Feedback Control Of Plasma Position Research Articles

A diagnostics slim cassette for reflectometry measurements in DEMO: Design and simulation studies

The reflectometry diagnostic proposed for DEMO aims to measure the radial edge density profile and to provide data for the feedback control of plasma position and shape. Microwave (MW) reflectometry measurements are foreseen at 16 poloidal locations (gaps) and require plasma-facing antennas and waveguides (WGs) to route the microwaves through the upper port (UP) to the diagnostic hall. The current integration proposal is based on the Diagnostics Slim Cassette (DSC) concept, a full, 25-cm thick poloidal sector dedicated to house the antennas and WGs. This paper translates the main findings from previous studies into an updated, coherent design of the DSC, while discussing integration aspects not considered before. Neutronics simulations, thermo-mechanical analyses and electromagnetic simulations are employed to evaluate the WG deformation under operating conditions and its effect on the reflectometry measurements. Although the estimated deformations are not expected to compromise the performance of the diagnostic, this work shows that the operating conditions must be considered when optimizing the shapes of the WGs to minimize losses.

Design and integration studies of a diagnostics slim cassette concept for DEMO

The diagnostics slim cassette (DSC) is a concept currently proposed for DEMO to host the microwave reflectometry diagnostic in a dedicated poloidal section, for the feedback control of plasma position and shape, also providing the radial edge density profile at several poloidal angles. This DSC, which is expected to house up to 100 antennas and waveguides, is to be integrated with the breeding blanket (BB) segments. Moreover, it is being designed with remote handling compatibility in mind to facilitate a ‘fast’ exchange by remote maintenance (RM) when needed, fulfiling one of the aims of the work package diagnostic and control. In this approach the pre-assembled (banana-shaped) DSC modules are inserted/removed from the upper port (UP) of the vacuum vessel (VV), at least when the BB segments are replaced. Here the main constraints related to the integration of the DSC with the BB, the UP and the equatorial port are identified and discussed in detail, with a strong focus on the required RM operations inside the VV (in-vessel), providing solutions that can be adapted to present and future blanket and UP designs.

Real-time plasma position reflectometry system development and integration on COMPASS tokamak

O-mode frequency-modulated continuous wave (FMCW) reflectometry provides an alternative to magnetic measurements in the determination of the plasma separatrix position for plasma position control. This type of measurement proves to be particularly attractive for the control of future fusion reactors where the harsh radiation environment may damage magnetic probes or induce non-compensable measurement drifts. Plasma position reflectometry (PPR), first demonstrated in ASDEX-Upgrade, is a control technique that is increasingly important to validate in diversified experimental devices and relevant plasma regimes. The COMPASS tokamak provides suitable conditions for such advanced demonstrations and regular PPR operation and development, thanks to its O-mode reflectometer and Multi-Threaded Application Real-Time executor (MARTe) based real-time control system. Herein we present the integration of a PPR system on COMPASS, both at hardware and software levels. Reflectometry swept measurements require signals to be acquired in bursts of data and streamed to the corresponding MARTe-PPR node through PCIe® fibre-optic links. The data transferred in real-time is used to reconstruct the radial density profiles from which the outer separatrix position is estimated. This estimate is then delivered to the central MARTe controller node via a dedicated Xilinx® Aurora® link at a rate matching COMPASS's 500 μs slow control cycle. The implemented system systematically met the required latency specifications, being able to deliver an estimation of the plasma radial position capable of successfully replacing the corresponding magnetic measurements in the plasma position feedback control loops.

Compensation of the pick-up fluxes on the plasma position control system when resonant magnetic perturbations are applied in J-TEXT.

In the J-TEXT tokamak the displacement coils, which serve as sensors for feedback control of plasma position, are significantly affected by the resonant magnetic perturbation (RMP) coils. To eliminate the deviation of plasma position caused by the RMP, the responses of the displacement coils to the RMP are studied with a view to compensate for the pick-up fluxes. Both the linear and eddy responses on the displacement coils are taken into account and adjusted accordingly. It was found that positional correction in the vertical direction was required when the RMP was operated in the poloidal mode m = 1 and horizontal positional correction was required when the RMP was operated in the m = 2 mode. When compensation was applied to the position control system, correction of plasma position was successfully achieved, as observed from soft X-ray imaging, as well as from the evolution of the current of the horizontal and vertical fields.

