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Local beryllium (Be) I and Be II line intensities were measured in the plasma–wall interaction region near an ICRH antenna in JET. The intent was to use these intensities as a measure of the formation of local Radio Frequency (RF) sheath potentials, through RF sheath rectification and potential build up at the end of field lines passing in front of the antenna. Experimentally, it was found that the Be I and Be II emission increase when using the antenna local to the spectroscopic measurement, and increase even more when using a remote antenna that is magnetically connected to the observation point. Magnetic field mapping indicates a magnetic connection between the observation location and the top corner region of the remote antenna and/or its protection limiter. These measurements can be used in support of RF sheath modeling that is an important part of the optimization of antenna design for next generation fusion energy devices, including ITER.

A set of dedicated JET experiments is described where the plasma elongation (k) and triangularity (δ) were varied separately in order to study the influence of plasma magnetic topology on the internal transport barrier (ITB). With low δ, type III ELMs were observed and ITBs readily generated. At the highest δ, large type I ELMs and ELM free phases were observed but, at best, only marginal ITBs. At fixed δ the increase of the elongation of internal magnetic surface have a beneficial effect on the transport quality of the ITB.

A set of discharges performed with the JET ITER-like wall is investigated with respect to control capabilities on tungsten sources and transport. In attached divertor regimes, increasing fueling by gas puff results in higher divertor recycling ion flux, lower divertor tungsten source, higher ELM frequency and lower core plasma radiation, dominated by tungsten ions. Both pedestal flushing by ELMs and divertor screening (including redeposition) are possibly responsible. For specific scenarios, kicks in plasma vertical position can be employed to increase the ELM frequency, which results in slightly lower core radiation. The application of ion cyclotron radio frequency heating at the very center of the plasma is efficient to increase the core electron temperature gradient and flatten electron density profile, resulting in a significantly lower central tungsten peaking. Beryllium evaporation in the main chamber did not reduce the local divertor tungsten source whereas core radiation was reduced by approximately 50%.

In optimized shear plasmas in the Joint European Torus [P. H. Rebut and B. E. Keen, Fusion Technol. 11, 13 (1987)], safety factor (q) profiles with negative magnetic shear are produced by applying lower hybrid (LH) waves during the plasma current ramp-up phase. These plasmas produce a barrier to the electron energy transport. The radius at which the barrier is located increases with the LH wave power. When heated with high power from ion cyclotron resonance heating and neutral beam injection, they can additionally produce transient internal transport barriers (ITBs) seen on the ion temperature, electron density, and toroidal rotation velocity profiles. Due to recent improvements in coupling, q profile control with LH current drive in ITB plasmas with strong combined heating can be explored. These new experiments have led to ITBs sustained for several seconds by the LH wave. Simulations show that the current driven by the LH waves peaks at the ITB location, indicating that it can act in the region of low magnetic shear.

Some models for the suppression of turbulence in the L to H transition, suggest that the width of the H-mode edge barrier is either proportional or is of the order of the thermal or the fast-ion poloidal Larmor radius. This would require that the width of the edge barrier should depend on the plasma current. This dependence has been clearly verified at JET in experiments designed to control the edge MHD stability of ELM-free hot-ion H-mode plasmas. The effects of isotopic mass and the applicability of several edge barrier models to the hot-ion H-mode plasmas were analysed in (Guo H Y et al 2000 Edge transport barrier in JET hot-ion H-modes Nucl. Fusion 40 69) using a large database containing both deuterium-only and deuterium-tritium plasmas. This database has now been enlarged to include discharges from a plasma shape scan, allowing one to study the dependence of the pedestal height on the edge shear. In addition, the range of plasma currents was extended up to 6 MA. It is shown that the edge data are best described by a model where the edge barrier width is determined by the fast ions weighted towards the components with largest poloidal Larmor radii. However, it is not possible to conclusively eliminate the thermal ion model.

