Ladungsaustauschspektroskopie mit Hilfe eines Wasserstoffdiagnostikstrahls am Tokamak TEXTOR-94

In this work the energy and impurity transport was investigated by means of the active chargeexchange recombination spectroscopy (CXRS). CXRS is a method to determine the ion temperature, plasma rotation and impurity density both space and time-resolved . It is based on the investigation of the spectral shape of the lines, which are emitted by the impurity ions after the CX processes with the neutral particles . The source of the neutral particles are high-energy beams (e.g. heating beams), which penetrate deeply into the plasma and therefore enable the measurements over the entire plasma radius. During this work a new CXRS diagnostics was installed at TEXTOR-94. The principal part of this diagnostics is the diagnostic hydrogen beam RUDI. The RUDI injector ensures an equivalent neutral current of 1 .1 A with an energy of 50 keV and a pulse length of 4 s, modulated with 500 Hz. The observation system covering the whole beam path and a low divergence of 0.6° of the RUDI beam lead to a good space resolution. Measurements using a modified CXRS diagnostics at the heating beam were performed for the characterisation of the plasma edge, particularly in discharges with impurity seeding and improved energy confinement (RI-mode) . With the measurements of the ion temperature and plasma rotation profiles it was proven, that in the RI-mode there is no transport barrier at the plasma edge, which is typical for another regime with the improved confinement, the H-mode. The ratio of the ion and electron temperature at the plasma edge varies between 4 at the low and 1 .5 at the high densities . The ratio TilT, becomes larger with increasing radiation level, because the electrons are cooled directly via inelastic collisions with the impurity ions . A correlation between the measured edge parameters and the global confinement characteristics was observed: the confinement degradation leads to the higher neutral particle densities at the edge, which slow down the toroidal rotation influenced by the flows in the scrape-off layer. Regarding CXRS the most important advantage of the diagnostic beam in relation to the heating beam is the possibility to measure under all discharge conditions. Measurements with RUDI took place to a large extent under conditions of ohmic plasma heating: the energy balance of ions and electrons was investigated for different plasma currents and densities ; in high density discharges the transition to improved ohmic confinement (IOC) was observed after switching off the external gas flow. In the standard high density ohmic plasmas (saturated ohmic confinement - SOC) the confinement time is independent of the plasma density . In contrast, it scales in the IOC regime linearly with the density . The SOC-IOC transition was investigated regarding the influence of the toroidal ITG instability driven by the ion temperature gradient. On the basis of the measured ion temperature distributions the q;-parameter (ratio of the density and ion temperature decay lengths) and the growth rate of the ITG instability were calculated. The ITG mode is destabilised, if ih is larger then a critical value depending on the scale length of the density profile. After the SOC-IOC transition i1i lies in a noticeably smaller radial region over the critical value. As the result, the IOC regime is characterised by a clear reduction of the growth rate y1To. The steepening of the plasma density profile after the reduction of the external gas flow leads to the suppression of the ITG instability and to the improvement of the confinement in the IOC regime. First measurements of the impurity densities in ohmic and additionally heated discharges were performed. Densities of C6+, Nel°+, Ney+, Nex' and Ox+ were determined. The measured density profiles show qualitatively a good agreement in their radial shape with the profiles calculated by the impurity transport code RITM. However, there are relatively large quantitative deviations of up to 50 %, which can be explained by inaccurate CX rate coefficients for the not completely ionised particles. For clarifying these discrepancies additional measurements of the impurity densities under different plasma conditions are needed

Surface morphology of tungsten exposed to helium plasma at temperatures below fuzz formation threshold 1073 K

Impact of crystal orientation on the surface morphology of the helium plasma exposed tungsten has been investigated on the linear device PSI-2. A nanoscale undulating surface structure having a periodic arrangement is formed for temperatures below 1073 K, in contrast to the fuzz nanostructure formation in a higher temperature range. The crests of undulation align with the $\langle 1\,0\,0\rangle$ direction. The interval of the undulation is narrowest at the crystal grain of {1 1 0} surface. The interval becomes wider as the crystal grain surface tilts away from the {1 1 0} surface, and the undulating surface structure is not formed near the {1 0 0} surface. The height of undulations is $\sim 8$ nm, independently of the interval of the undulations, and it corresponds to the depth of the layer heavily damaged due to helium plasma exposure

Bubble formation in ITER-grade tungsten after exposure to stationary D/He plasma and ELM like thermal shocks

Plasma-facing materials (PFMs) in the ITER divertor will be exposed to severe conditions, including exposure to transient heat loads from edge-localized modes (ELMs) and to plasma particles and neutrons. Tungsten is the material chosen as PFM for the ITER divertor. In previous tests, bubble formation in ITER-grade tungsten was detected when exposed to fusion relevant conditions. For this study, ITER-grade tungsten was exposed to simultaneous ELM-like transient heat loads and D/He (6%) plasma in the linear plasma device PSI-2. Bubble formation was then investigated via SEM micrographs and FIB cuts. It was found that for exposure to 100.000 laser pulses of 0.6 GWm−2 absorbed power density (Pabs), only small bubbles in the nanometer range were formed close to the surface. After increasing Pabs to 0.8 and 1.0 GWm−2, the size of the bubbles went up to about 1 µm in size and were deeper below the surface. Increasing the plasma fluence had an even larger effect, more than doubling bubble density and increasing bubble size to up to 2 µm in diameter. When using deuterium-only plasma, the samples showed no bubble formation and reduced cracking, showing such bubble formation is caused by exposure to helium plasma

Linear Plasma Device PSI-2 for Plasma-Material Interaction Studies

The linear plasma device PSI-2 serves as a pilot experiment for the development of components, operational regimes and control systems for the linear plasma device JULE-PSI, which will be located in the nuclear environment allowing studies of radioactive and toxic samples. PSI-2 is also used for fusion reactor relevant plasma-material interaction studies. This contribution describes the PSI-2 layout and parameters and summarizes the recent scientific and technical progress in the project, including the installation of a target station for the sample manipulation and analyses

Short-term retention in metallic PFCs: modelling in view of mass spectrometry and LIBS

Based on the conventional model of hydrogen retention in plasma-facing components, the question of hydrogen outgassing during and after plasma exposure is addressed in relation to mass spectrometry and laser-induced breakdown sprectroscopy (LIBS) measurements. Fundamental differences in retention and release data acquired by LIBS and by mass spectrometry are described analytically and by modelling. Reaction-diffusion simulations are presented that demonstrate possible thermal outgassing effects caused by LIBS. Advantages and limitations of LIBS as a tool for analysis of short term retention are discussed