174 results about "Tokamak" patented technology

The invention discloses a method and a device for cleaning a tokamak first mirror by vacuum ultraviolet laser. A dust impurity deposit layer on the first mirror is cleaned by adopting a laser ablation method; through controlling the output energy of a laser device, the energy density of an excimer laser bundle on the surface of a first mirror surface is at a cleaning threshold of the impurity deposition layer and the damages threshold of first mirror metal, and the first mirror metal material can not be damaged when the impurity is cleaned simultaneously. A laser with 193nm wavelength is adopted for the method and the device for cleaning the tokamak first mirror by the vacuum ultraviolet laser, in the mutual effect process of the laser and the material, the ablation effect is less, the surface uniformity is good, the dust is cleaned, and the effect is ideal.

The invention discloses an optical mass spectrometer diagnosis technology for distinguishing helium and deuterium Tokamak residual gas by using penning discharge, and the difference between the characteristic spectrums He-I and D-alpha of the low temperature plasma emission spectrums produced by D2 and He discharge is utilized. The D2 and He to be measured are introduced into an ionization chamber, and a penning discharge system installed at one side of the ionization chamber ionizes the gas to be measured to produce low temperature plasma; then the He-I spectral line and D-alpha spectrum generated by the plasma are analyzed, so as to measure the absolute content of the two gases (He and D2). The invention realizes the diagnosis technology for real-time monitoring of the evolution of He and D2 in Tokamak, thus solving the problem that a commercial mass spectrometer cannot distinguish He and D2 due to close molecular mass numbers (He:4.003, D2:4.028). The diagnosis technology provides direct engineering and physical experiences for ITER and possible future in situ measurement of He / D2 on DEMO.

The invention discloses a superhigh-temporal resolution laser Thomson scattering diagnosis system. A laser produces multi-pulsed laser with a pulse repetition rate of 4KHz and wavelength of 1064nm. A laser beam goes through a laser beam optical path transmission assembly and is focused and introduced into an EAST Tokamak device, in the EAST Tokamak device, laser and plasma interact and produce Thomson scattering, the scattered signal is imaged on end surfaces of multiple optical fibers through a large-aperture large-field angle imaging system, the scattered signal is respectively transmitted into 30 spectrometers through optical fiber transmission and is subjected to photovoltaic conversion under the action of the 30 spectrometers and the electric signal is respectively fed into a GHz sampling frequency data acquisition and processing system. The large energy high frequency continuous output laser is used as a laser Thomson scattering active diagnosis light source so that necessary conditions are provided for Thomson superhigh-temporal resolution diagnosis and the superhigh-temporal resolution laser Thomson scattering diagnosis is realized.

The invention discloses a simulation method for the evolution of an electron cyclotron current profile in a tokamak. In the process of numerically simulating the evolution of the electron cyclotron wave drive current profile in the tokamak device, a drive current source is firstly described by using a Gaussian function; the evolution of the drive current under a specific magnetic field configuration is calculated according to a convection equation determined by a Fisch-Boozer theory; the magnetic field configuration under the correction of the drive current is calculated; the evolution of thedrive current continues to be calculated under the magnetic field configuration; and the simulation effect of long-term evolution is achieved through repeatedly calculating. The method disclosed by the invention has the beneficial effects that the evolution of the electron cyclotron wave-driven current profile under a real three-dimensional magnetic field configuration is realized; the drive current profile can be obtained at any time; in the meanwhile, the spatial distribution situation of the drive current can be more accurately described through the convection equation determined by the Fisch-Boozer theory; the calculation efficiency is high; the numerical stability is strong; and the method is a stable and efficient numerical simulation method.

