79 results about "Tokamak" patented technology

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 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 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 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 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 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 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.

This invention discloses a tokamak plasma turbulence prediction method and system based on layer-theoretic topology and spectral neural operators. The method includes acquiring diagnostic and control data from a tokamak device and converting them into a set of physical field vectors, actuator vectors, and equilibrium parameter vectors; mapping the physical field vector set to a latent grid representation and fusing the actuator and equilibrium parameter vectors to obtain a conditional latent grid representation; using a multi-layer Fourier neural operator for time evolution to obtain the latent grid representation for the next time step; applying multi-region consistency constraints to the latent grid representation using a constraint mapping operator and a glue loss function; and using the constrained predicted latent grid representation for prediction, outputting the hazard level and remaining time prediction values ​​of plasma turbulence. Simultaneously, a projection and repair module applies physical constraints to the restored physical field prediction representation to obtain a corrected physical field prediction representation. This invention improves the accuracy of the results.

The invention discloses an auxiliary coil positioning device, and belongs to the technical field of nuclear power assembly.The auxiliary coil positioning device is used for assembling a coil and a wire box in a Tokamak device, an assembling hanger of the auxiliary coil positioning device comprises a connecting assembly and a guiding and limiting assembly, and the connecting assembly comprises a hanging part and a mounting part which are connected; the mounting part is detachably hung on the outer edge of the wire groove, the mounting part is attached to the inner side wall of one side of the wire groove, the guiding and limiting assembly comprises a guiding part, a connecting part and a limiting part, the connecting part is mounted on the mounting part, and the guiding part and the limiting part are both connected with the mounting part. According to the coil auxiliary positioning device, the size of the limiting piece can be correspondingly adjusted according to the size error of the contact position of the coil and the guiding piece, so that when the coil has the size error, the assembling precision of the coil and the wire box is improved, the magnetic field precision of a Tokamak device is improved, and stable constraint of the Tokamak device on plasma is guaranteed.

The application discloses a boundary plasma simulation method under a unit magnetic surface coordinate system and belongs to the technical field of magnetic confinement tokamak boundary plasma simulation. Firstly, a physical model is expressed by using a unit magnetic surface coordinate system, physical quantity and geometric effect are separated by using unitized base vectors, and singularity near an X point is avoided; secondly, a toroidal Fourier series representation is combined with difference along a magnetic field line, problem dimension is reduced, and numerical dissipation is reduced; thirdly, a generalized finite difference method and a transformed coordinate system are adopted to process the X point and adjacent points, and the X point can be directly contained in a grid; finally, a grid screening method is used to pre-calculate geometric quantities and difference coefficients under high-density grids, and the calculation precision of geometric effect is improved. The application can accurately simulate the influence of the X point on the stability of the boundary plasma, significantly reduces the consumption of computing resources, and improves the simulation efficiency and accuracy.

The application discloses a kind of nuclear fusion adaptive control system and method based on reinforcement learning, the control system includes physical layer and control layer, physical layer contains tokamak / stellarator device, Thomson scattering and so on diagnostic system and FPAD / PAM projectile injection system, responsible for collecting plasma density profile, temperature profile and other data and executing fuel supply;Control layer is reinforcement learning intelligent agent, contains state perception, strategy network and execution interface module, strategy network uses PPO / SAC algorithm, state perception is included line average density, density change rate and historical sequence data;The control method is trained offline by Torax simulator, is optimized by composite reward function, forms perception-decision-making-execution-feedback closed-loop control, supports real machine deployment and online fine tuning.The application solves the problem of traditional PID control lag, adaptive difference, the control precision is improved by about 40%, density fluctuation is reduced by about 50%, prolongs hardware maintenance cycle, adapts nuclear fusion long pulse operation demand.

A modular tokamak is disclosed. The modular tokamak according to the present invention may comprise a plurality of core chambers sequentially connected in the longitudinal direction and may also comprise extension chambers each connecting two adjacent core chambers.

