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.

A 20 MW/5GHz Lower Hybrid Current Drive (LHCD) system was initially due to be commissioned and used for the second mission of ITER, i.e. the Q=5 steady state target. Though not part of currently planned procurement phase, it is now under consideration for an earlier delivery. In this paper, both physics and technology conceptual designs are reviewed.

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

Since the installation of an ITER-like wall, the JET programme has focused on the consolidation of ITER design choices and the preparation for ITER operation, with a specific emphasis given to the bulk tungsten melt experiment, which has been crucial for the final decision on the material choice for the day-one tungsten divertor in ITER. Integrated scenarios have been progressed with the re-establishment of long-pulse, high-confinement H-modes by optimizing the magnetic configuration and the use of ICRH to avoid tungsten impurity accumulation. Stationary discharges with detached divertor conditions and small edge localized modes have been demonstrated by nitrogen seeding. The differences in confinement and pedestal behaviour before and after the ITER-like wall installation have been better characterized towards the development of high fusion yield scenarios in DT. Post-mortem analyses of the plasma-facing components have confirmed the previously reported low fuel retention obtained by gas balance and shown that the pattern of deposition within the divertor has changed significantly with respect to the JET carbon wall campaigns due to the absence of thermally activated chemical erosion of beryllium in contrast to carbon. Transport to remote areas is almost absent and two orders of magnitude less material is found in the divertor.

A new ITER-relevant lower hybrid current drive (LHCD) launcher, based on the passive-active-multijunction (PAM) concept, was brought into operation on the Tore Supra tokamak in autumn 2009. The PAM launcher concept was designed in view of ITER to allow efficient cooling of the waveguides, as required for long pulse operation. In addition, it offers low power reflection close to the cut-off density, which is very attractive for ITER, where the large distance between the plasma and the wall may bring the density in front of the launcher to low values. The first experimental campaign on Tore Supra has shown extremely encouraging results in terms of reflected power level and power handling. Power reflection coefficient <2% is obtained at low density in front of the launcher, i.e. close to the cut-off density, and very good agreement between the experimental results and the coupling code predictions is obtained. Long pulse operation at ITER-relevant power density has been demonstrated. The maximum power and energy reached so far is 2.7 MW during 78 s, corresponding to a power density of 25 MW m−2, i.e. its design value at f = 3.7 GHz. In addition, 2.7 MW has been coupled at a plasma–launcher distance of 10 cm, with a power reflection coefficient <2%. Finally, full non-inductive discharges have been sustained for 50 s with the PAM.