Duarte Borba

On 25 September 2006, experiments resumed on JET, with the start of Experimental Campaign C16 of the EFDA-JET 2006 Workprogramme. After successful completion of this Campaign on 13 October 2006, Campaign C17 began on 23 October 2006, and was completed on 15 December 2006.

These experimental campaigns took advantage of the recent upgrades to the JET facilities. These included a new configuration of the lower part of the machine’s first wall Рthe area designed to exhaust most of the heat from the plasma with a null in the magnetic configuration (divertor configuration) Рheating system upgrades, and several new and enhanced diagnostics. A strong focus was maintained on the preparation of the ITER detailed design and ITER exploitation. These include the preparation for the planned replacement of the carbon based protection tiles with beryllium and tungsten tiles, foreseen for installation on JET in 2008 and 2009.

Studies in the ITER baseline scenario (ELMy H-mode) included the exploration of different techniques to minimize the sudden heat loss induced by edge plasma instabilities known as Edge Localized Modes (ELMs). These instabilities cause temporary loss of confinement in the plasma edge, leading to significant heat fluxes to the plasma facing components. Enhanced heat loads increases the erosion of the plasma facing materials, limiting the lifetime of such components in ITER. The new bolometer diagnostic was able to resolve radiation patterns between and during ELMs and filamentary power deposition was detected with the new wide angle infrared camera. One approach to reduce the impact of ELMs, further studied in these experiments, is to explore configurations with plasmas shapes, density and plasma current profiles that minimize the energy lost during these edge instabilities. Another approach is to use active techniques such as externally applied magnetic perturbations. This technique was shown to be effective in experiments carried out in the DIII-D tokamak (San Diego, USA), when a resonant magnetic perturbation with three wave periods in the toroidal direction was applied. On JET the Error Field Correction Coils, capable of producing a magnetic perturbation with longer wave lengths such as one and two wave periods in the toroidal direction, were used in these studies. In addition, the application of vertical kicks to the plasma, to control the ELMs was investigated, employing similar techniques to the experiments carried out in the TCV (Switzerland) and ASDEX- Upgrade (Germany) tokamaks. Both techniques proved successful in controlling the frequency and amplitude of ELMs, showing their potential as mitigation tools for these instabilities.

A complementary approach to avoid high heat loads on the plasma facing components on ITER, is to surround the plasma by a radiating zone, which can be produced by impurity injection. The production of such a radiating zone has been studied with nitrogen injection with plasma currents of up to 3 MA. To preserve high fusion performance, it is important that the impurities injected for this purpose do not accumulate to high levels in the plasma core and degrade plasma core confinement. To understand impurity transport, systematic studies were conducted with injection of low to high Z impurities. These experiments will also be useful to understand transport of ITER eroded first-wall materials into the plasma.

Indeed material erosion from the plasma facing components is another important issue for ITER. It relates to the study of migration and retention of impurities in Tokamaks. Recent results highlighted the importance of the magnetic field configuration in determining the location and thickness of the deposited layers, and the fact that a few large ELMs lead to a much stronger erosion and subsequent deposition than small ELMs.

Advanced Tokamak regimes, aiming at producing more favourable plasma configurations with better confinement and stability properties, were also further developed during the recent experiments. These advanced scenarios will reduce the need for externally driven plasma current and allow longer plasma discharges in ITER, thereby offering the prospect of very long pulse or steady state operation. These scenarios were studied with the new plasma shape in JET, which can be reached with the new divertor. The new shape is very similar to the plasma shape foreseen in ITER, allowing JET to explore the improved stability and confinement properties of a more triangular plasma configuration. These studies employed high auxiliary heating power of up to 31MW and the edge stability controlled using the injection of impurities such as Neon.

However, real time control of plasma profiles proved essential for the optimisation of such advanced Tokamak configurations. Therefore, a significant effort was also devoted to integrated real time control of the plasma current and pressure profiles, aiming at robust regimes with significant improved confinement in the plasma core. The stability properties of these scenarios were also studied by measuring the response of the plasma to applied magnetic perturbations.

In ITER, a relatively large gap will be required between the plasma and first-wall components. This poses a problem for the coupling of the power to the plasma using radio frequency systems, such as lower hybrid heating (LH) and ion cyclotron resonance heating (ICRH). Gaps of up to 15cm between the plasma and the antennas are foreseen in ITER. During Campaigns C16 and C17, studies were conducted on JET in which an ITER-like separation was maintained between the plasma and the LH and ICRH antennas, using gas injection close to the antennas in order to improve plasmaantenna coupling.

The new JET data will allow significant progress in the understanding of the behaviour of plasma scenarios in ITER, compatible with acceptable erosion of first wall components, and will make valuable contributions to the development of real time control systems for ITER.