In order to simultaneously control the current and pressure profiles in high performance tokamak plasmas with internal transport barriers (ITB), a multi-variable model-based technique has been proposed. New algorithms using a truncated singular value decomposition (TSVD) of a linearised model operator and retaining the distributed nature of the system have been implemented in the JET control system. Their simplest versions have been applied to the control of the current density profile in reversed shear plasmas using three heating and current drive actuators (neutral beam injection, ion cyclotron resonant frequency heating and lower hybrid current drive). Successful control of the safety factor profile has been achieved in quasi steady state, on a time scale of the order of the current redistribution time. How the TSVD algorithm will be used in the forthcoming campaigns for the simultaneous control of the current profile and of the ITB temperature gradient is discussed in some detail, but this was not yet attempted in the present pioneering experiments.

The frequency spectral broadening of lower hybrid (LH) waves injected into tokamak plasmas is extensively analysed with reference mostly to experimental data from the ASDEX tokamak. The link between the magnitude of the pump spectral width and the degradation of the LH current drive efficiency (up to a factor of 2), pointed out in previous works, is explained. The experimental behaviour of LH power absorption is also well reproduced, even in situations when the access of the launched LH waves to the core plasma should be largely forbidden. Experiments are described that are aimed at determining whether parametric decay instabilities (PDIs) or scattering of LH waves by density fluctuations in the plasma edge are causes of the broadening of the LH pump frequency spectrum. Fluctuations emerge as the largely dominant process, while no signature of PDI processes is observed. Careful measurements of the density fluctuations in the ASDEX scrape-off layer plasma allow the analytical description given by Andrews and Perkins (1983) to be assumed as the appropriate model for LH scattering. Indeed, it supplies the correct magnitude for the frequency spectral width of the LH pump, and explains quantitatively, together with a ray tracing code, why the CD efficiency decreases with the broadening of the pump spectrum. It can also account for the observed LH power absorption coefficient

Advanced scenario plasmas must often be run at low densities and high power, leading to hot edge temperatures and consequent power handling issues at plasma-surface interaction zones. Experiments at JET are addressing this issue by exploring the use of extrinsic impurity seeding and D 2 puffing to reduce heat fluxes. The experiments presented in this paper continue the line of Advanced Tokamak scenario studies at high triangularity in JET by concentrating on the characterisation of the edge pedestal and the ELM behaviour with Deuterium and/or light impurity fuelling (Neon, Nitrogen). Both injection of extrinsic impurities and D 2 puffing are shown to have a significant impact on the edge pedestal in typical JET Advanced Tokamak conditions. The ELM energy loss, ∆W ELM /W dia , can be reduced to below 3% and the maximum ELM penetration depth can be limited to r/a > 0.7, thus enhancing the possibility for sustainable internal transport barriers at large plasma radius. These conditions can be achieved in two separate domains, either at a radiated power fraction (F rad) of 30% or at a fraction of > 50%. At the lower F rad the ELMs are Type I and a high pedestal pressure is maintained, but the occasional large ELM may still occur. At F rad > 50% the pedestal pressure is degraded by 30%-50%, but the ELMs are degraded to Type III. The intermediate regime at F rad ~ 40% is unattractive for ITB scenarios because large Type I ELMs occur intermittently during the predominantly type III ELM phases (compound Type I/III). F rad = 30%, can be obtained with D 2 fuelling alone, whereas neon or nitrogen seeding is needed to achieve F rad > 50%. Only a limited number of tests have been carried out with nitrogen seeding, with the preliminary conclusion that the plasma edge behaviour is similar to that with neon seeding once the radiated fraction is matched.

Liquid lithium as a plasma-facing material was tested for the first time on a high field medium size tokamak, FTU. A liquid Li reservoir supplies a mesh of capillaries that is movable from shot to shot in the scrape-off layer (SOL) plasma to act as a secondary limiter. An almost complete lithization of the vacuum vessel walls is obtained in about three discharges. Plasmas cleaner than boronization and titanization, with lower radiation losses and smaller impurity content are produced. The SOL electron temperature increases, T e ∼ 10 eV, while density (n e ) is less affected. The 2D multifluid code TECXY explains this only if a strong reduction of plasma recycling on the walls and main limiter occurs, consistent with the high Li hydrogen pumping capability. This property also permits a much tighter control of the plasma density. With the Li limiter inserted inside the vessel poloidal asymmetries develop in the SOL that TECXY explains with a local increase of radiation, caused by enhanced evaporation/sputtering of Li. New regimes can be produced in such conditions with a clear increase in |∇n e /n e | and of the peaking factor n e0 / < n e 2 at the Greenwald density limit ( ne ∼ 2 × 10 20 m −3 ), without any direct central particle fuelling.

A successful demonstration of real-time, model-based control of the plasma current density profile has been made in JET. The safety factor profile was reconstructed using magnetic and polarimetric signals. Various predefined q-profile targets have been reached-in the least square approximation-using lower hybrid current drive as the only actuator, with a feedback control loop using the measurements from five fixed normalized radii.

The first evidence of dust-impact ionization processes in the scrape-off layer of the Frascati tokamak upgrade (FTU) was reported in Nucl. Fusion 47 L5 (2007). In this work an extended data analysis (both of electrostatic probe signals and probe surface analysis) is presented, using results of empirical studies of hypervelocity impacts. Within the accuracy and limits of application of such results to tokamak plasma environment, the data are shown to be consistent with occurrence of impacts by micrometre size iron particles at velocities of the order of 10 km s−1.

A 'hybrid' scenario for ITER is defined through its objectives: a large fusion yield for a long time duration. In many tokamaks, discharges characterized by a stationary current density profile, enclosing a large volume of low magnetic shear with q 0 near 1, have achieved improved confinement and higher beta limits. Their extrapolation to ITER from existing data corresponds to the ITER hybrid scenario. These discharges are characterized by soft MHD events. Physics issues relevant to the existence and extrapolation of this scenario will be addressed. New JET experiments with a large component of RF heating have answered some of these issues: injected momentum is not essential, hybrid scenarios are achievable at low ρ * , hybrid regimes have been achieved with ITER-relevant T e /T i and they are compatible with a very low edge activity/low pressure pedestal. Data from pure RF discharges in other tokamaks (FTU, TS, TCV) seem to confirm that a large volume of low magnetic shear with q 0 close to 1 is the key to achieving hybrid scenarios. Issues needing resolution in the extrapolation to ITER are discussed. The present understanding provides encouraging prospects for the use of this scenario in ITER.

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.