The application discloses a tokamak discharge modeling system based on a bidirectional long short-term memory neural network, which is composed of multiple modules, and the modules are low-coupling functional units.The modules at least include a data transfer module, a Batch data access input module, a training self-defined parameter module, a bidirectional long short-term memory neural network modeling module and a data visualization module.The whole system architecture is mainly divided into two paths, namely data training and data modeling.Using the visualization technology, the tokamak discharge experiment personnel can refer to the discharge modeling results and visualize the model training process.The application can one-key type in the experiment proposal stage to model the whole process of the tokamak discharge curve in advance, so that the experiment personnel and the proposal design personnel can check the validity and rationality of the proposal.The application can also be used for assisting data discovery after the experiment.

The application discloses an electromagnetic emission armature recycler for a tokamak device. The recycler comprises a vacuum exhaust port, an armature inlet track, an armature recovery cavity, an armature transmission track, a spiral conveying rod, a driving motor, a recovery sand storage tank and a plurality of sealing devices. The vacuum exhaust port discharges air in the electromagnetic emission armature recycler. The sealing devices are arranged at each connection. One end of the armature inlet track is connected with the tokamak device to receive the armature, and the other end is connected with the armature recovery cavity. The armature recovery cavity is filled with buffer sand. One end of the armature transmission track is connected with the lower part of the armature recovery cavity, and the spiral conveying rod is arranged in the armature transmission track. One end of the spiral conveying rod is connected with the armature motor and rotates along with the driving motor to drive the mixed armature and buffer sand to move along the armature transmission track. The recovery sand storage tank is connected with the lower part of the other end of the armature transmission track to recover the armature and buffer sand. The armature can be efficiently recovered in a vacuum environment.

The application relates to a teaching laser speckle interferometry experimental device for measuring material micro-morphology, belonging to the technical field of real-time diagnosis of material morphology of a tokamak wall. The device comprises a laser, a beam expander, a collimating mirror, a sample holder, two groups of beam splitters, a piezoelectric phase-shifting mirror, an imaging lens, a CCD camera and an optical platform. The device has a simple structure, the sensitivity can be dynamically expanded and selected according to actual detection conditions, and speckle interference fringes are obtained by subtracting the intensity field distribution functions before and after the surface displacement or deformation of the measured object. Compared with traditional morphology detection methods, physical contact with the measured object is reduced, thereby avoiding secondary deformation of the measured object and causing system errors.

The application discloses a kind of particle evolution simulation methods based on multi-component plasma radiation effect in tokamak, belong to magnetic confinement controlled nuclear fusion technical field.Firstly, the evolution of impurity density distribution, electron density distribution and ion density distribution is calculated, then continuous radiation and linear radiation are calculated, and then a variety of plasma radiation is accurately added inside magnetic island.Calculating the magnetic field configuration under the plasma radiation correction, and continue to calculate the evolution of plasma total radiation and related parameters under advanced magnetic field configuration, so repeated calculation, achieve the simulation effect of long time evolution.The application realizes the calculation of impurity, electron and ion real-time dynamic distribution, and can also calculate the evolution of a variety of plasma radiation under real three-dimensional magnetic field configuration, and then obtain the plasma related parameter profile at any time;Meanwhile, the spatial distribution of plasma radiation and the situation of magnetohydrodynamic instability under advanced magnetic field configuration are more accurately described.

The invention discloses a Tokamak plasma fracture grading protection method and system, and the method achieves the combined prediction of the fracture occurrence probability and the hazard degrees of different hazard types through a multi-task deep learning model, can achieve the early warning of the fracture risk, and also can achieve the prediction of the hazard degrees of halo current, thermal load and escape current. Then, attribution analysis is carried out on the multi-modal diagnosis signals, key features in the multi-modal diagnosis signals are determined, then fracture inducements are determined based on the key features, and remaining time windows from the current time node to fracture occurrence are determined; in the stage of hierarchical protection decision making, in combination with reversibility determined based on fracture inducement, flow reduction stability determined based on the damage degree of each fracture damage type and flow reduction adequacy determined based on a remaining time window from a current time node to fracture occurrence, inhibition means with different intensities are constructed in a targeted manner, excessive response is avoided, and the reliability of the system is improved. Therefore, the operation efficiency of the Tokamak device is maintained to the maximum extent.

