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A set of 24 in-vessel saddle coils is planned for MHD control experiments in ASDEX Upgrade. These coils can produce static and alternating error fields for suppression of Edge Localised Modes, locked mode rotation control and, together with additional conducting wall elements, resistive wall mode excitation and feedback stabilisation experiments. All of these applications address critical physics issues for the operation of ITER. This extension is implemented in several stages, starting with two poloidally separated rings of eight toroidally distributed saddle coils above and below the outer midplane. In stages 2 and 3, eight midplane coils around the large vessel access ports and 12 AC power converters are added, respectively. Finally (stage 4), the existing passive stabilising loop (PSL), a passive conductor for vertical growth rate reduction, will be complemented by wall elements that allow helical current patterns to reduce the RWM growth rate for active control within the accessible bandwidth. The system is capable of producing error fields with toroidal mode number n = 4 for plasma edge ergodisation with core island width well below the neoclassical tearing mode seed island width even without rotational shielding. Phase variation between the three toroidal coil rings allows to create or avoid resonances with the plasma safety factor profile, in order to test the importance of resonances for ELM suppression.

The need for high I3 ( -i reactor economy) and high plasma currents ( --f good confinement)under the consmint of the safety factor at the plasma boundary % > 2 c a b for vertically elongated plasma cross-sections in a tokamak reactor, which are inherently unstable to vertical displacements. Furthermore. the creation of divertor configurations allowing particle and energy exhaust with reactor relevant poloidal field coil configurationsusing superconducting coils means that the main shaping coils have to be installed outside the toroidal field coilsalso leads to elongated plasma shapes.One of the main research areas of present divertor tokamaks is the position, shape and disruption control of noncircular plasmas withas far as possiblereactor relevant poloidal field coil configurations allowing merent divertor configurations ( S N single null D N double null) with an inherently large divertor region (figure 1). Passive structures close to the plasma are necessary to slow down the vertical plasma displacement which is unstable on the fast Alfven time scale by means of induced currents in these structures (section 2). The vertical position instability, now on the slower L/R time scale of the passive structure, can then be handled by active fast position control systems, these being increasingly designed as prototypes for a nextstep device [I]. Nevertheless, loss of control can occur (section 3). resulting in a "vertical displacement event" (VDE), first observed in JET discharge #1947 in 1984 121.The dangers associated with these M E s and the intimately linked plasma disruptions are twofold. Firstly, thermal loads are exerted on target plates and wall segments coming from the thermal plasma and magnetic energies. Secondly, electromagnetic forces act on structures resulting from the liberation of the magnetic self energy of the toroidal plasma current $ , and of the interaction energy of I , , with the external quadrupole currents needed for elongated plasma crosssections [;?-SI. This interaction energy drives the vertical instability and generates vertical $3 forces on the passive structures and their supports amounting to as much as 350 tons in JET, or vessel movements of up to 2 mm [3,4]. A substantial fraction of these forces can be due to halo currents, which are caused by the flux changes during VDEs and flow outside the closed flux surfaces in the scrape-off layer (SOL) closing through plasma facing vessel components (section 4). These halo currents also exert deforming forces on the vessel itself and the plasma facing invessel components which are large enough to damage these structures [4]. Arcing due to the voltages induced during fast plasma movements is a third cause of severe damage observed 1471.This review examines results from all non-circular tokamaks with a distinct emphasis on investigations in ASDEX-Upgrade. There a major fraction of the experimental time has been dedicated srudying VDEs of SN plasmas over a large range of q-values (section 5 ) in an attempt to obtain the scaling of both the displ...

Recent experiments at ASDEX Upgrade have achieved advanced scenarios with high β N (>3) and confinement enhancement over ITER98(y, 2) scaling, H H98y2 = 1.1-1.5, in steady state. These discharges have been obtained in a modified divertor configuration for ASDEX Upgrade, allowing operation at higher triangularity, and with a changed neutral beam injection (NBI) system, for a more tangential, off-axis beam deposition. The figure of merit, β N H ITER89-P , reaches up to 7.5 for several seconds in plasmas approaching stationary conditions. These advanced tokamak discharges have low magnetic shear in the centre, with q on-axis near 1, and edge safety factor, q 95 in the range 3.3-4.5. This q-profile is sustained by the bootstrap current, NBI-driven current and fishbone activity in the core. The off-axis heating leads to a strong peaking of the density profile and impurity accumulation in the core. This can be avoided by adding some central heating from ion cyclotron resonance heating or electron cyclotron resonance heating, since the temperature profiles are stiff in this advanced scenario (no internal transport barrier). Using a combination of NBI and gas fuelling line, average densities up to 80-90% of the Greenwald density are achieved, maintaining good confinement. The best integrated results in terms of confinement, stability and ability to operate at high density are obtained in highly shaped configurations, near double null, with δ = 0.43. At the highest densities, a strong reduction of the edge localized mode activity similar to type II activity is observed, providing a steady power load on the divertor, in the range of 6 MW m −2 , despite the high input power used (>10 MW).