Joint research on the tokamaks DIII-D and EAST demonstrates a successful integration control of divertor detachment with excellent core plasma performance, a milestone towards solving the critical Plasma-wall-interaction (PWI) issues for ITER and future reactors. DIII-D has achieved actively controlled fully detached divertor with low plasma temperature ($T_{e,div} \le $ 5 eV across the entire...
Substantial seeding of impurities into the divertor has been used for a long time in tokamaks to reduce the power and particle fluxes impacting on the divertor targets and is one of the main techniques to be utilised on ITER to facilitate stationary divertor operation [$1$]. There have been attempts to predict how the impurity concentration required for detachment should scale with different...
Power exhaust is one of the big challenges for future fusion reactors. In the EU programme, both conventional and alternative divertor approaches are studied. For a conventional divertor in EU-DEMO (1), more than 80% of the exhaust power needs to be dissipated before entering the SOL to keep the divertor in the detached regime, where the interaction of the plasma with the wall is significantly...
A fusion reactor based on a stellarator design has the advantage of an easier access to long pulse scenarios. In fact, one of the main goals of Wendelstein 7-X (W7-X), the largest advanced stellarator in the world, is to demonstrate the steady-state capabilities of the stellarator line. From 2022 onward, all plasma-facing components will be water-cooled, which should enable pulse duration of...
Introduction – Achieving safe power and particle exhaust compatible with a high-performance core plasma is one of the main challenges towards commercial fusion power. Currently, the most promising solution is to operate a diverted tokamak in detached conditions with high divertor neutral pressure, high volumetric power dissipation, and a strong temperature and pressure gradient along the...
A small angle slot (SAS) divertor concept [1] with a closed slot structure and appropriate target shaping in the near SOL has been developed in order to explore a potentially robust boundary solution with acceptable plasma surface interaction which is essential for fusion reactor plasma conditions, in particular for high-power steady-state operation. Recent experimental tests in DIII-D have...
