25 results about "Plasma current" patented technology

The invention provides a method and a system for measuring plasma current in a quasi-ring symmetrical star simulator, and relates to the technical field of star simulators, and the method comprises the steps: determining a sensor design index according to a bootstrap current simulation calculation result, and converting the design index into an actual arrangement scheme according to an installation environment constraint; all the magnet coils are independently excited in a plasma-free state, wall eddy current interference signals generated by the magnet coils are calibrated, and thus wall eddy current calibration voltage signals are output; a sensor in an actual arrangement scheme is used for synchronously collecting a total original induced voltage signal including plasma current and wall eddy current; an integral voltage signal representing the total current is output; and compensating the integral voltage signal by using the wall eddy current calibration voltage signal, and inverting the final measurement value of the plasma current in the quasi-ring symmetric star simulator. The device and the method have the beneficial effects that the interference of a stray field on a measurement result is weakened, and a high-precision diagnostic tool is provided for the plasma physical research of the star simulator.

A plasma power supply (10) and a control method therefor, and a fusion reaction system. The plasma power supply (10) comprises: a breakdown power supply module (101) and an extraction power supply module (102). The breakdown power supply module (101) is used for: outputting a first voltage to a plasma generator (20) which generates plasma under the application of the first voltage; and when the output voltage is decreased from the first voltage to a second voltage, maintaining the output of the second voltage to the plasma generator (20). The extraction power supply module (102) is used for forming an electric field between the plasma generator (20) and a housing (30) of a nuclear fusion reaction chamber, and when the plasma generator (20) generates the plasma, forming a plasma current between the housing (30) and the plasma generator (20) on the basis of the plasma and the electric field. The plasma power supply (10) allows for direct formation of the plasma current in the nuclear fusion reaction chamber, thereby simplifying the formation process of the plasma current.

The invention provides a plasma current profile distribution reconstruction method and system based on a high-precision magnetic field inclination angle. The method comprises the following steps: injecting a neutral beam to a specified point direction of a plasma core; a plurality of polarizers with different polarization directions are used for receiving polarized light beams transmitted from the plasma direction; measuring the intensity of polarized light penetrating through the polarizers, and calculating the inclination angle of a specified point magnetic field of the plasma core based on the intensity ratio of polarized light corresponding to each polarizer; acquiring magnetic measurement input data; and performing plasma core current profile distribution inversion based on the inclination angle of the magnetic field of the specified point of the plasma core and magnetic measurement input data. The system comprises a high-precision magnetic field inclination angle measurement system, a data acquisition system and a plasma configuration inversion system. By accurately measuring the inclination angle of the plasma core magnetic field, more accurate data support is provided for reconstruction of plasma current profile distribution, feedback control is optimized, and the stability and constraint efficiency of plasma are improved.

The invention discloses a spherical Tokamak device vacuum cavity wall eddy current measuring method and system, and the method comprises the steps: enabling two detection ends of a collector to be connected with the other end of a first sensing ring and the other end of a third sensing ring respectively, electrifying plasma, applying a current to a CS coil, and enabling the CS coil to be connected with the collector; and obtaining the measurement current measured by the collector, obtaining the plasma current in the vacuum cavity wall, and obtaining the size of the eddy current of the vacuum cavity wall of the spherical Tokamak device. According to the invention, the plurality of sensing rings are arranged, the magnetic fields among the plurality of sensing rings are mutually counteracted and the number of turns is adjusted, so that the error in the measurement process is compensated, the interference of CS coil current can be eliminated as much as possible when the eddy current of the vacuum cavity wall of the spherical Tokamak device is measured through the cooperative work of the plurality of sensing rings, and the measurement precision is greatly improved. The problem that a traditional measurement method is low in measurement precision in a narrow space and a strong interference environment is solved.

The invention discloses an underwater plasma arc cutting device for a large-diameter seamless steel pipe. The underwater plasma arc cutting device comprises a water tank, a displacement system, a multi-degree-of-freedom manipulator, a plasma arc cutting system and a visual sensing and control system. A pressure sensor, a temperature sensor, a hydrogen concentration sensor and a carbon monoxide concentration sensor are arranged in the water tank. A supporting rotating mechanism is arranged in the water tank and used for clamping and driving the large-diameter seamless steel pipe to rotate around the axis of the large-diameter seamless steel pipe. The displacement system is arranged on one side of the water tank, and the multi-degree-of-freedom mechanical arm is installed on the displacement system. A cutting torch of the plasma arc cutting system is installed at the output end of the multi-degree-of-freedom mechanical arm and used for extending into the water tank to cut the steel pipe. And the visual sensing and control system is used for monitoring the cutting process in real time and controlling the plasma arc cutting system, the multi-degree-of-freedom manipulator and the supporting and rotating mechanism to move. Parameters such as the cutting torch position, the cutting track, the plasma current, the gas flow and the cutting speed can be dynamically adjusted through the visual sensing and control system.

