203 results about "Tokamak" patented technology

According to the auxiliary supporting mechanism for assembling the internal parts of the Tokamak and the assembling method, all auxiliary stand columns and a center stand column are installed at the same center distance, and the center stand column is provided with an upper-layer disc-type turn buckle supporting mechanism and a lower-layer disc-type turn buckle supporting mechanism so as to be connected with the outer surface of a vertical section of a vacuum chamber ring petal; the outer side supporting mechanism is connected around the outer surface of the vacuum chamber ring valve set, the double-column type bottom supporting table is located below the vacuum chamber ring valve set, and the vacuum chamber ring valve set is supported from the outer side and the lower portion of the vacuum chamber ring valve set; a stand column connecting rod is connected between every two adjacent auxiliary stand columns, and a limiting mechanism capable of extending into a window in a vacuum chamber ring petal is installed on each stand column connecting rod. The U-shaped operation platform penetrates through the vacuum chamber ring petal set, and a plurality of small internal part assembling auxiliary tools are movably installed on the surface of the U-shaped operation platform. The method is wide in application range, can effectively solve the problems that transportation channels of internal parts are few and the assembly construction space is limited, and reduces the construction difficulty and the construction cost.

The present invention provides a multi-degree-of-freedom conductor depositing system for tokamak toroidal field coil winding packs, belongs to the field of development of toroidal field superconducting coils for nuclear fusion. The multi-degree-of-freedom conductor depositing system is mounted on a winding table and distributed along the contour of the toroidal field coil. The optical proximity switches detect the position signals between the winding table and the bending unit and transmit them to an automatic control system. The multi-degree-of-freedom conductor depositing system is distributed in a spiral shape from the top to the bottom from the position of the bending unit along the direction of coil winding. The conductor is deposited onto the track drive mechanism of the multi-degree-of-freedom conductor depositing device, which has a bidirectional moving and rotating function. This avoids the risk of insulation rupture due to friction between the conductor and the depositing tooling during the coil winding process. It also has the advantages of simple structure, convenient operation and high reliability.

The invention relates to the technical field of tokamak device parts, in particular to a supporting system for a polar field coil in a tokamak device. Comprising a toroidal field coil, two first polar field coils, two second polar field coils, two third polar field coils, two first supporting assemblies, two second supporting assemblies and a third supporting assembly, one first supporting assembly is used for fixedly connecting the top end of the toroidal field coil with one first polar field coil, and the other first supporting assembly is used for fixedly connecting the top end of the third polar field coil. And the other first supporting assembly is used for fixedly connecting the bottom end of the annular field coil and the other first polar field coil. Wherein one second supporting assembly is used for fixedly connecting the upper part of the toroidal field coil and one second polar field coil, and the other second supporting assembly is used for fixedly connecting the lower part of the toroidal field coil and the other second polar field coil. And the third supporting assembly is used for fixedly connecting the middle part of the annular field coil and the two third polar field coils.

The invention relates to the technical field of Tokamak nuclear fusion, and discloses an assembling method of a pre-assembled assembly of a nuclear fusion device and an installing method of the nuclear fusion device. The pre-assembled assembly comprises a vacuum chamber, a window neck tube in the vacuum chamber, a vacuum chamber cold shield and a TF magnet; the assembling method of the nuclear fusion device preassembled assembly comprises the following steps: supporting between a cold shield inner sector and a vacuum chamber through a first tool; a first tool is supported between the cold shield outer sector and the vacuum chamber; the TF magnet is connected with the middle window position of the vacuum chamber through a second tool, the upper arc-shaped section of the outer fan section of the cold shield is connected with the hoisting tool of the TF magnet through a third tool, and the lower window section of the outer fan section of the cold shield is connected with the second tool; and the middle window section and the annular field magnet supporting seat are connected through a fourth tool. According to the assembling method of the nuclear fusion device pre-assembled assembly, interference among installation of the vacuum chamber, the window neck tube in the vacuum chamber, the vacuum chamber cold shield and the TF magnet can be avoided, and the stability and functionality of the pre-assembled assembly structure are guaranteed.