After the change over from the C-wall to the ITER-like Be/W wall (ILW) in JET, the radiation losses during ICRF heating have increased and are now substantially larger than those observed with NBI at the same power levels, in spite of the similar global plasma energies reached with the two heating systems. A comparison of the NBI and ICRF performances in the JET-ILW experiments, based on a statistical analysis of ∼3000 L-mode discharges, will be presented.

The experimental program of the Princeton Beta Experiment‐Modified (PBX‐M) tokamak [Phys. Fluids B2, 1271 (1990)] is directed toward tailoring plasma profiles to achieve greater stability and confinement and to gain access to the second stability region. Modification of the current density profile has been achieved with lower‐hybrid current drive (LHCD), leading to a regime free of global magnetohydrodynamic modes, while raising the value of q(0) above unity. The diffusion of the fast electrons produced by LHCD has been examined using two‐dimensional hard x‐ray imaging. Ion Bernstein waves (IBW) have been used for ion heating: a preliminary analysis shows that ion heating was spatially localized and in agreement with theoretical calculations. Divertor biasing has modified the electric field inside the last closed surface, resulting in the formation of a transport barrier, which in turn has reduced the threshold power of neutral beam injection (NBI) for H‐mode transition by 25%.

Sawtooth control is a potentially important method for avoiding MHD instabilities triggered by sawtooth crashes, such as neoclassical tearing modes in a reactor‐size tokamak. The aim of the present study is to investigate the effect of localised current drive on Sawtooth activity. In recent experiments in the Tore Supra tokamak, Electron Cyclotron Current Drive has been observed to strongly affect the Sawtooth period, even at modest ECCD power levels. For instance, with 8 MW of ICRH long, fast ion stabilised, sawteeth were shortened from 0.6 s to 0.3 s by application of 0.6 MW of co‐ECCD. In these experiments it is estimated that only 2–3% of the total current was driven by ECCD, but this was sufficient to significantly affect the sawtooth period provided the driven current was localized accurately with respect to the q = 1 surface

In JET, advanced tokamak research mainly focuses on plasmas with internal transport barriers (ITBs) that are strongly influenced by the current density profile. A previously developed optimized shear regime with low magnetic shear in the plasma centre has been extended to deeply negative magnetic shear configurations. High fusion performance with wide ITBs has been obtained transiently with negative central magnetic shear configuration: HIPB98(y,2) ∼ 1.9, βN = 2.4 at IP = 2.5 MA. At somewhat reduced performance, electron and ion ITBs have been sustained in full current drive operation with 1 MA of bootstrap current: HIPB98(y,2) ∼ 1, βN = 1.7 at Ip = 2.0 MA. The ITBs were maintained for up to 11 s for the latter case. This duration, much larger than the energy confinement time (37 times larger), is already approaching a current resistive time. New real-time measurements and feedback control algorithms have been developed and implemented in JET for successfully controlling the ITB dynamics and the current density profile in the highly non-inductive current regime.

When the magnetic shear vanishes over a wide spatial region close to the q=2 surface, a strong MHD activity ascribed to a global tearing mode m/n=2/1 is observed, corresponding to a 15% reduction of the current drive efficiency, and a significant flattening of the HXR profile localized in the vicinity of the island. Calculations of the HXR bremsstrahlung and the LH current drive efficiency are reported based on Fokker-Planck calculations to assess the role played by the fastest electrons of the tail, and identify the fraction which is lost by the MHD perturbation.

Good coupling of the LH power during the whole internal transport barrier (ITB) phase has been achieved in JET by fitting the plasma shape to the LH antenna shape and by injecting CD4 gas from a pipe near to the LH grill. No adverse effect on the edge pedestal or on the ITB quality, or significant impurity influx is observed. The key is the low ionization potential of all the CD4 dissociation chain, which is completed inside the SOL. Promising perspectives are open for injecting safely LHCD with more power during an ITB.