The invention discloses a low-hybrid-wave antenna phase compensation method. Three lines of main waveguide of an antenna unit respectively comprises a section of linear transitional waveguide with the gradually-widened narrow edge. A main waveguide is divided into N / 2 sections of branch waveguide through E face branch waveguide of a first phase shift section, then, the main waveguide is divided into N sections of branch waveguide in a second phase shift section, and N is an even number. Phase shifters are arranged through two methods respectively, and the 180-degree phase difference brought by an output port in the middle line of a mode converter and the phase difference caused by the waveguide length difference which is caused by an antenna port arc notch can be eliminated. The low-hybrid-wave antenna phase compensation method can be applied to the design of a Tokamak low-hybrid-wave multi-junction waveguide antenna and an active and passive staggered multi-junction waveguide PAM antenna of a nuclear fusion experimental device and can be used for eliminating various phase differences of the antennas and improving the coupling performance and current drive efficiency of the antennas and plasmas.

The invention discloses a lower hybrid wave power control system in a Tokamak experiment, which comprises 24 high-power klystrons. The output end of each high-power klystron is provided with a directional coupler. Incident power coupled by the directional coupler is sent to a detector after direct current (DC)-blocking through a DC-block and filtering through a filter, detection voltage coming out of the output end of the detector is sent to an amplifier, amplified voltage output by the amplifier is collected through an analog input module, processed through a power control lower computer, and output by an analog output module, and the output end of the analog output module is connected to a personal identification number (PIN) electrically controlled attenuator of a microwave preceding stage. By means of collection of the power control lower computer, PID algorithm and output of the analog input module, the lower hybrid wave power control system in the Tokamak experiment controls the PIN electrically controlled attenuator of the preceding stage and controls lower hybrid wave preceding stage power so that output power of the klystrons is controlled. Meanwhile, by means of collection of output power and status signals of the klystrons, working status of the system is judged, PID control is adjusted according to requirements so as to avoid output overshoot of the klystrons, ensure stability of the lower hybrid wave system, and meet requirements of the Tokamak physical experiment. The power control lower computer of the system adopts a cRIO system, and enables the control system to have accuracy and reliability of a field programmable gate array (FPAG) and flexibility of an ordinary industrial personal computer.

The invention discloses a method for removing hydrogen detained on a first wall in a nuclear fusion device by direct current glow oxygen plasma, which is mainly used for creating optimized wall condition for effective operation of nuclear fusion device plasma physical experiment economy and reducing safely and environment problems due to radioactive tritium detained on the first wall of future device. The inventin is characterized in that according to carbon and hydrogen codeposition characteristic in the device, a plasma auxiliary chemical removing method is used, on a HT-7 superconducting tokmak and a EAST device having similar bit-type with ITER, and first wall temperature of 400-470K, a direct current glow plasma technology is employed, oxygen which is easy to react with carbon and hydrogen is employed to form radio frequency oxygen plasma for removing carbon hydrogen re-deposition layer on the first wall, thereby effectively removing hydrogen detained on the first wall of a device. The invention creates excellent wall condition for re-deposition layer removal on the first wall, reduction of hydrogen detained in the device, particle recirculation lowering and nuclear fusion device high efficiency operation, and reduces safely and environment problems of future nuclear fusion device due to radioactive tritium detained on the first wall.

The invention discloses a test particle simulation method for analyzing the confinement performance of high-energy particles in Tokamak, and belongs to the field of magnetic confinement controlled nuclear fusion. The method comprises the following steps: firstly, dividing a two-dimensional triangular mesh according to input parameters such as a balanced magnetic field and a plasma profile; secondly, loading high-energy particle information, which comprises information such as position and speed, into a phase space; then, according to the full-orbit equation and the guide center orbit equation,the evolution of the particles along with time is solved, and a Monte Carlo method is adopted in the collision module; and finally, outputting information of all particles, and outputting informationsuch as particle distribution, energy flow and current after the information is processed by the post-processing module. According to the method, physical processes such as transportation and loss ofhigh-energy particles can be simulated, the calculation efficiency is high, the numerical stability is high, and the method is a stable and efficient numerical simulation method.