Main characteristic of the invention is the exploitation of the very large flux of high energy neutrons generated in a nuclear fusion tokamak with beam-driven plasma for the purpose of producing radioisotopes. It specifically relates to the deployment of precursor materials into the core of a nuclear fusion tokamak with beam-driven plasma for the purpose of exposing these materials to the neutron flux generated in the plasma by the process of nuclear fusion of Deuterium (D) and Tritium (T) to generate radioactive isotopes, so-called radioisotopes. The precursor materials contained in metallic capsules are incorporated into the blankets that surround the plasma.

The invention discloses a concurrent data processing method and system suitable for Tokamak integrated simulation, and relates to the technical field of concurrent data process.The method comprises the steps that one-dimensional data in a Tokamak device is expanded into high-dimensional data based on an isomagnetic surface and geometric factors, and the high-dimensional data are stored in a database; obtaining first coordinate data corresponding to the one-dimensional data in different coordinate systems, wherein the high-dimensional data comprises two-dimensional data and three-dimensional data; based on the isomagnetic surface and the geometric factor, obtaining second coordinate data corresponding to the high-dimensional data in different coordinate systems and dimension reduction data obtained after the high-dimensional data is subjected to magnetic surface average dimension reduction operation; acquiring a data format requirement, and converting the high-dimensional data, the first coordinate data, the second coordinate data and the dimension reduction data into target format data corresponding to the data format requirement; unified conversion, classification and labeling of intermediate physical quantities and final output physical quantities of different calculation codes under the integrated simulation platform and traceability of calculation data codes are achieved.

The invention relates to the technical field of Tokamak plasma diagnosis and data fusion, in particular to a multi-modal fusion Tokamak high-frequency Thomson scattering reconstruction method. According to the technical scheme, the method comprises the following steps: acquiring multi-modal diagnosis data on a Tokamak device, wherein the multi-modal diagnosis data comprises high-frequency auxiliary diagnosis data and low-frequency Thomson scattering data; and performing feature engineering and physical enhancement processing on the high-frequency auxiliary diagnosis data. According to the method, through multi-modal diagnosis data fusion and interpretable KAN network modeling, the Thomson scattering time resolution is greatly improved on the premise that hardware is not transformed, and high-fidelity reconstruction of sub-millisecond transient processes such as ELM and magnetic island evolution is achieved; the method has physical interpretability, can reveal the implicit relationship between diagnostic quantities, enhances the robustness under complex experimental conditions through strategies such as modal loss, and provides key high-frequency diagnostic support for Tokamak high-performance operation and real-time control.

The invention discloses a multi-mode Tokamak device fracture early warning method and system, and belongs to the field of Tokamak device safety protection. The method comprises the following steps: step 1, collecting multi-modal data, and marking fracture data in the collected multi-modal data; 2, time axis alignment is carried out on the collected multi-modal data, time slices are divided, and a fracture early warning database is formed; 3, training the fracture early warning model by using the fracture early warning database, and performing real-time deployment on the optimal fracture early warning model obtained by training; and step 4, in the Tokamak experiment process, inputting 0-dimensional diagnosis data and two-dimensional image diagnosis data which are acquired in real time into the optimal fracture early warning model, and performing fracture relief and device protection according to an output fracture early warning result. According to the method, the fracture state of the Tokamak device can be predicted more accurately, the fracture processing actuator is triggered when the Tokamak device is about to be fractured, and the device is better prevented from being fractured and damaged.

The invention relates to the technical field of magnetic confinement nuclear fusion, and particularly discloses a Tokamak plasma configuration prediction system and method.The Tokamak plasma configuration prediction method comprises the steps that time sequence data of multi-source parameters influencing the Tokamak plasma state are collected through multiple sensors and measuring equipment; a neural network model based on deep learning is adopted to carry out time sequence feature extraction and deep correlation analysis on the multi-source state parameters so as to capture a time sequence correlation change mode among the multi-source state parameters and reveal a dynamic behavior rule of the plasma under the time sequence synergistic effect of the multi-source state parameters; therefore, intelligent configuration prediction of the plasma is realized. In this way, the accuracy and real-time performance of plasma configuration prediction can be effectively improved, and a more reliable control basis is provided for operation of the Tokamak device.

