First Engineering Commissioning of EAST Tokamak

Experimental Advanced Superconducting Tokamak （EAST） is the first fully superconducting tokamak. The first commissioning started on Feb. 1st of 2006 and finished on March 30TM of 2006 at the Institute of Plasma Physics, Chinese Academy of Sciences. It consists of leakage testing at both room temperature and low temperature, pumping down, cooling down all coils, current leads, bus bar and the thermal shielding, exciting all the coils, measuring magnetic configuration and warming up the magnets. The electromagnetic, thermal hydraulic and mechanical performance of EAST Toroidal Field （TF） and Poloidal Field （PF） magnets have also been tested. All sub-systems, including pumping system, cryogenic system, PF＆ TF power supply systems, magnet instrumentation system, quench detection and protection system, water cooling system, data acquisition system, main control system, plasma control system （PCS）, interlock and safety system have been successfully tested.     

Diagnostics of Internal Inductance in HT-7

1. IntroductionMagnetic measurements have been used exten-sively since the early days of tokamaks. Many prop-erties of a tokamak plasma can be determined byusing simple loops or coils of wire. Basic measurements aJre for the plasma current, loop voltage,plasma position and shape, stored plasma energy,and current distribution. InfOrmation about instabilities [l] is also obtained. For efficient operationsof these tokarnaks, it is essentiaJ to have a fast, y6tan accurate method for the deterndn…     

Recent Progress of HT-7U Superconducting Tokamak

HT-7U is a superconducting tokamak. which is being constructed in Institute of Plasma Physics, Chinese Academy of Sciences. The mission of the HT-7U project is to develop a scientific and engineering basis of the steady state operation of advanced tokamak.The engineering design of the device has been optimized. The R&D program is going on. Short samples of the conductor and a CS model coil were tested. All the TF and PF coils will be manufactured and tested in Institute of Plasma Physics. Therefore, a 600-meter long jacketing line for cable-in-conduit conductors along with two winding machines, a set of VPI equipment and a test facility for the TF and PF coils are ready in ASIPP now. In this paper, the recent progress of the HT-7U is described.     

Magnetic Sensorless Control of Plasma Position and Shape in a Tokamak

Magnetic sensorless sensing and control experiments with the plasma horizontal position have been carried out in the superconducting tokamak HT-7. The sensing is made to focus on the ripple frequency component of the power supply with thyristor and directly from them without time integration. There is no drift problem with the integrator of wagnetic sensors. Two kinds of control experiments have been carried out： to keep the position constant and swing the position in a triangular waveform, And magnetic sensorless sensing of plasma shape is discussed.     

Structural design of cryogenic component support for EAST （HT-7U）

EAST is a full superconducting tokamak with an elongated plasma cross-section. It consists of superconducting poloidal field (PF) magnet system, toroidal field (TF) magnet system, vacuum vessel with inner parts, thermal shields and cryostat vessel. The mission of the project is to widely investigate both physics and technologies of advanced tokamak operations, especially the mechanism of power and particle handling for steady-state operations. The cryogenic component is mainly composed of superconducting TF and superconducting PF coils that ensure the ability of sustaining magnetic field for plasma confinement, control and shaping in steady-state. This report describes the process of the structure design of cryogenic component support for EAST.     

New Current Control Method of DC Power Supply for Magnetic Perturbation Coils on J-TEXT

In order to advance the research on suppressing tearing modes and driving plasma rotation, a DC power supply （PS） system has been developed for dynamic resonant magnetic perturbation （DRMP） coils and applied in the J-TEXT experiment. To enrich experimental phe- nomena in the J-TEXT tokamak, applying the circulating current four-quadrant operation mode in the DRMP DC PS system is proposed. By using the circulating current four-quadrant oper- ation, DRMP coils can be smoothly controlled without the dead-time when the current polarity reverses. Essential circuit analysis, control optimization and simulation of desired scenarios have been performed for normal current. Relevant simulation and test results are also presented.     

Design of Controller for New EAST Fast Control Power Supply

The effectiveness of the magnetic confinement of plasma can be improved by elongat- ing the plasma cross-section in tokamak devices. But elongated plasma has vertical displacement instability, so a feedback control system is needed to restrain the plasma＇s vertical displacement. A fast control power supply is needed to excite the active feedback coils, which produces a magnetic field to control the plasma＇s displacement. With the development of EAST, the fast control power supply needs to keep on enhancing the fast response and output current. The structure of a new power supply is introduced in this paper. The method of multiple inverters paralleled with the current sharing reactor is presented to meet the need for large current and fast control. According to the design demands of the EAST fast control power supply, the adjuster of the current close loop is applied to the inverter, which can advance its ability to restrain the loop current in low frequency and DC output. The result of the experiment confirms the validity of the proposed scheme and control strategy.     