The application provides a kind of quasi-ring symmetry star simulator plasma current measurement method and system, it is related to star simulator technical field, including determining sensor design index according to bootstrap current simulation calculation result, and actual arrangement scheme is converted into design index according to installation environment constraint;In the state of no plasma, all magnet coils are excited separately, and the wall eddy current interference signal generated by the magnet coil is calibrated, so as to output the wall eddy current calibration voltage signal;The total original induction voltage signal including plasma current and wall eddy current is synchronously collected by using the sensor in the actual arrangement scheme;Output integral voltage signal representing total current;The integral voltage signal is compensated by using the wall eddy current calibration voltage signal, and the final measurement value of quasi-ring symmetry star simulator plasma current is inverted.The beneficial effects of the application are to weaken the interference of stray field on the measurement results, and to provide a high-precision diagnostic tool for star simulator plasma physics research.

The invention discloses an adjusting system and method for realizing steady-state billion-degree ion temperature, and relates to the technical field of magnetic confinement nuclear fusion, and the system comprises a central control unit, and a neutral beam injection unit, a central spiral pipe and an air supply unit of which the output ends are connected with plasma in a vacuum chamber. The neutral beam injection unit is used for heating high-energy neutral atoms emitted by the plasma; the central spiral tube is used for generating and controlling plasma current; the gas supply unit is used for conveying working gas into the vacuum chamber; and the central control unit can respectively adjust the neutral beam injection unit, the central spiral pipe and the air supply unit so as to obtain the required steady-state ion temperature. By adjusting the beam energy and power of the neutral beam injection unit, the plasma electron density and the current climbing rate, the plasma has very high heating power density, ion energy transportation is reduced, heating efficiency is improved, and therefore the steady-state ion temperature of hundred million degrees is obtained.

An existing terahertz detector is slow in response speed and narrow in response spectrum. The terahertz micro-cavity area array detector based on the bias plasma comprises a window and a photosensitive element, the photosensitive element comprises a first micro-cavity plate, a gas cavity layer and a second micro-cavity plate which are connected in sequence, the first micro-cavity plate is provided with a gas filling pipe, the gas filling pipe penetrates through the first micro-cavity plate, and the gas filling pipe is provided with a gas outlet. The gas filling pipe is communicated with the gas cavity layer, a discharge electrode is arranged on the second microcavity plate, the gas cavity layer is connected with a driving circuit through the discharge electrode, and the gas filling pipe is connected with a gas filling device. The terahertz detector has a spectral imaging function, adopts the principle that terahertz enters a gas cavity layer and the state of plasma changes, realizes detection of terahertz waves, can obtain information such as polarization and intensity of the terahertz waves, can solve the problems that an existing terahertz detector is low in response speed and narrow in response spectrum, is used for hidden target imaging, and has a wide application prospect.

The invention relates to the technical field of plasma current driving, in particular to a TE01 mode transmission elbow structure and an LHCD system. The elbow structure comprises a transition section and an elbow section, the whole transition section is of a circular barrel structure, and a plurality of steps are constructed in the transition section to enable the transition section to form a multi-diameter hole section structure, namely, an inner hole of the transition section has a plurality of hole diameters; the two elbow sections are connected end to end, the free ends of the two elbow sections are connected with a transition section, the bending angle of each elbow section is 45 degrees so that the two elbow sections can form a right-angle turning structure, a plurality of annular grooves are formed in the inner wall of each elbow section, and the annular grooves are distributed in the extending direction of the elbow sections at intervals. According to the invention, the high transmission rate of the TE01 mode can be ensured in a low-frequency environment.