ICRH on JET is used for a wide range of experiments: confinement and H-mode threshold scaling experiments, combined heating experiments at high plasma current and heating in high performance plasmas which will be presented here. Experiments of combined NBI and ICRH in ELM-free Hot Ion H-modes have been carried out in the JET MKII campaign in 3.8 MA/3.4 T plasmas with up to 9.5 MW of ICRH power together with 16 MW of NBI using minority hydrogen resonance. ICRH results in increased core electron temperature, increased core ion temperature and increased stored energy up to a record of 14.5 MJ. Various schemes have been used including polychromatic mode of operation. Data analysis suggests that a substantial part of the ICRH power is coupled to deuterons at the second harmonic resonance. This is the dominant heating mechanism in new experiments with shear optimisation where the power is injected during the current ramp-up phase of the discharge. Very peaked ion pressure profiles are achieved, up to 3.5 bar an...

During ICRF heating experiments, the new four‐strap A2 antennas on JET have coupled 16.5 MW of power into single null divertor plasmas (Fig. 1). In combined heating experiments a maximum of 32 MW has been achieved by adding 15 MW of hydrogen minority ICRH to 17 MW of neutral beam injection (NBI). This combined heating power is a record for the present campaign and was achieved in a high density, strongly radiating, reactor‐like scenario with ‘grassy’ ELMs. Central electron temperatures close to 15 keV have been obtained with both RF—only and combined heating in ELM—free plasmas which were created with high flux expansion in the divertor region and which were substantially triangular (δ≂0.3). Similar discharges were used to obtain the highest D‐D reactivity in JET with NBI. The addition of hydrogen minority ICRH to these plasmas increases both the reactivity and the H‐factor obtained by comparing the thermal energy content to the ITERH93‐P scaling. RF‐only H‐modes have been produced in plasmas with currents (Ip) up to 4.7 MA and the power thresholds are similar those with NBI. At Ip=4.7 MA and a toroidal field (Bt) of 3.2 T, a plasma energy content of 7 MJ was achieved with 8 MW of 51.3 MHz ICRH in 0π0π phasing. The heating efficiency, as determined from power modulation experiments, is close to 100% with 0π0π and 0ππ0 phasing. The efficiency decreases for phasings such as 00ππ and monopole which produce average values of the parallel wavevector less than 3.5 m−1. However, the recent introduction of a separator limiter at the center of one antenna has improved its heating efficiency for perpendicularly launched waves and such separators will be fitted to all antennas in the near future.

Successful current profile control experiments using lower hybrid current drive (LCHD) clearly require knowledge of (1) the location of the driven fast electrons and (2) the ability to maintain that location from spreading due to radial diffusion. These issues can be addressed by examining the data from the hard x-ray camera on PBX-M, a unique diagnostic producing two-dimensional, time resolved tangential images of fast electron bremsstrahlung. Using modeling, these line-of-sight images are inverted to extract a radial fast electron current density profile. We note that ``hollow`` profiles have been observed, indicative of off-axis current drive. These profiles can then be used to calculate an upper bound for an effective fast electron diffusion constant: assuming an extremely radially narrow lower hybrid absorption profile and a transport model based on Rax and Moreau, a model fast electron current density profile is calculated and compared to the experimentally derived profile. The model diffusion constant is adjusted until a good match is found. Applied to steady-state quiescent modes on PBX-M, we obtain an upper limit for an effective diffusion constant of about D*=1.1 m{sup 2}/sec.

Lower hybrid current drive (LHCD) is being explored as a means to control the current density profile on PBX‐M with the goal of raising the central safety factor q(0) to values of 1.5–2 to facilitate access to a full‐volume second stable regime. Initial experiments have been conducted with up to 400 kW of 4.6 GHz LH power in circular and indented plasmas with modest parameters. A tangential‐viewing two‐dimensional hard x‐ray imaging diagnostic has been used to observe the bremsstrahlung emission from the suprathermal electrons generated during LHCD. Hollow hard x‐ray images have indicated off‐axis localization of the driven current. A serious obstacle to the control of the current density profile with LHCD is the concomitant generation of MHD activity, which can seriously degrade the confinement of suprathermal electrons. By combining neutral beam injection with LHCD, an MHD‐free condition has been obtained where q(0) is raised above 1.

The energy of suprathermal electrons during lower‐hybrid current drive on PBX‐M is determined. (AIP)