressure piece are intensively arranged in different mounting cavities in the same box body to form an integrated equipment framework. All the functional units are rapidly connected through preset gas paths, circuits and signal paths, and the operation process is remarkably simplified.

Targetry coupled separation refers to enhancing the production of a predetermined radiation product through the selection of a target (including selection of the target material and the material's physical structure) and separation chemistry in order to optimize the recovery of the predetermined radiation product. This disclosure describes systems and methods for creating (through irradiation) and removing one or more desired radioisotopes from a target and further describes systems and methods that allow the same target to undergo multiple irradiations and separation operations without damage to the target. In contrast with the prior art that requires complete dissolution or destruction of a target before recovery of any irradiation products, the repeated reuse of the same physical target allowed by targetry coupled separation represents a significant increase in efficiency and decrease in cost over the prior art.

The invention discloses a circulating device of a spherical element and a pebble bed high-temperature gas cooled reactor, the circulating device of the spherical element comprises a shell provided with a ball receiving pipe used for receiving the spherical element and a ball outlet pipe used for conveying the spherical element, the axial direction of the ball receiving pipe is consistent with the falling direction of the spherical element, and the axial direction of the ball receiving pipe is consistent with the falling direction of the spherical element. An included angle is formed between the ball outlet direction of the spherical element and the axial direction of the ball receiving pipe, so that the spherical element can be conveyed back to the reactor core along the ball outlet pipe after circulating in the shell; a rotor is coaxially and rotatably arranged in the shell, the rotor is provided with a bearing assembly for bearing the spherical element, and the spherical element can synchronously rotate along with the rotor; the shell is internally provided with the stator fixed relative to the shell, the stator can be matched with the bearing assembly to convey the spherical element to the ball outlet pipe, in this way, after entering the bearing assembly, the spherical element rotates along with the rotor until the stator is matched with the bearing assembly to convey the spherical element to the ball outlet pipe, and circulating conveying of the spherical element is achieved; and the circular conveying stability of the spherical elements is improved.

The invention relates to a nuclear fuel cladding produced with a substrate (14) which is made of pure zirconium or of a zirconium based alloy and a multilayer protective coating (16) which covers a surface (14B) of the substrate (14), the protective coating (16) comprising a main layer (18) made of pure chromium and one or more additional layers (20), each additional layer (20) being made of pure chromium or from a material made of chromium and, additionally, oxygen and / or nitrogen, with the possible presence of unavoidable impurities.

The present application relates to a lifting and transfer device for a radioactive filter element, the lifting and transfer device comprising a grasping mechanism, a displacement mechanism and a transfer mechanism. The grasping mechanism comprises a swinging assembly, a first rotating assembly and a claw, the first rotating assembly being arranged at one end of the swinging assembly, the claw being arranged at the end of the first rotating assembly away from the swinging assembly, and the claw being configured to grasp a filter element handle. The displacement mechanism comprises a lifting component, which drives the claw to rise or lower, and on which the grasping mechanism is arranged. The displacement mechanism is arranged on the transfer mechanism, and the transfer mechanism is configured to transfer the filter element. That is, by means of the rising and lowering of the lifting assembly, the swinging of the swinging assembly and the 360° rotation of the rotating assembly, filter elements at different heights and different inclination angles and filter handles in different orientations can be grasped.

The invention discloses a spherical fuel element lifting device and method.The spherical fuel element lifting device comprises a lifting shell, a lifting body, a detection assembly and a control assembly, and the lifting shell is provided with a spherical fuel element inlet, a spherical fuel element outlet and a spherical fuel element lifting channel; the spherical fuel element inlet and the spherical fuel element outlet are both communicated with the spherical fuel element lifting channel, and a plurality of electromagnetic coils are arranged on the lifting shell in the vertical direction of the lifting shell; the lifting body is movably arranged in a lifting channel of the spherical fuel element; the detection assembly is arranged on the lifting shell and can detect the position of the spherical fuel element; all the electromagnetic coils and the detection assembly are connected with the control assembly. According to the lifting device for the spherical fuel element, the lifting body is lifted by sequentially electrifying the electromagnetic coils, so that the spherical fuel element is driven to be lifted, the lifting efficiency of the spherical fuel element is improved, and collision between the spherical fuel element and the lifting shell can be reduced.

Targetry coupled separation refers to enhancing the production of a predetermined radiation product through the selection of a target (including selection of the target material and the material's physical structure) and separation chemistry in order to optimize the recovery of the predetermined radiation product. This disclosure describes systems and methods for creating (through irradiation) and removing one or more desired radioisotopes from a target and further describes systems and methods that allow the same target to undergo multiple irradiations and separation operations without damage to the target. In contrast with the prior art that requires complete dissolution or destruction of a target before recovery of any irradiation products, the repeated reuse of the same physical target allowed by targetry coupled separation represents a significant increase in efficiency and decrease in cost over the prior art.