Poloidal Coil Currents Research Articles

Conceptual design of a heavy ion beam probe for the QUEST spherical tokamak.

A heavy ion beam probe (HIBP) has been designed for the QUEST spherical tokamak to measure plasma turbulence and the profiles of electric potential profiles. Using a cesium ion beam with an energy of several 10keV, the observable region covers most of the upper half of the plasma. Although the probe beam is deflected by the poloidal magnetic field produced by plasma current and poloidal coil currents, it can be detected under plasma current up to 150kA by modifying the trajectories with two electrostatic sweepers. According to the numerical estimation of the intensity of the detected beam, sufficient signal intensity for measuring plasma turbulence can be obtained over almost the measurable area when the electron density is up to 1 × 1019 m-3, which is larger than the cut-off density of electron cyclotron heating in QUEST. The performance of the designed HIBP is sufficient to explore the mechanisms of heat and particle transport in magnetically confined plasmas, including the influence of plasma wall interactions, which is a goal of the QUEST project.

Magnetic reconnection process in transient coaxial helicity injection

The physics of magnetic reconnection and fast flux closure in transient coaxial helicity injection experiments in NSTX is examined using resistive MHD simulations. These simulations have been performed using the NIMROD code with fixed boundary flux (including NSTX poloidal coil currents) in the NSTX experimental geometry. Simulations show that an X point is formed in the injector region, followed by formation of closed flux surfaces within 0.5 ms after the driven injector voltage and injector current begin to rapidly decrease. As the injector voltage is turned off, the field lines tend to untwist in the toroidal direction and magnetic field compression exerts a radial J × B force and generates a bi-directional radial Etoroidal×Bpoloidal pinch flow to bring oppositely directed field lines closer together to reconnect. At sufficiently low magnetic diffusivity (high Lundquist number), and with a sufficiently narrow injector flux footprint width, the oppositely directed field lines have sufficient time to reconnect (before dissipating), leading to the formation of closed flux surfaces. The reconnection process is shown to have transient Sweet-Parker characteristics.

Experimental Study of Repetitive ST Plasma Merging by use of Oscillating Poloidal Coil Currents

Repetitive merging of spherical tokamak (ST) plasma was demonstrated in the TS-4 device using oscillating poloidal field (PF) coil currents. The oscillating induction current is transformed into plasma currents of small ST plasmas, if its polarity satisfies the Shafranov condition. This natural rectification was found to form repetitively small ST plasmas with single helicity, and their merging gradually drives the plasma current of the main ST plasmas.

Plasma modelling results and shape control improvements for NSTX

New shape control implementations and dynamics studies on the National Spherical Torus eXperiment (NSTX) (Ono et al 2000 Nucl. Fusion 40 557–61) are summarized. In particular, strike point position, X-point height and squareness control, and two new system-identification methods/control-tuning algorithms were put into operation. The PID controller for the strike point was tuned by analysing the step response of the strike point position to the poloidal coil currents, employing the Ziegler–Nichols method. An offline system identification of the plasma response to the control inputs based on ARMAX (Ljung 1999 System Identification: Theory for the User (Englewood Cliffs, NJ: Prentice-Hall)) input–output models was implemented. With this tool, rough estimates of the improvements were realized and several control improvements were identified. An online automatic relay-feedback PID tuning algorithm, which has the advantage of tuning the controller in one shot, was implemented, thus optimizing the use of experimental time. Using these new capabilities, all four upper/lower/inner/outer strike points were simultaneous controlled and a combined X-point height, strike point radius control was implemented. The new and improved control with better accuracy and robustness enabled successful plasma operations with the liquid lithium divertor. Additionally this year, the first independent squareness control was developed. This will enable better optimization of the NSTX shape for stability and high performance in the future.