The invention provides a probe power supply over-current protection circuit and an over-current protection method thereof, relates to the technical field of power electronics, and solves the problems that periodic instantaneous over-current and continuous fault over-current are difficult to accurately distinguish and a control signal and a protection system lack a linkage mechanism in the prior art. In the circuit, a current detection unit generates a single over-current trigger signal when detecting a single over-current event; the central control unit sends a blocking signal after receiving the single overcurrent trigger signal, and sends a reset signal after preset time delay; meanwhile, cumulative counting is carried out on single overcurrent trigger signals, and whether an overcurrent fault occurs or not is judged according to a preset threshold number of times; the output control unit is used for controlling the closing and resetting of a probe power supply under the action of different signals; and the state feedback unit feeds back state information to the main control system when an overcurrent fault occurs. According to the invention, accurate response to the ELM overcurrent of the Tokamak device and reliable blocking of persistent faults are realized.

The invention discloses a Tokamak temperature diagnosis reconstruction method and device based on a data-driven model, and belongs to the technical field of machine learning and nuclear fusion physical crossover, and the method comprises the steps: firstly, carrying out the training of a data-driven model based on the fusion of a recurrent neural network and a self-attention mechanism through the large-scale Tokamak experiment data, and carrying out the reconstruction of the Tokamak temperature diagnosis. The complex correlation between various diagnosis signals and electronic temperature diagnosis is learned; secondly, an efficient data processing mode is provided, tokamak original signals are divided into equal-length sub-data in a window division mode, and the data are packaged and stored in an array form after being subjected to data processing so that the data can be directly used for further processing, reading, training and reasoning of subsequent machine learning; finally, the model can be seamlessly integrated into an existing Tokamak data management system by excluding a control reference signal. According to the invention, relatively accurate electronic temperature diagnosis reference can be provided.

The invention discloses an end execution joint of a heavy-load mechanical arm for Tokamak device maintenance, which comprises a first shell, a second shell, a driving mechanism and a detection mechanism, a mounting cavity is formed in the first shell, the driving mechanism comprises a driving assembly and a rotating part, the driving assembly is arranged in the mounting cavity, and the rotating part is arranged in the mounting cavity. The driving end of the driving assembly is connected with the first end of the rotating part, the second end of the rotating part is connected with the second shell, and the driving assembly is configured to drive the rotating part to rotate by a preset angle so as to drive the second shell to rotate; the detection mechanism is configured to detect the rotating angle of the rotating part and control the driving assembly to work or stop according to the detection result. According to the heavy-load mechanical arm for tokamak device maintenance, the tail end execution joint is designed in a modularized mode, the tail end execution joint is simple in structure, the occupied space of the joint is reduced, and the interior of the tokamak device can be conveniently maintained.

The invention discloses a Tokamak plasma multi-parameter cooperative configuration analysis method, and relates to the technical field of plasma multi-parameter cooperative configuration analysis, and the method comprises the following steps: constructing a parameter database to store plasma parameters; performing normalization processing on the plasma parameters to obtain normalized parameters of the plasma parameters; whether correlation exists between the two plasma parameters or not is analyzed, and an independent incidence relation is obtained; analyzing whether the plasma parameters are simultaneously influenced by more than two plasma parameters, and analyzing a comprehensive association relationship between the plasma parameters; when the configuration of the plasma is regulated and controlled, plasma parameters are cooperatively regulated and controlled based on the independent incidence relation and the comprehensive incidence relation; the plasma multi-parameter cooperative configuration analysis method is used for solving the problem that better cooperative control and analysis cannot be carried out on plasmas due to the fact that correlation analysis of different plasma parameters is not comprehensive enough in an existing plasma multi-parameter cooperative configuration analysis technology.

