The Present Status of HT-7U Superconducting Tokamak Project

The first Chinese superconducting tokamak HT-7 was reconstructed on the basis of a Russian tokamak T-7.The main purpose of reconstruction is to improve the accessibility of the device and to provide a possibility of long pulse operation with high performance. The reconstruction has been done successfully. The HT-7U project has been approved and funded as a National Project, the engineering design and R & D are under way.

Preliminary Engineering Design of Toroidal Field Magnet System for Superconducting Tokamak HT-7U

HT7U is a large fusion experimental device. It will be built in the Institute of Plasma Physics of Chinese Academy of Sciences. The mission of HT-7U is to develop the scientific basis for a continuously operating tokama-k fusion reactor. This paper describes only a toroidal field (TF) superconducting magnet system of HT7U. In this paper, design criteria of conductor and stability analysis, coil winding and support structure design of magnet system, mechanical calculation and stress analysis, heat load evaluation are given.

Soft x-ray Diode-array Diagnostic System for HT-7 Superconducting Tokamak

A diagnostic system of soft x-ray diode-array was set up for HT-7 superconducting tokamak. The system consists of two slot-aperture cameras and is capable of measuring the soft x-ray emission from the plasma on HT-7 device with a high resolution in space and a high response in time. Both cameras, located separately in a horizontal port and a vertical port each with thirty-seven detectors of An-Si surface-barrier diode (SBD) can view the same toroidal cross-section of the plasma from different poloidal chords. In this paper, the structure, principle and performance of the diagnostic system are described and some experimental results observed are presented.

Study on the Current-Sharing Control System of the TF Power Supply for a Superconducting Tokamak

The superconducting toroidal field （TF） plays an important role in a superconducting tokamak, whose power supply was developed based on the feedback control principle. In this paper, superconducting tokamaks in different countries are described, and the TF power supply of the International Thermonuclear Experimental Reactor （ITER） is taken as an example to study the current-sharing characteristics in the current-stabilized stage. Firstly, the mathematical model of the TF power supply is established, and then the 3-loop control method is put forward for achieving the stability and reliability of current-stabilization and current-sharing. Furthermore, further studies indicate that the current-sharing controller has no influence on the current-stabilized control, and current-stabilizing and current-sharing can be realized at the same time. All the work done provides valuable references for the current-sharing design of the TF power supply for a superconducting tokamak, and all these studies lay a solid foundation for developing superconducting tokamaks.

Experimental Research on Lower Hybrid Current Drive in Superconducting Tokamak HT-7

?Fundamental experiments on lower hybrid current drive (LHCD) have been undertaken on HT-7 superconducting tokamak. The experiments on LHCD efficiency reveal its depen- deuce on plasma density and the toroidal magnetic field. Furthermore, the experiments on HT-7 successfully demonstrate the ability for LHCD to sustain long pulse tokamak discharges, such as discharges with full non-inductive current drive for several seconds. The experimental study to improve plasma confinements by LHCD suggests that the improvement should be due to the change o f current profile. It has also been demonstrated by the experiments that the lower hybrid wave may lead to an enhanced ionization of particles in the region where the wave is deposited.

Development of Special Winding Machine for HT-7U Superconducting Tokamak

A special winding machine with high accuracy has just been developed and applied to the construction of HT-7U Tokamak. It is one of the critical facilities for R & D of HT-7U construction. The machine mainly consists of five parts, including a CICC pay-off spool, a fourroller correcting assembly, a four-roller forming/bending assembly, a continuous winding structure and a CNC control system with three-axis AC servo motors. The facility is used for Cable in Conduit Conductor (CICC) magnet fabrication of HT-7U. The main requirements of the winding machine are: continuous winding to reduce joints inside the coils; pre-forming CICC conductor to avoid winding with tension; suitable for all TF & PF coils of various coil shapes and within the dimension limit; improving the configuration tolerance and the special flatness of the CICC conductor. This paper emphasizes on the design and fabrication of the special winding machine for HT-7U. Some analyses and techniques in winding process for trial D-shape coil are also presented.

Runaway electron dynamics in Experimental Advanced Superconducting Tokamak helium plasmas

The generation of runaway electrons(REs)is observed during the low-density helium ohmic plasma discharge in the Experimental Advanced Superconducting Tokamak(EAST).The growth rate of hard x-ray(HXR)is inversely proportional to the line-average density.Besides,the RE generation in helium plasma is higher than that in deuterium plasma at the same density,which is obtained by comparing the growth rate of HXR with the same discharge conditions.The potential reason is the higher electron temperature of helium plasma in the same current and electron density plateau.Furthermore,two Alfvén eigenmodes driven by REs have been observed.The frequency evolution of the mode is not fully satisfied with the Alfvén scaling and when extension of the Alfvén frequency is towards 0,the high frequency branch is~50 kHz.The different spatial position of the two modes and the evolution of the helium concentration could be used to understand deviation between theoretical and experimental observation.

