Position Of Plasma Column Research Articles

The Heavy-Ion Beam Diagnostic of the ISTTOK Tokamak-Highlights and Recent Developments.

The unique arrangement of the heavy-ion beam diagnostic in ISTTOK enables one to measure the evolution of temperature, density and pressure-like profiles in normal and AC discharges. The fast chopping beam technique provided the possibility to reduce the noise on the measurements of the plasma pressure-like profile and for the precise control of the plasma column position in real time. The consequent improvements in S/N levels allowed the observation of the effects of runaway beam magnetic energy conversion into plasma local heating. In addition, it made it possible to follow the evolution of the quiescent plasma maintained during AC transitions when the plasma current is null. The use of a new operation mode in the cylindrical energy analyzer provided an improved resolution up to five times in determining the fluctuations of the plasma potential as compared to the normal operation mode. Such analyzer is extremely compact (250 mm × 250 mm × 120 mm) and provides a unique geometry in order to cover the whole plasma diameter. The detector configuration choice gives the possibility for the simultaneous measurements of plasma poloidal magnetic field, plasma pressure-like and plasma potential profiles together with their fluctuations.

Plasma column position measurements using magnetic diagnostics in ADITYA-U tokamak

Due to several forces acting on the plasma column in a tokamak, the plasma column tends to move horizontally and/or vertically leading to many adverse events including termination of plasma. Precise measurement of plasma column position throughout the discharge with good temporal resolution in the real time is hence necessary in order to restrict the column movement using the stabilizing equilibrium magnetic fields. The plasma column position is measured by several magnetic diagnostics such as magnetic pick-up probe, Mirnov coil, Sine-Cosine coil along with flux loops in ADITYA Upgrade tokamak (ADITYA-U). In spite of the simplicity of the measurement principle of these probes as well as of their construction, obtaining the position of the plasma column is not very straightforward. The diversity of all these magnetic diagnostics in terms of their structures, installation location, mounting scenario etc leads to improper position estimations as these probes are being affected differently by the unwanted magnetic pick-ups. These unwanted pickups, especially those arising from vessel eddies, are successfully removed by introducing a scaling method for all the above-mentioned position probes in ADITYA-U based on the in situ calibration of these probes. The correction factors are deduced by estimating the known positions of a central current carrying conductor correctly using the probe measurements during the in situ calibration. These correction factors are used for estimating the plasma column position in the tokamak experiment along with proper nullification of magnetic fields arising from other magnets during tokamak operation. The plasma column position estimation from different magnetic probes are compared with each other and also with those estimated from diagnostics other than the magnetic ones.

Real-time feedback control system for ADITYA-U horizontal plasma position stabilisation

The ADITYA-U tokamak (R0 = 0.75 m, a = 0.25 m) is designed to shape plasma column in both single and double null diverter configurations. It is quite well known that sustaining a shaped plasma in tokamak requires very good plasma column position control, both horizontal and vertical. An FPGA-based proportional–integral–derivative (PID) control system has been designed and operated to achieve horizontal plasma position control in ADITYA-U tokamak. The complete system has been rigorously tested with sample signals before implementing to the ADITYA-U plasma discharges. The control system is integrated and time-synchronized with the plasma discharge operation of ADITYA –U. Furthermore, the system has been trained to take appropriate actions during the disruption or plasma failure in the tokamak operation. Detailed experimental results have been obtained by the operation of the digital PID controller. The complete design, installation, operation, tuning of the system along with all the relevant testing and operating experience of the digital PID controller for real-time horizontal plasma position control is presented in the paper.

Extension of Shafranov’s Equilibrium Theory to the Description of Current Quenches Affected by Resistive Wall Dissipation in Tokamaks

An equation for the plasma column position in a tokamak is derived with account of the resistive wall reaction on the changes in the plasma. A similar problem was considered by V.D. Shafranov [J. Nucl. Energy C 5, 521 (1963)] but for a fixed plasma current. Here, this current is treated as time-dependent, which allows to cover the current quenches and expand the study to the analysis of disruptions. It is shown that the net toroidal current induced in the wall during the current quench (which is not accounted for in the mentioned and related existing analytical models) can strongly affect the plasma equilibrium. A general algorithm applicable to arbitrary tokamaks is presented. For a large-aspect-ratio “circular” tokamak a full cycle of the step-by-step reduction of a problem is carried out down to the equation for the plasma shift evolution. The technique and results are ready for use with standard equilibrium solvers and models.

Determination of the plasma position for its real-time control in the COMPASS tokamak

An efficient horizontal and vertical stabilization of the plasma column position are essential for a reliable tokamak operation. Plasma position is generally determined by plasma current, plasma pressure and external vertical and horizontal magnetic fields. Such fields are generated by poloidal field coils and proper algorithm for the current control have to by applied, namely, in case of fast feedback loops. This paper presents a real-time plasma position reconstruction algorithms developed for the COMPASS tokamak. Further, its implementation in the MARTe (Multithreaded Application Real-Time executor) is described and the first results from test of the algorithm for real-time control of horizontal plasma positions are presented.

Analysis of the direction of plasma vertical movement during major disruptions in ITER

The plasma movement in the upward direction (away from the X-point) after the thermal quench (TQ) of major disruptions in ITER is favourable for the machine design, since the downward movement causes larger electromagnetic (EM) load due to the induced eddy and halo currents. Vertical directions of plasma movement after the TQ in ITER are investigated using the predictive mode of the DINA code. Three dominant parameters in determining the direction of plasma movement are identified: (i) the rate of plasma current quench (plasma temperature after the TQ), (ii) the width of plasma current mixing area just after the TQ (change of the internal plasma inductance li) and (iii) the initial vertical position of plasma column before the TQ. It is shown that the reference ITER plasma moves upwards after the TQ, if the electron temperature after the TQ is less than 10 eV and the drop of li does not exceed 0.2 for the present reference initial vertical position (55.5 cm above the centre of the machine). It is also shown that the operational domain leading to the upward movement is considerably large for disruptions with fast current quench, which could generate quite severe EM load due to the induced eddy current combined with the induced halo current if the movement is downwards.

Shift in plasma column position produced by azimuthally asymmetric electron heating in a tandem mirror

Large shifts in the equilibrium position of the plasma column in the Phaedrus tandem mirror [Nucl. Fusion 25, 1213 (1985)] are observed when a 25 kW, 1 msec, 15.5 GHz microwave heating pulse (ω∼2ωce) is injected into one end cell of the device. The shift in the plasma column axial position occurs in all regions of the device, and is comparable in size to the initial plasma radius. The change in plasma position is a result of nonuniform electron heating produced by azimuthally asymmetric illumination of the plasma column by the microwave beam.