The invention discloses a tokamak inner cold shield steady state test device and method, the tokamak inner cold shield steady state test device comprises a test box and an inner cold shield arranged in the test box, and the bottom of an inner cavity of the test box is provided with a movable base. According to the Tokamak inner cold screen steady state test device and method, the inner cold screen is supported through the bearing assembly, the inner cold screen is transferred into the test box under the arrangement of the movable base, the accurate positioning of the inner cold screen is realized in cooperation with the matched arrangement of the steering driving assembly and the positioning assembly, and the steady state of the inner cold screen is tested in cooperation with the arrangement of the end face plugging assembly. The inner cold screen opening end face is plugged, vacuum helium leak detection of the inner cold screen can be carried out, operation is simple and convenient, meanwhile, efficient vacuum helium leak detection of the inner cold screen can be achieved, after a test is completed, the inner cold screen can be stably transferred to the outside of the test box, good adaptability is achieved, and meanwhile efficient and stable detection of the inner cold screen is achieved.

The invention discloses a Tokamak cold shield assembling system of a liquid nitrogen rapid cooling pipeline, and relates to the technical field of Tokamak cold shield combined machining. The Tokamak cold shield assembling system of the liquid nitrogen rapid cooling pipeline comprises a stepping motor, and a rotating part of the stepping motor is connected with a first limiting plate and a second limiting plate; a rotating part of the gear motor is connected with a third limiting plate and a fourth limiting plate, and the two ends of the third limiting plate and the two ends of the fourth limiting plate are connected with placing plates; an inner containing cavity is formed between the second limiting plate and the third limiting plate, and an outer containing cavity is formed between the first limiting plate and the fourth limiting plate. According to the Tokamak cold shield assembly system of the liquid nitrogen rapid cooling pipeline, the inner cold shield petal plate and the outer cold shield petal plate are placed through the placement rotating piece and rotate to enter and exit from the installation position, the rotating piece is placed through the rotation limiting piece in a matched mode, and the inner cold shield petal plate and the outer cold shield petal plate are limited; and assembling of the inner cold screen petal plate or the outer cold screen petal plate is completed at the two ends of the vacuum chamber petal plate.

The invention discloses a multi-stage isolation high vacuum obtaining system of a neutral beam injection device and a control method, and relates to the field of nuclear fusion Tokamak vacuum pumping design. The low-temperature pump unit comprises a plug-in low-temperature pump in a neutral beam vacuum chamber, and the plug-in low-temperature pump is used for refrigerating and cooling liquid nitrogen and liquid helium to adsorb gas to realize vacuum pumping; the low-temperature pump preceding-stage air extractor unit comprises a molecular pump unit, a roots pump unit and a mechanical pump unit; an inlet of the molecular pump set is connected with the neutral beam vacuum chamber through a gate valve, and an outlet of the molecular pump set is connected with an inlet of the roots pump set through a forestage valve. An inlet of the roots pump set is connected with the neutral beam vacuum chamber through a preceding stage valve and a pre-pumping valve, and an outlet of the roots pump set is connected with the mechanical pump set through the preceding stage valve. By adopting a multi-stage vacuum shielding technology and a step-by-step start-stop strategy, step-by-step transition from low vacuum to high vacuum and from high vacuum to low vacuum can be efficiently, stably and reliably realized, and an air extractor unit and a vacuum environment are protected from being polluted.

The invention discloses a Tokamak device bottom cold shield cooling pipeline weld joint surface treatment device and a Tokamak device bottom cold shield cooling pipeline weld joint surface treatment process, and relates to the technical field of pipeline weld joint surface treatment equipment.The Tokamak device bottom cold shield cooling pipeline weld joint surface treatment device comprises a grinding table and a first linear module installed at the top of the grinding table, and a supporting seat is installed at the top of the first linear module; a support is fixedly connected to the top of the supporting seat, a pipeline is placed on the top of the support, a fixing seat is fixedly connected to one side of the supporting seat, and a three-jaw chuck is installed on one side of the fixing seat. By arranging the angle adjuster, the angle of the grinding disc can be adjusted in the moving process of the moving seat, so that the inclination angle of the grinding disc is gradually increased in the process that the grinding disc grinds the position of a pipeline welding seam and moves towards the outermost edge of the welding seam, and the edge of the welding seam can be slightly trimmed and polished; and the removal amount of the material is controlled, so that the overall practicability of the device is improved, and then the grinding precision is improved.

