Table of contents

Deuterium atom beam injectors developed for ITER plasma heating and current drive are based on negative ion acceleration and further neutralization with a gas target. The maximal efficiency of the gas stripping process is 60%. The replacement of the gas neutralizer by a plasma one must increase the neutral yield to 80%. An overview of experimental studies of microwave discharges in a multicusp magnetic system chosen as the base device for plasma neutralizer (PN) realization and design development of PNs for ITER neutral beam injectors (NBIs) is presented. The experimental results achieved with the PN model PNX-U are discussed. Plasma confinement, gas flows and ionization degree were investigated. High density plasmas with n_e ∼10¹⁸ m^-3 with low electron and ion temperatures ( ≈ 5-6 eV) and a high ionization degree (not less than 40%) at its centre have been generated in operation with argon.

A kinetic model for second harmonic electron cyclotron breakdown in stellarators, based on the results obtained from non-linear wave-particle interaction theory, is presented. A set of particle balance equations that describes the H₂ ionization chain by means of an electron momentum space discretization is introduced and solved by numerical integration. The time evolution of the various species densities is determined and breakdown related predictions are obtained for different prefill pressure conditions and different values of injected power. Analysis of the TJ-II stellarator experimental data shows good agreement with the theoretical results. This is the first time, to the authors' knowledge, that precise predictions are directly compared with experimental behaviour.

The Full Scale Sector Model Project, which was initiated in 1995 as one of the Seven Large Projects for ITER R&D, has been continued with the joint effort of the ITER Joint Central Team and the Japanese, Russian Federation and United States Home Teams. The fabrication of a full scale 18° toroidal sector, which is composed of two 9° sectors spliced at the port centre, was successfully completed in September 1997 with a dimensional accuracy of ±3 mm for the total height and total width. Both sectors were shipped to the test site at the Japan Atomic Energy Research Institute and the integration test of the sectors was begun in October 1997. The integration test involves the adjustment of field joints, automatic narrow gap tungsten inert gas welding of field joints with splice plates and inspection of the joints by ultrasonic testing, as required for the initial assembly of the ITER vacuum vessel. This first demonstration of field joint welding and the performance test of the mechanical characteristics were completed in May 1998, and all the results obtained have satisfied the ITER design. In addition to these tests, integration with the midplane port extension fabricated by the Russian Home Team by using a fully remotized welding and cutting system developed by the US Home Team was completed in March 2000. The article describes the progress, achievements and latest status of the R&D activities for the ITER vacuum vessel.

Pellet injection has been used as a primary fuelling scheme in the Large Helical Device. With pellet injection, the operational region of NBI plasmas has been extended to higher densities while maintaining a favourable dependence of energy confinement on density, and several important values, such as plasma stored energy of 0.88 MJ, energy confinement time of 0.3 s, β of 2.4% at 1.3 T and density of 1.1 × 10²⁰ m ^-3, have been achieved. These parameters cannot be attained by gas puffing. Ablation and the subsequent behaviour of the plasma have been investigated. The measured pellet penetration depth estimated on the basis of the duration of the H_α emission is shallower than the depth predicted from the simple neutral gas shielding (NGS) model. It can be explained by the NGS model with inclusion of the effect of fast ions on the ablation. Just after ablation, the redistribution of the ablated pellet mass was observed on a short timescale (∼400 ms). The redistribution causes shallow deposition and low fuelling efficiency.

A complete system of transport equations with all the important perpendicular currents is derived for the simulation of tokamak edge plasma. These transport equations are implemented in the B2.5 code and solved for the parameters of the ASDEX Upgrade tokamak. The relative roles of different mechanisms of transverse conductivity in the formation of the potential profile are studied. It is demonstrated that a reasonable potential distribution in the tokamak edge plasma can be obtained without an ad hoc assumption of the existence of the anomalous perpendicular conductivity. The role of E × B drifts in the redistribution of edge plasma and closing of the currents in the plasma is analysed.

Different types of central relaxation oscillations are observed in the presence of ECH depending on the location of the deposited power. In the TCV tokamak, normal sawteeth, i.e. triangular sawteeth similar to ohmic sawteeth, and saturated sawteeth are observed with central ECH power deposition, while giant sawteeth and `humpback' oscillations occur when heating close to the sawtooth inversion surface of the local soft X ray emissivity. New measurements with high temporal resolution show that the crash phase of these sawtooth types is accompanied by a reconnection process associated with an m/n = 1 resistive internal kink mode. After the sawtooth crash, full magnetic reconnection is observed in normal and in saturated sawteeth, while for giant and humpback sawteeth the reconnection process is incomplete and poloidally asymmetric temperature profiles persist after the crash. The detailed dynamics of the magnetic island associated with the resistive internal kink mode are described by a displacement function which is inferred from the experimental data. In normal sawteeth, the kink mode is destabilized just before the crash, while in all other sawtooth types a magnetic island exists for a significant fraction of the sawtooth period. The different types of sawteeth have been simulated using a numerical code based on a theoretical model which describes the evolution of the electron temperature in the presence of localized heat sources and of a magnetic m/n = 1/1 island.