A microwave-supplying magnetron and an ion burner are provided in an incinerator body, microwave from the magnetron and ion flame from the ion burner are resonated with each other to heat the inside of the incinerator body to high temperature, and wastes in the incinerator body are decomposed and melted by activated ions (+)(-). A tokamak is provided on the outer side of the incinerator body to reflect charged particles (radiation) and electromagnetic waves for focussing to the center of the inside of the incinerator body, thereby increasing an on concentration, a plasma concentration and a decomposing efficiency. Both or one of crystal and an acceptor-level additive are mixed into the furnace wall of the incinerator body.

A fuel injection mode of future Tokamak fusion reactor is featured as transporting high pressure fuel gas of fusion reactor to high lift region of vacuum chamber through piping, setting a Lavel nozzle at tail end of piping to let gas stream speed be supersonic speed at outlet of nozzle and then entering plasma high lift region. The high efficiency of fuel injection can be realized by the designed mode.

The invention discloses a method for removing hydrogen detained on a first wall in a magnetic constraint device by ion involution oxygen plasma. According to carbon and hydrogen codeposition characteristic in the magnetic constraint device, a plasma auxiliary chemical removing method is used, on a HT-7 superconducting tokmak and a EAST device having similar bit-type with ITER, under condition of keeping longitudinal magnetic fields of 1.5-2.0T and wall temperature of 400-470K, a ion involution plasma technology is employed, oxygen which is easy to react with carbon and hydrogen is employed to form ion involution oxygen plasma for removing carbon hydrogen re-deposition layer on the first wall, thereby effectively removing hydrogen detained on the first wall of a device. The technology is a unique and effective technology, creates excellent wall condition for re-deposition layer removal on the first wall, reduction of hydrogen detained in the device, particle recirculation lowering and nuclear fusion device high efficiency operation, and reduces safely and environment problems of future nuclear fusion device due to radioactive tritium detained on the first wall. Simultaneously, the treatment for the first wall can be carried out under condition that longitudinal field coil is not demagnetized, and is more suitable for future international thermonuclear fusion experiment reactor ITER and future fusion reactor.

The invention discloses a real-time plasma configuration reconstructing system and a real-time plasma configuration reconstructing method for Tokamak. The real-time plasma configuration reconstructing system comprises a plasma diagnosis system, the plasma diagnosis system is connected with a real-time configuration acquiring industrial personal computer, and the real-time configuration acquiring industrial personal computer is provided with a data acquisition card and a corresponding reflection memory card. Plasma diagnosis signals of the plasma diagnosis system enters the data acquisition card of the real-time configuration acquiring industrial personal computer, a time sequence unifying system provides a unified time sequence signal which also enters the real-time configuration acquiring industrial personal computer, the real-time configuration acquiring industrial personal computer transmits numerical value information of the plasma diagnosis signals acquired by the data acquisition card to a reflection memory interchanger through the reflection memory card, the reflection memory interchanger synchronizes numerical values of the plasma diagnosis signals to a reflection memory card of a high-grade real-time configuration reconstructing work station, the high-grade real-time configuration reconstructing work station transmits calculated plasma configuration information to the reflection memory interchanger through the reflection memory card connected onto the high-grade real-time configuration reconstructing work station, and a configuration controls calculates according to the plasma configuration information.

The invention belongs to the technical field of annular magnetic confinement nuclear fusion experimental research devices, and particularly relates to a tokamak device vacuum chamber window assembly welding positioning device. The tokamak device vacuum chamber window assembly welding positioning device comprises lateral plates, rib plates, connecting plates, a bottom plate and a positioning shaft, wherein the two lateral plates are welded together through the rectangular rib plates; the positions of the seven connecting plates correspond to window flanges on fan-shaped sections, the number of windows in the fan-shaped sections is smaller than seven, and the connecting plates corresponding to the positions without windows are empty; the connecting plates are provided with bolt holes, and during assembly welding, the connecting plates are connected with the corresponding 5-7 window flanges on the fan-shaped sections through bolts to position a window assembly. The bottom plate is arranged on the outer lateral plate and used for fixing the window assembly welding positioning device. The two connecting plates corresponding two opposite windows are each provided with a shaft hole, and a positioning shaft penetrates through the shaft holes to be matched with inner holes of the opposite window flanges to position the fan-shaped sections. Through the design, window welding deformation is effectively controlled, and the window assembly welding efficiency is greatly improved.