The invention relates to the technical field of nuclear fusion, and discloses an ultrasonic molecular beam injection system for a fusion device.The ultrasonic molecular beam injection system comprises a gas source, a pressurizing unit, a storage unit and an injection unit, the pressurizing unit comprises a gas conveying pipeline, and the two ends of the gas conveying pipeline communicate with the gas source and the storage unit correspondingly; a booster pump is arranged on the gas conveying pipeline and used for pressurizing gas in the direction away from the gas source, and the storage unit can receive and store the pressurized gas so that the gas can reach a preset pressure value; the injection unit comprises an injection pipeline and a nozzle, one end of the injection pipeline is communicated with the storage unit, the other end of the injection pipeline is provided with the nozzle, the nozzle is used for stretching into a Tokamak vacuum chamber of the fusion device, and the injection pipeline can convey gas in the storage unit to the nozzle so as to form an ultrasonic molecular beam current through the nozzle. According to the invention, the injection function of the ultrasonic molecular beam in a large fusion device can be realized.

The invention provides a design method and device for a biological shielding wall of a fusion device and electronic equipment. The design method of the biological shielding wall of the fusion device comprises the steps that a biological shielding occupying model is segmented in the height direction to obtain a plurality of biological shielding occupying modules, and the biological shielding occupying modules at least comprise top shielding modules used for arranging embedded parts; the bottom shielding module is used for arranging the bottom of the Tokamak Dewar base and an annular support; based on the performance indexes of the biological shielding wall, performing structural design on each biological shielding occupation module to obtain a biological shielding engineering module; and dividing each biological shielding engineering module in the circumferential direction to obtain a plurality of prefabricated sub-modules, and obtaining the biological shielding wall formed by splicing the plurality of prefabricated sub-modules. According to the method, on-site high-altitude, high-strength and long-time complex operation is reduced, the construction safety risk is reduced, and modular removal of the biological shielding wall and waste treatment are more efficient.

The invention discloses a Tokamak plasma magnetic response model construction method, a Tokamak plasma magnetic response model construction system and a medium. Relates to the technical field of nuclear fusion plasma simulation. According to the scheme, improvement is carried out on the basis of the prior art, modeling is carried out on the magnetic response of the Tokamak plasma on the basis of basic data of the Tokamak plasma, and plasma behaviors are directly learned from experimental data; the simulator taking the plasma magnetic configuration evolution autoregression model as a carrier can process nonlinear terms and high-order terms, and is low in calculation cost, high in speed and capable of being applied in real time; according to the method, the behavior of the plasma under magnetic control can be well simulated, the plasma behavior is directly learned from experimental data without simplifying and assuming plasma properties, and the method can be used for off-line simulation and can also be used for control strategy optimization or fault detection in real time.

The invention discloses a superconducting coil, a Tokamak device applying the superconducting coil and a superconducting coil stacking process, the superconducting coil comprises a plurality of layers of sub-unit coil stacks, each layer of sub-unit coil is provided with two coil joints, and the joints of the adjacent layers are connected in a staggered manner to form a spirally rising conduction structure. The stacking process takes a leveling support platform as a reference, and coil posture adjustment is realized through a replaceable clamp: a first clamp is used for fixing a coil and detecting a contour and a joint position, and a second clamp is used for replacement after the coil is qualified; during hoisting, positioning is carried out through a characteristic line alignment technology, the profile is adjusted through a shape maintaining tool or the angle is finely adjusted through a circumferential adjusting device, and accurate alignment between layers is ensured; the circumferential adjusting device corrects the angle deviation through the telescopic column. According to the scheme, through multi-stage precision control, assembly precision control over stacking of the superconducting coils is achieved, the problems of interlayer dislocation and poor conduction of the large-size coils are solved, and the magnetic field uniformity and operation stability of the Tokamak device are improved.