Under typical conditions of high recycling divertor plasmas, the ion diamagnetic flow makes the largest contribution to the ion poloidal rotation in the SOL. Owing to toroidal effects, the net poloidal rotation in the SOL gives rise to ion parallel Pfirsch-Schlüter flows, which are sustained by an up-down pressure asymmetry which reverses with toroidal field (B_ϕ) reversal. The up-down pressure asymmetry in turn gives rise to net, surface averaged, radial currents _r arising from the surface averaging of local diamagnetic currents j_r = (1/rB)(∂p/∂θ) due to toroidal effects. The radial divergence of radial currents _r can exist in the SOL because of electrical contact with the target. The resulting net toroidal _r B_θ force in the direction of the main plasma current creates an in-out pressure asymmetry in favour of the ion drift side (which is the inner side in normal B_ϕ and the outer side in reversed B_ϕ plasmas). This pressure asymmetry should result in higher density and low temperature plasma in the inner divertor leg in normal B_ϕ discharges and should also make the distribution of density and temperature between the targets more equal in reversed B_ϕ discharges, consistent with the main trends observed in experiments with toroidal field reversal. The proposed mechanism can therefore provide an alternative explanation (to the effect of the E × B drifts analysed earlier) for the observed changes in target asymmetries associated with B_ϕ reversal.

The results of an extensive study on transport mechanisms and on improved confinement scenarios in RFX are reported. The scaling of the thermal conductivity in the core with the Lundquist number indicates that the magnetic field in this region is not fully stochastic, as proved by the existence of thermal barriers observed in single helicity configurations. The electrostatic transport at the edge has been proved to depend on the highly sheared E × B flow, which has been interpreted using fluid and Monte Carlo models. Regimes of improved confinement have been obtained in the core by poloidal current drive techniques, and the electrostatic transport has been reduced at the edge by biasing experiments. A radiation mantle caused by impurity seeding has been found to successfully reduce the local plasma–wall interaction without causing a significant deterioration in the plasma performance.

Linear and non-linear properties of slab drift waves in the negative sheared slab configuration modelling of the q_min surface region in negative shear tokamaks are studied, where q_min is the minimum value of the safety factor q. Linear calculations show that both the slab ion temperature gradient (ITG) driven mode and the slab electron temperature gradient (ETG) driven mode become strongly unstable around the q_min surface. Non-linear simulations are performed for ETG turbulence, which evolves on a much faster timescale than ITG turbulence. It is found that quasi-steady E_r × B zonal flows are generated by an inverse wave energy cascade process. Linear stability analyses of the electrostatic Kelvin-Helmholtz (KH) mode show that the quasi-steady E_r × B zonal flow profile is closely related to the q profile or to the magnetic shear, which has a stabilizing effect on the KH mode. It is shown that the microscopic quasi-steady E_r × B zonal flows arising from ETG turbulence have a strong stabilizing effect on the slab ITG mode.

High resolution Thomson scattering temperature profiles at the TJ-II stellarator are subjected to spectral analysis. The obtained k spectrum is very robust in shape. While its amplitude has been shown previously to scale inversely with the collisionality, here it is found that its shape is independent of plasma configuration (magnetic well, iota), global plasma parameters (temperature, density) and radial position (centre, edge). The shape is also found to be very similar to that of k spectra obtained with Langmuir probes in the edge of TJ-II and of k spectra obtained using other techniques in other devices. The spectral shape is characterized by a `knee' at k = 500-1000 m^-1 and a steep decay at large k. Self-similarity analysis reveals the existence of a high degree of self-organization dominated by structures of around 2.5 cm in size.

Active feedback of non-axisymmetric external modes in tokamaks is studied by means of combined MHD and control analysis. The MARS code for toroidal MHD stability analysis has been extended to compute transfer functions for the electromagnetic response of the plasma, the coils and the resistive wall. These transfer functions are used for controller design. The controller is designed to allow for the longest possible response time of the amplifier-feedback coil circuit, such that certain predetermined performance criteria are satisfied. Calculations are presented for high beta advanced tokamak equilibria with ITER-FEAT shape. With an array of feedback coils that has only one coil in the poloidal direction, and assuming a single resistive wall, control of modes with toroidal mode number n = 1 is found to be possible for β_N up to 5, or twice the no-wall limit. Sensors for the poloidal field are superior to sensors for the radial field. Feedback coils with a rather broad cross-section have significant advantages over thin-wire coils. The controller needs to have significant derivative action, but the requirements on the time constant of the amplifier-feedback coil circuit are moderate. This time needs to be less than a few resistive wall times, and broad strips allow an even longer time constant.