Table of contents

A `non-local' response of the plasma core triggered by peripheral plasma perturbations other than laser ablation is found in the RTP tokamak. Oblique pellet injection (OPI) has been used to induce fast cooling of the peripheral plasma. In response, an inward cold pulse (T<sub>e</sub> drop) and a slightly delayed core T<sub>e</sub> rise are observed. A somewhat similar `non-local' response is observed when the peripheral plasma is heated by modulated electron cyclotron heating or by fast current ramps, i.e. the core temperature drops in response to the peripheral heating. The plasma conditions for the occurrence of the `non-local' response have been investigated. The core T_e rise following OPI is associated with the formation of a large temperature gradient in the region 1 < q < 2. The T_e rise is largest at low electron density and for large pellet deposition radii. Above a critical density the T_e rise disappears and only the (weaker) drop in core T_e is observed. Time dependent transport simulations show that the propagation of the inward cold pulse is consistent with local transport, while the core T_e rise is a slower phenomenon requiring a large transient drop of χ_e in the region 1 < q < 2.

Combined neutral beam injection and fast wave heating at the fourth cyclotron harmonic produce an energetic deuterium beam ion tail in the DIII-D tokamak. When the concentration of thermal hydrogen exceeds ∼5%, the beam ion absorption is suppressed in favour of second harmonic hydrogen absorption. As theoretically expected, the beam absorption increases with beam ion gyro-radius; also, central absorption at the fifth harmonic is weaker than central absorption at the fourth harmonic. For central heating at the fourth harmonic, an energetic, perpendicular, beam population forms inside the q = 1 surface. The beam ion tail transiently stabilizes the sawtooth instability but destabilizes toroidicity induced Alfvén  eigenmodes (TAEs). Saturation of the central heating correlates with the onset of the TAEs. Continued expansion of the q = 1 radius eventually precipitates a sawtooth crash; complete magnetic reconnection is observed.

The prediction of the Mercier (Suydam) criterion for localized interchange modes in stellarators shows stability in cases in which global modes are unstable. One case is non-resonant pressure driven instabilities with low mode numbers which become unstable even if the mode resonant surface does not exist inside the plasma column. The other case is interchange instabilities when the pressure gradient vanishes at the mode resonant surface. If the pressure becomes flat in a narrow region around the mode resonant surface, high mode number instabilities are suppressed and the beta limit at the particular resonant surface increases. The radial mode structure at nearly marginal beta also changes significantly. The properties of the non-resonant mode and the transition from a resonant to a non-resonant mode are clarified with a cylindrical plasma model for a low shear stellarator with a magnetic hill.

The effect of mode conversion on antenna-plasma coupling in the ICRH range is studied. A collisionless, weakly inhomogeneous plasma in the `slab' approximation, in which mode conversion is described by the tunnelling equation with first harmonic majority (ω = 2Ω_cM) and fundamental minority (ω = Ω_cm) ion cyclotron absorption, is considered. Green's function, connecting the electric field on an ICRH antenna with the current flowing in the same antenna, has been evaluated, essentially in closed form, and has been employed to evaluate the antenna radiation resistance. Numerical results have been obtained, for the parameters of the IGNITOR machine and a deuterium-hydrogen plasma; in the cold plasma limit, in which mode conversion effects are disregarded, our results are in agreement with those of an established code. Mode conversion effects produce a decrease of the radiation resistance for the monopole antenna configuration, and much weaker effects on the dipole configuration. This is related to the behaviour of the real part of the Green's function, which tends to vanish rapidly for toroidal wavenumbers approaching zero as a result of reflection from the mode conversion layer.

Current driven by fast Alfvén waves is measured in H mode and VH mode plasmas on the DIII-D tokamak for the first time. Analysis of the poloidal flux evolution shows that the fast wave current drive profile is centrally peaked but sometimes broader than theoretically expected. Although the measured current drive efficiency is in agreement with theory for plasmas with infrequent ELMs, the current drive efficiency is an order of magnitude too low for plasmas with rapid ELMs. Power modulation experiments show that the reduction in current drive with increasing ELM frequency is due to a reduction in the fraction of centrally absorbed fast wave power. The absorption and current drive are weakest when the electron density outside the plasma separatrix is raised above the fast wave cut-off density by the ELMs, possibly allowing an edge loss mechanism to dissipate the fast wave power since the cut-off density is a barrier for fast waves leaving the plasma.

The large variety of plasma shapes produced in the TCV tokamak places unique demands on the plasma facing surfaces. In particular, the central column graphite armour tiles are solicited during the creation of all TCV plasmas and function as power handling surfaces for both limited and diverted discharges. The higher power flux densities accompanying the addition of electron cyclotron heating systems have necessitated a new, optimized, design for these tiles. The optimization process and the subsequent new tile design are described. A basic `two point' model of the scrape-off layer plasma in conjunction with TCV equilibrium reconstructions and a simplified representation of the local magnetic field line geometry are used to impose simulated power flux densities onto a parametric toroidal tile contour. The thermo-mechanical response of the tile is then investigated via full 3-D finite element simulations accounting for the non-linear temperature dependence of the graphite thermal diffusivity and radiation from the tile surface. The final design choice is a compromise between the requirements for adequate power handling for a range of magnetic configurations, the need to protect against tile edge misalignment in the presence of grazing field line angles of incidence and the space restrictions imposed by vacuum vessel design.

The surface loss probabilities of hydrocarbon radicals on the surface of amorphous hydrogenated carbon (C:H) films are investigated by depositing films inside a cavity with walls made from silicon substrates. This cavity is exposed to a discharge using different hydrocarbon source gases, and particles from the plasma can enter the cavity through a slit. The surface loss probability β is determined by analysis of the deposition profile inside the cavity. This surface loss probability corresponds to the sum of the probabilities of effective sticking on the surface and of formation of a non-reactive volatile product via surface reactions. By comparing the deposition profiles measured in CH₄, C₂H₂ and C₂H₄ discharges one obtains for C₂H radicals β = 0.90 ±0.05 and for C₂H_x>2 radicals β = 0.35 ± 0.1, whereas the surface reaction probability for CH₃ is below 10^-2, as known from the literature. The growth rate of C:H films is, therefore, very sensitive to any contribution of C₂H_x species in the impinging flux from a hydrocarbon discharge. The very same growth precursors can be formed in a divertor plasma and should therefore dominate the formation of re-deposited layers. A scenario for the occurrence of these re-deposited films in fusion experiments on the basis of typical divertor plasma and surface parameters is being discussed. Strategies are proposed for prevention of these re-deposited layers and for their removal.

High fusion performance has been achieved in optimized shear discharges in JET with the early appearance of internal transport barriers (ITBs) in the high power phase in both DD and DT plasmas. An ITB has been produced in a large variety of conditions, including LHCD only, ICRH only, NBI only and NBI + ICRH with LHCD preheat. An ITB may coexist with the edge barrier. The high pressure gradient produced by ICRH combined with high power NBI heating starting in a low density plasma with q(0) ≈ 1.5-2 is the main prerequisite for an efficient ITB in high performance discharges. Strong coupling between q profiles, MHD activity, energy confinement and fusion yield has been observed. The transition from L to H mode can be triggered by MHD events. The transition is accompanied by significant momentary changes in the density fluctuation spectrum in the peripheral plasma. The ITB may persist for some time in both ELM-free and ELMy H mode.