On-line plasma shape identification via magnetic measurements

This paper deals with the problem of designing a plasma current, position, and shape controller for the ITER-FEAT tokamak. Exploiting the features of the plant, we show that it is possible to design separately two control loops: a first loop which stabilizes the vertical position by means of a simple derivative action, and a second loop which drives the plasma current and the geometrical shape descriptors as close as possible to the reference values. The second loop is designed by assigning the dominant closed loop poles.

Real-time reflectometry – An ASDEX Upgrade DCS plugin App for plasma position and shape feedback control

Nuclear fusion is, in a sense, the opposite of nuclear fission. Fission, which is a mature technology, produces energy through the splitting of heavy atoms like uranium in controlled chain reactions. Unfortunately, the by-products of fission are highly radioactive and long lasting. On the other hand, fusion is the process by which the nuclei of two light atoms such as hydrogen are fused together to form a heavier (helium) nucleus, with energy produced as a by-product. Although controlled fusion is extremely technologically challenging, a fusion-power reactor would offer significant advantages over existing energy sources. This thesis is devoted to the control of tokamaks, magnetic confinement devices constructed in the shape of a torus (or doughnut). Tokamaks are the most promising of several proposed magnetic confinement devices. The need to improve the performance of modern tokamak operations has led to a further development of the plasma shape and position control systems. In particular, extremely elongated plasmas, with high vertical-instability growth rate, are envisaged to reach the required performance for ignition. This request for better performance from the experimentalists’ side has motivated the development of the new vertical-stabilization (VS) system at the JET tokamak, which has been proposed within the Plasma Control Upgrade project. This thesis presents the activity carried out to increase the capability of the VS system and to understand the operational limits in order to assess what can be done to improve the overall performance with the existing hardware and control system so as to minimize the impact on JET operation. The first objective of this work is the analysis of the new diagnostic system and the influence of the mechanical structure on the magnetic measurements used as diagnostics by the VS controller; the main focus is on the influence on the controller performance in the presence of large perturbations. The second objective is to design a new controlled variable to increase the performance of the VS system. The third objective is to provide an equivalent model of an ELM (Edge Localized Mode), in terms of internal plasma profile parameters via best fit of the vertical velocity estimation. The last objective is to obtain a reliable and accurate model of the overall system, based on the new platform MARTe, developed at JET and useful also for other devices.

マルチプロセッサシステムを用いた核融合プラズマ位置形状制御システムの開発

This paper describes a multi-winding transformer as a plasma initiator in tokamak devices and the application to the design of poloidal field coil power supply system. The scale of tokamak device has been considerably increased based on some experimentally obtained scalings of a energy confinement, because the plasma performance has been improved with the current and size parameters. Since large tokamaks require the hybrid functional poloidal field coil system, the each coil power supplies must prepare the high voltage generation circuit for the plasma breakdown, and it causes increase in the number of current interrupters, disconnectors and the resisters. Therefore, the plasma position and shape control tend to be difficult especially in the start-up phase of the discharge. The plasma initiator proposed in this paper makes the plasma initiation circuit to be simple by unification of the high voltage generation circuit. It has similar ampere-turn and permeance distribution with the tokamak device and produces the magnetic field necessary for plasma breakdown. An application to FER tokamak proposed by JAERI is also presented.

A new shape control algorithm has been tested by numerical simulation of a tokamak discharge in the Burning Plasma Experiment (BPX), using the TSC code. The algorithm controls the plasma shape and current according to a preprogrammed scenario without using any precalculated coil current waveforms. If the plasma parameters (heating scenario, density evolution, effective charge, etc.) change, the algorithm automatically maintains the given shape evolution and does not have to be readjusted. This allows independent programming of plasma current, density, heating power, shape and position

In order to provide efficient performance of tokamaks with vertically elongated plasma position, control systems for limited and diverted plasma configuration are required. The accuracy, stability, speed of response, and reliability of plasma position control as well as plasma shape and current control depend on the performance of the control system. Therefore, the problem of the development of such systems is an important and actual task in modern tokamaks. In this study, the measured signals from the magnetic loops and Rogowski coils are used to reconstruct the plasma equilibrium, for which linear models in small deviations are constructed. We apply methods of the H∞-optimization theory to the synthesize control system for vertical and horizontal position of plasma capable to working with structural uncertainty of the models of the plant. These systems are applied to the plasma-physical DINA code which is configured for the tokamak Globus-M plasma. The testing of the developed systems applied to the DINA code with Heaviside step functions have revealed the complex dynamics of plasma magnetic configurations. Being close to the bifurcation point in the parameter space of unstable plasma has made it possible to detect an abrupt change in the X-point position from the top to the bottom and vice versa. Development of the methods for reconstruction of plasma magnetic configurations and experience in designing plasma control systems with feedback for tokamaks provided an opportunity to synthesize new digital controllers for plasma vertical and horizontal position stabilization. It also allowed us to test the synthesized digital controllers in the closed loop of the control system with the DINA code as a nonlinear model of plasma.