The invention discloses a method and device for enhancing magnetic disturbance based on gas injection to inhibit an escape current. The method comprises the steps: mounting a gas injection device on aTokamak vacuum chamber; loading gas to be injected into the gas injection device in advance; connecting a trigger control end of a switching valve of the gas injection device to a Tokamak central control system; when a plasma stably operates, controlling the Tokamak central control system not to trigger the switching valve and the gas injection device not to be started; and when the plasma is tobe disrupted, controlling the Tokamak central control system to trigger the switching valve to start the gas injection device, and making the gas injection device inject a gas into the plasma. The gasinjection device comprises a gas storage part, the switching valve and an injection nozzle. According to the method, the gas is injected into the plasma when the plasma is to be disrupted, and the gas entering the plasma disturbs a local current of the plasma to result in magnetic field disturbance, so that the formation of the escape current is inhibited, and the damage of the escape current tothe device is avoided.

The invention belongs to the technical field of nuclear fusion, and particularly relates to a turbulent momentum transport probe array. A rectangular groove at the lower part of a pyrolytic graphite protecting sleeve A is connected with a connecting plate to form a rectangular cavity; a step-shaped cavity fixed connecting piece is arranged at the center of the lower side of the connecting plate; a two-way four-step symmetrical structure is arranged at the upper part of the pyrolytic graphite protecting sleeve A; the middle of the symmetrical structure is high and the two sides of the symmetrical structure are reduced gradually; twenty-one pyrolytic graphite probes are divided into seven rows equally and distributed on steps in parallel; the lower end of each pyrolytic graphite probe is connected with a lead copper tube; the lower ends of the lead copper tubes are collected together by cable leads; the collected cable leads penetrate through a center hole of the connecting plate, and enter the interior of the fixed connecting piece; the lower ends of the cable leads are connected with movable plugs; and insulation sleeves are arranged in the rectangular cavity formed by the pyrolytic graphite protecting sleeve A and the connecting plate, and arranged among the lead copper tubes. With the adoption of the probe array, a macroparameter of plasma of an HL-2A tokamak device can be measured in a discharging process.

The invention discloses a high-frequency magnetic probe diagnosis system for measuring radio frequency wave characteristics. The measuring system comprises a vacuum coaxial electrode, a high-frequencymagnetic probe, a power divider, a detector, a phase discriminator and a signal acquisition system, the vacuum coaxial electrode is used for transmitting radio frequency signals from the Tokamak to the outside of the Tokamak; the high-frequency magnetic probe is used for coupling the radio-frequency signals; the power divider is used for distributing the voltage probe and current probe signals into two paths with equal amplitude and equal phase; the detector is used for detecting the peak-to-peak value of the voltage probe and current probe radio frequency signals, and the result is output bya direct-current signal; the phase discriminator detects the phase difference between the voltage probe and current probe signals, and the result is output by a direct-current signal. The signal acquisition system stores the collected voltage signals to a computer.

The present invention discloses high power continuous wave DCN laser. It contains discharge tube, anode, cathode, vacuum-pumping interface and inspiration interface on discharge tube, oil cooling jacket, waveguide type resonant cavity adopting plane cavity shape, plane reflection mirror, metallic nickel silk screen having tungsten filament, adopting x quartz plate sealing for anode end discharge tube laser output window, working medium being deuterate methane, nitrogen gas and deuterium gas or helium gas mixed gas. Said invention can continue steady operate for 168 hr. under lambda=194.7 micrometer output power about 150 mw condition, providing reliable plasma electron density measurement.