Numerical investigations of axisymmetric equilibria with current holes

In this paper, we present numerical tools convenient for the investigation of tokamak discharges with current holes. These tools allow the computation of fixed- and free-boundary equilibria with current holes, their representation in Boozer co-ordinates, the determination of the corresponding magnetic fields, field line tracing and also tracing of guiding centre orbits inside the hole. The numerical methods are tested by using two example configurations with different hole sizes. Furthermore, an equilibrium reconstruction of an ASDEX Upgrade discharge is performed. The motional Stark effect polarization angles of this discharge, which indicate a small current hole in the plasma centre, are reproduced by a numerical fit of various equilibrium parameters including pressure profile, rotational transform profile and poloidal coil currents. The resulting plasma parameters lie well inside the experimental error margins.

Design study of power supplies for LHD superconducting magnets

The Large Helical Device (LHD) is designed to have six pairs of superconducting coils. In the first phase of the project, these will be energized by six steady-power supplies. These power supplies are thyristor types, which generate positive and negative voltages and large currents (up to 30 kA). Each of them has a smoothing filter which reduces the voltage ripple to less than 0.2 V, and a protection circuit to transfer rectifier current to a dump resistor and drop it rapidly during coil quenches. These six steady power supplies are powered from the commercial line. In the second phase of the project, three more power supplies powered by a motor generator (120 MVA) or superconducting energy storage will be added to the poloidal coil circuits. These will be used in experiments which require rapid changes in the poloidal coil currents during a plasma discharge of 5 s. A protection circuit is now under development, composed of disconnectors, a vacuum circuit breaker, a power fuse and diodes. It has worked well up to 21 kA, and will be tested up to 30 kA.

Analysis of tokamak configurations using the toroidal multipole method

An exact prediction of the plasma contour is important for the design of new tokamaks and for the definition of the operational scenarios of existing machines. Certain features of the plasma boundary, such as elongation and triangularity, and a detailed description of the X-point region are relevant for the plasma-limiter interaction. The toroidal multipolar moment method has been used, with fixed boundary conditions, to produce the required plasma shape. The resulting vacuum flux function has been reproduced, using the poloidal coil currents that minimize the magnetic energy as best parameters for the fitting. Examples are given for the single and double X-point configurations in JET and ITER.

Experimental Study on Electromagnetic Interactions between Plasmas and a Vacuum Vessel during Disruptions in the Hitachi Tokamak HT-2

Electromagnetic interactions between plasmas and a vacuum vessel during disruptions are examined experimentally in the Hitachi tokamak HT-2. Eddy currents which flow in the toroidal direction and poloidal coil currents are determined from the measured magnetic data. The currents enable calculation of the electromagnetic force on the vacuum vessel and resistively dissipated magnetic energy. Eddy currents and electromagnetic forces are mainly due to the plasma displacement (shell effect), not decay of the plasma current. Large plasma current quench rate -dIP/dt is associated with scraping of the plasma by the inner limiter through the rapid plasma radial movement, and the decay rate in circular plasma is twice as large as that in elongated plasma. The magnetic energy dissipation is mainly due to the eddy current of the net toroidal current mode which is induced by large current quench rate.

New Developments in the Computer Complex System for the JFT-2M Tokamak

The JFT-2M computer complex system of the Japan Atomic Energy Research Institute (JAERI) consists of a data acquisition system and a plasma control system. The data acquisition system is composed of a CAMAC serial highway, a front-end minicomputer, a super-minicomputer and their peripherals. The system constitutes a definite hierarchical structure. This provides for easy expansion and maintenance as well as for improved RAS (reliability, availability and serviceability) functions and man-machine interface capabilities. The system permits unified control and high quality analysis of experimental data from the JFT-2M tokamak, processing 1.2 megabytes of data in less than 3 minutes. The plasma control system is made up of a pre-programmed minicomputer for current form pattern generation, an analog computer for high-speed plasma control, a local data logger, and thyristor power supplies. This system controls poloidal coil current by real-time feedback method. With a learning control function added to it, the system can effectively correct steady-state deviations.