A multi-loop plasma position and current control system in the Globus-M tokamak (Ioffe Institute, St. Petersburg, Russia) has been developed and modelled on a new evolution plasma code designed in MATLAB. Robust PID controllers in the plasma position control loops were tuned by the Quantitative Feedback Theory (QFT) based on the identified linear model of a nonlinear single-phase thyristor current inverter as a self-oscillating actuator.

For the purpose of stable plasma equilibrium control and detailed analysis, it is essential to reconstruct an accurate plasma boundary on the poloidal cross section in tokamak devices. The Cauchy condition surface (CCS) method is a numerical approach for calculating the spatial distribution of the magnetic flux outside a hypothetical surface and reconstructing the plasma boundary from the magnetic measurements located outside the plasma. The accuracy of the plasma shape reconstruction has been assessed by comparing the CCS method and an equilibrium calculation in JT-60SA with a high elongation and triangularity of plasma shape. The CCS, on which both Dirichlet and Neumann conditions are unknown, is defined as a hypothetical surface located inside the real plasma region. The accuracy of the plasma shape reconstruction is sensitive to the CCS free parameters such as the number of unknown parameters and the shape in JT-60SA. It is found that the optimum number of unknown parameters and the size of the CCS that minimizes errors in the reconstructed plasma shape are in proportion to the plasma size. Furthermore, it is shown that the accuracy of the plasma shape reconstruction is greatly improved using the optimum number of unknown parameters and shape of the CCS, and the reachable reconstruction errors in plasma shape and locations of strike points are within the target ranges in JT-60SA.

Currently, the fast evolution of cameras in recent years has made them promising tools for diagnostics of Tokamak. The solution presented in this paper consists of a prototype of high-speed visible image acquisition and processing system (HVIAPs) dedicated for experimental advanced superconducting tokamak shape and position control. Graphics processing unit (GPU) and field-programmable gate array (FPGA) are typically used as accelerators or co-processors in addition to a CPU. Such a heterogeneous computing system can combine the advantages of its individual components. The main imaging equipment components used for this system are four high-resolution fast cameras equipped with fiber interface. Image data from the cameras are received by the frame grabber card based on the FPGA and transmitted to the GPU via the peripheral component interconnect express interface. The software support for the system includes low-level drivers and application programming interface libraries for all components and the algorithm developed for image processing. The offline image process results are compared to equilibrium fitting code, which is a commonly used reconstruction method, with an average error of 1.5 cm. The total processing time for one frame is less than 0.3 ms.

In tokamak and other fusion devices, magnetic control is the main tool that allows to regulate the plasma current, position and shape; it is in charge of actuating the desired plasma current waveform, steering the plasma position to a given set point and maintain the plasma shape close to a prescribed plasma equilibrium. This work describes the application of several physics concepts and computational tools in order to obtain a novel optimal controller for the plasma centroid position, which has been implemented and tested in the real-time plasma control system at the ISTTOK tokamak. A key point for the development of the new control system was the installation of a recently upgraded hardware, that numerically integrates in real-time the magnetic probes signals.

POSITION AND SHAPE CONTROL ON ASDEX-UPGRADE

Various systems of magnetic control of plasma position, current, and shape are considered in vertically elongated tokamaks including spherical tokamaks being in operation. The systems of Resistive Wall Modes suppression are described. Tokamaks constructions and cross-sections, structural schemes of plasma control systems are given; heed is paid to operating principles of plasma control systems; experimental results of plasma control in tokamaks are given. Various engineering realizations of magnetic plasma control systems in tokamaks are presented.

The need to improve the performance of modern tokamak operations has led to a further development of the plasma shape and position control systems. In particular, extremely elongated plasmas, with high vertical-instability growth rate, are envisaged to reach the required performance for ignition. This request for better performance from the experimentalists' side has motivated the development of the new vertical-stabilization (VS) system at the JET tokamak, which has been proposed within the Plasma Control Upgrade project. The main aim of the project is to enhance the capabilities of the plasma vertical position control system in order to operate with very highly elongated plasmas in the presence of large perturbations. This paper focuses on the new software architecture of the VS system, which relies on a highly configurable real-time framework. Due to its flexibility, the new VS system executes different control algorithms, and it schedules the one which maximizes the performance in each plasma phase.

A key component of the DIII-D Advanced Tokamak and Radiative Divertor Programs is the development and implementation of methods to actively control a large number of plasma parameters. These parameters include plasma shape and position, total stored energy, density, RF loading resistance, radiated power and more detailed control of the current profile. To support this research goal, a flexible and easily expanded digital control system has been developed and implemented. We have made parallel progress in modeling of the plasma, poloidal coils, vacuum vessel, and power system dynamics and in ensuring the integrity of diagnostic and command circuits used in control. Recent activity has focused on exploiting the mature digital control platform through the implementation of simple feedback controls of more exotic plasma parameters such as enhanced divertor radiation, neutral pressure and Marfe creation and more sophisticated identification and digital feedback control algorithms for plasma shape, vertical position, and safety factor on axis (q/sub 0/). A summary of recent progress in each of these areas will be presented.