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W7-X completed its plasma operation in hydrogen with island divertor and inertially cooled test divertor unit (TDU) made of graphite. A substantial set of plasma-facing components (PFCs), including in particular marker target elements, were extracted from the W7-X vessel and analysed post-mortem. The analysis provided key information about underlying plasma–surface interactions (PSI) processes, namely erosion, transport, and deposition as well as fuel retention in the graphite components. The net carbon (C) erosion and deposition distribution on the horizontal target (HT) and vertical target (VT) plates were quantified and related to the plasma time in standard divertor configuration with edge transform ι = 5/5, the dominant magnetic configuration of the two operational phases (OP) with TDU. The operation resulted in integrated high net C erosion rate of 2.8 mg s−1 in OP1.2B over 4809 plasma seconds. Boronisations reduced the net erosion on the HT by about a factor 5.4 with respect to OP1.2A owing to the suppression of oxygen (O). In the case of the VT, high peak net C erosion of 11 μm at the strike line was measured during OP1.2B which converts to 2.5 nm s−1 or 1.4 mg s−1 when related to the exposed area of the target plate and the operational time in standard divertor configuration. PSI modelling with ERO2.0 and WallDYN-3D is applied in an interpretative manner and reproduces the net C erosion and deposition pattern at the target plates determined by different post-mortem analysis techniques. This includes also the 13C tracer deposition from the last experiment of OP1.2B with local 13CH4 injection through a magnetic island in one half module. The experimental findings are used to predict the C erosion, transport, and deposition in the next campaigns aiming in long-pulse operation up to 1800 s and utilising the actively cooled carbon-fibre composite (CFC) divertor currently being installed. The CFC divertor has the same geometrical design as the TDU and extrapolation depends mainly on the applied plasma boundary. Extrapolation from campaign averaged information obtained in OP1.2B reveals a net erosion of 7.6 g per 1800 s for a typical W7-X attached divertor plasma in hydrogen.

Wendelstein 7-X (W7-X), the largest advanced stellarator, is built to demonstrate high power, high performance quasi-continuous operation. Therefore, in the recent campaign, experiments were performed to prepare for long pulse operation, addressing three critical issues: the development of stable detachment, control of the heat and particle exhaust, and the impact of leading edges on plasma performance. The heat and particle exhaust in W7-X is realized with the help of an island divertor, which utilizes large magnetic islands at the plasma boundary. This concept shows very efficient heat flux spreading and favourable scaling with input power. Experiments performed to overload leading edges showed that the island divertor yields good impurity screening. A highlight of the recent campaign was a robust detachment scenario, which allowed reducing power loads even by a factor of ten. At the same time, neutral pressures at the pumping gap entrance yielded the particle removal rate close to the values required for stable density control in steady-state operation.

It is shown by means of PIC (particle-in-cell) simulations that in magnetized plasmas with pressure gradients local charge separation occurs due to different ion and electron gyration radii. The dimensionless ratio X = n pl m/(∈ o B 2 ) being the dielectric susceptibility (with the plasma density n pl , the mass m and the magnetic field strength B) is the main parameter in the analysis.

The shielding of a test charge imbedded into a magnetized plasma has been investigated by means of a PIC (particle-in-cell) model. It is shown that the ratio of the gyration radius to the Debye length is the main parameter in the analysis. For small values of this parameter the plasma surrounding the test charge is not able to compensate this additional charge and shielding does not occur. For high values of this parameter the effect of the magnetic field vanishes and the electric forces dominate. Scaling rules are given in order to compare different sets of plasma parameters.

The next step in the Wendelstein stellarator line is the large superconducting device Wendelstein 7-X, currently under construction in Greifswald, Germany. Steady-state operation is an intrinsic feature of stellarators, and one key element of the Wendelstein 7-X mission is to demonstrate steady-state operation under plasma conditions relevant for a fusion power plant. Steady-state operation of a fusion device, on the one hand, requires the implementation of special technologies, giving rise to technical challenges during the design, fabrication and assembly of such a device. On the other hand, also the physics development of steady-state operation at high plasma performance poses a challenge and careful preparation. The electron cyclotron resonance heating system, diagnostics, experiment control and data acquisition are prepared for plasma operation lasting 30 min. This requires many new technological approaches for plasma heating and diagnostics as well as new concepts for experiment control and data acquisition.

The effect of secondary electron emission and sputtered impurity ions on the sheath potential drop has been studied. An analytical model has been developed which allows a self-consistent description of the above mentioned effects. In the analysis different materials (two kinds of graphite and tungsten) have been compared and a considerable dependence of the potential drop on the material choice has been observed. It is shown that the most of energy coming from the plasma to the target is carried by electrons because of both the decrease of potential drop caused by SEE and the enhancement of electron current to the plate caused by sputtering.

The main atomic processes occuring at the solid wall during plasma wall interaction are presented. The data are based on experimental results obtained under laboratory and plasma conditions and/or by computer simulations.

The magnetic confinement of a hot plasma is the most promising concept to realize controlled thermonuclear fusion on earth. In the last years of intense research activities in the frame of broad international collaboration, it became clear that on the way to a stationary operating fusion reactor not only questions of plasma heating, transport and stability but also the problems associated with the choice of plasma facing materials are decisive. These issues cannot be decoupled from each other [1]. It is demonstrated that both sides, the plasma and the wall, exhibit mutual dependences. Burning conditions will not be achieved without careful adaptation of the chosen materials to the developed plasma scenarios and vice versa. Integrated concepts are required.