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The chemical erosion of carbon in interaction with a hydrogen plasma has been studied in detail in ion beam experiments, and erosion yield values are available as a function of ion energy and surface temperature. However, the conditions in the ITER divertor cannot be simulated by ion beam experiments, especially as far as ion flux is concerned.

Therefore, a joint attempt was made through the EU Task Force on plasma–wall interaction and the international tokamak physics activity involving seven different fusion devices and plasma simulators to clarify the flux dependence. For each data point the local plasma conditions were normalized to an impact energy of 30 eV, care was taken to select data for a surface temperature close to the maximum yield or room temperature and the calibration of the diagnostic was performed in situ. Through this procedure the previous large scatter was significantly reduced, revealing a clear trend for a decreasing yield with increasing ion flux, Φ. After the attribution of an error to each data point a fit using Bayesian probability analysis was performed, yielding a decrease in the erosion yield with Φ^−0.54 at high ion fluxes.

Transport simulations of high density scenarios of ITER and other reactor-grade devices require a scaling law for the effective heat diffusivity, χ, in the ELMy H-mode regime. A comprehensive empirical scaling, χ_H98, compatible with the ITER reference scaling, ITERH-98P(y, 2), for the thermal energy confinement time has been set up. It follows from a power law ansatz for χ and integration of the single-fluid energy equation and recovers all the exponents of the global confinement law. The dependences on temperature and temperature gradient are consistent with the power degradation of confinement and the experimental χ profiles. The χ_H98 scaling is validated by JET, DIII-D, ASDEX Upgrade and ASDEX discharges covering a wide parameter range. Simulations of the inductive scenario of ITER with χ_H98 yield an energy confinement time which agrees with the global scaling prediction.

At TEXTOR, an O-mode heterodyne reflectometer system is installed and operated for the measurement of plasma density fluctuations and turbulence investigations. With two antenna arrays in the equatorial and top positions having two and three horn antennae, respectively, poloidal correlations are investigated under different plasma scenarios. From the amplitude, cross-phase and coherency spectrum, differences in the ohmic and auxiliary heated discharges are investigated. Furthermore the dynamic behaviour of the turbulence is studied in the SOC–IOC transition and in the precursor phase of a disruption. For the latter an increased integrated power spectral density was observed at the X-point of the mode compared with the O-point. Stationary m = 2 mode activity is observed for the first time at TEXTOR by reflectometry. The fluctuation level is calculated for different conditions and rises significantly increasing heating power which is consistent with the L-mode confinement degradation. Correlation measurements yield the measured phase delays which are used to calculate the poloidal phase velocity perpendicular to the magnetic field. In ohmic plasmas the turbulence rotates like a 'rigid body' with constant angular velocity inside the q = 2 surface. The rigid body rotation is broken up during tangential neutral beam injection. From the deduced poloidal wavenumber of the turbulence, most likely ion temperature gradient modes are the driving mechanism of the turbulence.

Magnetohydrodynamic (MHD) induced energetic ion loss in neutral beam heated H-mode discharges in the National Spherical Torus Experiment (NSTX) is discussed. After H-mode onset, the neutral particle analyser (NPA) spectrum usually exhibits a significant loss of energetic ions mainly for E > E_b/2 where E_b is the beam injection energy, although this loss occasionally extends to lower energy. The magnitude of the energetic ion loss diminishes with increasing tangency radius of the NPA sightline, increasing the toroidal field and the neutral beam injection energy. Modelling suggests that MHD-induced ion loss is enhanced during H-mode operation due to an evolution of the q and beam deposition profiles that feeds both passing and trapped ions into the region of the plasma affected by the low-n MHD activity. It must be emphasized that this loss mechanism is a pressure profile effect that can sometimes occur in unusually high density L-mode discharges, but almost always is observed in H-mode discharges because of the intrinsic broad electron density profile. Analysis of the particle interaction with a model magnetic perturbation supports the energy selectivity of the observed MHD-induced loss. Transport analysis using a fast-ion diffusion model to emulate MHD-induced energetic ion loss shows significant modifications of the heating produced by the neutral beam which changes the inferred power balance, thermal diffusivities and the thermal energy confinement time. In the case cited herein, MHD-induced energetic ion loss increased the thermal energy confinement, τ_E, by ∼15% and reduced the toroidal beta, β_T, by ∼7%. An accounting of energetic ion loss is therefore important for proper analysis of power balance and transport in plasmas exhibiting MHD-induced energetic ion loss.

W7-X, a five-period, fully optimized stellarator, currently under construction at IPP-Greifswald, Germany, is built with superconducting coils to show the steady state capability of stellarators. However, the steady state needs continuous equilibrium information for monitoring and controlling the discharge. Although the timescales are long compared with tokamaks, the computational effort for calculating three-dimensional magnetohydrodynamic equilibria is also orders of magnitude higher. This has led us to start the development of a fast equilibrium recovery for W7-X. As a starting point and also for investigating the richness of magnetic configurations, of which only nine physically interesting examples have been examined till now, a fast recovery of vacuum magnetic configurations, described by the flux surface geometry and profile parameters, is carried out using the function parametrization (FP) and artificial neural networks (ANN) methods. Additionally, we parametrize the detectable major magnetic island chains (5/6, 5/5, 5/4) in the form of their locations, , and their width, w^(is). The quality of FP recovery is compared with ANN, where the vacuum parameters are non-linearly regressed in terms of linear combinations of the coil currents. The results show that a quadratic FP model is generally sufficient for a good recovery of the parameters which are related to the magnetic axis. However, a cubic model is necessary for modelling accurately the magnetic island-related parameters. ANN models offer no improvement over the cubic FP model.

Advanced tokamak discharges on ASDEX Upgrade can generate an internal transport barrier (ITB) during the current ramp-up phase early in the discharge. Formation of the ITB has become more reliable with the discovery that a low density is necessary for it to form. These ITBs form in very low or negative central shear regions. There is no clear evidence of integer q magnetic surfaces triggering the ITB phase. The appearance of an integer q surface (usually q = 2) often leads to a second ITB (i.e. another steepening in the ion temperature gradient) inside the first ITB.

Investigation of possible turbulence suppression mechanisms suggests that something is missing from the generally accepted model of sufficient E × B sheared flow suppressing the turbulence since the E × B shearing rate was not greater than, or comparable with, the maximum linear growth rate in about half of the ITB discharges analysed. Thus, it does not explain the reduced radial transport observed in these discharges.

Experiments in the DIII-D tokamak show that the n = 1 ideal kink can be stabilized by a resistive wall if the plasma is rotating fast enough. A database of the onset of the n = 1 resistive wall mode as a function of the equilibrium toroidal magnetic field, the plasma density and the toroidal rotation has been assembled for plasmas with beta between the theoretically predicted no wall and ideal wall stability limits. The critical rotation frequency is found to scale as the inverse of the Alfvén time with ω _ϕ τ_A ≈ 0.02 (evaluated at the q = 2 surface at ρ ≈ 0.6) or ω_ϕ τ_S ≈ 0.7, where τ_S is the sound time. The dependence of ω_ϕ τ _A or ω_ϕ τ_S on β_N/β_{N,no wall} from 1–2 is weak and suggests the plasmas are in the 'intermediate dissipation' regime.