The Determination of Plasma Radial Shafranov Shift (ΔR) and Vertical Shift (ΔZ) experimentally using Magnetic probe and Flux loop Method for SST-1 Tokamak

The radial Shafranov shift (ΔR) and vertical shift (ΔZ) has been calculated for steady state superconducting tokamak (SST-1) experimentally using magnetic probes and flux loops. The SST-1 plasma at the present phases of operations is circular in shape and leans against the limiters. The radial and vertical shift formulated from Shafranov equation have been used for computation. We have selected best possible pairs of flux loops and magnetic probes location for the measurements. The comparison of results obtained from these two methods also has been performed for numerous numbers of shots for repeatability and reliability purpose. We have observed that the results of both the methods are in good agreement for almost all SST-1 experiment campaigns. Since the control of plasma position plays an important role in plasma confinement and optimizing tokamak operations, the computed ΔR and ΔZ could later on be used as a control parameter for feedback control of plasma position in SST-1.

Enhancement of EAST plasma control capabilities

In order to improve the plasma control performance and enhance the capability for advanced plasma control, new algorithms such as PEFIT/ISOFLUX plasma shape feedback control, quasi-snowflake plasma shape development and vertical control under new vertical control power supply, have been implemented and experimentally tested and verified in EAST 2014 campaign. P-EFIT is a rewritten version of EFIT aiming at fast real-time equilibrium reconstruction by using GPU for parallelized computation. Successful control using PEFIT/ISOFLUX was established in dedicated experiment. Snowfldivertor plasma shape has the advantage of spreading heat over the divertor target and a quasi-snowflake (QSF) configuration was achieved in discharges with Ip=0.25 MA and Bt=1.8T, κ∼1.9, by plasma position feedback control. The shape feedback control to achieve QSF shape has been preliminary implemented by using PEFIT and the initial experimental test has been done. For more robust vertical instability control, the inner coil (IC) and its power supply have been upgraded. A new control algorithm with the combination of Bang-bang and PID controllers has been developed. It is shown that new vertical control power supply together with the new control algorithms results in higher vertical controllability.

Optical boundary reconstruction of tokamak plasmas for feedback control of plasma position and shape

A new diagnostic is developed to reconstruct the plasma boundary using visible wavelength images. Exploiting the plasma's edge localized and toroidally symmetric emission profile, a new coordinate transform is presented to reconstruct the plasma boundary from a poloidal view image. The plasma boundary reconstruction is implemented in MATLAB and applied to camera images of Mega-Ampere Spherical Tokamak discharges. The optically reconstructed plasma boundaries are compared to magnetic reconstructions from the offline reconstruction code EFIT, showing very good qualitative and quantitative agreement. Average errors are within 2 cm and correlation is high. In the current software implementation, plasma boundary reconstruction from a single image takes 3 ms. The applicability and system requirements of the new optical boundary reconstruction, called OFIT, for use in both feedback control of plasma position and shape and in offline reconstruction tools are discussed.

Recent experiments in the EAST and HT-7 superconducting tokamaks

First divertor plasma configuration in Experimental Advanced Superconducting Tokamak (EAST) was obtained in the second campaign after the last IAEA meeting. To achieve long pulse diverted plasma discharges, new capabilities including the fully actively water cooled in-vessel components, current drive and heating systems, diagnostics and real-time plasma control algorithm were developed. Pre-programmed shape and feedback control of plasma position and current (RZIP) produced a variety of shaped plasma configurations, covering most of the configurations foreseen at the design stage of the machine. Control algorithm based on real-time equilibrium reconstruction and iso-flux control for the last closed magnetic flux surface (RTEFIT/ISOFLUX) has also been realized. A number of operational issues, such as plasma initiation and ramp up under constraints of superconducting coils were successfully investigated. First LHCD experiments demonstrated long pulse discharges longer than 20 s and nearly full non-inductive current drive. The physical engineering capability on the superconducting magnetic system was assessed by simulating discharges. Since the last IAEA meeting, experiments in HT-7 have been focusing on long pulse operation to support the EAST experiments on both physics and technical aspects. Long pulse discharges up to 400 s have now been achieved in HT-7. Investigation of sawtooth activities in ohmic and LHCD plasmas supports the turbulence model instead of the fast reconnection of the m = 1 magnetic island. Coexistence of electron mode and ion mode in high density ohmic plasmas has been observed by 2D ECE imaging (ECEI) in HT-7. The spectral characteristics of geodesic acoustic mode at the plasma boundary have been investigated by Langmuir probe arrays.