Prediction of multifaceted asymmetric radiation from the edge movement in density-limit disruptive plasmas on Experimental Advanced Superconducting Tokamak using random forest

Multifaceted asymmetric radiation from the edge(MARFE) movement which can cause density limit disruption is often encountered during high density operation on many tokamaks. Therefore, identifying and predicting MARFE movement is meaningful to mitigate or avoid density limit disruption for the steady-state high-density plasma operation. A machine learning method named random forest(RF) has been used to predict the MARFE movement based on the density ramp-up experiment in the 2022’s first campaign of Experimental Advanced Superconducting Tokamak(EAST). The RF model shows that besides Greenwald fraction which is the ratio of plasma density and Greenwald density limit, dβp/dt,H98and d Wmhd/dt are relatively important parameters for MARFE-movement prediction. Applying the RF model on test discharges, the test results show that the successful alarm rate for MARFE movement causing density limit disruption reaches ~ 85% with a minimum alarm time of ~ 40 ms and mean alarm time of ~ 700 ms. At the same time, the false alarm rate for non-disruptive and non-density-limit disruptive discharges can be kept below 5%. These results provide a reference to the prediction of MARFE movement in high density plasmas, which can help the avoidance or mitigation of density limit disruption in future fusion reactors.

Advanced Control Scenario of High-Performance Steady-State Operation for JT-60 Superconducting Tokamak

Plasma control on high-β_N steady-state operation for JT-60 superconductingmodification is discussed. Accessibility to high-β_N exceeding the free-boundary limit isinvestigated with the stabilising wall of reduced-activated ferritic steel and the active feedbackcontrol of the ill-vessel non-axisymmetric field coils. Taking the merit of superconducting magnet,advanced plasma control for steady-state high performance operation could be expected.

Comparison of different noble gas injections by massive gas injection on plasma disruption mitigation on Experimental Advanced Superconducting Tokamak

Massive gas injection(MGI)is a traditional plasma disruption mitigation method.This method directly injected massive gas into the pre-disruption plasma and had been developed on the Experimental Advanced Superconducting Tokamak(EAST).Different noble gas injection experiments,including He,Ne,and Ar,were performed to compare the mitigation effect of plasma disruption by evaluating the key parameters such as flight time,pre-thermal quench(pre-TQ),and current quench(CQ).The flight time was shorter for low atomic number(Z)gas,and the decrease in flight time by increasing the amount of gas was insignificant.However,both pre-TQ and CQ durations decreased considerably with the increase in gas injection amount.The effect of atomic mass on pre-TQ and CQ durations showed the opposite trend.The observed trend could help in controlling CQ duration in a reasonable area.Moreover,the analysis of radiation distribution with different impurity injections indicated that low Z impurity could reduce the asymmetry of radiation,which is valuable in mitigating plasma disruption.These results provided essential data support for plasma disruption mitigation on EAST and future fusion devices.

Removal of Particles by ICRF Cleaning in HT-7 Superconducting Tokamak

The ICRF (Ion Cyclotron Range Frequency) cleaning technique has been used as a routine wall cleaning method in the HT-7 superconducting tokamak. In a wide range of toroidal field, the removal rate of residual gas by ICRF cleaning was about twenty times higher than that of glow discharge cleaning (GDC). At different gas pressure and RF power levels, the ICRF cleaning is studied carefully. A good impurity cleaning effect and a very high hydrogen removal rate were obtained. The removal rate of hydrogen by 5 kW ICRF cleaning achieved was 1.6 × 10-5 Tirr.1/s.And the relationships among pressure P, outgassing rate Q, atomic layers L absorbed on surface and the cleaning mode were discussed briefly.

A Study on Main Breaker for Quench Protection of HT-7U Toroidal Superconducting Tokamak

This paper proposes a quench protection project of HT-7U toroidal superconducting tokamak through a forced commutation analysis of DC circuit breaker (DCCB) paralleling fuse. Based on the requirement of quench protection, main parameters are selected. Experimental results demonstrate the validity of this proposed project.

Multi-channel FIR Interferometer for Electron Density Measurement on HT-7 Superconducting Tokamak

A five-channel far-infrared (FIR) hydrogen cyanide (HCN) laser interferometer was developed to measure plasma electron density profile on the HT-7 superconducting tokamak. The principle and structure of the five-channel FIR laser interferometer is described. The laser source used in the interferometer was a continuous wave glow discharge HCN laser with a 3.4 m cavity length and a 100 mW power output at 337μm wavelength. The temporal resolution was 0.1 ms and the detection sensitivity was 1/12 fringe. Preliminary experimental results measured by the interferometer on HT-7 tokamak are reported.