The invention discloses a cooperative control method for improving Tokamak radio frequency wave current driving locality, and the method mainly comprises the following steps: S1, according to a radial position where locality needs to be enhanced, heating narrow resonance layer electrons at the corresponding radial position through electron cyclotron resonance heating (ECRH); s2, ion cyclotron resonance heating (ICRH) is adjusted for regional heating, so that the radial position of an ICRH driving current main peak value coincides with the radial position of an ECRH current peak value; s3, using GENRAY and CQL3D programs to simulate low hybrid wave (LHW) and high harmonic fast wave (HHFW) combined current driving, calculating a momentum region corresponding to a synergistic effect in a momentum space, performing correlation analysis on the synergistic momentum region, wave emission parameters and current radial distribution, feeding back the parameter adjustment direction, and performing repeated iteration until required local current distribution is achieved; according to the invention, current driving with consideration of driving efficiency and locality control is realized, and the method is of great significance in enhancing the current distribution control capability, improving the constraint performance of plasma and obtaining longer constraint time and higher fusion power.

The invention discloses a Tokamak plasma rupture prediction method based on cWGAN and TCN, which takes AUC as a measurement index to evaluate the effectiveness of a model, and adopts a conditional Wasserstein generative adversarial network (cWGAN)-based minority class sample generation method for solving the problem that the number of minority class samples is limited. Gradient penalty (GP) is introduced in a training stage to improve the quality of generated samples and training stability, and an inverse frequency mini-batch minority class equilibrium training strategy is adopted to enable a generator to fully learn distribution characteristics of broken samples, so that high-quality minority class samples are generated and a training set is expanded. A Transform structure is introduced on the basis of a time convolutional network (TCN), local time sequence features are efficiently extracted by using the TCN, and the Transform enhances the expression ability of a model for a long-time dependency relationship and global features, so that the prediction effect is improved.

The invention provides an assembly construction method of a biological shielding wall of a fusion device and the biological shielding wall. The assembly construction method of the biological shielding wall of the fusion device comprises the following steps: manufacturing a wall body framework of a prefabricated sub-module; the wall body frameworks of the prefabricated submodules are assembled in the circumferential direction to obtain an annular framework of the biological shielding wall module, and the biological shielding wall module at least comprises a top shielding module used for arranging embedded parts and a bottom shielding module used for arranging the bottom of a Tokamak Dewar base and an annular support; and the annular framework is hoisted to a pouring foundation, and the interior of the annular framework is poured after hoisting is completed every time. According to the method, on-site high-altitude, high-strength and long-time complex operation is reduced, and construction safety risks are reduced. In addition, if internal parts need to be replaced or overhauled in the future, corresponding structures on the specific prefabricated submodules can be directionally dismantled, and when the service life of the fusion device is over, modular dismantling of the biological shielding wall and waste treatment are more efficient.

The invention relates to the technical field of Tokamak magnetic probe fault detection, in particular to a real-time fault detection method for a Tokamak magnetic probe. According to the technical scheme, the method comprises the following steps that neural network model training data are collected to construct a training database, and physical parameters related to magnetic probe signals serve as model input; a neural network model with multiple coding layers and decoding layers is adopted to capture the evolution characteristic of the magnetic probe signal in time, and the magnetic probe signal in the subsequent time period is predicted based on the input time period; and inputting and transmitting the model to a trained neural network model to predict a magnetic probe signal in real time. According to the method, the fault detection of the magnetic probe diagnosis in the Tokamak device is realized by using the artificial intelligence algorithm, the calculation precision of the model is improved, and the calculation time of the model is greatly optimized, so that the real-time fault detection of the magnetic probe becomes feasible, and the method has relatively high universality.

The invention relates to the design, configuration and operation concept of a system, herein called radioisotope recovery system, to deploy and recover different precursor materials into and from the core of a nuclear fusion tokamak with beam-driven plasma for the purpose of exposing these materials to the neutron flux generated in the plasma by the process of nuclear fusion of D and T ions, which will cause the generation of radioactive isotopes (also called radioisotopes). Main characteristics of the invention are (i) a closed loop that circulates precursor materials into and out of a nuclear fusion tokamak with beam-driven plasma, (ii) the compliance of the system with the design and maintenance of a nuclear fusion tokamak with beam-driven plasma.