A magnetic field sensor for use in a tokamak power plant is disclosed. The sensor comprises a Hall sensor 4 made of a radiation-hard semiconductor such as diamond or silicon carbide with a doped regio

The invention discloses a Tokamak plasma feedback control system and method based on real-time calculation of a helium-deuterium ratio, and the system consists of an automatic spectrum collection module, a real-time calculation module of the helium-deuterium ratio, and a data transmission and feedback control module. And continuous framing acquisition of emission spectrum signals including helium spectral lines and deuterium spectral lines in the whole discharge process is realized. And a helium-deuterium ratio real-time calculation algorithm is embedded, smooth filtering, baseline correction and numerical integration are carried out on a preset helium and deuterium characteristic spectral line narrow wavelength window in each frame of spectrum, corresponding spectral line integration intensity is obtained, and the He / (He + D) fuel component ratio is calculated on line. A calculation result is read by a plasma control system and then input into a PID controller, the opening degree of a helium and deuterium inflation valve is automatically adjusted, closed-loop control over the boundary helium and deuterium ratio in the discharging process is achieved, and energy constraint deterioration and fusion performance reduction caused by too high He / D are restrained.

The application relates to the technical field of nuclear fusion, and discloses a connecting assembly of a tokamak device, which comprises a connecting bolt, a first connecting sleeve, a second connecting sleeve, a first spherical gasket, a second spherical gasket, a connecting nut and a limiting sleeve. The connecting bolt is in screw joint cooperation with the connecting nut, and the connecting nut can rotate around a first axis at least. The first connecting sleeve, the second connecting sleeve, the first spherical gasket and the second spherical gasket are arranged on the outer circumferential side of the connecting bolt and are arranged in a spaced mode. A first spherical surface is formed on the side of the first connecting sleeve away from the connecting nut, a second spherical surface is formed on the side of the second connecting sleeve facing the connecting nut, the first spherical gasket is arranged on the side of the first connecting sleeve away from the connecting nut, and the second spherical gasket is arranged on the side of the second connecting sleeve facing the connecting nut. The connecting assembly can reduce the shearing force borne by the connecting bolt and improve the reliability of the connecting structure between the vacuum chamber and the cladding.

The invention provides a process door for a Tokamak device boundary main equipment channel, and belongs to the technical field of nuclear fusion process doors, the process door comprises two vertical steel frame modules, top transverse steel frame modules are welded to the tops of the vertical steel frame modules, and process door modules are welded to the close sides of the two vertical steel frame modules; metal door leaves are arranged on the close sides of the two process door modules in a sliding mode, a hoisting module is installed on the top of the top transverse steel frame module through bolts, each process door module comprises an installation guide rail, the outer side of each installation guide rail is welded to the inner side of the corresponding vertical steel frame module, and two rectangular connecting flanges are arranged on the front side of each installation guide rail. And double inflatable sealing rings are arranged between the rectangular connecting flanges. After the metal door leaf is closed, the pressing device can drive the rectangular connecting flange to be tightly attached to the sealing face of the door leaf, and the sealing gap can be fully filled through the double-channel inflation sealing ring under the double effects of extrusion deformation and inflation expansion.

The application discloses a kind of tokamak plasma multi-parameter collaborative configuration analysis method, it is related to plasma multi-parameter collaborative configuration analysis technical field, including the following steps: constructing parameter database to store plasma parameter;Plasma parameter is normalized, and the normalized parameter of plasma parameter is obtained;Analysis whether there is relevance between two plasma parameters and obtains independent correlation;Analysis whether plasma parameter is simultaneously influenced by more than two plasma parameters, and analysis their comprehensive correlation;When the configuration of plasma is regulated, based on independent correlation and comprehensive correlation, plasma parameter is cooperatively regulated;The application is used to solve the existing plasma multi-parameter collaborative configuration analysis technique still exists between different plasma parameters The relevance analysis is not comprehensive enough, leading to unable to better cooperatively control and analyze plasma.