The invention discloses a phase-measuring device and identification method to entrance Tokamak low clutter antenna phase. That is to connect a dual directional coupler respectively at the root of every klystron output waveguide transmission line and near the antenna. Each dual directional coupler microwave signal is separated into two parts through filter and two straight power splitters respectively access Module A, B input. Phase two modules used to work, A module will phase output phase voltage curves of the same phase of the two phase corresponding phase combined output phase voltage module B identify a phase value. The invention measures and phases out amount of regulation quickly and accurately when Tokamak system sends the 2.45 GHz microwave signal to antenna phase entrance, provides accurate feedback phase of the phase control, provide a reliable technical support for studying microwave plasma coupling efficiency.

A high-power continuous wave klystron for Tokamak device is formed by three portions: the first portion is an electronic focusing system which comprises a negative electrode, a control electrode and a positive electrode and the like, where electron beams emitted from the negative electrode are converged by the control electrode and the positive electrode, then, form a round electron beam after being subjected to high voltage acceleration, and reach a beam-wave interaction system of the klystron; the second portion is a high-frequency beam-wave interaction system of the klystron, the high-frequency beam-wave interaction system being formed by a cylindrical reentrant resonant cavity comprising an input cavity, buncher cavities and an output cavity, wherein the high-frequency beam-wave interaction system realizes improvement of power capacity, efficiency and gain effectively, and high-frequency microwave produced in the cavity is output through an output cavity waveguide; and the third portion of a collector, wherein the collector helps to collect the energy of electron beam in beam-wave interaction. The high-power continuous wave klystron is simple and easy to carry out; the high-power continuous wave klystron has the advantages of easy processing, small process processing difficulty, high efficiency and high gain and the like; and the high-power continuous wave klystron can work more stably and reliably.

The invention relates to the field of nuclear fusion and discloses a method for measuring an ablation resistance characteristic of materials. According to the method, the laser-induced breakdown spectroscopy (LIBS) and the quartz crystal microbalance (QCM) are combined, wherein the LIBS is used for diagnosing types of species generated by ablation and the percentage of all the species; the QCM is used for calculating total ablation amount, and then, measuring results of the LIBS and the QCM are combined to obtain the ablation rate of all components. According to the method for measuring the ablation resistance characteristic of the materials, by combining the LIBS and the QCM, defects that LIBS quantitative analysis is comparatively difficult and substance distinguishing cannot be performed by the QCM are remedied; the types of species generated by the ablation and the percentage of all the species are diagnosed by using the LIBS, and the total ablation amount is calculated by the QCM; and then, the measuring results of the LIBS and the QCM are combined to obtain the ablation rate of all the components. A feasible detection means is realized for evaluating whether an already used material is suitable for Tokamak stable operations or not and seeking a first wall material which is suitable for the Tokamak stable operations.

The invention discloses a method and a device for cracking relief for electromagnetic energy guide-out of Tokamak plasma cracking. The method is characterized in that a group of energy transfer coils coupled with the plasma is utilized to couple one part of electromagnetic energy of the plasma at the cracking moment, and the electromagnetic energy is guided out of a vacuum chamber and is absorbed by an energy absorbing part, so the sum of the energy dissipated in the vacuum camber at the cracking moment can be reduced, and the damage caused by dissipation of huge energy of the plasma in the vacuum chamber at the cracking moment is relieved; the sum of the energy dissipated in the vacuum camber in the cracking process cannot be reduced by the existing cracking relief methods, such as MGI (massive gas injection), but the defect can be effectively overcome by the method, and the method can also be cooperated with the existing methods, such as MGI, so as to relieve the damage to device parts by the huge energy of the plasma cracking. The method for reliving the Tokamak plasma cracking has the advantages that the complicated technical means are not needed, the operation is simple and reliable, and the implementation cost is low.

This invention discloses a method for measuring corona current in tokmak device and its device, characterized in: equipping interconnect screw coil with high frequency response in tokmak device vacuum case inwall, via inducted current of screw coil, attaining changes of vacuum inner magnetic field, based on these, via high temperature plasmas, round vertical field magnet, pole field magnet and vacuum case body in the tokmak device, testing the screw coils to attain corona current, and the non-uniformation of current distribution along ring direction.