The invention discloses a Tokamak device plasma discharge starting method and system, and relates to the technical field of magnetic confinement fusion. The method comprises the following steps: generating initial plasma through the plasma pre-ionization system; the neutral beam injection system injects neutral particles into the initial plasma, and the injection energy is dynamically adjusted according to the plasma density monitored in real time; the electron cyclotron wave heating system injects microwave energy to drive electrons to generate plasma current; the diagnosis system monitors plasma parameters and neutral beam penetration depth in real time, and feeds back the parameters to the injection system and the heating system to adjust the parameters; and after the plasma current reaches the standard and is stable, parameters are optimized to maintain steady-state operation. The system comprises a pre-ionization module, an injection module, a heating module, a diagnosis module and a control module. According to the invention, the volt-second consumption and construction cost are reduced, the starting efficiency and stability are improved, the device is prevented from being damaged, and various magnetic confinement fusion devices are adapted.

The application discloses a kind of spherical tokamak's equilibrium field power supply, equilibrium power supply method, comprising: constant current source I3 with equilibrium field coil electric connection;Two groups of capacitor banks are respectively connected with equilibrium field coil, for supplying equilibrium field coil a steady-state low-level or high-level current balance high plasma current;Two groups of charge-discharge circuits are respectively connected with two groups of capacitor banks.Based on the equilibrium field power supply of spherical tokamak, the equilibrium power supply method of two groups of capacitors is used to supply power to the equilibrium field coil, so that the low plasma current is balanced by the equilibrium field current provided by constant current source I3, and the high plasma current is balanced by the equilibrium field current provided by two groups of capacitor banks.In the presence of low plasma current, ohmic field is put in, and plasma current is improved to a certain extent.In this way, the problems of limited volt-second product of spherical tokamak ohmic transformer, insufficient coil installation precision and stray field influence caused by vacuum wall eddy current can be compensated.

The application provides a method for preparing single-walled carbon nanotubes by a plasma-assisted chemical vapor deposition method, and relates to the technical field of nanometer material preparation. The method adds controllable plasma current and voltage to the condition of chemical vapor deposition. The plasma promotes the polymerization and growth of carbon atoms by providing a high-energy field and a catalyst activation effect. In the chemical vapor deposition method, iron sulfate and an iron-containing inorganic compound are used in combination. The addition of sulfur can promote the crystallization and arrangement of carbon atoms, and the combination of the two can reduce the surface defects of single-walled carbon nanotubes and reduce the use cost. The method for preparing single-walled carbon nanotubes by the plasma-assisted chemical vapor deposition method can prepare single-walled carbon nanotubes with a Raman G / D ratio of more than 120 while reducing the cost, and large-scale production is realized.

The invention discloses an online quality monitoring system for high-strength steel laser-plasma composite cutting, which belongs to the field of composite cutting online quality monitoring and comprises a data acquisition module, a splash characteristic threshold judgment module, a real-time path comparison module, a process curve comparison module and a bimodal factor diagnosis module. The invention discloses an on-line quality monitoring system for high-strength steel laser-plasma composite cutting, which is characterized in that a data acquisition module acquires laser power, plasma current and other process parameters in real time, and synchronously captures an original image of a cutting area through a three-channel camera; the spatter characteristic threshold determination module establishes a parameter-spatter mapping library and outputs a parameterized threshold; the real-time path comparison module is used for stripping the splashing point and comparing a preset path; the process curve comparison module judges a process abnormal point through the reference arc line; the bimodal factor diagnosis module outputs a quality problem judgment result, real-time monitoring and accurate quality diagnosis in the cutting process are achieved, and it is guaranteed that the cutting quality is stable.

The present application relates to the technical field of controllable nuclear fusion, and particularly relates to a device for controlling energy confinement time by using a limiter bias system. The device comprises a bias electrode head, a soft connection wire connected with the bias electrode head at one end and connected with a bias power supply system of an NCST spherical tokamak device at the other end, and a limiter main body installed on one side of the NCST spherical tokamak device and provided with the bias electrode head for restraining the escaped electrons generated by discharge of the NCST spherical tokamak device. The present application can restrain the escaped electrons generated by discharge of the NCST spherical tokamak device by installing the limiter bias system, improve the energy confinement time parameter of the NCST spherical tokamak device, significantly restrain the plasma instability, and thus increase the plasma current duration and improve the plasma current size.

The application belongs to the technical field of nuclear fusion, and discloses a tokamak plasma vertical displacement event disruption prediction method based on a random forest. The technical scheme comprises the following method steps: S1, obtaining tokamak plasma equilibrium parameters, shape control data and plasma current control data, and cleaning the data; S2, setting different segmentation points according to whether each parameter of each discharge gun is stable or not, and marking them as 0 and 1 respectively; S3, verifying whether each gun has separability by using a t-SNE algorithm. The application inputs different plasma parameters in the tokamak into a machine learning model, sets a threshold value for judgment, and outputs an early warning signal. The method of artificial intelligence is introduced into the EAST tokamak plasma vertical displacement disruption prediction, and the problem of low accuracy of the disruption prediction is solved.

