H. Grote

Transition From Construction to Operation Phase of the Wendelstein 7-X Stellarator

Assembly of the superconducting stellarator Wendelstein 7-X is well advanced, and commissioning of the device is being prepared. A first draft of the commissioning tasks has been developed and will be discussed in this paper.

Chemical sputtering yields of carbon based materials at high ion flux densities

Graphite and advanced carbon fiber composites (CFC) are widely used inside the vacuum vessel of magnetic fusion devices. However, erosion by chemical sputtering via hydrocarbon formation might limit their application as target material in future machines like ITER. The first system- atic study of the chemical erosion of graphite and different CFCs (including a silicon-doped one) as a function of ion flux density in the range of 1.4 × 10 21 -5 × 10 22 m -2 s -1 was performed in the plasma generator PSI-1. The results of three different analysis methods agree within about 40%. No differences in the chemical erosion yields between hydrogen and deuterium exposures are found for the various materials. In contrast, the erosion yields differ up to a factor of two for the different CFC-materials. In general, the chemical sputtering yields decrease with increasing ion flux density according to -0.6 reaching levels below 1% at the highest fluxes. Scanning electron microscopy (SEM) and energy dispersive X- ray analysis (EDX) show preferred erosion in the area between the carbon fibers.

Results from a double-sided langmuir probe in T-10, and an extended model of a probe in a streaming magnetized plasma

Abstract By means of a JANUS-type Langmuir probe operated in the scrape-off layer (SOL) in T-10 the temporal evolution and radial profiles of plasma density, electron temperature, and toroidal asymmetry of saturation currents were measured in dependence on discharge parameters. Of particular importance turned out to be the B T field direction and the radial probe position relative to the active limiter structure (rail and aperture limiter). The interpretation of the results is based on the local SOL structure in terms of connection lengths to the active limiter and special flow effects in T-10, and makes use of a theoretical probe model by Hutchinson. As to the latter model, a refinement concerning its basic principles is presented independently.

Physical Aspects and Design of the Wendelstein 7-X Divertor

Abstract For the Wendelstein 7-X stellarator, an “open divertor” was chosen as a first step in divertor development for the expected extended magnetic and plasma parameter range. Particularly, the three-dimensional (3-D) geometry of the boundary and the provided stationary operation are challenges for the design. So far, simplified models have been used to specify the geometry of the divertor and the performance of the high-heat-load surfaces. By applying the 3-D codes that are now available, the results concerning local heat load and particle exhaust can have more detailed evaluation and can be confirmed generally. Together with the development of improved high-heat-load components, a significant reduction of the target area in comparison with the previous design is possible. The new specifications will be characterized.

Overview of EU CFCs development for plasma facing materials

To improve the properties of carbon materials, the tritium inventory should be reduced, chemical erosion and RES have to be suppressed to increase the resistance to water/oxygen at elevated temperatures. In addition, in the next generation devices, i.e., the International Thermonuclear Experimental Reactor (ITER), plasma disruption, slow transients, and ELMs, which can occur as off-normal events as the result of a transition from detached divertor operation to attached operation causes extremely high heat loading to carbon protection material. Therefore, Carbon fiber composites (CFCs) with high thermal conductivity (300 W m -1 K -1 at 20°C, 145 W m -1 K -1 at 800°C) are favourable. In framework of European Fusion Technology program, a great effort has been made to develop CFCs to meet all requirements. This paper presents an overview in progress of EU CFCs development. The characteristics of CFCs with respect to thermal-mechanical properties, erosion by plasma, tritium retention. H 2 O/O 2 reactions, and neutron irradiation effects were reported.

Deposition probe investigations of the sol of T-10

Abstract Several hundred systematic deposition probe measurements were carried out in the SOL of T-10 to investigate the time development of the Fe-impurity flux at various radial positions. During these investigations the discharge conditions as well as the limiter configuration were varied over a wide range. Some qualitative guiding principles are given for deposition probe measurements and their interpretation. On the basis of the radial dependence of the iron flux during both the plateau and the end phase of the tokamak discharge, the influence of erosion processes as well as of local sources and enhanced transport on the iron deposition is discussed. The experiments show an immediate relation between the direction of the toroidal magnetic field and the evolution of the impurity flux during the plateau phase, no influence of various limiter arrangements could be detected.

On the distribution of metals deposited onto the limiter and the liner of tokamaks after long-term operation

Abstract Three inspections of the inner parts of the discharge vessels of T-10 and TM1-MH after long-term operation revealed that metals originating from the various construction materials are distributed inhomogeneously over the first wall of these tokamaks. This partially allows one to identify local metal sources and to indicate anisotropies of the transport. Different materials from inner structures, even if they were only used in earlier experiments, are observed at all limiter surfaces and as components of the debris consisting of macroparticles of different size, shape and elemental composition. There are metallic deposits of the form of structured films or of solidified droplets.

Comparison of chemical sputtering yields for different graphites at high ion flux densities

Graphite is widely used inside the vacuum vessel of magnetic fusion devices and is proposed as target material for future machines like ITER. There are, however, uncertainties concerning the erosion of the material by chemical sputtering via hydrocarbon formation at high ion flux densities. We report on experiments at the plasma generator PSI-1 using a stationary quasi-neutral plasma beam. The ion flux densities used cover the range from 4 · 1020 to 1.2 · 1022 m−2 s−1. They are thus filling the gap between the upper limit of ion beam experiments (1020 mm−2 s−1 and tokamak relevant values (> 1023 m−2 s−1). To suppress impurity-induced erosion the hydrogen discharges were carefully conditioned and checked for possible impurities, especially oxygen. Samples of different advanced carbon fiber composites (CFC) - including a silicon-doped one - were exposed to various plasma conditions. A calibrated mass spectrometer monitored the CxHy-formation in situ and the axial dependence of the CH-band intensities at 431 and 432.4 nm in front of the target was detected. We have studied the temperature dependence (250–700°C) of the erosion yields at ion flux densities up to more than 1022 m−2 s−1 in hydrogen discharges. Weight loss measurements and scans with an optical profilometer were used to determine the mass loss. For Si-doped CFC an erosion yield of 1% was found, which is a factor of two less than for pure CFC.

Evaluation of an Advanced Silicon Doped CFC for Plasma Facing Material

To improve the CFCs characteristics of chemical erosion, tritium retention, and H s O/air resistivity, an advanced silicon doped 3D CFC has been developed in the framework of the European Fusion Technology Programme. This paper presents the manufacturing procedure, the thermal - mechanical properties and the results of experimental investigations on: outgassing behaviours, erosion by plasma, H 2 O/air reaction, stability of doped Si-in bulk material under D + ion irradiation and DAT retention. The results have been critically analysed and the consequence is discussed.

Deposition probe measurements of impurities injected into a tokamak plasma

Impurity confinement behaviour has been studied by using a deposition probe in conjunction with pellet injection. Generally, an exponential decay of the impurity efflux and nearly symmetric ion/electron side toroidal flows have been observed. During phases of strong plasma disturbances, asymmetric flow is seen, indicative of edge transport and prompt recycling from local sources. The application of ECRH may cause such disturbances.