Table of contents

The effect of plasma shaping on the damping of radially extended Alfvén eigenmodes (AEs) is investigated experimentally during the limiter phase of JET plasma discharges. The AE damping rate is observed to increase with increasing elongation, triangularity and edge magnetic shear. These parameters could therefore be used to control the AE stability in real time.

Naturally occurring error fields in tokamaks, which arise from misalignments of the external field coils, may trigger the onset of tearing modes. The conditions under which a static error field is able to lock a rotating tearing mode and how this process takes place are discussed. The analysis presented contributes to a new understanding and interpretation of mode locking, given in terms of the superposition of a slipping layer (a radial layer of very fast mode phase variations) and the tearing layer, where reconnection takes place. In addition, a stabilizing operating window is found to exist, independent of the phase time evolution of the mode.

A new method of gas fuelling, pulsed supersonic molecular beam injection (SMBI), has been successfully developed and used in the HL-1M tokamak. SMBI is an attempt to enhance the penetration depth and the fuelling efficiency, as well as to reduce both the injected particle-wall surface interaction and the impurity content in the plasma. SMBI can be considered a significant improvement over conventional gas puffing. With a penetration depth of hydrogen particles greater than 15 cm, the rate of increase of electron density, d_e/dt, was up to 7.2 × 10²⁰m^-3 s^-1 without disruption, and the highest plasma density was _e = 8.2 × 10¹⁹ m^-3. The density profile peaking factor Q_n reached a maximum value of more than 1.67 after SMBI. The energy confinement time τ_E measured by diamagnetism is 10-30% longer than that with gas puffing with the other discharge conditions kept the same. SMBI has recently been improved to enhance the flux of the beam and to allow a survey of the cluster effect within the beam. A series of new phenomena show the interaction of the beam (including clusters) with the toroidal plasma, which indicates that hydrogen clusters may be produced in the beam according to the Hagena empirical scaling law of clustering onset, Γ* = kd^0.85P₀/T₀^2.29. If Γ* > 100, clusters will form. In the present experiment Γ^* is about 127.

A marked feature of the pebble divertor system is an effect caused by the use of a functional multilayer coated pebble, which consists of a surface plasma facing layer, an intermediate tritium permeation barrier layer and a kernel for heat removal. The dimensions, structure and irradiation conditions of the pebbles are the important issues for the development of the pebble divertor. From the viewpoint of resistance to the induced thermal stress, the pebbles are taken to be as small as possible. On the other hand, from the viewpoint of pumping performance, the suitable irradiation temperature range of the surface layer of the pebbles was estimated from experimental results and a numerical analysis. A pumping process enhanced by dynamic retention is available to extend the upper limit of the allowable irradiation temperature range from 900 to 1100 K. On taking the temperature rise limitation due to the pumping effect and the fracture strength due to the induced thermal stress limitation, it was found that a pebble diameter of 1-2 mm is possible in about 20 MW/m² for the SiC kernel and one of 2-3 mm in less than 30 MW/m² for the graphite kernel.

Results obtained in the initial experimental phase of Heliotron J are reported. Electron beam mapping of the magnetic surfaces at a reduced DC magnetic field has revealed that the observed surfaces are in basic agreement with the ones calculated on the basis of the measured ambient field around the device. For 53.2 GHz second harmonic ECH hydrogen plasmas, a fairly wide resonance range for breakdown and heating by the TE₀₂ mode has been observed in Heliotron J as compared with that in Heliotron E. With ECH injection powers up to ≈ 400 kW, diamagnetic stored energies up to ≈ 0.7 kJ were obtained without optimized density control.

A possibility of `advanced tokamak' operation based on RF heating and current drive is studied using 1-D transport modelling for long pulse tokamaks such as Tore Supra. A current drive with lower hybrid and fast magnetosonic waves and strong electron heating with ion cyclotron waves are applied in this scenario. The self-consistent evolution of the energy, particles and current densities is simulated, starting from an ohmic plasma and arriving at a non-inductive steady state equilibrium. An estimation of the RF power required for the setting up of this equilibrium characterized by an improved confinement due to the building up of particle and energy internal transport barriers in the reversed magnetic shear configuration is presented. The maximum bootstrap current fraction attained in this scenario is about 50%. A limit on the bootstrap current fraction in plasmas with an L mode edge and improved core confinement produced due to the stabilizing effect of a low or negative magnetic shear is found. The conventional operational techniques with fixed plasma current or controlled plasma flux and non-inductive current are compared with the operation with the current profile control algorithm, and the advantages of current profile control are shown. The results presented are relevant to present tokamaks operating with RF heating and current drive and possibly to future tokamak projects aimed at advanced operation.

