This paper reviews recent results of the study of dust in magnetic fusion devices.
Assessment of the role of dust in current fusion devices and ITER is presented. Dust …

ITER will use beryllium as a plasma-facing material in the main chamber, covering a total
surface area of about 620 m 2. Given the importance of beryllium erosion and co-deposition …

Abstract The Chinese Fusion Engineering Testing Reactor (CFETR), complementing the
ITER facility, is aiming to demonstrate fusion energy production up to 200 MW initially and to …

Edge-localized-modes (ELMs) are a ubiquitous feature of H-mode in tokamaks. When
gradients in the H-mode transport barrier grow to exceed the MHD stability limit the ELM …

A heuristic model for the plasma scrape-off width in low-gas-puff tokamak H-mode plasmas
is introduced. Grad B and curv B drifts into the scrape-off layer (SOL) are balanced against …

The basic physics of the processes playing the most important role in divertor plasma
detachment is reviewed. The models used in the two-dimensional edge plasma transport …

The heating of tungsten monoblocks at the ITER divertor vertical targets is calculated using
the heat flux predicted by three-dimensional ion orbit modelling. The monoblocks are …

High-performance and long-pulse operation is a crucial goal of current magnetic fusion
research. Here, we demonstrate a high-confinement plasma regime known as an H-mode …

The findings of the EU 'Materials Assessment Group'(MAG), within the 2012 EU Fusion
Roadmap exercise, are discussed. MAG analysed the technological readiness of structural …

An improved energy confinement regime, I-mode, is studied in Alcator C-Mod, a compact
high-field divertor tokamak using ion cyclotron range of frequencies (ICRFs) auxiliary …