Design,Analysis and R＆D of the EAST In-Vessel Components

In-vessel components are important parts of the EAST superconducting tokamak. They include the plasma facing components, passive plates, cryo-pumps, in-vessel coils, etc. The structural design, analysis and related R＆D have been completed. The divertor is designed in an up-down symmetric configuration to accommodate both double null and single null plasma operation. Passive plates are used for plasma movement control. In-vessel coils are used for the active control of plasma vertical movements. Each cryo-pump can provide an approximately 45 m^3/s pumping rate at a pressure of 10^-1 Pa for particle exhaust. Analysis shows that, when a plasma current of 1 MA disrupts in 3 ms, the EM loads caused by the eddy current and the halo current in a vertical displacement event （VDE） will not generate an unacceptable stress on the divertor structure. The bolted divertor thermal structure with an active cooling system can sustain a load of 2 MW/m^2 up to a 60 s operation if the plasma facing surface temperature is limited to 1500 ℃. Thermal testing and structural optimization testing were conducted to demonstrate the analysis results.     

The Principle of Plasma Boundary Identification on HL-2A Tokamak

Plasma boundary identification is a basic task for studies on equilibrium and confinement in a divertor tokamak. With the progress on the experiments after engineering experiments, the boundary identification becomes an important issue for HL-2A. In order to satisfy the requirements of preciseness, simplified measurements and quickness, the filament current method instead of solving the equilibrium equations is used to identify plasma boundary on HL-2A. The involved principle, mathematics and the progresses, which have been made with this method, are given.     

Technical Design of Arc-Discharge and Deceleration Power Supply for MW Level NBI System on HL-2A Tokamak

Neutral beam injection (NBI) is one of the most effective ways to heat and drive plasma in a tokamak.The mega watt level neutral beam injector on the HL-2A tokamak con-sists of four high-power ion sources.Each source is supplied by discharge,beam extraction and auxiliary power supplies.Some circuit topologies and control sequences designed for the system are presented in this paper.Some important technologies such as the notching circuit,insulated gate bipolar transistor (IGBT) series-connected switch,high-frequency switching power supply and control system based on a digital signal processor (DSP) have been used.The system can be effectively used for high current ion beam extraction,protection,ion optics and so on.The power system has been safely used in HL-2A high-parameter NBI experiments for three years.The power of NBI can be kept at higher than 0.75 MW for 1 second and the ion beam power extracted from the ion source is higher than 2 MW.The ion temperature of the plasma center is close to 2.0 keV.These results show that the design of this power system is reasonable and reliable,and it can fully meet the system requirements for NBI of the HL-2A tokamak.     

First experimental results with the Current Limit Avoidance System at the JET tokamak

The Current Limit Avoidance System (CLA) has been recently deployed at the JET tokamak to avoid current saturations in the poloidal field (PF) coils when the eXtreme Shape Controller is used to control the plasma shape. In order to cope with the current saturation limits, the CLA exploits the redundancy of the PF coils system to automatically obtain almost the same plasma shape using a different combination of currents in the PF coils. In the presence of disturbances it tries to avoid the current saturations by relaxing the constraints on the plasma shape control. The CLA system has been successfully implemented on the JET tokamak and fully commissioned in 2011. This paper presents the first experimental results achieved in 2011–2012 during the restart and the ITER-like wall campaigns at JET.     

Operation with 1 MA plasma current in EAST

Discharge with a plasma current of 1 MA at a line-averaged density of 1.8×1019m-3 was realized in EAST, a fully superconducting tokamak. The key issues to achieve the discharge with 1 MA plasma current include both early shaping and LHCD assistance during start-up phase to extend the voltage margin of PF coils for easier plasma control, an optimization of the control methodology for PF coils to avoid over-current fault and a very good wall condition. A better wall condition was achieved mainly by extensive Lithium coating. Both stationary H-mode and diverted plasma discharge of 100 s were also obtained.     

Conceptual Design of Alborz Tokamak Poloidal Coils System

The Alborz tokamak is a D-shape cross section tokamak that is under construction in Amirkabir University of Technology. One of the most important parts of tokamak design is the design of the poloidal field system. This part includes the numbers, individual position, currents and number of coil turns of the magnetic field coils. Circular cross section tokamaks have Vertical Field system but since the elongation and triangularity of plasma cross section shaping are important in improving the plasma performance and stability, the poloidal field coils are designed to have a shaped plasma configuration. In this paper the design of vertical field system and the magnetohydrodynamic equilibrium of axisymmetric plasma, as given by the Grad-Shafranov equation will be discussed. The poloidal field coils system consists of 12 circular coils located symmetrically about the equator plane, six inner PF coils and six outer PF coils. Six outer poloidal field coils (PF) are located outside of the toroidal field coils (TF), and six inner poloidal field coils are wound on the inner legs and are located outside of a vacuum vessel.     

The magnetic field configuration measurement on EAST tokamak

The magnetic field configurations of poloidal field (PF) and toloidal field (TF) are the base of tokamak plasma operation. They are determined by the parameters such as positions and structures of PF and TF coils. Parameters of TF and PF coils of a new fully superconducting tokamak with non-circular cross-section EAST will change when the coils are cooled down from the ambient temperature to 4 K. Because of the cryogenic and refrigerator system, these parameters cannot be measured directly. Using magnetic probes signals, we measured and reconstructed magnetic field configuration of TF and PF coils. Parameters such as the positions of PF coils, the profile of the toloidal field in radial direction, the ripple and error field of toloidal field are obtained from the measurements.     