The characteristics of internal transport barrier (ITB) structures are studied and active ITB control has been developed in JT-60U reversed shear plasmas. The following results are found. Outward propagation of ITBs with steep T_i gradients is limited to the minimum safety factor location ρ_qmin. However, ITBs with reduced T_i gradients can move to the outside of ρ_qmin. The lower boundary of the ITB width is proportional to the ion poloidal gyroradius at the ITB centre. Furthermore, active control of the ITB strength based on modification of the radial electric field shear profile is successfully demonstrated by toroidal momentum injection in different directions or an increase of heating power by neutral beams.

The process of fuelling by injection of a spheromak-like compact toroid (SCT) is investigated by using MHD numerical simulations, where the SCT is injected into a magnetized target plasma region corresponding to a fusion device. In a previous study, the authors proposed a theoretical model to determine the penetration depth of the SCT into the target region on the basis of simulation results; in that study the SCT is decelerated not only by the magnetic pressure force but also by the magnetic tension force. However, since both ends of the target magnetic field are fixed on the boundary wall in the simulation, the deceleration caused by the magnetic tension force would be overestimated. In the present study, the dependence of the boundary condition of the target magnetic field on the SCT penetration process is examined. On the basis of these results, the theoretical model previously proposed is improved to include the effect that the fluctuations of the target magnetic field propagate with the Alfvén velocity. In addition, by carrying out the simulation with the torus domain, it is confirmed that the theoretical model can be applied to estimate the penetration depth of the SCT under such conditions. The dependence of the injection position (side injection and top/bottom injection) on the penetration process is also examined.

The role of turbulence stabilization by sheared E × B flow is analysed for different types of internal transport barrier (ITB) in JET discharges. After the ITB formation, the shearing rate ω_s is found to consistently exceed a rather crude estimate for the ion temperature gradient driven turbulence. The roles played by toroidal plasma rotation and by density and temperature gradients in both the radial electric field E_r and the shearing rate ω_s are evaluated. It turns out that the major contribution in triggering the positive feedback leading to ITB onset is given by the toroidal rotation.

The rich phenomenology of internal transport barriers (ITBs) observed in tokamaks includes a poloidal spin-up precursor for balanced injection neutral beam heating and stepwise expansion of the barrier for unbalanced injection. Examples of numerical simulations of these phenomena are presented. Two drift wave based predictive transport models (GLF23 and multimode) are used. Both models include the suppression of ion temperature gradient modes as the E × B shear approaches the computed maximum linear growth rate. Modelling of discharges with ITBs from the DIII-D, JET and TFTR tokamaks are compared.

The energy confinement and thermal transport characteristics of net current free plasmas in regimes with much smaller gyroradii and collisionality than previously studied have been investigated in the Large Helical Device (LHD). The inward shifted configuration, which is superior from the point of view of neoclassical transport theory, has revealed a systematic confinement improvement over the standard configuration. Energy confinement times are improved over the International Stellarator Scaling 95 by a factor of 1.6 ±0.2 for an inward shifted configuration. This enhancement is primarily due to the broad temperature profile with a high edge value. A simple dimensional analysis involving LHD and other medium sized heliotrons yields a strongly gyro-Bohm dependence (τ_EΩ ∝ ρ^*-3.8) of energy confinement times. It should be noted that this result is attributed to a comprehensive treatment of LHD for systematic confinement enhancement and that the medium sized heliotrons have narrow temperature profiles. The core stored energy still indicates a dependence of τ_EΩ ∝ ρ^*-2.6 when data only from LHD are processed. The local heat transport analysis of discharges dimensionally similar except for ρ^* suggests that the heat conduction coefficient lies between Bohm and gyro-Bohm in the core and changes towards strong gyro-Bohm in the peripheral region. Since the inward shifted configuration has a geometrical feature suppressing neoclassical transport, confinement improvement can be maintained in the collisionless regime where ripple transport is important. The stiffness of the pressure profile coincides with enhanced transport in the peaked density profile obtained by pellet injection.