A process chamber and a semiconductor process equipment are disclosed. The process chamber comprises a chamber body having an opening at a top portion thereof, an upper electrode assembly comprising an upper electrode cover plate covering the opening and an electrode assembly disposed at a bottom surface of the upper electrode cover plate, wherein an annular gap is defined between the electrode assembly and a sidewall of the chamber body, and an impedance ring assembly sleeved outside the electrode assembly, wherein the impedance ring assembly comprises dielectric rings and conductive rings which are alternately stacked, and the conductive rings are spaced apart from the electrode assembly and the sidewall of the chamber body. Each of the conductive rings is connected to at least one electrical connector. The electrical connector electrically connects the corresponding conductive ring to the chamber body in a first state, and disconnects the electrical connection between the corresponding conductive ring and the chamber body in a second state. The thickness of the impedance ring assembly can be adjusted without disassembling the upper electrode assembly, so as to adjust the distribution of the plasma current. The operation is convenient, and the production efficiency can be improved.

This invention relates to the interdisciplinary field of quantum computing and nuclear fusion plasma profile reconstruction, and particularly to a tokamak plasma current density reconstruction system based on a coherent Ising machine. The technical solution includes an electromagnetic diagnostic data acquisition unit and a Bayesian posterior modeling processor. The electromagnetic diagnostic data acquisition unit is used to acquire external electromagnetic measurement data of the tokamak device in real time, forming a measurement vector. The Bayesian posterior modeling processor is connected to the electromagnetic diagnostic data acquisition unit and is used to construct a posterior accuracy matrix and a linear term coefficient vector based on a conditional autoregressive prior model and the electromagnetic measurement likelihood probability. This invention organically integrates Bayesian inference modeling, spatial block decomposition, block coordinate descent iterative control, and coherent Ising machine quantum solving, achieving high-precision, rapid reconstruction of the tokamak plasma current density profile with uncertainty quantification within the constraints of existing quantum hardware bit capacity.

A system for a thermonuclear fusion reactor having a feeder configured to introduce a fuel projectile into a contained plasma is described. The system includes one or more lasers for adjusting a propulsion force of the fuel projectile by laser ablation and a controller configured to operate in conjunction with the feeder and the one or more lasers to controllably propel the fuel projectile toward a target location in the contained plasma. This contrasts with existing systems (e.g., fuel projectiles simply follow the magnetic field gradient of the plasma), and may enhance plasma control to achieve improved stability and generation of thermonuclear fusion energy, modulate the fusion rate and the amount of energy generated by the reactor, and improve the performance of the reactor. Plasma current is adjusted to alter the magnetic field and generate electrical power by magnetic induction, and / or have other advantages alone or in combination with existing systems.

Disclosed herein are systems and methods for delivering nitric oxide gas. According to an embodiment, a nitric oxide gas delivery system can include: a first nitric oxide (NO) gas generator, wherein the first NO gas generator includes a plasma chamber enclosing two electrodes separated by a gap; a plurality of port lights surrounding a plurality of corresponding gas outlet ports, wherein each of the port lights are configured to display a color; and an electronic control circuit, wherein the electronic control circuit is configured to enable the plurality of port lights to display the color based on a status of a plasma current across the two electrodes.

The invention relates to the technical field of controllable nuclear fusion, in particular to a device for controlling energy constraint time by using a limiter bias system. Comprising a bias electrode tip; one end of the flexible connection wire is connected with the bias electrode tip, and the other end of the flexible connection wire is connected with a bias power supply system of the NCST spherical Tokamak device; the limiter main body is mounted on one side of the NCST spherical Tokamak device, and a bias electrode head is mounted on the limiter main body and used for inhibiting escape electrons generated by discharging of the NCST spherical Tokamak device. By installing the limiter bias system, escape electrons generated by discharging of the NCST spherical Tokamak device are restrained, the energy constraint time parameter of the NCST spherical Tokamak device is improved, plasma instability can be remarkably restrained, then the plasma current duration time is prolonged, and the plasma current size is improved.