The invention relates to a magnetic coil arrangement (1) having at least one magnetic coil (MS1, MS2, MSN), in particular a stellarator coil or tokamak coil for a nuclear fusion system, wherein: the magnetic coil (MS1, MS2, MSN) comprises a superconducting conductor and one or more annular plate layers which are made of electrically conductive material and in each of which a plurality of windings of the superconducting conductor, which extends along a local longitudinal direction, are accommodated; the windings are arranged in succession in a local transverse direction; and the superconducting conductor of the magnetic coil (MS1, MS2, MSN) is part of a superconducting circuit of the magnetic coil arrangement (1), characterized in that at least one of the plate layers is part of an additional current path (17) which is connected in parallel with respect to at least one portion of the superconducting conductor and, with this, forms a discharge circuit (20), and in that a first switching element (S1) for interrupting the superconducting circuit is present in the superconducting circuit, wherein the switching element (S1) is arranged outside the discharge circuit (20). By means of the magnetic coil arrangement according to the invention, when local heating of the conductor occurs, firstly the magnetic coil can be quickly discharged and secondly the voltage prevailing across the conductor and relative to ground can be limited to an acceptable level.

The application discloses a coil auxiliary positioning device, and belongs to the technical field of nuclear power assembly. The coil auxiliary positioning device is used for the assembly of a coil and a coil box in a tokamak device. The assembly hanging frame of the coil auxiliary positioning device comprises a connecting assembly and a guiding and limiting assembly. The connecting assembly comprises a hanging piece and a mounting piece which are connected. The hanging piece is detachably hung on the outer edge of a coil slot. The mounting piece is attached to the inner side wall of one side of the coil slot. The guiding and limiting assembly comprises a guide piece, a connecting piece and a limiting piece. The connecting piece is installed on the mounting piece. The guide piece and the limiting piece are connected to the mounting piece. According to the coil auxiliary positioning device, the size of the limiting piece can be adjusted according to the size error of the contact position of the coil and the guide piece. When the coil has a size error, the assembly precision of the coil and the coil box is improved, so that the magnetic field precision of the tokamak device is improved, and the stable confinement of the plasma is ensured.

The invention relates to the technical field of nuclear fusion, in particular to a spherical Tokamak feeding system based on ultrasonic molecular beam injection. Comprising the steps that a gas tank, a hydrogen tank, a pressure reducing valve, a needle valve and the like form a common gas supply loop to serve as an initial feeding channel so as to meet the feeding requirement of daily operation, and an ultrasonic molecular beam injection gas path is additionally arranged on the basis to serve as a secondary supplementing feeding channel in the plasma discharging process and used for improving the plasma density; and compared with a traditional air supply mode for secondary feeding, the response time is shorter, and the feeding efficiency is higher. Finally, the ultrasonic molecular beam injection gas path and the initial feeding channel are connected through a gas pressure measuring assembly so as to monitor the gas pressure of the two gas paths; based on the air pressure of the two air paths, the amount of air fed every time can be accurately measured. According to the invention, staged fuel gas injection can be carried out on the spherical tokamak, the injection efficiency of the fuel gas is greatly improved, and the wall circulation is reduced.

The invention discloses an E / / B neutral particle analyzer detector for a Tokamak device, which belongs to the technical field of high-temperature plasma diagnosis and comprises a magnet structure for generating a specific magnetic field structure and a vacuum chamber structure for passage of charged ions. The electrode plate structures are used for guiding the charged ions to deflect; the detector main body structure is used for receiving the charged ions; the magnet structure is vertically fixed on a plane, the vacuum chamber structure is fixed and tightly attached in a gap of the magnet structure, the electrode plate structure is fixed in the vacuum chamber structure, and the detector main body structure is fixed in the vacuum chamber structure. While the target magnetic field structure and the energy detection range are met, the overall size and mass of the device are effectively optimized.

