Understanding impurity transport in the scrape-off layer (SOL) plasma in tokamak fusion devices is essential in predicting the retention of the fusion fuel on the plasma-facing surfaces – especially in ITER and future reactors, where the fuel mixture contains radioactive tritium. Therefore, diagnostics methods for measuring the SOL plasma flows and monitoring the amount and spatial distribution of the retained fuel are required. Plasma flows were inferred indirectly in the high-field side (HFS) SOL of the ASDEX Upgrade tokamak by means of Doppler spectroscopy of injected nitrogen (N) ions. The measurements were performed in low-power L-mode discharges with different degrees of HFS divertor detachment. The results suggest impurity flows away from the HFS divertor in the near SOL in high-recycling conditions and towards the HFS divertor across the SOL in detached conditions. The physical phenomena driving the SOL flows of deuterium (D) on the HFS were studied by the SOLPS 5.0 code in a generic density and power scan. The near-SOL flows were found to correlate with the poloidal SOL pressure asymmetries. As measured for N ions, reversal of the near-SOL flow was observed in high-recycling conditions, coinciding with the dominance of HFS pressure over the low-field side (LFS) pressure. The HFS pressure peak was built up as a combination of localized ionization of the recycled D and drift-driven radial transport. At high input powers, the reversal of the flow spread also into the far SOL due to radial momentum transport. Despite the qualitative agreement between the measured N and simulated D flow patterns, ERO simulations of the N2 injection suggested poor entrainment of the N ions with the D flow and inadequate quantitativity of the measurements. The description of the detachment of the HFS divertor in SOLPS 5.0 was improved by applying a convection-dominated radial ion transport model, mimicking non-diffusive radial transport phenomena. This resulted in improved confinement of the high-field side high-density front in the HFS divertor in agreement with spectroscopic measurements, allowing to increase the neutral fuelling with the D2 injection in the experiment. Consequently, the significant decrease of the HFS target ion flux in detachment was reproduced, consistent with the target probe measurements. Applicability of laser-induced breakdown spectroscopy (LIBS) for in situ monitoring of fuel retention in ITER was investigated by studying different ITER-relevant samples containing beryllium (Be) with an experimental set-up built at VTT Technical Research Centre of Finland. The retained D was distinguished from hydrogen in vacuum and at low ambient pressures, while the distinction could not be made at atmospheric pressures, planned for measurements in ITER, due to line broadening. LIBS was also successfully used for determining the relative concentrations of Be and W in a mixed coating and the qualitative deposition profiles of Be and D across the HFS divertor of the JET tokamak in agreement with secondary ion mass spectrometry (SIMS).
|Translated title of the contribution||Spectroscopic measurements of impurity migration, deposition and fuel retention in fusion devices|
|Publication status||Published - 2018|
|MoE publication type||G5 Doctoral dissertation (article)|
- scrape-off layer
- plasma flow
- fuel retention