The invention discloses a magnetic field enhanced electromagnetic launching device for relieving Tokamak fracture, and belongs to the technical field of electromagnetic track launching. The invention designs a magnetic field enhanced track structure aiming at the problems of how to enhance the electromagnetic force of a projectile, improve the speed of an armature, separate the projectile from the armature and the like. Comprising an electromagnetic acceleration track and an electromagnetic deceleration track which are arranged in parallel, the structures of the electromagnetic acceleration track and the electromagnetic deceleration track are in mirror symmetry, and each of the electromagnetic acceleration track and the electromagnetic deceleration track comprises a reinforced track and a through track; a through rail of the electromagnetic acceleration rail makes contact with an armature and is used for launching the armature, an enhancing rail is used for enhancing the in-chamber magnetic field intensity and increasing armature stress, and electromagnetic force is used for propelling the armature carrying the projectile to reach the set speed. A through rail of the electromagnetic speed reduction rail makes contact with the armature and is used for generating reverse acceleration through electromagnetic force to act on the armature so that the armature can be decelerated, and the reinforcing rail is used for enhancing the in-chamber magnetic field intensity, increasing armature stress and enabling the armature to be separated from the projectile.

The invention relates to a method for solving the problem that neutrons impact the inner wall of a reactor in the operation process of a Tokamak device, which can be called as a circulating liquid protection structure. The method specifically comprises the following steps: embedding two annular space sprayers into the inner wall above a middle column of a reactor, continuously spraying high-temperature-resistant liquid to the inner wall, enabling a layer of liquid continuously flowing downwards to be attached to the surface of the inner wall of the whole reactor under the action of gravity, and enabling the liquid to pass through a hole formed in the bottom of the inner wall of the reactor and flow into a liquid storage tank below the device; and then the liquid in the liquid storage tank is pressed into the two annular space nozzles by a pump through an external pipeline and is sprayed to the periphery. Therefore, the high-temperature-resistant liquid which circularly flows in the completely-closed vacuum structure can cover the surface of the inner wall of the whole reactor and continuously resist impact from neutrons, so that the inner wall of the reactor is fundamentally protected.

The invention relates to the application field of a negative neutral beam ion source system in a magnetic confinement nuclear fusion Tokamak device, in particular to a supporting and adjusting structure and method for a beam source of a negative neutral beam system in a fusion device. According to the technical scheme, the device comprises a beam source device and a stainless steel flat plate base located below the beam source device, and further comprises a welding corrugated pipe which is fixedly installed at one end of the beam source device and used for connecting the beam source device with a bunching vacuum chamber. The beam source three-dimensional space adjusting function is achieved through the characteristic that the welding corrugated pipe is small in rigidity, the position of the beam source can be accurately adjusted through the three-dimensional jacking jack, and therefore the position of a beam spot of a beam can be accurately adjusted, the anion beam spot can be kept at the center of a target, and the problem that the beam spot of an ion beam in a negative neutral beam system in a fusion device is difficult to adjust is solved. And an acting point is provided for the three-dimensional jacking jack through the arc-shaped supporting piece, stress concentration is reduced, point discharge is prevented, and safety in the high-voltage environment is guaranteed.

The present invention discloses a method, device and medium for constructing a compact copper conductor tokamak, which relates to the technical field of index evaluation. By obtaining the evaluation indexes of the tokamak device, a quantitative evaluation system is constructed; based on the device operation conditions, the parameter thresholds of the evaluation parameters in the quantitative evaluation system are determined; based on the parameter thresholds, the major radius and minor radius thresholds of the compact device are determined; according to the major radius threshold and minor radius threshold, the boundary of the plasma is determined; and based on the plasma boundary, the compact device is constructed. By establishing a quantitative index system for the comprehensive engineering technology capabilities of the tokamak, the ranges of core parameters such as the major radius and minor radius of the plasma can be quickly determined, the dimension chain can be determined more accurately, the conflict between physical requirements and engineering implementation caused by large dimension chain errors can be reduced, and the engineering capabilities can meet the standards when the device is miniaturized and compactly designed.