Stellerator concept improvement goals and centers around five major themes
1. Divertor characterization and control.
The W7-X island divertor configuration offers the best near-term opportunity to advance the physics of 3D divertors. U.S. and IPP scientists are collaborating in the application of state-of-the-art edge transport modeling tools, e.g. the EMC3-EIRENE code, to design experiments and make predictions by simulating heat loading of plasma facing components and impurity transport in the core plasma during the first W7-X operating campaign
2. Core turbulence and transport.
3. Error fields and island physics.
4. Equilibrium reconstruction.
5. Energetic particle confinement.
STELLARATORS are magnetic confinement fusion devices using external coils to create the confining magnetic field. Tokamaks, on the other hand, rely on the transformer principle to induce a plasma current which creates one of the magnetic field components. The tokamak configuration is today’s most advanced concept and it is thus used for ITER and will probably be used for DEMO. However, stellarators offer intrinsic advantages over tokamaks: they have the inherent capability for steady-state operation, because they do not use transformer action. Stellarators are also less prone to plasma instabilities and do not develop disruptions, both of which are potentially damaging plasma events. Up to now, stellarator plasmas have shown higher energy and particle loss than tokamak plasmas. As a result, fusion research focussed more on tokamaks and less on stellarators. The advanced stellarator W7-X addresses these issues by employing optimised magnetic field shapes to overcome the lower, stellarator specific energy confinement.