An experimental analysis is discussed of different phases of ECRH H mode discharges on COMPASS-D, from L-H transition triggering to steady state. Comprehensive high resolution measurements in the transport barrier region have enabled significant progress to be made in assessing H mode transition models, together with additional experiments aimed at resolving controlling trigger mechanisms. The application of four possible transition models to the data indicates that all demonstrate limiting or critical parameter values at the transition. One of them, based on Alfvén drift wave turbulence suppression, exhibits precursor behaviour in the stability regime diagram and predicts the density and magnetic field dependences of the H mode power threshold observed on COMPASS-D. Analysis of the evolution of the local radial electric field and its shear indicates that significant shear only develops after the L-H transition. Stationary ELMy H modes are achieved at high ρ^* with good confinement, and dimensionless scaling over a range of ν^* has been carried out, providing valuable confinement data in a regime where heat is deposited primarily to the electrons.

The article reports the results of experimental campaigns on plasma ohmic heating performed during 1999-2000 on the spherical tokamak Globus-M. Later experimental results with the tokamak fed by thyristor rectifiers are presented in detail. The toroidal magnetic field and plasma pulse duration in these experiments were significantly increased. The method of stray magnetic field compensation is described. The technology of vacuum vessel conditioning, including boronization of the vessel performed at the end of the experiments, is briefly discussed. Specific features of neutral gas column breakdown in spherical tokamaks by applied inductive voltage are outlined. Also discussed is the influence of ECR preionization on the breakdown conditions. Experimental data on plasma column formation and current ramp-up in different regimes of operation with the magnetic flux of the central solenoid limited to ∼100 mWb are presented. A significant reduction of the background density after boronization (below 2 × 10¹⁸m^-3) allows the density to be completely controlled with external gas puffing and makes the influence of the wall negligible. The magnetic flux consumption efficiency is discussed. The results of magnetic equilibrium simulations are presented and compared with experiment. Ramp-up of the plasma current of 0.25 MA for a time interval of about 0.03 s with about 0.02 s flat-top at a toroidal field strength of 0.35 T allows the conclusion that the target design parameters of Globus-M could be achieved in a double swing regime.

While beam steering has been used for control of the location of the ECRH power absorption in tokamaks, alternative or complementary techniques are worthy of investigation to improve the control of power deposition and to make the launcher system more resistant against the harsh environment of a reactor. Here, frequency tuning of ECRH power for reactor scale tokamak plasma heating is investigated. Particular emphasis is put on the adjustment of the power deposition near the m/n = 2 tearing mode resonance. Mode stabilization is studied by combining the ECRH absorption and MHD stability models with transport code calculations of the plasma evolution. The positioning of the power deposition should be performed with a moderate speed of frequency tuning (1 GHz/s) over a sufficient frequency range ( ≈ ±5 GHz). This is found to be within the reach of present gyrotron technology. The inductance and the width of the power deposition set lower limits for the step duration and power as well as an upper limit for the frequency step which is required in tuning for stabilization. For further fine tuning of the power deposition and current density profiles, fast frequency tuning by a properly programmed frequency sweeping of the order of 10 ms can be effective. A review of available methods to tune the frequency of gyrotrons is provided and their suitability for performing MHD mode stabilization is discussed.

Within the framework of the Adaptive Plasma Experiment (APEX) conceptual project, a trap with closed magnetic field lines, the Experimental Pseudo-Symmetric Closed Trap (EPSILON), is examined. The APEX project is aimed at theoretical and experimental development of the physical foundations for a steady state thermonuclear reactor designed on the basis of an alternative magnetic trap with tokamak-like large β plasma confinement. A discussion is given of the fundamental principle of pseudo-symmetry, which a magnetic configuration with tokamak-like plasma confinement should satisfy. Examples are given of calculations in the paraxial approximation of pseudo-symmetric curvilinear elements with a poloidal modulus B isoline. The EPSILON trap, consisting of two direct axisymmetric mirrors linked by two curvilinear pseudo-symmetric elements, is considered. To increase the equilibrium β, the plasma currents are short-circuited within curvilinear equilibrium elements. An untraditional scheme of MHD stabilization for a trap with closed field lines by use of axisymmetric mirrors with a divertor is analysed. The experimental installation EPSILON-One Mirror Element (OME), which is under construction for experimental investigation of stabilization by divertor, is discussed. The opportunity for applying the ECR method of plasma production in EPSILON-OME in conditions of high density and low magnetic field is examined.

Vertical stability and control are examined for a tokamak configuration intended to be a generic representation of next step devices. Vertical stability calculations show that a critical resistive wall location can be determined for realistic structures and that the introduction of small amounts of low resistivity material to an all steel structure can significantly reduce the vertical instability growth rate. Vertical control simulations show that internal control coils require significantly less feedback power than external coils, and that low resistivity materials can allow either very low feedback powers or the coils to be located externally with reasonable feedback powers.