Comparative measurements of plasma position using coils, hall probes, and bolometers on CASTOR tokamak

Plasma position on CASTOR tokamak is measured by several systems based on different experimental methods. Vertical and horizontal plasma displacements are deduced from data of 4 Mirnov coils spaced poloidaly by 45°. Output of the Mirnov coils is routinely used as an input for automatic feed-back control of plasma position on CASTOR. Independently, we used an array of 16 Hall sensors, which measure the required magnetic field directly. Additional information on plasma position is obtained from two arrays of bolometers that measure horizontal and vertical profile of plasma radiation. Principle design of the all used diagnostics is given. Comparison of horizontal and vertical displacements deduced using various experimental methods is presented.

Plasma current and position feedback control in ADITYA Tokamak

Ohmic heating power supply is used in ADITYA Tokamak for plasma breakdown and current drive. Presently the ohmic coils store a magnetic flux of 0.40 V s and high d i/d t is created for the required loop voltage. After the plasma breakdown the plasma current control is made active with the difference of reference plasma current and measured plasma current to drive the PI regulator, which regulates the ohmic falling current. The vertical field current is also made proportional to plasma current by feedback control at appropriate time. The control scheme adopted and the results with the feedback control are described in this paper. We could improve the plasma current as well as plasma current flattop by employing the plasma current feed back control. Further to correct the radial position in fast time scale, four-quadrant transistor switched bridge inverter power supply by plasma position feedback control is proposed. On line ±Δ X (by position coils) position signal is given to PI control and polarity detector accordingly the current is amplified by the bridge inverter and fed to the four set of feedback coils which shifts the plasma position on radial directions.

Vacuum compatible, variable cross-section magnetic coil diagnostic used in digital feedback control of plasma position in TEXT-Upgrade

A magnetic pickup coil diagnostic set is used to measure the position of the plasma column in the Texas Experimental Tokamak Upgrade (TEXT-U) project. The output from this coil set is used in a digital feedback system to control the plasma position. To provide a fast time response for the feedback system, one complete coil set is located on the interior of the vacuum vessel. Another set with a slower time response is located on the outside of the vessel. To simplify and speed up data processing, the coils are constructed so that the X and Y coordinates of the plasma current centroid are each determined using the signals from only two separate coils. For each coordinate one coil is used to measure a tangential (relative to the coil surface) magnetic field component, while the second coil measures a normal field component. Due to physical constraints, the coils are not continuous around the vacuum vessel. The presence of gaps in the coils causes pickup of the external current flowing in the divertor coil windings during TEXT-U diverted discharges. This pickup has been successfully nulled out by adding a divertor current Rogowski coil to the X position coil circuit. The data indicate that these coils, along with the digital feedback system, are useful tools for flexible position control over a wide range of TEXT-U plasma parameters.

Stability limit of feedback control of vertical plasma position in the JFT-2M tokamak

Stability diagrams and time responses of the feedback control for vertical instability are obtained in the JFT-2M tokamak. The experimental results are in good agreement with a feedback theory based on the filament current plasma model taking account of the non-axisymmetric effect of the vacuum vessel. The vertical instability, with a growth rate of about 500 s−1, is successfully stabilized by feedback control of the plasma position with ordinary 12-phase thyristor power supplies; the elongation factor achieved amounts to up to κ = 1.7.

Passive Feedback Control of Plasma Position in a Tokamak

The effect of many passive coils on the positional instabilities in tokamaks is numerically investigated in the case of t ≪τ e (effective skin time of the coils), assuming the rigid shift of the plasma column. In order to make the positional stability region of the decay index broader, the passive coils must be located as close as to the plasma surface with the optimum position. The skin time of the coils is determined mainly by the number, cross section area and the conductivity of the coils. Application of the analysis to two small tokamaks shows the fairly good agreements with the experimental results.

Feedback Control of Plasma Position in JIPP T-II Tokamak

A step response of a plasma column to the desired plasma position is examined for several types of feedback control by solving the equation of motion of a plasma column in a resistive shell. Experimental results in JIPP T-II are also described.

Plasma equilibrium position dynamics of ISX-A tokamak

Equations of plasma equilibrium position perturbation dynamics of a tokamak have been developed. Neglecting secondary effects, an approximated equation of plasma high frequency oscillation is obtained and a low frequency dynamics function of plasma motion is derived. A symbolic form of the dynamics function allows a stability study of the tokamak's plasma equilibrium position. Knowledge of the dynamics function will permit an ordered analytical synthesis of plasma position feedback control systems which are necessary for a successful operation of a tokamak.