Fast Thermography for HT-7U Superconducting Tokamak

A 30 channel fast IR pyrometry array has been proposed for HT-7U superconduct- ing tokamak, which has a 0.5 μs time response, a 10 mm diameter spatial resolution and a 5℃ temperature resolution. The temperature range is from 250℃ to 1200℃. The two dimensional temperature profiles of the first wall during both major and minor disruptions could be measured with an accuracy of about 1% of measuring temperature, which is adequate for tokamak experi- ments. This offers a very useful tool for disruption study, especially for the research of divertor physics and edge heat flux on HT-7U superconducting tokamak.

Design of the Pumping Stations and the Gas Puffing Systems for the Superconducting Tokamak HT-7U

The design of the pumping stations and the gas injection systems for the inner Chamber and the cryostat on HT-7U superconducting tokamak is described in this paper.

Eddy Currents and Electro-magnetic Forces on the Lower Hybrid Ware Launching Antenna on the Superconducting Tokamak HT-7

This paper analyzes the eddy currents and the electro-magnetic forces on the lower hybrid wave (LHW) launching antenna on the superconducting Tohamak HT-7 by using a finite element circult method. A new iterative algorithm is developed to analyze the coupled magnetic fields Which are very difficult to be calculated. The method and results obtained are helpful to study the eddy currents and electro-magnetic forces on metal plates which are placed in a rather complicated electro-magnetic environment.

The Design of Circuit for Checking Short in HT-7U Superconducting Tokamak Device

The tokamak HT-7U project has been funded as a Chinese national project since 1998. The main object of the project is to build a nuclear fusion experimental device with divertor configuration, which is designed by the Institute of Plasma Physics, the Chinese Academy of Sciences (ASIPP). It is a full superconducting device, consisting of superconducting toroidal field (TF) coils and superconducting poloidal field (PF) coil. During the operation of the device, the operational parameter of device should be checked by technical diagnosis. This paper describes the design of circuit for checldng short between every two parts of the HT7U device. The main contents of design include circuit of data acquisition and data processing of computer.

Superconducting Busline Design for HT-7U Tokamak

Superconducting (SC) tokamak HT-7U has seven pairs of buslines connecting toroidal/poloidal coils and the current leads. These SC buslines (SCBLs) share a common cryostat and are made of the cable in conduit conductors (CICCs) arranged as a decoupling configuration. In order to reduce the heat loads conducted from the seven current leads with a capacity of 15 kA during the magnets cooldown, the buslines with a much lower thermal conduction were employed in comparison with the current leads, and a special cooling loop was designed.

First results of CO_(2) dispersion interferometer on EAST tokamak

A dispersion interferometer(DI)has been installed and operates on the Experimental Advanced Superconducting Tokamak(EAST).This DI system utilizes a continuous-wave 9.3μm CO_(2)laser source to measure line-averaged electron densities accurately.In contrast to conventional interferometers,the DI does not require substantial vibration isolations or compensating systems to reduce the impact of vibrations in the optical path.It also employs a ratio of modulation amplitudes,ensuring it remains immune to the variations in detected intensities.Without a variation compensation system,the DI system on EAST reaches a density resolution of less than1.8×10^(-2)πrad and a temporal resolution of 20μs.The measurements made by the POlarimeterINTerferometer(POINT)system and the far-infrared hydrogen cyanide(HCN)interferometer are remarkably consistent with the DI’s results.The possibility of fringe jumps and the impact of refraction in high-density discharge can be significantly decreased using a shorter wavelength laser source.A rapid density change of 3×10^(19)m^(-3)during 0.15 s has been measured accurately in shot No.114755 of EAST.Additionally,the DI system demonstrates dependability and stability under 305 s long-pulse discharges in shot No.122054.

The Experiment of Modulated Toroidal Current on HT-7 and HT-6M Tokamak

The Experiments of Modulated Toroidal Current were done on the HT-6M tokamak and HT-7 superconducting tokamak. The toroidal current was modulated by programming the Ohmic heating field. Modulation of the plasma current has been used successfully to suppress MHD activity in discharges near the density limit where large MHD m = 2 tearing modes were suppressed by sufficiently large plasma current oscillations. The improved Ohmic confinement phase was observed during modulating toroidal current (MTC) on the Hefei Tokamak-6M (HT-6M) and Hefei superconducting Tokamak-7 (HT-7). A toroidal frequency-modulated current, induced by a modulated loop voltage, was added on the plasma equilibrium current. The ratio of A.C. amplitude of plasma current to the main plasma current Ip/Ip is about 12% ~ 30%. The different formats of the frequency-modulated toroidal current were compared.