Progress in thermonuclear fusion energy research based on deuterium plasmas magnetically confi ned in toroidal tokamak devices requires the development of effi cient current drive methods. Previous experiments have shown that plasma current can be driven effectively by externally launched radio frequency power coupled to lower hybrid plasma waves. However, at the high plasma densities required for fusion power plants, the coupled radio frequency power does not penetrate into the plasma core, possibly because of strong wave interactions with the plasma edge. Here we show experiments performed on FTU (Frascati Tokamak Upgrade) based on theoretical predictions that nonlinear interactions diminish when the peripheral plasma electron temperature is high, allowing signifi cant wave penetration at high density. The results show that the coupled radio frequency power can penetrate into high-density plasmas due to weaker plasma edge effects, thus extending the effective range of lower hybrid current drive towards the domain relevant for fusion reactors.

Combined analysis of divertor thermocouple and IR camera measurements during JET disruptions can provide valuable information on the distribution of the energy loads, even if the stored energy of the JET plasmas is small compared to that foreseen for the next-generation tokamaks. Typically the energy collected at the divertor represents a small fraction of the pre-disruption plasma energy; this is consistent with the high level of radiation observed and with part of the magnetic energy being transferred to the plasma-coupled conductors.The data for this paper are taken from the whole set of disruptive plasmas of JET operation in the years 2000 and 2001. In most of the MkIIGB disruptions, the plasma displaces upwards (away from the divertor); therefore, only a small number of downward events are available for analysis. However, divertor heat loads seem to be more strongly correlated to the delay of the loss of the X-point with respect to the thermal quench than the direction of the plasma displacement.When the plasma thermal energy is lost with the plasma still in X-point configuration, the septum and the tiles wetted by the strike-points, often more than one tile per strike-point, experience a sharp increase in temperature, equivalent to up to 1 MJ m −2 . When the thermal quench occurs at the same time as, or after, the loss of plasma vertical control, no significant divertor tile temperature increase can be observed for both upwards and downwards events. Most of the disruptions purposely made to produce runaway electrons went towards the divertor and, although not systematically, lead to local (mostly at the septum) temperature increase equivalent to a load up to 2 MJ m −2 , often toroidally asymmetric.

FAST is a new machine proposed to support ITER experimental exploitation as well as to anticipate DEMO relevant physics and technology. FAST is aimed at studying, under burning plasma relevant conditions, fast particle (FP) physics, plasma operations and plasma wall interaction in an integrated way. FAST has the capability to approach all the ITER scenarios significantly closer than the present day experiments using deuterium plasmas. The necessity of achieving ITER relevant performance with a moderate cost has led to conceiving a compact tokamak (R = 1.82 m, a = 0.64 m) with high toroidal field (B T up to 8.5 T) and plasma current (I p up to 8 MA). In order to study FP behaviours under conditions similar to those of ITER, the project has been provided with a dominant ion cyclotron resonance heating system (ICRH; 30 MW on the plasma). Moreover, the experiment foresees the use of 6 MW of lower hybrid (LHCD), essentially for plasma control and for non-inductive current drive, and of electron cyclotron resonance heating (ECRH, 4 MW) for localized electron heating and plasma control. The ports have been designed to accommodate up to 10 MW of negative neutral beams (NNBI) in the energy range 0.5-1 MeV. The total power input will be in the 30-40 MW range under different plasma scenarios with a wall power load comparable to that of ITER (P /R ∼ 22 MW m −1). All the ITER scenarios will be studied: from the reference H mode, with plasma edge and ELMs characteristics similar to the ITER ones (Q up to ≈1.5), to a full current drive scenario, lasting around 170 s. The first wall (FW) as well as the divertor plates will be of tungsten in order to ensure reactor relevant

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In this paper we present the fusion advanced studies torus (FAST) plasma scenarios and equilibrium configurations, designed to reproduce the ITER ones (with scaled plasma current) and suitable to fulfil plasma conditions for integrated studies of plasma-wall interaction, burning plasma physics, ITER relevant operation problems and steady state scenarios. The attention is focused on FAST flexibility in terms of both performance and physics that can be investigated: operations are foreseen in a wide range of parameters from high performance H-mode (toroidal field, B T , up to 8.5 T; plasma current, I P , up to 8 MA) to advanced tokamak (AT) operation (I P = 3 MA) as well as full non-inductive current scenario (I P = 2 MA). The coupled heating power is provided with 30 MW delivered by an ion cyclotron resonance heating system (30-90 MHz), 6 MW by a lower hybrid system (3.7 or 5 GHz) for the long pulse AT scenario, 4 MW by an electron cyclotron resonant heating system (170 GHz − B T = 6 T) for MHD and localized electron heating control and, eventually, with 10 MW by a negative neutral ion beam (NNBI), which the ports are designed to accommodate. In the reference H-mode scenario FAST preserves (with respect to ITER) fast ion induced as well as turbulence fluctuation spectra, thus addressing the cross-scale couplings issue of micro-to meso-scale physics. The non-inductive scenario at I P = 2 MA is obtained with 60-70% of bootstrap current and the remaining by LHCD. Predictive simulations of the H-mode scenarios have been performed by means of the JETTO code, using a semi-empirical mixed Bohm/gyro-Bohm transport model. Plasma position and shape control studies are also presented for the reference scenario.