The invention relates to the technical field of plasma energy engineering, and discloses a multi-pole magnetic trap magnetic flux distribution prediction method, device and equipment, and the method comprises the steps: constructing a neural network architecture comprising a magnetic flux neural network, a plasma current neural network and a convolution differential operator, and outputting total magnetic flux distribution through the magnetic flux neural network, the plasma current neural network is utilized to obtain the plasma current density from the total magnetic flux distribution and obtain a first constraint condition, meanwhile, a convolution differential operator is combined to obtain a second constraint condition from the total magnetic flux distribution, and a prediction result is determined by comparing convergence errors of the two constraint conditions. Physical constraints are introduced through a convolution differential operator, the problem of large deviation between magnetic flux distribution prediction and reality caused by lack of physical constraints of a common neural network is solved, and meanwhile, through feedback coupling between the magnetic flux neural network and the plasma current neural network, the convergence and stability of the model are improved, and the prediction precision is improved. And the method is closer to the actual multi-pole magnetic trap magnetic flux distribution prediction effect.

The application provides a MOS transistor, comprising: a substrate; a P-type well region and an N-type well region which are located in the substrate and are adjacently arranged; a floating gate located on the P-type well region, a bulk terminal located in the P-type well region and a first protection diode, the floating gate is communicated with the N pole of the first protection diode, and the P pole of the first protection diode is communicated with the bulk terminal; a second protection diode located in the N-type well region, the second protection diode is connected with the first protection diode in parallel, the N pole of the first protection diode is communicated with the P pole of the second protection diode, and the P pole of the first protection diode is communicated with the N pole of the second protection diode. If the current generated by the plasma is positive, the current passes through the second protection diode, so that the gate is protected from being damaged by the plasma current. If the current is negative, the current passes through the first protection diode, so that the gate is also protected from being damaged by the plasma current.

The invention belongs to the field of magnetic confinement nuclear fusion, and discloses a cladding integrated conductive loop system for a fusion reactor and a control method thereof.The system is integrated in a cladding area of a Tokamak device, cladding blocks are electrically connected to form a closed conductive loop, and under the condition that an independent excitation power source is not needed, the fusion reactor can be electrically connected with the closed conductive loop. The system enables a conductive loop to generate induced current in the rapid change process of plasma current, and comprises a vacuum chamber, plasma, cladding blocks and conductive connection blocks arranged between the adjacent cladding blocks. According to the integrated passive coil system and the control method, a non-axisymmetric disturbed magnetic field can be effectively excited during plasma rupture through cladding integrated design and reliable external switch control, efficient suppression of escape electrons is realized, and the reliability of the system is improved. And an effective technical scheme is provided for passive inhibition of escaped electrons of the reactor-grade Tokamak device.

The application discloses an ohmic field coil power supply of a spherical tokamak, and belongs to the technical field of magnetic confinement nuclear fusion. The application comprises a first power supply unit, a second power supply unit, an auxiliary power supply unit and a control unit; the auxiliary power supply unit is composed of super capacitors and is connected in parallel to a discharge branch of the first power supply unit. The control unit is configured to perform a three-stage current control strategy: first, control the first and / or second power supply unit to make the ohmic field coil current rapidly climb to a preset peak value; then, start the auxiliary power supply unit to make the current stably maintain in a flat-top stage; after the flat-top stage ends, control the current to rapidly decrease. The equilibrium field coil is started in the current flat-top stage, and by taking advantage of the characteristic that the current variation is extremely small during the flat-top stage, the mutual coupling influence between the ohmic field coil and the equilibrium field coil is completely eliminated. The application has strong compatibility and controllable cost, effectively eliminates the limitation of mutual coupling on the plasma current, and significantly improves the operation performance of the spherical tokamak device.

The application discloses a fusion device high-temperature superconducting magnet quench detection method based on deep learning, which comprises the following steps: obtaining a magnet quench voltage noise signal, a magnet current signal, a plasma current signal and an eddy current signal from historical operation data of the fusion device, taking the signals as a data set to train a deep learning model; continuously and real-timely collecting the magnet current signal, the plasma current signal and the eddy current signal during the operation of the device, using the offline trained deep learning model to real-timely calculate the magnet quench voltage noise signal; deducting the real-timely calculated magnet quench voltage noise signal from the real-timely collected magnet quench voltage detection signal to obtain a real magnet quench voltage signal, and thus realizing accurate detection of the quench of the high-temperature superconducting magnet of the fusion device. With the strong learning ability of the deep learning, the method can timely and accurately detect the occurrence of the quench of the high-temperature superconducting magnet, and guarantee the safe and stable operation of the high-temperature superconducting magnet in the fusion device.