The application relates to the field of new energy and energy-saving technology, in particular to a front mixed water heat exchange device of an improved TOKAMAK test system, which comprises a mixed water heat exchange tank and a jet pump, a rotating injection water distribution mechanism, a shunt cylinder and a heat exchange inclined fin are arranged in the innermost sleeve of the mixed water heat exchange tank. The application is based on the original chilled water and chilled water circuit, a balance cooling water injection circuit is branched from the chilled water, the end pressure of the chilled water main circuit is balanced, the cooling water is pre-cooled, an auxiliary circuit is branched from the cooling water to sufficiently absorb the cold energy of the chilled water in the injection circuit, and the three circuits of the main cooling circuit of the mixed cooling water and chilled water are cooled and mixed through the improved mixed water heat exchange tank. The concentrated water flow of the original device into the water tank is divided into multiple water flows through the triangular weir shunt cylinder, the noise in the operation of the original device is reduced, the reliability of the original device is increased, the mixed water efficiency is improved, and the device is convenient to maintain.

The invention discloses a method for measuring the service life of a first wall boron film of a Tokamak device, and belongs to the technical field of high-temperature plasma diagnos.The method comprises the steps that in the operation stage of a Tokamak plasma experiment, the mass change of a material is monitored in real time through a quartz crystal microbalance system located in a magnetic shadow area; based on monitoring data of the quartz crystal microbalance system, an evolution curve of accumulated mass change along with accumulated discharge time is drawn; performing local fitting and slope inspection on the slope of the evolution curve of the cumulative mass change along with the cumulative discharge time, and when the local fitting slope is remarkably changed, extracting plasma parameters under corresponding discharge conditions and confirming that the plasma parameters are in a stable state; and judging the end-of-life point of the first wall boron film based on the slope characteristic of the evolution curve of the cumulative mass change along with the cumulative discharge time. According to the invention, in-situ, real-time and accurate life measurement can be realized in the plasma discharge process.

The application relates to the technical field of fusion device installation, and discloses an installation method of an SSAT tool for a tokamak, which comprises the following steps: embedding a rigid component on the ground; installing a first central column and a guide rail assembly on the rigid component; installing a first cantilever beam on the top of the first central column; installing a second central column on the top of the first central column, and arranging a plurality of second cantilever beams on the second central column in an axial direction; installing an outer side column assembly on the side of the guide rail assembly far from the first central column; installing a connecting cross beam between the second central column and the outer side column assembly; installing magnet support columns on the opposite sides of the guide rail assembly; installing a cold screen support column on the side of the guide rail assembly far from the first central column; and installing a rotating frame on the guide rail assembly, and the rotating frame is connected with the cantilever beam assembly. The installation method of the SSAT tool for the tokamak can meet the final installation precision requirement of the SSAT tool.

The invention relates to the technical field of deuterium-tritium fusion fuel measurement, in particular to a laser spectrum and mass spectrum combined fuel retention in-situ measurement method. Aiming at the problems of incapability of real-time in-situ monitoring, limited measurement precision, incapability of providing fuel retention space distribution characteristics and the like, the invention provides the following technical scheme: step 1, realizing external trigger time sequence synchronous control of an instrument; 2, acquiring a mass spectrum signal and determining a calibration coefficient; step 3, acquiring the distance between the endoscope and the position to be measured and feeding back the distance; 4, focusing the laser to a to-be-measured position; 5, synchronously collecting spectrum and mass spectrum signals; step 6, converting the mass spectrum signal into the total number of fuel particles in unit area on the wall surface; step 7, combining the spectrum and the mass spectrum to obtain depth distribution of fuel particle number at the position to be measured; 8, remote scanning of different areas of the inner wall of the Tokamak is achieved. Quantitative information of local fuel particles is provided, and overall distribution characteristics are obtained.