Design Features and Commissioning of the Versatile Experiment Spherical Torus (VEST) at Seoul National University

A new spherical torus called VEST (Versatile Experiment Spherical Torus) is designed,constructed and successfully commissioned at Seoul National University.A unique design feature of the VEST is two partial solenoid coils installed at both vertical ends of a center stack,which can provide sufficient magnetic fluxes to initiate tokamak plasmas while keeping a low aspect ratio configuration in the central region.According to initial double null merging start-up scenario using the partial solenoid coils,appropriate power supplies for driving a toroidal field coil,outer poloidal field coils,and the partial solenoid coils are fabricated and successfully commissioned.For reliable start-up,a preionization system with two cost-effective homemade magnetron power supplies is also prepared.In addition,magnetic and spectroscopic diagnostics with appropriate data acquisition and control systems are well prepared for initial operation of the device.The VEST is ready for tokamak plasma operation by completing and commissioning most of the designed components.     

Feedback Control of Plasma Current and Horizontal Position in HT-7

1. IntroductionA superconducting tokamak HT-7 has been estab-lished at ASIPP, Hefei, China. The machine.was de-signed to mainly investigate the reactor-relevant ls-sues, such as edvanced operation modes and plasmawall interastions in the near-steady-state condition.Its poloidal fie1d coils include ohmic heatlng coi1s'bias field coils' vertical field coils and horizontalfie1d coi1s (See Fig.1), being connected to indlvldualpower supplies which are all the thyristor--controlledrectifier unlt…     

Multivariable Online Adaptive PID Controller for Plasma Current,Shape, and Position in Tokamaks

Nuclear fusion is one of the newest and most promising clean and safe energies hence, it imposes a new research area of control. In this paper, the design of a multivariable adaptive proportional-integral-derivative (PID) controller for the control of the plasma current, shape and position to ensure the safe operation of the fusion reactor is successfully developed. The recursive least square algorithm is used in an alternative way as an adaptation mechanism for tuning PID controller gains. Since stability is a vital issue in the evaluation of control systems, therefore stability analysis of the proposed controller is developed using the Lyapunov stability theory. The main objective of plasma current, shape and position controller in fusion reactors is to improve the stability and the performance of tokamak magnetic systems without contravening the limits imposed by the actuating coils voltages physical limitations. The proposed APID (adaptive PID) controller tunes online its parameters to cope with the presence of the disturbance or any parameters changes occur during the operation. The results of the proposed APID on a simulation code of a tokamak show a noteworthy improvement with respect to those obtained with other control techniques in the cases of changing the initial controller gains, adding disturbance signal and variation in the reactor model parameters.     

Plasma profile and shape optimization for the advanced tokamak power plant, ARIES-AT

An advanced tokamak plasma configuration is developed based on equilibrium, ideal MHD stability, bootstrap current analysis, vertical stability and control, and poloidal field coil analysis. The plasma boundaries used in the analysis are forced to coincide with the 99% flux surface from the free-boundary equilibrium. Using an accurate bootstrap current model and external current drive profiles from ray tracing calculations in combination with optimized pressure profiles, β_N values above 7.0 have been obtained. The minimum current drive requirement is found to lie at a lower β_N of 6.0. The external kink mode is stabilized by a tungsten shell located at 0.33 times the minor radius and a feedback system. Plasma shape optimization has led to an elongation of 2.2 and triangularity of 0.9 at the separatrix. Vertical stability could be achieved by a combination of tungsten shells located at 0.33 times the minor radius and feedback control coils located behind the shield. The poloidal field coils were optimized in location and current, providing a maximum coil current of 8.6 MA. These developments have led to a simultaneous reduction in the power plant major radius and toroidal field from those found in a previous study [S.C. Jardin, C.E. Kessel, C.G. Bathke, D.A. Ehst, T.K. Mau, F. Najmabadi, T.W. Petrie, the ARIES Team, Physics basis for a reversed shear tokamak power plant, Fusion Eng. Design 38 (1997) 27].     

Present Status of the KSTAR Superconducting Magnet System Development

The mission of Korea Superconducting Tokamak Advanced Research (KSTAR) project is to develop an advanced steady-state superconducting tokamak for establishing a scientific and technological basis for an attractive fusion reactor. Because one of the KSTAR mission is to achieve a steady-state operation, the use of superconducting coils is an obvious choice for the magnet system. The KSTAR superconducting magnet system consists of 16 Toroidal Field (TF) coils and 14 Poloidal Field (PF) coils. Internally-cooled Cable-In-Conduit Conductors (CICC) are put into use in both the TF and PF coil systems. The TF coil system provides a field of 3.5 T at the plasma center and the PF coil system is able to provide a flux swing of 17 V-sec. The major achievement in KSTAR magnet-system development includes the development of CICC,the development of a full-size TF model coil, the development of a coil system for background magnetic-field generation , the construction of a large-scale superconducting magnet and CICC test facility. TF and PF coils are in the stage of fabrication to pave the way for the scheduled completion of KSTAR by the end of 2006.