The invention discloses a method applicable to a fusion reactor for triggering a high-frequency edge localized mode by low-hybrid-wave modulation, which is characterized in that the current density orpressure intensity distribution of a stylobate region of the fusion reactor is directly or indirectly changed by using high-frequency, high-power and low-hybrid-wave modulation injection under the parameter background that a low-frequency large-amplitude edge localized mode breaks out in the operation of a high confinement mode of a Tokamak fusion reactor, a high-frequency edge localized mode istriggered, the amplitude of the edge localized mode and the harm brought about by the edge localized mode for a wall material of the fusion reactor are reduced. The method disclosed by the invention has flexibility in technical adjustment and adaptability for plasma parameter changes, and has little disturbance to plasma discharge during steady-state operations of the fusion reactor.

The invention discloses a foreign particle concentration measuring system for a J-TEXT Tokamak device. The foreign particle concentration measuring system comprises an X-ray optical path module, an X-ray detection module, a pulse shaping amplification module and an MCA data processing module. When the Tokamak device is cracked, a foreign gas is injected in order to suppress high-energy high-speed escape electrons, foreign particles may generate radiation and the impurity concentration is calibrated by measuring the impurity radiation intensity. Because a radiation energy pulse signal is weak, a high-sensitivity detector is used to detect the pulse signal and the high-gain pulse shaping amplification module is used to shape and amplify the pulse signal in order to output a voltage pulse signal with amplified amplitude. The MCA data processing module, according to the collected voltage pulse signal, performs energy spectrum analysis. Because different impurities have different radiation pulse peaks, the concentration of the various impurities can be determined according to the energy spectrum.

The invention discloses a high-temperature superconduction cable for strong magnetic field and fusion reactor tokamak superconduction magnet. The high-temperature superconduction cable comprises a support structure, wherein a plurality of grooves are formed in the support structures in an axial direction, semi-circular low-temperature cooling grooves are formed in the bottom and a side wall of each groove, a high-temperature superconduction strip stack is fixedly welded with the interior of each groove by a heat conduction and electric conduction welding material, a filler is arranged on the high-temperature superconduction strip stack and is used for fixing the strip stack, a central cooling hole is formed in the center of the support structure, and a protection shell sleeves an outer side of the support structure. Two cooling modes of conduction cooling and direct cooling are employed. The cable employs a spirally-twisting high-temperature superconduction stack to eliminate the anisotropy of the high-temperature superconduction strip, the mechanical property of the cable is improved, and the stability of the cable under a large-current and complicated strong magnetic field fusionenvironment is improved.

The invention discloses a hard X-ray flux detection system of a J-TEXT Tokamak device. The hard X-ray flux detection system comprises a hard X-ray detection module, a detection circuit and an AD data acquisition module; the J-TEXT Tokamak device generates a large number of high-speed escape electrons when disruption is operated; the high-speed escape electrons bombard the material of the first wall of the device so as to directly damage the material; great importance should be attached to the density monitoring of the escape electrons; when the escape electrons damage the first wall, thick target bremsstrahlung occurs between the escape electrons and the material, and hard X-rays are generated, and the energy of the hard X-rays can be up to 0.5 to 10MeV; the more escape electrons strike the wall of the device, the higher the energy of the generated hard X-rays is, and the higher pulse voltage outputted by the detector is; hard X-ray optical signals are converted to pulse voltage signals through the hard X-ray detection module; the peak signals of the pulse signals are detected by the detection circuit; the peak signals are transmitted to a computer terminal through the data acquisition card; the computer terminal stores the peak signals; radiation intensity can be determined; and since the radiation intensity is proportional to the amplitude of the pulse signals, the distribution condition of the escape electrons can be obtained.

