Abstract
A key element in future toroidal magnetic fusion machines like ITER is the design of a divertor, which allows for safe particle and power exhaust in parallel with high bulk plasma performance. Correspondingly, the definition and design of an optimized divertor is a major task within the ongoing international ITER research and development effort. In order to provide a profound physics basis for such a divertor optimization, different divertor geometries are being tested on major tokamaks. This paper describes the effects of these divertor modifications on plasma performance in ASDEX Upgrade and in JET. In conclusion, increasing closure improves divertor performance without limiting the core plasma performance in ELMy H-modes.
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