The invention belongs to the technical field of Tokamak device, particularly relates to a scanning probe system on a high field side of a Tokamak device, which solves the measurement problem of plasma parameters at the edge on the high field side of the Tokamak device. A shaft is assembled on a fixed bracket through a bearing and a bearing sleeve; a coil, a coil framework and a probe fixed bracket connecting pipe are assembled on the shaft respectively; the coil and the plane surface of the coil framework are mutually perpendicular to the probe fixed bracket connecting pipe in the axial direction; a probe fixed bracket is arranged on the probe fixed bracket connecting pipe; a probe socket is arranged on the probe fixed bracket; probes are arranged on the probe socket and divided into two groups with one group arranged in a circular direction and another group arranged in an electrode direction; and lead wires connected with the probes are led out through the probe fixed bracket connecting pipe. In the scanning probe system provided by the invention, the probes are made of pyrolysis graphite material, and digital raster displacement is adopted for automatic measuring and controlling the movement displacement; and the temperature, the density, and the distribution of electric potential and the field of the electronics at the edge, which are vital parameters in the study of edge physics, can be simultaneously measured within one discharge of the controlled nuclear fusion study device.

The invention provides a device for cleaning a first mirror for a tokamak device by a direct-current cascade arc plasma torch. The device for cleaning the first mirror for the tokamak device by the direct-current cascade arc plasma torch comprises a cascade arc plasma torch generation system, a cooling water supplying system, a supporting system and a reflectivity detection system, wherein the cascade arc plasma torch generation system comprises a vacuum chamber, a cascade source, a direct-current power supply, a vacuum unit, a gas supply unit and a cooling unit; the cascade source is arranged at one end of the vacuum chamber; the direct-current power supply is electrically connected with the cascade source; the gas supply unit is connected with the cascade source through a pipeline; the vacuum unit is connected with the vacuum chamber; the cooling unit is arranged in the cascade source. The device for cleaning the first mirror for the tokamak device by the direct-current cascade arc plasma torch is simple, reasonable and compact in structure; the device is based on a direct-current cascade arc plasma torch technique; on the premise that an impurity is not introduced, the large-area and uniform cleaning of an impurity sedimentary formation on the surface of a sample of the first mirror can be realized; the impurity sedimentary formation is removed quickly.

The invention belongs to the field of nuclear fusion plasma diagnosis, and particularly relates to a Tokamak density fluctuation and electromagnetic field fluctuation correlation measuring device. TheTokamak density fluctuation and electromagnetic field fluctuation correlation measuring device comprises a microwave emitting source and a microwave intrinsic source, and further comprises a microwave emitting link connected with the microwave emitting source, a microwave intrinsic link connected with the microwave intrinsic source, a microwave receiving antenna and an orthogonal mode converter,wherein an angle rotating device is installed on the microwave receiving antenna. By means of the Tokamak density fluctuation and electromagnetic field fluctuation correlation measuring device, strictsimultaneous measurement of plasma density fluctuation and magnetic fluctuation can be achieved, and therefore correlation calculation between local measurement physical quantities is achieved, and the system has the advantages of being high in integration level, easy, convenient and flexible to use, low in cost and easy to maintain.

The invention discloses a Tokamak plasma disruption energy processing device and processing method. The Tokamak plasma disruption energy processing device comprises a voltage dependent resistor, a pulse capacitor unit, a resistance unit, an invert unit and a control unit; the voltage dependent resistor, the pulse capacitor unit and the resistance unit are parallelly connected in sequence between the positive pole and the negative pole of a direct current bus, and the positive pole and the negative pole of the pulse capacitor unit and the positive pole and the negative pole of the direct current bus are correspondingly connected; the positive pole of an input end of a direct current side of the invert unit is connected with the positive pole of the direct current bus, and the negative pole of the input end of the direct current side of the invert unit is connected with the negative pole of the direct current bus. The Tokamak plasma disruption energy processing device and processing method are mainly applied to the controllable nuclear fusion Tokamak plasma disruption energy processing, a multiprocessing function of processing energy storage, energy absorption, energy consumption and energy feedback recycle is achieved, the plasma disruption energy transmitted by a plasma coupled coil can be effectively absorbed and processed, and the Tokamak plasma disruption protection and the energy recycling are achieved.